ML17290A005

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Draft Revised Model Safety Evaluation of TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - RITSTF Initiative 4B
ML17290A005
Person / Time
Site: Technical Specifications Task Force
Issue date: 05/01/2018
From: Victor Cusumano
NRC/NRR/DSS/STSB
To:
Technical Specifications Task Force
Honcharik M
Shared Package
ML17290A003 List:
References
TSTF-505, Rev 2
Download: ML17290A005 (32)


Text

May 1, 2018 Technical Specifications Task Force 11921 Rockville Pike, Suite 100 Rockville, MD 20852

SUBJECT:

DRAFT REVSIED MODEL SAFETY EVALUATION FOR TRAVELER TSTF-505, PROVIDE RISK-INFORMED EXTENDED COMPLETION TIMES - RITSTF INITIATIVE 4B

Dear Members of the Technical Specifications Task Force:

The availability of Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 1, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, and a model safety evaluation (SE), were announced in the Federal Register (77 FR 15399) on March 15, 2012.

The U.S. Nuclear Regulatory Commission (NRC) staff identified areas requiring further review related to TSTF-505, Revision 1, during its review of plant-specific license amendment requests to adopt a risk-informed completion time (RICT) program. The NRC staff notified the TSTF of its concerns in a letter dated November 15, 2016, and suspended its approval of Revision 1 at that time (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16281A021). The TSTF submitted a response to the NRC staffs identified issues in a letter dated September 27, 2017 (ADAMS Package Accession No. ML17290B229).

The NRC Staff has prepared a draft revised traveler denoted as TSTF-505, Revision 2 (Enclosure 1). Attached to the revised traveler are drafts of a table of revised retained TS actions (Table 1) and a table listing the TS actions that require additional justification if an applicant elects to include them in the scope of its RICT Program (Table 2). The NRC staff has enclosed a draft revised model SE (Enclosure 2), which supersedes the model SE from 2012 and a revised model application (Enclosure 3).

Sixty calendar days are provided to you to comment on any factual errors or clarity concerns contained in the enclosed documents. The final versions of the documents will be issued after making any necessary changes. To facilitate the NRC staff's review of your comments, please provide a marked-up copy of the documents showing proposed changes and provide a summary table of the proposed changes.

If you have any questions, please contact Michelle Honcharik at 301-415-1774 or via e-mail at Michelle.Honcharik@nrc.gov.

Sincerely,

/RA/

Victor G. Cusumano, Chief Technical Specifications Branch Division of Safety Systems Office of Nuclear Reactor Regulation Project No. 753

Enclosures:

1. Draft Traveler TSTF-505, Revision 2
2. Draft Revised Model SE
3. Revised Model Application cc: See next page

Package: ML17290A003, Draft TSTF-505, Revision 2 (Enclosure 1): ML17290A082, Draft Revised Model SE (Enclosure 2): ML17290A005, Draft TSTF-505, Table 1, Revised Retained TS Actions (Attachment 1 to Encl. 1): ML17290A097, Draft TSTF-505, Table 2, TS Action Requiring Plant-Specific Justification (Attachment 2 to Encl. 1):

ML17339A168 Revised Model Application: ML18115A482

  • concurred via e-mail NRR-106 OFFICE DORL/LSPB/LA*

DE/EICB/BC DE/EEOB/BC DSS/SRXB/BC DORL/LPL2-1/BC NAME JBurkhardt MWaters JQuichocho JWhitman MMarkley DATE 11/17/2017 11/21/2017 12/4/2017 12/4/17 11/27/2017 OFFICE DRA/APLA/BC DSS/SCPB/BC*

OGC DSS/STSB/PM DSS/STSB/BC NAME SRosenberg RDennig DRoth MHoncharik VCusumano DATE 1/26/2018 11/17/2017 4/24/2018 04/30/2018 05/01/2018

Technical Specifications Task Force Project No. 753 cc:

Technical Specifications Task Force c/o EXCEL Services Corporation 11921 Rockville Pike, Suite 100 Rockville, MD 20852 Attention: Brian D. Mann E-mail: brian.mann@excelservices.com James P. Miksa Entergy Nuclear Operations, Inc.

Palisades Nuclear Power Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Email: jmiksa@entergy.com Jordan L. Vaughan Duke Energy EC2ZF / P.O. Box 1006 Charlotte, NC 28202 Email: jordan.vaughan@duke-energy.com Lisa L. Williams Energy Northwest Columbia Generating Station PO Box 968 Mail Drop PE20 Richland, WA 99352-0968 E-mail: llwilliams@energy-northwest.com David M. Gullott Exelon Generation 4300 Winfield Road Warrenville IL 60555 Email: David.Gullott@exeloncorp.com Wesley Sparkman Southern Nuclear Operating Company 42 Inverness Center Parkway / Bin B237 Birmingham, AL 35242 Email: wasparkm@southernco.com

General Directions: This Model SE provides the format and content to be used when preparing 1

the plant-specific SE of an LAR to adopt TSTF-505. The bolded bracketed information shows 2

text that should be filled in for the specific amendment; individual licensees would furnish 3

plant-specific nomenclature or values for these bracketed items. The italicized wording provides 4

guidance on what should be included in each section and should not be included in the SE.

5 6

DRAFT REVISED MODEL SAFETY EVALUATION 7

BY THE OFFICE OF NUCLEAR REACTOR REGULATION 8

OF TSTF-505, 9

PROVIDE RISK-INFORMED EXTENDED COMPLETION TIMES - RITSTF INITIATIVE 4B 10 11

1.0 INTRODUCTION

12 13 By application dated [enter date], (Agencywide Documents Access and Management System 14 (ADAMS) Accession No. [MLXXXXXXXXX]), [name of licensee] (the licensee) proposed 15 changes to the technical specifications (TSs) for the [name of facility and applicable units 16 (abbreviated name)]. Specifically, the licensee requested changes to the TS to adopt Traveler 17 TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.

18

[Variations from TSTF-505, are described in Section 2.2.4 of this safety evaluation (SE).]

19 20 The licensee requested the proposed changes to the TSs in accordance with Section 50.90 of 21 Title 10 of the Code of Federal Regulations (10 CFR). [The supplemental letters dated [enter 22 date(s)], provided additional information that clarified the application, did not expand the 23 scope of the application as originally noticed, and did not change the U.S. Nuclear 24 Regulatory Commission (NRC) staffs original proposed no significant hazards 25 consideration determination as published in the Federal Register on [enter date] (cite FR 26 reference).]

27 28 The proposed amendment(s) would modify TS requirements to permit the use of risk-informed 29 completion times (RICTs) for actions to be taken when limiting conditions for operation (LCOs) 30 are not met. The methodology is based on the Nuclear Energy Institute (NEI) Topical 31 Report 06-09, Revision 0-A (hereafter referred to as NEI 06-09-A), Risk-Informed Technical 32 Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, 33 (ADAMS Accession No. ML12286A322 (part of ADAMS Package Accession 34 No. ML122860402)). The NEI developed the guidance in NEI 06-09-A as a methodology to 35 evaluate and extend TS LCO Required Action Completion Times (CTs). The NRC staffs SE 36 dated May 17, 2007 (ADAMS Accession No. ML071200238), found the guidance in NEI 06-09, 37 Revision 0, to be acceptable, with clarification from the NRC staff positions, limitations, and 38 conditions in the SE dated May 17, 2007. In its letter dated October 12, 2012 (ML12286A321),

39 which provided the NRC with NEI 06-09, Rev. 0-A, the NEI stated that [t]his version 40 incorporates NRCs final safety evaluation, dated May 17, 2007, and is designated as the []A[]

41 version. The 93-page submittal (ADAMS Accession No. ML12286A322) included the entirety 42 of the NRCs final SE (pages 6/93 to 32/93 of ADAMS Accession No. ML12286A322), and 43 NEI 06-09, Revision 0, dated November 2006 (pages 33/93 to 93/93 of ADAMS Accession 44 No. ML12286A322). The October 12, 2012, NEI letter did not include a marked-up version of 45 NEI 06-09, Revision 0, that could have shown how NEI 06-09, Revision. 0 needed to be 46 changed to fit within the scope of the NRCs approval. This was in keeping with the topical 1

report process at the time.

2 3

Traveler TSTF-505 was developed to provide a generic model for implementing the TS changes 4

supported by the methodology in NEI 06-09-A. The availability of TSTF-505, Revision 1, was 5

announced in the Federal Register (77 FR 15399) on March 15, 2012. The NRC staff identified 6

concerns with TSTF-505, Revision 1, during its review of plant-specific license amendment 7

requests (LARs) requesting adoption of a RICT program. The NRC staff determined that the 8

precautions and limitations on the use of NEI 06-09-A were not appropriately reflected in 9

Traveler TSTF-505. The NRC staff notified the TSTF of its concerns in a letter dated 10 November 15, 2016, and suspended its approval of Revision 1 at that time (ADAMS Accession 11 No. ML16281A021). The TSTF responded via letter dated September 27, 2017 (ADAMS 12 Accession Package No. ML17290B229).

13 14 The NRC staff reviewed the changes described in the TSTF letter. The NRC staff developed 15 Traveler TSTF-505, Revision 2, incorporating resolution of the issues; a table of revised 16 retained TS actions (Table 1); a table of TS actions requiring plant-specific justification (Table 2) 17 and a revised model SE. The draft Traveler TSTF-505, Revision 2, draft Tables 1 and 2, the 18 draft revised model application, and the draft revised model SE are available in ADAMS at 19 Accession Nos. ML17290A082, ML17290A097, ML17339A168, ML18115A482 and 20 ML17290A005, respectively.

21 22

2.0 REGULATORY EVALUATION

23 24

2.1 DESCRIPTION

OF RISK-INFORMED COMPLETION TIME PROGRAM 25 26 The TS LCOs are the lowest functional capability or performance levels of equipment required 27 for safe operation of the facility. When an LCO is not met, the licensee must shut down the 28 reactor or follow any remedial or required action (e.g., testing, maintenance, or repair activity) 29 permitted by the TSs until the condition can be met. The remedial actions (i.e., ACTIONS) 30 associated with an LCO contain Conditions that typically describe the ways in which the 31 requirements of the LCO can fail to be met. Specified with each stated Condition are Required 32 Action(s) and CTs. The CTs are referred to as the front stops in the context of this SE. For 33 certain Conditions, the TS require exiting the Mode of Applicability of an LCO.

34 35

{NOTE: This paragraph may be used for facilities that have not converted to STS.}

36

[The licensees TS are not presented in the STS format. The term Action Statement is 37 conventionally used to describe ways in which the requirements of the LCO can fail to be 38 met (i.e., Condition) and the necessary Required Actions. Throughout this SE, the terms 39 Condition and Required Actions are used to describe Action Statements. The term 40 Allowed Outage Time is conventionally used to describe the length of time that 41 equipment is permitted to be inoperable. For the purposes of this SE, the terms CT 42 and Allowed Outage Time are used interchangeably.]

43 44 The Topical Report NEI 06-09-A provides a methodology for extending existing CTs and 45 thereby delay exiting the operational mode of applicability or taking Required Actions if risk is 46 assessed and managed within the limits and programmatic requirements established by a RICT 47 Program.

48 49

2.2 DESCRIPTION

OF TS CHANGES 1

2 The licensees submittal requested approval to add a RICT Program to the Administrative 3

Controls section of the TS [, add new conditions and associated actions in some TSs], and 4

modify selected CTs to permit extending the CTs, provided risk is assessed and managed as 5

described in NEI 06-09-A. The licensees application for the changes proposed to use 6

NEI 06-09-A and included documentation regarding the technical adequacy of the probabilistic 7

risk assessment (PRA) models for the RICT Program, consistent with the guidance of 8

Regulatory Guide (RG) 1.200, Revision 2, An Approach for Determining the Technical 9

Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009 10 (ADAMS Accession No. ML090410014).

11 12 2.2.1 Use and Application Example 13 14 The licensee has proposed to add the following example to the TSs as Example 1.3-8:

15 16

{NOTE: This is quoted from the TSTF letter dated September 27, 2017 (ADAMS Package 17 Accession No. ML17290B229). Be sure it matches what the licensee submitted.}

18 19 ACTIONS 20 CONDITION REQUIRED ACTION COMPLETION TIME A. One subsystem inoperable.

A.1 Restore subsystem to OPERABLE status.

7 days OR In accordance with the Risk Informed Completion Time Program B. Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours 21 When a subsystem is declared inoperable, Condition A is entered.

22 The 7 day Completion Time may be applied as discussed in 23 Example 1.3-2. However, the licensee may elect to apply the Risk 24 Informed Completion Time Program which permits calculation of a 25 Risk Informed Completion Time (RICT) that may be used to 26 complete the Required Action beyond the 7 day Completion Time.

27 The RICT cannot exceed 30 days. After the 7 day Completion 28 Time has expired, the subsystem must be restored to OPERABLE 29 status within the RICT or Condition B must also be entered.

30 31 The Risk Informed Completion Time Program requires 32 recalculation of the RICT to reflect changing plant conditions. For 33 planned changes, the revised RICT must be determined prior to 34 implementation of the change in configuration. For emergent 1

conditions, the revised RICT must be determined within the time 2

limits of the Required Action Completion Time (i.e., not the RICT) 3 or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

4 5

If the 7 day Completion Time clock of Condition A has expired and 6

subsequent changes in plant condition result in exiting the 7

applicability of the Risk Informed Completion Time Program 8

without restoring the inoperable subsystem to OPERABLE status, 9

Condition B is also entered and the Completion Time clocks for 10 Required Actions B.1 and B.2 start.

11 12 If the RICT expires or is recalculated to be less than the elapsed 13 time since the Condition was entered and the inoperable 14 subsystem has not been restored to OPERABLE status, 15 Condition B is also entered and the Completion Time clocks for 16 Required Actions B.1 and B.2 start. If the inoperable subsystems 17 are restored to OPERABLE status after Condition B is entered, 18 Conditions A and B are exited, and therefore, the required actions 19 of Condition B may be terminated.

20 21 2.2.2 Technical Specification [5.5.15/5.5.18] Risk-Informed Completion Time Program 22 23 Technical Specification [5.5.15/5.5.18], which describes the RICT Program, would be added to 24 the TS and reads as follows:

25 26

{NOTE: With the exception of items b. and e. below, this is quoted from the TSTF letter dated 27 September 27, 2017 (ADAMS Package Accession No. ML17290B229), Attachment 1 (ADAMS 28 Accession No. ML17290B238), pages 14 and 15. Be sure it matches what the licensee 29 submitted. The wording in item b. was revised to reflect the modes of operation for BWRs. The 30 wording in item e. below is acceptable to provide the appropriate administrative controls, which 31 differs from the TSTF letter.}

32 33 Risk Informed Completion Time Program 34 35 This program provides controls to calculate a Risk Informed 36 Completion Time (RICT) and must be implemented in accordance 37 with NEI 06-09-A, Revision 0, Risk-Managed Technical 38 Specifications (RMTS) Guidelines. The program shall include the 39 following:

40 41

a.

The RICT may not exceed 30 days; 42 43

{NOTE: The RICT is only applicable in MODES supported by the licensees PRA. Licensees 44 applying the RICT Program to MODES other than Modes 1 and 2 must demonstrate that they 45 have the capability to calculate a RICT in those MODES or that the risk indicated by their MODE 46 1 and 2 PRA model is bounding with respect to the lower MODE conditions.}

47 48

b.

A RICT may only be utilized in MODE 1, 2 [, and 3, and MODE 4 49 while relying on steam generators for heat removal][, and 50 MODE 3 while relying on the main condenser for heat 1

removal];

2 3

c.

When a RICT is being used, any change to the plant 4

configuration, as defined in NEI 06-09-A, Appendix A, must 5

be considered for the effect on the RICT.

6 7

1.

For planned changes, the revised RICT must be 8

determined prior to implementation of the change in 9

configuration.

10 11

2.

For emergent conditions, the revised RICT must be 12 determined within the time limits of the Required 13 Action Completion Time (i.e., not the RICT) or 14 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, 15 whichever is less.

16 17

3.

Revising the RICT is not required if the plant 18 configuration change would lower plant risk and 19 would result in a longer RICT.

20 21

d.

For emergent conditions, if the extent of condition 22 evaluation for inoperable structures, systems, or 23 components (SSCs) is not complete prior to exceeding the 24 Completion Time, the RICT shall account for the increased 25 possibility of common cause failure (CCF) by either:

26 27

1.

Numerically accounting for the increased possibility 28 of CCF in the RICT calculation; or 29 30

2.

Risk Management Actions (RMAs) not already 31 credited in the RICT calculation shall be 32 implemented that support redundant or diverse 33 SSCs that perform the function(s) of the inoperable 34 SSCs, and, if practicable, reduce the frequency of 35 initiating events that challenge the function(s) 36 performed by the inoperable SSCs.

37 38

e.

The risk assessment approaches and methods shall be 39 acceptable to the NRC. The plant PRA shall be based on 40 the as-built, as-operated, and maintained plant; and reflect 41 the operating experience at the plant, as specified in 42 Regulatory Guide 1.200, Revision 2. Methods to assess 43 the risk from extending the completion times must be PRA 44 methods used to support this license amendment, or other 45 methods approved by the NRC for generic use; and any 46 change in the PRA methods to assess risk that are outside 47 these approval boundaries require prior NRC approval.

48 49 2.2.3 Application of the RICT Program to Existing LCOs and Conditions 1

2 The typical CT is modified by the application of the RICT Program as shown in the following 3

example. The changed portion is indicated in italics.

4 5

ACTIONS 6

CONDITION REQUIRED ACTION COMPLETION TIME A. One subsystem inoperable.

A.1 Restore subsystem to OPERABLE status.

7 days OR In accordance with the Risk Informed Completion Time Program 7

8 Where necessary, conforming changes are made to CTs to make them accurate following use 9

of a RICT. For example, most TSs have requirements to close/isolate containment isolation 10 devices if one or more containment penetrations have inoperable devices. This is followed by a 11 requirement to periodically verify the penetration is isolated. By adding the flexibility to use a 12 RICT to determine a time to isolate the penetration, the periodic verifications must then be 13 based on the time following isolation.

14 15 Individual LCO Required Actions and CTs modified by the proposed change are identified 16 below.

17 18

{NOTE: TSTF-505, Revision 2, Table 1 (ADAMS Accession No. ML17290A097), contains a list 19 of the Required Actions and CTs from the STS that are included in TSTF-505. Insert a list of 20 the Required Actions and CTs associated with each LCO that are proposed to be included in 21 the RICT Program for the plant-specific submittal.

22 23 The suggested format is 24 25 LCO 3.X.X Title of LCO 3.X.X 26 Required Action X.1 (Describe Condition)}

27 28

[2.2.4 Variations from TSTF-505]

29 30

[2.2.4.1 Application of the RICT Program to Modified Conditions, Required 31 Actions, and Completion Times 32 33 The following Conditions are modified to permit the application of a RICT:]

34 35

{NOTE: These are Conditions that are applicable when one or more subsystems/channels are 36 inoperable and there is no TS loss of function. The CT of these specific ACTIONS are modified 37 to accommodate a RICT. Example:

38 39 LCO 3.x.x Title of LCO 3.x.x 40 41 The existing ACTIONS requirement states:

1 2

ACTIONS 3

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more

[channel/subsystem/

train] inoperable.

A.1 Restore [channel/

subsystem/train] to OPERABLE status.

[24 hours]

4 The revised ACTIONS requirement states:

5 6

ACTIONS 7

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more

[channel/subsystem/

train] inoperable.

A.1 Restore [channel/

subsystem/train] to OPERABLE status.

[24 hours]

OR


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

Not applicable when

[all/two/four/both] required

[channel/subsystem/train]

are inoperable.

In accordance with the Risk Informed Completion Time Program

}

8 9

[2.2.4.2 Application of the RICT to Additional ACTIONS Requirements]

10 11

{NOTE: TSTF-505, Revision 2, Table 2 (ADAMS Accession No. ML17339A168) lists the 12 Conditions that should be evaluated on a plant-specific basis to confirm that the Condition does 13 not represent a TS loss of function and to confirm that the Condition is appropriately modeled in 14 the facilitys PRA.

15 16 The suggested format is:

17 18 LCO 3.X.X Title of LCO 3.X.X 19 Required Action X.1 (Describe Condition)}

20 21

[2.2.4.3 Additional Variations from TSTF-505]

22 23

{NOTE: List any additional variations from TSTF-505 24 25 The suggested format is:

26 27 LCO 3.X.X Title of LCO 3.X.X 28 Required Action X.1 (Describe Condition)}

29 30 2.3 REGULATORY REVIEW 1

2 2.3.1 Applicable Regulations 3

4 Under 10 CFR 50.90, whenever a holder of a license wishes to amend the license, including 5

technical specifications in the license, an application for amendment must be filed, fully 6

describing the changes desired. Under 10 CFR 50.92(a), determinations on whether to grant an 7

applied-for license amendment are to be guided by the considerations that govern the issuance 8

of initial licenses or construction permits to the extent applicable and appropriate.

9 10 The regulation under 10 CFR 50.36(c)(2) requires that TSs contain LCOs, which are the lowest 11 functional capability or performance levels of equipment required for safe operation of the 12 facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor 13 or follow any remedial action permitted by the TSs until the LCO can be met. Typically, the TSs 14 require restoration of equipment in a timeframe commensurate with its safety significance, along 15 with other engineering considerations. In determining whether the proposed TSs remedial 16 actions should be granted, the Commission will apply the reasonable assurance standards of 17 10 CFR 50.40(a) and 50.57(a)(3). The regulation at 10 CFR 50.40(a) states that in determining 18 whether to grant the licensing request, the Commission will be guided by, among other things, 19 consideration about whether the processes to be performed, the operating procedures, the 20 facility and equipment, the use of the facility, and other technical specifications, or the 21 proposals, in regard to any of the foregoing collectively provide reasonable assurance that the 22 applicant will comply with the regulations in this chapter, including the regulations in part 20 of 23 this chapter, and that the health and safety of the public will not be endangered.

24 25 The regulation under 10 CFR 50.36(c)(5) states that administrative controls are the provisions 26 relating to organization and management, procedures, recordkeeping, review and audit, and 27 reporting necessary to assure operation of the facility in a safe manner.

28 29 The regulation under 10 CFR 50.55a(h), Protection and safety systems, states in part 30 Protection systems of nuclear power reactors of all types must 31 meet the requirements specified in this paragraph. Each combined 32 license for a utilization facility is subject to the following conditions.

33 34 (2) Protection systems. For nuclear power plants with construction 35 permits issued after January 1, 1971, but before May 13, 1999, 36 protection systems must meet the requirements in 37 IEEE Std 279-1968, "Proposed IEEE Criteria for Nuclear Power 38 Plant Protection Systems," or the requirements in 39 IEEE Std 279-1971, "Criteria for Protection Systems for Nuclear 40 Power Generating Stations," or the requirements in 41 IEEE Std 603-1991, "Criteria for Safety Systems for Nuclear 42 Power Generating Stations, and the correction sheet dated 43 January 30, 1995. For nuclear power plants with construction 44 permits issued before January 1, 1971, protection systems must 45 be consistent with their licensing basis or may meet the 46 requirements IEEE Std. 603-1991 and the correction sheet dated 47 January 30, 1995.

48 (3) Safety systems. Applications filed on or after May 13, 1999, for 1

construction permits and operating licenses under this part, and 2

for design approvals, design certifications, and combined licenses 3

under part 52 of this chapter, must meet the requirements for 4

safety systems in IEEE Std. 603-1991 and the correction sheet 5

dated January 30, 1995.

6 7

Both IEEE 279 and IEEE 603 stipulate aspects of diversity and defense-in-depth; for example, 8

both require the protection system to include means for manual initiation of each automatically 9

initiated protective action (i.e., an independent and diverse means of initiating the protective 10 action).

11 12 Section 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at 13 nuclear power plants (i.e., the Maintenance Rule), requires licensees to monitor the 14 performance or condition of SSCs against licensee-established goals in a manner sufficient to 15 provide reasonable assurance that these SSCs are capable of fulfilling their intended functions.

16 The regulation under 10 CFR 50.65(a)(4) requires the assessment and management of the 17 increase in risk that may result from a proposed maintenance activity.

18 19 The plants design criteria are set forth in the current licensing basis of the plant, as documented 20 in the updated Final Safety Analysis Report (FSAR). The plants design criteria define minimum 21 requirements that achieve aspects of the defense-in-depth philosophy; as a consequence, even 22 a compromise of the intent of those design criteria can directly result in a significant reduction in 23 the effectiveness of one or more of the layers of defense. When evaluating the effect of the 24 proposed application of risk-informed completion times, the NRC staff evaluated continued 25 adherence to the intent of the plants design criteria.

26 27 2.3.2 Commission Policy 28 29 The NRC provided details concerning the use of PRA in the Final Policy Statement: Use of 30 Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities, published in the 31 Federal Register (60 FR 42622; August 16, 1995). In this publication, the Commission wrote, in 32 part:

33 34 The Commission believes that an overall policy on the use of PRA 35 methods in nuclear regulatory activities should be established so 36 that the many potential applications of PRA can be implemented 37 in a consistent and predictable manner that would promote 38 regulatory stability and efficiency. In addition, the Commission 39 believes that the use of PRA technology in NRC regulatory 40 activities should be increased to the extent supported by the 41 state-of-the-art in PRA methods and data and in a manner that 42 complements the NRCs deterministic approach.

43 44 PRA addresses a broad spectrum of initiating events by assessing 45 the event frequency. Mitigating system reliability is then 46 assessed, including the potential for multiple and common cause 47 failures. The treatment therefore goes beyond the single failure 48 requirements in the deterministic approach. The probabilistic 49 approach to regulation is, therefore, considered an extension and 50 enhancement of traditional regulation by considering risk in a 1

more coherent and complete manner.

2 3

Therefore, the Commission believes that an overall policy on the 4

use of PRA in nuclear regulatory activities should be established 5

so that the many potential applications of PRA can be 6

implemented in a consistent and predictable manner that 7

promotes regulatory stability and efficiency. This policy statement 8

sets forth the Commissions intention to encourage the use of 9

PRA and to expand the scope of PRA applications in all nuclear 10 regulatory matters to the extent supported by the state-of-the-art 11 in terms of methods and data.

12 13 Therefore, the Commission adopts the following policy statement 14 regarding the expanded NRC use of PRA:

15 16 (1) The use of PRA technology should be increased in all 17 regulatory matters to the extent supported by the 18 state-of-the-art in PRA methods and data and in a manner that 19 complements the NRCs deterministic approach and supports 20 the NRCs traditional defense-in-depth philosophy.

21 22 (2) PRA and associated analyses (e.g., sensitivity studies, 23 uncertainty analyses, and importance measures) should be 24 used in regulatory matters, where practical within the bounds 25 of the state-of-the-art, to reduce unnecessary conservatism 26 associated with current regulatory requirements, regulatory 27 guides, license commitments, and staff practices. Where 28 appropriate, PRA should be used to support the proposal for 29 additional regulatory requirements in accordance with 30 10 CFR 50.109 (Backfit Rule). Appropriate procedures for 31 including PRA in the process for changing regulatory 32 requirements should be developed and followed. It is, of 33 course, understood that the intent of this policy is that existing 34 rules and regulations shall be complied with unless these rules 35 and regulations are revised.

36 37 (3) PRA evaluations in support of regulatory decisions should be 38 as realistic as practicable and appropriate supporting data 39 should be publicly available for review.

40 41 (4) The Commissions safety goals for nuclear power plants and 42 subsidiary numerical objectives are to be used with 43 appropriate consideration of uncertainties in making regulatory 44 judgments on the need for proposing and backfitting new 45 generic requirements on nuclear power plant licensees.

46 47 2.3.3 Regulatory Guidance 48 49 Revision 3 of RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed 50 Decisions on Plant-Specific Changes to the Licensing Basis, May 2011 (ADAMS Accession 51 No. ML100910006), describes an acceptable risk-informed approach for assessing the nature 1

and impact of proposed permanent licensing basis changes by considering engineering issues 2

and applying risk insights. This regulatory guide also provides risk acceptance guidelines for 3

evaluating the results of such evaluations.

4 5

Revision 1 of RG 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking:

6 Technical Specifications, May 2011 (ADAMS Accession No. ML100910008), describes an 7

acceptable risk-informed approach specifically for assessing proposed TS changes. This 8

regulatory guide identifies a three-tiered approach for a licensees evaluation of the risk 9

associated with a proposed TS CT change, as follows.

10 11 Tier 1 assesses the risk impact of the proposed change in accordance with acceptance 12 guidelines consistent with the Commissions Safety Goal Policy Statement, as 13 documented in RG 1.174 and RG 1.177. The first tier assesses the impact on plant risk 14 as expressed by on the change in core damage frequency (CDF) and change in large 15 early release frequency (LERF). It also evaluates plant risk while equipment covered 16 by the proposed CT is out-of-service, as represented by incremental conditional core 17 damage probability (ICCDP) and incremental conditional large early release probability 18 (ICLERP). The limits for ICCDP and ICLERP are consistent with the criteria for 19 incremental core damage probability (ICDP) and incremental large early release 20 probability (ILERP) from the Nuclear Management and Resources Council 21 (NUMARC) 93-01, Revision 4A, Industry Guideline for Monitoring the Effectiveness of 22 Maintenance at Nuclear Power Plants, April 2011 (ADAMS Accession 23 No. ML11116A198), guidance for managing the risk of on-line maintenance activities.

24 ICDP and ILERP are the limits on which licensee will base the RICT. This guidance was 25 endorsed by the NRC staff in RG 1.160, Revision 3, Monitoring the Effectiveness of 26 Maintenance at Nuclear Power Plants, May 2012 (ADAMS Accession 27 No. ML113610098), for compliance with the Maintenance Rule, 10 CFR 50.65(a)(4).

28 Tier 1 also addresses PRA quality, including the technical adequacy of the licensees 29 plant-specific PRA for the subject application. Cumulative risk of the proposed TS 30 change is considered with uncertainty/sensitivity analysis with respect to the 31 assumptions related to the proposed TS change.

32 33 Tier 2 identifies and evaluates any potential risk-significant plant equipment outage 34 configurations that could result if equipment, in addition to that associated with the 35 proposed license amendment, is removed from service simultaneously, or if other 36 risk-significant operational factors, such as concurrent system or equipment testing, are 37 also involved. The purpose of this evaluation is to ensure that there are appropriate 38 restrictions in place such that risk-significant plant equipment outage configurations will 39 not occur when equipment associated with the proposed CT is implemented.

40 41 Tier 3 addresses the licensees configuration risk management program (CRMP) to 42 ensure that adequate programs and procedures are in place for identifying 43 risk-significant plant configurations resulting from maintenance or other operational 44 activities and appropriate compensatory measures are taken to avoid risk-significant 45 configurations that may not have been considered when the Tier 2 evaluation was 46 performed. Compared with Tier 2, Tier 3 provides additional coverage to ensure 47 risk-significant plant equipment outage configurations are identified in a timely manner 48 and that the risk impact of out-of-service equipment is appropriately evaluated prior to 49 performing any maintenance activity over extended periods of plant operation. Tier 3 50 guidance can be satisfied by the Maintenance Rule, which requires a licensee to assess 51 and manage the increase in risk that may result from activities such as surveillance 1

testing and corrective and preventive maintenance, subject to the guidance provided in 2

RG 1.177, Section 2.3.7.1 and the adequacy of the licensees program and PRA model 3

for this application. The CRMP ensures that equipment removed from service prior to or 4

during the proposed extended CT will be appropriately assessed from a risk perspective.

5 6

Revision 2 of RG 1.200 describes an acceptable approach for determining whether the quality 7

of the PRA, in total or the parts that are used to support an application, is sufficient to provide 8

confidence in the results, such that the PRA can be used in regulatory decision making for 9

light-water reactors. This RG provides guidance for assessing the technical adequacy of a 10 PRA. Revision 2 of RG 1.200, endorses, with clarifications and qualifications, the use of the 11 American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Standard, 12 RA-Sa-2009, Addenda to ASME RA-S-2008 Standard for Level 1/Large Early Release 13 Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (i.e., the PRA 14 Standard).

15 16 As discussed in RG 1.177, Revision 1, and RG 1.174, Revision 3, a risk-informed application 17 should be evaluated to ensure that the proposed changes meet the following key principles:

18 19

1.

The proposed change meets the current regulations unless it is explicitly 20 related to a requested exemption; 21 22

2.

The proposed change is consistent with the defense-in-depth philosophy; 23 24

3.

The proposed change maintains sufficient safety margins; 25 26

4.

When proposed changes result in an increase in core damage frequency 27 or risk, the increases should be small and consistent with the intent of the 28 Commissions Safety Goal Policy Statement; and 29 30

5.

The impact of the proposed change should be monitored using 31 performance measurement strategies.

32 33

3.0 TECHNICAL EVALUATION

34 35

{NOTE: This SE can only be used when there are no TS or PRA loss of function conditions 36 included in the RICT program.}

37 38 The licensees adoption of TSTF-505, Revision 2, provides for the addition of a RICT Program 39 to the Administrative Controls section of the TS and modifies selected Required Action CTs to 40 permit extending the CTs, provided risk is assessed and managed as described in NEI 06-09-A.

41 In accordance with NEI 06-09-A, PRA methods are used to justify each extension to a Required 42 Action CT based on the specific plant configuration which exists at the time of the applicability of 43 the Required Action and are updated when plant conditions change. The licensees application 44 for the changes proposed in TSTF-505, Revision 2, included documentation regarding the 45 technical adequacy of the PRA models used in the CRMP, consistent with the requirements of 46 RG 1.200.

47 48 Most TS identify one or more Conditions for which the LCO may not be met, to permit a licensee 49 to perform required testing, maintenance, or repair activities. Each Condition has an associated 50 Required Action for restoration of the LCO or for other actions, each with some fixed time 51 interval, referred to as the CT, which identifies the time interval permitted to complete the 1

Required Action. Upon expiration of the CT, the licensee is required to shut down the reactor or 2

follow the Required Action(s) stated in the ACTIONS requirements. The RICT Program 3

provides the necessary administrative controls to permit extension of CTs and thereby delay 4

reactor shutdown or Required Actions, if risk is assessed and managed within specified limits 5

and programmatic requirements. The specified safety function or performance level of TS 6

required equipment is unchanged, and the Required Action(s), including the requirement to shut 7

down the reactor are also unchanged, only the CTs for the Required Actions are extended by 8

the RICT Program.

9 10 3.1 REVIEW OF KEY PRINCIPLES 11 12 Revision 1 of RG 1.177 and RG 1.174, Revision 3, identify five key safety principles to be 13 applied to risk-informed changes to the TSs. Each of these principles are addressed in 14 NEI 06-09-A. The NRC staffs evaluation of the licensees proposed use of RICTs against these 15 key safety principles is discussed below.

16 17 3.1.1 Key Principle 1: Evaluation of Compliance with Current Regulations 18 19 As stated in 10 CFR 50.36(c)(2), [l]imiting conditions for operation are the lowest functional 20 capability or performance levels of equipment required for safe operation of the facility. When a 21 limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the 22 reactor or follow any remedial action permitted by the technical specifications until the condition 23 can be met.

24 25 When the necessary redundancy is not maintained (e.g., one train of a two-train system is 26 inoperable), the TSs permit a limited period of time to restore the inoperable train to OPERABLE 27 status and/or take other remedial measures. If these actions are not completed within the CT, 28 the TSs normally require that the plant exit the mode of applicability for the LCO. With one train 29 of a two-train system inoperable, the TS safety function is accomplished by the remaining 30 OPERABLE train. In the current TSs, the CT is specified as a fixed time period (termed the 31 front stop). The addition of the option to determine the CT in accordance with the RICT 32 Program would allow an evaluation to determine a configuration-specific CT. The evaluation 33 would be done in accordance with the methodology prescribed in NEI 06-09-A and [TS 5.5.18].

34 The RICT is limited to a maximum of 30 days (termed the back stop) and can only be used 35 when there is no TS or PRA loss of function. The CTs in the current TSs were established 36 using experiential data, risk insights, and engineering judgement. The RICT Program provides 37 the necessary administrative controls to permit extension of CTs and thereby delay reactor 38 shutdown or Required Actions, if risk is assessed and managed appropriately within specified 39 limits and programmatic requirements.

40 41 When the necessary redundancy is not maintained and the system loses the capability to 42 perform its safety function(s) without any further failures (e.g., two trains of a two-train system 43 are inoperable), there is a TS loss of function and the plant must exit the mode of applicability 44 for the LCO, or take remedial actions, as specified in the TSs. A configuration-specific RICT 45 may not be determined and used following a TS loss of function because the system has lost 46 the capability to perform its safety function(s). With the incorporation of the RICT Program, the 47 required performance levels of equipment specified in LCOs are not changed. Only the 48 required CT for the Required Actions are modified by the RICT Program.

49 50 3.1.1.1 Key Principle 1 Conclusions 1

2 Based on the discussion provided above, the NRC staff finds that the proposed changes meet 3

the first key safety principle of RG 1.174, Revision 3, and RG 1.177, Revision 1.

4 5

3.1.2 Key Principle 2: Evaluation of Defense-in-Depth 6

7 Defense-in-depth is an approach to designing and operating nuclear facilities that prevents and 8

mitigates accidents that release radiation or hazardous materials. The key is creating multiple 9

independent and redundant layers of defense to compensate for potential human and 10 mechanical failures so that no single layer, no matter how robust, is exclusively relied upon.

11 Defense-in-depth includes the use of access controls, physical barriers, redundant and diverse 12 key safety functions, and emergency response measures.

13 14 As discussed throughout RG 1.174, consistency with the defense-in-depth philosophy is 15 maintained by the following:

16 17 Preserve a reasonable balance among the layers of defense.

18 19 Preserve adequate capability of design features without an overreliance on 20 programmatic activities as compensatory measures.

21 22 Preserve system redundancy, independence, and diversity commensurate with 23 the expected frequency and consequences of challenges to the system, including 24 consideration of uncertainty.

25 26 Preserve adequate defense against potential CCFs.

27 28 Maintain multiple fission product barriers.

29 30 Preserve sufficient defense against human errors.

31 32 Continue to meet the intent of the plants design criteria.

33 34 The proposed change represents a robust technical approach that preserves a reasonable 35 balance among avoidance of core damage, avoidance of containment failure, and consequence 36 mitigation. The three-tiered approach to risk-informed TS CT changes provides additional 37 assurance that defense-in-depth will not be significantly impacted by such changes to the 38 licensing basis. The licensee is proposing no changes to the design of the plant or any 39 operating parameter, no new operating configurations, and no new changes to the design-basis 40 in the proposed changes to the TS.

41 42 The effect of the proposed changes when implemented will be that the RICT Program will allow 43 CTs to vary based on the risk significance of the given plant configuration (i.e., the equipment 44 out-of-service at any given time) provided that the system(s) retain(s) the capability to perform 45 the applicable safety function(s) without any further failures (e.g., one train of a two-train system 46 is inoperable). A configuration-specific RICT may not be determined and used following a TS 47 loss of function because the system has lost the capability to perform its safety function(s).

48 These restrictions on TS loss of function or inoperability of all required trains of a system ensure 49 that consistency with the defense-in-depth philosophy is maintained by following existing 50 guidance when the capability to perform TS safety function(s) is lost.

51 1

The proposed RICT Program uses plant-specific operating experience for component reliability 2

and availability data. Thus, the allowances permitted by the RICT Program are directly 3

reflective of actual component performance in conjunction with component risk significance. In 4

some cases, the RICT Program may use compensatory actions to reduce calculated risk in 5

some configurations. Where credited in the PRA, these actions are incorporated into station 6

procedures or work instructions and have been modeled using appropriate human reliability 7

considerations. Application of the RICT Program determines the risk significance of plant 8

configurations. It also permits the operator to identify the equipment that has the greatest effect 9

on the existing configuration risk. With this information, the operator can manage the 10 out-of-service duration and determine the consequences of removing additional equipment from 11 service.

12 13

{NOTE: This paragraph is only included if Section 3.1.2.2 is needed.}

14

[The application of the RICT Program places high value on key safety functions and 15 works to ensure they remain a top priority over all plant conditions. The RICT will be 16 applied to extend CTs on key electrical power distribution systems. Failures in electrical 17 power distribution systems can simultaneously affect multiple safety functions; 18 therefore, potential degradation to defense-in-depth during the extended CTs is 19 discussed further below.]

20 21 3.1.2.1 Use of Compensatory Measures to Retain Defense-in-Depth 22 23 Application of the RICT Program provides a structure to assist the operator in identifying 24 effective compensatory actions for various plant maintenance configurations to maintain and 25 manage acceptable risk levels. Topical Report NEI 06-09-A addresses potential compensatory 26 actions and RMA measures by stating, in generic terms, that compensatory measures may 27 include but are not limited to the following:

28 29 Reduce the duration of risk-sensitive activities.

30 Remove risk-sensitive activities from the planned work scope.

31 Reschedule work activities to avoid high risk-sensitive equipment outages or 32 maintenance states that result in high-risk plant configurations.

33 Accelerate the restoration of out-of-service equipment.

34 Determine and establish the safest plant configuration.

35 36 Topical Report NEI 06-09-A requires that compensatory measures be initiated when the PRA 37 calculated RMA time (RMAT) is exceeded, or for preplanned maintenance for which the RMAT 38 is expected to be exceeded, RMAs shall be implemented at the earliest appropriate time.

39 40 3.1.2.2 Evaluation of Electrical Power Systems 41 42 According to the [PLANT] Updated FSAR, the plant is designed such that the safety functions 43 are maintained assuming a single failure within the electrical power system. By incorporating an 44 electrical power supply perspective, this concept is further reflected in a number of principal 45 design criteria for [PLANT]. Single failure requirements are typically suspended for the time 46 that a plant is not meeting an LCO (i.e., in an ACTION statement). This section considers the 1

plant configurations from a defense-in-depth perspective.

2 3

[Insert description of the facilitys electrical power system design.]

4 5

The licensee has requested to use the RICT Program to extend the existing CT for the following 6

TS 3.8, Electrical Power Systems, condition(s). The NRC staffs evaluation of the proposed 7

changes considered a number of potential plant conditions allowed by the proposed RICTs.

8 The staff also considered the available redundant or diverse means to respond to various plant 9

conditions. In these evaluations, the NRC staff examined the safety significance of different 10 plant conditions resulting in both shorter and longer CTs. The plant conditions evaluated are 11 discussed in more detail below.

12 13

[Insert a discussion of the plant conditions being evaluated as well as the criteria used to 14 evaluate the condition. At a minimum, the evaluation of the plant condition shall include 15 (a) the design success criteria for accomplishing safety functions, (b) the verification of 16 remaining credited subsystem(s) (e.g., power source, inverter, etc.), (c) if applicable, the 17 availability of additional power source(s)/SSCs, and (d) examples of the compensatory 18 measures or RMAs.]

19 20 The NRC staff reviewed the licensees proposed TS changes and supporting documentation.

21 Based on the evaluations above, the staff finds that while the redundancy is not maintained 22 (e.g., one train of a two train system is inoperable), the CT extensions in accordance with the 23 RICT Program are acceptable because (a) the capability of the systems to perform their safety 24 functions (assuming no additional failures) is maintained, and (b) the licensees demonstration 25 of identifying and implementing compensatory measures or RMAs, in accordance with the RICT 26 Program, are appropriate to monitor and control risk.

27 28 3.1.2.3 Evaluation of Instrumentation and Control Systems 29 30

{NOTE: Include this section of the SE if the licensee proposed to include instrumentation and 31 control TS in the RICT Program.}

32 33 The licensee has requested to use the RICT Program to extend the existing CT for the following 34 TS conditions. The NRC staffs evaluation of the proposed changes considered a number of 35 potential plant conditions allowed by the new TSs and considered what redundant or diverse 36 means were available to assist the licensee in responding to various plant conditions. The plant 37 conditions evaluated are discussed in more detail below.

38 39

[Insert a discussion of the plant conditions being evaluated as well as the criteria used to 40 evaluate the conditions in light of the RICT Program. The evaluation of the plant 41 condition(s) shall include the basis for the evaluation including the design success 42 criteria, the capability of the instrumentation and control systems to perform their safety 43 functions, and diverse means to accomplish the safety functions.]

44 45 Since the licensee did not propose any changes to the design basis, the independency and the 46 fail-safe principle remain unchanged. The licensee stated in the LAR that the proposed 47 changes did not include any TS loss of function conditions. However, it is recognized that while 48 in an ACTION statement, redundancy of the given protective feature will be temporarily 49 reduced, and, accordingly, the system reliability will be reduced. In the LAR, the licensee stated 50 in the description of proposed changes to the instrumentation and control systems that at least 51 one redundancy or diverse means (e.g., other automatic features or manual action) to 1

accomplish the safety functions (e.g., reactor trip, safety injection, or containment isolation) 2 remain available during the use of the RICT. The NRC staff reviewed the licensees proposed 3

TS changes to assess the availability of the diverse means to accomplish the safety function(s).

4 The NRC staff finds that the availability of diverse protective features provide sufficient 5

defense-in-depth to accomplish the safety functions, allowing for the extension of CTs in 6

accordance with the RICT Program.

7 8

The NRC staff reviewed the licensees proposed TS changes and supporting documentation.

9 The NRC staff finds that while the instrumentation and control redundancy is reduced, the CT 10 extensions implemented in accordance with the RICT Program are acceptable because: (a) the 11 capability of the instrumentation and control systems to perform their safety functions is 12 maintained, (b) diverse means to accomplish the safety functions exist, and (c) the licensee will 13 identify and implement risk management actions to monitor and control risk in accordance with 14 the RICT Program.

15 16 3.1.2.4 Key Principle 2 Conclusions 17 18 The LAR proposes to modify the TS requirements to permit extending selected CTs using the 19 RICT Program in accordance with NEI 06-09-A. The NRC staff has reviewed the licensees 20 proposed TS changes and supporting documentation. The NRC staff finds that extending the 21 selected CTs with the RICT Program following loss of redundancy, but maintaining the 22 capability of the system to perform its safety function, is an acceptable reduction in 23 defense-in-depth provided that the licensee identifies and implements compensatory measures 24 as appropriate during the extended CT. Therefore, quantitative risk analysis, the qualitative 25 considerations, and the prohibition on loss of all trains of a required system assure a reasonable 26 balance of defense-in-depth is maintained to ensure protection of public health and safety. The 27 NRC staff finds that this proposed change meets the second key safety principle of RG 1.177 28 and is, therefore, acceptable.

29 30 Based on the preceding evaluation, the NRC staff concludes that the proposed changes are 31 consistent with the defense-in-depth philosophy as described in RG 1.174.

32 33 3.1.3 Key Principle 3: Evaluation of Safety Margins 34 35 Section 2.2.2 of RG 1.177, Revision 1, states, in part, that sufficient safety margins are 36 maintained when:

37 38 Codes and standards or alternatives approved for use by the NRC are 39 met...

40 Safety analysis acceptance criteria in the final safety analysis report 41 (FSAR) are met or proposed revisions provide sufficient margin to account 42 for analysis and data uncertainties.

43 44 In Section [x.x] of its submittal, the licensee confirmed that the use of the RICT Program to 45 determine a RICT will not affect [PLANT] commitment to the codes and standards used in the 46 design of [PLANT]. Further the licensee is not proposing in this application to change any 47 quality standard, material, or operating specification. Acceptance criteria for operability of 48 equipment are not changed and use of the RICT only when the system(s) retain(s) the capability 49 to perform the applicable safety function(s) ensure that the current safety margins are retained.

50 Safety margins are also maintained if PRA functionality is determined for the inoperable train 51 which would result in an increased CT. Credit for PRA functionality, as described in 1

NEI 06-09-A, is limited to the inoperable train, loop, or component. The reduced but available 2

functionality may support further increase in the CT consistent with available safety margin. The 3

specified safety function is still being met by the operable train and therefore requires no 4

evaluation of PRA functionality to meet the design basis success criteria.

5 6

3.1.3.1 Key Principle 3 Conclusions 7

8 As discussed above, the NRC staff finds that the design-basis analyses for [PLANT] remain 9

applicable. Although the licensee will be able to have design-basis equipment out-of-service 10 longer than the current TS allow and the likelihood of successful fulfillment of the function will be 11 decreased when redundant train(s) are not be available, the capability to fulfill the function will 12 be retained when the available equipment functions as designed. Any increase in unavailability 13 because less equipment is available for a longer time is included in the RICT evaluation.

14 Therefore, safety margins are not affected adversely by the implementation of the RICT 15 Program. Based on the above, the NRC staff concludes that the proposed change meets the 16 third key safety principle of RG 1.177 and is acceptable.

17 18 3.1.4 Key Principle 4: Change in Risk Consistent with the Safety Goal 19 Policy Statement 20 21 In Section [x.x] of its submittal, the licensee described the guidelines that will be used to 22 determine acceptable changes in risk. The NRC staff evaluated whether the change in risk from 23 the proposed changes was small and consistent with the intent of the Commissions Safety Goal 24 Policy Statement, as discussed below. The NRC staff evaluated the licensees proposed 25 changes against the three-tiered approach in RG 1.177, Revision 1, for the licensees evaluation 26 of the risk associated with a proposed TS CT change.

27 28 3.1.4.1 Tier 1: PRA Capability and Insights 29 30 The first tier evaluates the impact of the proposed changes on plant operational risk. The Tier 1 31 review involves two aspects: (1) the technical acceptability of the PRA models and their 32 application to the proposed changes, and (2) a review of the PRA results and insights described 33 in the licensees application.

34 35 3.1.4.1.1 PRA Quality 36 37 The objective of the PRA quality review is to determine whether the [PLANT] PRA used to 38 implement the RICT Program is of sufficient scope, level of detail, and technical adequacy for 39 this application.

40 41 The NRC staff evaluated the PRA quality information provided by the licensee in Section [x.x] of 42 its submittal, including industry peer review results and the licensees self-assessment of the 43 plant PRA models for internal and external events, including fires [seismic, other external 44 hazards] against the requirements of the currently applicable revision of RG 1.200, 45

[Revision 2].

46 47

[Insert the plant-specific evaluation of each PRA model. This is a detailed discussion of 48 the peer reviews and other internal self-assessments to determine the conformance of 49 the PRA models to capability Category II of the relevant PRA standards. Failure of a PRA 50 model to conform to one or more supporting requirements of a standard at capability 1

Category II should be dispositioned for acceptability for use in the RICT Program.]

2 3

Based on the NRC staffs review of the licensees submittal and assessments, the NRC staff 4

determined that the [PLANT] PRA models for internal and external events, fires [seismic, other 5

external hazards] used to implement the RICT Program satisfy the guidance of RG 1.200.

6 7

[Insert discussion of capability categories contrasted with NRC staff SE of NEI 06-09-A 8

direction that all SRs adequately conform to capability Category II of the American 9

Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) standard for the 10 supporting requirements.]

11 12 Based on the review of the provided information, the [PLANT] PRA models were determined to 13 be of sufficient technical adequacy to support implementation of the RICT Program.

14

[Therefore, the NRC staff finds that the licensee has satisfied the intent of RG 1.177, 15 Revision 1 (Sections2.3.1, 2.3.2, and 2.3.3), and RG 1.174, Revision 3 16 (Sections 2.3 and 2.5); and that the quality of the [PLANT] PRA is sufficient to implement 17 RMTS in accordance with the RICT Program and NEI 06-09-A.]

18 19 The NRC staff has reviewed the results of the peer reviews to assess whether the PRA is 20 adequate to support the RICT Program. [Insert discussion of conclusions.] The issues have 21 been resolved satisfactorily [or will be resolved before implementation of the RICT 22 Program]. {NOTE: If using bracketed option, staff should consider making this a license 23 condition, discussing it in the cover letter, and including it in the implementation requirements.}

24 25 The licensee has also established a periodic update and review process for the PRA and 26 associated CRMP model. {NOTE: Verify there are no changes to the change control processes 27 for PRA methods. If so, insert a discussion of the change control process that is in the license 28 condition or TS Section 5 requirement. The addition to TS Section 5 paragraph (e) requires that 29 RICTs be calculated using NRC accepted methods. The NRC documents acceptance of PRA 30 methods in a number of different ways including plant-specific SEs, topical reports, SEs, facts 31 and observations (F&O) closures, FAQs, and through the proposed vetting panel process.}

32 33 The licensee (1) has reviewed the PRA using endorsed guidance and adequately resolved all 34 identified issues, (2) has established a periodic update and review process to update the PRA 35 and associated CRMP model to incorporate changes made to the plant, and (3) will calculate 36 RICTs using NRC-accepted PRA methods. Therefore, the NRC staff concludes that the 37 licensee has and will maintain a PRA that is technically adequate to support implementation of 38 the RICT Program.

39 40 3.1.4.1.2 Scope of the PRA 41 42 Topical Report NEI 06-09-A requires a quantitative assessment of the potential impact on risk 43 due to impacts from internal and external events, including internal fires, floods, and other 44 significant external events. As discussed in Section 3.2.4.1.1, the [PLANT] PRA used for the 45 RICT Program includes contributions from internal and external events, including internal fires 46 and floods[, seismic events, and other external events]. In addition, the NRC staff finds that 47 the seismic and other external hazard analyses (i.e., do not have seismic margins analysis or 48 seismic PRA models) provide a bounding approach for the RICT Program consistent with the 49 NEI 06-09-A guidance on bounding analyses.

50 51

{NOTE: NRC would expect to establish additional requirements in the Administrative Controls 1

TS and/or a license condition for incomplete PRAs that rely on bounding analysis.}

2 3

{NOTE: Provide a summary of how the PRA used for the RICT Program addresses seismic 4

events and other external hazards if a full-scope plant-specific PRA model is not used. This 5

may be a justification that the contribution from these hazards is not significant to the RICT 6

calculations, or a justification for the use of bounding quantitative analyses.}

7 8

Because the RICT Program is not applicable in Modes [4 and 5/5 and 6], risk evaluations for 9

these modes are not relevant to the proposed change.

10 11

[Based on the above, the NRC staff finds that the licensee has satisfied the intent of 12 RG 1.177, Revision 1 (Section 2.3.2), and RG 1.174, Revision 3 (Section 2.3), and that the 13 scope of the PRA model is appropriate for this application.]

14 15 3.1.4.1.3 PRA Modeling 16 17 To evaluate a RICT for a given Required Action, the specific systems or components involved 18 should be modeled in the PRA. For each TS LCO for which the RICT Program is proposed to 19 apply, for any of its Required Actions, the licensee identified that: (1) the system is included in 20 the PRA models, or has addressed systems not in the PRA either in the LAR or in response to 21 an RAI; (2) the success criteria parameters used to determine PRA Functional determination 22 are the same as the design basis success criteria parameters or, if different, a plant-specific 23 analyses used to support the PRA are justified; (3) CCFs and surrogate identification[, and 24 plant-specific PRA modelling issues] are appropriately addressed; and (4) the CRMP 25 provides the capability to select the system as out of service in order to calculate a RICT, and 26 the CRMP is maintained consistent with the baseline PRA model with modifications to the 27 CRMP model to reflect the current plant versus the average plant.

28 29

[Insert a summary of the PRA system modeling and how the licensee provides that 30 (1) the system is included in the PRA models, or has addressed systems not in the PRA 31 either in the LAR or in response to a request for additional information (RAI); (2) the 32 success criteria parameters used to determine PRA Functional determination are the 33 same as the design basis success criteria parameters or, if different, a plant-specific 34 analyses used to support the PRA are justified; and (3) common-cause failures, 35 surrogates identification and plant specific PRA modelling issues (e.g., instruments) if 36 any.]

37 38 With respect to Item (4), the PRA model serves as the model used by the CRMP tool, which is 39 used to perform the RICT calculations. [Insert discussion of tool.] The tool used to perform 40 the RICT calculations provides a user interface which supports the RICT Program by providing 41 a method to evaluate the plant configuration.

42 43 The NRC staff reviewed the licensees information and concluded that the PRA modelling used 44 to support the RICT Program is able to treat alignments of components during periods when the 45 RICT will be calculated. [Therefore, the NRC staff finds that the licensee has satisfied the 46 intent of RG 1.177, Revision 1 (Section 2.3.3), and RG 1.174, Revision 3 (Section 2.3), and 47 that the PRA modeling is appropriate for this application.]

48 49 3.1.4.1.4 Assumptions 1

2 Using PRAs to evaluate TS changes requires consideration of a number of assumptions made 3

within the PRA that can have a significant influence on the ultimate acceptability of the proposed 4

changes. With regard to changes to CTs, the following assumptions were evaluated:

5 6

{NOTE: Insert the plant-specific PRA assumptions and disposition of each for the RICT 7

Program. This should include a description of the methods used to identify assumptions.}

8 9

[Based on the identification and disposition of the significant PRA assumptions 10 described above, the NRC staff finds that the licensee has satisfied the intent of 11 RG 1.177, Revision 1 (Section 2.3.4), and that the assumptions for risk evaluation of 12 extended CTs are appropriate for this application.]

13 14 3.1.4.1.5 Sensitivity and Uncertainty Analyses 15 16 Risk-informed analyses of TS changes can be affected by uncertainties regarding the 17 assumptions made during the PRA models development and application. Typically, the risk 18 resulting from TS CT changes is relatively insensitive to most uncertainties because the 19 uncertainties tend to affect similarly both the base case and the changed case. The licensee 20 considered PRA modeling uncertainties and their potential impact on the RICT Program and 21 identified, as necessary, the applicable RMAs to limit the impact of these uncertainties. In 22 Section [x.x] of its submittal, the licensee discussed sources of uncertainty and assumptions.

23 24 The licensee performed an evaluation of its PRA model for [PLANT] to identify the key 25 assumptions and sources of uncertainty for this application consistent with the RG 1.200 26 definitions, using sensitivity and importance analyses to place bounds on uncertain processes, 27 to identify alternate modeling strategies, and to provide information to users of the PRA.

28 29

{NOTE: Insert the plant-specific PRA uncertainties and disposition of each for the RICT 30 program. This should also include a description of the methods used to identify uncertainties.}

31 32 The NRC staffs review indicates the licensee performed an adequate assessment to identify 33 the potential sources of uncertainty, and the identification of the key assumptions and sources 34 of uncertainty was appropriate and consistent with RG 1.174, Revision 3. The licensees 35 evaluation of the potential impact of these sources of uncertainty on the RICT Program is 36 acceptable.

37 38

[Therefore, the NRC staff finds that the licensee has satisfied the intent of RG 1.177, 39 Revision 1 (Section 2.3.5), and RG 1.174, Revision 3 (Section 2.2), and that the treatment 40 of model uncertainties for risk evaluation of extended CTs is appropriate for this 41 application and consistent with the guidance identified in NEI 06-09-A.]

42 43 3.1.4.1.6 PRA Results and Insights 44 45 The proposed change implements a process to determine TS RICTs rather than specific 46 changes to individual TS CTs. Topical Report NEI 06-09-A requires periodic assessment of the 47 risk incurred due to operation beyond the front stop CTs due to implementation of a RICT 48 Program and comparison to the guidance of RG 1.174, Revision 3, for small increases in risk.

49 As with other unique risk-informed applications, supplemental risk acceptance guidelines that 50 complement the RG 1.174 guidance are appropriate.

51 1

Further, NEI 06-09-A requires that configuration risk be assessed to determine the RICT, and 2

establishes the criteria for ICDP and ILERP on which to base the RICT. An ICDP of 1E-5 and 3

an ILERP of 1E-6 are used as the risk measures for calculating individual RICTs. These limits 4

are consistent with NUMARC 93-01, Revision 4A. The use of these limits in NEI 06-09-A aligns 5

the TS CTs with the risk management guidance used to support plant programs for the 6

Maintenance Rule, and the NRC staff accepted these supplemental risk acceptance guidelines 7

for RMTS programs in its approval of NEI 06-09-A.

8 9

Topical Report NEI 06-09-A requires that the cumulative impact of implementation of an RMTS 10 be periodically assessed and shown to result in: (1) a total risk impact below 1E-5/year for 11 changes to core damage frequency (CDF), (2) a total risk impact below 1E-6/year for changes 12 to large early release frequency (LERF), and (3) the total CDF and total LERF must be 13 reasonably shown to be less than 1E-4/year and 1E-5/year, respectively. The licensee 14 indicated in [Enclosure X] of its submittal that the estimated total CDF and LERF meet the 15 1E-4/year CDF and 1E-5/year LERF criteria of RG 1.174 consistent with the guidance in 16 NEI 06-09-A and that these guidelines be satisfied whenever a RICT is implemented.

17 18 The licensee has incorporated NEI 06-09-A in the RICT Program of TS [5.5.15/5.5.18],

19 calculates the RICT consistently with its criteria, and assesses the RICT Program to assure any 20 risk increases are small per the guidance of RG 1.174.

21 22 Based on satisfying the intent of RG 1.177, Revision 1 (Section 2.4), and RG 1.174, Revision 3 23 (Sections 2.4 and 2.5), the NRC staff finds the proposed changes are acceptable.

24 25 3.1.4.1.7 Implementation of the RICT Program 26 27 Because NEI 06-09-A involves the real-time application of PRA results and insights by the 28 licensee, the NRC staff reviewed the licensees description of programs and procedures 29 associated with implementation of the RICT Program in Section [x.x] of its submittal. The 30 administrative controls on the PRA and on changes to the PRA should provide confidence that 31 the PRA results are reasonable, and the administrative controls on the plant personal using the 32 RICT should provide confidence that the RICT program will be applied appropriately.

33 34 The quality assurance practices for the PRA models include meeting the ASME/ANS PRA 35 standards and RG 1.200, which includes guidance for performing peer reviews and 36 focused-scope peer reviews. The quality assurance practices for the PRA models are 37 discussed by the licensee in [Enclosure X] of its submittal. According to Section [x.x] of its 38 submittal, for maintenance of the baseline PRA model, changes made to the baseline PRA 39 model in translation to the on-line model, and changes made to the on-line model configuration 40 files are controlled and documented by plant procedures.

41 42

[Insert a summary of the process used to convert the baseline to the on-line PRA 43 models.]

44 45 In Section [x.x] of its submittal, the licensee indicated that those procedures are intended to 46 specify an acceptance test to be performed after every on-line model update. This test verifies 47 proper translation of the baseline PRA models and acceptance of all changes made to the 48 baseline PRA models into the on-line model. This test also verifies correct mapping of plant 49 components into the on-line model.

50 51

[Insert a summary of these programs, procedures, and training.]

1 2

[The NRC staff found that the licensee has established appropriate programmatic and 3

procedural controls for its RICT Program, consistent with the guidance of NEI 06-09-A.]

4 NEI 06-09-A requires that stations implementing a RMTS program shall provide training in the 5

programmatic requirements associated with the RMTS program and of the individual RICT 6

evaluations to personnel responsible for determining TS operability decisions or conducting 7

RICT assessments. Training of plant personnel shall be provided for those organizations with 8

functional responsibilities for performing or administering the CRMP commensurate with each 9

positions responsibilities, in accordance with 10 CFR 50.120(b)(3) and other applicable 10 regulations, within the RICT Program, as described in NEI 06-09-A. In [Enclosure 9] of its 11 submittal, the licensee described its program for providing training to its staff. The NRC staff 12 reviewed the description of the training program provided in the license amendment request, 13 and concluded that the program, if properly implemented, would be consistent with the training 14 requirements set for the in NEI 06-09-A.

15 16 Therefore, the NRC staff finds that the licensee has proposed acceptable administrative controls 17 on the PRA and on the personnel that will use the RICT Program.

18 19 3.1.4.2 Tier 2: Avoidance of Risk-Significant Plant Configurations 20 21 The second tier provides that a licensee should provide reasonable assurance that 22 risk-significant plant equipment outage configurations will not occur when specific plant 23 equipment is taken out-of-service in accordance with the proposed TS change.

24 25 Topical Report NEI 06-09-A does not permit voluntary entry into high-risk configurations, which 26 would exceed instantaneous CDF and LERF limits of 1E-3/year and 1E-4/year, respectively. It 27 further requires implementation of RMAs when the actual or anticipated risk accumulation 28 during a RICT will exceed one-tenth of the ICDP or ILERP limit. Such RMAs may include 29 rescheduling planned activities to lower risk periods or implementing risk-reduction measures.

30 The limits established for entry into a RICT and for RMA implementation are consistent with the 31 guidance of NUMARC 93-01, Revision 4A, endorsed by RG 1.160, Revision 3, as applicable to 32 plant maintenance activities. The RICT Program requirements and criteria are consistent with 33 the principle of Tier 2 to avoid risk-significant configurations.

34 35 Based on the licensees incorporation of NEI 06-09-A in the TS as discussed in Section [x.x] of 36 its submittal, and because the proposed changes are consistent with the guidance of RG 1.174, 37 Revision 3, and RG 1.177, Revision 1, the NRC staff finds the licensees Tier 2 program is 38 acceptable and supports the proposed implementation of the RICT Program.

39 40 3.1.4.3 Tier 3: Risk-Informed Configuration Risk Management 41 42 The third tier provides that a licensee should develop a program that ensures that the risk 43 impact of out-of-service equipment is appropriately evaluated prior to performing any 44 maintenance activity.

45 46 Topical Report NEI 06-09-A addresses Tier 3 guidance by requiring assessment of the RICT to 47 be based on the plant configuration of all SSCs that might impact the RICT, including 48 safety-related and non-safety-related SSCs. A plant configuration is considered risk-significant 49 when the ICDP or the ILERP exceeds one-tenth of the risk on which the RICT is based, 50 generally 1E-5 and 1E-6 ICDP and ILERP, respectively. If a risk-significant plant configuration 51 exists, then NEI 06-09-A via the RICT Program in the TS, would require the licensee to 1

implement compensatory measures and RMAs. Therefore, the NRC staff determined that the 2

RICT Program provides a methodology to assess and address risk-significant configurations.

3 The NRC staff also determined that the proposed changes will require reassessment of any 4

plant configuration changes to be completed in a timely manner based on the more restrictive 5

limit of any applicable TS action requirement or a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the configuration 6

change occurs.

7 8

Based on the licensees incorporation of NEI 06-09-A in the TS, as discussed in Section [x.x] of 9

its submittal, and because the proposed changes are consistent with the Tier 3 guidance of 10 RG 1.177, Revision 1, the NRC staff finds that the proposed changes are acceptable.

11 12 3.1.4.4 Key Principle 4 Conclusions 13 14 The licensee has demonstrated the technical adequacy and scope of its PRA models, and that 15 the models can support implementation of the RICT Program for determining CTs. Proper 16 consideration of key assumptions and sources of uncertainty have been made. The risk metrics 17 are consistent with the approved methodology of NEI 06-09-A and the RICT Program is 18 controlled administratively through plant procedures and training. The RICT Program follows 19 the NRC-approved methodology in NEI 06-09-A. The NRC staff concludes that the RICT 20 Program satisfies the fourth key safety principle of RG 1.177 and is, therefore, acceptable.

21 22 3.1.5 Key Principle 5: Performance Measurement Strategies - Implementation 23 and Monitoring Program 24 25 Revision 1 of RG 1.177 and RG 1.174, Revision 3, establish the need for an implementation 26 and monitoring program to ensure that extensions to TS CTs do not degrade operational safety 27 over time and that no adverse degradation occurs due to unanticipated degradation or common 28 cause mechanisms. An implementation and monitoring program is intended to ensure that the 29 impact of the proposed TS change continues to reflect the reliability and availability of SSCs 30 impacted by the change. Revision 3 of RG 1.174 states that monitoring performed in 31 conformance with the Maintenance Rule, 10 CFR 50.65, can be used when the monitoring 32 performed is sufficient for the SSCs affected by the risk-informed application. [According to 33

[Enclosure X] of the submittal, the SSCs in the scope of the RICT Program are also in the 34 scope of the Maintenance Rule.]

35 36 Section 3.3.3 of NEI 06-09-A requires that the licensee track the risk associated with all entries 37 beyond the front stop CT, and Section 2.3.1 provides a requirement for assessing cumulative 38 risk, including a periodic evaluation of any increase in risk due to the use of the RMTS program 39 to extend the CTs. According to [Enclosure X] of its submittal, the licensee calculates 40 cumulative risk at least every refueling cycle, but the recalculation period does not exceed 41 24 months, which is consistent with an NEI 06-09-A program. The licensee converts the 42 cumulative ICDP and the ILERP into average annual values which are then compared to the 43 limits of RG 1.174. If any limits are exceeded, corrective actions are taken to ensure future 44 plant operational risk is within the acceptance guidance. This evaluation assures that RMTS 45 program implementation meets RG 1.174 guidance for small risk increases. The licensee is 46 implementing NEI 06-09-A via the RICT Program and therefore complies with this RMTS 47 program.

48 49 The NRC staff concludes that the RICT Program satisfies the fifth key safety principle of 50 RG 1.177, Revision 1, and is therefore, acceptable.

51 1

[3.2 VARIATIONS FROM TSTF-505 2

3

[Insert an evaluation of variations discussed in Section 2.2.4 of this SE. This would 4

include any variations related to the treatment of new PRA methods as described in the 5

RICT Program.]

6 7

[The traveler discusses the applicable regulatory requirements and guidance, including 8

the 10 CFR Part 50, Appendix A, General Design Criteria. [PLANT] was not licensed to 9

the 10 CFR Part 50, Appendix A, GDC. The [PLANT] equivalents to the referenced GDC 10 are [discussion from licensee's application.] These differences do not alter the 11 conclusion that the proposed change is applicable to [PLANT].]

12 13 3.3 TECHNICAL SPECIFICATION ADMINISTRATIVE CONTROLS SECTION 14 15 The NRC staff reviewed the licensees proposed addition of a new program, the RICT Program, 16 to the Administrative Controls section of the TS. The NRC staff evaluated the elements of the 17 new program to ensure alignment with the requirements in 10 CFR 50.36(c)(5) and to ensure 18 the programmatic controls are consistent with the RICT Program described in NEI 06-09-A.

19 20

[TS 5.5.15/TS 5.5.18] requires that the RICT Program be implemented in accordance with 21 NEI 06-09-A. This is acceptable because NEI 06-09-A establishes a framework for an 22 acceptable RICT Program.

23 24 The TS states that a RICT may not exceed 30 days. The NRC staff determined that 30-day 25 backstop is appropriate because it allows sufficient time to restore SSCs to operable status 26 while avoiding excessive out of service times for TS SSCs.

27 28 The TS states that the RICT may only be used in Mode 1, 2[, and 3, and 4 while relying on 29 steam generators for decay heat removal][, and Mode 3 while relying on the main 30 condenser for heat removal]. This provision ensures that the RICT is only used for 31 determination of CDF and LERF for modes of operation modelled in covered by the PRA.

32 33 The TS requires that while in a RICT, any change in plant configuration as defined in 34 NEI 06-09-A be considered for the effect on the RICT. The TS also specifies time limits for 35 determining the effect on the RICT. These time limitations are consistent with those specified in 36 NEI 06-09-A.

37 38 The TS contains requirements for the treatment of CCFs for emergent conditions in which the 39 common cause evaluation is not complete. The requirements are to either numerically account 40 for the increased probability of CCF or to implement RMAs that support redundant or diverse 41 SSCs that perform the functions of the inoperable SSCs and, if practicable, reduce the 42 frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

43 Key Principle 2 of risk-informed decision making is that the change is consistent with 44 defense-in-depth philosophy. The seven considerations supporting the evaluation of the impact 45 of the change on defense-in-depth are discussed in RG 1.174, including one to preserve 46 adequate defense against potential CCF. The NRC staff finds that numerically accounting for 47 an increased probability of failure will shorten the estimated RICT based on the particular SSCs 48 involved thereby limiting the time when a CCF could affect risk. Alternatively, implementing 49 actions that can increase the availability of other mitigating SSCs or decrease the frequency of 50 demand on the affected SSCs will decrease the likelihood that a CCF could affect risk. The 51 NRC staff concludes that both the quantitative and the qualitative actions minimize the impact of 1

CCF and therefore support meeting Key Principle 2 as described in RG 1.174. These methods 2

either limit the exposure time, help ensure the availability of alternate SSCs, or decrease the 3

probability of plant conditions requiring the safety function to be performed. The NRC staff finds 4

that these methods contribute to maintaining defense-in-depth because the methods limit the 5

exposure time or ensure the availability of alternate SSCs.

6 7

The TS contains a provision that risk assessment approaches and methods shall be acceptable 8

to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; 9

and reflect the operating experience at the plant, as specified in RG 1.200, Revision 2. Methods 10 to assess the risk from extending the CTs must be PRA methods used to support this LAR, or 11 other methods approved by the NRC for generic use, and any change in the PRA methods to 12 assess risk that are outside these approval boundaries require prior NRC approval. As stated in 13 the NRC staffs SE of NEI 06-09-A:

14 15 TR NEI 06-09, Revision 0, requires an evaluation of the PRA 16 model used to support the RMTS against the requirements of 17 RG 1.200, Revision 1, and AMSE RA-S-2002, Standard for 18 Probabilistic Risk Assessment for Nuclear power Plant 19 Applications, for capability Category II. This assures that the PRA 20 model is technically adequate for use in the assessment of 21 configuration risk. This capability category of PRA is sufficient to 22 support the evaluation of risk associated with out of service SSCs 23 and establishing risk-informed CTs.

24 25

[TS 5.5.15/5.5.18] was updated to reflect the current revision of RG 1.200. RG 1.200 26 incorporates ASME RA-S-2002 by reference.

27 28 The NRC staffs SE of NEI 06-09-A also states:

29 30 As part of its review and approval of a licensees application 31 requesting to implement the RMTS, the NRC staff intends to 32 impose a license condition that will explicitly address the scope of 33 the PRA and non-PRA methods approved by the NRC staff for 34 use in the plant-specific RMTS program. If a licensee wishes to 35 change its methods, and the change is outside the bounds of the 36 license condition, the licensee will need NRC approval, via a 37 license amendment, of the implementation of the new method in 38 its RMTS program. The focus of the NRC staffs review and 39 approval will be on the technical adequacy of the methodology 40 and analyses relied upon for the RMTS application.

41 42 This limitation and condition is being relocated from a license condition to the Administrative 43 Controls section of TS. [TS 5.5.15/5.5.18] restates this limitation and condition from the NRC 44 staffs SE in language appropriate for the Administrative Controls Section of TS. The staff finds 45 that this requirement is appropriately reflected in the Administrative Controls section of TS.

46 47 The regulations in 10 CFR 50.36(c)(5) require the TS to contain administrative controls 48 providing provisions relating to organization and management, procedures, recordkeeping, 49 review and audit, and reporting necessary to assure operation of the facility in a safe manner.

50 for the contents of the Administrative Controls section of the TS. The NRC staff has determined 51 that Administrative Controls section of the TS is will assure operation of the facility in a safe 1

manner when the facility is using the RICT program. Therefore, the NRC staff has determined 2

that the requirements of 10 CFR 50.36(c)(5) are satisfied.

3 4

4.0 ADDITIONAL CHANGES TO THE OPERATING LICENSE 5

6

[Insert a discussion of any license conditions or implementation requirements.]

7 8

5.0

SUMMARY

9 10 The NRC staff finds that the licensees proposed implementation of the RICT Program for the 11 identified scope of Required Actions is consistent with the guidance of NEI 06-09-A[, subject to 12 the limitations and conditions evaluated in Section 4.0 of this SE]. The licensees 13 methodology for assessing the risk impact of extended CTs, including the individual CT 14 extension impacts in terms of ICDP and ILERP, and the overall program impact in terms of 15 CDF and LERF, is accomplished using PRA models of sufficient scope and technical 16 adequacy based on consistency with the guidance of RG 1.200, Revision 2. [For external 17 hazards which do not have PRA models, the licensee will use bounding analyses in 18 accordance with NEI 06-09-A guidance and Administrative Control TS and/or license 19 condition provided in this SE]. The RICT calculation uses the PRA model as translated into 20 the CRMP tool, and the licensee has an acceptable process in place to ensure the quality of the 21 translation. In addition, the NRC staff finds that the proposed implementation of the RICT 22 Program addresses the RG 1.177 defense-in-depth philosophy and safety margins to ensure 23 that they are adequately maintained, and includes adequate administrative controls as well as 24 performance monitoring programs.

25 26 The regulation at 10 CFR 50.36(a)(1) states, in part: A summary statement of the bases or 27 reasons for such specifications other than those covering administrative controls shall also be 28 included in the application, but shall not become part of the technical specifications.

29 Accordingly, along with the proposed TS changes, the licensee also submitted TS Bases 30 changes that corresponded to the proposed TS changes to provide the reasons for the TSs.

31 The TS bases changes were consistent with the bases changes in the model application.

32 33

6.0 STATE CONSULTATION

34 35 This section is to be prepared by the plant project manager.

36 37 In accordance with the Commission's regulations, the [Name of State] State official was notified 38 of the proposed issuance of the amendment(s) on [date]. The State official had [no]

39 comments. [If comments were provided, they should be addressed here.]

40 41

7.0 ENVIRONMENTAL CONSIDERATION

42 43 This section is to be prepared by the plant project manager in accordance with current 44 procedures.

45 46

8.0 CONCLUSION

47 48 This section is to be prepared by the plant project manager.

49 50 The Commission has concluded, based on the considerations discussed above, that: (1) there 1

is reasonable assurance that the health and safety of the public will not be endangered by 2

operation in the proposed manner, (2) there is reasonable assurance that such activities will be 3

conducted in compliance with the Commission's regulations, and (3) the issuance of the 4

amendment(s) will not be inimical to the common defense and security or to the health and 5

safety of the public.

6 7

9.0 REFERENCES

8 9

Optional section to be prepared by the PM and primary reviewers. If document is publicly 10 available, the ADAMS Accession No. should be listed for all references.

11 12 13

{NOTE: These are the principal contributors for the model SE of the traveler. Replace these 14 names with those you prepared the plant-specific SE.}

15 Principal Contributors:

M. Biro 16 N. Carte 17 M. Chernoff 18 S. Dinsmore 19 J. Evans 20 M. Li 21 K. Nguyen 22 23 Date of issuance:

24