TMI-10-072, Response to Request for Additional Information, Application for Technical Specifications Change Risk-Informed Justification for Relocation of Specific Surveillance Frequency Requirements to Licensee Controlled Program

Response to Request for Additional Information, Application for Technical Specifications Change Risk-Informed Justification for Relocation of Specific Surveillance Frequency Requirements to Licensee Controlled Program

ML102110459
Person / Time
Site:	Crane
Issue date:	07/29/2010
From:	David Helker Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TMI-10-072
Download: ML102110459 (15)

Text

Exein 11 2

rxelon Vva Karl neSt Square. 24 19348 10 CFR 50.90 TMI-1 0-072 July 29, 2010 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

Response to Request for Additional Information, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)

References:

1.

Letter from Pamela B. Cowan, Exelon Generation Company, LLC, to U.S.

Nuclear Regulatory Commission, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3), dated March 24, 2010.

2.

Letter from Peter Bamford, U.S. Nuclear Regulatory Commission, to Michael J. Pacilio, Exelon Nuclear, Three Mile Island Nuclear Station Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Relocation of Surveillance Frequencies to a Licensee Controlled Program (TAC No. ME3587), dated July 2, 2010.

In Reference 1, Exelon Generation Company, LLC (Exelon) submitted a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1 (TMI Unit 1). The proposed amendment would modify TMI Unit 1 TS by relocating selected Surveillance Requirement frequencies to a licensee-controlled program. The NRC reviewed the license amendment request and identified the need for additional information in order to complete their evaluation of the amendment request. On June 17, 2010, draft questions were sent to Exelon to ensure that the questions were understandable, the regulatory basis for the questions was clear, and to determine if the information was previously docketed. On June 23, 2010, a teleconference was held between the NRC and Exelon to further discuss the additional information requested by the NRC. In Reference 2, the NRC formally issued the request for additional information. to this letter provides a restatement of the questions along with Exelons responses.

In addition, TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control RITSTF [Risk-Informed Technical Specifications Task Force] Initiative 5b, dated March 18, 2009, provided an optional insert to existing TS Bases to facilitate adoption of the TSTF traveler.

The TSTF-425 TS Bases insert states as follows:

Exelon Nuclear 20 0 Exelon Way Kennett Square. PA19348 TMI-10-072 July 29, 2010 www.exeloncorp.com Exelon.

Nuclear 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289

References:

Response to Request for Additional Information, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency ReqUirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) 1.

Letter from Pamela B. Cowan, Exelon Generation Company, LLC, to U.S.

Nuclear Regulatory Commission, "Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)," dated March 24, 2010.

2.

Letter from Peter Bamford, U.S. Nuclear Regulatory Commission, to Michael J. Pacilio, Exelon Nuclear, "Three Mile Island Nuclear Station -

Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Relocation of Surveillance Frequencies to a Licensee Controlled Program (TAC No. ME3587)," dated July 2, 2010.

In Reference 1, Exelon Generation Company, LLC (Exelon) submitted a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1 (TMI Unit 1). The proposed amendment would modify TMI Unit 1 TS by relocating selected Surveillance Requirement frequencies to a licensee-controlled program. The NRC reviewed the license amendment request and identified the need for additional information in order to complete their evaluation of the amendment request. On June 17, 2010, draft questions were sent to Exelon to ensure that the questions were understandable, the regulatory basis for the questions was clear, and to determine if the information was previously docketed. On June 23, 2010, a teleconference was held between the NRC and Exelon to further discuss the additional information requested by the NRC. In Reference 2, the NRC formally issued the request for additional information. to this letter provides a restatement of the questions along with Exelon's responses.

In addition, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -

RITSTF [Risk-Informed Technical Specifications Task Force] Initiative 5b," dated March 18, 2009, provided an optional insert to existing TS Bases to facilitate adoption of the TSTF traveler.

The TSTF-425 TS Bases insert states as follows:

Response to Request for Additional Information LAR

- Adoption of TSTF-425, Revision 3 Docket No. 50-289 July 29, 2010 Page 2 The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

Recently several licensees submitting license amendment requests (LARs) for adoption of TSTF-425 have identified a need to deviate from this statement because it only applies to Surveillance Frequencies that have been changed in accordance with the Surveillance Frequency Control Program (SFCP) and does not apply to Surveillance Frequencies that are relocated to the SFCP but not changed. For Surveillance Frequencies relocated to the SFCP but not changed, the existing TS Bases description provides a valid description of the bases for the unchanged Surveillance Frequencies.

Therefore, upon implementation of the proposed change, where appropriate, the existing TS Bases information describing the bases for the Surveillance Frequencies will be relocated to the SFCP. This will ensure that the information describing the bases for unchanged Surveillance Frequencies is maintained. Also, relative to the Bases insert, Exelon proposes to replace the TSTF-425 Bases insert specified above with a revised insert that reads The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program, as indicated on revised proposed TS/Bases pages provided in Attachment 2.

Exelon has concluded that the information provided in this response does not impact the conclusions provided in the original submittal (Reference 1).

This response to the request for additional information contains no regulatory commitments.

If you have any questions or require additional information, please contact Glenn Stewart at 610-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the day of July 2010.

Respectfully, David P. Helker Manager, Licensing & Regulatory Affairs Exelon Generation Company, LLC : Response to Request for Additional Information : Revised Proposed Technical Specifications/Bases Pages cc:

Regional Administrator

- NRC Region I w/attachments NRC Senior Resident Inspector TMI Unit 1 NRC Project Manager, NRR TMI Unit 1 Director, Bureau of Radiation Protection PA Department of Environmental Resources Chairman, Board of County Commissioners of Dauphin County Chairman, Board of Supervisors of Londonderry Township Response to Request for Additional Information LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-289 July 29, 2010 Page 2 "The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program."

Recently several licensees SUbmitting license amendment requests (LARs) for adoption of TSTF-425 have identified a need to deviate from this statement because it only applies to Surveillance Frequencies that have been changed in accordance with the Surveillance Frequency Control Program (SFCP) and does not apply to Surveillance Frequencies that are relocated to the SFCP but not changed. For Surveillance Frequencies relocated to the SFCP but not changed, the existing TS Bases description provides a valid description of the bases for the unchanged Surveillance Frequencies.

Therefore, upon implementation of the proposed change, where appropriate, the existing TS Bases information describing the bases for the Surveillance Frequencies will be relocated to the SFCP. This will ensure that the information describing the bases for unchanged Surveillance Frequencies is maintained. Also, relative to the Bases insert, Exelon proposes to replace the TSTF-425 Bases insert specified above with a revised insert that reads "The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program," as indicated on revised proposed TS/Bases pages provided in Attachment 2.

Exelon has concluded that the information provided in this response does not impact the conclusions provtded in the original submittal (Reference 1).

This response to the request for additional information contains no regulatory commitments.

If you have any questions or require additional information, please contact Glenn Stewart at 610-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 29th day of July 2010.

Respectfully, David P. Helker Manager, Licensing & Regulatory Affairs Exelon Generation Company, LLC : Response to Request for Additional Information : Revised Proposed Technical Specifications/Bases Pages cc:

Regional Administrator - NRC Region I NRC Senior Resident Inspector - TMI Unit 1 NRC Project Manager, NRR - TMI Unit 1 Director, Bureau of Radiation Protection - PA Department of Environmental Resources Chairman, Board of County Commissioners of Dauphin County Chairman, Board of Supervisors of Londonderry Township w/attachments

ATTACHMENT 1 License Amendment Request Three Mile Island Nuclear Station, Unit 1 Docket No. 50-289 Application for Technical Specification Change Regarding Risk Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)

Response to Request for Additional Information ATTACHMENT 1 License Amendment Request Three Mile Island Nuclear Station, Unit 1 Docket No. 50-289 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)

Response to Request for Additional Information

Docket No. 50-289 Page 1 of 9 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION APPLICATION FOR TECHNICAL SPECIFICATION CHANGE REGARDING RISK INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM (ADOPTION OF TSTF-425, REVISION 3)

In Reference 1, Exelon Generation Company, LLC (Exelon) submitted a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1 (TMI Unit 1). The proposed amendment would modify TMI Unit 1 TS by relocating selected Surveillance Requirement frequencies to a licensee-controlled program. The NRC reviewed the license amendment request and identified the need for additional information in order to complete their evaluation of the amendment request. On June 17, 2010, draft questions were sent to Exelon to ensure that the questions were understandable, the regulatory basis for the questions was clear, and to determine if the information was previously docketed. On June 23, 2010, a teleconference was held between the NRC and Exelon to further discuss the additional information requested by the NRC.

In Reference 2, the NRC formally issued the request for additional information (RAI).

The questions are restated below along with Exelons responses.

RAI-1

The LAP states that the changes presented are consistent with TSTF-425 and also includes a discussion of the differences in the application that result primarily from the custom TMI-1 TSs as compared to the STSs presented in TSTF-425 and NUREG-1430. The LAP, Attachment 4, TSTF-425 (NUREG-1430) vs. TMI Unit 1 Cross-Reference, is provided to aid in the determination of consistency of the surveillances proposed for relocation as compared to TSTF 425. In order to verify that the surveillances proposed for relocation are consistent with TSTF 425 as the LAP asserts, the NRC staff requests that the licensee provide corresponding TSTF 425 cross references for the following surveillance frequencies proposed for relocation: Table 4.1-1, Instrument Surveillance Requirements, Channel Description Nos. 11, 15, 17, 19e, 19f, 45, and 46.

RESPONSE

The corresponding TSTF-425 cross-references for the specified TMI TS Table 4.1-1 instrument channel descriptions are provided in the table below.

TMI TS Table 4.1-1 TSTF4251 NUREG-1 430 Comments Item Description Equivalent 11 Reactor Coolant Pressure-SR 3.3.1.1 STS Table 3.3.1-1, Item 5 Temperature Comparator SR 3.3.1.4 SR 3.3.1.5 15 High Pressure Injection Analog SR 3.3.5.1 STS Table 3.3.5-1, Items 1 & 2 Channels SR 3.3.5.2 SR 3.3.5.3 17 Low Pressure Injection Analog SR 3.3.5.1 STS Table 3.3.5-1, Items 1 & 2 Channels SR 3.3.5.2 SR 3.3.5.3 Docket No. 50-289 Page 1 of 9 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION APPLICATION FOR TECHNICAL SPECIFICATION CHANGE REGARDING RISK-INFORMED JUSTIFICATION FOR THE RELOCATION OF SPECIFIC SURVEILLANCE FREQUENCY REQUIREMENTS TO A LICENSEE CONTROLLED PROGRAM (ADOPTION OF TSTF-425, REVISION 3)

In Reference 1, Exelon Generation Company, LLC (Exelon) submitted a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1 (TMI Unit 1). The proposed amendment would modify TMI Unit 1 TS by relocating selected Surveillance Requirement frequencies to a licensee-controlled program. The NRC reviewed the license amendment request and identified the need for additional information in order to complete their evaluation of the amendment request. On June 17, 2010, draft questions were sent to Exelon to ensure that the questions were understandable, the regulatory basis for the questions was clear, and to determine if the information was previously docketed. On June 23, 2010, a teleconference was held between the NRC and Exelon to further discuss the additional information requested by the NRC. In Reference 2, the NRC formally issued the request for additional information (RAI).

The questions are restated below along with Exelon's responses.

The LAR states that the changes presented are consistent with TSTF-425 and also includes a discussion of the differences in the application that result primarily from the custom TMI-1 TSs as compared to the STSs presented in TSTF-425 and NUREG-1430. The LAR, Attachment 4, "TSTF-425 (NUREG-1430) vs. TMI Unit 1 Cross-Reference," is provided to aid in the determination of consistency of the surveillances proposed for relocation as compared to TSTF-425. In order to verify that the surveillances proposed for relocation are consistent with TSTF-425 as the LAR asserts, the NRC staff requests that the licensee provide corresponding TSTF-425 cross references for the following surveillance frequencies proposed for relocation: Table 4.1-1, "Instrument Surveillance Requirements," Channel Description Nos. 11, 15, 17, 1ge, 19f, 45, and 46.

RESPONSE

The corresponding TSTF-425 cross-references for the specified TMI TS Table 4.1-1 instrument channel descriptions are provided in the table below.

TMI TS Table 4.1-1 TSTF-425/

NUREG-1430 Comments Item Description Equivalent 11 "Reactor Coolant Pressure-SR 3.3.1.1 STS Table 3.3.1-1, Item 5 Temperature Comparator" SR 3.3.1.4 SR 3.3.1.5 15 "High Pressure Injection Analog SR 3.3.5.1 STS Table 3.3.5-1, Items 1 & 2 Channels" SR 3.3.5.2 SR 3.3.5.3 17 "Low Pressure Injection Analog SR 3.3.5.1 STS Table 3.3.5-1, Items 1 & 2 Channels" SR 3.3.5.2 SR 3.3.5.3

Response to Request for Additional Information LAR

- Adoption of TSTF-425, Revision 3 Page 2 of 9 Docket No. 50-289 TMI TS Table 4 TSTF-425/

NUREG-1 430 Comments Item Description Equivalent 19e Reactor Bldg. Purge Line High SR 3.3.15.1 Radiation (AH-V-1AJD)

SR 3.3.15.2 SR 3.3.15.3 19f Line break isolation signal (ICCW SR 3,3,5.1 Line break isolation is a

& NSCCW)

SR 3.3.5.2 diverse method for Reactor SR 3.3.5.3 Building Isolation. This is a TMI-specific signal which is redundant to a signal on Reactor Building high pressure and accomplishes the same function as STS Table 3.3.5-1, Item 3 45 Loss of Feedwater Reactor Trip SR 3.3.1.1 STS Table 3.3.1-1, Item 10 SR 3.3.1.4 SR 3.3.1.5 46 Turbine Trip / Reactor Trip SR 3.3.1.1 STS Table 3.3.1-1, Item 9 SR 3.3.1.4 SR 3.3.1.5

RAI-2

With reference to the LAR, Attachment 2, Table 2-1, each of the findings in the following table identified an issue or gap that, individually, might not significantly impact the results from a surveillance test interval (STI) risk evaluation performed via the NEI 04-10 methodology, but, when taken cumulatively, could prove significant. The NRC staffs concern associated with each is highlighted in italics. Please address whether, when taken cumulatively, their effects could prove significant to the risk evaluation for an STI TS change and, if not, why not.

RESPONSE

Subsequent to the LAR submittal, several of the gaps identified in this RAI were addressed and resolved. The following gaps have been resolved as described in the table below:

lE-A5-01 lE-A7-01 LE-E4-01 Additionally, responses for the following three gaps are provided in the table below:

lE-A4a-01 QU-D5-01 SC-C2-01 Based on the discussions provided in the table, these gaps are still considered to not impact the results of an STI evaluation.

Response to Request for Additional Information LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-289 Page 2 of 9 TMI TS Table 4.1-1 TSTF-425/

NUREG-1430 Comments Item Description Equivalent tse "Reactor Bldg. Purge Line High SR 3.3.15.1 Radiation (AH-V-1A1D)"

SR 3.3.15.2 SR 3.3.15.3 191 "Line break isolation signal (ICCW SR 3.3.5.1 Line break isolation is a

& NSCCW)"

SR 3.3.5.2 diverse method for Reactor SR 3.3.5.3 Building Isolation. This is a TMI-specific signal which is redundant to a signal on Reactor Building high pressure and accompl ishes the same function as STS Table 3.3.5-1, Item 3 45 "Loss of Feedwater Reactor Trip" SR 3.3.1.1 STS Table 3.3.1-1, Item 10 SR 3.3.1.4 SR 3.3.1.5 46 "Turbine Trip / Reactor Trip" SR 3.3.1.1 STS Table 3.3.1-1, Item 9 SR 3.3.1.4 SR 3.3.1.5 With reference to the LAR, Attachment 2, Table 2-1, each of the findings in the following table identified an issue or gap that, individually, might not significantly impact the results from a surveillance test interval (STI) risk evaluation performed via the NEI 04-10 methodology, but, when taken cumulatively, could prove significant. The NRC staff's concern associated with each is highlighted in italics. Please address whether, when taken cumulatively, their effects could prove significant to the risk evaluation for an STI TS change and, if not, why not.

RESPONSE

Subsequent to the LAR submittal, several of the gaps identified in this RAI were addressed and resolved. The following gaps have been resolved as described in the table below:

IE-A5-01 IE-A7-01 LE-E4-01 Additionally, responses for the following three gaps are provided in the table below:

IE-A4a-01 QU-D5-01 SC-C2-01 Based on the discussions provided in the table, these gaps are still considered to not impact the results of an STI evaluation.

Response to Request for Additional Information LAR

- Adoption of TSTF-425, Revision 3 Page 3 of 9 Docket No. 50-289 A sensitivity calculation was performed to address LE-C8a-01. The sensitivity shows that there is no impact on the base model results, but additional sensitivities will be performed, if necessary, to support specific STI evaluations. Also, for QU-F5-01, the technical adequacy associated with this gap is accounted for in the NEI-04-10 process (see discussion in table).

This leaves only gap IE-A6-01 as not addressed; however, this is not expected to have an impact as described in the table below.

Since three of the gaps identified in the RAI are resolved, there are six open gaps remaining; four of these have no impact, and two will be addressed by sensitivities required by the NEI 04-10 methodology. As a result, there is no cumulative impact of these open gaps.

Response to Request for Additional Information LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-289 Page 3 of 9 A sensitivity calculation was performed to address LE-C8a-01. The sensitivity shows that there is no impact on the base model results, but additional sensitivities will be performed, if necessary, to support specific STI evaluations. Also, for QU-F5-01, the technical adequacy associated with this gap is accounted for in the NEI-04-10 process (see discussion in table).

This leaves only gap IE-A6-01 as not addressed; however, this is not expected to have an impact as described in the table below.

Since three of the gaps identified in the RAI are resolved, there are six open gaps remaining; four of these have no impact, and two will be addressed by sensitivities required by the NEI 04-10 methodology. As a result, there is no cumulative impact of these open gaps.

Response to Request for Additional Information LAR

- Adoption of TSTF-425, Revision 3 Page 4 of 9 Docket No. 50-289 Finding Issue/Gap Status of Issue/Gap lE-A4a-01 The potential for common cause failures [CCF5] was The text of the comment provided for IE-A4a-01 in the included in examination of potential initiating events LAR was misleading. In fact, the examination for resulting from the systematic evaluation for potential potential initiating events did include common cause initiating events. As recommended per (Regulatory failures from routine system alignments that could Guide] RG 1.200, Rev. 2, for Supporting Requirement result from preventive or corrective maintenance.

(SR) IE-A6 (Capability Category [(CC)]-ll), this Therefore, the italicized item is not an issue in the examination should also include CCFs from routine performance of STI evaluations.

system alignments that could result from preventive and corrective maintenance.

IE-A5-01 No documentation was found of incorporating: (a)

Subsequent to the LAR submittal, a new review was events that have occurred at conditions other than at-performed and documented for events meeting either power operation (i.e., during low-power or shutdown (a> or (b) in SR IE-A7, The review covered events conditions), and for which it is determined that the from January 1, 1990 to December 31, 2009. No new event could also occur during at-power operation; (b) initiators were identified from this review. Therefore, events resulting in a controlled shutdown that this gap is resolved.

includes a scram prior to reaching low-power conditions, unless it is determined that an event is not applicable to at-power operation. SR IE-A7 requires that, even if not documented, these events have to be incorporated.

Response to Request for Additional Information LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-289 Page 4 of 9 Finding Issue/Gap Status of Issue/Gap IE-A4a-01 "The potential for common cause failures [CCFs] was The text of the comment provided for IE-A4a-01 in the included in examination of potential initiating events LAR was misleading. In fact, the examination for resulting from the systematic evaluation for potential potential initiating events did include common cause initiating events." As recommended per [Regulatory failures from routine system alignments that could Guide] RG 1.200, Rev. 2, for Supporting Requirement result from preventive or corrective maintenance.

(SR) IE-A6 (Capability Category [(CC)]-II), this Therefore, the italicized item is not an issue in the examination should also include CCFs from routine performance of STI evaluations.

system alignments that could result from preventive and corrective maintenance.

IE-A5-01 "No documentation was found of incorporating: (a)

Subsequent to the LAR submittal, a new review was events that have occurred at conditions other than at-performed and documented for events meeting either power operation (Le., during low-power or shutdown (a) or (b) in SR IE-A?

The review covered events conditions), and for which it is determined that the from January 1, 1990 to December 31,2009. No new event could also occur during at-power operation; (b) initiators were identified from this review. Therefore, events resulting in a controlled shutdown that this gap is resolved.

includes a scram prior to reaching low-power conditions, unless it is determined that an event is not applicable to at-power operation." SR IE-A7 requires that, even if not documented, these events have to be incorporated.

Response to Request for Additional Information LAR

- Adoption of TSTF-425, Revision 3 Page 5 of 9 Docket No. 50-289 Finding Issue/Gap Status of Issue/Gap lE-A6-01 No documentation was found of interviews with plant Subsequent to the LAR submittal, a new review was and personnel (e.g., operations, maintenance, performed and documented for precursor events. The lE-A7-01 engineering, safety analysis) to determine if potential review covered events from January 1, 1990 to initiating events have been overlooked

... No December 31, 2009. No new initiators were identified documentation of the review of plant-specific from this review. Therefore, gap lE-A7-01 is resolved, operating experience for initiating event precursors was found in the [probabilistic risk assessment] PRA Recent interviews with plant personnel (lE-A6-01) are notebooks. Even if not documented, CC-Il for both still outstanding. Based on completion and of these SRs requires that the interviews (SR I&A8 documentation of the review of plant-specific operating

[CC-Il], with finding IE-A6-O1) and reviews (SR lE-A9 experience for precursors, previous (undocumented)

[CC-Il], with finding lE-A7-O1) have been conducted.

plant personnel interviews, and other initiating event identification methods used for the TM! PRA, the likelihood of plant personnel interviews identifying additional potential plant-specific initiating events is low.

SC-C2-01 SR SC-C2 requires that, even if not documented (or For success criteria that were developed for the PRA, else still in the process of being documented),

generally MAAP4 is used instead of using design computer code limitations or potential basis success criteria. The overall conclusion from conservatisms have to be addressed.

the EPRI MAAP Thermal-Hydraulic Qualification Studies was that MAAP had a wide range of applicability; however, a few limitations were identified.

The current position on MAAP code limitations can be found on the MAAP4 web site. The significant limitation of MAAP for PWRs is Large LOCA behavior prior to reflood. The TM! PRA uses design basis criteria for Large LOCAs, so this limitation of MAAP4 has been addressed.

Response to Request for Additional Information LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-289 Page 5 of 9 Finding Issue/Gap Status of Issue/Gap IE-A6-01 "No documentation was found of interviews with plant Subsequent to the LAR submittal, a new review was and personnel (e.g., operations, maintenance, performed and documented for precursor events. The IE-A7-01 engineering, safety analysis) to determine if potential review covered events from January 1, 1990 to initiating events have been overlooked... No December 31, 2009. No new initiators were identified documentation of the review of plant-specific from this review. Therefore, gap IE-A7-01 is resolved.

operating experience for initiating event precursors was found in the [probabilistic risk assessment] PRA Recent interviews with plant personnel (IE-A6-01) are notebooks." Even if not documented, CC-II for both still outstanding. Based on completion and of these SRs requires that the interviews (SR IE-AB documentation of the review of plant-specific operating

[CC-II], with finding IE-A6-01) and reviews (SR IE-A9 experience for precursors, previous (undocumented)

[CC-II], with finding IE-A7-01) have been conducted.

plant personnel interviews, and other initiating event identification methods used for the TMI PRA, the likelihood of plant personnel interviews identifying additional potential plant-specific initiating events is low.

SC-C2-01 SR SC-C2 requires that, even if not documented (or For success criteria that were developed for the PRA, else still in the process of being documented),

generally MAAP4 is used instead of using design computer code "limitations or potential basis success criteria. The overall conclusion from conservatisms" have to be addressed.

the EPRI MAAP Thermal-Hydraulic Qualification Studies was that MAAP had a wide range of applicability; however, a few limitations were identified.

The current position on MAAP code limitations can be found on the MAAP4 web site. The significant limitation of MAAP for PWRs is Large LOCA behavior prior to reflood. The TMI PRA uses design basis criteria for Large LOCAs, so this limitation of MAAP4 has been addressed.

Response to Request for Additional Information LAR

- Adoption of TSTF-425, Revision 3 Page 6 of 9 Docket No. 50-289 Finding Issue/Gap Status of Issue/Gap QU-D5-01 Some SSCs [structures, systems, and components]

Significant contributors to initiating events were that are significant contributors to initiating events, but identified through a review of support system initiating not to mitigation, are not explicitly identified in the event cutsets, but the individual contributors and documentation of significant contributors CC-Il for cutsets were omitted from the quantification notebook.

SR QU-D6, against which this finding is cited, It should be noted that initiating event fault trees are requires that significant contributors to core damage re-quantified for any application affecting the frequency, including initiating events, and SSCs and components or configurations represented by these operator actions that contribute to initiating event fault trees, frequencies, be identified. While not explicitly identified in the documentation, were these significant contributors to initiating events actually identified butjust omitted from the documentation? If they were not identified, how were they known to be significant and to what extent?

QU-F5-01

[Q]ther than the [large early release frequency] LERF LERF truncation is the only identified limitation to the truncation limitation, no evaluations of limitations TMI PRA model for applications. Additional limitations were presented

, [including] limitations of the model may exist (e.g., STl components not modeled in the as they may apply to applications. As implied by SR PRA), but the NEI 04-10 process (Step 8) requires an QU-F5, these limitations need to have been assessment of whether the STI change can be addressed.

adequately characterized by the PRA.

Response to Request for Additional Information LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-289 Page 6 of 9 Finding Issue/Gap Status of Issue/Gap QU-D5-01 "Some SSCs [structures, systems, and components]

Significant contributors to initiating events were that are significant contributors to initiating events, but identified through a review of support system initiating not to mitigation, are not explicitly identified in the event cutsets, but the individual contributors and documentation of significant contributors." CC-II for cutsets were omitted from the quantification notebook.

SR QU-D6, against which this finding is cited, It should be noted that initiating event fault trees are requires that significant contributors to core damage re-quantified for any application affecting the frequency, including initiating events, and SSCs and components or configurations represented by these operator actions that contribute to initiating event fault trees.

frequencies, be identified. While "not explicitly identified" in the documentation, were these significant contributors to initiating events actually identified but just omitted from the documentation? If they were not identified, how were they known to be significant and to what extent?

QU-F5-01

"[O]ther than the [large early release frequency] LERF LERF truncation is the only identified limitation to the truncation limitation, no evaluations of limitations TMI PRA model for applications. Additional limitations were presented..., [including] limitations of the model may exist (e.g., STI components not modeled in the as they may apply to applications." As implied by SR PRA), but the NEI 04-10 process (Step 8) requires an QU-F5, these limitations need to have been assessment of whether the STI change can be addressed.

adequately characterized by the PRA.

Response to Request for Additional Information LAR

- Adoption of TSTF-425, Revision 3 Page 7 of 9 Docket No. 50-289 Finding Issue/Gap Status of Issue/Gap LE-C8a-01 The Reactor Building fan coolers are undersized at In response to this RAI, a sensitivity analysis was TMI and have a little to no impact on containment performed to determine the impact on the base

[CNMTj pressure and temperature with respect to (average) PRA model assuming the Reactor Building early containment failure. SR LE-C9 (CC-Il) requires fan coolers were not available following core damage.

justification for any credit taken for equipment There was no change to the LERF results (i.e.,

survivability under adverse environmental conditions identical large early release cutsets and frequency).

It such that, even if the fan coolers were assumed to be is expected that this conclusion will be the same for failed, there would be little to no impact on CNMT most applications. However, there is still a potential pressure and temperature with respect to early CNMT that the assumption for the Reactor Building fan failure.

coolers surviving adverse environmental conditions may impact a specific STI evaluation being performed.

To address this potential, the comment for this gap in the LAR states that it will be evaluated via a sensitivity analysis per NEI 04-10, if applicable to the STI.

LE-E4-01 The level 2 results with the flag file are expected to Subsequent to the LAR submittal, this F&Q/gap was be conservative. When the cutsets were reviewed, it resolved. A test was performed that was similar to that was determined that there appears to be non-minimal done by the peer reviewer. It determined that the cutsets in the level 2 model as quantified without the reason for the higher FTREX results is because of the flag file

... Some sensitivities have been performed, way that CAFTA calculates the total value of cutsets although a conclusive determination has not been using Mm Cut Upper Bound. The LERF results from made regarding the current method for quantifying FTREX without using the flag file have a significant LERF... ({T]he TMI model uses Forte 3.Oc as the number of events greater than or equal to 0.9. Using quantifier). SR LE-E4 requires that LERFbe the EPRI Acube (beta) software, it was shown that the quantified consistently as with core damage sum of the cutset values calculated without the flag file frequency. This implies that the LERF quantification was less than when using the flag file. This is the result be conclusively determined as conservative, e.g., by expected. Therefore, the method utilizing the flag file quantifying LERF using Forte 3.Oc at a greater is conservative and acceptable.

truncation value just to assess whether the use of the flag file produces conservative results.

Response to Request for Additional Information LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-289 Page 7 of 9 Finding Issue/Gap Status of Issue/Gap LE-C8a-01 "The Reactor Building fan coolers are undersized at In response to this RAI, a sensitivity analysis was TMI and have a little to no impact on containment performed to determine the impact on the base

[CNMT] pressure and temperature with respect to (average) PRA model assuming the Reactor Building early containment failure." SR LE-C9 (CC-II) requires fan coolers were not available following core damage.

justification for any credit taken for equipment There was no change to the LERF results (Le.,

survivability under adverse environmental conditions identical large early release cutsets and frequency). It such that, even if the fan coolers were assumed to be is expected that this conclusion will be the same for failed, there would be "little to no impact" on CNMT most applications. However, there is still a potential pressure and temperature with respect to early CNMT that the assumption for the Reactor Building fan failure.

coolers surviving adverse environmental conditions may impact a specific STI evaluation being performed.

To address this potential, the comment for this gap in the LAR states that it will be evaluated via a sensitivity analysis per NEt 04-10, if applicable to the STI.

LE-E4-01 "The level 2 results with the flag file are expected to Subsequent to the LAR submittal, this F&O/gap was be conservative. When the cutsets were reviewed, it resolved. A test was performed that was similar to that was determined that there appears to be non-minimal done by the peer reviewer. It determined that the cutsets in the level 2 model as quantified without the reason for the higher FTREX results is because of the flag file... Some sensitivities have been performed, way that CAFTA calculates the total value of cutsets although a conclusive determination has not been using Min Cut Upper Bound. The LERF results from made regarding the current method for quantifying FTREX without using the flag file have a significant LERF... ([T]he TMI model uses Forte 3.0c as the number of events greater than or equal to 0.9. Using quantifier)." SR LE-E4 requires that LERF be the EPRI Acube (beta) software, it was shown that the quantified consistently as with core damage sum of the cutset values calculated without the flag file frequency.

This implies that the LERF quantification was less than when using the flag file. This is the result be conclusively determined as conservative, e.g., by expected. Therefore, the method utilizing the flag file quantifying LERF using Forte 3.0c at a greater is conservative and acceptable.

truncation value just to assess whether the use of the flag file produces conservative results.

Response to Request for Additional Information LAR

- Adoption of TSTF-425, Revision 3 Page 8 of 9 Docket No. 50-289

RAI-3

With reference to the LAR, Attachment 2, Table 2-2, Finding DA-B2-01 states: There is no evidence that the intent of this SR was met. Although the component failure rates are grouped by system and component type, that does not guarantee that outliers are not included in a group. SR DA-B2 (CC-Il> requires exclusion of outliers in the definition of system/component failure groups. Were outliers appropriately excluded from group definitions?

If not, will their exclusion be part of the sensitivity analysis for an STI evaluation?

RESPONSE

A review of the component grouping has subsequently been performed. There is no indication of outliers due to testing or operational characteristics (except potentially for manual valves), nor due to poor performance of certain components or systems. Operational characteristics (normal position and frequency of manipulation) for manual valves was not taken into account (i.e., for failure rate purposes, all manual valves were grouped together). However, the risk significance of manual valves is negligible; therefore, no impact on the results would be expected if they were grouped differently.

RAI-4

With reference to LAR, Attachment 2, Table 2-2, Finding IFEV-A5-01 states: Several requirements in establishing flood initiating event frequencies are not met. Specifically cited are SRs IFEV-A5 through IFEV-A7, which require inclusion of plant-specific information and consideration of human-induced floods during maintenance (CC-Il). Are any of the valves that may be assigned new STIs potential flooding sources, such that increasing the STI could increase the frequency of a flood due to miscalibration, etc., of one of these valves?

RESPONSE

Which valves, if any, are assigned new STI5 using the NEI-04-10 process is unknown at this time. However, as indicated in the LAR submittal for this item, the methodology requires sensitivities for assumptions in the PRA model that may affect the results of the analysis or of any gaps to Capability Category II. This would lead to these issues being appropriately addressed for any valves associated with a surveillance interval change analysis.

REFERENCES:

1.

Letter from Pamela B. Cowan, Exelon Generation Company, LLC, to U.S. Nuclear Regulatory Commission, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3),

dated March 24, 2010.

Response to Request for Additional Information LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-289 Page 8 of 9 With reference to the LAR, Attachment 2, Table 2-2, Finding DA-B2-01 states: "There is no evidence that the intent of this SR was met. Although the component failure rates are grouped by system and component type, that does not guarantee that outliers are not included in a group." SR DA-B2 (CC-II) requires exclusion of outliers in the definition of system/component failure groups. Were outliers appropriately excluded from group definitions? If not, will their exclusion be part of the sensitivity analysis for an STI evaluation?

RESPONSE

A review of the component grouping has subsequently been performed. There is no indication of outliers due to testing or operational characteristics (except potentially for manual valves), nor due to poor performance of certain components or systems. Operational characteristics (normal position and frequency of manipulation) for manual valves was not taken into account (Le., for failure rate purposes, all manual valves were grouped together). However, the risk significance of manual valves is negligible; therefore, no impact on the results would be expected if they were grouped differently.

With reference to LAR, Attachment 2, Table 2-2, Finding IFEV-A5-01 states: "Several requirements in establishing flood initiating event frequencies are not met." Specifically cited are SRs IFEV-A5 through IFEV-A7, which require inclusion of plant-specific information and consideration of human-induced floods during maintenance (CC-II). Are any of the valves that may be assigned new STls potential flooding sources, such that increasing the STI could increase the frequency of a flood due to miscalibration, etc., of one of these valves?

RESPONSE

Which valves, if any, are assigned new STls using the NEI-04-10 process is unknown at this time. However, as indicated in the LAR submittal for this item, the methodology requires sensitivities for assumptions in the PRA model that may affect the results of the analysis or of any gaps to Capability Category II. This would lead to these issues being appropriately addressed for any valves associated with a surveillance interval change analysis.

REFERENCES:

1.

Letter from Pamela B. Cowan, Exelon Generation Company, LLC, to U.S. Nuclear RegUlatory Commission, "Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3),"

dated March 24, 2010.

Response to Request for Additional Information LAR

- Adoption of TSTF-425, Revision 3 Page 9 of 9 Docket No. 50-289 2.

Letter from Peter Bamford, U.S. Nuclear Regulatory Commission, to Michael J. Pacilio, Exelon Nuclear, Three Mile Island Nuclear Station

• Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Relocation of Surveillance Frequencies to a Licensee Controlled Program (TAO No. ME3587), dated July 2, 2010.

Response to Request for Additional Information LAR - Adoption of TSTF-425, Revision 3 Docket No. 50-289 Page 9 of 9 2.

Letter from Peter Bamford, U.S. Nuclear Regulatory Commission, to Michael J. Pacilio, Exelon Nuclear, "Three Mile Island Nuclear Station - Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Relocation of Surveillance Frequencies to a Licensee Controlled Program (TAC No. ME3587)," dated July 2,2010.

ATTACHMENT 2 License Amendment Request Three Mile Island Nuclear Station, Unit 1 Docket No. 50-289 Application for Technical Specification Change Regarding Risk Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)

Revised Proposed Technical Specifications/Bases Pages 4-2a 4-47 ATIACHMENT2 License Amendment Request Three Mile Island Nuclear Station, Unit 1 Docket No. 50-289 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)

Revised Proposed Technical Specifications/Bases Pages 4-2a 4-47

Bases (Contd)

The 600 ppmb limit in Item 4, Table 4.1-3 is used to meet the requirements of Section 5.4. Under other circumstances the minimum acceptable boron concentration would have been zero ppmb.

Calibration Calibration shall be performed to assure the presentation and acquisition of accurate information.

The nuclear flux (power range) channels amplifiers shall be checked in accordance with the Surveillance Frequency Control Program against a heat balance standard and calibrated if necessary, every shift against-a heat balance standard. The frequency of heat balance checks will assure that the difference between the out-of-core instrumentation and the heat balance remains less than 4%.

Channels subject only to drift errors induced within the instrumentation itself can tolerate longer intervals between calibrations. Process system instrumentation errors induced by drift can be expected to remain within acceptance tolerances if recalibration is performed at the intervals of each refueling periodspecified in the Surveillance Frequency Control Program.

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

Thus, minimum calibration frequencies set forth in the Surveillance Frequency Control Program are considered acceptable.

Testing On-line testing of reactor protection channels is required semi annually in accordance with the Surveillance Frequency Control Program on a rotational basis. The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systematic test errors being introduced into each redundant channel (Reference 1). Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.

The rotation schedule for the reactor nrptection channel-s is as follows:

a)

Deleted b)

Semi-annually with one channel being tested every 16 days on a WIIWIUOUS sequential rotation.

The reactor protection system instrumentation test cycle is continued with one channels instrumentation tested every 46 days. The frequency of every 46 days on a continuous sequential rotation is consistent with the calculations of Reference 2 that indicate-the RPS retains a high level of reliability for this Upon detection of a failure that prevents trip action in a channel, the instrumentation associated with the protection parameter failure will be tested in the remaining channels. If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting.

The protection channels coincidence logic, the control rod drive trip breakers and the regulating control rod power SCRs electronic trips, are trip tested in accordance with the Surveillance Frequency Control Programquarterly with one channel being tested every 23 days on a continuous sequential rotation. Calculations have shown that the frequency of every 23 days maintains a high level of reliability of the Reactor Trip System in Reference 4. The trip test checks all logic combinations and is to be performed on a rotational basis.

Discovery of a failure that prevents trip action requires the testing of the instrumentation associated with the protection parameter failure in the remaining channels.

For purposes of surveillance, reactor trip on loss of feedwater and reactor trip on turbine trip are considered reactor protection system channels.

4-2a Amendment No. 78, 157, 181, 200, 216, 255 Bases (Cont'd)

The 600 ppmb limit in Item 4, Table 4.1-3 is used to meet the requirements of Section 5.4. Under other circumstances the minimum acceptable boron concentration would have been zero ppmb.

Calibration Calibration shall be performed to assure the presentation and acquisition of accurate information.

The nuclear flux (power range) channels amplifiers shall be checked In accordance with the Surveillance Frequency Control Program against a heat balance standard and calibrated if necessary, every shift against a heat balance standard. The frequency of heat balance checks will assure that the difference between the out-of-core instrumentation and the heat balance remains less than 4%.

Channels subject only to "drift" errors induced within the instrumentation itself can tolerate longer intervals between calibrations. Process system instrumentation errors induced by drift can be expected to remain within acceptance tolerances if recalibration is performed at the intervals of each refueling periodspecified in the Surveillance Frequency Control Program.

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

Thus, minimum calibration frequencies set forth in the Surveillance Frequency Control Program are considered acceptable.

Testing On-line testing of reactor protection channels is required semi annually in accordance with the Surveillance Frequency Control Program on a rotational basis. The rotation scheme is designed to reduce the probability of an undetected failure existing within the system and to minimize the likelihood of the same systematic test errors being introduced into each redundant channel (Reference 1). Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.

The rotation schedule for the roastor protestion channels is as follows:

a)

Deleted b)

Semi annually with one channel being tested every 46 days on a continuous sequential rotation.

The reactor protection system instrumentationtest cycle is continued with one channel's instrumentation tosted e'tfory 46 days. The frequency of every 46 days on a continuous soquontial rotation is consistent

'....ith the calculations of Reforence 2 that indicate the RPS retains a high level of reliability for this interval.

Upon detection of a failure that prevents trip action in a channel, the instrumentation associated with the protection parameter failure will be tested in the remaining channels. If actuation of a safety channel occurs, assurance will be required that actuation was within the limiting safety system setting.

The protection channels coincidence logic, the control rod drive trip breakers and the regulating control rod power SCRs electronic trips, are trip tested in accordance with the Surveillance Frequency Control Programquarterly with one channel being tested every 23 days on a continuous sequential rotation. Calculations havo shown that the frequency of every 23 days maintains a high le'lel of reliability of the Reactor Trip System in Reference 4. The trip test checks all logic combinations and is to be performed on a rotational basis.

Discovery of a failure that prevents trip action requires the testing of the instrumentation associated with the protection parameter failure in the remaining channels.

For purposes of surveillance, reactor trip on loss of feedwater and reactor trip on turbine trip are considered reactor protection system channels.

4-2a Amendment No. 78,157,181,200,216,255

d.

The battery will be subjected to a load test en-a efue14g4r4tEwva basis in accordance with the Surveillance Frequency Control Program.

(1)

Verify battery capacity exceeds that required to meet design loads.

(2)

Any battery which is demonstrated to have less than 85% of manufacturers ratings during a capacity discharge test shall be replaced during the subsequent refueling outage.

4.6.3 Pressurizer Heaters a.

The following tests shall be conducted at least once each refueling in accordance with the Surveillance Frequency Control Program:

(1)

Pressurizer heater groups 8 and 9 shall be transferred from the normal power bus to the emergency power bus and energized. Upon completion of this test, the heaters shall be returned to their normal power bus.

(2)

Demonstrate that the pressurizer heaters breaker on the emergency bus cannot be closed until the safeguards signal is bypassed and can be closed following bypass.

(3)

Verify that following input of the Engineered Safeguards Signal, the circuit breakers, supplying power to the manually transferred loads for pressurizer heater groups 8 and 9, have been tripped.

Bases The tests specified are designed to demonstrate that one diesel generator will provide power for operation of safeguards equipment. They also assure that the emergency generator control system and the control systems for the safeguards equipment will function automatically in the event of a loss of normal a-c station service power or upon the receipt of an engineered safeguards Actuation Signal. The intent of the monthly periodic tests is to demonstrate the diesel capability to carry design basis accident (LOOP/LOCA) load. The test should not exceed the diesel 2000-hr. rating of 3000 kW. The automatic tripping of manually transferred loads, on an Engineered Safeguards Actuation Signal, protects the diesel generators from a potential overload condition. The testing frequency specified is intended to identify and permit correction of any mechanical or electrical deficiency before it can result in a system failure. The fuel oil supply, starting circuits, and controls are continuously monitored and any faults are alarmed and indicated. An abnormal condition in these systems would be signaled without having to place the diesel generators on test.

Precipitous failure of the station battery is extremely unlikely. The Ssurveillance specified is that which has been demonstrated over the years to provide an indication of a cell becoming unserviceable long before it fa3lFrequencies are controlled under the Surveillance Frequency Control Program.

The PORV has a remotely operated block valve to provide a positive shutoff capability should the relief valve become inoperable. The electrical power for both the relief valve and the block valve is supplied from an ESF power source to ensure the ability to seal this possible RCS leakage path.

The requirement that a minimum of 107 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation.

4-47 Amendment No. 78, 157, 167, 175, ECR TM 07-00119 d.

The battery will be subjected to a load test on a refueling intePJal 9a&i& in accordance with the Surveillance Frequency Control Program.

(1)

Verify battery capacity exceeds that required to meet design loads.

(2)

Any battery which is demonstrated to have less than 85% of manufacturers ratings during a capacity discharge test shall be replaced during the subsequent refueling outage.

4.6.3 Pressurizer Heaters a.

The following tests shall be conducted at least ense eash refueling in accordance with the Surveillance Frequency Control Program:

(1)

Pressurizer heater groups 8 and 9 shall be transferred from the normal power bus to the emergency power bus and energized. Upon completion of this test, the heaters shall be returned to their normal power bus.

(2)

Demonstrate that the pressurizer heaters breaker on the emergency bus cannot be closed until the safeguards signal is bypassed and can be closed following bypass.

(3)

Verify that following input of the Engineered Safeguards Signal, the circuit breakers, supplying power to the manually transferred loads for pressurizer heater groups 8 and 9, have been tripped.

Bases The tests specified are designed to demonstrate that one diesel generator will provide power for operation of safeguards equipment. They also assure that the emergency generator control system and the control systems for the safeguards equipment will function automatically in the event of a loss of normal a-c station service power or upon the receipt of an engineered safeguards Actuation Signal. The intent of the monthly periodic tests is to demonstrate the diesel capability to carry design basis accident (LOOP/LOCA) load. The test should not exceed the diesel 2000-hr. rating of 3000 kW. The automatic tripping of manually transferred loads, on an Engineered Safeguards Actuation Signal, protects the diesel generators from a potential overload condition. The testing frequency specified is intended to identify and permit correction of any mechanical or electrical deficiency before it can result in a system failure. The fuel oil supply, starting circuits, and controls are continuously monitored and any faults are alarmed and indicated. An abnormal condition in these systems would be signaled without having to place the diesel generators on test.

Precipitous failure of the station battery is extremely unlikely. The Ssurveillance spesified is that 'Nhish has been demonstrated over the years to provide an indisation of a sell becoming unsePJiceable long before it fa#sFrequencies are controlled under the Surveillance Frequency Control Program.

The PORV has a remotely operated block valve to provide a positive shutoff capability should the relief valve become inoperable. The electrical power for both the relief valve and the block valve is supplied from an ESF power source to ensure the ability to seal this possible RCS leakage path.

The requirement that a minimum of 107 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation.

4-47 Amendment No. 78,157,167,175, ECR TM 0700119