ML19255J324
| ML19255J324 | |
| Person / Time | |
|---|---|
| Site: | Technical Specifications Task Force |
| Issue date: | 10/24/2019 |
| From: | Victor Cusumano NRC/NRR/DSS/STSB |
| To: | |
| Honcharik M, NRR/DSS, 301-415-1774 | |
| Shared Package | |
| ML19253A044 | List: |
| References | |
| EPID L-2019-PMP-0178, TSTF-541, Rev 2 | |
| Download: ML19255J324 (20) | |
Text
Enclosure 1 DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1
TECHNICAL SPECIFICATIONS TASK FORCE TRAVELER 2
TSTF-541, REVISION 2 3
ADD EXCEPTIONS TO SURVEILLANCE REQUIREMENTS FOR VALVES 4
AND DAMPERS LOCKED IN THE ACTUATED POSITION 5
USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS 6
(EPID L-2019-PMP-0178) 7 8
1.0 INTRODUCTION
9 10 By letter dated August 28, 2019 (Agencywide Documents Access and Management System 11 (ADAMS) Accession No. ML19240A315), the Technical Specifications Task Force (TSTF) 12 submitted to the U.S. Nuclear Regulatory Commission (NRC) Traveler TSTF-541, Revision 2, 13 Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated 14 Position. Traveler TSTF-541, Revision 2, proposes changes to the Standard Technical 15 Specifications (STS) for Babcock & Wilcox (B&W), Westinghouse, Combustion Engineering 16 (CE), and General Electric (GE) plant designs. These changes would be incorporated into 17 future revisions of NUREG-1430, NUREG-1431, NUREG-1432, NUREG-1433, and 18 NUREG-1434.1 This traveler would be made available to licensees for adoption through the 19 consolidated line item improvement process.
20 21 The proposed changes would revise certain Surveillance Requirements (SRs) in the STS by 22 adding an exception to the SRs for automatic valves or dampers that are locked, sealed, or 23 otherwise secured in the actuated position.
24 25 1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Babcock and Wilcox Plants, NUREG-1430, Volume 1, Specifications, and Volume 2, Bases, Revision 4.0, 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, 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, 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, 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, April 2012 (ADAMS Accession Nos. ML12104A195 and ML12104A196, respectively).
1.1 Reason for the Proposed Change 1
2 As described in the Commissions Final Policy Statement on Technical Specifications 3
Improvements for Nuclear Power Reactors published in the Federal Register on July 22, 1993 4
(58 FR 39132), the NRC and industry task groups for new STS recommend that improvements 5
include greater emphasis on human factors principles in order to add clarity and understanding 6
to the text of the STS, and provide improvements to the Bases of STS, which provides the 7
purpose for each requirement in the specification. The improved vendor-specific STS were 8
developed and issued by the NRC in September 1992.
9 10 NUREG-1430 through 1434 contain the NRC staffs guidance for one method the NRC staff 11 finds acceptable to comply with the requirements in Section 50.36 of Title 10 of the Code of 12 Federal Regulations (10 CFR) for B&W, Westinghouse, CE, and GE plant designs. A defined 13 term common to NUREG-1430 through 1434 is OPERABLE - OPERABILITY which means:
14 15 A system, subsystem, [train/division], component, or device shall be OPERABLE 16 or have OPERABILITY when it is capable of performing its specified safety 17 function(s) and when all necessary attendant instrumentation, controls, normal or 18 emergency electrical power, cooling and seal water, lubrication, and other 19 auxiliary equipment that are required for the system, subsystem, [train/division],
20 component, or device to perform its specified safety function(s) are also capable 21 of performing their related support function(s).
22 23 In the STSs, Limiting Conditions for Operation (LCOs) are generally expressed in statements 24 such as Two trains of the X System shall be OPERABLE. The OPERABLE - OPERABILITY 25 definition is used to evaluate whether an LCO is met. To determine which systems, 26 subsystems, trains/divisions, components, or devices might have their operability affected by a 27 given structure, system, or component (SSC), knowledge of whether the SSC is required for the 28 system, subsystem, train/division, component, or device to perform its specified safety 29 function(s) is required.
30 31 STS LCO 3.0.1 through LCO 3.0.9 establish the rules of usage applicable to all Specifications 32 and apply at all times, unless otherwise stated. STS LCO 3.0.2 establishes that upon discovery 33 of a failure to meet an LCO, the associated Required Actions shall be met. The Required 34 Actions establish those remedial measures that must be taken within specified Completion 35 Times when the requirements of an LCO are not met.
36 37 The STS SRs 3.0.1 through 3.0.4 establish the rules of usage for SRs and apply at all times, 38 unless otherwise stated. SR 3.0.1 establishes the requirement that SRs must be met during the 39 MODES or other specified conditions in the Applicability for which the requirements of the LCO 40 apply, unless otherwise specified in the individual SRs. This usage rule ensures that 41 Surveillances are performed to verify the OPERABILITY of systems and components, and that 42 variables are within specified limits. STS SR 3.0.1 states:
43 44 SRs shall be met during the MODES or other specified conditions in the 45 Applicability for individual LCOs, unless otherwise stated in the SR. Failure to 46 meet a Surveillance, whether such failure is experienced during the performance 47 of the Surveillance or between performances of the Surveillance, shall be failure 48 to meet the LCO. Failure to perform a Surveillance within the specified 49 Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.
50 Surveillances do not have to be performed on inoperable equipment or variables 1
outside specified limits.
2 3
For SRs lacking an explicit exception, the sentence Failure to meet a Surveillance, whether 4
such failure is experienced during the performance of the Surveillance or between 5
performances of the Surveillance, shall be failure to meet the LCO, requires that when an SR is 6
not met, the LCO is not met. Per the usage rules, when an LCO is not met, Required Actions 7
must be met within specified Completion Times.
8 9
For some cases, an individual SSC may not be capable of meeting an SR, but the system, 10 subsystem, or train/division to which it belongs may still be capable of performing its specified 11 safety function. In these cases, declaring the LCO not met may not be necessary because the 12 system, subsystem, or train/division to which the SSC belongs may be OPERABLE. The 13 current version of the STS contains explicit exceptions in the text of a limited number of SRs to 14 avoid unnecessarily declaring the LCO not met when an SSC is not capable of meeting an SR 15 but the system, subsystem, or train/division to which it belongs is still capable of performing its 16 specified safety function.
17 18 The TSTF reviewed the STS and identified SRs that do not have explicit exceptions but for 19 which exceptions would be appropriate to avoid unnecessary entry into Conditions and 20 Required Actions. The TSTF proposed the changes described in Section 2.4 based on this 21 review. The NRC staff deems the attempt to clarify its current guidance on acceptable methods 22 to meet 10 CFR 50.36, given the situation created by current STS rules, worthwhile since the 23 resulting clarification of a licensees licensing basis aligns with the intent of the Commission as 24 discussed in the Final Policy Statement on TS Improvements for Nuclear Power Reactors.
25 Specifically, one of the expectations for the implementation of the STS is to reduce action 26 statement induced plant transients.
27 28 Since 2008, the TSTF and NRC staff have been collaborating to develop an acceptable 29 approach to providing exceptions for the situation described above. By letter dated 30 October 14, 2008 (ADAMS Accession No. ML082880503), the TSTF submitted TSTF-512, 31 Revision 0, Revise SR 3.0.3 to Address SRs that Cannot be Performed or are Not Met, which 32 proposed changes that the NRC staff found unacceptable, as documented in the staffs letter 33 dated May 1, 2009 (ADAMS Accession No. ML090230254). The initial revision of TSTF-541 34 (ADAMS Accession No. ML13253A390) was submitted for NRC staff review in 2013. Due to 35 the lack of NRC staff resources during the response to Fukushima-related issues, the review 36 was delayed until 2015. Upon review of the initial version of TSTF-541, the staff had questions 37 regarding the acceptability of the approach. The NRC staff provided requests for additional 38 information (RAIs) to the TSTF by letters dated August 13, 2015 (ADAMS Accession 39 No. ML15208A287), and February 25, 2016 (ADAMS Accession No. ML16012A427).
40 Revision 2 of TSTF-541 was developed based on TSTF and NRC staff interaction through a 41 series of public meetings; the most recent of which was on February 21, 2019 (ADAMS 42 Package Accession No. ML19056A435) 43 44
2.0 REGULATORY EVALUATION
45 46 2.1 System Descriptions 47 48 The STS use generic nomenclature for systems that may go by different names at an actual 49 plant; however, regardless of the specific names, the functions of the systems are similar. The 50 text below provides a high-level description of the systems affected by the proposed change as 51 they are named in the respective STS.
1 2
For NUREG-1430, B&W Plants:
3 4
The spray additive system is a subsystem of the containment spray system that assists in 5
reducing the iodine fission product inventory in the containment atmosphere resulting from a 6
design-basis accident (DBA). In the event of an accident such as a loss-of-coolant accident 7
(LOCA), the spray additive system will be automatically actuated upon a high containment 8
pressure signal by the engineered safety features actuation system (ESFAS). The purpose of 9
SR 3.6.7.4 is to verify that each automatic valve in the spray additive system flow path actuates 10 to its correct position upon receipt of an actual or simulated actuation signal.
11 12 The emergency ventilation system (EVS) filters air from the area of the active emergency core 13 cooling system (ECCS) components during the recirculation phase of a LOCA. Ductwork, 14 valves or dampers, and instrumentation also form part of the system. During emergency 15 operations, the EVS dampers are realigned, and fans are started to begin filtration. Upon 16 receipt of the actuation signal(s), normal air discharges from the negative pressure area are 17 isolated, and the stream of ventilation air discharges through the system filter trains. The 18 prefilters remove any large particles in the air, and any entrained water droplets present, to 19 prevent excessive loading of the high-efficiency particulate air (HEPA) filters and charcoal 20 adsorbers. The purpose of SR 3.7.12.3 is to verify proper actuation of all train components, 21 including dampers, on an actual or simulated actuation signal. The purpose of SR 3.7.12.5 is to 22 ensure that the system is functioning properly by operating the EVS filter bypass damper.
23 24 The fuel storage pool ventilation system (FSPVS) provides negative pressure in the fuel storage 25 area, and filters airborne radioactive particulates from the area of the fuel pool following a fuel 26 handling accident. The FSPVS consists of portions of the normal fuel handling area ventilation 27 system (FHAVS), the station EVS, ductwork bypasses, and dampers. The portion of the normal 28 FHAVS used by the FSPVS consists of ducting between the spent fuel pool and the normal 29 FHAVS exhaust fans or dampers, and redundant radiation detectors installed close to the 30 suction end of the FHAVS exhaust fan ducting. The purpose of SR 3.7.13.3 is to verify proper 31 actuation of all train components, including dampers, on an actual or simulated actuation signal.
32 The purpose of SR 3.7.13.5 is to ensure that the system is functioning properly by operating the 33 FSPVS filter bypass damper.
34 35 The control room emergency ventilation system (CREVS) provides a protected environment 36 from which occupants can control the unit following an uncontrolled release of radioactivity, 37 hazardous chemicals, or smoke. The purpose of SR 3.7.10.3 is to verify that each 38 train/subsystem starts and operates on an actual or simulated actuation signal.
39 40 For NUREG-1431, Westinghouse Plants:
41 42 The control room emergency filtration system (CREFS) provides a protected environment from 43 which occupants can control the unit following an uncontrolled release of radioactivity, 44 hazardous chemicals, or smoke. The purpose of SR 3.7.10.3 is to verify that each 45 train/subsystem starts and operates on an actual or simulated actuation signal.
46 47 The shield building air cleanup system (SBACS) is required to ensure that radioactive materials 48 that leak from the primary containment into the shield building (secondary containment) 49 following a DBA are filtered and adsorbed prior to exhausting to the environment. The 50 containment has a secondary containment called the shield building, which is a concrete 51 structure that surrounds the steel primary containment vessel. Between the containment vessel 1
and the shield building inner wall is an annular space that collects any containment leakage that 2
may occur following a LOCA. The SBACS establishes a negative pressure in the annulus 3
between the shield building and the steel containment vessel. Filters in the system then control 4
the release of radioactive contaminants to the environment. The SBACS consists of two 5
separate and redundant trains. Each train includes a heater, cooling coils, a prefilter, moisture 6
separators, a HEPA filter, an activated charcoal adsorber section for removal of radioiodine, and 7
a fan. Ductwork, valves and/or dampers, and instrumentation also form part of the system. The 8
system initiates and maintains a negative air pressure in the shield building by means of filtered 9
exhaust ventilation of the shield building following receipt of a safety injection signal. The 10 purpose of SR 3.6.13.3 is to verify proper actuation of all train components, including dampers, 11 on an actual or simulated actuation signal. The purpose of SR 3.6.13.4 is to ensure that the 12 system is functioning properly by operating the filter bypass damper.
13 14 The iodine cleanup system (ICS) is provided to reduce the concentration of fission products 15 released to the containment atmosphere following a postulated accident. The ICS would 16 function together with the containment spray and cooling systems following a DBA to reduce the 17 potential release of radioactive material, principally iodine, from the containment to the 18 environment. The ICS consists of two 100-percent capacity, separate, independent, and 19 redundant trains. Each train includes a heater, cooling coils, a prefilter, a demister, a HEPA 20 filter, an activated charcoal adsorber section for removal of radioiodine, and a fan. Ductwork, 21 valves and/or dampers, and instrumentation also form part of the system. The system initiates 22 filtered recirculation of the containment atmosphere following receipt of a safety injection signal.
23 The purpose of SR 3.6.11.3 is to verify proper actuation of all train components, including 24 dampers, on an actual or simulated actuation signal. The purpose of SR 3.6.11.4 is to ensure 25 that the system is functioning properly by operating the ICS filter bypass damper.
26 27 The emergency core cooling system pump room exhaust air cleanup system (ECCS PREACS),
28 in conjunction with other normally operating systems, also provides environmental control of 29 temperature and humidity in the ECCS pump room area and the lower reaches of the auxiliary 30 building. Ductwork, valves or dampers, and instrumentation also form part of the system, as 31 well as demisters functioning to reduce the relative humidity of the air stream. During 32 emergency operations, the ECCS PREACS dampers are realigned, and fans are started to 33 begin filtration. Upon receipt of the actuating ESFAS signal(s), normal air discharges from the 34 ECCS pump room isolate, and the stream of ventilation air discharges through the system filter 35 trains. The prefilters or demisters remove any large particles in the air, and any entrained water 36 droplets present, to prevent excessive loading of the HEPA filters and charcoal adsorbers. The 37 purpose of SR 3.7.12.3 is to verify proper actuation of all train components, including dampers, 38 on an actual or simulated actuation signal. The purpose of SR 3.7.12.5 is to ensure that the 39 system is functioning properly by operating the ECCS PREACS filter bypass damper.
40 41 The fuel building air cleanup system (FBACS) filters airborne radioactive particulates from the 42 area of the fuel pool following a fuel handling accident or LOCA. The FBACS, in conjunction 43 with other normally operating systems, also provides environmental control of temperature and 44 humidity in the fuel pool area. The FBACS consists of two independent and redundant trains.
45 Each train consists of a heater, a prefilter or demister, a HEPA filter, an activated charcoal 46 adsorber section for removal of gaseous activity (principally iodines), and a fan. Ductwork, 47 valves or dampers, and instrumentation also form part of the system, as well as demisters, 48 functioning to reduce the relative humidity of the airstream. The system initiates filtered 49 ventilation of the fuel handling building following receipt of a high-radiation signal. The FBACS 50 is a standby system, parts of which may also be operated during normal plant operations. Upon 51 receipt of the actuating signal, normal air discharges from the building, the fuel handling building 1
is isolated, and the stream of ventilation air discharges through the system filter trains. The 2
purpose of SR 3.7.13.3 is to verify proper actuation of all train components, including dampers, 3
on an actual or simulated actuation signal. The purpose of SR 3.7.13.5 is to ensure that the 4
system is functioning properly by operating the FBACS filter bypass damper.
5 6
The penetration room exhaust air cleanup system (PREACS) filters air from the penetration 7
area between containment and the auxiliary building. The PREACS consists of two 8
independent and redundant trains. Each train consists of a heater, a prefilter or demister, a 9
HEPA filter, an activated charcoal adsorber section for removal of gaseous activity (principally 10 iodines), and a fan. Ductwork, valves or dampers, and instrumentation, as well as demisters, 11 functioning to reduce the relative humidity of the air stream, also form part of the system. The 12 PREACS is a standby system, parts of which may also operate during normal unit operations.
13 Upon receipt of the actuating signal(s), the PREACS dampers are realigned and fans are 14 started to initiate filtration. The purpose of SR 3.7.14.3 is to verify proper actuation of all train 15 components, including dampers, on an actual or simulated actuation signal. The purpose of 16 SR 3.7.14.5 is to ensure that the system is functioning properly by operating the PREACS filter 17 bypass damper.
18 19 For NUREG-1432, CE Plants:
20 21 The control room emergency air cleanup system (CREACS) provides a protected environment 22 from which occupants can control the unit following an uncontrolled release of radioactivity, 23 hazardous chemicals, or smoke. The purpose of SR 3.7.11.3 is to verify that each 24 train/subsystem starts and operates on an actual or simulated actuation signal.
25 26 The shield building exhaust air cleanup system (SBEACS) is required to ensure that radioactive 27 materials that leak from the primary containment into the shield building (secondary 28 containment) following a DBA are filtered and adsorbed prior to exhausting to the environment.
29 The containment has a secondary containment called the shield building, which is a concrete 30 structure that surrounds the steel primary containment vessel. Between the containment vessel 31 and the shield building inner wall is an annular space that collects any containment leakage that 32 may occur following a LOCA. The SBEACS establishes a negative pressure in the annulus 33 between the shield building and the steel containment vessel. Filters in the system then control 34 the release of radioactive contaminants to the environment. The SBEACS consists of two 35 separate and redundant trains. Each train includes a heater, cooling coils, a prefilter, moisture 36 separators, a HEPA filter, an activated charcoal adsorber section for removal of radioiodine, and 37 a fan. Ductwork, valves and/or dampers, and instrumentation also form part of the system. The 38 system initiates and maintains a negative air pressure in the shield building by means of filtered 39 exhaust ventilation of the shield building following receipt of a safety injection signal. The 40 purpose of SR 3.6.8.3 is to verify proper actuation of all train components, including dampers, 41 on an actual or simulated actuation signal. The purpose of SR 3.6.8.4 is to ensure that the 42 system is functioning properly by operating the filter bypass damper.
43 44 The ICS is provided to reduce the concentration of fission products released to the containment 45 atmosphere following a postulated accident. The ICS would function together with the 46 containment spray and cooling systems following a DBA to reduce the potential release of 47 radioactive material, principally iodine, from the containment to the environment. The ICS 48 consists of two 100-percent capacity, separate, independent, and redundant trains. Each train 49 includes a heater, cooling coils, a prefilter, a demister, a HEPA filter, an activated charcoal 50 adsorber section for removal of radioiodine, and a fan. Ductwork, valves and/or dampers, and 51 instrumentation also form part of the system. The system initiates filtered recirculation of the 1
containment atmosphere following receipt of a containment isolation actuation signal. The 2
purpose of SR 3.6.10.3 is to verify proper actuation of all train components, including dampers, 3
on an actual or simulated actuation signal. The purpose of SR 3.6.10.4 is to ensure that the 4
system is functioning properly by operating the ICS filter bypass damper.
5 6
The ECCS PREACS, in conjunction with other normally operating systems, also provides 7
environmental control of temperature and humidity in the ECCS pump room area and the lower 8
reaches of the auxiliary building. Ductwork, valves or dampers, and instrumentation also form 9
part of the system, as well as demisters functioning to reduce the relative humidity of the air 10 stream. During emergency operations, the ECCS PREACS dampers are realigned, and fans 11 are started to begin filtration. Upon receipt of the actuating ESFAS signal(s), normal air 12 discharges from the ECCS pump room isolate, and the stream of ventilation air discharges 13 through the system filter trains. The prefilters or demisters remove any large particles in the air, 14 and any entrained water droplets present, to prevent excessive loading of the HEPA filters and 15 charcoal adsorbers. The purpose of SR 3.7.13.3 is to verify proper actuation of all train 16 components, including dampers, on an actual or simulated actuation signal. The purpose of 17 SR 3.7.13.5 is to ensure that the system is functioning properly by operating the ECCS 18 PREACS filter bypass damper.
19 20 The FBACS filters airborne radioactive particulates from the area of the fuel pool following a fuel 21 handling accident or LOCA. The FBACS, in conjunction with other normally operating systems, 22 also provides environmental control of temperature and humidity in the fuel pool area. The 23 FBACS consists of two independent and redundant trains. Each train consists of a heater, a 24 prefilter or demister, a HEPA filter, an activated charcoal adsorber section for removal of 25 gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and 26 instrumentation also form part of the system, as well as demisters, functioning to reduce the 27 relative humidity of the airstream. The system initiates filtered ventilation of the fuel handling 28 building following receipt of a high-radiation signal. The FBACS is a standby system, parts of 29 which may also be operated during normal plant operations. Upon receipt of the actuating 30 signal, normal air discharges from the building, the fuel handling building is isolated, and the 31 stream of ventilation air discharges through the system filter trains. The purpose of SR 3.7.14.3 32 is to verify proper actuation of all train components, including dampers, on an actual or 33 simulated actuation signal. The purpose of SR 3.7.14.5 is to ensure that the system is 34 functioning properly by operating the FBACS filter bypass damper.
35 36 The PREACS filters air from the penetration area between containment and the auxiliary 37 building. The PREACS consists of two independent and redundant trains. Each train consists 38 of a heater, a prefilter or demister, a HEPA filter, an activated charcoal adsorber section for 39 removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and 40 instrumentation, as well as demisters, functioning to reduce the relative humidity of the air 41 stream, also form part of the system. The PREACS is a standby system, parts of which may 42 also operate during normal unit operations. Upon receipt of the actuating signal(s), the 43 PREACS dampers are realigned and fans are started to initiate filtration. The purpose of 44 SR 3.7.15.3 is to verify proper actuation of all train components, including dampers, on an 45 actual or simulated actuation signal. The purpose of SR 3.7.15.5 is to ensure that the system is 46 functioning properly by operating the PREACS filter bypass damper.
47 48 The essential chilled water (ECW) system provides a heat sink for the removal of process and 49 operating heat from selected safety-related air handling systems during a DBA or transient. The 50 ECW system is a closed-loop system consisting of two independent trains. Each 100-percent 51 capacity train includes a heat exchanger, surge tank, pump, chemical addition tank, piping, 1
valves, controls, and instrumentation. An independent 100-percent capacity chilled water 2
refrigeration unit cools each train. The ECW system is actuated on a safety injection actuation 3
signal and supplies chilled water to the heating, ventilation, and air conditioning units in 4
engineered safety feature equipment areas (e.g., the main control room, electrical equipment 5
room, and safety injection pump area). The purpose of SR 3.7.10.2 is to verify proper automatic 6
operation of the ECW system components and that the ECW pumps will start in the event of any 7
accident or transient that generates a safety injection actuation signal. This SR also ensures 8
that each automatic valve in the flow paths actuates to its correct position on an actual or 9
simulated safety injection actuation signal.
10 11 For NUREG-1433, GE BWR/4 Plants:
12 13 The main control room environmental control (MCREC) provides a protected environment from 14 which occupants can control the unit following an uncontrolled release of radioactivity, 15 hazardous chemicals, or smoke. The purpose of SR 3.7.4.3 is to verify that each 16 train/subsystem starts and operates on an actual or simulated actuation signal.
17 18 The ECCS is designed to limit the release of radioactive materials to the environment following 19 a LOCA and consists of the high-pressure coolant injection system, the core spray system, the 20 low-pressure coolant injection mode of the residual heat removal (RHR) system, and the 21 automatic depressurization system. The purpose of SR 3.5.1.10 is to verify the automatic 22 initiation logic of high-pressure coolant injection, core spray, and low-pressure coolant injection 23 will cause the systems or subsystems to operate as designed, including actuation of the system 24 throughout its emergency operating sequence, automatic pump startup, and actuation of all 25 automatic valves to their required positions on receipt of an actual or simulated actuation signal.
26 27 The function of the reactor core isolation cooling (RCIC) system is to respond to transient 28 events by providing makeup coolant to the reactor. The purpose of SR 3.5.3.5 is to verify the 29 system operates as designed, including actuation of the system throughout its emergency 30 operating sequence; that is, automatic pump startup and actuation of all automatic valves to 31 their required positions on receipt of an actual or simulated actuation signal.
32 33 The plant service water (PSW) system and ultimate heat sink are designed to provide cooling 34 water for the removal of heat from equipment, such as the diesel generators, RHR pump 35 coolers and heat exchangers, and room coolers for ECCS equipment, required for a safe 36 reactor shutdown following a DBA or transient. The PSW system also provides cooling to unit 37 components, as required, during normal shutdown and reactor isolation modes. During a DBA, 38 the equipment required only for normal operation is isolated and cooling is directed to only 39 safety-related equipment. The purpose of SR 3.7.2.6 is to verify the systems will automatically 40 switch to the position to provide cooling water exclusively to safety-related equipment during an 41 accident.
42 43 The function of the standby gas treatment (SGT) system is to ensure that radioactive materials 44 that leak from the primary containment into the secondary containment following a DBA are 45 filtered and adsorbed prior to exhausting to the environment. The purpose of SR 3.6.4.3.3 is to 46 verify that each SGT subsystem starts on receipt of an actual or simulated initiation signal. The 47 purpose of SR 3.6.4.3.4 is to verify that the filter cooler bypass damper can be opened, and the 48 fan started. This ensures that the ventilation mode of SGT system operation is available.
49 50 For NUREG-1434, GE BWR/6 Plants:
1 2
The control room fresh air (CRFA) system provides a protected environment from which 3
occupants can control the unit following an uncontrolled release of radioactivity, hazardous 4
chemicals, or smoke. The purpose of SR 3.7.3.3 is to verify that each train/subsystem starts 5
and operates on an actual or simulated actuation signal.
6 7
The ECCS is designed to limit the release of radioactive materials to the environment following 8
a LOCA and consists of the high-pressure core spray (HPCS) system, the low-pressure core 9
spray system, the low pressure coolant injection mode of the RHR system, and the automatic 10 depressurization system. The purpose of SR 3.5.1.5 is to verify the automatic initiation logic of 11 HPCS, low pressure core spray, and low-pressure coolant injection will cause the systems or 12 subsystems to operate as designed, including actuation of the system throughout its emergency 13 operating sequence, automatic pump startup, and actuation of all automatic valves to their 14 required positions on receipt of an actual or simulated actuation signal.
15 16 The function of the RCIC system is to respond to transient events by providing makeup coolant 17 to the reactor. The purpose of SR 3.5.3.5 is to verify the system operates as designed, 18 including actuation of the system throughout its emergency operating sequence; that is, 19 automatic pump startup and actuation of all automatic valves to their required positions on 20 receipt of an actual or simulated actuation signal.
21 22 The standby service water (SSW) system and ultimate heat sink are designed to provide cooling 23 water for the removal of heat from equipment, such as the diesel generators, RHR pump 24 coolers and heat exchangers, and room coolers for ECCS equipment, required for a safe 25 reactor shutdown following a DBA or transient. The SSW system also provides cooling to unit 26 components, as required, during normal shutdown and reactor isolation modes. During a DBA, 27 the equipment required only for normal operation is isolated and cooling is directed to only 28 safety-related equipment. The purpose of SR 3.7.1.6 is to verify the systems will automatically 29 switch to the position to provide cooling water exclusively to safety-related equipment during an 30 accident.
31 32 The RHR containment spray system is designed to mitigate the effects of primary containment 33 bypass leakage and low energy line breaks. The purpose of SR 3.6.1.7.3 is to verify that each 34 RHR containment spray subsystem automatic valve actuates to its correct position upon receipt 35 of an actual or simulated automatic actuation signal.
36 37 The function of the SGT system is to ensure that radioactive materials that leak from the primary 38 containment into the secondary containment following a DBA are filtered and adsorbed prior to 39 exhausting to the environment. The purpose of SR 3.6.4.3.3 is to verify that each SGT 40 subsystem starts on receipt of an actual or simulated initiation signal. The purpose of 41 SR 3.6.4.3.4 is to verify that the filter cooler bypass damper can be opened, and the fan started.
42 This ensures that the ventilation mode of SGT system operation is available.
43 44 The high-pressure core spray service water system (HPCS SWS) provides cooling water for the 45 removal of heat from components of the Division 3 HPCS system. The purpose of SR 3.7.2.3 is 46 to verify that the automatic valves of the HPCS SWS will automatically switch to the safety or 47 emergency position to provide cooling water exclusively to the safety related equipment on an 48 actual or simulated initiation signal.
49 2.2 Proposed Changes to the Standard Technical Specifications 1
2 The proposed changes to the STS would revise certain SRs by adding exceptions to the SR for 3
automatic valves or dampers that are locked, sealed or otherwise secured in the actuated 4
position.
5 6
The following list denotes the proposed changes to the SRs for all plant designs (B&W, 7
Westinghouse, CE, and GE plants, NUREG-1430 through NUREG-1434, respectively). The 8
proposed new text containing the exception is shown in italics.
9 10 For NUREG-1430:
11 12 SR 3.6.7.4 Verify each spray additive automatic valve in the flow path actuates 13 to the correct position on an actual or simulated actuation signal, except for 14 valves that are locked, sealed, or otherwise secured in the actuated position.
15 16 SR 3.7.10.3 Verify [each CREVS train actuates] [or the control room isolates] on 17 an actual or simulated actuation signal, except for dampers and valves that are 18 locked, sealed, or otherwise secured in the actuated position.
19 20 SR 3.7.12.3, Verify each EVS train actuates on an actual or simulated actuation 21 signal, except for dampers and valves that are locked, sealed, or otherwise 22 secured in the actuated position.
23 24 SR 3.7.12.5 Verify each EVS filter cooling bypass damper can be opened, 25 except for dampers that are locked, sealed, or otherwise secured in the open 26 position.
27 28 SR 3.7.13.3 Verify each FSPVS train actuates on an actual or simulated 29 actuation signal, except for dampers and valves that are locked, sealed, or 30 otherwise secured in the actuated position.
31 32 SR 3.7.13.5 Verify each FSPVS filter bypass damper can be opened, except for 33 dampers that are locked, sealed, or otherwise secured in the open position.
34 35 For NUREG-1431:
36 37 SR 3.6.11.3 Verify each ICS train actuates on an actual or simulated actuation 38 signal, except for dampers and valves that are locked, sealed, or otherwise 39 secured in the actuated position.
40 41 SR 3.6.11.4 Verify each ICS filter bypass damper can be opened, except for 42 dampers that are locked, sealed, or otherwise secured in the open position.
43 44 SR 3.6.13.3 Verify each SBACS train actuates on an actual or simulated 45 actuation signal, except for dampers and valves that are locked, sealed, or 46 otherwise secured in the actuated position.
47 48 SR 3.6.13.4 Verify each SBACS filter bypass damper can be opened, except for 49 dampers that are locked, sealed, or otherwise secured in the open position.
50 51 SR 3.7.10.3 Verify each CREFS train actuates on an actual or simulated 1
actuation signal, except for dampers and valves that are locked, sealed, or 2
otherwise secured in the actuated position.
3 4
SR 3.7.12.3 Verify each ECCS PREACS train actuates on an actual or 5
simulated actuation signal, except for dampers and valves that are locked, 6
sealed, or otherwise secured in the actuated position.
7 8
SR 3.7.12.5 Verify each ECCS PREACS filter bypass damper can be closed, 9
except for dampers that are locked, sealed, or otherwise secured in the closed 10 position.
11 12 SR 3.7.13.3 Verify each FBACS train actuates on an actual or simulated 13 actuation signal, except for dampers and valves that are locked, sealed, or 14 otherwise secured in the actuated position.
15 16 SR 3.7.13.5 Verify each FBACS filter bypass damper can be closed, except for 17 dampers that are locked, sealed, or otherwise secured in the closed position.
18 19 SR 3.7.14.3 Verify each PREACS train actuates on an actual or simulated 20 actuation signal, except for dampers and valves that are locked, sealed, or 21 otherwise secured in the actuated position.
22 23 SR 3.7.14.5 Verify each PREACS filter bypass damper can be closed, except for 24 dampers that are locked, sealed, or otherwise secured in the closed position.
25 26 For NUREG-1432:
27 28 SR 3.6.8.3 Verify each SBEACS train actuates on an actual or simulated 29 actuation signal, except for dampers and valves that are locked, sealed, or 30 otherwise secured in the actuated position.
31 32 SR 3.6.8.4 Verify each SBEACS filter bypass damper can be opened, except for 33 dampers that are locked, sealed, or otherwise secured in the open position.
34 35 SR 3.6.10.3 Verify each ICS train actuates on an actual or simulated actuation 36 signal, except for dampers and valves that are locked, sealed, or otherwise 37 secured in the actuated position.
38 39 SR 3.6.10.4 Verify each ICS filter bypass damper can be opened, except for 40 dampers that are locked, sealed, or otherwise secured in the open position.
41 42 SR 3.7.10.2 Verify the proper actuation of each ECW System component on an 43 actual or simulated actuation signal, except for valves that are locked, sealed, or 44 otherwise secured in the actuated position.
45 46 SR 3.7.11.3 Verify each CREACS train actuates on an actual or simulated 47 actuation signal, except for dampers and valves that are locked, sealed, or 48 otherwise secured in the actuated position.
49 50 SR 3.7.13.3 Verify each ECCS PREACS train actuates on an actual or 1
simulated actuation signal, except for dampers and valves that are locked, 2
sealed, or otherwise secured in the actuated position.
3 4
SR 3.7.13.5 Verify each ECCS PREACS filter bypass damper can be opened, 5
except for dampers that are locked, sealed, or otherwise secured in the open 6
position.
7 8
SR 3.7.14.3 Verify each FBACS train actuates on an actual or simulated 9
actuation signal, except for dampers and valves that are locked, sealed, or 10 otherwise secured in the actuated position.
11 12 SR 3.7.14.5 Verify each FBACS filter bypass damper can be opened, except for 13 dampers that are locked, sealed, or otherwise secured in the open position.
14 15 SR 3.7.15.3 Verify each PREACS train actuates on an actual or simulated 16 actuation signal, except for dampers and valves that are locked, sealed, or 17 otherwise secured in the actuated position.
18 19 SR 3.7.15.5 Verify each PREACS filter bypass damper can be opened, except 20 for dampers that are locked, sealed, or otherwise secured in the open position.
21 22 For NUREG-1433:
23 24 SR 3.5.1.10 Verify each ECCS injection/spray subsystem actuates on an actual 25 or simulated automatic initiation signal, except for valves that are locked, sealed, 26 or otherwise secured in the actuated position.
27 28 SR 3.5.3.5 Verify the RCIC System actuates on an actual or simulated 29 automatic initiation signal, except for valves that are locked, sealed, or otherwise 30 secured in the actuated position.
31 32 SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual or simulated 33 initiation signal, except for dampers that are locked, sealed, or otherwise secured 34 in the actuated position.
35 36 SR 3.6.4.3.4 Verify each SGT filter cooler bypass damper can be opened and 37 the fan started, except for dampers that are locked, sealed, or otherwise secured 38 in the open position.
39 40 SR 3.7.2.6 Verify each [PSW] subsystem actuates on an actual or simulated 41 initiation signal, except for valves that are locked, sealed, or otherwise secured in 42 the actuated position.
43 44 SR 3.7.4.3 Verify each [MCREC] subsystem actuates on an actual or simulated 45 initiation signal, except for dampers and valves that are locked, sealed, or 46 otherwise secured in the actuated position.
47 48 For NUREG-1434:
1 2
SR 3.5.1.5 Verify each ECCS injection/spray subsystem actuates on an actual 3
or simulated automatic initiation signal, except for valves that are locked, sealed, 4
or otherwise secured in the actuated position.
5 6
SR 3.5.3.5 Verify the RCIC System actuates on an actual or simulated 7
automatic initiation signal, except for valves that are locked, sealed, or otherwise 8
secured in the actuated position.
9 10 SR 3.6.1.7.3 Verify each RHR containment spray subsystem automatic valve in 11 the flow path actuates to its correct position on an actual or simulated automatic 12 initiation signal, except for valves that are locked, sealed, or otherwise secured in 13 the actuated position.
14 15 SR 3.6.4.3.3 Verify each SGT subsystem actuates on an actual or simulated 16 initiation signal, except for dampers that are locked, sealed, or otherwise secured 17 in the actuated position.
18 19 SR 3.6.4.3.4 Verify each SGT filter cooler bypass damper can be opened and 20 the fan started, except for dampers that are locked, sealed, or otherwise secured 21 in the open position.
22 23 SR 3.7.1.6 Verify each [SSW] subsystem actuates on an actual or simulated 24 initiation signal, except for valves that are locked, sealed, or otherwise secured in 25 the actuated position.
26 27 SR 3.7.2.3 Verify the HPCS SWS actuates on an actual or simulated initiation 28 signal, except for valves that are locked, sealed, or otherwise secured in the 29 actuated position.
30 31 SR 3.7.3.3 Verify each [CRFA] subsystem actuates on an actual or simulated 32 initiation signal, except for dampers and valves that are locked, sealed, or 33 otherwise secured in the actuated position.
34 35 In Volume 2 of each NUREG, where the reason for each particular SR is described, the 36 following text would be added:
37 38 The SR excludes automatic dampers and valves that are locked, sealed, or 39 otherwise secured in the actuated position. The SR does not apply to dampers 40 or valves that are locked, sealed, or otherwise secured in the actuated position 41 since the affected dampers or valves were verified to be in the actuated position 42 prior to being locked, sealed, or otherwise secured. Placing an automatic valve 43 or damper in a locked, sealed, or otherwise secured position requires an 44 assessment of the operability of the system or any supported systems, including 45 whether it is necessary for the valve or damper to be repositioned to the 46 non-actuated position to support the accident analysis. Restoration of an 47 automatic valve or damper to the non-actuated position requires verification that 48 the SR has been met within its required Frequency.
49 50 The traveler would also correct errors in the descriptions of the reasons for NUREG-1430, 1
SR 3.7.12.5; NUREG-1432, SR 3.7.13.5; NUREG-1432, SR 3.7.14.5; and NUREG-1432, 2
SR 3.7.15.5 in Volume 2 of each respective NUREG. The descriptions erroneously state that 3
operability is verified if the damper can be closed. The description should state operability is 4
verified if the damper can be opened.
5 6
2.3 Applicable Regulatory Requirements and Guidance 7
8 Section IV, The Commission Policy, of the Final Policy Statement on TS Improvements for 9
Nuclear Power Reactors states, in part:
10 11 The purpose of Technical Specifications is to impose those conditions or 12 limitations upon reactor operation necessary to obviate the possibility of an 13 abnormal situation or event giving rise to an immediate threat to the public health 14 and safety by identifying those features that are of controlling importance to 15 safety and establishing on them certain conditions of operation which cannot be 16 changed without prior Commission approval.
17 18
[T]he Commission will also entertain requests to adopt portions of the 19 improved STS [(e.g., TSTF-541)], even if the licensee does not adopt all STS 20 improvements. The Commission encourages all licensees who submit 21 Technical Specification related submittals based on this Policy Statement to 22 emphasize human factors principles.
23 24 In accordance with this Policy Statement, improved STS have been developed 25 and will be maintained for each NSSS [nuclear steam supply system] owners 26 group. The Commission encourages licensees to use the improved STS as the 27 basis for plant-specific Technical Specifications. [I]t is the Commission intent 28 that the wording and Bases of the improved STS be used to the extent 29 practicable.
30 31 The Summary section of the Final Policy Statement on TS Improvements for Nuclear Power 32 Reactors states, in part:
33 34 Implementation of the Policy Statement through implementation of the improved 35 STS is expected to produce an improvement in the safety of nuclear power 36 plants through the use of more operator-oriented Technical Specifications, 37 Improved Technical Specification Bases, reduced action statement induced plant 38 transients, and more efficient use of NRC and industry resources.
39 40 The regulation under 10 CFR 50.36(a)(1) requires that:
41 42 Each applicant for a license authorizing operation of a production or utilization 43 facility shall include in his application proposed technical specifications in 44 accordance with the requirements of this section. A summary statement of the 45 bases or reasons for such specifications, other than those covering 46 administrative controls, shall also be included in the application, but shall not 47 become part of the technical specifications.
48 49 The regulation under 10 CFR 50.36(b) requires that:
1 2
Each license authorizing operation of a utilization facility will include 3
technical specifications. The technical specifications will be derived from the 4
analyses and evaluation included in the safety analysis report, and amendments 5
thereto, submitted pursuant to [10 CFR] 50.34 [Contents of applications; 6
technical information]. The Commission may include such additional technical 7
specifications as the Commission finds appropriate.
8 9
The categories of items required to be in the TS are listed in 10 CFR 50.36(c).
10 11 The regulation under 10 CFR 50.36(c)(2) states that LCOs are the lowest functional capability 12 or performance levels of equipment required for safe operation of the facility. The regulation 13 requires that when an LCO of a nuclear reactor is not met, the licensee shall shut down the 14 reactor or follow any remedial action permitted by the TS until the condition can be met.
15 16 SRs are defined in 10 CFR 50.36(c)(3) as requirements relating to test, calibration, or 17 inspection to assure that the necessary quality of systems and components is maintained, that 18 facility operation will be within safety limits, and that the limiting conditions for operation will be 19 met.
20 21 The regulation under 10 CFR 50.36(c)(5) requires TS to include administrative controls, which 22 are the provisions relating to organization and management, procedures, recordkeeping, 23 review and audit, and reporting necessary to assure operation of the facility in a safe manner.
24 25 The regulation under 10 CFR 50.59, Changes, tests, and experiments, contains requirements 26 for the process by which licensees, under certain conditions, may make changes to their 27 facilities and procedures as described in the Final Safety Analysis Report (FSAR) (as updated),
28 without prior NRC approval. The process requires licensees to request a license amendment 29 via 10 CFR 50.90 for any change that would require NRC approval.
30 31 Section 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at 32 nuclear power plants, requires licensees to monitor the performance or condition of SSCs, 33 against licensee-established goals, in a manner sufficient to provide reasonable assurance that 34 these SSCs, as defined in paragraph (b) of this section, are capable of fulfilling their intended 35 functions.
36 37 The regulation under 10 CFR 50.65(a)(4) states:
38 39 Before performing maintenance activities (including but not limited to 40 surveillance, post-maintenance testing, and corrective and preventive 41 maintenance), the licensee shall assess and manage the increase in risk that 42 may result from the proposed maintenance activities. The scope of the 43 assessment may be limited to structures, systems, and components that a risk-44 informed evaluation process has shown to be significant to public health and 45 safety.
46 47 The regulation under 10 CFR 50.65(b) states:
1 2
The scope of the monitoring program specified in paragraph (a)(1) of this section 3
shall include safety related and nonsafety related structures, systems, and 4
components, as follows:
5 6
(1) Safety-related structures, systems and components that are relied upon to 7
remain functional during and following design basis events to ensure the integrity 8
of the reactor coolant pressure boundary, the capability to shut down the reactor 9
and maintain it in a safe shutdown condition, or the capability to prevent or 10 mitigate the consequences of accidents that could result in potential offsite 11 exposure comparable to the guidelines in [10 CFR] 50.34(a)(1),
12
[10 CFR] 50.67(b)(2), or [10 CFR] 100.11 of this chapter, as applicable.
13 14 (2) Nonsafety related structures, systems, or components:
15 16 (i) That are relied upon to mitigate accidents or transients or are used in plant 17 emergency operating procedures (EOPs); or 18 19 (ii) Whose failure could prevent safety-related structures, systems, and 20 components from fulfilling their safety-related function; or 21 22 (iii) Whose failure could cause a reactor scram or actuation of a safety-related 23 system.
24 25 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing 26 Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, 27 establishes quality assurance requirements for the operation of nuclear power plant 28 safety-related SSCs.
29 30 NRC Regulatory Guide (RG) 1.33, Revision 2, Quality Assurance Program Requirements 31 (Operation), with Appendix A, Typical Procedures for Pressurized Water Reactors and Boiling 32 Water Reactors, dated February 1978 (ADAMS Accession No. ML003739995), describes a 33 method acceptable to the NRC staff for complying with the Commissions regulations with 34 regard to overall quality assurance program requirements for the operation phase of nuclear 35 power plants. Section 8.b of RG 1.33, Appendix A, states that implementing procedures are 36 required for each surveillance test, inspection, or calibration listed in the technical 37 specifications. Section 9.e of RG 1.33, Appendix A, states that General procedures for the 38 control of maintenance, repair, replacement, and modification work should be prepared before 39 reactor operation is begun. Section 9.e.1 states that the procedures should include information 40 such as methods for obtaining permission and clearance for operation personnel to work and for 41 logging such work.
42 43 STS 5.4.1.a in the Administrative Controls section of NUREG-1430 through 1434 contains 44 requirements that written procedures shall be established, implemented, and maintained 45 covering the applicable procedures recommended in RG 1.33, Revision 2, Appendix A, 46 February 1978.
47 48 STS 5.5.11/5.5.8, Ventilation Filter Testing Program (VFTP), in the Administrative Controls 49 section of NUREG-1430 through 1434 contains requirements to identify any filter degradation 50 and ensures the ability of the filters to perform in a manner consistent with the licensing basis 1
for the facility.
2 3
The NRC staffs guidance for the review of TS is in Chapter 16.0, Revision 3, Technical 4
Specifications, dated March 2010 (ADAMS Accession No. ML100351425) of NUREG-0800, 5
Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:
6 LWR [Light-Water Reactor] Edition (SRP). As described therein, as part of the regulatory 7
standardization effort, the NRC staff has prepared STS for each of the LWR nuclear designs.
8 Accordingly, the NRC staffs review includes consideration of whether the proposed changes 9
are consistent with the applicable reference STS (i.e., the current STS), as modified by 10 NRC-approved travelers. In addition, the guidance states that comparing the change to 11 previous STS can help clarify the TS intent.
12 13
3.0 TECHNICAL EVALUATION
14 15 The NRC staff reviewed Traveler TSTF-541, Revision 2, which proposed changes to 16 NUREG-1430, NUREG-1431, NUREG-1432, NUREG-1433, and NUREG-1434. The regulatory 17 framework the NRC staff used to determine the acceptability of the proposed changes consists 18 of the requirements and guidance listed in Section 2.3 of this safety evaluation. The NRC staff 19 reviewed the changes to determine whether the proposed changes to the STS meet the 20 standards for TS in 10 CFR 50.36, as well as conform to the Final Policy Statement on TS 21 Improvements for Nuclear Power Reactors. The NRC staff also used the SRP to determine 22 whether the proposed changes to the STS would clarify the intent of the STS.
23 24 In NUREG-1430 through 1434, the NRC staff-accepted format for SRs is text which states that 25 certain SSCs or systems (subsystems, trains, etc.) of components must be verified to be able to 26 actuate or function. Each verification must be performed at a given frequency. The rules 27 governing SRs are explicitly stated in the STS in SR 3.0.1 through SR 3.0.4. While SR 3.0.1 28 through SR 3.0.4 are explicit with respect to when SRs are to be met and performed, the text of 29 the individual SRs in the STS, and typically in a plant-specific TS, does not contain more detail 30 than a system name or component name. The details of how the licensee will implement SRs 31 are contained in licensee-controlled procedures.
32 33 During its reviews of previous proposals to address the issue (i.e., TSTF-512 and earlier 34 revisions of TSTF-541), the NRC staff had concerns regarding the acceptability of providing 35 exceptions to the SRs for SSCs. The NRC staff was concerned that locking or securing SSCs 36 in position could have inadvertent effects on system OPERABILITY, SSC quality, clarity of a 37 plants licensing basis, and the validity of a plants current radiological consequence analyses if 38 exceptions to the SRs for SSCs were adopted. The technical evaluation section of TSTF-541, 39 Revision 2, contains justification for the current proposed change and states:
40 41 These allowances permit components to be exempted from testing under the SR.
42 However, the proposed change does not permit a system that is inoperable to be 43 considered operable. As stated in the SR 3.0.1 Bases, Nothing in this 44 Specification, however, is to be construed as implying that systems or 45 components are OPERABLE when: a. The systems or components are known to 46 be inoperable, although still meeting the SRs.
47 48 Placing a component in a condition not consistent with the design requires 49 consideration of the effect on the operability of the associated system or any 50 supported systems under the licensees administrative processes, such as the 51 operability determination process. The model application requires licensees to 1
verify that their administrative processes require assessing the operability of the 2
system or any supported systems when utilizing the SR allowances. The 3
operability assessment will consider whether movement of the affected valves or 4
dampers following an event is assumed in the safety analysis (i.e., the analysis of 5
design basis accidents, anticipated operational occurrences, and transients).
6 7
As stated in the proposed TS Bases, the automatic valve or damper is verified to 8
be in the correct position prior to locking, sealing, or securing it in position.
9 Valves and dampers that are locked, sealed, or otherwise secured are entered 10 into the licensees tagging program, which is routinely inspected by the NRC 11 under various 71111 procedures in the NRC Inspection Manual. While in the 12 actuated position, verification of automatic actuation or valve isolation time is not 13 necessary as the specified safety function is assured. However, as with the 14 existing similar SR allowances, the SR must be verified to be met within its 15 required Frequency after removing the valve or damper from the locked, sealed 16 or otherwise secured status.
17 18 These allowances and the proposed change do not permit changing the plant 19 design, which must be evaluated under 10 CFR 50.59, and the Final Safety 20 Analysis Report (FSAR) must be updated per 10 CFR 50.71(e). If the valve or 21 damper is locked, sealed, or otherwise secured to support plant operation (such 22 as changing modes, or removing or placing systems in operation), restoration to 23 the design condition is controlled by plant procedures, changes to which are also 24 governed by 10 CFR 50.59. If the valve or damper is locked, sealed, or 25 otherwise secured to facilitate maintenance, restoration is governed by 26 10 CFR 50, Appendix B, Criterion XVI, and 10 CFR 50.65. If the SR exception is 27 utilized to not test the actuation of a valve or damper and the specified 28 Frequency of the SR is exceeded without testing the component, the SR must be 29 performed on the component when it is returned to service in order to meet the 30 SR.
31 32 Under the proposed change, the affected valves and dampers may be excluded 33 from testing when locked, sealed or otherwise secured in the actuated position.
34 However, if the exception is used the operability of the system or any supported 35 systems must be assessed, including whether the safety analysis assumes 36 movement from the actuated position following an event. If the system cannot 37 perform its specified safety function it is inoperable regardless of whether the SR 38 is met. Therefore, the proposed allowance has no effect on the ability to satisfy 39 the safety analysis assumptions.
40 41 The above justification was developed during TSTF and NRC discussions regarding previous 42 revisions of TSTF-541. The NRC staff agrees with the statements for the reasons described in 43 the following paragraphs.
44 45 In the technical evaluation section of TSTF-541, Revision 2, quoted above, the traveler states 46 that safety analysis is the analysis of design basis accidents, anticipated operational 47 occurrences, and transients. It is noted that in the proposed changes to the STS Bases, this is 48 referred to as the accident analysis. The NRC staff notes that while accidents are a specific 49 category of all design basis events, the terms safety analysis and accident analysis are 50 considered equivalent in this context.
51 1
The procedures for how a licensee will implement SRs are discussed in Section 8.b of 2
Appendix A to RG 1.33, Revision 2, which is a requirement of STS 5.4. The procedures for 3
general maintenance and equipment work clearances and logging discussed in Section 9.e of 4
Appendix A to RG 1.33, Revision 2, are also requirements of TS 5.4. Since SR procedures 5
along with maintenance, equipment work clearance, and logging procedures are 6
licensee-controlled documents, changes to the procedure details must be done in accordance 7
with 10 CFR 50.59. If the change would require NRC approval, 10 CFR 50.59 would require the 8
licensee to submit an amendment request to the NRC per 10 CFR 50.90. SSCs with SRs are 9
scoped into the requirements of 10 CFR 50.65 and 10 CFR 50.65(a)(4) contains the 10 requirement to assess and manage the risk of maintenance. Therefore, a licensee must further 11 evaluate the effect of any maintenance on SSCs for which the exception is employed. Given 12 the stipulations of 10 CFR 50.59 and 10 CFR 50.65, the NRC staff has reasonable assurance 13 that a licensee will assess the impact of using the exception in the SR for the SSCs and 14 systems involved. If a licensee failed to make the proper assessments, enforcement actions 15 related to the stated regulations could be taken.
16 17 Since 10 CFR 50.59 and 10 CFR 50.65 require a licensee to evaluate and document a change, 18 the exception is acceptable because there is reasonable assurance that placing the component 19 in a given position will not inadvertently impact the operability of required SSCs. The NRC staff 20 determined that there is reasonable assurance that the change will not have inadvertent effects 21 on system OPERABILITY or SSC quality.
22 23 The traveler contained a model license amendment request (LAR) that a licensee would use to 24 propose adoption of the TSTF-541, Revision 2, changes to its TS via 10 CFR 50.90. The model 25 LAR contains the following statements a licensee would make to propose adoption of the 26 changes to its TS:
27 28 While the proposed exceptions permit automatic valves and dampers that are 29 locked, sealed, or otherwise secured in the actuated position to be excluded from 30 the SR in order to consider the SR met, the proposed changes will not permit a 31 system that is made inoperable by locking, sealing, or otherwise securing an 32 automatic valve or damper in the actuated position to be considered operable.
33 As stated in the [SR 3.0.1] Bases, Nothing in this Specification, however, is to be 34 construed as implying that systems or components are OPERABLE when: a. The 35 systems or components are known to be inoperable, although still meeting the 36 SRs.
37 38 39 40
[LICENSEE] acknowledges that under the proposed change, the affected valves 41 and dampers may be excluded from the SR when locked, sealed or otherwise 42 secured in the actuated position. However, if the safety analysis assumes 43 movement from the actuated position following an event, or the system is 44 rendered inoperable by locking, sealing, or otherwise securing the valve or 45 damper in the actuated position, then the system cannot perform its specified 46 safety function and is inoperable regardless of whether the SR is met.
47 48
[LICENSEE] acknowledges for components for which the SR allowance can be 49 utilized, the SR must be verified to have been met within its required Frequency 50 after removing the valve or damper from the locked, sealed or otherwise secured 51 status. If the SR exception is utilized to not test the actuation of a valve or 1
damper and the specified Frequency of the SR is exceeded without testing the 2
component, the SR must be performed on the component when it is returned to 3
service in order to meet the SR.
4 5
Given the statements a licensee would provide on the docket to adopt the TSTF-541, 6
Revision 2, changes, the NRC staff determined that there is reasonable assurance that the 7
change will not have inadvertent effects on the clarity of a plants licensing basis.
8 9
The NRC staff determined that the STS, as amended by the TSTF-541, Revision 2, changes will 10 continue to provide an acceptable way to meet 10 CFR 50.36(c)(3) because the STS SRs will 11 continue to provide assurance that the necessary quality of systems and components is 12 maintained and that the LCOs will be met.
13 14 The NRC staff also determined that when the exception is used, the radiological consequences 15 for the accidents previously evaluated are not changed since the system is still capable of 16 performing the specified safety function assumed in the accident analyses and the associated 17 TS actions are followed if the system cannot perform its specified safety function. Additionally, 18 the licensee is required to perform filter testing in accordance with the Ventilation Filter Testing 19 Program as stated in the accompanying STSs SRs, as these SRs are not affected by this 20 proposed change. The Ventilation Filter Testing Program in STS 5.5.11/5.5.8 would identify any 21 filter degradation and ensure the ability of the filters to perform in a manner consistent with the 22 licensing basis for the facility.
23 24
4.0 CONCLUSION
25 26 The NRC staff reviewed Traveler TSTF-541, Revision 2, which proposed changes to 27 NUREG-1430, NUREG-1431, NUREG-1432, NUREG-1433, and NUREG-1434. The NRC staff 28 determined that the proposed changes to the STS meet the standards for TS in 10 CFR 50.36.
29 The proposed changes and the STS, as revised, continue to specify the appropriate SRs for 30 tests and inspections to ensure the necessary quality of affected SSCs is maintained and that 31 the LCOs are met.
32 33 Additionally, the changes to the STS were reviewed and found to be technically clear and 34 consistent with customary terminology and format in accordance with SRP Chapter 16.0.
35 36 The NRC staff reviewed the proposed changes against the regulations and concludes that the 37 changes continue to meet the requirements of 10 CFR 50.36, for the reasons discussed above, 38 and thus provide reasonable assurance that a licensee adopting these changes will have the 39 requisite requirements and controls to operate safely. Therefore, the NRC staff concludes that 40 the proposed STS changes are acceptable.
41 42 Principal Contributors: Matthew Hamm, NRR/DSS 43 Kristy Bucholtz, NRR/DSS 44 Robert Beaton, NRR/DSS 45 46 Date:
47