ML19017A217
| ML19017A217 | |
| Person / Time | |
|---|---|
| Site: | Technical Specifications Task Force |
| Issue date: | 05/29/2019 |
| From: | Pete Snyder NRC/NRR/DSS/STSB |
| To: | |
| Honcharik M, NRR/DSS, 301-415-1774 | |
| Shared Package | |
| ML19017A219 | List: |
| References | |
| EPID L-2018-PMP-0002 | |
| Download: ML19017A217 (7) | |
Text
1 General Directions: This Model safety evaluation (SE) provides the format and content to be 2 used when preparing the plant-specific SE of a license amendment request to adopt TSTF-569, 3 Revision 1. The bolded bracketed information shows text that should be filled in for the specific 4 amendment; individual licensees would furnish site-specific nomenclature or values for these 5 bracketed items. The italicized wording provides guidance on what should be included in each 6 section and should not be included in the SE.
7 8 DRAFT MODEL SAFETY EVALUATION 9 BY THE OFFICE OF NUCLEAR REACTOR REGULATION 10 TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER 11 TSTF-569, REVISION 1 12 REVISE RESPONSE TIME TESTING DEFINITION 13 USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS 14 (EPID [insert EPID number])
15 16 17
1.0 INTRODUCTION
18 19 By application dated [enter date] (Agencywide Documents Access and Management System 20 (ADAMS) Accession No. [MLXXXXXXXXX]), [as supplemented by letters dated [enter 21 date(s) and ADAMS Accession Nos.)), [name of licensee] (the licensee) submitted a license 22 amendment request (LAR) for [name of facility (abbreviated name), applicable units]. The 23 amendment would revise technical specification (TS) definitions for [for Westinghouse-designed 24 plants use: engineered safety feature (ESF) response time and reactor trip system (RTS) 25 response time OR for CE-designed plants use: engineering safety feature actuation system 26 (ESFAS) response time and reactor protection system (RPS) response time] that are 27 referenced in Surveillance Requirements (SRs), hereafter referred to as response time testing 28 (RTT).
29 30 The proposed changes are based on Technical Specifications Task Force (TSTF) traveler 31 TSTF-569, Revision 1, Revise Response Time Testing Definition, dated October 31, 2018 32 (ADAMS Accession No. ML18304A026). The U.S. Nuclear Regulatory Commission (NRC or 33 the Commission) issued a final safety evaluation (SE) approving TSTF-569, Revision 1, on 34 [enter date] (ADAMS Accession No. ML19[XXXXXXX]).
35 36 [The licensee has proposed variations from the TS changes described in TSTF-569, 37 Revision 1. The variations are described in Section [2.2.1] of this SE and evaluated in 38 Section [3.3)). OR [The licensee is not proposing any variations from the TS changes 39 described in TSTF-569, Revision 1, or the applicable parts of the NRC staffs SE of 40 TSTF-569, Revision 1.))
41 42 [The supplemental letter[s] dated [enter date(s)], provided additional information that 43 clarified the application, did not expand the scope of the application as originally 44 noticed, and did not change the NRC staffs original proposed no significant hazards Enclosure 2
1 consideration determination as published in the Federal Register on [enter date] (cite FR 2 reference).]
3 4
2.0 REGULATORY EVALUATION
5 6
2.1 DESCRIPTION
OF RESPONSE TIME TESTING 7
8 The [RTS OR RPS] initiates a unit shutdown, based on the values of selected unit parameters, 9 to protect against violating the core fuel design limits and the reactor coolant system pressure 10 boundary during anticipated operational occurrences and to assist the [ESF OR ESFAS] in 11 mitigating accidents. The [ESF OR ESFAS] initiates necessary safety systems, based on the 12 values of selected unit parameters, to protect against violating core design limits and the reactor 13 coolant system pressure boundary, and to mitigate accidents.
14 15 RTT verifies that the individual channel or train actuation response times are less than or equal 16 to the maximum values assumed in the accident analyses. The RTT acceptance criteria are 17 under licensee control, in the [Technical Requirements Manual or equivalent document].
18 Individual component response times are not modeled in the accident analyses. The analyses 19 model the overall or total elapsed time, from the point at which the parameter exceeds the trip 20 setpoint value at the sensor to the point at which the equipment reaches the required functional 21 state (e.g., control and shutdown rods fully inserted in the reactor core).
22 23
2.2 PROPOSED CHANGE
S TO THE TECHNICAL SPECIFICATIONS 24 25 The licensee proposed to revise the RTT TS definitions in Section [1.1] of the TS. Specifically, 26 the proposed changes would revise the TS definitions to eliminate the requirement for prior 27 NRC review and approval of the response time verification of new pressure sensor components 28 {NOTE: this may be used interchangeably with the phrase pressure transmitter within this SE 29 due to the usage of these terms in TSTF-569, Revision 1.} and protection channel components, 30 while still requiring verification to be performed using the standard methodology contained in 31 NRC-approved TSTF-569, Revision 1, Attachment 1, Methodology to Eliminate Pressure 32 Sensor and Protection Channel (for Westinghouse Plant only) Response Time Testing. The 33 proposed change would allow the licensee to verify the response time of similar/comparable 34 component types to those components being replaced without prior NRC approval for each set 35 of different components being installed.
36 37 {NOTE: For Westinghouse-designed plants use:}
38 [The proposed change would revise the following TS definitions in Section [1.1]:
39 40
- Engineered Safety Feature (ESF) Response Time and 41
- Reactor Trip System (RTS) Response Time.
42 43 The definitions would be revised to state the following (with changes underlined):
44 45 Engineered Safety Feature (ESF) Response Time 46 47 The ESF RESPONSE TIME shall be that time interval from when the 48 monitored parameter exceeds its actuation setpoint at the channel sensor 49 until the ESF equipment is capable of performing its safety function (i.e.,
50 the valves travel to their required positions, pump discharge pressures 51 reach their required values, etc.). Times shall include diesel generator
1 starting and sequence loading delays, where applicable. The response 2 time may be measured by means of any series of sequential, overlapping, 3 or total steps so that the entire response time is measured. In lieu of 4 measurement, response time may be verified for selected components 5 provided that the components and methodology for verification have been 6 previously reviewed and approved by the NRC, and the components have 7 been evaluated in accordance with an NRC approved methodology.
8 9 Reactor Trip System (RTS) Response Time 10 11 The RTS RESPONSE TIME shall be that time interval from when the 12 monitored parameter exceeds its RTS trip setpoint at the channel sensor 13 until loss of stationary gripper coil voltage. The response time may be 14 measured by means of any series of sequential, overlapping, or total steps 15 so that the entire response time is measured. In lieu of measurement, 16 response time may be verified for selected components provided that the 17 components and methodology for verification have been previously 18 reviewed and approved by the NRC, and the components have been 19 evaluated in accordance with an NRC approved methodology.]
20 21 {NOTE: For CE-designed plants use:}
22 The proposed change would revise the following TS definitions in Section [1.1]:
23 24
- Engineered Safety Feature (ESF) Response Time and 25
- Reactor Protection System (RPS) Response Time.
26 27 The definitions would be revised to state the following (with changes underlined):
28 29 Engineered Safety Feature (ESF) Response Time 30 31 The ESF RESPONSE TIME shall be that time interval from when the 32 monitored parameter exceeds its ESF actuation setpoint at the channel 33 sensor until the ESF equipment is capable of performing its safety function 34 (i.e., the valves travel to their required positions, pump discharge 35 pressures reach their required values, etc.). Times shall include diesel 36 generator starting and sequence loading delays, where applicable. The 37 response time may be measured by means of any series of sequential, 38 overlapping, or total steps so that the entire response time is measured. In 39 lieu of measurement, response time may be verified for selected 40 components provided that the components and methodology for 41 verification have been previously reviewed and approved by the NRC, and 42 the components have been evaluated in accordance with an NRC approved 43 methodology.
44 45 Reactor Protection System (RPS) Response Time 46 47 The RPS RESPONSE TIME shall be that time interval from when the 48 monitored parameter exceeds its RPS trip setpoint at the channel sensor 49 until electrical power to the CEAs drive mechanism is interrupted. The 50 response time may be measured by means of any series of sequential, 51 overlapping, or total steps so that the entire response time is measured. In
1 lieu of measurement, response time may be verified for selected 2 components provided that the components and methodology for 3 verification have been previously reviewed and approved by the NRC, and 4 the components have been evaluated in accordance with an NRC approved 5 methodology.]
6 7 The proposed change would be supported by changes to the TS Bases. Similar to the RTT 8 definitions, the Bases would state that for components that have been evaluated in accordance 9 with a methodology approved by the NRC, the response time can be verified in lieu of being 10 measured. The proposed change would revise the Bases to be consistent with the proposed 11 definition change.
12 13 [2.2.1 Variations from TSTF-569, Revision 1 14 15 Insert description of any variations here.]
16 17 2.3 APPLICABLE REGULATORY REQUIREMENTS AND GUIDANCE 18 19 Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) requires each 20 applicant for a license authorizing operation of a utilization facility to include in the application 21 proposed TSs.
22 23 The regulation at 10 CFR 50.36(b) states that:
24 25 The technical specifications will be derived from the analyses and evaluation 26 included in the safety analysis report, and amendments thereto, submitted 27 pursuant to [10 CFR] 50.34 [Contents of applications; technical information].
28 The Commission may include such additional technical specifications as the 29 Commission finds appropriate.
30 31 The regulation at 10 CFR 50.40(a) states, in part, that the TSs shall provide reasonable 32 assurance that the health and safety of the public will not be endangered.
33 34 Appendix A to 10 CFR Part 50 provides General Design Criteria (GDC) for nuclear power 35 plants. Plant-specific design criteria are described in the plants Updated Final Safety Analysis 36 Report (UFSAR).
37 The regulation at 10 CFR Part 50, Appendix A, GDC 13, Instrumentation and Control, states:
38 39 Instrumentation shall be provided to monitor variables and systems over their 40 anticipated ranges for normal operation, for anticipated operational occurrences, 41 and for accident conditions as appropriate to assure adequate safety, including 42 those variables and systems that can affect the fission process, the integrity of 43 the reactor core, the reactor coolant pressure boundary, and the containment 44 and its associated systems. Appropriate controls shall be provided to maintain 45 these variables and systems within prescribed operating ranges.
46
1 The regulation at 10 CFR Part 50, Appendix A, GDC 21, Protection System Reliability and 2 Testability, states:
3 4 The protection system shall be designed for high functional reliability and 5 inservice testability commensurate with the safety functions to be performed.
6 Redundancy and independence designed into the protection system shall be 7 sufficient to assure that (1) no single failure results in loss of the protection 8 function and (2) removal from service of any component or channel does not 9 result in loss of the required minimum redundancy unless the acceptable 10 reliability of operation of the protection system can be otherwise demonstrated.
11 The protection system shall be designed to permit periodic testing of its 12 functioning when the reactor is in operation, including a capability to test 13 channels independently to determine failures and losses of redundancy that may 14 have occurred.
15 16 The NRC staffs guidance for the review of TSs is in Chapter 16.0, Revision 3, Technical 17 Specifications, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis 18 Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), March 2010 19 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory 20 standardization effort, the NRC staff has prepared Standard Technical Specifications (STS) for 21 each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of 22 whether the proposed changes are consistent with the applicable reference STS, as modified by 23 NRC-approved travelers. The STS applicable to [abbreviated name of facility] is 24 25 {NOTE: Choose applicable STS}
26 [NUREG-1431, Revision 4.0, Standard Technical Specifications, Westinghouse Plants, 27 April 2012, Volume 1, Specifications (ADAMS Accession No. ML12100A222), and 28 Volume 2, Bases (ADAMS Accession No. ML12100A228).
29 30 NUREG-1432, Revision 4.0, Standard Technical Specifications, Combustion Engineering 31 Plants, April 2012, Volume 1, Specifications (ADAMS Accession No. ML12102A165),
32 and Volume 2, Bases (ADAMS Accession No. ML12102A169).]
33 34 Regulatory Guide (RG) 1.118, Revision 3, Periodic Testing of Electric Power and Protection 35 Systems, April 1995 (ADAMS Accession No. ML003739468), endorses the Institute of 36 Electrical and Electronics Engineers, Inc. (IEEE) Std. 338-1987, IEEE Standard Criteria for the 37 Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems, which 38 was approved on March 3, 1988, by the American National Standards Institute.
39 40 Branch Technical Position (BTP) 7-17, Guidance on Self-Test and Surveillance Test 41 Provisions, August 23, 2016 (ADAMS Accession No. ML16019A316), states, in part:
42 43 Failures detected by hardware, software, and surveillance testing should be 44 consistent with the failure detectability assumptions of the single-failure analysis 45 and the failure modes and effects analysis.
46
1
3.0 TECHNICAL EVALUATION
2 3
3.1 PROPOSED CHANGE
S TO THE RESPONSE TIME TESTING DEFINITION 4
5 The proposed change to TS Section [1.1] would eliminate required direct measurement RTT for 6 selected [For CE plants: pressure transmitter/sensor OR for Westinghouse plants: pressure 7 transmitter/sensor and protection channel] components but does not eliminate required 8 surveillance testing for the entirety of an instrument channel or the system as a whole (e.g.,
9 RTS). Therefore, the NRC staff finds that the proposed change is consistent with the 10 surveillance testing requirements of 10 CFR 50.36.
11 12 The NRC staff confirmed that the proposed change has no effect on the design, fabrication, use, 13 or methods of testing of the instrumentation and will not affect the ability of the instrumentation 14 to perform the functions assumed in the safety analysis. Therefore, compliance with the 15 [design criteria GDC 13 and GDC 21 OR plant-specific criteria] is not affected.
16 17 RG 1.118, Revision 3, describes acceptable methods for complying with NRC regulations 18 pertaining to periodic testing of protection systems and power systems. TSTF-569, Revision 1, 19 states the following regarding applicable design criteria:
20 21 Clause 6.3.4 of IEEE 338-1987, "Criteria for the Periodic Surveillance Testing of 22 Nuclear Power Generating Station Safety Systems," states response time testing 23 shall be required only on safely systems or subsystems to verify that the 24 response times are within the limits given in the Safety Analysis Report including 25 Technical Specifications. Response time testing of all safety-related equipment is 26 not required if, in lieu of response time testing, the response time of safety 27 system equipment is verified by functional testing, calibration checks, or other 28 tests, or both. This is acceptable if it can be demonstrated that changes in 29 response time beyond acceptable limits are accompanied by changes in 30 performance characteristics that are detectable during routine periodic tests.
31 32 Section 5.3.4, "Response time verification tests," of IEEE Standard 338-2012, 33 "IEEE Standard for Criteria for the Periodic Surveillance Testing of Nuclear 34 Power Generating Station Safety Systems," Item c) states response time testing 35 of all safety-related equipment is not required if, in lieu of response time testing, 36 the response time of safety system equipment is verified by functional testing, 37 calibration checks, or other tests. This is acceptable if it can be demonstrated 38 that changes in response time beyond acceptable limits are accompanied by 39 changes in performance characteristics that are detectable during routine 40 periodic tests.
41 42 The traveler states that system operation, design basis, and capability for testing will remain 43 unchanged as the replacement components comply with these design criteria. The NRC staff 44 found that the traveler provided an adequate technical basis and that replacement components 45 can continue to perform the same design functions as the original components. The NRC staff 46 found that the methodologies contained in Attachment 1 to the traveler provide adequate criteria 47 for ensuring that replacement components degraded response time issues or failures would be 48 captured. Therefore, conformance with IEEE 338-2012 and 338-1987 design criteria is not 49 affected, since the licensee is adopting TSTF-569, Revision 1.
50
1 3.2
SUMMARY
2 3 The NRC staff reviewed the proposed changes against the regulations and determined that, 4 with the proposed changes, the TS will continue to meet the requirements of 10 CFR 50.36(b) 5 and, consistent with 10 CFR 50.40, will continue to provide reasonable assurance that the 6 health and safety of the public will not be endangered. Additionally, the NRC staff determined 7 that the proposed changes are technically clear and consistent with customary terminology and 8 format in accordance with SRP Chapter 16.0. Therefore, the NRC staff concludes that the 9 proposed changes are acceptable.
10 11 [3.3 VARIATIONS FROM TSTF-569, REVISION 1 12 13 The licensee described variations from TSTF-569, Revision 1, in Section [2.2] of the LAR.
14 The licensee provided justification for the proposed variations. The NRC staff reviewed 15 the justifications and determined that the variations are [not] acceptable because.
16 17 The [Name of facilitys] TSs utilize different [numbering][and][titles] than the STS on 18 which TSTF-569, Revision 1, was based. The NRC staff determined that these differences 19 are editorial and do not affect the applicability of TSTF-569, Revision 1, to the proposed 20 LAR.]
21 22
4.0 STATE CONSULTATION
23 24 {This section is to be prepared by the plant project manager.}
25 26 In accordance with the Commissions regulations, the [Name of State] State official was notified 27 of the proposed issuance of the amendment(s) on [date]. The State official had [no]
28 comments. [If comments were provided, they should be addressed here.]
29 30
5.0 ENVIRONMENTAL CONSIDERATION
31 32 {This section is to be prepared by the plant project manager in accordance with current 33 procedures.}
34 35
6.0 CONCLUSION
36 37 {This section is to be prepared by the plant project manager.}
38 39 The Commission has concluded, based on the considerations discussed above, that: (1) there 40 is reasonable assurance that the health and safety of the public will not be endangered by 41 operation in the proposed manner, (2) there is reasonable assurance that such activities will be 42 conducted in compliance with the Commissions regulations, and (3) the issuance of the 43 amendment(s) will not be inimical to the common defense and security or to the health and 44 safety of the public.
45 46 Principal Contributors: [PM Name, NRR/DORL 47 C. Tilton, NRR/DSS]
48 49 Date: May 29, 2019