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