ML24031A644
| ML24031A644 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/14/2024 |
| From: | Thomas Wengert Plant Licensing Branch IV |
| To: | Entergy Operations |
| Wengert T | |
| References | |
| EPID L-2023-LLA-0048 | |
| Download: ML24031A644 (1) | |
Text
March 14, 2024 ANO Site Vice President Arkansas Nuclear One Entergy Operations, Inc.
N-TSB-58 1448 S.R. 333 Russellville, AR 72802
SUBJECT:
ARKANSAS NUCLEAR ONE, UNIT 1 - ISSUANCE OF AMENDMENT NO. 282 TO MODIFY TECHNICAL SPECIFICATION 3.3.1, REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION, TURBINE TRIP FUNCTION ON LOW CONTROL OIL PRESSURE (EPID L-2023-LLA-0048)
Dear Site Vice President:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 282 to Renewed Facility Operating License No. DPR-51 for Arkansas Nuclear One, Unit 1 (ANO-1). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 30, 2023.
The amendment revises ANO-1 TS 3.3.1, Reactor Protection System (RPS) Instrumentation, table 3.3.1-1, Reactor Protection System Instrumentation, to reflect plant modifications to the RPS instrumentation associated with reactor trip on main turbine trip on low fluid oil pressure.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Thomas J. Wengert, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-313
Enclosures:
- 1. Amendment No. 282 to DPR-51
- 2. Safety Evaluation cc: Listserv
ENTERGY OPERATIONS, INC.
DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 282 Renewed License No. DPR-51
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee), dated March 30, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and 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. DPR-51 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 282, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications.
- 3.
This amendment is effective as of its date of issuance and shall be implemented prior to startup from refueling outage 1R31 (spring 2024), coincident with the plant modifications to be performed in 1R31.
FOR THE NUCLEAR REGULATORY COMMISSION Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. DPR-51 and the Technical Specifications Date of Issuance: March 14, 2024 Jennivine K. Rankin Digitally signed by Jennivine K. Rankin Date: 2024.03.14 15:01:45 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 282 RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313 Replace the following pages of Renewed Facility Operating License No. DPR-51 and the 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 3
3 Technical Specifications REMOVE INSERT 3.3.1-5 3.3.1-5
(5)
EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6)
EOI, pursuant to the Act and 10 CFR Parts 30 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 following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 282, are hereby incorporated in the renewed license.
EOI shall operate the facility in accordance with the Technical Specifications.
(3)
Safety Analysis Report The licensees SAR supplement submitted pursuant to 10 CFR 54.21(d),
as revised on March 14, 2001, describes certain future inspection activities to be completed before the period of extended operation. The licensee shall complete these activities no later than May 20, 2014.
(4)
Physical Protection EOI shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans, including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: Arkansas Nuclear One Physical Security Plan, Training and Qualifications Plan, and Safeguards Contingency Plan, as submitted on May 4, 2006.
Renewed License No. DPR-51 Amendment No. 282 Revised by letter dated July 18, 2007
RPS Instrumentation 3.3.1 ANO-1 3.3.1-5 Amendment No. 215,264, Table 3.3.1-1 Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS CONDITIONS REFERENCED FROM REQUIRED ACTION C.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- 1. Nuclear Overpower -
a.
High Setpoint b.
Low Setpoint 1,2(a),3(d) 2(b),3(b) 4(b),5(b)
D E
SR 3.3.1.1 SR 3.3.1.2 SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 104.9% RTP 5% RTP
- 2. RCS High Outlet Temperature 1,2 D
SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 618 °F
- 3. RCS High Pressure 1,2(a),3(d)
D SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 2355 psig
- 4. RCS Low Pressure 1,2(a)
D SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 1800 psig
- 5. RCS Variable Low Pressure 1,2(a)
D SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 As specified in the COLR
- 6. Reactor Building High Pressure 1,2,3(c)
D SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 18.7 psia
- 7. Reactor Coolant Pump to Power 1,2(a)
D SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 55% RTP with one pump operating in each loop.
- 8. Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE 1,2(a)
D SR 3.3.1.1 SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.5 As specified in the COLR
- 9. Main Turbine Trip (Hydraulic Oil Pressure) 45% RTP F
SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 700 psig
SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 55.5 psig
- 11. Shutdown Bypass RCS High Pressure 2(b),3(b) 4(b),5(b)
E SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.5 1720 psig (a)
When not in shutdown bypass operation.
(b)
During shutdown bypass operation with any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.
(c)
With any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.
(d)
With any CRD trip breaker in the closed position, the CRD system capable of rod withdrawal, and not in shutdown bypass operation.
282
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 282 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313
1.0 INTRODUCTION
By "1CAN032301, License Amendment Request to Modify the Arkansas Nuclear One, Unit 1, Technical Specification 3.3.1, Reactor Protection System Instrumentation, [[Equipment trip" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Function on Low Control Oil Pressure|letter dated March 30, 2023]] (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23089A261), Entergy Operations, Inc. (the licensee) proposed changes to the Technical Specifications (TSs) for Arkansas Nuclear One, Unit 1 (ANO-1).
The proposed changes would revise ANO-1 TS 3.3.1, Reactor Protection System (RPS)
Instrumentation, table 3.3.1-1, Reactor Protection System Instrumentation, to reflect plant modifications to the RPS Instrumentation associated with reactor trip on main turbine trip on low fluid oil pressure. The proposed change to the TS is due to the replacement and relocation of the pressure switches from the main turbine low pressure auto-stop (AST) oil header that operates at approximately 100 pounds per square inch gauge (psig) to the high pressure turbine electrohydraulic control (EHC) oil header that operates at approximately 1800 psig to 1900 psig.
The licensee stated that the main turbine trip oil pressure TS allowable value must be increased to be consistent with and to accommodate the higher EHC oil pressure.
2.0 REGULATORY EVALUATION
2.1
System Description
In section 3.1, System Description, of the enclosure to its license amendment request (LAR),
the licensee provided the following description of the system:
The Main Turbine oil pressure trip function anticipates the loss of heat removal capabilities of the secondary system following a turbine trip. This trip function acts to minimize the pressure and temperature transient on the reactor and RCS
[reactor coolant system]. A turbine trip from a power level below 45% [percent]
power will not directly initiate a reactor trip. Four pressure switches (one per each RPS channel) [referred to as anticipatory reactor trip system (ARTS) pressure switches] monitor the ETH [emergency trip header] pressure in the Main Turbine EHC system. A low pressure condition of [less than or equal to] 850 psig sensed by two-out-of-four pressure switches with reactor power [greater than
or equal to] 45% rated thermal power (RTP) will initiate a reactor trip. These pressure switches do not provide any input to the non-nuclear instruments or integrated control system.
The EHC system supplies hydraulic control oil fluid to the turbine throttle valves, governor valves, reheat stop valves, and reheat intercept valves. The existing turbine protection system consists of the auto-stop oil system and the EHC system. Auto-stop oil is composed of the same oil used for lubrication of the main turbine and provides a trip interface between critical turbine functions and the EHC system. On a turbine trip signal, the auto-stop oil system line is depressurized by the actuation of protective devices, solenoid trip valves, or an emergency trip valve when a turbine trip condition exists. [However, the licensee proposes to change the sensing of a turbine trip by monitoring the ETH pressure rather than the turbine auto-stop header].
The EHC fluid is an incompressible fluid and when the solenoid/emergency trip
[solenoid valve trip block assembly] valves are opened, the dump valves at each turbine governor valve, throttle valve, reheat stop valve, and reheat intercept valve are depressurized due to the high-pressure EHC fluid to these valves actuators being released to drain back to the EHC fluid reservoir. The new
[Westinghouse] Ovation TCS [turbine control system] uses a separate set of three pressure transmitters to sense ETH pressure and initiate a turbine trip if EHC pressure decreases to 1000 psig. Two Testable Dump Manifolds (TDMs) replace the previously installed protective device trip interfaces. The trip signals will be sent to the TDMs which open to depressurize the EHC system. The governor, throttle, and reheat stop and intercept valves are spring actuated closed such that when the high pressure electrohydraulic (EH) fluid is removed from the valve actuators, the valves close instantaneously [under the spring force.]
The EHC fluid is provided by skid-mounted hydraulic pumps that maintain operating pressure at approximately 1800 - 1900 psig. The EHC fluid pressure is sensed by the new ARTS pressure switches. When the decreasing ETH pressure reaches 850 psig, as sensed by the pressure switches at 45% RTP, a reactor trip signal is initiated by at least two-out-of-four RPS channels. The circuitry and logic associated with the ARTS pressure switches and RPS operates in the same manner as the current auto-stop oil pressure switches, and it remains independent of the Ovation TCS.
The reactor trip on turbine trip is an anticipatory trip input signal to the RPS. This trip is anticipatory in that it is not assumed to occur in any of the [ANO-1] Safety Analysis Report (SAR) Chapter 6 or Chapter 14 accident analyses. This trip meets the requirements of IEEE [Institute of Electrical and Electronics Engineers] 279-1971, Criteria for Nuclear Power Plant Protection Systems,
[June 12, 1971] including [requirements for] separation, redundancy, single failure, and testability. The ARTS pressure switches are treated [by the licensee]
as safety-related and are seismically qualified, although the location in the non-seismic Turbine Building does not permit full qualification.
2.2 Pressure Switch Configuration In sections 3.2 and 3.3 of the enclosure to its LAR, the licensee provided a description of the current pressure switch configuration and the proposed new configuration, as described, in part, in the following summary:
The modification to the EHC system removes the low pressure auto-stop oil system, including the existing ARTS low auto-stop oil pressure switches. To support this modification, the RPS trip function will now be performed by four new pressure switches located on the Main Turbine ETH. Connected to the ETH is a Testable Dump Manifold (TDM) which depressurizes the ETH upon turbine trip.
A separate TDM is installed on the Overspeed Protection Control (OPC) header which serves as a backup for turbine protection. As with the original pressure switches, each of the four new pressure switches has an output contact that provides an input to its corresponding RPS instrumentation channel via a contact buffer located in the RPS channel's cabinet. The RPS logic is not affected by the proposed change. The new ARTS pressure switches will continue to initiate a reactor trip on a turbine trip when reactor power is 45% RTP as sensed on the power range nuclear instruments (NIs).
Additional details provided by the licensee can be found in sections 3.2 and 3.3 of the enclosure to the LAR.
2.3 Regulatory Requirements The regulations in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications, sets forth the regulatory requirements for the content of TSs. The regulation requires, in part, that the TSs include items in the following categories: (1) safety limits, limiting safety systems settings (LSSSs), and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. However, the regulation does not specify the particular requirements to be included in TSs. The regulation at 10 CFR 50.36(c)(1)(ii)(A) states, in part, that, Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.
Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen such that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor.
The regulation in 10 CFR 50.55a(h), Protection and safety systems, incorporates by reference the criteria of IEEE 279-1968, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, or IEEE 279-1971 for protection systems. The RPS is within the scope of this rule and the U.S. Nuclear Regulatory Commission (NRC) staff has performed an evaluation to determine whether the requirements will still be met following this upgrade.
The regulation in 10 CFR 50.65 (the Maintenance Rule) sets forth the regulatory requirements for monitoring the effectiveness of maintenance at nuclear power plants. The existing turbine control system, including the pressure switches that initiate the reactor trip on a turbine trip, are
within the scope of the Maintenance Rule and are monitored at the plant level. The RPS is also within the scope of the rule and is monitored accordingly.
ANO-1 was not licensed to the General Design Criteria (GDC) in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants. ANO-1 was originally designed to comply with the Atomic Energy Commission Proposed General Design Criteria for Nuclear Power Plant Construction Permits, published in July 1967 (1967 Proposed GDC). However, the licensee assessed the standards applied to the design, the plant design features, and plant procedures compared to most of the 10 CFR Part 50, Appendix A, GDC as discussed in section 1.4 of Amendment 31 of the ANO-1 SAR (ML23180A110). The Appendix A GDC relevant to this LAR are:
Criterion 13 - Instrumentation and control, states:
Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
Criterion 20 - Protection system functions, states:
The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
Criterion 22 - Protection system independence, states:
The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis.
Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.
Criterion 23 - Protection system failure modes, states:
The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air),
or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.
2.4 Regulatory Guidance Regulatory Guide (RG) 1.105, Setpoints for Safety-Related Instrumentation, Revision 4, dated February 2021 (ML20330A329), describes a method acceptable to the NRC staff for complying with the NRC's regulations for ensuring that setpoints for safety-related instrumentation are initially within and remain within the technical specification limits.
Regulatory Issue Summary (RIS) 2006-17, NRC Staff Position on the Requirements of 10 CFR 50.36, Technical Specifications, Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels, dated August 24, 2006 (ML051810077), discusses issues that could occur during testing of LSSSs and which, therefore, may have an adverse effect on equipment operability. This RIS also presents an approach, found acceptable to the NRC staff, for addressing these issues for use in licensing actions that require prior NRC staff approval.
2.5 Proposed Technical Specification Changes The licensee proposed the following changes to Function 9 in ANO-1 TS table 3.3.1-1:
The existing TS states (in part):
FUNCTION ALLOWABLE VALUE
- 9. Main Turbine Trip (Oil Pressure) 40.5 psig The proposed TS would state (in part):
FUNCTION ALLOWABLE VALUE
- 9. Main Turbine Trip (Hydraulic Oil Pressure) 700 psig As shown above, for main turbine trip on low fluid oil pressure, the allowable value would be revised from 40.5 psig to 700 psig and the functions title would be revised to specify hydraulic oil pressure.
3.0 TECHNICAL EVALUATION
3.1 Evaluation of the New Pressure Switch Configuration The proposed amendment to modify the allowable value for Main Turbine Trip (Oil Pressure) supports the replacement of the current analog electrohydraulic control system to a new digital EHC, Ovation TCS. The existing turbine trip function operates using pressure switches located at the AST oil header, which operates at 100 psig decreasing pressure with the turbine latched.
The modification to the EHC system removes the AST, which includes the existing low-pressure
switches. As a result of the modification, the RPS trip function will now be performed utilizing new pressure switches measuring pressure at the ETH, where it is maintained at approximately 1800-1900 psig under normal operations. Similar to the original pressure switches, when decreasing ETH pressure reaches the proposed setpoint of 850 psig decreasing, a reactor trip signal is initiated on a two-out-of-four RPS channel logic from the output contacts of the pressure switches causing a reactor trip. This RPS trip remains independent from the Ovation TCS and maintains the same logic and operating conditions as the existing configuration. The NRC staff finds the instrumentation and setpoint change discussed above acceptable to meet plant-specific GDC 20 because the protection system is not affected by the change, and therefore will continue to perform its design function to sense a turbine trip and automatically initiate an anticipatory reactor trip to minimize the RCS pressure and temperature transient for a loss of load transient.
As described above, in addition to the new pressure switch configuration, the new Ovation TCS will utilize a separate set of three pressure transmitters that will be used to initiate a turbine trip on low EHC header pressure. In case a leak on the EHC system takes place where ETH pressure gradually decreases, these pressure transmitters will sense the ETH pressure transient and initiate a turbine trip based on a two-out-of-three logic configuration when reaching a proposed setpoint of 1000 psig decreasing ETH pressure. This turbine trip setpoint has been selected sufficiently far from the RPS trip setpoint of 850 psig decreasing to ensure that there is enough separation margin between the potential maximum reactor protection trip initiation setpoint pressure and the minimum possible ETH low pressure setpoint to assure the anticipatory trip does not initiate a spurious reactor trip signal before a turbine trip is assured to have been initiated. The licensee stated that there is sufficient separation margin of 23.76 psig to assure the anticipatory trip does not initiate a reactor trip signal before a turbine trip has been initiated, inclusive of trip channel uncertainties.
The anticipatory RPS reactor trip, like the previous configuration, will be performed by four new pressure switches. These new switches have a higher measurement range than the previously installed switches, to be compatible with the ETH header pressure range instead of the AST header pressure range. Each pressure switch contains a dual-pole, dual-throw dry contact monitored by Class-1E contact buffer in its corresponding RPS cabinet. The new pressure switches will measure EHC fluid pressure located on the ETH, with a proposed decreasing setpoint of 850 psig to initiate the RPS reactor trip, where reaching that setpoint in the ETH is an indication that the turbine has tripped. The licensee stated that the sensing line taps for the current pressure switch configuration have been modified for the new pressure switches to sense ETH pressure. The NRC staff finds the instrumentation and setpoint change discussed above acceptable to meet the plant-specific GDC 13 because instrumentation is provided to monitor variables and systems over their anticipated ranges for the anticipated operational occurrence and includes appropriate controls to maintain them within prescribed operating ranges. The new pressure switch configuration will keep the same two-out-of-four logic and condition for the reactor trip at 45 percent RTP, as indicated by the power range neutron flux instrumentation and as found in table 3.3.1-1, Function 9 of the ANO-1 TSs.
The reactor trip on a sensed turbine trip is an anticipatory trip input signal to the RPS. In section 3.1 of the enclosure to the LAR, the licensee states, in part: This trip is anticipatory in that it is not assumed to occur in any of the Safety Analysis Report (SAR) Chapter 6 or Chapter 14 accident analyses. Therefore, this trip is not listed in table 7-13, Relationship Between Tested Trip Response Times and Trip Response Times Assumed in the Safety Analysis, of the ANO-1 SAR, and there are no response timing requirements for the initiation of this trip, given that it is not credited in the accident analysis. In its LAR, the licensee stated that there will be no
changes to the RPS logic and that the current trip response time for the RPS reactor trip remains unchanged at 150 milliseconds. Attachment 4, CALC-22-E-0007-01, Unit 1 Main Turbine Anticipatory Reactor Trip Setpoint Basis, of the enclosure to the LAR includes a technical manual (attachment1), which contains the vendor specification sheet associated with the proposed new pressure switches. The NRC staff notes that the vendor specification sheet for the new pressure switches, the SOR Incorporated, Measurement and Control (Model 9N6-BB45-U1-C1A-JJTTNQ), confirms that they are qualified to meet IEEE 323-1974 and -1983 IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations and IEEE 344-1974 and -1987 IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations. A pressure switch failure would result in a contact opening and input provided to the RPS channel in the same manner as an EHC header pressure drop below its specified setpoint, failing into a fail-safe state. The NRC staff finds the instrumentation and design discussed above acceptable to meet plant-specific GDC 23 because if a new pressure switch were to fail, the contacts would open and fail to a safe state, as designed, and therefore, would provide actuation input to the associated RPS channel.
The reactor RPS trip logic remains the same with a two-out-of-four logic functionality as is currently in use. The new pressure switches are qualified to IEEE 323 and 344, and the modification utilizing the existing sensing line taps to measure pressure now at the ETH.
Accordingly, the NRC staff finds that the RPS reactor trip will continue to meet the criteria of clause 4.4 of IEEE 279-1971. The licensee also stated that separation for each RPS channel and its associated wiring is maintained according to IEEE-279-1971. Given the above, the NRC staff finds this acceptable to meet the plant-specific GDC 22 because the change maintains the protective function of the reactor trip into RPS from low fluid oil pressure conditions.
To determine whether the new ARTS pressure switch setpoint has been selected to ensure that an automatic protective action will cause an anticipatory trip of the reactor before a reactor safety limit is exceeded during a transient that causes pressure depressurization, the NRC staff evaluated the licensees setpoint analysis information provided in the LAR (CALC-22-E-0007-01). In its calculations, the licensee considered parameters such as drift, calibration interval, as-found and as-left tolerances, minimum and maximum error values, reference accuracy, repeatability, component drift, temperature, pressure, measurement, and test equipment error for both the reactor trip associated with the pressure switches and the TCS turbine trip associated with the pressure transmitters to determine the setpoints. The establishment of both setpoints accounts for the total error using the parameters above, as detailed in the licensees calculations.
The licensee provided its setpoint determination calculations for both the turbine and reactor trips. Figure 1, as shown below, illustrates the relationship between the turbine trips provided by the transmitters/TCS, and the reactor trips provided by the pressure switches. This figure also indicates that there is a separation margin between the worst case (lowest likely) turbine trip and the worst case (highest likely) reactor trip setpoint of approximately 23.76 psig, which provides for the avoidance of spurious trips.
The regulation in 10 CFR 50.36(c)(1)(ii)(A) states, in part, that setpoints shall be selected to ensure that a safety limit is not exceeded. However, in this LAR the licensee did not identify an Analytical Limit associated with this anticipatory trip. The NRC staff evaluated the licensees reactor trip setpoint calculation to determine whether there is adequate margin between the nominal trip setpoint and the TS allowable value to account for expected drift, accuracy; measurement and test equipment uncertainty; and setting tolerance. The regulation in 10 CFR 50.36(c)(1)(ii)(A) also states, in part:
Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor.
During TS-required surveillances, the licensee will verify each reactor trip channel is functioning as required through conducting functional tests or calibration surveillances. Licensees can comply with the regulation by following approved setpoint methodologies in accordance with the guidance of RG 1.105; American National Standards Institute (ANSI)/ International Society of Automation (ISA) 67.04.01, Setpoints for Nuclear Safety-Related Instrumentation; and RIS 2006-17. A licensee following these guidance documents will evaluate the expected normal performance of a channel which accounts for reference accuracy, drift, and measurement and test equipment uncertainties. The NRC staff notes that the licensee calculated a total loop uncertainty of +/-107 psig that is being maintained between the selected nominal trip setpoint of 850 psig and the TS proposed allowable value of 700 psig, which allows for a margin of 43 psig.
This margin helps to ensure that a reactor trip has occurred prior to reaching the proposed Allowable Value. The setpoint determination for the new pressure switch configuration is based on the minimum required EHC fluid oil pressure, and is consistent with NUREG-1430, Standard
Technical Specifications Babcock and Wilcox Plants, Revision 5, dated September 2021 (ML21272A363), STS table 3.3.1-1 Function 9.
3.2 Technical Evaluation Summary Based on the above, the NRC staff finds that the proposed changes to the ANO-1 TS table 3.3.1-1, Function 9, Main Turbine Trip (Hydraulic Oil Pressure), to modify the Allowable Value from 40.5 psig to 700 psig, is acceptable. With the new pressure switch configuration, the low pressure sensed on the EHC high pressure header following a turbine trip will initiate an anticipatory reactor trip upon meeting the setpoint criteria. The staff also finds it acceptable to add the terms Hydraulic, for TS table 3.3.1-1, Function 9.
The NRC staff finds that the licensees setpoint methodology meets the methods acceptable for complying with the regulations for ensuring that setpoints for safety-related instrumentation are initially established and remain within the TS limits as described within RIS 2006-17 and RG 1.105, Revision 4. The licensee does not credit this anticipatory reactor trip for protection of fission product barriers. The staff also finds that the applicable regulatory requirements listed in section 2.3 of this safety evaluation and 10 CFR 50.36(c)(1)(ii)(A) will continue to be met.
Based on the review above, the NRC staff finds that with the proposed new pressure switch configuration, the applicable requirements listed in the plant-specific GDC as referenced in the ANO-1 SAR and the applicable requirements of 10 CFR 50.36(c)(1)(ii)(A) for TSs are met. The anticipatory trip would minimize the effects of a reactor coolant pressure and temperature transient for a loss of load transient. The setpoint calculations for the proposed changes were reviewed and are found to be acceptable, given they are conservative and follow the approved setpoint methodology outlined above, while also considering the separation between the turbine trip initiation and reactor trip setpoints to minimize the likelihood of any spurious reactor trips from occurring. The NRC staff concludes that the proposed amendment meets the regulatory requirements set forth in section 2.3 of this safety evaluation and is consistent with the guidance discussed in section 2.4, and therefore is acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Arkansas State official was notified of the proposed issuance of the amendment on January 30, 2024. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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 the Federal Register on July 11, 2023 (88 FR 44166), 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: G. Blas Rodriguez, NRR D. Rahn, NRR Date: March 14, 2024
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