ML20029E406
| ML20029E406 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/05/2020 |
| From: | Klos L Plant Licensing Branch IV |
| To: | Diya F Union Electric Co |
| Klos L | |
| References | |
| EPID L-2019-LLA-0063, TSTF-545, Rev 3 | |
| Download: ML20029E406 (32) | |
Text
March 5, 2020 Mr. Fadi Diya Senior Vice President and Chief Nuclear Officer Ameren Missouri Callaway Energy Center 8315 County Road 459 Steedman, MO 65077
SUBJECT:
CALLAWAY PLANT, UNIT NO. 1 - ISSUANCE OF AMENDMENT NO. 222 RE:
ADOPTION OF TECHNICAL SPECIFICATIONS TASK FORCE (TSTF)
TRAVELER TSTF-545, REVISION 3 (EPID L-2019-LLA-0063)
Dear Mr. Diya:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 222 to Renewed Facility Operating License No. NPF-30 for Callaway Plant, Unit No. 1. The amendment consists of changes to the technical specifications (TSs) in response to your application dated March 22, 2019, as supplemented by letter dated May 21, 2019.
The amendment deletes TS 5.5.8, Inservice Testing Program. A new defined term INSERVICE TESTING PROGRAM, is add to TS Section 1.0, Definitions. In addition, existing uses of the term Inservice Testing Program in the TSs are capitalized throughout to indicate that it is now a defined term. These changes are based on NRC-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications Change Traveler TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing, dated October 21, 2015.
A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
L. John Klos, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483
Enclosures:
- 1. Amendment No. 222 to NPF-30
- 2. Safety Evaluation cc: Listserv
UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 222 Renewed License No. NPF-30
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Union Electric Company (UE, the licensee),
dated March 22, 2019, as supplemented by letter dated May 21, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-30 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan*
The Technical Specifications contained in Appendix A, as revised through Amendment No. 222 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This amendment is effective as of its date of issuance, and shall be implemented within 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License No. NPF-30 and Technical Specifications Date of Issuance: March 5, 2020
ATTACHMENT TO LICENSE AMENDMENT NO. 222 CALLAWAY PLANT, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Replace the following pages of the Renewed Facility Operating License No. NPF-30 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 1.1-3 1.1-3 1.1-4 1.1-4 1.1-5 1.1-5 1.1-6 1.1-6 3.4-21 3.4-21 3.4-25 3.4-25 3.4-35 3.4-35 3.5-5 3.5-5 3.5-6 3.5-6 3.6-14 3.6-14 3.6-20 3.6-20 3.7-2 3.7-2 3.7-8 3.7-8 3.7-11 3.7-11 3.7-14 3.7-14 3.7-47 3.7-47 5.0-10 5.0-10
- (3)
UE, pursuant to the Act and 1 O CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)
UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
( 1)
Maximum Power Level UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal ( 100% power) in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan*
(3)
The Technical Specifications contained in Appendix A, as revised through Amendment No. 222 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Environmental Qualification {Section 3.11. SSER #3)**
Deleted per Amendment No. 169.
Amendments 133, 134, & 135 were effective as of April 30, 2000 however these amendments were implemented on April 1, 2000.
The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Renewed License No. NPF-30 Amendment No. 222
1.1 Definitions (continued)
DOSE EQUIVALENT XE-133 ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE CALLAWAY PLANT Definitions 1.1 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-87, Kr-88, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity.
The determination of DOSE EQUIVALENT XE-133 shall be performed using the effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water, and Soil", 1993.
The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.
The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.SSa(f).
LEAKAGE shall be:
- a. Identified LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or (continued) 1.1-3 Amendment 222
1.1 Definitions LEAKAGE (continued)
MASTER RELAY TEST MODE OPERABLE - OPERABILITY CALLAWAY PLANT Definitions 1.1
- 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
- b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water leakoff) that is not identified LEAKAGE;
- c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
1.1-4 (continued)
Amendment 222
1.1 Definitions (continued)
PHYSICS TESTS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
QUADRANT POWER TILT RATIO (QPTR)
RATED THERMAL POWER (RTP)
REACTOR TRIP SYSTEM (RTS) RESPONSE TIME CALLAWAY PLANT Definitions 1.1 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
- a. Described in Chapter 14 of the FSAR;
- b. Authorized under the provisions of 10 CFR 50.59; or
- c. Otherwise approved by the Nuclear Regulatory Commission.
The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and REPORT (PTLR) cooldown rates, the power operated relief valve (PORV) lift settings, and the Cold Overpressure Mitigation System (COMS) arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6.
QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3565 Mwt.
The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.
1.1-5 (continued)
Amendment 222
1.1 Definitions (continued}
Definitions 1.1 SHUTDOWN MARGIN (SDM}
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
SLAVE RELAY TEST STAGGERED TEST BASIS THERMAL POWER TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT}
CALLAWAY PLANT
- a. All rod cluster control assemblies (RCCAs} are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
- b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the hot zero power temperatures.
A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.
1.1-6 Amendment 222
Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SURVEILLANCE Verify each pressurizer safety valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, lift settings shall be within +/- 1 % of 2460 psig.
CALLAWAY PLANT 3.4-21 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Amendment No. 222
SURVEILLANCE REQUIREMENTS SR 3.4.11.1 SR 3.4.11.2 SURVEILLANCE
~--------~--------~-----NOTE------~------~------------
Not required to be performed with block valve closed in accordance with the Required Actions of this LCO.
Perform a complete cycle of each block valve.
Perform a complete cycle of each PORV.
CALLAWAY PLANT 3.4-25 Pressurizer PORVs 3.4.11 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM Amendment No. 222
SURVEILLANCE REQUIREMENTS SR 3.4.14.1 SURVEILLANCE
-~--------~--------~--------~ N()TES ------------~------------
- 1.
Not required to be performed in M()DES 3 and 4.
- 2.
Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.
- 3.
RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.
Verify leakage from each RCS PIV is equivalent to
~ 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure~ 2215 psig and~ 2255 psig.
CALLAWAY PLANT 3.4-35 RCS PIV Leakage 3.4.14 FREQUENCY In accordance with the INSERVICE TESTING PR()GRAM, In accordance with the Surveillance Frequency Control Program Prior to entering M()DE 2 whenever the unit has been in M()DE 5 for 7 days or more and if leakage testing has not been performed in the previous 9 months (continued)
Amendment No. 222
SURVEILLANCE REQUIREMENTS SR 3.5.2.1 Number SURVEILLANCE Verify the following valves are in the listed position with power to the valve operator removed.
Position Function BNHV8813 Open Safety Injection to RWST Isolation Valve EMHV8802A Closed SI Hot Legs 2 & 3 Isolation Valve EMHV8802B Closed SI Hot Legs 1 & 4 Isolation Valve EMHV8835 Open Safety Injection Cold Leg Isolation Valve EJHV8840 Closed RHR/SI Hot Leg Recirc Isolation Valve EJHV8809A Open RHR to Accum Inject Loops 1 & 2 Isolation Valve EJHV8809B Open RHR to Accum Inject Loops SR 3.5.2.2 SR 3.5.2.3 3 & 4 Isolation Valve
~--------~--------~-----NOTE--------------~------------
Not required to be met for system vent flow paths opened under administrative control.
Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
Verify ECCS locations susceptible to gas accumulation are sufficiently filled with water.
CALLAWAY PLANT 3.5-5 ECCS - Operating 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program (continued)
Amendment No. 222
SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.4 SR 3.5.2.5 SR 3.5.2.6 SR 3.5.2.7 SR 3.5.2.8 SURVEILLANCE Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.
Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.
Verify each ECCS pump starts automatically on an actual or simulated actuation signal.
Verify, for each ECCS throttle valve listed below, each mechanical position stop is in the correct position.
EMV0095 EMV0096 EMV0097 EMV0098 Valve Number EMV0107 EMV0108 EMV0109 EMV0110 EMV0089 EMV0090 EMV0091 EMV0092 Verify, by visual inspection, each ECCS train containment sump suction inlet is not restricted by debris and the suction inlet strainers show no evidence of structural distress or abnormal corrosion.
CALLAWAY PLANT 3.5-6 ECCS - Operating 3.5.2 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 222
Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.3.4 SR 3.6.3.5 SR 3.6.3.6 SURVEILLANCE
NOTE ---------------------------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify each containment isolation manual valve and blind flange that is located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.
Verify the isolation time of each automatic power operated containment isolation valve is within limits.
NOTE---------------------------
Only required to be performed when containment shutdown purge valve blind flanges are installed.
Perform leakage rate testing for containment shutdown purge valves with resilient seals and associated blind flanges.
CALLAWAY PLANT 3.6-14 FREQUENCY Prior to entering MODE 4from MODE 5 if not performed within the previous 92 days In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Following each reinstallation of the blind flange (continued)
Amendment No. 222
Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.6.3 Verify each containment cooling train cooling water In accordance flow rate is ~ 2200 gpm.
with the Surveillance Frequency Control Program SR 3.6.6.4 Verify each containment spray pump's developed In accordance with head at the flow test point is greater than or equal to the INSERVICE the required developed head.
TESTING PROGRAM SR 3.6.6.5 Verify each automatic containment spray valve in the In accordance flow path that is not locked, sealed, or otherwise with the secured in position, actuates to the correct position on Surveillance an actual or simulated actuation signal.
Frequency Control Program SR 3.6.6.6 Verify each containment spray pump starts In accordance automatically on an actual or simulated actuation with the signal.
Surveillance Frequency Control Program SR 3.6.6.7 Verify each containment cooling train starts In accordance automatically and minimum cooling water flow rate is with the established on an actual or simulated actuation Surveillance signal.
Frequency Control Program SR 3.6.6.8 Verify each spray nozzle is unobstructed.
In accordance with the Surveillance Frequency Control Program (continued)
CALLAWAY PLANT 3.6-20 Amendment No. 222
ACTIONS (continued)
CONDITION REQUIRED ACTION B.
(continued) 8.2 NOTE Only required in Mode 1.
Reduce the Power Range Neutron Flux -
High Trip setpoints to less than or equal to the Maximum Allowable %
RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVS.
C.
Required Action and C.1 Bein MODE 3 associated Completion Time not met.
AND OR C.2 BeinMODE4 One or more steam generators with ~ 4 MSSVs inoperable.
SURVEILLANCE REQUIREMENTS SR 3.7.1.1 SURVEILLANCE
~~~~~~~~NOTE~~~~~~-
Only required to be performed in MODES 1 and 2.
Verify each required MSSV lift setpoint per Table
- 3. 7.1-2 in accordance with the INSERVICE TESTING PROGRAM. Following testing, lift setting shall be within+/- 1%.
CALLAWAY PLANT 3.7-2 MSSVs 3.7.1 COMPLETION TIME 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 6 hours 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> FREQUENCY In accordance with the INSERVICE TESTING PROGRAM Amendment No. 222
MSIVs, MSIVBVs, and MSLPDIVs 3.7.2 SURVEILLANCE REQUIREMENTS SR 3.7.2.1 SR 3.7.2.2 SR 3.7.2.3 SURVEILLANCE Verify isolation time of each MSIV is within limits.
Verify each MSIV, each MSIVBV, and each MSLPDIV actuates to the isolation position on an actual or simulated actuation signal.
Verify isolation time of each MSIVBV and MSLPDIV is within limits.
CALLAWAY PLANT 3.7-8 FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PROGRAM Amendment No. 222
MFIVs, MFRVs, and MFRVBVs 3.7.3 SURVEILLANCE REQUIREMENTS SR 3.7.3.1 SR 3.7.3.2 SR 3.7.3.3 SURVEILLANCE FREQUENCY
N()TE ---------------------------
()nly required to be performed in M()DES 1 and 2.
Verify the closure time of each MFRV and MFRVBV is In accordance with within limits.
the INSERVICE TESTING PR()GRAM
N ()TE ---------------------------
For the MFRVs and MFRVBVs, only required to be performed in M()DES 1 and 2.
Verify each MFIV, MFRV and MFRVBV actuates to the isolation position on an actual or simulated actuation signal.
Verify the closure time of each MFIV is within limits.
In accordance with the Surveillance Frequency Control Program In accordance with the INSERVICE TESTING PR()GRAM CALLAWAY PLANT 3.7-11 Amendment 222
SURVEILLANCE REQUIREMENTS SR 3.7.4.1 SR 3.7.4.2 SURVEILLANCE
~--------~--------~----- N()TE ----------------------------
()nly required to be performed in M()DES 1 and 2.
Verify one complete cycle of each ASD.
Verify one complete cycle of each ASD manual isolation valve.
CALLAWAY PLANT 3.7-14 ASDs 3.7.4 FREQUENCY In accordance with the INSERVICE TESTING PR()GRAM In accordance with the INSERVICE TESTING PR()GRAM Amendment No. 222
ACTIONS (continued)
CONDITION B.
Required Action and Associated Completion Time not met.
B.1 AND B.2 REQUIRED ACTION Be in MODE 3.
Be in MODE4.
SURVEILLANCE REQUIREMENTS SR 3.7.19.1 SR 3.7.19.2 SURVEILLANCE Verify the isolation time of each automatic SSIV is within limits.
Verify each automatic SSIV in the flow path actuates to the isolation position on an actual or simulated actuation signal.
CALLAWAY PLANT 3.7-47 SSIVs 3.7.19 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours FREQUENCY In accordance with the INSERVICE TESTING PROGRAM In accordance with the Surveillance Frequency Control Program Amendment No. 222
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.8 5.5.9 Not Used Steam Generator <SG} Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:
- a.
Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing (continued)
CALLAWAY PLANT 5.0-10 Amendment No. 222
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 222 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483
1.0 INTRODUCTION
By application dated March 22, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19081A173), as supplemented by letter dated May 21, 2019 (ADAMS Package Accession No. ML19141A249), Union Electric Company, dba Ameren Missouri (the licensee) requested changes to the technical specifications (TSs) for Callaway Plant, Unit No. 1 (Callaway). Specifically, the licensee requested changes to the TSs consistent with Technical Specifications Task Force (TSTF) Standard Technical Specifications (STSs)
Change Traveler TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR [Surveillance Requirement] Usage Rule Application to Section 5.5 Testing, dated October 21, 2015 (ADAMS Accession No. ML15294A555).
The licensees proposed changes would delete Callaway TS 5.5.8, Inservice Testing Program, and add a new defined term, INSERVICE TESTING PROGRAM, to the TSs. All existing references to the Inservice Testing Program in the Callaway TS SRs would be replaced with INSERVICE TESTING PROGRAM so that the SRs refer to the new definition in lieu of the deleted program.
The supplemental letter dated May 21, 2019, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on June 4, 2019 (84 FR 25841).
2.0 REGULATORY EVALUATION
An inservice test is a test to assess the operational readiness of a structure, system, or component after first electrical generation by nuclear heat. The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) provides requirements for inservice testing of certain components in light-water nuclear power plants. The ASME OM Code identifies the components subject to the testing (i.e., pumps,
valves, pressure relief devices, and dynamic restraints), responsibilities, methods, intervals, parameters to be measured and evaluated, criteria for evaluating results, corrective actions, personnel qualification, and recordkeeping. Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(f)(4), Inservice testing standards requirement for operating plants, requires, in part, that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the inservice test requirements (except design and access provisions) set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in 10 CFR 50.55a(f)(2) and (3) and that are incorporated by reference in 10 CFR 50.55a(a)(1)(iv), to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulation at 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, includes inservice examination and testing requirements for dynamic restraints (snubbers) depending on the applicability of the ASME OM Code or the ASME Boiler and Pressure Vessel Code (BPV Code). The facilitys TSs also prescribe inservice testing requirements and frequencies for ASME Code Class 1, 2, and 3 components.
The regulation in 10 CFR 50.55a(f)(5)(ii), IST [Inservice testing] program update: Conflicting IST Code requirements with technical specifications, states, in part, If a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program. TSTF-545, Revision 3, provides guidance to licensees on how to request license amendments that would eliminate conflicting requirements between 10 CFR 50.55a, Codes and standards, and the TSs. TSTF-545, Revision 3, proposes elimination of the Inservice Testing Program from the Administrative Controls section of the TSs. The TSs contain surveillances that require testing or test intervals in accordance with the Inservice Testing Program. The elimination of the Inservice Testing Program from the TSs could cause uncertainty regarding the correct application of these SRs. Therefore, TSTF-545, Revision 3, also proposes adding a new definition, INSERVICE TESTING PROGRAM, to the TSs, which would be defined as the licensee program that fulfills the requirements of 10 CFR 50.55a(f). TSTF-545, Revision 3, proposes replacement of existing uses of the term, Inservice Testing Program, with the defined term, as denoted by capitalized letters, throughout the TSs.
The NRC approved TSTF-545, Revision 3, by letter dated December 11, 2015 (ADAMS Package Accession No. ML15317A071) and published a notice of availability in the Federal Register on March 28, 2016 (81 FR 17208).
2.2 Proposed Technical Specifications Changes The licensee requested to delete TS 5.5.8 from the Administrative Controls section of TSs and replace it with the words Not Used. TS 5.5.8 currently states:
This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:
- a.
Testing frequencies applicable to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:
ASME OM Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days
- b.
The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities;
- c.
The provisions of SR 3.0.3 are applicable to inservice testing activities; and
- d.
Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.
SR 3.0.2 allows an extension of inservice testing intervals by up to 25 percent. If it is discovered that a surveillance associated with an inservice testing activity was not performed within the required interval, SR 3.0.3 allows the licensee to delay declaring the associated limiting condition for operation not met in order to perform the missed surveillance. The licensee did not request changes to SR 3.0.2 or SR 3.0.3.
The licensee requested to revise Callaway TS Section 5, Administrative Controls, Table of Contents by removing the term Inservice Testing Program and replacing it with Not Used.
This Table of Contents for Section 5 is not subject to review since it is not contained in the current licensing basis. This change is administrative in nature and therefore not discussed further in this SE.
The licensee requested to revise the Definitions section of TSs by adding the term, INSERVICE TESTING PROGRAM, with the following definition: The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f). The licensee also requested that all existing occurrences of Inservice Testing Program in TS SRs be replaced with INSERVICE TESTING PROGRAM, so that the SRs refer to the new definition in lieu of the deleted program.
2.3 Regulatory Requirements and Guidance The NRC staff considered the following regulatory requirements, guidance, and licensing information during its review of the proposed changes:
Technical Specifications Paragraph 50.36(c) of 10 CFR requires TSs to include the following categories: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. Section 50.36(c)(3) of 10 CFR states that [s]urveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Section 50.36(c)(5) of 10 CFR states that [a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
The NRC staffs guidance for review of the TSs is in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]
Edition, Chapter 16, Section 16, Technical Specifications, Revision 3, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared improved STSs for each of the LWR nuclear steam supply systems and associated balance-of-plant equipment systems. The licensees proposed amendment is based on TSTF-545, Revision 3, which is an NRC-approved change to the improved STSs. The NRC staffs review includes consideration of whether the proposed changes are consistent with TSTF-545, Revision 3. Special attention is given to TS provisions that depart from the improved STSs, as modified by NRC-approved TSTF travelers, to determine whether proposed differences are justified by uniqueness in plant design or other considerations so that 10 CFR 50.36 is met. In addition, the guidance states that comparing the change to previous STSs can help clarify the TS intent.
Inservice Testing Pursuant to 10 CFR 50.54, Conditions of licenses, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include inservice testing of pumps and valves at nuclear power reactors in accordance with the ASME OM Code as specified in 10 CFR 50.55a(f). The NRC regulations in 10 CFR 50.55a(f) state, in part:
Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements for preservice and inservice testing (referred to in this paragraph (f) collectively as inservice testing) of the ASME BPV Code and ASME OM Code as specified in this paragraph (f). Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions [referring to 10 CFR 50.55a(f)(1) through (f)(6)]....
The ASME OM Code is a consensus standard, which is incorporated by reference into 10 CFR 50.55a. During the incorporation process, the NRC staff reviewed the ASME OM Code requirements for technical sufficiency and found that the ASME OM Code inservice testing program requirements were suitable for incorporation into the NRCs rules.
The applicable NRC regulations are found in 10 CFR 50.55(a)(f)(5)(ii) which addresses a plants IST program.
NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves and Inservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants, Final Report, October 2013 (ADAMS Accession No. ML13295A020) provides guidance for the inservice testing of pumps valves, and dynamic restraints.
NUREG-0800, Chapter 3, Section 3.9.6, Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints, Revision 3, March 2007 (ADAMS Accession No. ML070720041), provides guidance and acceptance criteria for the NRC staff review of the inservice testing program for pumps, valves and dynamic restraints.
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensees application to determine if the proposed changes are consistent with the guidance, regulations, and licensing information discussed in Section 2.3 of this safety evaluation. In determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Among the considerations are whether the TSs, as amended, would provide the necessary administrative controls per 10 CFR 50.36(c)(5)
(i.e., provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner). In making its determination as to whether to amend the license, the NRC staff considered those regulatory requirements that are automatically conditions of the license through 10 CFR 50.54.
Where the regulations already condition the license, there is no need for a duplicative requirement in the TSs; the regulations provide the necessary reasonable assurance of the health and safety of the public.
3.1 Deletion of the Inservice Testing Program from the TSs TS 5.5.8 requires the licensee to have an inservice testing program that provides controls for inservice testing of ASME Code Class 1, 2, and 3 components (i.e., pumps and valves).
Through 10 CFR 50.54, the applicable requirements of 10 CFR 50.55a are conditions of every nuclear power reactor operating license issued under 10 CFR Part 50. These requirements include 10 CFR 50.55a(f), which specifies the requirements for the inservice testing of pumps and valves. Therefore, requiring the licensee to have an inservice testing program in TSs is duplicative of the license condition in 10 CFR 50.54. Thus, with the proposed TS changes, the licensee will still be required to maintain an inservice testing program in accordance with the ASME OM Code, as specified in 10 CFR 50.55a(f). For the reasons explained below, it is not necessary to have additional administrative controls in the TSs relating to the inservice testing program to assure operation of the facility in a safe manner.
Consideration of TS 5.5.8.a The ASME OM Code requires testing to normally be performed within certain time periods.
TS 5.5.8.a sets inservice testing frequencies more precisely than those specified in the ASME OM Code and applicable addenda (e.g., at least once per 31 days contrasted with monthly).
However, the NRC staff determined that the more precise inservice testing frequencies are not necessary to assure operation of the facility in a safe manner.
Consideration of TS 5.5.8.b TS 5.5.8.b allows the licensee to extend, by up to 25 percent, the interval between inservice testing activities, as required by TS 5.5.8.a and for other normal and accelerated frequencies specified as 2 years or less in the inservice testing program. Similar to TS 5.5.8.b, the NRC authorization of ASME Code Case OMN-20, Inservice Test Frequency, which permits the licensee to extend the inservice testing intervals specified in the ASME OM Code by up to 25 percent, has been published in the Federal Register on July 18, 2017 (82 FR 32934).
Additionally, in August 2017, the NRC incorporated paragraph (x), OM condition: ASME OM Code Case OMN-20, in 10 CFR 50.55a(b)(3), Conditions on ASME OM Code, to allow licensees to implement ASME OM Code Case OMN-20 with their applicable ASME OM Code of record. In early 2020, the NRC is planning to remove 10 CFR 50.55a(b)(3)(x) as unnecessary with the acceptance of Code Case OMN-20 in NRC Regulatory Guide (RG) 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code, and the incorporation by reference of that revision of RG 1.192 into 10 CFR 50.55a.
The NRC staff determined that the TS 5.5.8.b allowance to extend inservice testing intervals is not needed to assure operation of the facility in a safe manner. Therefore, the NRC staff determined that deletion of TS 5.5.8.b is acceptable. The deletion of TS 5.5.8.b does not impact the licensees ability to extend inservice testing intervals using Code Case OMN-20, as authorized by the NRC.
Consideration of TS 5.5.8.c TS 5.5.8.c allows the licensee to use SR 3.0.3 when it discovers that an SR associated with an inservice test was not performed within its specified frequency. SR 3.0.3 allows the licensee to delay declaring a limiting condition for operation not met in order to perform the missed surveillance. The use of SR 3.0.3 for inservice tests is limited to those inservice tests required by an SR. In accordance with 10 CFR 50.55a, the licensee may also request relief from the ASME OM Code requirements to address issues associated with a missed inservice test.
Deletion of TS 5.5.8.c does not change any of these requirements, and SR 3.0.3 will continue to apply to those inservice tests required by SRs. Based on the above, the NRC staff determined that deletion of TS 5.5.8.c is acceptable.
Consideration of TS 5.5.8.d TS 5.5.8.d states that nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. However, the NRC regulations in 10 CFR 50.55a(f)(5)(ii) address what to do if a revised inservice testing program for a facility conflicts with the TSs for the facility; they require the licensee to apply for an amendment to the TSs to conform the TSs to the revised program at least 6 months prior to the start of the period for which the provisions become applicable. Accordingly, there is no need for a TS stating how to address conflicts between the TSs and the inservice testing program because the regulations specify how conflicts must be resolved.
Conclusion Regarding Deletion of TS 5.5.8 The NRC staff determined that the requirements currently in TS 5.5.8 are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the NRC staff concludes that deletion of TS 5.5.8 from the licensees TSs is acceptable, because TS 5.5.8 is not required by 10 CFR 50.36(c)(5).
3.2 Definition of INSERVICE TESTING PROGRAM and Revision to SRs The licensee proposes to revise the TS Definitions section to include the term, INSERVICE TESTING PROGRAM, with the following definition: The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f). The proposed definition of the INSERVICE TESTING PROGRAM is consistent with the definition in TSTF-545, Revision 3. The definition is acceptable to the NRC staff because it correctly refers to the inservice testing requirements in 10 CFR 50.55a(f).
The licensee requested that all existing references to the Inservice Testing Program in SRs be revised to INSERVICE TESTING PROGRAM to reference the new TS defined term in lieu of the deleted program. The proposed change is consistent with the intent of TSTF-545, Revision 3, to replace the current references in SRs with the new definition. The NRC staff verified that for each SR reference to the Inservice Testing Program, the licensee proposed to change the reference to INSERVICE TESTING PROGRAM. The proposed change does not alter how the SR testing is performed. However, the inservice testing frequencies could change because the TSs will no longer include the more precise test frequencies in TS 5.5.8.a. As discussed in Section 3.1 of this safety evaluation, the NRC staff determined that the TSs do not need to include the more precise testing frequencies currently in TS 5.5.8.a. Based on its review, the NRC staff determined that revising the SRs to refer to the new definition is acceptable because these SRs will continue to be performed in accordance with the requirements of 10 CFR 50.55a(f). The NRC staff also determined that, with the proposed changes that allow less-precise testing frequencies, 10 CFR 50.36(c)(3) will continue to be met because the SRs will continue to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Missouri State official was notified of the proposed issuance of the amendment on January 28, 2020. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.
The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration published in Federal Register on June 4, 2019 (84 FR 25841), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Caroline Tilton, NRR Date: March 5, 2020
- via e-mail OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/STSB/BC NRR/DEX/EMIB/
(A)BC*
NAME JKlos PBlechman VCusumano TScarbrough for SBailey DATE 01/28/2020 02/19/2020 11/27/2019 02/03/2020 OFFICE OGC/NLO NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME AGhosh JDixon-Herrity JKlos DATE 03/03/2020 03/04/2020 03/05/2020