TMI-10-021, 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)

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)

ML100840205
Person / Time
Site:	Three Mile Island
Issue date:	03/24/2010
From:	Cowan B Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TMI-10-021
Download: ML100840205 (89)

Text

Exelon Nuclear wwwexeloncorp.com 200 Exelon Way Nuclear Kennett Square, PA 19348 10 CFR 50.90 TMI-10-021 March 24, 2010 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289

SUBJECT:

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)

In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR 50.90),

"Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) is submitting 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 specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies."

The changes are consistent with NRC-approved Industry Technical Specifications Task Force (TSTF)

Standard Technical Specifications (STS) change TSTF-425, "Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b, Revision 3, (ADAMS Accession No. ML090850642). The Federal Register notice published on July 6,2009 (74 FR 31996), announced the availability of this TS improvement. provides a description of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides documentation of Probabilistic Risk Assessment (PRA) technical adequacy. Attachment 3 provides the eXisting TMI Unit 1 TS and TS Bases pages marked up to show the proposed changes. Attachment 4 provides a TSTF-425 (NUREG-1430) versus TMI Unit 1 TS Cross-Reference. Attachment 5 provides the proposed No Significant Hazards Consideration.

There are no regulatory commitments contained in this letter.

License Amendment Request Adoption of TSTF-425, Rev. 3 Docket No. 50-289 March 24, 2010 Page 2 Exelon requests approval of the proposed license amendment by March 24, 2011, with the amendment being implemented within 120 days.

These proposed changes have been reviewed by the Plant Operations Review Committee and approved in accordance with Nuclear Safety Review Board procedures.

In accordance with 10 CFR 50.91, "Notice for Public Comment; State Consultation," a copy of this application, with attachments, is being provided to the designated State Official.

If you have any questions regarding this submittal, please contact Glenn Stewart at 610-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 24th day of March 2010.

Respectfu IIy, C;l{* P~'--- _

Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Description and Assessment

2. Documentation of PRA Technical Adequacy

3. Proposed Technical Specification and Bases Page Changes

4. TSTF-425 (NUREG-1430) vs. TMI Unit 1 Cross-Reference

5. Proposed No Significant Hazards Consideration cc: Regional Administrator, Region I, USNRC wi attachments USNRC Project Manager, TMI Unit 1 II USNRC Senior Resident Inspector, 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

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)

Description and Assessment

LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-289 Page 1 of 4 DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

The proposed amendment would modify the Three Mile Island Nuclear Station, Unit 1 (TMI Unit

1) Technical Specifications (TS) by relocating specific surveillance frequencies to a Iicensee-controlled program with the adoption of Technical Specification Task Force (TSTF)-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b" (Ref. 1). Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 6, Administrative Controls.

The changes are consistent with NRC-approved IndustryffSTF Standard Technical Specifications (STS) change TSTF-425, Revision 3, (ADAMS Accession No. ML090850642). The Federal Register notice published on July 6,2009 (74 FR 31996) (Ref. 2), announced the availability of this TS improvement.

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation Exelon Generation Company, LLC (Exelon) has reviewed the NRC staff's model safety evaluation for TSTF-425, Revision 3, dated July 6, 2009. This review included a review of the NRC staff's model safety evaluation, TSTF-425, Revision 3, and the requirements specified in NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. ML071360456) (Ref.

3). includes Exelon's documentation with regard to Probabilistic Risk Assessment (PRA) technical adequacy consistent with the requirements of Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," (ADAMS Accession No. ML070240001) (Ref. 4), Section 4.2, and describes any PRA models without NRC-endorsed standards, including documentation of the quality characteristics of those models in accordance with Regulatory Guide 1.200.

Exelon has concluded that the justifications presented in the TSTF proposal and the NRC staff's model safety evaluation prepared by the NRC staff are applicable to TMI Unit 1 and justify this amendment to incorporate the changes to the TMI Unit 1 TS.

2.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3; however, Exelon proposes variations or deviations from TSTF-425, as identified below, which includes differing Surveillance numbers.

1. Revised (clean) TS pages are not included in this amendment request given the number of TS pages affected, the straightforward nature of the proposed changes, and outstanding TMI Unit 1 amendment requests that will impact some of the same TS pages.

Providing only mark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site

LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-289 Page 2 of 4 permit," (Ref. 5) in that the mark-ups fully describe the changes desired. This is an administrative deviation from the NRC staff's model application dated July 6,2009 (74 FR 31996) with no impact on the NRC staff's model safety evaluation published in the same Federal Register Notice. As a result of this deviation, the contents and numbering of the attachments for this amendment request differ from the attachments specified in the NRC staff's model application. Also, since the Bases for the TMI Unit 1 Surveillance Requirements are intermingled throughout the TS Surveillance Requirement sections, mark-ups of both the proposed TS changes and the proposed TS Bases changes are provided together in Attachment 3.

2. Attachment 4 provides a cross-reference between the NUREG-1430 Surveillances included in TSTF-425 versus the TMI Unit 1 Surveillances proposed to be relocated as part of this amendment request. Attachment 4 includes a summary description of the referenced TSTF-425 (NUREG-1430)ffMI Unit 1 TS Surveillances, which is provided for information purposes only and is not intended to be a verbatim description of the TS Surveillances. This cross-reference highlights the following:

a. NUREG-1430 Surveillances included in TSTF-425 and corresponding TMI Unit 1 Surveillances that have differing Surveillances numbers,

b. NUREG-1430 Surveillances included in TSTF-425 that are not contained in the TMI Unit 1 TS, and

c. TMI Unit 1 plant-specific Surveillances that are not contained in NUREG-1430 and, therefore, are not included in the TSTF-425 mark-ups.

Concerning the above, TMI Unit 1 TS are custom TS for a pressurized water reactor (PWR) plant. As a result, the applicable TMI Unit 1 TS and associated Bases numbers differ from the Standard Technical Specifications (STS) presented in NUREG-1430 and TSTF-425, Revision 3. Although the majority of the TMI Unit 1 Surveillances that correspond to the NUREG-1430 Surveillances are located in TMI Unit 1 TS Section 4.0, "Surveillance Standards," there are some instances where the TMI Unit 1 Surveillances that correspond to the NUREG-1430 Surveillances are located in TMI Unit 1 TS Section 3.0, "Limiting Conditions for Operation." For example, TMI Unit 1 TS 3.5.2.7.1 (axial power imbalance) and TS 3.5.2.4.g (quadrant power tilt) correspond to NUREG-1430 Surveillance Requirements 3.2.3.1 and 3.2.4.1, respectively. These are identified in the Attachment 4 cross-reference. This is an administrative deviation from TSTF-425 with no impact on the NRC staff's model safety evaluation dated July 6, 2009 (74 FR 31996).

In addition, there are Surveillances contained in NUREG-1430 that are not contained in the TMI Unit 1 TS. Therefore, the NUREG-1430 mark-ups included in TSTF-425 for these Surveillances are not applicable to TMI Unit 1. Also, the TMI Unit 1 TS do not contain a definition for Staggered Test Basis. These are administrative deviations from TSTF-425 with no impact on the NRC staff's model safety evaluation dated July 6,2009 (74 FR 31996).

Furthermore, the TMI Unit 1 TS include plant-specific Surveillances that are not contained in NUREG-1430 and, therefore, are not included in the NUREG-1430 mark-ups provided in TSTF-425. For example, TMI Unit 1 TS 3.14.1.1 concerns periodic inspection of the dikes around TMI which does not appear in NUREG-1430. This plant-specific Surveillance and the others are specified in the Attachment 4 cross-reference. Exelon has determined that the relocation of the Frequencies for these TMI Unit 1 plant-specific

LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-289 Page 3 of 4 Surveillances is consistent with TSTF-425, Revision 3, and with the NRC staff's model safety evaluation dated July 6, 2009 (74 FR 31996), including the scope exclusions identified in Section 1.0, "Introduction," of the model safety evaluation. Changes to the Frequencies for these plant-specific Surveillances would be controlled under the Surveillance Frequency Control Program (SFCP). The SFCP provides the necessary administrative controls to require that Surveillances related to testing, calibration and inspection are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the Limiting Conditions for Operation will be met. Changes to Frequencies in the SFCP would be evaluated using the methodology and probabilistic risk guidelines contained in NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. ML071360456),

as approved by NRC letter dated September 19, 2007 (ADAMS Accession No. ML072570267). The NEI 04-10, Revision 1 methodology includes qualitative considerations, risk analyses, sensitivity studies and bounding analyses, as necessary, and recommended monitoring of the performance of systems, components, and structures (SSCs) for which Frequencies are changed to assure that reduced testing does not adversely impact the SSCs. In addition, the NEI 04-10, Revision 1 methodology satisfies the five key safety principles specified in Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated August 1998 (ADAMS Accession No. ML003740176) (Ref. 6), relative to changes in Surveillance Frequencies. Therefore, the proposed relocation of the TMI Unit 1 plant-specific Surveillance Frequencies is consistent with TSTF-425 and with the NRC staff's model safety evaluation dated July 6,2009 (74 FR 31996).

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Exelon has reviewed the proposed no significant hazards consideration (NSHC) determination published in the Federal Register dated July 6,2009 (74 FR 31996). Exelon has concluded that the proposed NSHC presented in the Federal Register notice is applicable to TMI Unit 1, and is provided as Attachment 5 to this amendment request, which satisfies the requirements of 10 CFR 50.91 (a), "Notice for public comment; State consultation" (Ref. 7).

3.2 Applicable Regulatory Reguirements A description of the proposed changes and their relationship to applicable regulatory requirements is provided in TSTF-425, Revision 3 (ADAMS Accession No. ML090850642) and the NRC staff's model safety evaluation published in the Notice of Availability dated July 6, 2009 (74 FR 31996). Exelon has concluded that the relationship of the proposed changes to the applicable regulatory requirements presented in the Federal Register notice is applicable to TMI Unit 1.

3.3 Precedence This application is being made in accordance with the TSTF-425, Revision 3 (ADAMS Accession No. ML090850642). Exelon is not proposing significant variations or deviations from the TS changes described in TSTF 425 or in the content of the NRC staff's model safety evaluation published on July 6,2009 (74 FR 31996). The NRC has previously approved amendments to

LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-289 Page 4 of 4 the TS as part of the pilot process for TSTF-425, including: Amendment Nos. 186 and 147 for Limerick Generating Station, Units 1 and 2, respectively (TAC Nos. MC3567 and MC3568) dated September 28, 2006; Amendment Nos. 200 and 201 for Diablo Canyon Power Plant, Units 1 and 2, respectively (TAC Nos. MD8911 and MD8912), dated October 30,2008; and Amendment Nos. 188 and 175 for South Texas Project, Units 1 and 2, respectively (TAC Nos. MD7058 and MD7059), dated October 31, 2008. The subject License Amendment Request proposes to relocate periodic surveillance frequencies to a licensee-controlled program and add a new program (the Surveillance Frequency Control Program) to the Administrative Controls section of TS in accordance with TSTF-425 and as discussed in the previously approved amendments.

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

4.0 ENVIRONMENTAL CONSIDERATION

Exelon has reviewed the environmental consideration included in the NRC staff's model safety evaluation published in the Federal Register on July 6,2009 (74 FR 31996). Exelon has concluded that the staff's findings presented therein are applicable to TMI Unit 1, and the determination is hereby incorporated by reference for this application.

5.0 REFERENCES

1. TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b," March 18,2009 (ADAMS Accession Number: ML090850642).

2. NRC Notice of Availability of Technical Specification Improvement to Relocate Surveillance Frequencies to Licensee Control- Risk-Informed Technical Specification Task Force (RITSTF) Initiative 5b, Technical Specification Task Force - 425, Revision 3, published on July 6,2009 (74 FR 31996).

3. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession Number:

ML071360456).

4. Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," January 2007 (ADAMS Accession Number: ML070240001).

5. 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit."

6. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications," dated August 1998 (ADAMS Accession No. ML003740176).

7. 10 CFR 50.91 (a), "Notice for public comment; State consultation."

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)

Documentation of PRA Technical Adequacy

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page i of i Documentation of PRA Technical Adequacy TABLE OF CONTENTS Section 2.1 Overview 1 2.2 Technical Adequacy of the PRA Model.. 3 2.2.1 Plant Changes Not Yet Incorporated into the PRA Model .4 2.2.2 Applicability of Peer Review Findings and Observations 4 2.2.3 Consistency With Applicable PRA Standards 5 2.2.4 Identification of Key Assumptions 5 2.3 External Events Considerations 21 2.3.1 Fire PRA 21 2.4 Summary 22 2.5 References 22

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 1 of 23 Documentation of PRA Technical Adequacy 2.1 Overview The implementation of the Surveillance Frequency Control Program (also referred to as Tech Spec Initiative 5b) at Three Mile Island (TMI) will follow the guidance provided in NEI 04-10, Revision 1 [Ref. 1] in evaluating proposed surveillance test interval (STI) changes.

The following steps of the risk-informed STI revision process are common to proposed changes to all STls within the proposed licensee-controlled program.

• Each STI revision is reviewed to determine whether there are any commitments made to the NRC that may prohibit changing the interval. If there are no related commitments, or the commitments may be changed using a commitment change process based on NRC endorsed guidance, then evaluation of the STI revision would proceed. If a commitment exists and the commitment change process does not permit the change, then the STI revision would not be implemented.

• A qualitative analysis is performed for each STI revision that involves several considerations as explained in NEI 04-10.

• Each STI revision is reviewed by an Expert Panel, referred to as the Integrated Decision-making Panel (lOP), which is normally the same panel as is used for Maintenance Rule implementation, but with the addition of specialists with experience in surveillance tests and system or component reliability. If the lOP approves the STI revision, the change is implemented and documented for future audits by the NRC. If the lOP does not approve the STI revision, the STI value is left unchanged.

• Performance monitoring is conducted as recommended by the lOP. In some cases, no additional monitoring may be necessary beyond that already conducted under the Maintenance Rule. The performance monitoring helps to confirm that no failure mechanisms related to the revised test interval become important enough to alter the information provided for the justification of the interval changes.

• The lOP is responsible for periodic review of performance monitoring results.

If it is determined that the time interval between successive performances of a surveillance test is a factor in the unsatisfactory performances of the surveillance, the lOP returns the STI back to the previously acceptable STI.

• In addition to the above steps, the PRA is used when possible to quantify the effect of a proposed individual STI revision compared to acceptance criteria in Figure 2 of NEI 04-10. Also, the cumulative impact of all risk-informed STI revisions on all PRAs (Le., internal events, external events and shutdown) is also compared to the risk acceptance criteria as delineated in NEI 04-10.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 2 of 23 For those cases where the STI cannot be modeled in the plant PRA (or where a particular PRA model does not exist for a given hazard group), a qualitative or bounding analysis is performed to provide justification for the acceptability of the proposed test interval change.

The NEI 04-10 methodology endorses the guidance provided in Regulatory Guide 1.200, Revision 1 [Ref. 2]. "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." The guidance in RG-1.200 indicates that the following steps should be fOllowed when performing PRA assessments:

1. Identify the parts of the PRA used to support the application.

- SSCs, operational characteristics affected by the application and how these are implemented in the PRA model.

- A definition of the acceptance criteria used for the application.

2. Identify the scope of risk contributors addressed by the PRA model.

- If not full scope (Le. internal and external), identify appropriate compensatory measures or provide bounding arguments to address the risk contributors not addressed by the model.

3. Summarize the risk assessment methodology used to assess the risk of the application.

- Include how the PRA model was modified to appropriately model the risk impact of the change request.

4. Demonstrate the Technical Adequacy of the PRA.

- Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does not impact the PRA results used to support the application.

- Document peer review findings and observations that are applicable to the parts of the PRA required for the application, and for those that have not yet been addressed justify why the significant contributors would not be impacted.

- Document that the parts of the PRA used in the decision are consistent with applicable standards endorsed by the Regulatory Guide (currently, RG-1.200 Revision 1 includes only internal events PRA standard).

Provide justification to show that where specific requirements in the standard are not adequately met, it will not unduly impact the results.

- Identify key assumptions and approximations relevant to the results used in the decision-making process.

Because of the broad scope of potential Initiative 5b applications and the fact that the impact of such assumptions differs from application to application, each of the issues encompassed in Items 1 through 3 will be covered with the preparation of each individual PRA assessment made

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 3 of 23 in support of the individual STI interval requests. The purpose of the remaining portion of this appendix is to address the requirements identified in item 4 above.

2.2 Technical Adequacy of the PRA Model The TM1080 version of the TMI PRA model is the most recent evaluation of the risk profile at TMI for internal event challenges, including internal flooding. The TMI PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the TMI PRA is based on the event tree / fault tree methodology, which is a well-known methodology in the industry.

Exelon Generation Company, LLC (Exelon) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating Exelon nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the TMI PRA.

PRA Maintenance and Update The Exelon risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in the Exelon Risk Management program, which consists of a governing procedure (ER-AA-600, "Risk Management") and subordinate implementation procedures. Exelon procedure ER-AA-600-1015, "FPIE PRA Model Update," delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating Exelon nuclear generation sites. The overall Exelon Risk Management program, including ER-AA-600-1015, defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operating experience), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plants, the following activities are routinely performed:

• Design changes and procedure changes are reviewed for their impact on the PRA model.

• New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

• Maintenance unavailabilities are captured, and their impact on CDF is trended.

• Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated approximately every four years.

In addition to these activities, Exelon risk management procedures provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance includes:

• Documentation of the PRA model, PRA products, and bases documents.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 4 of 23

• The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.

• Guidelines for updating the futl power, internal events PRA models for Exelon nuclear generation sites.

• Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10CFR50.65(a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximately 4-year cycle; longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant. Exelon completed the TM1080 update to the TMI PRA model in June 2009, which was the result of a regularly scheduled update to the previous PRA model.

As indicated previously, RG-1.200 also requires that additional information be prOVided as part of the LAR submittal to demonstrate the technical adequacy of the PRA model used for the risk assessment. Each of these items (plant changes not yet incorporated in to the PRA model, relevant peer review findings, consistency with applicable PRA Standards, and the identification of key assumptions) will be discussed in turn.

2.2.1 Plant Changes Not Yet Incorporated into the PRA Model A PRA updating requirements evaluation (URE) Exelon PRA model update tracking database is created for all issues that are identified that could impact the PRA model. The URE database includes the identification of those plant changes that could impact the PRA model.

As part of the PRA evaluation for each STI change request, a review of open items in the URE database for TMI will be performed and an assessment of the impact on the results of the application will be made prior to presenting the results of the risk analysis to the lOP. If a non-trivial impact is expected, then this may include the performance of additional sensitivity studies or model changes to confirm the impact on the risk analysis.

2.2.2 Applicability of Peer Review Findings and Observations Several assessments of technical capability have been made, and continue to be planned, for the TMI PRA model. These assessments are as follows and further discussed in the paragraphs below.

• An independent PRA peer review was conducted under the auspices of the B&W Owners Group in 2000, following the Industry PRA Peer Review process [Ref. 3]. This peer review included an assessment of the PRA model maintenance and update process.

• A limited scope gap assessment was performed in 2005 to support the Mitigating Systems Performance Indicator (MSPI) implementation. Additionally, the TMI Unit 1

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 5 of 23 PRA model results were evaluated in the B&W Owners Group PRA cross-comparisons study performed in support of implementation of the MSPI process.

• A RG-1.200 Peer Review was conducted in October 2008 against the ASME PRA Standard, Addenda RA-Sb-2005 and RA-Sc-2007 [Ref. 4]. The DA and IF elements (Data and Internal Flooding) were not reviewed at this time.

A summary of the disposition of the PRA Peer Review facts and observations (F&Os) for the TMI PRA model was documented as part of the statement of PRA capability for MSPI in the TMI MSPI Basis Document [Ref. 5]. As noted in that document, the one significance level A F&O and all but one significance level B F&Os from that peer review have been addressed and closed out as of the TMI 2004 Revision 1 PRA model. The remaining issue was resolved in the TMI 2004 Revision 2 PRA model.

2.2.3 Consistency with Applicable PRA Standards As indicated above, a PRA model update was completed in 2009, resulting in the TM1080 updated model. In updating the PRA, changes were made to the PRA model to address most of the identified gaps from the peer review, as well as to address other open UREs. Open findings from the peer review are summarized in Table 2-1.

The 2008 peer review did not cover the DA or IF elements. For Internal Flooding, all the B F&Os associated with the IF technical element from the 2000 peer review have been dispositioned. The Data Analysis has been upgraded since the 2000 peer review. A self-assessment against the Standard was performed for both DA and IF; the results are provided in Table 2-2.

All remaining gaps will be reviewed for consideration for the next periodic PRA model update, but are judged to have low impact on the PRA model or its ability to support a full range of PRA applications. The remaining gaps are documented in the URE database so that they can be tracked and their potential impacts accounted for in applications where appropriate.

Each item will be reviewed as part of each STI change assessment that is performed and an assessment of the impact on the results of the application will be made prior to presenting the results of the risk analysis to the lOP. If a non-trivial impact is expected, then this may include the performance of additional sensitivity studies or model changes to confirm the impact on the risk analysis.

2.2.4 Identification of Key Assumptions The overall Initiative 5b process is a risk-informed process with the PRA model results providing one of the inputs to the lOP to determine if an STI change is warranted. The methodology recognizes that a key area of uncertainty for this application is the standby failure rate utilized in the determination of the STI extension impact. Therefore, the methodology requires the performance of selected sensitivity studies on the standby failure rate of the component(s) of interest for the STI assessment.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 6 of 23 The results of the standby failure rate sensitivity study plus the results of any additional sensitivity studies identified during the performance of the reviews as outlined in 2.2.1 and 2.2.3 above for each STI change assessment will be documented and included in the results of the risk analysis that goes to the lOP.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 7 of 23 Table 2-1 Open Peer Review Findings Title Description of Gap Applicable Current Comment SRs [Ref. Status 61 IE-M-01 The list of systems examined seems to be generated IE-A5 Open Table 5 in the Initiating Event Notebook (TMI-PRA-002, from a high level PRA system significance standpoint Rev.1) shows the results of a systematic review of all and does not seem to provide a complete list of all the systems in the PRA. The impact of failure of plant systems. systems not modeled in the PRA that cause an IE are subsumed in other events (e.g., reactor trip, Loss of offsite power, LOCA, etc.). However, this is not explicitly documented in the IE Notebook.

Therefore, this is considered a documentation issue not affectinQ the technical adequacy of the PRA model.

IE-Ma-01 For the systematic evaluation required in IE-A4, the IE-A6 Open The potential for common cause failures was included in examination of potential initiating events resulting from the systematic evaluation for potential initiating events.

common cause failures is not documented. This is a documentation issue not affecting the technical adequacy of the PRA model.

IE-A5-01 No documentation was found of incorporating: (a) IE-A7 Open This is a documentation issue not affecting the technical events that have occurred at conditions other than at- adequacy of the PRA model.

power operation (Le., during low-power or shutdown conditions), and for which it is determined that the event could also occur during at-power operation; (b) events resulting in a controlled shutdown that includes a scram prior to reaching low-power conditions, unless it is determined that an event is not applicable to at-power operation.

IE-A6-01 No documentation was found of interviews with plant IE-A8 Open This is a documentation issue not affecting the technical personnel (e.g., operations, maintenance, adequacy of the PRA model.

engineering, safety analysis) to determine if potential initiatinq events have been overlooked.

IE-A7-01 No documentation of the review of plant-specific IE-A9 Open This is a documentation issue not affecting the technical operating experience for initiating event precursors adequacy of the PRA model.

was found in the PRA notebooks. Review plant-specific operating experience for initiating event precursors for potential initiators.

IE-ClO-01 No comparison of the initiating event fault tree results IE-C12 Open The initiating event fault tree results were compared to with generic data has been identified. generic industry frequencies and with the PWROG Compare plant initiating event fault tree results to database. However, the results of the review are not generic frequency sources (Le., NUREGlCR-5750, documented; NUREG/CR-6928, WaG PSA Database, etc.) and Therefore, this is a documentation issue not affecting explain differences. the technical adequacy of the PRA model.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 8 of 23 Table 2-1 Open Peer Review Findings Title Description of Gap Applicable Current Comment SRs [Ref. Status 6]

IE-D1-01 The initiating event analysis has not been IE-D1 Open The IE notebook was updated, although several documented in a manner that facilitates PRA documentation-related IE F&Os remain open.

applications, upgrades, and peer review. Therefore, this F&O remains open, but it is a The IE analysis is very difficult to trace and relies documentation issue not affecting the technical heavily on the ABS 2003 documentation, without adequacy of the PRA model.

proper reference in the IE notebook.

AS-C1-01 Much of the AS-related documentation is located in AS-C1 Open The Event Tree Notebook was updated, although the ABS 2003 report, with updates identified in the several documentation-related AS F&Os remain open.

Event Tree notebook. In many cases, bases could Therefore, this F&O remains open, but it is a not be verified without support of the TMI PRA documentation issue not affecting the technical personnel to aid in tracking down the documentation. adequacy of the PRA model.

To facilitate reviews, upgrades, etc., it is necessary to either include all the documentation in the event tree notebook or to reference the material in other documents..

AS-C2-01 The process used to develop the accident sequences AS-C2, Open The process for developing accident sequences used is not provided. Incorporation of plant specific AS-A4, plant-specific information such as procedures.

information is therefore not demonstrated. AS-A5 This is a documentation issue not affecting the technical Provide the process description and include the adequacy of the PRA model.

discussion of use of orocedures, etc.

SC-B2-01 Tables 3-1 through 3-8 of TMI PRA-003 include SC-B2 Open Expert judgment was NOT used in determining the several instances of use of "Judgment" as the basis success criteria.

for success criteria. These applications of judgment Therefore, this is a documentation issue not affecting do not use section 4.3 of the ASME std. to attain CCII the technical adequacy of the PRA model.

and are not discussed in the report as required by SC-C2 to attain CCI.

Do not use judgment as basis for success criteria or apply para. 4.3 of the ASME standards when implementinq expert iudqment.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 9 of 23 Table 2-1 Open Peer Review Findings Title Description of Gap Applicable Current Comment SRs [Ref. Status 6]

SC-B4-01 Many of the T/H success criteria were developed SC-B4 Open The documentation is misleading. MAAP was not used using the MAAP computer code, including large break to develop the success criteria for Large LOCAs.

LOCAs greater than 10* diameter (see Table 3-3 of Therefore, this is a documentation issue not affecting Success Criteria notebook). However, FAI has the technical adequacy of the PRA model.

identified a limitation/precaution using MAAP for the large break LOCA analyses. "...the results of the code should not be used for a definitive determination of the primary system pressure response, mass and energy releases, and peak cladding temperatures during this time frame."

Do not use MAAP to develop large LOCA success criteria due to limitations associated with the code.

SC-B5-01 No documentation of a check for the reasonableness SC-B5 Open Reasonableness and acceptability of the results were and acceptability of the results (Le., comparison with checked; this is a documentation issue not affecting the results of the same analyses performed for similar technical adequacy of the PRA model.

plants, accounting for differences in unique plant features). Compare TMI results with results of the same analyses performed for similar plants, accounting for differences in unique plant features SC-C1-01 Documentation does not facilitate PRA application, SC-C1 Open This is a documentation issue not affecting the technical upgrades, or peer review. Though it appears the adequacy of the PRA model.

information exists, one must piece together information in multiple notebooks and calculations with no correlation reference to understand how success criteria were evaluated or developed in the model. In application and model upgrade one could easily make an error due to the disconnected nature of the documentation.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 10 of 23 Table 2-1 Open Peer Review Findings Title Description of Gap Applicable Current Comment SRs[Ref. Status 6]

SC-C2-01 There is an implied process in the latest Success SC-C2 Open The HRA and its notebook were revised to include Criteria Notebook, but not a clear process for documented bases for times available to perform evaluating and documenting success criteria, this can operator actions and times needed to perform the be easily related to the other SC-C criteria not being actions.

met. Documentation of core damage could be Expert judgment was NOT used in the development of clarified. Though calculations and other references success criteria.

are used to develop success criteria they are not The Success Criteria NB still requires updating, so this easily found in the documentation. Computer codes F&O remains open.

are identified in some cases, however there is no This is a documentation issue not affecting the technical description of limitations or potential conservatisms. adequacy of the PRA model.

The use of expert judgment is used without rational or basis. There is in many cases no basis for the time given for human actions such as operator interviews or simulator runs or MAAP analysis. There is no summery of success criteria for mitigating systems and HEP's used, SC-C3-01 Documentation of sources of uncertainty has not been SC-C3 Open Sources of uncertainty and their impact on this accomplished. This is a recognized/acknowledged application will be addressed by sensitivities per NEI 04-gap for the TMI PRA. 10, if applicable to the specific STI evaluation.

SY-A20-01 In general, the system notebooks do not discuss room SY-A22, Open This is a documentation issue not affecting the technical cooling. The EFW system considered the impact of a SY-B6, adequacy of the PRA model.

steam line break and the diesel generators are SY-B?,

assumed to require the room fan for success. AS-B?

However, other notebooks (e.g. HPI, DHRW/CCW, LPI/DHR) do not mention room cooling. HVAC systems are discussed in Appendix D to the 2003 TMI update, which presents the TMI responses to the 2000 peer review. In that document, the response to F&O DE-2 presents a review of various HVAC systems. Some PRA component areas are excluded with a good basis (e.g. NSCCW pump areas reference results from loss of ventilation tests).

However, it is difficult to evaluate each area's HVAC requirements by reading the responses to the F&Os.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 11 of 23 Table 2-1 Open Peer Review Findings Title Description of Gap Applicable Current Comment SRs [Ref. Status 6]

SY-Cl-0l This SR is not met due to SY-C2 and SY-C3 not being SY-Cl, Open This is a documentation issue not affecting the technical met. In general the system notebooks did not supply SY-C2, adequacy of the PRA model.

sufficient information to evaluate SY effectively. SY-C3, Continue developing system documentation as a SY-A2 stand-alone document representing the current model.

It is recommended that system notebooks like the Electrical Systems be broken up to discuss specific systems in more detail, for example the Diesel Generators, 4.160Kv, 480VAC, DC etc.

SY-C2-01 Documentation of the systems analysis was not SY-C2 Open This is a documentation issue not affecting the technical sufficient reasonably assess the associated adequacy of the PRA model.

supporting requirements.

QU-BS-Ol Logic loops have been broken, as none appear in the QU-B5 Open This is a documentation issue not affecting the technical TMll042 model. However, no record can be found of adequacy of the PRA model.

how the logic loops were broken. Document how logic loops were identified and broken.

QU-Cl-0l Multiple HFE identification only considers HFE in the QU-Cl, Open Only three recoveries are used in the TMI PRA.

quantified model fault tree. Recovery event applied HR-H3 Dependency with other HFEs is considered for two of post quantification by the recovery tree were not them, but not the third. Determining the dependency addressed. Include recovery tree event for and applying it to the recovery will be addressed by dependency identification. sensitivities per NEI 04-10, if applicable to the specific STI evaluation.

QU-D3-01 Comparison of the results of the model with other QU-D4 Open A comparison of the model results with other plants was similar plants was not documented. Perform the performed at a high level for other B&W plants and at a comparison. more detailed level for ANO-2. This review is not documented.

Therefore, this is a documentation issue not affecting the technical adequacy of the PRA model nor this application.

QU-D5-01 Contribution to CDF of SSCsioperator actions are not QU-D6 Open Some SSCs that are significant contributors to initiating provided in a manner to distinguish between initiating events, but not to mitigation, are not explicitly identified events vs. event mitigation. Expand the results in the documentation of significant contributors.

discussion to include additional discussion of However, this is a documentation issue not affecting the contributors at lower level of resolution and provide technical adequacy of the PRA model nor this the contributions for IEs and for mitigation. application.

QU-E4-01 There is no evidence that an evaluation was QU-E4 Open This will be addressed by sensitivities per NEI 04-10, if performed of the sensitivity of the results to key model applicable to the specific STI evaluation.

uncertainties and key assumptions.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 12 of 23 Table 2-1 Open Peer Review FindinQs Title Description of Gap Applicable Current Comment SRs [Ref. Status 6]

QU-Fl-0l The documentation of the quantification included only QU-Fl Open The Quantification Notebook has been updated and minimal information. Many of the requirements of the significantly improved. However, several stated in the SRs are not included. (Examples: documentation-related QU F&Os remain open.

Reviews are not documented. Contributors are very Therefore, this F&O remains open, but it is a minimally documented.) The documentation does not documentation issue not affecting the technical meet the minimum requirements of the ASME adequacy of the PRA model.

standard and thus does not facilitate applications/upgrades/reviews. The documentation does not describe the approach for identification and breaking of logic loops in the model. Revise the quantification documentation to include the requirements of the SRs.

QU-F2-01 The documentation of the quantification is missing QU-F2 Open Many of the sections listed in the F&O have been added significant sections such as reviews, sequence to the Quantification Notebook, but not all items in QU-discussions, lower level results, uncertainty analyses. F2 have been documented and this F&O remains open.

Update the results documentation to include all However, it is a documentation issue not affecting the needed information. Use the SR to provide guidance technical adequacy of the PRA model.

regarding needed and suggested content.

QU-F5-01 In the quantification notebook, other than the LERF QU-F5, Open This is a documentation issue not affecting the technical truncation limitation, no evaluations of limitations were LE-G5 adequacy of the PRA model.

presented. Explicitly consider limitations of the model as they may apply to applications.

LE-Bl-0l The LERF contributors from Table 4.5.9-3 of the LE-Bl Open LE-Bl does not meet Capability Category II, but is ASME Standard are considered in the TMI considered generally adequate for this application. This Containment Event Tree. Of the items applicable for will be addressed by sensitivities per NEI 04-10, if Large, Dry Containments such as TMI, containment applicable to the specific STI evaluation.

isolation is addressed in CET heading B, ISLOCA, SGTR, and induced SGTR in heading A, and HPMEIcore debris impingement in heading E. The item "In-vessel recovery" is considered in preventing late containment failures (per TMI-PRA-015.2, page 5-109), but no credit is given (failure event set to 1.0). It is conservative to take no credit for in-vessel recovery; the conservative modeling in the late analysis does not impact LERF, but failure to consider in the early analysis could potentially overstate impact of early containment failure after vessel breach.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 13 of 23 Table 2-1 Open Peer Review Findings Title Description of Gap Applicable Current Comment SRs [Ref. Status 61 LE-C2-01 This F&O applies to several LE SRs that involve LE-C3, Open LE-C3, LE-C10 and LE-C11 do not meet Capability reviewing significant LERF sequences for potential LE-C10, Category II, but they are treated conservatively. This is credit for equipment repair, additional recovery LE-C11 considered to be conservative relative to this actions, engineering evaluations, etc. There is no application. However, it will be addressed by formal record of a review of the LERF results for such sensitivities per NEI 04-10, if applicable to the specific items. Document reviews of the significant accident STI evaluation.

progression sequences that result in a large, early release to determine if repair, additional recoveries, additional engineering evaluations, etc. can be credited. If any credit is given, provide justification for the credit.

LE-C7-01 System level operator actions are described in the LE-C7 Open Conservative screening values are used for the CET Levell System Analysis notebooks. Offsite power HEPs. Therefore, the impact of these HEPs is recovery data is consistent with the Levell analysis. considered to be conservative relative to this Other human actions in the TMI CET were estimated application. This will be addressed by sensitivities per using qualitative judgment in Table 5-1 of the CET NEI 04-10, if applicable to the specific STI evaluation.

notebook. This qualitative evaluation is acceptable for some uncertain phenomenological issues, but more detailed HRA analyses are needed for actions that can be quantified, as per the requirements of the ASME Standard paragraph 4.5.5. Identify operator actions in the Level 2 for which a more detailed HRA is possible. One example would be comparing the time at which PORVs can be opened to reduce RCS pressure to the time at which an induced SGTR might occur. Consider sensitivity analyses on uncertain parameters.

LE-C8a-01 Equipment survivability is considered for the LE-C9 Open The Reactor Building fan coolers are undersized at TMI containment fans in Section 5 of the CET notebook. and have a little to no impact on containment pressure For before, soon after, and long after vessel failure and temperature with respect to early containment containment conditions, the fans are assumed to have failure.

a 0% chance of failure due to the accident However, this will be addressed by sensitivities per NEI environment. As a basis, the analysis states that the 04-10, if applicable to the specific STI evaluation.

Oconee fans are expected to remain functional throughout an accident. The Oconee reference is from 1990, and may have been updated since that time. As the fans are important in controlling containment pressure and temperature, which impacts the EARLY evaluation, more detailed justification should be examined to credit their survivabilitv.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 14 of 23 Table 2-1 Open Peer Review Findings Title Description of Gap Applicable Current Comment SRs [Ref. Status 6]

LE-Dlb-Ol The TMI containment comparison to the Oconee LE-D2 Open This will be addressed by sensitivities per NEI 04-10, if containment evaluation (Appendix B of TMI-PRA- applicable to the specific STI evaluation.

015.2, Rev. 0) provided a good basis for utilizing the Oconee analyses of the personnel airlocks and purge penetrations. However, the report identified that additional analyses were necessary for evaluation of the equipment hatch, mechanical penetrations and electrical penetrations. A discussion of a qualitative evaluation of these is provided in the latter portion of Appendix C of TMI-PRA-015.2, Rev. O. A detailed evaluation of the TMI equipment hatch, personnel airlock and containment purge valves are performed, providing good plant-specific basis for their evaluation.

However, the evaluation of the electrical and mechanical penetrations is very subjective, stating simply that it is assumed that their failure pressures will be higher than the containment structure. While these assumptions are likely true, some additional basis should be provided.

LE-D4-01 The secondary side isolation is evaluated in the Level LE-D5 Open All accident progression sequences involving SGTR 1 analysis. However, the SG relief valve was (either as an initiator or induced following core damage) evaluated only for the pre-eore damage failures to are assumed to be LERF. No credit is taken for SG isolate. Should core damage occur, the relief valve isolation for any SGTR accident progression sequence.

would experience many additional challenges (either Therefore, SGTR is treated conseNatively. This is also passing steam or water depending on whether or not considered to be conseNative relative to this there is FW flow to the SG). The Level 2 analysis application. However, it will be addressed by does not account for this elevated potential for a stuck sensitivities per NEI 04-10, if applicable to the specific open relief valve. STI evaluation.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 15 of 23 Table 2-1 Open Peer Review Findings Title Description of Gap Applicable Current Comment SRs [Ref. Status 61 LE-D5-01 Induced SGTR is considered in CET top event node LE-D6 Open The operator action to clear seals was determined to be Bypass, but it does not appear that a specific ISGTR considerably less likely than previously assumed, based methodology was utilized. on a review of the latest TMI SAMG guidance. This The most significant issue with the ISGTR model is reduced ISGTR contribution.

the assumption that operators would start the RCPs This F&O is still open, although the excessive with dry SGs. The CET notebook states that conservatism relating to ISGTR has been removed.

operators are directed to do so with no caution about Still, the representation of ISGTR is considered to be SG status. Clearing the loop seal results in significant conservative relative to this application. This will be convective heat transfer to the SG tubes, yielding the addressed by sensitivities per NEI 04-10, if applicable to assumed 0.9 conditional probability of ISGTR. the specific STI evaluation.

However, the current TMI SAMG guidance (ER-TM-TSC-0010, Rev. 1) directs operators to turn on the RCPs as a SAMG action but has a caution on the step 3.3 that turning on the RCPs when the SGs are dry can result in an induced SGTR. The caution states that if the SG cannot be adequately protected, then don't turn on the RCPs.

LE-E2-01 The TMI CET parameter estimates are conservative in LE-E2 Open The TMI CET parameter estimates are considered to be general. The probabilities of early containment failure conservative relative to this application. This will be from DCH, rapid steam generation, and combustible addressed by sensitivities per NEI 04-10, if applicable to gas burns are conservative and are all based on the specific STI evaluation.

references from 1992 and earlier. Studies since that time (e.g. NUREG/CR-6075, NUREG/CR-6109 and NUREG/CR-6338) have recommended greatly reduced probabilities or even eliminated early containment HPME failures from large, dry containments.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 16 of 23 Table 2-1 Open Peer Review Findings Title Description of Gap Applicable Current Comment SRs [Ref. Status 6]

LE-E4-01 The TMI documentation states the level 2 model was LE-E4 Open Some sensitivities have been periormed, although a not quantifiable in a reasonable time frame without conclusive determination has not been made regarding use of the level 2 flag file. The peer review team had the current method for quantifying LERF. The access to FTREX and was able to quantify the level 2 reviewer's use of FTREX is not necessarily applicable LERF file at a truncation of 1E-9 without the flag file. since this model has never been benchmarked against The result without the flag file was 3.126E-6 and with that quantifier (the TMI model uses Forte 3.OC as the the flag file was 1.966E-6. Based on TMI quantifier).

documentation, this was not expected; see section 5.7 Therefore, this F&O remains open and will be of the quantification notebook. The level 2 results with addressed by sensitivities per NEI 04-10, if applicable to the flag file are expected to be conservative. When the specific STI evaluation.

the cutsets were reviewed, it was determined that there appears to be non-minimal cutsets in the level 2 model as quantified without the flag file. Based on these results, the peer review team was unable to determine the validity of the flag file approach of setting level 2 split fractions with a probability of .9 and greater to true. When applying a simplified quantification approach such as the one used in TMI for level 2, an assessment should be periormed to show validity of the results to confirm that no valid cutsets were inadvertently omitted or minimalized into another cutset.

LE-F2-01 The CET document notes some MAAP sensitivity LE-F3 Open This will be addressed by sensitivities per NEI 04-10, if analyses that were periormed to aid in determining applicable to the specific STI evaluation.

the split fractions in the CET. The sensitivity analyses were not specifically referenced, but were periormed to address some of the MAAP uncertainties. No sensitivities on the other phenomenological Level 2 uncertainties (e.g. induced SGTR assumptions and probabilities) have been periormed. No uncertainty calculation was documented in the Level 2 notebooks.

Periorm LERF uncertainty and sensitivity calculations.

Characterize LERF uncertainties consistent with the applicable requirements of ASME Standard tables 4.5.8-2(d) and 4.5.8-2(e).

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 17 of 23 Table 2-2 Identified Gaps to Capability Category II for the DA and IF Technical Elements Title Description of Gap Applicable Current Comment SRs [Ref. Status 61 DA-B2-01 There is no evidence that the intent of this SR was DA-B2 Open This will be addressed by sensitivities per NEI 04-10, if met. Although the component failure rates are applicable to the specific STI evaluation.

grouped by system and component type, that does not guarantee that outliers are not included in a group.

DA-C2-01 Plant specific data is collected for component failures DA-C2 Open This will be addressed by sensitivities per NEI 04-10, if and success for all components in the scope of applicable to the specific STI evaluation.

MSPI. Although this is a smaller set of component types and failure modes than all the significant basic events (e.g., F-V >.005 or RAW >2), it is considered an acceptable scope of data for a model update.

Unavailability data is collected for all MR equipment for which unavailability data is maintained.

DA-C4-01 The MSPI rules are used for data collection of DA-C4 Open This is considered a documentation issue not affecting failures, as described in the Data Notebook. The the technical adequacy of the PRA model.

failure definitions are generally consistent with the PRA failure definitions. However, that is currently an assumption, since there is no documented basis.

DA-C7-01 The number of Surveillance Tests are estimated DA-C7 Open Although the number of Surveillance Tests are based on plant requirements. Data is obtained from estimated, the estimation is expected to be very close to the MSPI Derivation Reports as described in Section the actual value. If actual numbers of tests were used, 2.4 of the Data Notebook. the final failure rates should not be significantly different from the failure rates calculated with estimated demands.

This will be addressed by sensitivities per NEI 04-10, if applicable to the specific STI evaluation.

DA-C8-01 Time that the component is in standby is estimated DA-C8 Open Although the standby time of components is estimated, based on plant requirements (e.g., nominal time the estimation is expected to be very close to the actual between surveillance tests). Documentation of value. If actual standby times were used, the final standby mission times is lacking in the current data failure probabilities should not be significantly different notebook. from the failure probabilities calculated with estimated times.

This will be addressed by sensitivities per NEI 04-10, if applicable to the specific STI evaluation.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 18 of 23 Table 2-2 Identified Gaps to Capability Category II for the DA and IF Technical Elements TItle Description of Gap Applicable Current Comment SRs [Ref. Status 61 DA-C9-01 Operational times are estimated based on DA-C9 Open Operational times for standby components were surveillance tests and operational practices. Actual estimated using Surveillance Test practices. However, operational data is not used; it is estimated based on operational times for normally operating components operations and system engineer input. are estimated; the estimation is conservative and expected to be reasonably close to the actual value. If actual operation times were used, the final failure rates should be lower but not significantly different from the failure rates calculated with estimated times.

This will be addressed by sensitivities per NEI 04-10, if applicable to the specific STI evaluation.

DA-Cl0-0l Surveillance tests were reviewed to determine DA-Cl0 Open Although the demands and operational time for demands and operational time. Successes were components is estimated, the estimation is conservative estimated based on surveillance schedules. and expected to be reasonably close to the actual value. If actual operation times and successes were used, the final failure rates should be lower but not significantly different from the failure rates calculated with estimated times.

This will be addressed by sensitivities per NEI 04-10, if applicable to the specific STI evaluation.

DA-C12-01 A review of support system unavailability to ensure DA-C12 Open This is a documentation issue not affecting the technical that double counting did not occur was performed. adequacy of the PRA model.

However, the documentation is lacking in the Data Notebook.

DA-C14-01 A review of coincident unavailability was performed. DA-C14 Open This is a documentation issue not affecting the technical However, the documentation is lacking in the adequacy of the PRA model.

notebook.

DA-Dl-0l The decision was made to only update MSPI DA-Dl Open This will be addressed by sensitivities per NEI 04-10, if components with plant-specific data. However, this applicable to the specific STI evaluation.

does not meet the requirements to update all Significant BEs, since there are significant BEs that are not within the scope of MSPI. For the BEs within the scope of MSPI, CC II is Met. However, the full scope of significant BEs does not use both generic and plant-specific data in a Bayes process.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 19 of 23 Table 2-2 Identified Gaps to Capability Category II for the DA and IF Technical Elements Title Description of Gap Applicable Current Comment SRs [Ref. Status 6]

DA-D4-01 Although a Bayesian Approach is used, there is no DA-D4 Open This is a documentation issue not affecting the technical evidence that a check of the posterior distribution adequacy of the PRA model.

was made as required by this SR. On review, it can be seen that the type codes which were updated with plant specific information have reasonable values, but there is no documentation of the check.

DA-E2-01 There are several holes in the Data Notebook which DA-E2 Open This is a documentation issue not affecting the technical make this SR Not Met: adequacy of the PRA model.

(1) Data sources for some type codes <<10%) are from previous model documentation and should be explicitly documented in the current Data Notebook.

(2) The use of MSPI for failure and success data collection needs to be better documented.

Specifically, justification that MSPI data collection "rules' are applicable for the PRA (SRs DA-C4 and DA-C5).

(3) Documentation of standby mission times (SR DA-C8) needs to be updated in the Data Notebook.

(4) Documentation is needed to describe how unavailability data was analyzed to prevent double counting (from support systems) and to account for coincident maintenance.

(5) Documentation needed to reconcile independent vs. CCF generic data component boundaries.

(6) Documentation is needed to describe how beta and gamma distributions are combined.

DA-E3-01 This SR is not met. Parametric uncertainty values DA-E3 Open This will be addressed by sensitivities per NEI 04-10, if are provided, but sources of model uncertainty and applicable to the specific STI evaluation.

related assumptions are not.

IFPP-A2-01 Documentation is lacking in details for several parts IFPP-A2, Open This is a documentation issue not affecting the technical of the flood analysis, such as flood area IFSO-B2, adequacy of the PRA model.

determination and screening criteria and results. IFSN-B2 IFQU-B2 IFPP-B3-01 Documentation and evaluation of sources of IFPP-B3, Open This will be addressed by sensitivities per NEI 04-10, if uncertainty has not been accomplished. This is a IFSO-B3 applicable to the specific STI evaluation.

recognized/acknowledged gap for the TMI PRA. IFSN-B3 IFEV-B3

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 20 of 23 Table 2*2 Identified Gaps to Capability Category II for the DA and IF Technical Elements Title Description of Gap Applicable Current Comment SRs [Ref. Status 6]

IFEV-A5-01 Several requirements in establishing flood initiating IFEV-A5, Open This will be addressed by sensitivities per NEI 04-10, if event frequencies are not met. IFEV-A6, applicable to the specific STI evaluation.

1) Recent pipe data is not used IFEV-A?

2) Effect of plant specific features and experience are not factored into the initiating event frequencies

3) Human-induced flooding does not appear to be evaluated.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 21 of 23 2.3 External Events Considerations External hazards were evaluated in the TMI Individual Plant Examination for External Events (IPEEE) submittal in response to the NRC IPEEE Program (Generic Letter 88-20 Supplement 4). The IPEEE Program was a one-time review of external hazard risk and was limited in its purpose to the identification of potential plant vulnerabilities and the understanding of associated severe accident risks.

The results of the TMI IPEEE study are documented in the TMI IPEEE Main Report [Ref. 7].

Each of the TMI external event evaluations were reviewed as part of the Submittal by the NRC and compared to the requirements of NUREG-1407.

Consistent with Generic Letter 88-20, the TMI IPEEE submittal does not screen out seismic or fire hazards, but provides quantitative analyses. The seismic risk analysis provided in the TMI Individual Plant Examination for External Events is based on a detailed Seismic Probabilistic Risk Assessment, or Seismic PRA. The internal fire events were addressed by using a combination of the Fire Induced Vulnerability Evaluation (FIVE) methodology [Ref. 8] and Fire PRA.

The TMI Seismic PRA study is a detailed analysis that, like the internal event analysis, uses quantification and model elements (e.g., system fault trees, event tree structures, random failure rates, common cause failures, etc.) consistent with those employed in the internal events portion of the TMI IPE study. TMI currently does not maintain a Seismic PRA.

The Fire IPEEE analysis used the FIVE methodology to screen fire areas and Fire PRA to evaluate unscreened areas.

As such, there are no comprehensive CDF and LERF values available from the IPEEE to support the STI risk assessment.

Other External Hazards The other external hazards are assessed to be non-significant contributors to plant risk:

• Extreme Winds / Tornadoes: The TMI IPEEE study concluded that neither tornado wind loads nor a tornado generated missile would exceed the NUREG 1407 screening criteria.

• Offsite / Transportation Hazards: The IPEEE identifies that the frequency of aircraft impact, transportation and nearby facility accidents is concluded to be acceptable low. Transportation and nearby hazards were screened from further consideration in the IPEEE.

• External Floods: The TMI site is situated in the Susquehanna River.

Several external flood heights were evaluated and a CDF estimated in the IPEEE. This evaluation has not been maintained and would be used for qualitative insights only.

2.3.1 Fire PRA Since the performance of the IPEEE, an updated Fire PRA model was developed in 2005 and updated in 2007; it is currently based on the TM1042 Full Power Internal events PRA model.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 22 of 23 The TMI Fire PRA was developed and has been updated using a graded approach. The 2007 TMI Fire PRA is an interim implementation of NUREG/CR-6850 [Ref. 9]; that is, not all tasks identified in NUREG/CR-6850 are yet completely addressed or implemented due to the changing state-of-the-art of industry at the time of the TMI Fire PRA development. In addition, the TMI Fire PRA has not undergone a PRA Peer Review; therefore, the TMI Fire PRA model is used in a limited manner to obtain additional insights for risk applications and provide qualitative and bounding quantitative assessments.

The NEI 04-10 methodology allows for STI change evaluations to be performed in the absence of quantifiable PRA models for all external hazards. For those cases where the STI cannot be modeled in the plant PRA (or where a particular PRA model does not exist for a given hazard group), a qualitative or bounding analysis is performed to provide justification for the acceptability of the proposed test interval change.

Therefore, in performing the assessments for the other hazard groups, the qualitative or bounding approach will be utilized in most cases. The fire PRA model will be exercised to obtain quantitative fire risk insights when appropriate but refinements may need to be made on a case-by-case basis. This approach is consistent with the accepted NEI 04-10 methodology (refer to Figure 2 of NEI 04-10).

2.4 Summary The TMI PRA maintenance and update processes and technical capability evaluations described above proVide a robust basis for concluding that the PRA is suitable for use in risk-informed processes such as that proposed for the implementation of a Surveillance Frequency Control Program. As indicated above, in addition to the standard set of sensitivity studies required per the NEI 04-10 methodology, open items for changes at the site and remaining gaps to specific requirements in the PRA standard will be reviewed to determine which, if any, would merit application-specific sensitivity studies in the presentation of the application results.

2.5 References

[1] Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, NEI 04-10, Revision 1, April 2007.

[2] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, Revision 1, January 2007.

[3] Framatome Technologies, Inc., PSA Peer Review Certification Process: PSA Self-Assessment Process, 47-5005658-00, September 1999.

[4] American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, (ASME RA-S-2002), Addenda RA-Sb-2005, and Addenda RA-Sc-2007, August 2007.

[5] TMI MSPI Basis Document, TMI-2006-004 Rev. 2, September 2009.

[6] ASME Committee on Nuclear Risk Management in collaboration with ANS Risk Informed Standards Committee, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASMEIANS RA-Sa-2009, March 2009.

[7] GPU Nuclear Corporation, Three Mile Island Individual Plant Examination for External Events, Main Report, December 1994.

LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-289 Page 23 of 23

[8] Professional Loss Control, Inc., Fire-Induced Vulnerability Evaluation (FIVE)

Methodology Plant Screening Guide, EPRI TR-100370, Electric Power Research Institute, April 1992.

[9] EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, EPRI 1011989, NUREG/CR-6850, Final Report, September 2005.

ATTACHMENT 3 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)

Proposed Technical Specification and Bases Page Changes v 4-5a 4-30 4-52a 3-34a 4-6 4-39 4-54 3-35a 4-7 4-41 4-55 3-59 4-7a 4-42 4-55a 4-2 4-8 4-43 4-55f 4-2a 4-9 4-44 4-55g 4-2b 4-10 4-45 4-59 4-2d 4-10a 4-46 4-86 4-3 4-10b 4-47 6-30 4-4 4-10c 4-48 4-5 4-29 4-52

TABLE OF CONTENTS Section Page 5 DESIGN FEATURES 5-1 5.1 SITE 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 5.3 REACTOR 5-4 5.3.1 REACTOR CORE 5-4 5.3.2 REACTOR COOLANT SYSTEM 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 NEW FUEL STORAGE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.2.1 CORPORATE 6-1 6.2.2 UNIT STAFF 6-1 6.3 UNIT STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 DELETED 6-3 6.5.1 DELETED 6-4 6.5.2 DELETED 6-5 6.5.3 DELETED 6-7 6.5.4 DELETED 6-8 6.6 REPORTABLE EVENT ACTION 6-10 6.7 SAFETY LIMIT VIOLATION 6-10 6.8 PROCEDURES AND PROGRAMS 6-11 6.9 REPORTING REQUIREMENTS 6-12 6.9.1 ROUTINE REPORTS 6-12 6.9.2 DELETED 6-14 6.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6-17 6.9.4 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6-18 6.9.5 CORE OPERATING LIMITS REPORT 6-19 6.9.6 STEAM GENERATOR TUBE INSPECTION REPORT 6-19 6.10 RECORD RETENTION 6-20 6.11 RADIATION PROTECTION PROGRAM 6-22 6.12 HIGH RADIATION AREA 6-22 6.13 PROCESS CONTROL PROGRAM 6-23 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6-24 6.15 DELETED 6-24 6.16 DELETED 6-24 6.17 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-25 6.18 TECHNICAL SPECIFICATION (TS) BASES CONTROL PROGRAM 6-25 6.19 STEAM GENERATOR (SG) PROGRAM 6-26 6-20 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM 6-29 6.21 SURVEILLANCE FREQUENCY CONTROL PROGRAM 6-30

-v-Amendment No. 11, 47, 72, 77, 129, 150, 173, 212, 252, 253, 256, 261, 264,269

2. The protection system reactor power/imbalance envelope trip setpoints shall be reduced 2 percent in power for each 1 percent tilt, in excess of the tilt limit, or when thermal power is equal to or less than 50% full power with four reactor coolant pumps running, set the nuclear overpower trip setpoint equal to or less than 60% full power.

3. The control rod group withdrawal limits in the CORE OPERATING LIMITS REPORT shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.

4. The operational imbalance limits in the CORE OPERATING LIMITS REPORT shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.

f. Except for physics or diagnostic testing, if quadrant tilt is in excess of the maximum tilt limit defined in the CORE OPERATING LIMITS REPORT and using the applicable detector system defined in 3.5.2.4.a, b, and c above, reduce thermal power to 15% FP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Diagnostic testing during power operation with a quadrant tilt is permitted provided that the thermal power allowable is restricted as stated in 3.5.2.4.d above.

g. Quadrant tilt shall be monitored on a minimum frequency of once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the QPT alarm is inoperable and every 7 daysin accordance with the Surveillance Frequency Control Program when the alarm is operable during power operation above 15 percent of rated power. When QPT has been restored to steady state limit, verify hourly for 12 consecutive hours, or until verified acceptable at 95% FP.

3-34a Amendment No. 29, 38, 39, 40, 45, 50, 120, 126, 142, 150, 152, 211

e. If an acceptable axial power imbalance is not achieved within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, reactor power shall be reduced to 40% FP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

f. Axial power imbalance shall be monitored on a minimum frequency of once every 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sin accordance with the Surveillance Frequency Control Program when axial power imbalance alarm is OPERABLE, and every 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when imbalance alarm is inoperable during power operation above 40 percent of rated power.

3.5.2.8 A power map shall be taken at intervals not to exceed 31 effective full power days using the incore instrumentation detection system to verify the power distribution is within the limits shown in the CORE OPERATING LIMITS REPORT.

Bases The axial power imbalance, quadrant power tilt, and control rod position limits are based on LOCA analyses which have defined the maximum linear heat rate. These limits are developed in a manner that ensures the initial condition LOCA maximum linear heat rate will not cause the maximum clad temperature to exceed 10 CFR 50 Appendix K. Operation outside of any one limit alone does not necessarily constitute a situation that would cause the Appendix K Criteria to be exceeded should a LOCA occur. Each limit represents the boundary of operation that will preserve the Acceptance Criteria even if all three limits are at their maximum allowable values simultaneously. The effects of the APSRs are included in the limit development. Additional conservatism included in the limit development is introduced by application of:

a. Nuclear uncertainty factors

b. Thermal calibration uncertainty

c. Fuel densification effects

d. Hot rod manufacturing tolerance factors

e. Postulated fuel rod bow effects

f. Peaking limits based on initial condition for Loss of Coolant Flow transients.

The incore instrumentation system uncertainties used to develop the axial power imbalance and quadrant tilt limits accounted for various combinations of invalid Self Powered Neutron Detector (SPND) signals. If the number of valid SPND signals falls below that used in the uncertainty analysis, then another system shall be used for monitoring axial power imbalance and/or quadrant tilt.

For axial power imbalance and quadrant power tilt measurements using the incore detector system, the minimum incore detector system consists of OPERABLE detectors configured as follows:

Axial Power Imbalance

a. Three detectors in each of three strings shall lie in the same axial plane with one plane in each axial core half.

b. The axial planes in each core half shall be symmetrical about the core mid-planes.

c. The detectors shall not have radial symmetry.

Quadrant Power Tilt

a. Two sets of four detectors shall lie in each core half. Each set of four shall lie in the same axial plane. The two sets in the same core half may lie in the same axial plane.

b. Detectors in the same plane shall have quarter core radial symmetry.

3-35a Amendment No. 17, 29, 38, 39, 50, 120, 126, 142, 150, 157, 168, 211

3.14 FLOOD 3.14.1 PERIODIC INSPECTION OF THE DIKES AROUND TMI Applicability Applies to inspection of the dikes surrounding the site.

Objective To specify the minimum frequency for inspection of the dikes and to define the flood stage after which the dikes will be inspected.

Specification 3.14.1.1 The dikes shall be inspected at least once every six monthsin accordance with the Surveillance Frequency Control Program and after the river has returned to normal, following the condition defined below:

a. The level of the Susquehanna River exceeds flood stage; flood stage is defined as elevation 307 feet at the Susquehanna River Gage at Harrisburg.

Bases The earth dikes are compacted to provide a stable impervious embankment that protects the site from inundation during the design flood of 1,100,000 cfs.

The rip-rap, provided to protect the dikes from wave action and the flow of the river, continues downward into natural ground for a minimum depth of two feet to prevent undermining of the dike (References 1 and 2).

Periodic inspection, and inspection of the dikes and rip-rap after the river has returned to normal from flood stage, will assure proper maintenance of the dikes, thus assuring protection of the site during the design flood.

References (1) UFSAR, Section 2.6.5 - Design of Hydraulic Facilities (2) UFSAR, Figure 2.6 Typical Dike Section 3-59 Amendment No. 157, 182

4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.

Specification 4.1.1 The minimum frequency and type of surveillance required for reactor protection system, engineered safety feature protection system, and heat sink protection system instrumentation when the reactor is critical shall be as stated in Table 4.1-1. The frequency of surveillance required for the instrumentation shown in Table 4.1-1 is specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.1-1.

4.1.2 Equipment and sampling test shall be performed as detailed in Tables 4.1-2, 4.1-3, and 4-1-5 at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Tables 4.1-2, 4.1-3, and 4-1-5.

4.1.3 Each post-accident monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the check, test and calibration at the frequencies specified in the Surveillance Frequency Control Program unless otherwise notedshown in Table 4.1-4.

4.1.4 Each remote shutdown system function shown in Table 3.5-4 shall be demonstrated OPERABLE by the performance of the following check, test, and calibration at the frequencies specified in the Surveillance Frequency Control Program:

a) Perform CHANNEL CHECK for each required instrumentation channel that is normally energized every 31 days.

b) Verify each required control circuit and transfer switch is capable of performing the intended function every refueling interval.

c) Perform CHANNEL CALIBRATION for each required instrumentation channel every refueling interval (excludes source range flux).

Bases Check Failures such as blown instrument fuses, defective indicators, or faulted amplifiers which result in "upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action. Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in surveillance. The acceptance criteria for the daily check of the Makeup Tank pressure instrument will be maintained within the error used to develop the plant operating limit. Based on experience in operation of both conventional and nuclear systems, when the unit is in operation, the minimum checking frequency stated in the Surveillance Frequency Control Program is deemed adequate for reactor system instrumentation.

4-2 Amendment No. 78, 123, 138, 156, 157,158 , 181, 216, 225, 227

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 based on operating experience, equipment reliability, and plant risk, and are controlled under the Surveillance Frequency Control Program.

The rotation schedule for the reactor protection 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 instrumentation test cycle is continued with one channel's 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 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 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 equipment testing and system sampling frequencies specified in the Surveillance Frequency Control ProgramTables 4.1-2, 4.1-3, and 4.1-5 are considered adequate to maintain the equipment and systems in a safe operational status.

The primary to secondary leakage surveillance in TS Table 4.1-2, Item 12, verifies that the sum of the primary to secondary leakage from both SGs is less than or equal to 144 gallons per day.

Satisfying the primary to secondary leakage limit ensures that the operational leakage performance criterion in the Steam Generator Program is met. If this surveillance is not met, compliance with TS 3.1.1.2, Steam Generator (SG) Tube Integrity, and TS 3.1.6.3, should be evaluated. The 144 gallons per day limit is measured at room temperature. The operational leakage rate limit applies to the sum of the leakage through both SGs.

The TS Table 4.1-2 primary to secondary leakage surveillance is modified by a Note, which states that the initial surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

The TS Table 4.1-2 primary to secondary leakage surveillance frequency specified in the Surveillance Frequency Control Programof 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable interval to trend primary to secondary leakage and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary leakage is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref.

5).

The surveillance test procedures for the Variable Low Pressure Trip Setpoint do not compare the as-found Trip Setpoint (TSP) to the previous surveillance test as-left TSP. Basing operability determinations for the as-found TSP on the Nominal Setpoint (NSP) is acceptable because:

1. The NSP as-left tolerance specified in the surveillance test procedures is less than or equal to the calculated NSP as-left tolerance.

2. The NSP as-left tolerance is not included in the Total Loop Uncertainty (TLU) calculation. This is acceptable because the NSP as-left tolerance specified in the surveillance test procedures is less than half of the calculated NSP as-left tolerance.

This prevents masking of excessive drift from one side of the tolerance band to the other.

3. The pre-defined NSP as-found tolerance is based on the square root of the sum of the square of the instrument accuracy, M&TE accuracy and drift. The NSP as-left tolerance is not included in this calculation.

Credible uncertainties for the Variable Low Pressure Trip Setpoint include instrument uncertainties during normal operation including drift and measurement and test equipment uncertainties. In no case shall the pre-defined as-found acceptance criteria band overlap the Allowable Value. If one end of the pre-defined as-found acceptance criteria band is truncated due to its proximity to the Allowable Value, this does not affect the other end of the pre-defined as-found acceptance criteria band. If equipment is replaced, such that the previous as-left value is not applicable to the current configuration, the as-found acceptance criteria band is not applicable to calibration activities performed immediately following the equipment replacement.

4-2b Amendment No. 181, 225, 255, 261, 262

Bases (Cont'd)

The TSP is stored in wire mesh baskets placed inside the containment at the 281 ft elevation.

Any quantity of TSP between 18,815 lb and 28,840 lb. will result in a pH in the desired range.

If it is discovered that the TSP in the containment building is not within limits, action must be taken to restore the TSP to within limits. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for restoring the TSP within limits, where possible, because 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is the same time allowed for restoration of other ECCS components.

Surveillance Testing Periodic determination of the mass of TSP in containment must be performed due to the possibility of leaking valves and components in the containment building that could cause dissolution of the TSP during normal operation. A Refueling FrequencyThe surveillance is required to determine that 18,815 lbs and 28,840 lbs are contained in the TSP baskets.

This requirement ensures that there is an adequate mass of TSP to adjust the pH of the post LOCA sump solution to a value 7.3 and 8.0. The periodic verification is required every refueling outagein accordance with the Surveillance Frequency Control Program.

Operating experience has shown this Surveillance Frequency to be acceptable due to the margin in the mass of TSP placed in the containment building.

Periodic testing is performed to ensure the solubility and buffering ability of the TSP after exposure to the containment environment. Satisfactory completion of this test assures that the TSP in the baskets is "active." Adequate solubility is verified by submerging a representative sample, taken via a sample thief or similar instrument, of TSP from one of the baskets in containment in un-agitated borated water heated to a temperature representing post-LOCA conditions; the TSP must completely dissolve within a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period. The test time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is to allow time for the dissolved TSP to naturally diffuse through the un-agitated test solution.

Agitation of the test solution during the solubility verification is prohibited, since an adequate standard for the agitation intensity (other than no agitation) cannot be specified. The agitation due to flow and turbulence in the containment sump during recirculation would significantly decrease the time required for the TSP to dissolve. Adequate buffering capability is verified by a measured pH of the sample solution, following the solubility verification, between 7.3 and 8.0.

The sample is cooled and thoroughly mixed prior to measuring pH. The quantity of the TSP sample, and quantity and boron concentration of the water are chosen to be representative of post-LOCA conditions. A sampling Frequency of every refueling outage is specified. Operating experience has shown this Surveillance Frequency to be acceptable.

REFERENCE (1) UFSAR, Section 7.1.2.3(d) - "Periodic Testing and Reliability" (2) NRC SER for BAW-10167A, Supplement 1, December 5, 1988.

(3) BAW-10167, May 1986.

(4) BAW-10167A, Supplement 3, February 1998.

(5) EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

4-2d Amendment No. 261, 263

TABLE 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS Amendment No. 46, 103, 123, 137, 175, 199, 255 CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS

1. Protection Channel NA Q NA Coincidence Logic

2. Control Rod Drive Trip NA Q NA (1) Includes independent testing of shunt Breaker and Regulating trip and undervoltage trip features.

Rod Power SCRs

3. Power Range Amplifier D(1) NA (2) (1) When reactor power is greater than 15%.

(2) When above 15% reactor power run a heat balance check once per shiftin accordance with the Surveillance Frequency Control Program. Heat balance calibration shall be performed whenever heat balance exceeds indicated neutron power by more than two percent.

4. Power Range Channel S S/A M(1)(2) (1) When reactor power is greater than 60% verify imbalance using incore instrumentation.

4-3 (2) When above 15% reactor power calculate axial offset upper and lower chambers after each startup if not done within the previous seven days.

5. Intermediate Range Channel S(1) P S/U NA (1) When in service.

6. Source Range Channel S(1) P S/A NA (1) When in service.

7. Reactor Coolant Temperature S S/A F Channel

TABLE 4.1-1 (Continued)

Amendment No. 47, 137, 149, 157, 175 225, 247, 255, 262 CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS

8. High Reactor Coolant S S/A R Pressure Channel

9. Low Reactor Coolant S S/A R Pressure Channel

10. Flux-Reactor Coolant Flow S S/A F Comparator

11. Reactor Coolant Pressure-Temperature S S/A R See Notes (a) and (b).

Comparator

12. Pump Flux Comparator S S/A R

13. High Reactor Building S S/A F Pressure Channel 4-4 14. High Pressure Injection NA Q NA Logic Channels

15. High Pressure Injection Analog Channels

a. Reactor Coolant S(1) M R (1) When reactor coolant system is pressurized Pressure Channel above 300 psig or Tave is greater than 200°F

16. Low Pressure Injection NA Q NA Logic Channel

17. Low Pressure Injection 0 Analog Channels

a. Reactor Coolant S(1) M R (1) When reactor coolant system is pressurized Pressure Channel above 300 psig or Tave is greater than 200°F

18. Reactor Building Emergency NA Q NA Cooling and Isolation System Logic Channel

TABLE 4.1-1 (Continued)

CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS

19. Reactor Building Emergency 4-5 Cooling and Isolation System Analog Channels

a. Reactor Building S(1) M(1) F (1) When CONTAINMENT INTEGRITY is 4 psig Channels required.

b. RCS Pressure 1600 psig S(1) M(1) NA (1) When RCS Pressure > 1800 psig.

c. Deleted

d. Reactor Bldg.30 psi S(1) M(1) F (1) When CONTAINMENT INTEGRITY is pressure switches required.

e. Reactor Bldg. Purge W(1) M(1)(2) F (1) When CONTAINMENT INTEGRITY is Line High Radiation required.

Amendment No. 24, 78, 156, 157, 175, 189, 200, 225 (AH-V-1A/D)

f. Line Break Isolation W(1) M(1) R (1) When CONTAINMENT INTEGRITY is Signal (ICCW & NSCCW) required.

20. Reactor Building Spray NA Q NA System Logic Channel

21. Reactor Building Spray NA M F 30 psig pressure switches

22. Pressurizer Temperature S NA R Channels

23. Control Rod Absolute Position S(1) NA R (1) Check with Relative Position Indicator

24. Control Rod Relative Position S(1) NA R (1) Check with Absolute Position Indicator

25. Core Flooding Tanks

a. Pressure Channels Coolant NA NA F

b. Level Channels NA NA F

26. Pressurizer Level Channels S NA R

TABLE 4.1-1 (Continued)

Page 4-5a CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS

27. Makeup Tank Instrument Channels:

a. Level D(1) NA R (1) When Makeup and Purification System is in operation.

b. Pressure D(1) NA R Amendment Nos. 24, 78, 100, 108, 156, 161, 175, 197, 212,

28. Radiation Monitoring Systems*

a. DELETED (1) Using the installed check source when background is less than twice the expected

b. DELETED increase in cpm which would result from the check source alone. Background readings

c. DELETED greater than this value are sufficient in themselves to show that the monitor is 227, 260

d. RM-A2P (RB Atmosphere particulate)

W(1)(4) M(4) E(4) functioning.

e. RM-A21 (RB Atmosphere iodine)

W(1)(4) M(4) Q(4) (2) DELETED

f. RM-A2G (RB Atmosphere gas)

W(1)(4) M(4) E(4) (3) DELETED (4) RM-A2 operability requirements are given in T.S. 3.1.6.8 Corrected by letter dated July 8, 1999

29. High and Low Pressure N/A N/A R Injection Systems:

Flow Channels

• Includes only monitors indicated under this item. Other T.S. required radiation monitors are included in specifications 3.5.5.2, 4.1.3, Table 3.5-1 item C.3.f, and Table 4.1-1 item 19e.

TABLE 4.1-1 (Continued)

CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS Amendment Nos. 175, 212, 225, 227

30. Borated Water Storage W NA R Tank Level Indicator

31. DELETED

32. DELETED

33. Containment Temperature NA NA F

34. Incore Neutron Detectors M(1) NA NA (1) Check functioning; including functioning of computer readout or recorder readout when reactor power is greater than 15%.

Page 4-6

35. Emergency Plant Radiation M(1) NA F (1) Battery Check.

Instruments

36. (DELETED)

37. Reactor Building Sump NA NA R Level

TABLE 4.1-1 (Continued)

Amendment No. 70, 78, 80, 124, 135, 175, 224, 255, 263 CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS

38. OTSG Full Range Level W NA R

39. Turbine Overspeed Trip NA R NA

40. Deleted

41. Deleted

42. Diesel Generator NA NA R Protective Relaying

43. 4 KV ES Bus Undervoltage Relays (Diesel Start)

a. Degraded Grid NA M(1) A (1) Relay operation will be checked by local test pushbuttons.

b. Loss of Voltage NA M(1) R (1) Relay operation will be checked by 4-7 local test pushbuttons.

44. Reactor Coolant Pressure S(1) M R (1) When reactor coolant system is DH Valve Interlock Bistable pressurized above 300 psig or Tave is greater than 200°F.

45. Loss of Feedwater Reactor Trip S(1) S/A(1) R (1) When reactor power exceeds 7%

power.

46. Turbine Trip/Reactor Trip S(1) S/A(1) F (1) When reactor power exceeds 45%

power.

47. a. Pressurizer Code Safety Valve S(1) NA F (1) When Tave is greater than 525°F.

and PORV Tailpipe Flow Monitors

b. PORV - Acoustic/Flow NA M(1) R (1) When Tave is greater than 525°F.

48. PORV Setpoints NA M(1) R (1) Per Specification 3.1.12 excluding valve operation.

TABLE 4.1-1 (Continued)

CHANNEL DESCRIPTION CHECK(c) TEST(c) CALIBRATE(c) REMARKS Amendment No. 78, 105, 124, 135, 137, 147, 175, 182, 191, 262

49. Saturation Margin Monitor S(1) M(1) R (1) When Tave is greater than 525°F.

50. Emergency Feedwater Flow NA M(1) F (1) When Tave is greater than 250°F.

Instrumentation

51. Heat Sink Protection System

a. EFW Auto Initiation (1) Includes logic test only.

Instrument Channels

1. Loss of both Feedwater NA Q(1) F Pumps

2. Loss of All RC Pumps NA Q(1) R

3. Reactor Building NA Q F Pressure

4. OTSG Low Level W Q R 4-7a

b. MFW Isolation OTSG Low NA Q R Pressure

c. EFW Control Valve Control System

1. OTSG Level Loops W Q R

2. Controllers W NA R

d. HSPS Train Actuation Logic NA Q(1) R

52. Backup Incore Thermocouple M(1) NA R (1) When Tave is greater than 250°F.

Display

53. Deleted

54. Reactor Vessel Water Level NA NA R Notes (a) If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found tolerance then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. Enter condition into Corrective Action Program.

(b) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conversative than the NSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures to confirm channel performance. The NSP and the methodologies used to determine the as-found and the as-left tolerances are specified in a document incorporated by reference into the UFSAR.

(c) Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency

1. Control Rods Rod drop times of all Note 1Each Refueling shutdown full length rods

2. Control Rod Movement of each rod Note 1Every 92 days, when Movement reactor is critical

3. Pressurizer Setpoint In accordance with the Safety Valves Inservice Testing Program

4. Main Steam Setpoint In accordance with the Safety Valves Inservice Testing Program

5. Refueling System Functional Start of each Interlocks refueling period

6. (Deleted) -- --

7. Reactor Coolant Evaluate Note 1Daily, when reactor System Leakage coolant system temperature is greater than 525 degrees F (Not applicable to primary-to-secondary leakage.)

8. (Deleted) -- --

9. Spent Fuel Functional Each refueling period Cooling System prior to fuel handling

10. Intake Pump (a) Silt Accumulation - Note 1Not to exceed 24 months House Floor Visual inspection (Elevation of Intake Pump 262 ft. 6 in.) House Floor (b) Silt Accumulation Note 1Quarterly Measurement of Pump House Flow

11. Pressurizer Block Functional* Note 1Quarterly Valve (RC-V2)

12. Primary to Secondary Evaluate Note 1Every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (Note: Not required Leakage to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.)

• Function shall be demonstrated by operating the valve through one complete cycle of full travel.

Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

4-8 Amendment No. 55, 68, 78, 149, 175, 198, 211, 246, 261

TABLE 4.1-3 MINIMUM SAMPLING FREQUENCY Amendment No. 62, 95, 108, 204, 211, 272 Item Check Frequency

1. Reactor Coolant a. Verify reactor coolant DOSE EQUIVALENT Xe-133 i) At least once each 7 days Note 1 (during all plant conditions except specific activity is less than or equal to 797 REFUELING SHUTDOWN and COLD SHUTDOWN).

microcuries/gram.

ii) One Sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a one hour period during all plant conditions except REFUELING SHUTDOWN and COLD SHUTDOWN.

b. Isotopic Analysis for DOSE EQUIVALENT i) 1 per 14 days Note 1 (during power operations).

I-131 Concentration ii) One Sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a one hour period during all plant conditions except REFUELING SHUTDOWN and COLD SHUTDOWN.

iii) # Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the specific activity exceeds 4-9 0.35 Ci/gram DOSE EQUIVALENT I-131 during all plant conditions except REFUELING SHUTDOWN and COLD Corrected by ltr dtd 07/08/99 SHUTDOWN.

c. Deleted

d. Chemistry (Cl, F and O2) 5 times/week Note 1 (when Tavg is greater than 200°F).

e. Boron concentration 2 times/weekNote 1

f. Tritium Radioactivity MonthlyNote 1

2. Borated Water Boron concentration Weekly Note 1 and after each makeup when reactor coolant Storage Tank system pressure is greater than 300 psig or Tavg is greater Water Sample than 200°F.

3. Core Flooding Tank Boron concentration Monthly Note 1 and after each makeup when RCS pressure is Water Sample greater than 700 psig.

TABLE 4.1-3 Contd Item Check Frequency Amendment No. 62, 80, 95, 108, 115, 138, 200, 225, 263

4. Spent Fuel Pool Boron Concentration greater than WeeklyNote 1 Water Sample or equal to 600 ppmb

5. Secondary Coolant Isotopic analysis for DOSE At least once per 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sNote 1 (when EQUIVALENT I-131 concentration reactor coolant system pressure is greater than 300 psig or Tav is greater than 200°F.)

6. Deleted

7. Deleted

8. Deleted

9. Deleted

10. Deleted 4-10

11. Deleted

12. Deleted

1. Until the specific activity of the primary coolant system is restored within its limits.

• Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

• • Deleted

• • • Deleted Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

TABLE 4.1-4 POST ACCIDENT MONITORING INSTRUMENTATION Amendment No. 100, 144, 175, 240 FUNCTION INSTRUMENTS CHECK(a) TEST(a) CALIBRATE(a) REMARKS 1 Noble Gas Effluent

a. Condenser Vacuum Pump Exhaust (RM-A5-Hi) W M F (1) Using the installed check source when background is less that twice the expected increase in cpm which would result from the check source alone. Background readings greater than this value are sufficient in themselves to show that this monitor is functioning.

b. Condenser Vacuum Pump W(1) M F Exhaust (RM-G25)

c. Auxiliary and Fuel Handling W M F Building Exhaust (RM-A8-Hi) 4-10a d. Reactor Building Purge W M F Exhaust (RM-A9-Hi)

e. Reactor Building Purge W(1) M F Exhaust (RM-G24)

f. Main Steam Lines W(1) M F Radiation (RM-G26/RM-G27)

2. Containment High Range W M R Radiation (RM-G22/G23)

3. Containment Pressure W N/A F

4. Containment Water Level W N/A R

5. DELETED

6. Wide Range Neutron Flux W N/A F

TABLE 4.1-4 (Continued)

POST ACCIDENT MONITORING INSTRUMENTATION Amendment No. 100, 144, 175, 240 FUNCTION INSTRUMENTS CHECK(a) TEST(a) CALIBRATE(a) REMARKS

7. Reactor Coolant System Cold Leg W N/A R Water Temperature (TE-959, 961; TI-959A, 961A)

8. Reactor Coolant System Hot Leg W N/A R (TE-958, 960; TI-958A, 960A)

9. Reactor Coolant System Pressure W N/A R (PT-949, 963; PI-949A, 963)

10. Steam Generator Pressure W N/A R (PT-950, 951, 1180, 1184; PI-950A, 951A, 1180, 1184) 4-10b

11. Condensate Storage Tank Water W N/A F Level (LT-1060, 1061, 1062, 1063; LI-1060, 1061, 1062, 1063)

(a) Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

TABLE 4.1-5 SYSTEM SURVEILLANCE REQUIREMENTS Item Test Frequency

1. Core Flood Tank a. Verify two core flood tanks Note 1S each contain 940 +/- 30 ft3 borated water.

b. Verify that two core flood Note 1S tanks each contain 600 +/- 25 psig.

c. Verify CF-V-1A&B are fully open. Note 1S

d. Verify power is removed from Note 1M CF-V-1A&B and CF-V-3A&B valve operators

2. Reactor Building a. Verify the TSP baskets Note 1R Emergency Sump contain 18,815 lbs and pH Control 28,840 lbs of TSP.

System

b. Verify that a sample from Note 1R the TSP baskets provides adequate pH adjustment of borated water.

Note 1 Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

4-10c Amendment No. 225, 263

4.4 REACTOR BUILDING 4.4.1 CONTAINMENT LEAKAGE TESTS Applicability Applies to containment leakage.

Objective To verify that leakage from the reactor building is maintained within allowable limits.

Specification 4.4.1.1 Integrated Leakage Rate Testing (ILRT) shall be conducted in accordance with the Reactor Building Leakage Rate Testing Program at test frequencies established in accordance with the Reactor Building Leakage Rate Testing Program.

4.4.1.2 Local Leakage Rate Testing (LLRT) shall be conducted in accordance with the Reactor Building Leakage RateTesting Program. LLRT shall be performed at a pressure not less than peak accident pressure Pac with the exception that the airlock door seal tests shall normally be performed at 10 psig and the periodic containment airlock tests shall be performed at a pressure not less than Pac . LLRT frequencies shall be in accordance with the Reactor Building Leakage Rate Testing Program.

4.4.1.3 Operability of the personnel and emergency air lock door interlocks and the associated control room annunciator circuits shall be determined at least once per six months in accordance with the Surveillance Frequency Control Program. If the interlock permits both doors to be open at the same time or does not provide accurate status indication in the control room, the interlock shall be declared inoperable, except as provided in Technical Specification Section 3.8.6.

Bases (1)

The Reactor Building is designed to limit the leakage rate to 0.1 percent by weight of contained atmosphere in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design internal pressure of 55 psig with a coincident temperature of 281°F at accident conditions. The peak calculated Reactor Building pressure for the design basis loss of coolant accident, Pac, is 50.6 psig. The maximum allowable Reactor Building leakage rate, La, shall be 0.1 weight percent of containment atmosphere per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pac. Containment Isolation Valves are addressed in the UFSAR (Reference 2).

4-29 Amendment No. 63, 167, 201, 236

4.4 REACTOR BUILDING (Continued)

The Reactor Building will be periodically leakage tested in accordance with the Reactor Building Leakage Rate Testing Program (See Section 6.8.5). This program is contained in the surveillance procedures for Reactor Building inspection, Integrated Leak Rate Testing, and Local Leak Rate Testing. These periodic testing requirements verify that Reactor Building leakage rate does not exceed the assumptions used in the safety analysis. At 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C leakage, and 0.75 La for overall Type A leakage. At all other times between required leakage tests, the acceptance criteria is based on an overall Type A leakage limit of 1.0 La.

Periodic surveillance of the airlock interlock systems (Reference 4) is specified to assures continued operability and precludes instances where one or both doors are inadvertently left open. When an airlock is inoperable and containment integrity is required, local supervision of airlock operation is specified.

References (1) UFSAR, Chapter 5.7.4 - Post Operational Leakage Rate Tests (2) UFSAR, Tables 5.7-1 and 5.7-3 (3) DELETED (4) UFSAR, Table 5.7-2 4-30 (Pages 4-31 through 4-34, 4-34a, and 4-34b deleted)

Amendment No. 27, 167, 201

4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER, EMERGENCY CORE COOLING SYSTEM & REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 Emergency Loading Sequence Applicability: Applies to periodic testing requirements for safety actuation systems.

Objective: To verify that the emergency loading sequence and automatic power transfer is operable.

Specifications:

4.5.1.1 Sequence and Power Transfer Test

a. During each refueling intervalIn accordance with the Surveillance Frequency Control Program, a test shall be conducted to demonstrate that the emergency loading sequence and power transfer is operable.

b. The test will be considered satisfactory if the following pumps and fans have been successfully started and the following valves have completed their travel on preferred power and transferred to the emergency power.

-M. U. Pump

-D. H. Pump and D. H. Injection Valves and D. H. Supply Valves

-R. B. Cooling Pump

-R. B. Ventilators

-D. H. Closed Cycle Cooling Pump

-N. S. Closed Cycle Cooling Pump

-D. H. River Cooling Pump

-N. S. River Cooling Pump

-D. H. and N. S. Pump Area Cooling Fan

-Screen House Area Cooling Fan

-Spray Pump. (Initiated in coincidence with a 2 out of 3 R. B.

30 psig Pressure Test Signal.)

-Motor Driven Emergency Feedwater Pump

c. Following successful transfer to the emergency diesel, the diesel generator breaker will be opened to simulate trip of the generator then re-closed to verify block load on the reclosure.

4.5.1.2 Sequence Test

a. At intervals not to exceed 3 monthsIn accordance with the Surveillance Frequency Control Program, a test shall be conducted to demonstrate that the emergency loading sequence is operable, this test shall be performed on either preferred power or emergency power.

b. The test will be considered satisfactory if the pumps and fans listed in 4.5.1.1b have been successfully started and the valves listed in 4.5.1.1b have completed their travel.

4-39 Amendment No. 70, 78, 149, 167, 212

4.5.2 EMERGENCY CORE COOLING SYSTEM Applicability: Applies to periodic testing requirement for emergency core cooling systems.

Objective: To verify that the emergency core cooling systems are operable.

Specification 4.5.2.1 High Pressure Injection

a. During each refueling interval In accordance with the Surveillance Frequency Control Program and following maintenance or modification that affects system flow characteristics, system pumps and system high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable.

b. The test will be considered satisfactory if the valves (MU-V-14A/B

& 16A/B/C/D) have completed their travel and the make-up pumps are running as evidenced by system flow. Minimum acceptable injection flow must be greater than or equal to 431 gpm per HPI pump when pump discharge pressure is 600 psig or greater (the pressure between the pump and flow limiting device) and when the RCS pressure is equal to or less than 600 psig.

c. Testing which requires HPI flow thru MU-V16A/B/C/D shall be conducted only under either of the following conditions:

1) Indicated RCS temperature shall be greater than 329°F.

2) Head of the Reactor Vessel shall be removed.

4.5.2.2 Low Pressure Injection

a. During each refueling period In accordance with the Surveillance Frequency Control Program and following maintenance or modification that affects system flow characteristics, system pumps and high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable. The auxiliaries required for low pressure injection are all included in the emergency loading sequence specified in 4.5.1.

b. The test will be considered satisfactory if the decay heat pumps listed in 4.5.1.1b have been successfully started and the decay heat injection valves and the decay heat supply valves have completed their travel as evidenced by the control board component operating lights. Flow shall be verified to be equal to or greater than the flow assumed in the Safety Analysis for the single corresponding RCS pressure used in the test.

4-41 Amendment No. 19, 57, 68, 149, 203, 225, 234

c. When the Decay Heat System is required to be operable, the correct position of DH-V-19A/B shall be verified by observation within four hours of each valve stroking operation or valve maintenance which affects the position indicator.

4.5.2.3 Core Flooding

a. During each refueling periodIn accordance with the Surveillance Frequency Control Program, a system test shall be conducted to demonstrate proper operation of the system. Verification shall be made that the check and isolation valves in the core cooling flooding tank discharge lines operate properly.

b. The test will be considered satisfactory if control board indication of core flooding tank level verifies that all valves have opened.

4.5.2.4 Component Tests

a. At intervals not to exceed 3 monthsIn accordance with the Surveillance Frequency Control Program, the components required for emergency core cooling will be tested.

b. The test will be considered satisfactory if the pumps and fans have been successfully started and the valves have completed their travel as evidenced by the control board component operating lights, and a second means of verification, such as: the station computer, verification of pressure/flow, or control board indicating lights initiated by separate limit switch contacts.

Bases The emergency core cooling systems (Reference 1) are the principal reactor safety features in the event of a loss of coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.

The low pressure injection pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the bypass valves in the borated water storage tank fill line. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank.

The minimum acceptable HPI/LPI flow assures proper flow and flow split between injection legs.

With the reactor shutdown, the valves in each core flooding line are checked for operability by reducing the reactor coolant system pressure until the indicated level in the core flood tanks verify that the check and isolation valves have opened.

Reference (1) UFSAR, Section 6.1 - "Emergency Core Cooling System" 4-42 Amendment No.57, 68, 149, 157, 167, 225

4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM Applicability Applies to testing of the reactor building cooling and isolation systems.

Objective To verify that the reactor building cooling systems are operable Specification 4.5.3.1 System Tests

a. Reactor Building Spray System

1. At each refueling interval In accordance with the Surveillance Frequency Control Program and simultaneously with the test of the emergency loading sequence, a reactor building 30 psi high pressure test signal will start the spray pump. Except for the spray pump suction valves, all engineered safeguards spray valves will be closed.

Water will be circulated from the borated water storage tank through the reactor building spray pumps and returned through the test line to the borated water storage tank.

The operation of the spray valves will be verified during the component test of the R. B. Cooling and Isolation System.

The test will be considered satisfactory if the spray pumps have been successfully started.

2. Compressed air will be introduced into the spray headers to verify each spray nozzle is unobstructed at least every ten yearsin accordance with the Surveillance Frequency Control Program.

b. Reactor Building Cooling and Isolation Systems

1. During each refueling periodIn accordance with the Surveillance Frequency Control Program, a system test shall be conducted to demonstrate proper operation of the system.

2. The test will be considered satisfactory if measured system flow is greater than accident design flow rate.

4-43 Amendment No. 167, 198, 212, 225

4.5.3.2 Component Tests

a. At intervals not to exceed three monthsIn accordance with the Surveillance Frequency Control Program, the components required for Reactor Building Cooling and Isolation will be tested.

b. The test will be considered satisfactory if the valves have completed their expected ravel as evidenced by the control board component operating lights, and a second means of verification, such as: the station computer, local verification, verification of pressure/flow, or control board component operating lights initiated by separate limit switch contacts.

Bases The Reactor Building Cooling and Isolation Systems and Reactor Building Spray System are designed to remove the heat in the containment atmosphere to prevent the building pressure from exceeding the design pressure (References 1 and 2).

The delivery capability of one Reactor Building Spray Pump at a time can be tested by opening the valve in the line from the borated water storage tank, opening the corresponding valve in the test line, and starting the corresponding pump.

With the pumps shut down and the Borated Water Storage Tank outlet closed, the Reactor Building spray injection valves can each be opened and closed by the operator action. With the Reactor Building spray inlet valves closed, low pressure air can be blown through the test connections of the Reactor Building spray nozzles to demonstrate that the flow paths are open.

The equipment, piping, valves and instrumentation of the Reactor Building Cooling System are arranged so that they can be visually inspected. The cooling units and associated piping are located outside the secondary concrete shield.

Personnel can enter the Reactor Building during power operations to inspect and maintain this equipment.

The Reactor Building fans are normally operating periodically, constituting the test that these fans are operable.

Reference (1) UFSAR, Section 6.2 - "Reactor Building Spray System" (2) UFSAR, Section 6.3 - "Reactor Building Emergency Cooling System" 4-44 Amendment No. 68, 149, 157, 167

4.5.4 ENGINEERED SAFEGUARDS FEATURE (ESF) SYSTEMS LEAKAGE Applicability Applies to those portions of the Decay Heat, Building Spray, and Make-Up Systems, which are required to contain post accident sump recirculation fluid, when these systems are required to be operable in accordance with Technical Specification 3.3.

Objective To maintain a low leakage rate from the ESF systems in order to prevent significant off-site exposures and dose consequences.

Specification 4.5.4.1 The total maximum allowable leakage into the Auxiliary Building from the applicable portions of the Decay Heat, Building Spray and Make-Up System components as measured during refueling interval tests in Specification 4.5.4.2 shall not exceed 15 gallons per hour.

4.5.4.2 Once each refueling interval In accordance with the Surveillance Frequency Control Program the following tests of the applicable portions of the Decay Heat Removal, Building Spray and Make-Up Systems shall be conducted to determine leakage:

a. The applicable portion of the Decay Heat Removal System that is outside containment shall be leak tested with the Decay Heat pump operating, except as specified in b.

b. Piping from the Reactor Building Sump to the Building Spray pump and Decay Heat Removal System pump suction isolation valves shall be pressure tested at no less than 55 psig.

c. The applicable portion of the Building Spray system that is outside containment shall be leak tested with the Building Spray pumps operating and BS-V-1A/B closed, except as specified in b above.

c. The applicable portion of the Make-Up system on the suction side of the Make-Up pumps shall be leak tested with a Decay Heat pump operating and DH-V-7A/B open.

d. The applicable portion of the Make-Up system from the Make-Up pumps to the containment boundary valves (MU-V-16A/D, 18, and 20) shall be leak tested with a Make-Up pump operating.

f. Visual inspection shall be made for leakage from components of these systems. Leakage shall be measured by collection and weighing or by another equivalent method.

Bases The leakage rate limit of 15 gph (measured in standard room temperature gallons) for the accident recirculation portions of the Decay Heat Removal (DHR), Building Spray (BS), and Make-Up (MU) systems is based on ensuring that potential leakage after a loss-of-coolant accident will not result in off-site dose consequences in excess of those calculated to comply with the 10 CFR 50.67 limits (Reference 1 and 2). The test methods prescribed in 4.5.4.2 above for the applicable portions of the DH, BS and MU systems ensure that the testing results account for the highest pressure within that system during the sump recirculation phase of a design basis accident.

References (1) UFSAR, Section 6.4.4 - "Design Basis Leakage" (2) UFSAR, Section 14.2.2.5(d) - "Effects of Engineered Safeguards Leakage During Maximum Hypothetical Accident" 4-45 Amendment No. 157, 205, 215, Corrected by letter dated: 9/24/99, 235

4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS Applicability: Applies to periodic testing and surveillance requirement of the emergency power system Objective: To verify that the emergency power system will respond promptly and properly when required.

Specification:

The following tests and surveillance shall be performed as stated:

4.6.1 Diesel Generators

a. Manually-initiate start of the diesel generator, followed by manual synchronization with other power sources and assumption of load by the diesel generator up to the name-plate rating (3000 kw). This test will be conducted every month in accordance with the Surveillance Frequency Control Program on each diesel generator. Normal plant operation will not be effected.

b. Automatically start and loading the emergency diesel generator in accordance with Specification 4.5.1.1.b/c including the following. This test will be conducted every refueling interval in accordance with the Surveillance Frequency Control Program on each diesel generator.

(1) Verify that the diesel generator starts from ambient condition upon receipt of the ES signal and is ready to load in 10 seconds.

(2) Verify that the diesel block loads upon simulated loss of offsite power in 30 seconds.

(3) The diesel operates with the permanently connected and auto connected load for 5 minutes.

(4) The diesel engine does not trip when the generator breaker is opened while carrying emergency loads.

(5) The diesel generator block loads and operates for 5 minutes upon reclosure of the diesel generator breaker.

c. Deleted.

4.6.2 Station Batteries

a. The voltage, specific gravity, and liquid level of each cell will be measured and recorded:

(1) every 92 days in accordance with the Surveillance Frequency Control Program (2) once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a battery discharge <105 V (3) once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a battery overcharge >150 V (4) If any cell parameters are not met, measure and record the parameters on each connected cell every 7 days thereafter until all battery parameters are met.

b. The voltage and specific gravity of a pilot cell will be measured and recorded weekly in accordance with the Surveillance Frequency Control Program. If any pilot cell parameters are not met, perform surveillance 4.6.2.a on each connected cell within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and every 7 days thereafter until all battery parameters are met.

c. Each time data is recorded, new data shall be compared with old to detect signs of abuse or deterioration.

4-46 Amendment No.70, 149, 157,200, 232,243

d. The battery will be subjected to a load test on a refueling interval 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 surveillance specified is that which has been demonstrated over the years to provide an indication of a cell becoming unserviceable long before it failsintervals are based on operating experience, equipment reliability, and plant risk, and are controlled under the Surveillance Frequency Control Program (SFCP).

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

4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS Applicability Applies to the surveillance of the control rod system.

Objective To assure operability of the control rod system.

Specification 4.7.1.1 The control rod trip insertion time shall be measured for each control rod at either full flow or no flow conditions following each refueling outage prior to return to power. The maximum control rod trip insertion time for an operable control rod drive mechanism, except for the axial power shaping rods (APSRs), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1.66 seconds at hot reactor coolant full flow conditions or 1.40 seconds for the hot no flow conditions (Reference 1). For the APSRs it shall be demonstrated that loss of power will not cause rod movement. If the trip insertion time above is not met, the rod shall be declared inoperable.

4.7.1.2 If a control rod is misaligned with its group average by more than an indicated nine inches, the rod shall be declared inoperable and the limits of Specification 3.5.2.2 shall apply. The rod with the greatest misalignment shall be evaluated first. The position of a rod declared inoperable due to misalignment shall not be included in computing the average position of the group for determining the operability of rods with lesser misalignments.

4.7.1.3 If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indications or in or out limit lights, the rod shall be declared to be inoperable.

Bases The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has actuated the 25% withdrawn reference switch during insertion from the fully withdrawn position. The specified trip time is based upon the safety analysis in UFSAR, Chapter 14 and the Accident Parameters as specified therein.

Each control rod drive mechanism shall be exercised by a movement of a minimum of 3%

of travel every 92 daysin accordance with the Surveillance Frequency Control Program.

This requirement shall apply to either a partial or fully withdrawn control rod at reactor operating conditions. Exercising the drive mechanisms in this manner provides assurance of reliability of the mechanisms.

4-48 Amendment No. 157, 211

4.9 DECAY HEAT REMOVAL (DHR) CAPABILITY - PERIODIC TESTING Applicability Applies to the periodic testing of systems or components which function to remove decay heat.

Objective To verify that systems/components required for DHR are capable of performing their design function.

Specification 4.9.1 Reactor Coolant System (RCS) Temperature greater than 250 degrees F.

4.9.1.1 Verify each Emergency Feedwater (EFW) Pump is tested in accordance with the requirements and acceptance criteria of the Inservice Test Program.

Note: This surveillance is not required to be performed for the turbine-driven EFW Pump (EF-P-1) until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 750 psig.

4.9.1.2 DELETED 4.9.1.3 At least once per 31 daysIn accordance with the Surveillance Frequency Control Program, each EFW System flowpath valve from both Condensate Storage Tanks (CSTs) to the OTSGs via the motor-driven pumps and the turbine-driven pump shall be verified to be in the required status.

4.9.1.4 On a refueling interval basisIn accordance with the Surveillance Frequency Control Program:

a) Verify that each EFW Pump starts automatically upon receipt of an EFW test signal.

b) Verify that each EFW control valve responds upon receipt of an EFW test signal.

c) Verify that each EFW control valve responds in manual control from the control room and remote shutdown panel.

4.9.1.5 Prior to STARTUP, following a REFUELING SHUTDOWN or a COLD SHUTDOWN greater than 30 days, conduct a test to demonstrate that the motor driven EFW Pumps can pump water from the CSTs to the Steam Generators.

4-52 Amendment No. 78, 119, 124, 172, 242, 266

4.9 DECAY HEAT REMOVAL (DHR) CAPABILITY-PERIODIC TESTING (Continued) 4.9.1.6 Acceptance Criteria These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly except for the tests required by Specification 4.9.1.1.

4.9.2 RCS Temperature less than or equal to 250 degrees F.*

4.9.2.1 On a daily basisIn accordance with the Surveillance Frequency Control Program, verify operability of the means for DHR required by Specification 3.4.2 by observation of console status indication.

• These requirements supplement the requirements of Specifications 4.5.2.2 and 4.5.4.

Bases The ASME Code specifies requirements and acceptance standards for the testing of nuclear safety related pumps. The quarterly EFW Pump test frequency specified by the ASME Code will be sufficient to verify that the turbine-driven and both motor-driven EFW Pumps are operable. Compliance with the normal acceptance criteria assures that the EFW Pumps are operating as expected. The surveillance requirements ensure that the overall EFW System functional capability is maintained.

Deferral of the requirement to perform IST on the turbine-driven EFW Pump is necessary to assure sufficient OTSG pressure to perform the test using Main Steam.

Daily Periodic verification of the operability of the required means for DHR ensures that sufficient DHR capability will be maintained.

4-52a Amendment No. 78, 119, 124, 172, 242, 266

4.11 REACTOR COOLANT SYSTEM VENTS Applicability Applies to Reactor Coolant System Vents.

Objective To ensure that Reactor Coolant System vents are able to perform their design function.

Specification 4.11.1 Each reactor coolant system vent path shall be demonstrated OPERABLE once per refueling interval in accordance with the Surveillance Frequency Control Program by cycling each power operated valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING.

BASES Frequency of tTests specified above are necessary to ensure that the individual Reactor Coolant System Vents will perform their functions. It is not advisable to perform these tests during Plant Power Operation, or when there is significant pressure in the Reactor Coolant System. Tests are, therefore, to be performed during either Cold Shutdown or Refueling.

4-54 Amendment No. 95, 97

4.12 AIR TREATMENT SYSTEM 4.12.1 EMERGENCY CONTROL ROOM AIR TREATMENT SYSTEM Applicability Applies to the emergency control room air treatment system and associated components.

Objective To verify that this system and associated components will be able to perform its design functions.

Specification 4.12.1.1 At least every refueling intervalIn accordance with the Surveillance Frequency Control Program, the pressure drop across the combined HEPA filters and charcoal adsorber banks of AH-F3A and 3B shall be demonstrated to be less than 6 inches of water at system design flow rate (+/-10%).

4.12.1.2 a. The tests and sample analysis required by Specification 3.15.1.2 shall be performed initially and at least once per year in accordance with the Surveillance Frequency Control Program for standby service or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and following significant painting, steam, fire or chemical release in any ventilation zone communicating with the system that could contaminate the HEPA filters or charcoal adsorbers.

b. DOP testing shall be performed after each complete or partial replacement of the HEPA filter bank or after any structural maintenance on the system housing which could affect the HEPA filter bank bypass leakage.

c. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of the charcoal adsorber bank or after any structural maintenance on the system housing which could effect the charcoal adsorber bank bypass leakage.

d. Each AH-E18A and B (AH-F3A and B) fan/filter circuit shall be operating at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every monthat the frequency specified in the Surveillance Frequency Control Program.

4.12.1.3 At least once per refueling intervalIn accordance with the Surveillance Frequency Control Program, automatic initiation of the required Control Building dampers for isolation and recirculation shall be demonstrated as operable.

4.12.1.4 An air distribution test shall be performed on the HEPA filter bank initially, and after any maintenance or testing that could affect the air distribution within the system . The air distribution across the HEPA filter bank shall be uniform within

+/-20%. The test shall be performed at 40,000 cfm (+/-10%) flow rate.

4.12.1.5 Control Room Envelope unfiltered air inleakage testing shall be performed in accordance with the Control Room Envelope Habitability Program.

4-55 Amendment No.55, 68, 149, 175, 223, 264

BASES Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be determined at least once per refueling cycle in accordance with the Surveillance Frequency Control Program to show system performance capability.

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon shall be performed in accordance with approved test procedures.

Replacement adsorbent should be qualified according to ASTM D3803-1989. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples.

Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable all adsorbent in the system shall be replaced. Tests of the HEPA filters with DOP aerosol shall also be performed in accordance with approved test procedures. Any HEPA filters found defective should be replaced with filters qualified according to Regulatory Guide 1.52 March 1978.

Operation of the system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month at the frequency specified in the Surveillance Frequency Control Program will demonstrate operability of the filters and adsorber system and remove excessive moisture built up on the adsorber.

If significant painting, steam, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign materials, the same tests and sample analysis shall be performed as required for operational use. The determination of significance shall be made by the Vice President-TMI Unit 1.

Demonstration of the automatic initiation of the recirculation mode of operation is necessary to assure system performance capability. Dampers required for control building isolation and recirculation are specified in UFSAR Sections 7.4.5 and 9.8.1.

Control Room Envelope unfiltered air inleakage testing verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.

The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous chemicals and smoke. Air inleakage testing verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, Section 3.15.1.5 must be entered. The required actions allow time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident. Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 1) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 2).

These compensatory measures may also be used as mitigating actions as required by Section 3.15.1.5. Temporary analytical methods may also be used as compensatory measures to 4-55a Amendment No. 55, 179, 218, 223, 226, 264

4.12.4 FUEL HANDLING BUILDING ESF AIR TREATMENT SYSTEM Applicability Applies to Fuel Handling Building (FHB) ESF Air Treatment System and associated components.

Objective To verify that this system and associated components will be able to perform its design functions.

Specification 4.12.4.1 Each refueling interval prior to movement of irradiated fuel:

a. The pressure drop across the entire filtration unit shall be demonstrated to be less than 7.0 inches of water at 6,000 cfm flow rate (+/-10%).

b. The tests and sample analysis required by Specification 3.15.4.2 shall be performed.

4.12.4.2 Testing necessary to demonstrate operability shall be performed as follows:

a. The tests and sample analysis required by Specification 3.15.4.2 shall be performed following significant painting, steam, fire, or chemical release in any ventilation zone communicating with the system that could contaminate the HEPA filters or charcoal adsorbers.

b. DOP testing shall be performed after each complete or partial replacement of a HEPA filter bank, and after any structural maintenance on the system housing that could affect the HEPA filter bank bypass leakage.

c. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a charcoal adsorber bank, and after any structural maintenance on the system housing that could affect charcoal adsorber bank bypass leakage.

4.12.4.3 Each filter train shall be operated at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month at the frequency specified in the Surveillance Frequency Control Program.

4.12.4.4 An air flow distribution test shall be performed on the HEPA filter bank initially and after any maintenance or testing that could affect the air flow distribution within the system. The distribution across the HEPA filter bank shall be uniform within

+/-20%. The test shall be performed at 6,000 cfm +/- 10% flow rate.

4-55f Amendment No. 122

Bases The FHB ESF Air Treatment System is a system which is normally kept in a "standby" operating status.

Tests and sample analysis assure that the HEPA filters and charcoal adsorbers can perform as evaluated.

The charcoal adsorber efficiency test procedure should allow for the removal of a sample from one adsorber test canister. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. The in-place test criteria for activated charcoal will meet the guidelines of ANSI-N510-1980. The laboratory test of charcoal will be performed in accordance with ASTM D3803-1989. If laboratory test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified in accordance with ASTM D3803-1989. Any HEPA filters found defective will be replaced with filters qualified in accordance with ANSI-N509-1980.

Pressure drop across the entire filtration unit of less than 7.0 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter.

Operation of the system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month at the frequency specified in the Surveillance Frequency Control Program will demonstrate operability of the filters and adsorber system and remove excessive moisture buildup on the adsorbers and HEPA filters.

If significant painting, steam, fire, or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational movement of irradiated fuel. The determination of what is significant shall be made by the Vice President-TMI Unit 1.

4-55g Amendment No. 122, 157, 179, 218, 226

4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE Applicability Applies to Reactor Internals Vent Valves.

Objective To verify that no reactor internals vent valve is stuck in the open position and that each valve continues to exhibit freedom of movement.

Specification Item Test Frequency 4.16.1 Reactor Internals Demonstrate Operability Each Refueling Vent Valves By: ShutdownIn accordance with the Surveillance Frequency Control Program

a. Conducting a remote visual inspection of visually accessible sur-faces of the valve body and disc sealing faces and evaluating any observed surface irregu-larities.

b. Verifying that the valve is not stuck in an open position, and

c. Verifying through manual actuation that the valve is fully open with a force of 400 lbs. (applied vertically upward).

Bases Verifying vent valve freedom of movement insures that coolant flow does not bypass the core through reactor internals vent valves during operation and therefore insures the conservatism of Core Protection Safety limits as delineated in Figures 2.1-1 and 2.1-3, and the flux/flow trip setpoint.

4-59 Amendment No. 65, 149

4.20 REACTOR BUILDING AIR TEMPERATURE Applicability This specification applies to the average air temperature of the primary containment during power operations.

Objective To assure that the temperatures used in the safety analysis of the reactor building are not exceeded.

Specification 4.20.1 When the reactor is critical, the reactor building temperature will be checked once each twenty-four (24) hours in accordance with the Surveillance Frequency Control Program. If any detector exceeds 130°F (120°F below elevation 320) the arithmetic average will be computed to assure compliance with Specification 3.17.1.

4-86 Amendment No. 41, 47

6.21 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Definition 1.25 and Surveillance Requirement 4.0.2 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

6-30 Amendment No.

ATTACHMENT 4 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)

TSTF-425 (NUREG-1430) vs. TMI Unit 1 Cross-Reference

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 1 of 10 TSTF-425 (NUREG-1430) vs. TMI Unit 1 Cross-Reference Technical Specification Section Title/Surveillance Description* TSTF-425 TMI Unit 1 Definitions (Staggered Test Basis) 1.1 -----

Shutdown Margin (SDM) 3.1.1 ------

Verify SDM is within limits 3.1.1.1 ---------------

Reactivity Balance 3.1.2 ------

Verify core reactivity balance within + 1% ~k/k 3.1.2.1 ---------------

Control Rod Group Alignment Limits 3.1.4 ------

Verify control rod positions 3.1.4.1 4.1.1/

Table 4.1-1/

Items 23 &

24 Verify control rod freedom of movement 3.1.4.2 4.1.21 Table 4.1-21 Item 2 Verify control rod drop times --------------- 4.1.21 Table 4.1-21 Item 1 Safety Rod Insertion Limits 3.1.5 ------

Verify each safety rod fully withdrawn 3.1.5.1 ---------------

Axial Power Shaping Rod (APSR) Alignment Limits 3.1.6 -------

Verify position of each APSR 3.1.6.1 ---------------

Position Indicator Channels 3.1.7 ----

Verify absolute and relative position indicator channels aQree 3.1.7.1 ---------------

Physics Test Exceptions - Mode 1 3.1.8 -------

Verify Thermal Power < 85% RTP 3.1.8.1 ---------------

Perform SR 3.2.5.1 3.1.8.2 ---------------

Verify nuclear overpower trip setpoint 3.1.8.3 ---------------

Verify SDM is within limits 3.1.8.4 ---------------

Physics Test Exceptions - Mode 1 3.1.9 -------

Verify Thermal Power < 5% RTP 3.1.9.1 ---------------

Verify nuclear overpower trip setpoint 3.1.9.2 ---------------

Verify SDM is within limits 3.1.9.3 ---------------

Regulating Rod Insertion Limits 3.2.1 -------

Verify reQulatinQ rod qroups within sequence and overlap limits 3.2.1.1 ---------------

Verify reQulatinQ rod qroups meet insertion limits 3.2.1.2 ---------------

APSR Insertion Limits 3.2.2 ------

Verify APSRs are within acceptable limits 3.2.2.1 ---------------

Axial Power Imbalance Operating Limit 3.2.3 3.5.2.7 Verify Axial Power Imbalance is within limits 3.2.3.1 3.5.2.7.f Quadrant Power Tilt (QPn 3.2.4 3.5.2.4 Verify OPT is within limits 3.2.4.1 3.5.2.4.q Reactor Protection System (RPS) Instrumentation 3.3.1 4.1.1 Channel Check 3.3.1.1 Table 4.1-1 Calorimetric heat balance calc. to power range channel output 3.3.1.2 Table 4.1-1/

Item 3

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 2 of 10 Technical Specification Section Title/Surveillance Description

• TSTF-425 TMI Unit 1 Out of core to incore measured Axial Power Imbalance 3.3.1.3 Table 4.1-1/

Item 4 Channel Functional Test 3.3.1.4 Table 4.1-1 Channel Calibration 3.3.1.5 Table 4.1-1 Verify RPS Response Time is within limits 3.3.1.6 ---------------

RPS - Reactor Trip Module (RTM) 3.3.3 4.1.1 Channel Functional Test 3.3.3.1 Table 4.1-1/

Item 1 Control Rod Drive (CRD) Trip Devices 3.3.4 4.1.1 Channel Functional Test 3.3.4.1 Table 4.1-1/

Item 2 Engineered Safety Feature Actuation System (ESFAS) 3.3.5 4.1.1 Instrumentation Channel Check 3.3.5.1 Table 4.1-1 Channel Functional Test 3.3.5.2 Table 4.1-1 Channel Calibration 3.3.5.3 Table 4.1-1 Verify ESFAS Response Time is within limits 3.3.5.4 ----------------

ESFAS Manual Initiation 3.3.6 ------

Channel Functional Test 3.3.6.1 ---------------

ESFAS Automatic Actuation Logic 3.3.7 4.1.1 Channel Functional Test 3.3.7.1 Table 4.1-1/

Items 14, 16, 18 & 20 Emergency Diesel Generator (EDG) Loss of Power Start (LOPS) 3.3.8 ------

Channel Check 3.3.8.1 ---------------

Channel Functional Test 3.3.8.2 4.1.1/

Table 4.1-1/

Item 43 Channel Calibration 3.3.8.3 4.1.1/

Table 4.1-1/

Item 43 Source Range Neutron Flux 3.3.9 4.1.1 Channel Check 3.3.9.1 Table 4.1-1/

Item 6 Channel Calibration 3.3.9.2 Table 4.1-1/

Item 6 Intermediate Range Neutron Flux 3.3.10 4.1.1 Channel Check 3.3.10.1 Table 4.1-1/

Item 5 Channel Calibration 3.3.10.2 Table 4.1-1/

Item 5 Emergency Feedwater Initiation and Control (EFIC) System 3.3.11 4.1.1 Instrumentation Channel Check 3.3.11.1 Table 4.1-1/

Item 51 Channel Functional Test 3.3.11.2 Table 4.1-1/

Item 51

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 3 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 TMI Unit 1 Channel Calibration 3.3.11.3 Table 4.1-1/

Item 51 Verify EFIC Response Time is within limits 3.3.11.4 ---------------

EFIC Manual Initiation Channel Functional Test 3.3.12 3.3.12.1 EFIC Logic 3.3.13 4.1.1 Channel Functional Test 3.3.13.1 Table 4.1-1/

Item 50 EFIC - Emergency Feedwater (EFW) - Vector Valve Logic 3.3.14 -------

Channel Functional Test 3.3.14.1 ---------------

Reactor Building (RB) Purge Isolation - High Radiation 3.3.15 ------

Channel Check 3.3.15.1 ---------------

Channel Functional Test 3.3.15.2 ---------------

Channel Calibration 3.3.15.3 ---------------

Control Room Isolation - High Radiation 3.3.16 -----

Channel Check 3.3.16.1 ---------------

Channel Functional Test 3.3.16.2 ------------- .. -

Channel Calibration 3.3.16.3 ---------------

Post Accident Monitoring (PAM) Instrumentation 3.3.17 4.1.3 Channel Check 3.3.17.1 Table 4.1-4 Channel Calibration 3.3.17.2 Table 4.1-4 Channel Test --------------- Table 4.1-4 Reactor Vessel Water Level --------------- 4.1.1/

Table 4.1-1/

Item 54 Remote Shutdown System 3.3.18 4.1.4 Channel Check 3.3.18.1 4.1.4.a Verify control circuit/transfer switch capable of intended function 3.3.18.2 4.1.4.b Channel Calibration 3.3.18.3 4.1.4.c Additional TMI Instrumentation ----- 4.1.1 Core flooding tanks - pressure/level channels --------------- Table 4.1-1/

Items 25a &

25b Pressurizer level channels ... -------------- Table 4.1-1/

Item 26 Makeup tank instrument channels - level/pressure --------------- Table 4.1-1/

Items 27a &

27b Borated Water Storage Tank level indicator --------------- Table 4.1-1/

Item 30 Containment temperature --------------- Table 4.1-1/

Item 33 Incore neutron detectors --------------- Table 4.1-1/

Item 34 Emergency plant radiation instruments --------------- Table 4.1-1/

Item 35

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 4 of 10 Technical Specification Section TitlelSurveiliance Description* TSTF-425 TMI Unit 1 OTSG full range level -----_ ... _------- Table 4.1-1/

Item 38 Turbine overspeed trip --------------- Table 4.1-1/

Item 39 Diesel generator protective relaying --------------- Table 4.1-1/

Item 42 Reactor coolant pressure DH valve interlock bistable --------------- Table 4.1-1/

Item 44 Pressurizer code safety valves and PORV tailpipe flow monitors --------------- Table 4.1-1/

Item 47a PORV - acoustic flow --------------- Table 4.1-1/

Item 47b PORV setpoints --------------- Table 4.1-1/

Item 48 Saturation margin monitor --------------- Table 4.1-1/

Item 49 Backup incore thermocouple display _... ------------- Table 4.1-1/

Item 52 Reactor Coolant System (ReS) Pressure, Temperature, and 3.4.1 ..

Flow Departure from Nucleate Boiling (DNB) Limits Verify RCS loop pressure (with four or three RCPs operating) 3.4.1.1 ---------------

Verify RCS hot leg temperature 3.4.1.2 ---------------

Verify RCS total flow (with four or three RCPs operating) 3.4.1.3 ---------------

Verify RCS total flow rate within limits 3.4.1.4 ---------------

RCS Minimum Temperature for Criticalitv 3.4.2 --- ... -

Verify RCS Tavg in each loop 3.4.2.1 ---------------

RCS Pressure and Temperature (pm Limits 3.4.3 .

Verify RCS pressure, temperature, heatup and cooldown rates 3.4.3.1 ... --------------

within limits RCS Loops* MODES 1 and 2 3.4.4 ._--

Verify required RCS loops in operation 3.4.4.1 ---------------

RCS Loops* MODE 3 3.4.5 Verify one RCS loop in operation 3.4.5.1 ---------------

Verify correct breaker alignment and indicated power available 3.4.5.2 ---------------

RCS Loops* MODE 4 3.4.6 . _.

Verify required Decay Heat Removal (DHR) or ReS loop in 3.4.6.1 ---------------

operation Verify correct breaker alignment and indicated power available 3.4.6.2 RCS Loops* MODE 5, Loops Filled 3.4.7 Verr v required DHR loop in operation 3.4.7.1 ---------------

Veri v steam aenerator (SG) secondary side water levels 3.4.7.2 ---------------

Ver~ v correct breaker alignment and indicated power available 3.4.7.3 ---------------

RCS Loops* MODE 5, Loops Not Filled 3.4.8 ---

Verify required DHR loop in operation 3.4.8.1 ---------------

Verify correct breaker alianment and indicated power available 3.4.8.2 ---------------

Reactor Internals Vent Valves -------------- 4.16.1 Pressurizer 3.4.9 .

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 5 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 TMI Unit 1 Verify pressurizer water level 3.4.9.1 ---------------

Verify pressurizer heaters are capable of being powered from an 3.4.9.2 4.6.3.a emergency power supply Verify emergency power supply for pressurizer heaters is operable 3.4.9.3 ---------------

Pressurizer Power Operated Relief Valve (PORV) 3.4.11 ------

Perform one complete cycle of block valve 3.4.11.1 4.1.21 Table 4.1-21 Item 11 Perform one complete cycle of the PORV 3.4.11.2 ---------------

Verify PORV and block valve capable of being powered from an 3.4.11.3 --_ -----------

emergency power source Low Temperature Overpressure Protection (LTOP) System 3.4.12 ------

Verify max. of [one] makeup pump capable of injecting into RCS. 3.4.12.1 ---------------

Verify High Pressure Injection (HPI) is deactivated 3.4.12.2 ---------------

Verify each Core Flood Tank (CFT) is isolated 3.4.12.3 ---------------

Verify pressurizer level 3.4.12.4 ---------------

Verify PORV block valve is open 3.4.12.5 ---------------

Verify required RCS vent is open 3.4.12.6 ---------------

Perform Channel Functional Test for PORV 3.4.12.7 ---------------

Perform Channel Calibration for PORV 3.4.12.8 ---------------

RCS Operational Leakage 3.4.13 4.1.2 Verify RCS operational leakage is within limits 3.4.13.1 Table 4.1-21 Item 7 Verify primary to secondary leakage through anyone SG 3.4.13.2 Table 4.1-21 Item 12 Engineered Safeguards Feature (ESF) System Leakage -------------- 4.5.4.2 RCS Pressure Isolation Valve (PIV) Leakage 3.4.14 -----

Verify leakage from each RCS PIV 3.4.14.1 ---------------

Verify DHR System auto closure interlock prevents the valves from 3.4.14.2 ---------------

being opened on RCS pressure signal Verify DHR System auto closure interlock causes the valves to 3.4.14.3 ---------------

close automatically on RCS pressure signal RCS Leakage Detection Instrumentation 3.4.15 4.1.1 Channel Check (cont.atmosphere rad monitor) 3.4.15.1 Table 4.1-1/

Item 28 Channel Functional Test (cont.atmosphere rad monitor) 3.4.15.2 Table 4.1-1/

Item 28 Channel Calibration (containment sump monitor) 3.4.15.3 Table 4.1-1/

Item 37 Channel Calibration (cont.atmosphere rad monitor) 3.4.15.4 Table 4.1-1/

Item 28 RCS Specific Activity 3.4.16 4.1.2 Verify reactor coolant gross specific activity 3.4.16.1 Table 4.1-3/

Item 1a Verify reactor coolant Dose Equivalent 1-131 specific activity 3.4.16.2 Table 4.1-3/

Item 1b Determine E bar 3.4.16.3 ---------------

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 6 of 10 Technical Specification Section Title/Surveillance Descriptlon* TSTF-425 TMI Unit 1 Chemistry (CI, F and 02) --------------- Table 4.1-3/

Item 1d Boron concentration --------------- Table 4.1-3/

Item 1e Tritium radioactivity --------------- Table 4.1-3/

Item 1f Core Flood Tanks (CFTs) 3.5.1 4.1.2 Verify each CFT isolation valve is fully open 3.5.1.1 Table 4.1-5/

Item 1.c Verify borated water volume in each CFT 3.5.1.2 Table 4.1-5/

Item 1.a Verify nitrogen cover pressure in each CFT 3.5.1.3 Table 4.1-5/

Item 1.b Verify boron concentration in each CFT 3.5.1.4 Table 4.1-3/

Item 3 Verify power is removed from each CFT isolation valve operator 3.5.1.5 Table 4.1-5/

Item 1.d Emergency Core Cooling Systems (ECCS) . Operating 3.5.2 -.

Verify valves are in the listed position with power removed 3.5.2.1 ---------------

Verify each valve in flow path is in correct position 3.5.2.2 ---------------

Verify ECCS pipinq is full of water. 3.5.2.3 ---------------

Verify each ECCS pump's developed head 3.5.2.4 ---------------

Verify each ECCS automatic valve actuates to the correct position 3.5.2.5 4.5.2.1.a/

4.5.2.2.a Verify each ECCS pump starts automatically 3.5.2.6 4.5.2.1.a/

4.5.2.2.a Verify the correct settinqs of stops for HPI stop check valves: 3.5.2.7 ---------------

Verify the flow controllers for low pressure injection (LPI) throttle 3.5.2.8 ---------------

valves operate properly Verify, by visual inspection, each ECCS train containment sump 3.5.2.9 ---------------

suction inlet is not restricted Core Flooding system operability --------------- 4.5.2.3 Emerqency core coolinq component tests --------------- 4.5.2.4 Borated Water Storage Tank (BWSn 3.5.4 ------

Verify BWST borated water temperature 3.5.4.1 ---------------

Verify BWST borated water volume 3.5.4.2 ---------------

Verify BWST boron concentration 3.5.4.3 4.1.2/

Table 4.1-3/

Item 2 Containment Air Locks 3.6.2 ..-----

Verify only one door in air lock can be opened 3.6.2.2 4.4.1.3 Containment Isolation Valves 3.6.3 ------

Verify each [481 inch purqe valve is sealed closed 3.6.3.1 ---------------

Verify each [8] inch purge valve is closed 3.6.3.2 ---------------

Verify each containment isolation manual valve and blind flange 3.6.3.3 ---------------

that is located outside containment is closed

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 7 of 10 Technical Specification Section Title/Surveillance Description

• TSTF-425 TMI Unit 1 Verify the isolation time of each automatic power operated 3.6.3.5 -.-------------

containment isolation valve is within limits Perform leakage rate testing for containment purge valves with 3.6.3.6 ---------------

resilient seals Verify each automatic containment isolation valve actuates to the 3.6.3.7 ---------------

isolation position Verify each [ ] inch containment purge valve is blocked to restrict 3.6.3.8 ----.----------

the valve from opening Containment Pressure 3.6.4 -----

Verify containment pressure is within limits 3.6.4.1 ---------------

Containment Air Temperature 3.6.5 4.20 Verify containment average air temperature is within limit 3.6.5.1 4.20.1 Containment Spray and Cooling Systems 3.6.6 ------

Verify each containment spray valve is in the correct position 3.6.6.1 ---------------

Operate each containment cooling train fan unit 3.6.6.2 ---------------

Verify each containment coolina train coolina water flow rate 3.6.6.3 4.5.3.1.b.1 Verify each automatic containment spray valve actuates to the 3.6.6.5 4.5.3.1.a.1 correct position Verify each containment spray pump starts automatically 3.6.6.6 4.5.3.1 .a.1 Verify each containment cooling train starts automatically 3.6.6.7 4.5.3.1.b.1/

4.5.3.2.a Verify each spray nozzle is unobstructed 3.6.6.8 4.5.3.1.a.2 Spray Additive System 3.6.7 ------

Verify each spray additive valve is in the correct position 3.6.7.1 ---------------

Verify spray additive tank solution volume 3.6.7.2 ---------------

Verify spray additive tank fNaOHl solution concentration 3.6.7.3 ---------------

Verify each spray additive automatic valve actuates to the correct 3.6.7.4 ---------------

position Verify Spray Additive System flow from each solution's flow path 3.6.7.5 ---------------

Reactor Building Emergency Sump pH Control System ----- 4.1.2 Verify TSP baskets contain Ibs of TSP --_ .. _--------- Table 4.1-5/

Item 2a Verify TSP basket sample provides adequate pH adjustment -------------- Table 4.1-5/

Item 2b Main Steam Isolation Valves (MSIVs) 3.7.2 -------

Verify each MSIV actuates to the isolation position 3.7.2.2 ---------------

Main Feedwater Stop Valves (MFSVs), Main Feedwater Control 3.7.3 ------

Valves (MFCVs), and Associated Startup Feedwater Control Valves (SFCVs)

Verify each [MFSVJ, [MFCV], and [SFCV] actuates to the isolation 3.7.3.2 ---------------

position Atmospheric Vent Valves (AVVs) 3.7.4 ----

Verify one complete cycle of each AVV 3.7.4.1 ---------------

Verify one complete cycle of each AVV block valve 3.7.4.2 ---------------

Emergency Feedwater (EFW) System [DHR Capabilityl Verify each EFW valve is in the correct position 3.7.5 3.7.5.1 4.9.1.3 Verify each EFW automatic valve actuates to the correct position 3.7.5.3 4.9.1.4

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 8 of 10 Technical Specification Section Title/Surveillance Description

• TSTF-425 TMI Unit 1 Verify each EFW pump starts automatically 3.7.5.4 4.9.1.4 Perform a Channel Functional Test for the EFW pump suction 3.7.5.6 ---------------

pressure interlocks Perform a Channel Calibration for the EFW pump suction pressure 3.7.5.7 ---------------

interlocks Verify operability of means for DHR -------------- 4.9.2.1 Reactor Coolant system Vents - 4.11 Reactor coolant system vent path operable -------------- 4.11.1 Condensate Storage Tank (CST) 3.7.6 --------

Verify CST level 3.7.6.1 ---------------

Component Cooling Water (CCW) System 3.7.7 Verify each CCW valve is in the correct position 3.7.7.1 ---------------

Verify each CCW automatic valve actuates to the correct position 3.7.7.2 ---------------

Verify each CCW pump starts automatically 3.7.7.3 ---------------

Service Water System (SWS) 3.7.8 ------

Verify each SWS valve is in the correct position 3.7.8.1 ---------------

Verify each SWS automatic valve actuates to the correct position 3.7.8.2 ---------------

Verify each SWS pump starts automatically 3.7.8.3 ---------------

Ultimate Heat Sink (UHS) 3.7.9 ------

Verify water level of UHS 3.7.9.1 ---------------

Verify averaQe water temperature of UHS 3.7.9.2 ---------------

Operate each cooling tower fan 3.7.9.3 ---------------

Intake Pump House Floor -- ---- 4.1.2 Silt accumulation - visual inspection --------------- Table 4.1-21 Item 10(a)

Silt accumulation measurement of pump house flow --------------- Table 4.1-21 Item 10(b)

Control Room Emergency Ventilation System (CREVS) 3.7.10 4.12.1 Operate each CREVS train 3.7.10.1 4.12.1.2.d CREVS filter testing [NUREG-1430 - Ventilation filter Testing .------------- 4.12.1.2.a ProQram]

Verify [each CREVS train actuates] [or the control room isolates] 3.7.10.3 4.12.1.3 Verify one CREVS train can maintain a positive pressure relative 3.7.10.4 ---------------

to the adjacent [area] during the [pressurization] mode of operation Verify the system makeup flow rate when supplying the control 3.7.10.5 ---------------

room with outside air Pressure drop across HEPA filter and charcoal adsorber -------------- 4.12.1.1 Control Room Emergency Air Temperature Control System 3.7.11 -----

(CREATCS)

Verify each CREATCS train has the capability to remove the 3.7.11.1 ---------------

assumed heat load Emergency Ventilation System (EVS) 3.7.12 -------

Operate each EVS train 3.7.12.1 ---------------

Verify each EVS train actuates 3.7.12.3 ---------------

Verify one EVS train can maintain pressure relative to atmospheric 3.7.12.4 ---------------

pressure durinQ the [post accident] mode of operation Verify each EVS filter coolinQ bypass damper can be opened 3.7.12.5 ---------------

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 9 of 10 Technical Specification Section Title/Surveillance Description

• TSTF-425 TMI Unit 1 Fuel Storage Pool Ventilation System (FSPVS) 3.7.13 4.12.4 Operate each FSPVS train 3.7.13.1 4.12.4.3 Verify each FSPVS train actuates 3.7.13.3 ---------------

Verify one FSPVS train can maintain pressure with respect to 3.7.13.4 ---------------

atmospheric pressure during the rpost accidentl mode of operation Verify each FSPVS filter bypass damper can be opened 3.7.13.5 ---------------

Fuel Storage Pool Water Level 3.7.14 ---

Verify the fuel storage pool water level 3.7.14.1 ---------------

Spent Fuel Pool Boron Concentration 3.7.15 4.1.2 Verify the spent fuel pool boron concentration is within limit 3.7.15.1 Table 4.1-3/

Item 4 Secondary Specific Activity 3.7.17 4.1.2 Verify the specific activity of the secondary coolant Dose 3.7.17.1 Table 4.1-3/

Equivalent 1-131. Item 5 Steam Generator Level 3.7.18 ----

Verify steam generator water level to be within limits 3.7.18.1 ---------------

AC Sources - Operating 3.8.1 ---_..- -----

Verify correct breaker alignment and power availability 3.8.1.1 ---------------

Verify each DG starts from standby conditions/steady state 3.8.1.2 ---------------

Verify each DG is synchronized and loaded 3.8.1.3 4.6.1.a Verify each day tank level 3.8.1.4 ---------------

Check for and remove accumulated water from day tank 3.8.1.5 ---------------

Verify fuel oil transfer system operates 3.8.1.6 ---------------

Verify each DG starts from standby conditions/quick start 3.8.1.7 ---------------

Verify transfer of power from offsite circuit to alternate circuit 3.8.1.8 ---------------

Verify DG rejects load greater than single largest load 3.8.1.9 ---------------

Verify DG maintains load following load reject 3.8.1.10 ---------------

Verify on loss of offsite power signal 3.8.1.11 ---------------

Verify DG starts on Engineered Safety Feature actuation signal 3.8.1.12 ---------------

Verify DG automatic trips bypassed on ESF actuation signal 3.8.1.13 ---------------

Verify each DG operates for> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.8.1.14 ---------------

Verify each DG starts from standby conditions/quick restart 3.8.1.15 ---------------

Verify each DG synchronizes with offsite power 3.8.1.16 ---------------

Verify ESF actuation signal overrides test mode 3.8.1.17 ---------------

Verify interval between each sequenced load block 3.8.1.18 ---------------

Verify on LOOP in conjunction with ESF actuation signal 3.8.1.19 4.6.1.b/

4.5.1.1.a Verify simultaneous DG starts 3.8.1.20 ---------------

Emergency loading sequence test -------------- 4.5.1.2.a Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 -------

Verify fuel oil storage tank volume 3.8.3.1 ---------------

Verify lube oil inventory 3.8.3.2 ---------------

Verify each DG air start receiver pressure 3.8.3.4 ---------------

Check/remove accumulated water from fuel oil storage tank 3.8.3.5 ---------------

DC Sources - Operating/Battery Parameters 3.8.413.8.6 -----

Verify battery terminal voltage 3.8.4.1 ---------------

Verify each battery charger supplies amperage 3.8.4.2 ---------------

LAR - Adoption of TSTF-425, Revision 3 Attachment 4 Docket No. 50-289 Page 10 of 10 Technical Specification Section Title/Surveillance Description

• TSTF-425 TMI Unit 1 Verify battery capacity during battery service test 3.8.4.3 4.6.2.d Verify batterv capacity durina performance discharae test 3.8.6.6 ---------------

Verify batterv float current 3.8.6.1 ---------------

Verify batterv pilot cell voltaae 3.8.6.2 4.6.2.b Verify battery pilot cell specific gravity --------------- 4.6.2.b Verify batterv connected cell electrolvte level 3.8.6.3 4.6.2.a(1)

Verify batterv pilot cell temperature 3.8.6.4 ---------------

Verify battery connected cell voltaae 3.8.6.5 4.6.2.a(1)

Verify battery connected cell specific aravitv --------------- 4.6.2.a(1)

Inverters - Operatina 3.8.7 Verify correct inverter voltaae, freauency and alianment 3.8.7.1 ---------------

Inverters - Shutdown 3.8.8 ------

Verify correct inverter voltaae, freauencv and alianment 3.8.8.1 ---------------

Distribution Systems - Operatina 3.8.9 -- -

Verify correct breaker alianmentlvoltaae to distribution subsystems 3.8.9.1 ---------------

Distribution Systems - Shutdown 3.8.10 -----

Verify correct breaker alianmentlvoltaae to distribution subsvstems 3.8.10.1 ---------------

Boron Concentration 3.9.1 -- -

Verify boron concentration is within the limit specified in the COLR 3.9.1.1 ---------------

Nuclear Instrumentation 3.9.2 ------

Channel Check 3.9.2.1 ---------------

Channel Calibration 3.9.2.2 ---------------

Containment Penetrations 3.9.3 -----

Verify each reauired containment penetration is in reauired status 3.9.3.1 ---------------

Verify each required containment purge and exhaust valve 3.9.3.2 ---------------

actuates to the isolation position DHR and Coolant Circulation - Hiah Water Level 3.9.4 -----

Verify one DHR loop is in operation 3.9.4.1 ---------------

DHR and Coolant Circulation - Low Water Level 3.9.5 -------

Verify one DHR loop is in operation 3.9.5.1 ---------------

Verify correct breaker alianment and indicated power available 3.9.5.2 ---------------

Refueling Canal Water Level 3.9.6 -----

Verify refuelina canal water level 3.9.6.1 ---------------

Flood/Periodic Inspection of the Dikes ----- 3.14.1 Dike inspection -------------- 3.14.1.1 Programs (Surveillance Frequency Control Program [SFCPl) 5.5.18 6.21

• The Technical Specification Section Title/Surveillance Description portion of this attachment is a summary description of the referenced TSTF-425 (NUREG-1430)/TMI Unit 1 TS Surveillances which is provided for information purposes only and is not intended to be a verbatim description of the TS Surveillances.

ATTACHMENT 5 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)

Proposed No Significant Hazards Consideration

LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-289 Page 1 of 2 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION Description of Amendment Request: This amendment request involves the adoption of approved changes to the standard technical specifications (STS) for Babcock and Wilcox (B&W) plants (NUREG-1430), to allow relocation of specific TS surveillance frequencies to a Iicensee-controlled program. The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML090850642) related to the Relocation of Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b and are described in the Notice of Availability published in the Federal Register on July 6,2009 (74 FR 31996).

The proposed changes are consistent with NRC-approved Industry/ TSTF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control- RITSTF Initiative 5b."

The proposed changes relocate surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP). The changes are applicable to licensees using probabilistic risk guidelines contained in NRC-approved NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. 071360456).

Basis for proposed no significant hazards consideration: As required by 10 CFR 50.91 (a),

the Exelon analysis of the issue of no significant hazards consideration is presented below:

1. Do the proposed changes involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No.

The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program.

Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.

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

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

Response: No.

No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (Le., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the

LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-289 Page 2 of 2 changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.

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

3. Do the proposed changes involve a significant reduction in the margin of safety?

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Exelon will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP.

NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based upon the above, Exelon concludes that the requested changes do not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), "Issuance of Amendment."