ML20154K606

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Draft Safety Evaluation of TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements
ML20154K606
Person / Time
Site: Technical Specifications Task Force
Issue date: 10/30/2020
From: Victor Cusumano
NRC/NRR/DSS/STSB
To:
Technical Specifications Task Force
Honcharik,M., NRR/DSS, 301-415-1774
Shared Package
ML20304A138 List:
References
EPID L-2019-PMP-0181, TSTF-554, Rev 1
Download: ML20154K606 (12)
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1 DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 2 TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER 3 TSTF-554, REVISION 1, REVISE REACTOR COOLANT LEAKAGE REQUIREMENTS 4 USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS 5 (EPID L-2019-PMP-0181) 6 7

1.0 INTRODUCTION

8 9 By letter dated January 16, 2020 (Agencywide Documents Access and Management System 10 (ADAMS) Accession No. ML20016A233), the Technical Specifications Task Force (TSTF) 11 submitted traveler TSTF-554, Revision 1, Revise Reactor Coolant Leakage Requirements, 12 (TSTF-554) to the U.S. Nuclear Regulatory Commission (NRC). TSTF-554 proposes changes 13 to the Standard Technical Specifications (STSs) for pressurized water reactor and boiling water 14 reactor (BWR) plants. Upon approval, these changes will be incorporated into future revisions 15 of NUREG-1430 through NUREG-1434, and NUREG-21941 and this traveler will be available to 16 licensees for adoption through the consolidated line item improvement process.

17 18 The proposed changes would revise the technical specifications (TSs) related to reactor coolant 19 system (RCS) operational leakage by adding a new Condition and Required Action when 20 pressure boundary leakage exists and revise the definition of the term LEAKAGE.

21 1U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Babcock and Wilcox Plants, NUREG-1430, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, dated April 2012 (ADAMS Accession Nos. ML12100A177 and ML12100A178, respectively).

U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, dated April 2012 (ADAMS Accession Nos. ML12100A222 and ML12100A228, respectively).

U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Combustion Engineering Plants, NUREG-1432, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, dated April 2012 (ADAMS Accession Nos. ML12102A165 and ML12102A169, respectively).

U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR/4 Plants NUREG-1433, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, dated April 2012 (ADAMS Accession Nos. ML12104A192 and ML12104A193, respectively).

U.S. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric BWR/6 Plants NUREG-1434, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, dated April 2012 (ADAMS Accession Nos. ML12104A195 and ML12104A196, respectively).

U.S. Nuclear Regulatory Commission, "Standard Technical Specifications, Westinghouse Advanced Passive 1000 (AP1000) Plants," NUREG-2194, Volume 1 Specifications, and Volume 2, Bases, Revision 0, dated April 2016 (ADAMS Accession Nos. ML16110A277 and ML16110A369, respectively).

2 1

2.0 REGULATORY EVALUATION

2 3 2.1 Reactor Coolant System Description 4

5 Components that contain or transport the coolant to or from the reactor core make up the RCS.

6 Materials can degrade as a result of the complex interaction of the materials, the stresses they 7 encounter, and through operational wear or mechanical deterioration during normal and upset 8 operating environments. Such material degradation could lead to leakage of reactor coolant 9 into containment buildings.

10 11 RCS leakage falls under two main categories - identified leakage and unidentified leakage.

12 Identifying sources of leakage is necessary for prompt identification of potentially adverse 13 conditions, assessment of safety significance of the leakage, and quick corrective action. A 14 limited amount of leakage from the reactor coolant pressure boundary (RCPB) directly into the 15 containment/drywell atmosphere is expected as the RCS and other connected systems cannot 16 be made 100 percent leak tight. This leakage is detected, located, and isolated from the 17 containment atmosphere, so as to not interfere with measurement of unexpected RCS leakage 18 detection.

19 20 Leakage from the RCPB inside the primary containment/drywell is detected by independently 21 monitored parameters, such as sump level changes and containment/drywell gaseous and 22 particulate radioactivity levels. Plant TSs identify at least two independent and diverse means 23 and/or methods of detection. The primary means of quantifying significant leakage in the 24 containment/drywell is the containment/drywell sump monitoring system. The containment 25 atmosphere particulate and gaseous radioactivity monitors are sensitive to radioactivity in any 26 RCS leakage, but do not provide a reasonably accurate means of quantifying leakage.

27 28 The containment/drywell sump monitoring system monitors the liquid collected in the sump.

29 This liquid consists of leakage from RCS, leakage from other systems inside primary 30 containment (e.g., component cooling water), and condensation of steam released from the 31 RCS or other high-temperature systems that is condensed by the containment/drywell coolers 32 and directed to the sump. The containment sump instrumentation measures the rate of liquid 33 accumulation in the sump, displays results in the main control room, and provides for an alarm 34 for high rates of liquid accumulation. The rate of liquid accumulation may be determined by 35 changes in measured level in the sump or by the time between periodic pump operation to drain 36 the sump between known sump levels.

37 38 Gaseous and/or particulate primary containment atmospheric radioactivity monitors 39 continuously monitor the containment atmosphere during reactor operation for indications of 40 leakage. The RCS contains radioactivity that, when released to the primary containment, can 41 be detected by the gaseous or particulate primary containment atmospheric radioactivity 42 monitor. Radioactivity detection systems are included for monitoring particulate and/or gaseous 43 activities because of their sensitivities and rapid responses to RCS leakage. Reactor coolant 44 radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until 45 activated corrosion products have been formed and fission products have been released from 46 fuel elements. To enhance detection capability, radioactivity alarm settings are typically set to 47 provide the most sensitive response without causing an excessive number of spurious alarms.

48 49 The safety significance of RCS leakage varies widely depending on its source, rate, and 50 duration. Therefore, detecting and monitoring RCS leakage into the containment area is 51 necessary. Separation of identified leakage from unidentified leakage, provides quantitative

3 1 information to the operators, allowing them to take corrective action should leakage occur that is 2 detrimental to the safety of the unit and the public.

3 4 2.2 Proposed Changes to Standard Technical Specifications 5

6 TSTF-554 proposes to revise the definition of "LEAKAGE" and STS [3.4.13] RCS Operational 7 Leakage. Mark-ups of the proposed changes are provided in TSTF-554 (ADAMS Accession 8 No. ML20016A233). The proposed changes would revise the requirements when pressure 9 boundary leakage is detected and add a Required Action when pressure boundary leakage is 10 identified. The proposed change is applicable to the following STSs:

11 12

  • NUREG-1430, "Standard Technical Specifications, Babcock & Wilcox Plants,"

13 Section 1.1, "Definitions", and STS 3.4.13, "RCS Operational LEAKAGE" 14

  • NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants,"

17 Section 1.1, "Definitions", and STS 3.4.13, "RCS Operational LEAKAGE" 18

19 Section 1.1, "Definitions", and STS 3.4.4, "RCS Operational LEAKAGE" 20

21 Section 1.1, "Definitions", and STS 3.4.5, "RCS Operational LEAKAGE" 22

  • NUREG-2194, "Standard Technical Specifications, Westinghouse Advanced Passive 23 1000 (AP1000) Plants," Section 1.1, "Definitions", and STS 3.4.7, "RCS Operational 24 LEAKAGE" 25 26 2.2.1 Section 1.1, Definitions 27 28 The identified LEAKAGE definition (a.2) would be revised to no longer exclude pressure 29 boundary leakage from identified leakage by deleting the phrase not to be pressure boundary 30 LEAKAGE.

31 32 The pressure boundary LEAKAGE definition [(c)] would be revised to delete the word 33 "nonisolable."2 The sentence, "LEAKAGE past seals, packing, and gaskets is not pressure 34 boundary LEAKAGE," would be relocated from the STS Bases and added to the definition.

35 36 Additionally, the LEAKAGE definition would be revised by other editorial and punctuation 37 changes.

38 39 2.2.2 Reactor Coolant System Operational Leakage, STS [3.4.13]3 40 41 The ACTIONS section of STS [3.4.13] RCS Operational LEAKAGE, would be revised to add a 42 new Condition A to isolate pressure boundary leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

43 44 In NUREGs-1430, -1431, -1432, 1433, and -1434, existing Condition B would be revised to be 45 applicable should any Action of limiting conditions for operation (LCO) [3.4.13] not be met by 46 deleting of Condition A [or B].

2 The section number is bracketed to indicate that the numbering convention varies by NUREG.

NUREGs-1430, -1431, -1432, and -2194 use c while NUREGs-1433 and -1434 use d.

3 The STS number is bracketed to indicate that the numbering convention varies by NUREG. STS 3.4.13 is used in NUREGs-1430, -1431, and -1432; STS 3.4.4 is used in NUREG-1433; STS 3.4.5 is used in NUREG-1434; STS 3.4.7 is used in NURE-2194. STS [3.4.13] will be used throughout this safety evaluation.

4 1 In NUREGs-1430, -1431, -1432, and -2194, existing Conditions A and B would be renumbered 2 as Conditions B and C to reflect new Condition A. The existing Condition B would be revised to 3 delete the condition pressure boundary leakage exists because pressure boundary leakage 4 would be addressed by the new Condition A. Finally, the Required Actions associated with 5 existing Conditions A and B would be renumbered accordingly.

6 7 In NUREGs-1433 and -1434, existing Conditions A, B, and C would be renumbered to reflect 8 new Condition A. The existing Condition C would be revised to delete to the condition pressure 9 boundary leakage exists because pressure boundary leakage would be addressed by the new 10 Condition A. Finally, the Required Actions Associated with existing Conditions A and B would 11 be renumbered accordingly.

12 13 2.2.3 STS Bases Changes 14 15 The STS Bases would be revised to reflect the changes to the STS. The Bases for 16 LCO [3.4.13] would be revised to reflect changes in the definitions of RCS operational 17 LEAKAGE. Mark-ups of the proposed changes to the Bases for LCO [3.4.13] are provided in 18 TSTF-554 (ADAMS Accession No. ML20016A233).

19 20 Additionally, the TSTF proposed Bases for the new Action A.1 of LCO [3.4.13] Condition A.

21 The Bases for Action A.1 are:

22 23 If pressure boundary LEAKAGE exists, the affected component, 24 pipe, or vessel must be isolated from the RCS by a closed manual 25 valve, closed and de-activated automatic valve, blind flange, or 26 check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. While in this condition, structural 27 integrity of the system should be considered because the 28 structural integrity of the part of the system within the isolation 29 boundary must be maintained under all licensing basis conditions, 30 including consideration of the potential for further degradation of 31 the isolated location. Normal LEAKAGE past the isolation device 32 is acceptable as it will limit RCS LEAKAGE and is included in 33 identified or unidentified LEAKAGE. This action is necessary to 34 prevent further deterioration of the RCPB.

35 36 2.3 Applicable Regulatory Requirements and Guidance 37 38 As described in the Commissions Final Policy Statement on Technical Specifications 39 Improvements for Nuclear Power Reactors, (58 Federal Register 39132, dated July 22, 1993),

40 the NRC and industry task groups for new STS recommended that improvements include 41 greater emphasis on human factors principles in order to add clarity and understanding to the 42 text of the STS, and should provide improvements to the Bases Section of TSs, which provides 43 the purpose for each requirement in the specification. The improved vendor-specific STS were 44 developed and issued by the NRC in September 1992.

45 46 Section IV, The Commission Policy, of the Final Policy Statement on TS states, in part:

47 48 The purpose of Technical Specifications is to impose those 49 conditions or limitations upon reactor operation necessary to 50 obviate the possibility of an abnormal situation or event giving rise 51 to an immediate threat to the public health and safety by

5 1 identifying those features that are of controlling importance to 2 safety and establishing on them certain conditions of operation 3 which cannot be changed without prior Commission approval.

4 5 [T]he Commission will also entertain requests to adopt portions 6 of the improved STS [(e.g., TSTF-554)], even if the licensee does 7 not adopt all STS improvements. The Commission encourages 8 all licensees who submit Technical Specification related submittals 9 based on this Policy Statement to emphasize human factors 10 principles.

11 12 In accordance with this Policy Statement, improved STS have 13 been developed and will be maintained for each NSSS [nuclear 14 steam supply system] owners group. The Commission 15 encourages licensees to use the improved STS as the basis for 16 plant-specific Technical Specifications. [I]t is the Commission 17 intent that the wording and Bases of the improved STS be used 18 to the extent practicable.

19 20 The Summary section of the Final Policy Statement on TS states, in part that:

21 22 Implementation of the Policy Statement through implementation of 23 the improved STS is expected to produce an improvement in the 24 safety of nuclear power plants through the use of more 25 operator-oriented Technical Specifications, improved Technical 26 Specification Bases, reduced action statement induced plant 27 transients, and more efficient use of NRC and industry resources.

28 29 The Final Policy Statement on TS provides the following description of the scope and the 30 purpose of the STS Bases:

31 32 Each LCO, Action, and Surveillance Requirement should have 33 supporting Bases. The Bases should at a minimum address the 34 following questions and cite references to appropriate licensing 35 documentation (e.g., Updated Final Safety Analysis Report 36 (FSAR), Topical Report) to support the Bases.

37 38 1. What is the justification for the Technical Specification, i.e.,

39 which Policy Statement criterion requires it to be in the Technical 40 Specifications?

41 42 2. What are the Bases for each LCO, i.e., why was it determined 43 to be the lowest functional capability or performance level for the 44 system or component in question necessary for safe operation of 45 the facility and, what are the reasons for the Applicability of the 46 LCO?

47 48 3. What are the Bases for each Action, i.e., why should this 49 remedial action be taken if the associated LCO cannot be met; 50 how does this Action relate to other Actions associated with the 51 LCO; and what justifies continued operation of the system or

6 1 component at the reduced state from the state specified in the 2 LCO for the allowed time period?

3 4 4. What are the Bases for each Safety Limit?

5 6 5. What are the Bases for each Surveillance Requirement and 7 Surveillance Frequency; i.e., what specific functional requirement 8 is the surveillance designed to verify? Why is this surveillance 9 necessary at the specified frequency to assure that the system or 10 component function is maintained, that facility operation will be 11 within the Safety Limits, and that the LCO will be met?

12 13 Note: In answering these questions the Bases for each number 14 (e.g., Allowable Value, Response Time, Completion Time, 15 Surveillance Frequency), state, condition, and definition (e.g.,

16 operability) should be clearly specified. As an example, a number 17 might be based on engineering judgment, past experience, or 18 PSA [probabilistic safety assessment] insights; but this should be 19 clearly stated.

20 21 The regulation under 10 CFR 50.36(a)(1) requires that:

22 23 Each applicant for a license authorizing operation of a 24 utilization facility shall include in his application proposed technical 25 specifications in accordance with the requirements of this section.

26 A summary statement of the bases or reasons for such 27 specifications, other than those covering administrative controls, 28 shall also be included in the application, but shall not become part 29 of the technical specifications.

30 31 The regulation under 10 CFR 50.36(b) requires that:

32 33 Each license authorizing operation of a utilization facility will 34 include technical specifications. The technical specifications will 35 be derived from the analyses and evaluation included in the safety 36 analysis report, and amendments thereto, submitted pursuant to 37 [10 CFR] 50.34 [Contents of applications; technical information].

38 The Commission may include such additional technical 39 specifications as the Commission finds appropriate.

40 41 The categories of items required to be in the TSs are listed in 10 CFR 50.36(c). The regulation 42 at 10 CFR 50.36(c)(2) requires that TSs include LCOs. LCOs are the lowest functional 43 capability or performance levels of equipment required for safe operation of the facility. The 44 regulation also requires that when an LCO of a nuclear reactor is not met, the licensee shall 45 shut down the reactor or follow any remedial action permitted by the TS until the condition can 46 be met.

47 48 The regulation at 10 CFR 50.36(c)(2)(ii)(B) requires that an LCO be established for a process 49 variable, design feature, or operating restriction that is an initial condition of a design basis 50 accident or transient analysis that either assumes the failure of or presents a challenge to the 51 integrity of a fission product barrier. Operational leakage is an input to containment pressure

7 1 calculations during a loss of offsite power. It is also an input to containment heat load 2 calculations for containment coolers assuring the containment bulk air temperature remains 3 below the initial temperature assumed for containment pressure response during a 4 loss-of-coolant accident evaluation. Operational leakage limits constitute an initial condition of 5 transient analyses supporting exclusion of pipe rupture accident dynamic effects from the 6 design basis. The STS contain an LCO that requires operational leakage be maintained within 7 limits associated with conditions assumed in the analyses. The unidentified leakage limit is set 8 as a fraction of the leakage rate calculated for a RCPB critical crack (i.e., a crack size 9 approaching instability with the potential for rapid growth, as determined through experimental 10 observations and application of fracture mechanics principles). The margin between the 11 unidentified leakage limit and the critical crack leak rate provides time for operator action to shut 12 down and cool down the reactor before hazardous degradation to the RCPB develops. The 13 identified leakage limit ensures that the identified leakage would be low enough to avoid 14 masking unidentified leakage and prevent increased identified leakage from causing significant 15 pressure boundary degradation. The prohibition against pressure boundary leakage ensures 16 that operation does not continue with existing significant degradation of the RCPB 17 18 The RCPB is defined in 10 CFR 50.2 as:

19 20 all those pressure-containing components of boiling and 21 pressurized water-cooled nuclear power reactors, such as 22 pressure vessels, piping, pumps, and valves, which are:

23 24 (1) Part of the reactor coolant system, or 25 26 (2) Connected to the reactor coolant system, up to and 27 including any and all of the following:

28 29 (i) The outermost containment isolation valve in 30 system piping which penetrates primary reactor 31 containment, 32 33 (ii) The second of two valves normally closed during 34 normal reactor operation in system piping which does not 35 penetrate primary reactor containment, 36 37 (iii) The reactor coolant system safety and relief valves.

38 39 For nuclear power reactors of the direct cycle boiling water type, 40 the reactor coolant system extends to and includes the outermost 41 containment isolation valve in the main steam and feedwater 42 piping.

43 44 The NRC staffs guidance for the review of TSs is in Chapter 16.0, Technical Specifications, of 45 NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for 46 Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), March 2010 (ADAMS 47 Accession No. ML100351425). As described therein, as part of the regulatory standardization 48 effort, the NRC staff has prepared STSs for each of the LWR nuclear designs. Accordingly, the 49 NRC staffs review includes consideration of whether the proposed changes are consistent with 50 the applicable referenced STS, as modified by NRC-approved travelers. In addition, the SRP 51 states that comparing the change to previous STSs can help clarify the STS intent.

8 1 Regulatory Guide (RG) 1.45, Revision 1, Guidance on Monitoring and Responding to Reactor 2 Coolant System Leakage, dated May 2008 (ADAMS Accession No. ML073200271), states:

3 4 The safety significance of leakage from the RCS can vary widely, 5 depending on the source of the leakage as well as the leakage 6 rate and duration. Operating experience and research have 7 indicated that very low levels of leakage could cause (or indicate) 8 material degradation arising, for example, as a result of boric acid 9 corrosion, primary water stress-corrosion cracking, and 10 intergranular stress-corrosion cracking. Such forms of 11 degradation could potentially compromise the integrity of a 12 system, leading to a loss-of-coolant accident. To minimize the 13 probability of rapidly propagating failure attributable to material 14 degradation and gross rupture of the RCPB, plants should keep 15 the leakage to a level that is as low as practical and take prompt 16 action in responding to leakage to limit the safety consequences.

17 18

3.0 TECHNICAL EVALUATION

19 20 3.1 LEAKAGE Definition, STS 1.1 21 22 The change to the definition of identified leakage applies to leakage from an RCS component 23 that would be released directly into the containment/drywell atmosphere where the leakage 24 would be detectable by the RCS leakage detection systems. The revised definition of identified 25 leakage removes the existing exclusion of leakage known to be pressure boundary leakage. It 26 also clarifies that identified leakage is known to not impair leakage detection system operation, 27 such as by masking other leakage. Therefore, all RCS leakage that is specifically located and 28 known to not interfere with the operation of leakage detection systems would be considered 29 identified leakage, regardless of the source of leakage.

30 31 Section B. Discussion Leakage Separation of RG 1.45 provides the following related to 32 separation between identified and unidentified leakage:

33 34 Procedures for separating the sources of leakage (i.e., leakage 35 from an identified source versus leakage from an unidentified 36 source) are necessary for prompt identification of potentially 37 adverse conditions, assessment of the safety significance of the 38 leakage, and quick corrective action.

39 40 The reactor vessel closure seals and safety and relief valves 41 should not have significant leakage; however, if leakage occurs 42 through these paths or through pump and valve seals, it should be 43 detectable and collectable, and the system should isolate it from 44 the containment atmosphere to the extent practical so as not to 45 mask any potentially serious leakage that may occur. This 46 leakage is identified leakage, and it should discharge to tanks or 47 sumps so that the plant operator can measure or calculate, 48 monitor, and analyze the flow rate and trend in flow rate during 49 plant operation.

50

9 1 Leakage to the containment atmosphere, which is not collected 2 (such as from valve stem packing glands and other sources),

3 increases the humidity of the containment. The moisture removed 4 from the atmosphere by air coolers, together with any associated 5 liquid leakage to the containment, is unidentified leakage, and 6 the system should collect it in tanks or sumps separate from the 7 identified leakage so that the plant operator can establish, 8 monitor, and analyze the flow rate and the trend in flow rate of the 9 unidentified leakage during plant operation.

10 11 Thus, the distinction between identified and unidentified leakage is the capability to collect and 12 measure identified leakage such that it does not impair the leakage detection system function to 13 monitor unidentified leakage. The source of the leakage is not relevant to this capability 14 provided that separate, appropriate limits on pressure boundary leakage have been established.

15 Therefore, the proposed change to the definition of identified leakage is acceptable.

16 17 The proposed change to the definition of pressure boundary leakage deletes the word 18 nonisolable and adds a sentence clarifying that pressure boundary leakage does not include 19 leakage past seals, packing, and gaskets. RG 1.45 defines RCPB leakage as:

20 21 leakage from a nonisolable fault in the material of an RCS component, 22 pipe wall (including welds), or vessel wall. Leakage from seals, gaskets, 23 and mechanical connections (e.g., bolts, valve seals) is not considered 24 RCPB leakage although these components are part of the RCPB, as 25 defined in 10 CFR 50.2, Definitions ... Thus, RCPB leakage is indicative 26 of degradation of pressure-retaining components that could ultimately 27 result in a loss of component structural integrity.

28 29 The word nonisolable has been interpreted inconsistently in the definition of pressure 30 boundary leakage. In some interpretations, it has been considered a means of emphasizing 31 that the leakage fault is in the base material of the pressure boundary and, therefore, the 32 leakage cannot be stopped by adjusting packing or seals. In such a case, the fault represents 33 degradation of the pressure boundary material that could result in a loss of structural integrity.

34 Another interpretation is that leakage through a fault in portions of the pressure boundary that 35 can be separated from the RCS by an isolation device (typically an installed valve) need not be 36 considered as pressure boundary leakage once the isolation device is performing its isolation 37 function. This would allow certain small sections of the RCPB between the outermost two 38 valves to be removed from consideration as RCPB leakage when the inner valve is closed.

39 40 Regardless of the interpretation, deletion of the word nonisolable does not alter the 41 fundamental meaning that pressure boundary leakage represents degradation that could 42 ultimately result in a loss of structural integrity. Therefore, the NRC staff finds that removing the 43 term nonisolable provides a clearer definition of pressure boundary leakage and does not 44 conflict with the RCPB definition in 10 CFR 50.2. Additionally, the NRC staff finds that the 45 additional sentence LEAKAGE past seals, packing, and gaskets is not pressure boundary 46 LEAKAGE, is consistent with the RG 1.45 discussion that pressure boundary leakage is not 47 leakage from seals, gaskets, or packing. The proposed punctuation changes to leakage 48 definition are considered editorial and provide further clarity to the definitions of leakage in 49 accordance with SRP Chapter 16.0. Therefore, the NRC staff finds the proposed change to the 50 definition of leakage acceptable.

51

10 1 3.2 Reactor Coolant System Operational Leakage, STS [3.4.13]

2 3 The traveler proposes to add a new Condition A to STS [3.4.13]. This condition applies if 4 pressure boundary leakage exists. The new Required Action A.1 requires isolation of the 5 affected component, pipe, or vessel from the RCS by use of a closed manual valve, closed and 6 de-activated automatic valve, blind flange, or check valve, within a Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

7 This action may only be completed when the component is located where an existing isolation 8 device or feature can be configured to provide the isolation function within the 4-hour 9 Completion Time. If the 4-hour Completion Time cannot be met, the plant must initiate 10 shutdown in accordance with existing Condition C (previously Condition B).

11 12 The proposed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for new Condition A is consistent with the existing 13 Completion Time for Condition B (previously Condition A) for when the identified or unidentified 14 leakage exceeds limit as defined in LCO [3.4.13]. If the leak cannot be isolated within the 15 allotted time period, the plant will begin shutdown activities before any significant damage to the 16 RCPB can take place. Therefore, the NRC staff finds that the 4-hour Completion Time is 17 acceptable because it is a reasonable time frame for a leak to be isolated and provides a 18 reasonable period to isolate the flaw while avoiding further damage to the RCPB 19 20 The new proposed Required Action requires the flaw to be isolated from the reactor coolant 21 pressure source to prevent further degradation of the flaw, which could result in additional 22 leakage. If the Required Action A.1 cannot be completed within the 4-hour Completion Time ,

23 STS [3.4.13] Condition C (previously Condition B) requires that the reactor be brought to lower 24 pressure conditions to reduce the severity of the LEAKAGE and its potential consequences (i.e.,

25 be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />). Therefore, the NRC staff finds that 26 proposed the new Condition A, including its associated Required Action A.1 and Completion 27 Time, is acceptable because it continues to meet the requirements of 10 CFR 50.36(c)(2)(i), by 28 providing remedial actions and shutting down the reactor if the remedial actions cannot be met.

29 30 Additionally, formatting and numbering changes were proposed to LCO [3.4.13] caused by the 31 addition of a new Condition A. These NRC staff finds these proposed changes acceptable 32 because they are editorial clarifications and do not substantively change the TS requirements.

33 34 3.3 Standard Technical Specification Bases 35 36 As discussed in Section 2.3 of this SE, the Final Policy Statement on TS describes the scope 37 and purpose of the STS Bases. It does so by listing five questions the STS Bases must 38 address. While the STS Bases as a whole must address these questions, not every question 39 will be relevant to every change to the STS Bases. The NRC staff reviewed the proposed 40 STS Bases changes in TSTF-554. The first, fourth, and fifth questions are not relevant to this 41 evaluation because the STS changes proposed in TSTF-554, as evaluated above, do not 42 affect the justification for the STSs, safety limits, or surveillance requirements.

43 44 The proposed addition of STS Bases supporting the new action statement of Condition A are 45 discussed above in Section 2.2.3. As these proposed STS Bases support a new action 46 statement, question three is relevant to the changes.

47 48 The NRC staff reviewed the proposed changes to the STS Bases. The STS Bases state that if 49 leakage is detected the licensee must isolate the affected component. In doing so, the 50 proposed addition describes the remedial actions to be taken if the associated LCO cannot be 51 met. The proposed bases also state that normal leakage past the isolation device is acceptable

11 1 because it limits RCS leakage and is included in identified or unidentified leakage. It also 2 explains that this action prevents further deterioration of the RCPB. Finally, the proposed 3 addition also explains that, while in this condition, licensees should consider the systems 4 structural integrity because structural integrity must be maintained. This justifies continued 5 operation at the reduced state from the state specified in the LCO because structural integrity of 6 the system will be maintained by isolating the leak within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and any normal leakage past 7 the isolated component would be included as identified or unidentified leakage. This means that 8 continued operation while the LCO is not met is acceptable because the structural integrity of 9 the system will be maintained since the leak will be isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and any normal 10 leakage past the isolated component will be accounted for in the identified or unidentified 11 leakage limits.

12 13 With regard to the 4-hour completion time, the NRC staff notes revised Action B.14 has an 14 identical completion time.5 The NRC staff has reviewed the information in the STS Bases about 15 the Action B.1s completion timei.e. it allows time to verify leakage rates and either identify 16 unidentified LEAKAGE or reduce LEAKAGE to within limits to prevent further deterioration of the 17 RCPB; if that cannot be done, then the reactor must be shutdown. The NRC staff finds that 18 those statements apply equally to new Action A.1s completion time. Further, the STS Bases do 19 not exclude applying those statements as the basis for Action A.1s competition time.

20 Therefore, the proposed additions to the STS Bases need not address that question with regard 21 to new Action A.1 because the STS Bases already include a relevant discussion. Finally, the 22 STS Bases already explain how the new remedial action relates to other actions associate with 23 the LCO, meaning the changes to the STS Bases need not provide an explanation. Therefore, 24 the NRC staff finds the proposed STS Bases changes adequately address Question 3 with 25 regard to the new action statement.

26 27 The other proposed changes would revise the existing STS Bases consistent with the 28 associated proposed Definition and LCO changes; renumber action statements to account for 29 the proposed new action statement; revise the basis for an action statement to reflect that fact 30 that an action has been removed; and correct typographical errors. Updating the STS Bases 31 consistent with the proposed definition and LCO changes ensures that they properly explain 32 why this LCO was chosen as the lowest functional capability or performance level for the 33 system in question. As the proposed changes do not affect the applicability of the LCO, the 34 STS Bases, therefore, address Question 2 as it relates to TSTF-554s proposed changes. In 35 addition, the correction of typographical errors improves the clarity of the STS Bases. Finally, 36 administrative changes like renumbering action statements and deleting sentences that are no 37 longer applicable ensure the STS Bases adequately describe the STS. Therefore, the NRC 38 staff finds that the proposed revisions to the STS Bases are consistent with the Final Policy 39 Statement on TS and 10 CFR 50.36.

40 41 Furthermore, the NRC staff review determined that the proposed STS Bases changes enhance 42 and/or clarify the current STS Bases, which follow the guidance of RG 1.45, 43 Revision 1. Regulatory Guide 1.45 describes acceptable methods for selecting leakage 44 detection systems. The NRC staffs review concluded that the proposed STS Bases changes 45 adhere to the Final Policy Statement and 10 CFR 50.36 specified above, and therefore the 46 changes are acceptable.

4 Note that in NUREGs-1433 and -1434 the condition under which the relevant action applies is different. It applies when unidentified LEAKAGE is not within the limit or total LEAKAGE is not within the limit. This difference does not change the analysis of the completion time justification.

5 The TSTF-554 Traveler notes that new Action A.1s completion time was chosen for consistency with revised Action B.1s completion time.

12 1

2 4.0

SUMMARY

3 4 The NRC staff finds that the changes to STS 1.1 Definition and STS [3.4.13] correctly specify 5 the lowest functional capability or performance levels of equipment required for safe operation 6 of the facility in accordance with 10 CFR 50.36(c)(2)(i). Also, the remedial actions to be taken 7 until each LCO can be met provide protection to the health and safety of the public, thereby 8 satisfying 10 CFR 50.36(c)(2)(i). Additionally, the NRC staff determined that the changes are 9 technically clear and consistent with customary terminology and format in accordance with SRP 10 Chapter 16.0.

11 12 Therefore, the NRC staff concludes that all the proposed changes in TSTF-554 are acceptable 13 and thus, approved.

14 15 Principal Reviewers: Ravinder Grover, NRR/DSS/STSB 16 Ian Young, NRR/DNRL/NPHB 17 Steve Jones, NRR/DSS/SCPB 18 19 Date:

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