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:	Three Mile Island
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

Exelon Nuclear www.exeloncorp.com 11 Exein Exelon.

Exelon Vva 22 0 0 rxelon Way Nuclear Ken neSt Karl net t Square.

Squa re. 24 PA19348 19348 10 CFR 50.90 10 50.90 TMI-10-072 TMI-1 0-072 July 29,29, 2010 U.S. Nuclear Regulatory Commission U.S.

ATTN:

ATIN: Document Control Desk Washington, D.C. D.C. 20555-0001 Three Mile Island Nuclear Station, Unit 11 Renewed Facility Operating License No. No. DPR-50 NRC Docket No. No. 50-289

Subject:

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

References:

1. Letter from Pamela B.

1. Cowan,, Exelon Generation Company, B. Cowan Company, LLC,LLC, to U.S.

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

2010 .

Bamford,, U.S. Nuclear Regulatory Commission,

2. Letter from Peter Bamford Commission, to Michael J. Pacilio, Exelon Nuclear, "ThreeThree Mile Island Nuclear Station - -

Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Relocation of Surveillance Frequencies to a Licensee ME3587), dated July 2, 2010.

Controlled Program (TAC No. ME3587),"

In Reference 1, Generation Company, LLC (Exelon) submitted a request for an 1, Exelon Generation amendment to the Technical Specifications (TS), (TS), Appendix A of Renewed Facility Operating License No. DPR-50 for Three Mile Island Nuclear Station, (TMI Unit 1).

Station, Unit 11 (TMI 1). The proposed amendment would modify TMI Unit 11 TS by relocating selected Surveillance Requirement Surveillance frequencies to a licensee licensee-controlled

-controlled program. The NRC reviewed the license amendment request and identified the need for additional information in order to complete complete their evaluation of the amendment request. On June 17, 17, 2010 2010,, draft questions were sent to Exelon to ensure that questions the quest regulatory basis for the questions was clear, and to ions were understandable, the regulatory determine if the information was previously docketed. On June 23, 2010 2010,, a teleconference was held between the NRC and Exelon to further requested by the further discuss the additional information requested NRC. In Reference 2, formally issued the request for additional 2, the NRC formally information..

additional information Attachment 1 1 to this letter provides a restatement of the questions along with Exelon's Exelons responses.

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

Relocate Surveillance -

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

facilitate adoption of the TSTF traveler.

The TSTF-425 TS Bases insert states as follows:

Response to Response to Request Request for for Additional Addit ional Information Information LAR LAR -

- Adoption Adoption of of TSTF-425, TSTF-425, Revision Revision 33 Docket No.

Docket No. 50-28950-289 July 29, July 29, 2010 2010 Page 22 Page "The Surveillance The Surveillance Frequency based on Frequency isis based experience, equipment operating experience, on operating reliability , and equipment reliability, and plant risk plant risk and and is is controlled controlled under under the the Surveillance Surveillance Frequency Frequency Control Control Program.

Program."

Recently several Recently several licensees licensees submitting SUbmitting license (LARs) for requests (LARs) amendment requests license amendment adoption of for adoption of TSTF-425 have TSTF-425 have identified identified aa need need to deviate from to deviate statement because this statement from this only applies to because it only applies it to Surveillance Frequencies Surveillance Frequencies that that have have been changed in been changed accordance with in accordance the Surveillance with the Surveillance Frequency Control Frequency Control Program Program (SFCP)

(SFCP) and and does does notnot apply apply toto Surveillance Frequencies that Surveillance Frequencies that are are relocated to to the the SFCP SFCP but but not not changed.

changed. For For Surveillance relocated to Frequencies relocated Surveillance Frequencies the SFCP to the SFCP but not but not changed, changed , the existing existing TS description provides aa valid TS Bases description description of valid description the bases for of the the unchanged unchanged Surveillance Surveillance Frequencies.

Frequencies.

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

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

1).

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

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

610-765-5529.

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

Respectfully, Respectfully, David David P. P. Helker Helker Manager, Manager, Licensing Licensing && Regulatory Regulatory Affairs Affairs Exelon Generation Exelon Generation Company, Company, LLC LLC Attachment Attachment 1: 1: Response Response to to Request Request for for Additional Information Additional Information Attachment Attachment 2: 2: Revised Revised Proposed Proposed Technical Specifications/Bases Pages Technical Specifications/Bases Pages cc:

cc: Regional Regional Administrator Administrator - NRC- Region II NRC Region w/attachments w/attachments NRC NRC Senior Senior Resident Resident Inspector Inspector - TMI Unit 11 TMI Unit NRC NRC Project Project Manager, Manager, NRR NRR - TMI TMI UnitUnit 11 Director, Director, Bureau Bureau of of Radiation Radiation Protection Protection - PA Department of PA Department of Environmental Resources Environmental Resources Chairman, Chairman, Board Board ofof County Commissioners of County Commissioners of Dauphin County Dauphin County Chairman Chairman,, Board Board ofof Supervisors Supervisors of of Londonderry Township Londonderry Township

ATTACHMENT 1 1 License Amendment Request Three Mile Island Nuclear Station, Unit 1 1

Docket No. 50-289 Application for Technical Specification Change Regarding Risk-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 Attachment 11 Page 11 of 9 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION APPLICATION FOR TECHNICAL SPECIFICATION CHANGE REGARDING RISK- 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, 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.No. DPR-50 for Three Mile Island Nuclear Station, Unit 11 (TMI Unit 1). 1). The proposed amendment would modify TMI Unit 1 1 TS by relocating selected Surveillance Requirement program.. The NRC reviewed the license amendment frequencies to a licensee-controlled program request and identified the need for additional information in in order to complete their evaluation of the amendment request. On June 17, 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, 2, the NRC formally issued the request for additional information (RAI).

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

RAI-1

The LARLAP 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-1TMI-1 TSs as compared to the STSs presented in TSTF TSTF-425

-425 and NUREG-1430. The LAR, LAP, Attachment 4, "TSTF-425 TSTF-425 (NUREG-1430) vs. vs. TMI Unit 1 1 Cross-Reference,"

Cross-Reference, is provided to aid in the determination of consistency of the surveillances proposed for relocation as compared to TSTF- TSTF 425. In In order to verify that the surveillances proposed for relocation are consistent with TSTF-TSTF 425 as the LAR LAP asserts, the NRC staff requests that the licensee provide corresponding TSTF TSTF-425 cross references for the following surveillance frequencies proposed for relocation: Table 4.1-1, "Instrument Requirements, Channel Description Nos. 11, Instrument Surveillance Requirements," 11, 15, 17, 1ge, 15, 17, 19e, 19f, 45, and 46.

RESPONSE

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

TMI TS Table 4.1-1 TSTF4251 TSTF-425/

NUREG-1 430 NUREG-1430 Comments Item Description Equivalent 11 11 "Reactor Pressure--

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

Response to Request for Additional Information Attachment 11 LAR - Adoption of TSTF-425, Revision 3

- Page 2 of 9 Docket No.No. 50-289 TMI TS Table 4 4.1-1 TSTF-425/

NUREG-1 430 NUREG-1430 Comments Item Description Equivalent 19e tse Reactor Bldg. Purge Line High "Reactor SR 3.3.15.1 3.3.15.1 (AH-V-1AJD)

Radiation (AH-V-1A1D)" SR 3.3.15.2 SR 3.3.15.3 3.3.15 .3 191 19f Line break isolation signal (ICCW "Line 3,3,5.1 SR 3.3.5.1 Line break isolation is a NSCCW)

& 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 accompl ishes the same function as STS Table 3.3.5-1, Item 3 45 Loss Trip "Loss of Feedwater Reactor Trip" SR 3.3.1.1 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" "Turbine Trip SR 3.3.1.1 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, methodology, but, when taken cumulatively, could prove significant. The NRC staff'sstaffs concern associated with italics. Please address whether, when taken cumulatively, their effects each is highlighted in italics.

could prove significant to the risk evaluation for an STI TS change and, if not,not, why not.

RESPONSE

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

• lE-A5-01 IE-A5-01

• lE-A7-01 IE-A7-01

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

Additionally,

• lE-A4a-01 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 Response Information Attachment 11 LAR LAR - Adoption

- TSTF-425, Revision 33 Adopt ion of TSTF-425, Page 33 of 99 Docket No.

No. 50-289 50-289 A sensitivity calculation was performed to address LE-C8a-01. The sensitivity shows that there A

no impact on is no is on the base model results, performed, if will be performed, results , but additional sensitivities will necessary, to support specific STI necessary, STI evaluations.

evaluations. Also, QU-F5-01 , the technical adequacy Also, for QU-F5-01, associated with this gap is is accounted for in in the NEI-04-10 process (see discussion in in table).

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

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

Response to Request for Additional Response Additional Information Attachment 11 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 IE-A4a-01 The potential for common cause failures [CCFs]

"The [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 events. As recommended per [Regulatory initiating events." (Regulatory failures from routine system alignments that could 1.200, Rev. 2, for Supporting Requirement Guide] RG 1.200, result from preventive or corrective maintenance.

(SR) IE-A6 (Capability Category [(CC)]-ll),

[(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)

"No 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 (i.e., during low-power or shutdown power operation (Le., (a> or (b) in SR IE-A?

(a) IE-A7, The review covered events conditions), and for which it is determined that the from January 1, 1, 1990 to December 31,2009.

31, 2009. No new event could also occur during at-power operation; (b) initiators were identified from this review. Therefore, events resulting inin 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." IE-A7 operation. SR IE-A 7 requires that, even if not documented, these events have to be incorporated.

Response to Request for Additional Information Response Information Attachment 11 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 IE-A6-01 No documentation was found of interviews with plant "No Subsequent to the LAR submittal, a new review was and operations, maintenance, personnel (e.g., operations, performed and documented for precursor events events.. The lE-A7-01 IE-A7-01 engineering, safety analysis) to determine if potential review covered events from January 1, 1, 1990 to initiating events have been overlooked ... No ... December 31, 2009. No new initiators were identified documentation of the review of plant-specific review.. Therefore from this review Therefore,, gap IE-A7-01 lE-A7-01 is resolved.

resolved, experience for initiating event precursors operating experience was found in the [probabilistic risk assessment] PRA (lE-A6-01) are Recent interviews with plant personnel (IE-A6-01) notebooks. Even if not documented, CC-II notebooks." CC-Il for both still outstanding. Based on completion and of these SRs requires that the interviews (SR IE-AB I&A8 documentation of the review of plant-specific operating

[CC-Il],

[CC -II], with finding IE-A6-01)

IE-A6-O1) and reviews (SR IE-A9 lE-A9 experience for precursors, previous (undocumented)

[CC-Il], with finding IE-A7-01)

[CC-II], lE-A7-O1) have been conducted.

conducted. plant personnel interviews, and other initiating event identification methods used for the TMI TM! PRA, 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 limitations or potential criteria.. The overall conclusion from basis success criteria conservatisms have to be addressed.

conservatisms" 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!TMI PRA uses design basis criteria for Large LOCAs, so this limitation of MAAP4 has been addressed.

Response to Request for Additional Information Attachment 11 LAR - Adoption of TSTF-425, Revision 3

- Page 6 of 9 No. 50-289 Docket No.

Finding Issue/Gap Status of Issue/Gap QU-D5-01 Some SSCs [structures, systems, and components]

"Some 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 contributors CC-II 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 requires re-quantified for any application affecting the frequency, including initiating events, and SSCs and represented by these components or configurations represented operator actions that contribute to initiating event trees, fault trees.

frequencies, be identified. While "notnot explicitly identified in the documentation, were these identified" significant contributors to initiating 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 [Q]ther than the [large early release frequency] LERF

"[O]ther 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

, STl components not modeled in the may exist (e.g., STI as they may apply to applications."

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 Attachment 11 LAR - Adoption of TSTF-425, Revision 3

- Page 7 of 9 No. 50-289 Docket No.

Finding Issue/Gap Status of Issue/Gap LE-C8a-01 The Reactor Building fan coolers are undersized at "The 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

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

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

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

failure. coolers surviving adverse environmental conditions performed.

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 NEI 04-10, if applicable to the STI.

analysis per NEt LE-E4-01 The level 2 results with the flag file are expected to "The F&Q/gap was Subsequent to the LAR submittal, this F&O/gap 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 Mm Cut Upper Bound. The LERF results from using Min made regarding the current method for quantifying FTREX without using the flag file have a significant

({T]he TMI model uses Forte 3.0c LERF ... ([T]he

... 3.Oc as the number of events greater than or equal to 0.9. Using quantifier).

quantifier)." SR LE-E4 requires that LERFbe 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.Oc 3.0c at a greater is conservative and acceptable.

acceptable.

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

Response to Request for Additional Information Attachment 1 1 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 states:: "There 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. (CC-Il> requires exclusion of outliers in the definition of system/component group ." SR DA-B2 (CC-II) 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 systemssystems.. Operational characteristics (normal position and frequency of manipulation) for manual valves was not taken into account (Le., (i.e., for failure rate purposes, all manual valves were grouped together). However, the risk significance of manual valves is negligible; therefore 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 Several requirements in establishing flood initiating event frequencies are not met." 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). (CC-Il). Are any of the valves that may be assigned new STls 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, valves , if any, are assigned new STI5 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. 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.

1. B. Cowan, Exelon Generation Company, LLC, to U.S. Nuclear Regulatory Commission, "Application RegUlatory 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)," 3),

dated March 24, 2010.

Response to Request for Additional Information Information Attachment 11 LAR - TSTF-425 , Revision 33 of TSTF-425, LAR - Adoption of Page 9 of of 99 Docket No.

No. 50-289 Bamford , U.S.

2. Letter from Peter Bamford,

2. U.S. Nuclear Regulatory Commission, to Michael J. J. Pacilio, Exelon Nuclear, Three Nuclear Station - Request for Additional Information "Three Mile Island Nuclear

• Information TSTF-425, Relocation of Surveillance Regarding License Amendment Request to Adopt TSTF-425, Frequencies to aa Licensee Controlled Program (TAO Frequencies No. ME3587),

(TAC No. ME3587) ," dated July 2, 2010.

2,2010.

ATIACHMENT2 ATTACHMENT 2 License Amendment Request 1

Three Mile Island Nuclear Station, Unit 1 Docket No. 50-289 Application for Technical Specification Change Regarding Risk-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

(Cont'd)

Bases (Contd)

The 600 ppmb limit in 600 ppmb Item 4, in Item Table 4.1-3 4, Table 4.1-3 is used to is used the requirements of Section 5.4.

to meet the 5.4. Under other circumstances the the minimum acceptable boron boron concentration would have been been zero ppmb.ppmb.

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

and acquisition information .

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

Channels subject only to drift "drift" errors induced within the instrumentation itself can can tolerate longer calibrations . Process system instrumentation errors induced intervals between calibrations. induced by drift can 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 aa channel (essentially aa channel failure) will Substantial be revealed during routine will be checking and testing procedures.

checking procedures .

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

Thus, considered acceptable.

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 1).

Program.

The rotation reactor nrptection rotation schedule for the roastor channels is as follows:

protestion channel-s follows:

a) Deleted b) Semi annually with one channel being tested every 16 Semi-annually 46 days on a WIIWIUOUS continuous sequential rotation.

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

'with calculations of Reforence

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

Upon detection failure that prevents trip detection of a failure trip action action in channel, the instrumentation in a channel, instrumentation associated associated withwith the protection parameter failure will be tested in protection in the remaining channels. If actuation of a safety channel remaining channels.

occurs, assurance will be required that actuation was within occurs, limiting safety system setting.

within the limiting setting.

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

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

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

d.

d. The battery The battery willwill be subjected to be subjected to aa load load test test on en-a a refueling intePJal efue14g4r4tEwva basis in 9a&i& in accordance accordance with with the the Surveillance Surveillance Frequency Frequency Control Control Program Program..

(1)

(1) Verify battery Verify battery capacity capacity exceeds exceeds that that required required to to meet meet design design loads.

loads.

(2)

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

4.6.3 Heaters Pressurizer Heaters

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

Program:

(1) Pressurizer heater groups 8 and 9 shall be transferred transferred from the normal power bus to the emergency power bus and energized. energized. Upon 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 bypass..

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

Bases The tests specified are designed to demonstrate that one diesel generator will will provide power for operation of safeguards equipment. They also assure that the emergency generator control system and the control contro l systems for the safeguards equipment willwill function automatically in in the event of aa 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 diesel capability to carry design basis accident (LOOP/LOCA) load. The test should not exceed the diesel 2000-hr. rating of 3000 kW. kW. The automatic tripping of manually transferred loads, loads, onon an Engineered Safeguards Actuation Signal, Signal, protects the diesel generators from aa potential overload condition. The testing frequency specified is is intended to identify and permit correction of any mechanical or electrical deficiency before itit can result inin aa system failure. The fuel oil oil supply, starting circuits, circuits, and controls are continuously monitored and any faults faults are are alarmed and indicated.

indicated . An abnormal condition in in these systems would be signaled without having to place the diesel generators on on test.

Precipitous failure of the station battery is is extremely unlikely. The Ssurveillance specified spesified is is that which

'Nhish has has been been demonstrated over the years to to provide an an indication indisation of aa cell sell becoming becoming unserviceable unsePJiceable long long before before itit fa3lFrequencies fa#sFrequencies are are controlled controlled under under the Surveillance Frequency Frequency Control Control Program.

Program .

The PORV has The has aa remotely operated block block valve to to provide aa positive shutoffshutoff capability should should the relief valve become become inoperable. The The electrical power for both both the relief valve valve and and thethe block valve valve isis supplied from from anan ESF ESF power source to ensure the to the ability ability to to seal seal this this possible possible RCSRCS leakage leakage path.

path.

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

circulation.

4-47 4-47 Amendment No. 78, Amendment No. 157, 167, 175, ECR 78,157,167,175, ECR TM TM 07-00119 0700119