ML20087L803
| ML20087L803 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/08/2020 |
| From: | Thomas Wengert NRC/NRR/DORL/LPL4 |
| To: | Entergy Operations |
| Tom Wengert NRR/DORL 301-415-4037 | |
| References | |
| EPID L-2019-LLA-0285 | |
| Download: ML20087L803 (25) | |
Text
April 8, 2020 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 2 - ISSUANCE OF AMENDMENT NO. 320 TO REVISE CONTROL ELEMENT ASSEMBLY DROP TIMES (EPID L-2019-LLA-0285)
Dear Sir or Madam:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 320 to Renewed Facility Operating License No. NPF-6 for Arkansas Nuclear One, Unit 2 (ANO-2). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated December 18, 2019, as supplemented by letter dated March 11, 2020.
The amendment increases both the individual and average control element assembly drop time limits by 0.2 seconds to establish margin impacted by installation of new high temperature upper gripper coils associated with the control element drive mechanism for each control element assembly.
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/
Thomas J. Wengert, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-368
Enclosures:
- 1. Amendment No. 320 to NPF-6
- 2. Safety Evaluation cc: Listserv
ENTERGY OPERATIONS, INC.
DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 320 Renewed License No. NPF-6
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Entergy Operations, Inc. (the licensee), dated December 18, 2019, as supplemented by letter dated March 11, 2020, 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.
Enclosure 1
- 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-6 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 320, are hereby incorporated in the renewed license. The licensee 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 within 60 days from 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-6 and Technical Specifications Date of Issuance: April 8, 2020
ATTACHMENT TO LICENSE AMENDMENT NO. 320 RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ARKANSAS NUCLEAR ONE, UNIT 2 DOCKET NO. 50-368 Replace the following pages of the Renewed Facility Operating License No. NPF-6 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 3/4 1-23 3/4 1-23
3 (4) EOI, pursuant to the Act and 10 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; (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 or instrument calibration or associated with radioactive apparatus or components; and (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 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; 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 EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 3026 megawatts thermal. Prior to attaining this power level EOI shall comply with the conditions in Paragraph 2.C.(3).
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 320, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
Exemptive 2nd paragraph of 2.C.2 deleted per Amendment 20, 3/3/81.
(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following issuance of the renewed license or within the operational restrictions indicated.
The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.
2.C.(3)(a) Deleted per Amendment 24, 6/19/81.
Renewed License No. NPF-6 Amendment No. 320
REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual CEA drop time, from a fully withdrawn position, shall be d 3.9 seconds and the arithmetic average of the CEA drop times of all CEAs, from a fully withdrawn position, shall be d 3.4 seconds from when the electrical power is interrupted to the CEA drive mechanisms until the CEAs reach their 90 percent insertion positions with:
- a. Tavg t 525 °F, and
- b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2.
ACTION:
- a. With the CEA drop times determined to exceed either of the above limits, restore the CEA drop times to within the above limits prior to proceeding to MODE 1 or 2.
- b. With the CEA drop times within limits but determined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of all CEAs shall be demonstrated through measurement prior to reactor criticality:
- a. For all CEAs following each removal of the reactor vessel head,
- b. For specifically affected individuals CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and
- c. In accordance with the Surveillance Frequency Control Program.
ARKANSAS - UNIT 2 3/4 1-23 Amendment No. 84,94,100,169,275, 315,
Correction Letter dated 10/24/95
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 320 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT 2 DOCKET NO. 50-368
1.0 INTRODUCTION
By application dated December 18, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19352F266), as supplemented by letter dated March 11, 2020 (ADAMS Package Accession No. ML20076C412), Entergy Operations, Inc. (the licensee),
requested changes to the Technical Specifications (TSs) for Arkansas Nuclear One, Unit 2 (ANO-2).
The proposed changes would increase both the individual and average control element assembly (CEA) drop time limits by 0.2 seconds. The licensee stated that these TS changes are necessary to establish margin impacted by installation of new high temperature upper gripper coils associated with the control element drive mechanism (CEDM) for each CEA.
The supplemental letter dated March 11, 2020, 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 or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on February 25, 2020 (85 FR 10732).
2.0 REGULATORY EVALUATION
2.1 System Description At ANO-2, the CEDM control system is designed to control reactivity changes and shut down the reactor by regulating CEAs. The CEAs consist of five elements, each containing boron carbide pellets in a hollow tube. The CEA elements are clustered into groups of five fingers sharing a common CEDM. Each CEA is positioned via a magnetic jack CEDM mounted on the reactor vessel closure head. The magnetic jack CEDM uses induced magnetic fields to operate a mechanism for moving a CEA. Coils mounted in a coil stack assembly slide over the mechanism pressure housing and rest upon a locating shoulder. These coils provide the Enclosure 2
magnetic flux that operates the mechanical parts of the drive within the pressure housing.
Linear motion of these parts causes the operation of latching devices, which translate the motion of these parts to the CEA drive shaft. Driving and holding of the CEA occurs when power is sequentially applied to the coils. Each gripper sub-assembly moves between upper and lower stationary stops. The upper gripper is used to perform the lift part of the step and is engaged during holding. The lower gripper assembly holds the CEA during repositioning of the upper grippers.
All CEAs can be inserted rapidly by gravity into the core by interrupting power to the gripper coils within the CEDMs, which hold the individual CEAs. An automatic or a manual reactor trip signal could interrupt CEDM power, causing CEAs insertion into the core and adding negative reactivity.
2.2 Proposed Changes to the Technical Specifications The licensees proposed changes to TS Limiting Condition for Operation (LCO) 3.1.3.4, CEA Drop Time, would increase both the individual and average CEA drop time limits by 0.2 seconds.
TS LCO 3.1.3.4 currently reads:
The individual CEA drop time, from a fully withdrawn position, shall be [less than or equal to] 3.7 seconds and the arithmetic average of the CEA drop times of all CEAs, from a fully withdrawn position, shall be 3.2 seconds from when the electrical power is interrupted to the CEA drive mechanisms until the CEAs reach their 90 percent insertion positions with:
- a. Tavg [greater than or equal to] 525 °F [degrees Fahrenheit], and
- b. All reactor coolant pumps operating.
Revised TS LCO 3.1.3.4 would read (changes are in bold text):
The individual CEA drop time, from a fully withdrawn position, shall be 3.9 seconds and the arithmetic average of the CEA drop times of all CEAs, from a fully withdrawn position, shall be 3.4 seconds from when the electrical power is interrupted to the CEA drive mechanisms until the CEAs reach their 90 percent insertion positions with:
- a. Tavg 525 °F, and
- b. All reactor coolant pumps operating.
TS LCO 3.1.3.4 specifies the maximum permitted CEA drop times. The associated Surveillance Requirement 4.1.3.4 requires that the CEA drop time of all CEAs be demonstrated through measurement prior to reactor criticality. Measuring drop time is required for all CEAs following each removal and reinstallation of the reactor vessel head, for specifically affected individual CEAs following any maintenance on or modification to the CEA drive system that could affect the drop time of those specific CEAs, and at each refueling outage. Should the average CEA drop time or an individual CEA drop time exceed the LCO limits of TS 3.1.3.4, the reactor unit is not allowed to proceed to power operation (Mode 1) or startup (Mode 2).
2.3 Applicable Regulatory Requirements The regulation in Title 10 of the Code of Federal Regulations (10 CFR)
Section 50.36(a)(1) requires each applicant for a license authorizing operation of a utilization facility to include proposed TSs in the application.
The regulation in 10 CFR 50.36(b) states, in part, that:
Each license authorizing operation of a utilization facility will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 [Contents of applications; technical information]. The Commission may include such additional technical specifications as the Commission finds appropriate.
The categories of items required to be in TSs are provided in 10 CFR 50.36(c). One such category is LCOs. LCOs are defined in 10 CFR 50.36(c)(2)(i), which states, in part, that:
Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.
The regulation in 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether an LCO is required to be included in TSs. The following two criteria are applicable to CEA drop times and state:
Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The regulation in 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, requires that a loss-of-coolant accident (LOCA) analysis must meet the emergency core cooling system performance acceptance criteria.
ANO-2 was designed and constructed to meet the intent of the Atomic Energy Commissions (AECs) general design criteria (GDC), as originally proposed in July 1967, and thus, the design and construction were initiated and proceeded to a significant extent based upon the criteria proposed in 1967. The ANO-2 Safety Analysis Report (SAR), Section 3.1, Conformance with AEC General Design Criteria, describes the manner in which the ANO-2 GDC meet the intent of the corresponding GDC published as Appendix A of 10 CFR Part 50 in 1971. The regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, include the following GDC applicable to CEA design requirements, as described in this license amendment request:
GDC 10, Reactor design, of 10 CFR Part 50, Appendix A, states that:
The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
GDC 15, Reactor coolant system design, of 10 CFR Part 50, Appendix A states that:
The reactor coolant system and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
3.0 TECHNICAL EVALUATION
In support of the proposed TS changes, the licensee provided, in Section 3.0, Technical Evaluation, of the license amendment request (LAR) dated December 18, 2019, the technical analysis, including the evaluation and reanalysis of the applicable Chapter 15 and non-Chapter 15 events discussed in the ANO-2 SAR Amendment 28 (ADAMS Accession No. ML19282B426), for NRC staff review and approval.
The NRC staff reviewed the proposed TS changes and the associated technical analyses provided by the licensee. The NRC staffs review was conducted to confirm that the licensees analyses for applicable events described in the ANO-2 SAR Chapter 15 and non-Chapter 15 events would remain acceptable for the conditions with the proposed revised CEA drop times.
The NRC staffs evaluation of the licensees analyses is provided in Sections 3.1.1 through 3.1.8 of this safety evaluation (SE).
3.1 SAR Chapter 15 Analyses The insertion of CEAs on a reactor trip signal, which would add negative reactivity to the core to shut down the unit, is credited in the analyses of several design-basis events discussed in SAR Chapter 15, Accident Analysis. In Section 3.1, SAR Chapter 15 Evaluations, of the Enclosure to the LAR, the licensee assessed all of the SAR Chapter 15 analyses to determine the impact of the proposed CEA drop times. Based on the licensees impact assessment, the SAR Chapter 15 events can be classified into the following categories:
Category 1 - Events that are not applicable to ANO-2; Category 2 - Events for which a reactor trip does not occur or is not credited; Category 3 - Events for which consequences are bounded by other Chapter 15 events; Category 4 - Events for which consequences are bounded by the Chapter 15 analysis of the same events; and Category 5 - Events for which consequences are affected by the CEA drop time changes.
For the events in Categories 1 through 4, discussed below, the licensee determined that the SAR analyses are not affected by the TS changes and provided a rationale for each event supporting its determination that a reanalysis was not needed. For Category 5 events, the licensee determined that the SAR analyses would be affected by the TS changes and provided the reanalysis of Category 5 events for the NRC to review and approve.
3.1.1 Category 1 Events The NRC staff determined that the following events are not applicable to ANO-2, as discussed below:
- 1. SAR Section 15.1.6 - Idle Loop Startup The reactor coolant system (RCS) at ANO-2 consists of two loops, with two reactor coolant pumps (RCPs) in each loop, connected in parallel to the reactor vessel. The idle loop startup event is defined as the startup of both RCPs in the same loop that were not in operation. This mode of operation is prohibited by LCO 3.4.1.1 in ANO-2 TS 3/4 4.1, Reactor Coolant Loops and Coolant Circulations Startup and Power Operation, which states that Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation. Therefore, the NRC staff determined that this event is not applicable to ANO-2.
The above three events are only applicable to BWRs. ANO-2 is a pressurized water reactor; therefore, these events are not applicable to ANO-2.
- 5. SAR Section 15.1.22 - Break in Instrument Line or Other Lines from Reactor Coolant Pressure Boundary that Penetrate Containment There are no instrument lines from the RCS that penetrate containment. Therefore, this event is not applicable to ANO-2.
3.1.2 Category 2 Events The NRC staff verified that the analyses for the following events are not affected by the proposed revised CEA drop times because a reactor trip does not occur or is not credited.
- 3. SAR Section 15.1.15 - Inadvertent Loading of a Fuel Assembly into the Improper Position
- 4. SAR Section 15.1.16 - Waste Gas Decay Tank Leakage or Rupture
- 5. SAR Section 15.1.23 - Fuel Handling Accident
- 6. SAR Section 15.1.24 - Small Spills or Leaks of Radioactive Material Outside Containment
- 7. SAR Section 15.1.25 - Fuel Cladding Failure Combined with Steam Generator Leak
- 8. SAR Section 15.1.26 - Control Room Uninhabitability
- 9. SAR Section 15.1.27 - Failure or Overpressurization of Low-Pressure Residual Heat Removal System
- 10. SAR Section 15.1.30 - Loss of Service Water System
- 12. SAR Section 15.1.32 - Inadvertent Operation of ECCS [Emergency Core Cooling System] during Power Operation
- 13. SAR Section 15.1.33 - Turbine Trip with Failure of Generator Breaker to Open
- 14. SAR Section 15.1.34 - Loss of Instrument Air System
- 15. SAR Section 15.1.35 - Malfunction of Turbine Gland Sealing System 3.1.3 Category 3 Events The NRC staff reviewed the respective Category 3 events and confirmed that the analyses are not affected by the proposed TS changes because they are bounded by other SAR Chapter 15 events, as discussed and evaluated below:
- 1. SAR Section 15.1.5 - Total and Partial Loss of Reactor Coolant Forced Flow The analysis of the partial loss of reactor coolant forced flow event would be bounded by that of the total loss of reactor coolant forced flow (loss of flow (LOF)) event discussed in SAR Section 15.1.5 because the reactor coolant flow coastdown and its effect on the thermal margin degradation of this event would be less severe than that for a total LOF event. In addition, the revised CEA drop times would not change the flow coastdown characteristics.
SAR Section 15.1.4.2.6, Dilution During Critical Operation, discusses the boron dilution events in Modes 1 and 2 and shows that the events are bounded by CEA withdrawal (CEAW) events because the rate of reactivity addition (and corresponding power increase) during a dilution event in Modes 1 and 2 operation is less than that of a CEA Bank Withdrawal event. The revised CEA drop time would not change the event characteristics. Therefore, the NRC staff concludes that the dilution events in Modes 1 and 2 would remain bounded by that of the CEAW events.
- 3. SAR Section 15.1.9 - Loss of All Normal and Preferred AC [Alternating Current] Power to the Station Auxiliaries SAR Section 15.1.9 discusses the analysis of the event with loss of all normal and preferred AC power to the station auxiliaries and indicates that the analysis would be bounded by the analysis of the loss of reactor coolant forced flow events for minimum departure from nucleate boiling ratio (DNBR) specified acceptable fuel design limit (SAFDL) (SAR Section 15.1.5), loss of external load (LOL) and/or turbine trip event for peak RCS and steam generator (SG) pressures (SAR Section 15.1.7), and the loss of normal feedwater flow event for minimum SG liquid inventory (SAR Section 15.1.8).
Therefore, because the revised CEA drop times would not change the event characteristics, the NRC staff concludes that the analysis of this event would be bounded by that of these three SAR events.
- 4. SAR Section 15.1.28 - Loss of Condenser Vacuum SAR Section 15.1.28 discusses the analysis of the loss of condenser vacuum event and indicates that the analysis would be comparable to or bounded by the LOL and/or turbine trip event in SAR Section 15.1.7. Since the revised CEA drop times would not change the event characteristics, the NRC staff concludes that this event would continue to be bounded by the LOL and/or turbine trip event.
- 5. SAR Section 15.1.29 - Turbine Trip with Coincident Failure of Turbine Bypass Valves to Open SAR Section 15.1.29 indicates that the analysis of the event of the turbine trip with coincident failure of turbine bypass valves to open would be bounded by the analysis for the SAR Section 15.1.7 event, the loss of external load and/or turbine trip event. Since the revised CEA drop times would not change the event characteristics, the NRC staff concludes that this event would continue to be bounded by the LOL and/or turbine trip event.
3.1.4 Category 4 Events The analysis of some Category 4 events credits the following reactor trips: (1) the core protection calculator (CPC) variable overpower trip (VOPT); (2) low SG level (LSGL) trip; (3) high pressurizer pressure trip (HPPT); (4) CPC RCP shaft trip; (5) CPC Tsat trip (CPC Hot-Leg (saturation) temperature); and (6) CPC Tcold trip (CPC trip). The licensee stated that the existing analysis of the Category 4 events remained valid for the proposed increase in the CEA drop time of 0.2 seconds (from 3.2 seconds to 3.4 seconds), because the trip response time modeled in the analysis for each trip was greater than the actual response time by at least 0.2 seconds. In its March 11, 2020, supplemental response to the NRC staffs request for additional information the current safety analysis modeled trip response times and actual response times for the above six trips are summarized as follows:
Current Safety Worst (Actual)
Analysis Modeled Measured Response Response time 2003 - 2018 Time Frame (seconds) (seconds)
CPC VOPT 0.6 0.14 Low SG Level Trip 1.3 0.78 HPPT 0.9 0.37 CPC RCP Shaft Speed Trip 1.0 0.29 CPC Tsat Trip 3.0 0.17 CPC Tcold Trip 0.6 0.16 The NRC staffs evaluation of the Category 4 events is provided below:
- 1. SAR Section 15.1.2 - Uncontrolled CEA Withdrawal from Critical Conditions SAR Section 15.1.2 discusses the analysis of the uncontrolled CEAW from critical conditions, which would add reactivity to the reactor core, causing the core power level to increase. A CEAW event could result from a malfunction in the CEDM control system or by operator error.
The existing analyses of the CEAW events credited the CPC VOPT for reactor protection to assure that the SAFDLs would not be exceeded. The CPC VOPT modeled a response time of 0.6 seconds.
The modeled trip response time of 0.6 seconds is greater than the actual response time of 0.14 seconds by 0.46 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times. Therefore, the NRC staff concludes that the existing analysis of this event remains valid.
- 2. SAR Section 15.1.8 - Loss of Normal Feedwater Flow SAR Section 15.1.8 discusses the analysis of the loss of normal feedwater flow event, which credits the LSGL trip with the trip response time of 1.3 seconds.
The modeled trip response time of 1.3 seconds is greater than the actual response time of 0.78 seconds by 0.52 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times. Therefore, the NRC staff concludes that the existing analysis of this event remains valid.
- 3. SAR Section 15.1.10 - Excess Heat Removal Due to Secondary System Malfunction SAR Section 15.1.10 discusses the analysis of the excess heat removal due to a secondary system malfunction event, which credits the CPCs VOPT. The CPCs VOPT modeled a response time of 0.6 seconds.
The modeled trip response time of 0.6 seconds is greater than the actual response time of 0.14 seconds by 0.46 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times. Therefore, the NRC staff concludes that the existing analysis of this event remains valid.
- 4. SAR Section 15.1.11 - Failure of the Regulating Instrumentation SAR Section 15.1.11 assesses malfunctions or failures of regulating/control systems that could result in deviations of plant process parameters from the prescribed values. Such deviations would initiate a reactor trip in the event that a core safety limit was approached. The revised CEA drop times would not impact the event characteristics, because a failure of the regulating instrumentation is not a SAR Chapter 15 stand-alone design-basis event. For example, a failure in the control instrumentation of the auxiliary feedwater system could result in the loss of feedwater event, which is a design-basis event. The change in the CEA drop times would not affect those event characteristics.
The information provided by the licensee in the LAR demonstrates that an increase in the CEA drop times does not impact all credible design-basis events. Therefore, the NRC staff concludes that the existing analysis would remain valid.
- 5. SAR Section 15.1.12 - Internal and External Events Including Major and Minor Fires, Floods, Storms, and Earthquakes The evaluation in SAR Section 15.1.12 indicates that the subject phenomena would not prevent safe shutdown of the plant. The proposed revised CEA drop times would not impact the characteristics of the subject phenomena to prevent safe shutdown of the
plant. Therefore, the NRC staff concludes that the existing evaluation of these phenomena would remain valid.
- 6. SAR Section 15.1.13 - Major Rupture of Pipes Containing Reactor Coolant up to and Including Double-Ended Rupture of Largest Pipe in the Reactor Coolant System (Loss of Coolant Accident)
On page 10 of the Enclosure to the LAR, the licensee discusses the effects of the revised CEA drop times on the analyses of the LOCA events, including LOCA, post-LOCA long-term cooling and small-break LOCA. For the large-break LOCA and post-LOCA long-term cooling, the analyses did not credit core reactivity due to CEA insertion. Therefore, the NRC staff concludes that the existing analyses would remain valid.
For the small-break LOCA, the analysis credits the CEA drop reactivity. The licensee compared the reactivity-versus-time curve used in the small-break LOCA analysis with the reactivity insertion curve representing the proposed revised CEA drop time, and confirmed that the current small-break LOCA analysis reactivity curve bounds that of the revised CEA drop time curve. Therefore, the NRC staff concludes that the existing analysis would remain valid.
As for the LOCA alternative source term dose releases, the analyses did not model the CEA drop time. Therefore, the NRC staff concludes that the existing analyses would not be impacted by the revised CEA drop times.
- 7. SAR Section 15.1.14 - Major Secondary System Pipe Breaks with or without a Concurrent Loss of AC Power SAR Section 15.1.14 discusses the analysis of the secondary system pipe breaks, which contains analyses for: (1) the steam line break (SLB) accident, and (2) the feedwater line break (FWLB) accident.
For the SLB accidents, the cases analyzed included different break locations for inside and outside containment, with and without a loss of offsite power, combined with different single failures at hot zero power and hot full power conditions. The analysis shows that the SLB post-trip accidents are dominated by the total scram reactivity added and the rate of the primary-side cooldown, accounting for the associated moderator and fuel feedback effects from the coolant and fuel temperature changes. The total reactivity added would not be changed due to the revised CEA drop times. Also, the primary-side cooldown was directly dependent on the secondary-side cooldown. Since the SG blowdown would not be impacted by the revised CEA drop times, there would be no change in the secondary-side cooldown. As there is no change to the secondary-side cooldown, there would be no change in the primary-side cooldown. Therefore, the NRC staff concludes that the revised CEA drop times would not have a significant impact on the SLB post-trip analysis.
The existing analysis of the FWLB event credits the LSGL and the HPPT. The LSGL trip modeled a trip response time of 1.3 seconds. The modeled response time of 1.3 seconds is greater than the actual response time of 0.78 seconds by 0.52 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times. For the HPPT, the FWLB analysis modeled a trip response time of 0.9 seconds. The modeled
response time of 0.9 seconds is greater than the actual response time of 0.37 seconds by 0.53 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times.
Both the LSGL trip response time and HPPT response time used in the FWLB analysis have adequate margin to offset the 0.2-second increase in the CEA drop times.
Therefore, the NRC staff concludes that the FWLB analysis would remain valid for the proposed revised CEA drop time conditions.
- 8. SAR Section 15.1.18 - Steam Generator Tube Rupture with or without a Concurrent Loss of AC Power On page 12 of the enclosure to the LAR, the licensee discusses the effects of the proposed revised CEA drop times on the existing analyses of the SG tube rupture (SGTR) accidents, which credit: (1) the CPCs RCP shaft speed - Low trip for the cases without AC power available, and (2) the CPCs Hot-Leg (saturation) temperature (Tsat) -
High trip for the cases with AC power available.
The CPC RCP shaft low speed trip modeled a response time of 1.0 seconds. The modeled trip response time of 1.0 seconds is greater than the actual response time of 0.29 seconds by 0.71 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times. Therefore, the NRC staff concludes that the existing analysis of the SGTR without AC power available cases remains valid for the proposed revised CEA drop time conditions.
The CPC Tsat trip modeled a response time of 3.0 seconds. The modeled trip response time of 3.0 seconds is greater than the actual response time of 0.17 seconds by 2.83 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times. Therefore, the NRC staff concludes that the existing analysis of the SGTR with AC power available cases and the SGTR event alternative source term dose analysis remain valid.
- 9. SAR Section 15.1.20 - Control Element Assembly Ejection SAR Section 15.1.20 discusses the analysis of the non-LOCA CEA ejection accidents, which credit the CPCs VOPT. The CPCs VOPT modeled a response time of 0.6 seconds. The modeled trip response time of 0.6 second is greater than the actual CPC VOPT response time of 0.14 seconds by 0.46 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times. Therefore, the NRC staff concludes that the existing analysis of this event remains valid.
- 10. SAR Section 15.1.36 - Transients Resulting from the Instantaneous Closure of a Single MSIV [Main Steam Isolation Valve]
SAR Section 15.1.36 discusses the analysis of the events resulting from the instantaneous closure of a single MSIV or LOL to one SG, which credits the CPC Tcold trip. The trip modeled a response time of 0.6 seconds. The modeled trip response time of 0.6 seconds is greater than the actual CPC Tcold response time of 0.16 seconds by 0.44 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times. Therefore, the NRC staff concludes that the existing analysis of this event remains valid.
3.1.5 Category 5 Events The following Category 5 events are impacted by the CEA drop time change and were reanalyzed by the licensee to demonstrate compliance with the applicable SAR Chapter 15 acceptance criteria.
- 1. SAR Section 15.1.1 - Uncontrolled Control Element Assembly Withdrawal from a Subcritical Condition
- 2. SAR Section 15.1.5 - Total and Partial Loss of Reactor Coolant Forced Flow
- 3. SAR Section 15.1.7 - Loss of External Load and/or Turbine Trip The NRC staffs evaluation of the Category 5 events is provided in Section 3.1.7 of this SE.
3.1.6 Summary of NRC Staff Evaluation of Licensees Basis for Event Categorization The NRC staff has reviewed the bases used for categorization of each event and found that all the applicable SAR Chapter 15 events were considered, and that the bases supporting the events classified in each of the five categories correctly reflects the effects of an increase in the CEA drop times. In addition, the NRC staff determined that the licensee adequately considered the results of the SAR Chapter 15 analysis regarding credit of a reactor trip, sensitivity of the CEA drop time, and bounding event identification. The NRC staff, therefore, concludes that the bases for determining the event impact is appropriate.
3.1.7 Reanalysis of Category 5 Events CEA Insertion Curves used in the Reanalysis SAR Section 15.1.0.2.3, Shutdown CEA Reactivity, indicates that the CEA insertion reactivity curves in SAR Figure 15.1.0-1D (representing CEA reactivity versus (vs.) insertion time) and Figure 15.1.0-1E (representing CEA insertion position vs. insertion time) were used in the transients and accidents analyses applicable to Cycle 13 and subsequent cycles. The curves were developed based on the assumption that the arithmetic average drop time to 90 percent inserted was 3.2 seconds, of which 0.6 seconds is attributed to holding coil delay time.
In its supplemental letter dated March 11, 2020, the licensee clarified that the SAR Figures 15.1.0-1D and 15.1.0-1E CEA insertion curves were used in the reanalysis, unchanged except that the time of the upper gripper coil release of the CEA was delayed by 0.2 seconds.
The NRC staff determined that the licensees use of the CEA insertion time curves discussed above in the reanalysis is acceptable, because: (1) the reanalysis of three events discussed in SE Sections 3.1.7.1, 3.1.7.2, and 3.1.7.3 below demonstrate that the results meet the applicable acceptance criteria, (2) prior to each fuel cycle, the licensee will review the key input parameters relative to the SAR Chapter 15 analyses, (3) pre-startup testing of the CEA drop times will ensure that the revised TS acceptance criteria are met, (4) upon NRC approval of the revised TS acceptance criteria, the licensee will revise station procedures and licensing basis documents (including the SAR) to reflect the reactor protection system response time changes upon which the revised CEA drop time acceptance criteria relied, and (5) the licensee will revise the reactor protection system response time acceptance criteria in the associated surveillance procedures.
3.1.7.1 Reanalysis of the SAR Section 15.1.1 Event In its letter dated March 11, 2020, the licensee clarified that the Cycle 16 core was the first core at the power uprate conditions.1 The analysis of this event is discussed in SAR Section 15.1.1.4.3, Cycle 16 CEA Withdrawal from Subcritical Conditions. Selected analysis assumptions are listed in SAR Table 15.1.1-7, Assumptions for the Cycle 16 Uncontrolled CEA Withdrawal from a Subcritical Condition. Subsequent to the power uprate, the fuel transition to next generation fuel (NGF) accounted for the NGF WSSV-T critical heat flux (CHF) correlation (ADAMS Accession No. ML080840015) over Cycles 20 (SAR Section 15.1.1.4.4) and 21 (SAR Section 15.1.1.4.5). The transition to NGF did not affect the analysis assumptions listed in SAR Table 15.1.1-7. The CEAW reanalysis started from the NGF analyses and revised only the CEA drop times discussed in SE Section 3.1.7 above. The reanalysis used the same methods and computer codes discussed in SAR Section 15.1.1.4.5, and performed the following two SAR limiting CEAW cases from subcritical conditions contained in SAR Table 15.1.1-7:
Case 1 -- Assumes the maximum reactivity insertion rate (RIR) of 2.5 x 10-4 /sec (a unit of reactivity insertion rate) and the total nuclear heat flux factor (Fq) of 6.8.
Case 2 -- Assumes the minimum RIR of 2.0 x 10-4 /sec and Fq of 9.0.
The differences between the above two limiting cases are the CEA RIRs and the total (maximum) nuclear heat flux factors (Fq) as seen above. These combinations of RIR and Fq would define the limits on the allowable core design, and the reload process will confirm that these limits were met in each cycle. These limits would not be affected by the change in CEA drop time.
The reanalysis shows that the results would not violate the safety minimum departure from DNBR limit of 1.23 or the fuel center line melting temperature limit of 4681 °F.
The SAR limiting CEAW events were reanalyzed using NRC-approved methods, computer codes, and CHF correlation, the same initial conditions for the SAR limiting cases, and the results were within the safety limits of the acceptable fuel performance. Therefore, the NRC staff concludes that the CEAW reanalysis is acceptable for the revised CEA drop times.
3.1.7.2 Reanalysis of the SAR Section 15.1.5 The LOF analyses in the SAR consists of the analyses of: (1) loss of forced reactor coolant flow and, (2) seized RCP rotor. The reanalysis of both events is discussed and evaluated in the following section.
- a. Reanalysis of Loss of Forced Reactor Coolant Flow Event The licensee previously performed, and the NRC approved, the analysis of the loss of forced reactor coolant flow event for the 2002 power uprate and the 2008 updates to account for the NGF WSSV-T CHF correlation. The reanalysis started from the NGF analyses and revised only the CEA drop time discussed in SE Section 3.1.7 above. The same methods and computer codes discussed in SAR Sections 15.1.5.2.3.7, Cycle 16 Loss of Coolant Flow Resulting from an Electrical Failure, and 15.1.5.2.3.9, Cycle 21 1
The NRC staff approved the ANO-2 extended power uprate amendment (Amendment No. 244) on April 24, 2002 (ADAMS Accession No. ML021130826).
Loss of Coolant Flow Resulting from an Electrical Failure, and initial conditions in SAR Tables 15.1.5-12, Four Reactor Coolant Pump Flow Coastdown Resulting from an Electrical Failure, and 15.1.5-13, Assumptions for the Cycle 16 Loss Of Coolant Flow Analysis, were used in the reanalysis.
For the LOF reanalysis discussed on page 7 of the enclosure to the LAR, the licensee indicated that an additional 2 percent DNBR margin of the 7 percent difference in analysis required overpower margin (ROPM) and core operating limits supervisory system ROPM was credited to partially offset the reduced DNBR margin due to the revised CEA drop times. The licensee clarified that ROPM was the quantity of overpower margin that is defined in the proprietary, non-public, version of the licensees letter dated March 11, 2020, set aside to ensure that the DNBR SAFDL would not be violated by an anticipated operational occurrence. The limiting transient analyses defined ROPM requirements, and the reload process will confirm that the ROPM requirements were met in each cycle. The limiting ROPM event initiated from 100 percent power was the seized rotor event, whereas the LOF was a non-limiting event with respect to ROPM. The difference in ROPMs between the seized rotor and the LOF was the 7 percent value referenced above. The increase in CEA drop time would increase the thermal margin degradation resulting from the LOF. To provide assurance that the DNBR SAFDL was not violated, the ROPM for the LOF was increased by the 2 percent DNBR margin referenced above. Although the difference between the LOF and the limiting seized rotor event is reduced, the LOF continues to be a non-limiting event with respect to the setting of the ROPM.
The reanalysis showed that the results would not violate the safety DNBR limit of 1.23.
Since the SAR limiting case was reanalyzed using the NRC-approved methods, computer codes, and CHF correlation, and the same initial conditions for the SAR limiting cases, and the results were within the safety DNBR limits, the NRC staff concludes that the reanalysis of the loss of forced reactor coolant flow event is acceptable for the revised CEA drop times.
- b. Reanalysis of Seized RCP Rotor Event As indicated in SAR Section 15.1.5, the licensees analysis conservatively assumed an instantaneous reduction in flow to the rate for three RCS pump flow. Therefore, only the three-pump flow fraction was important in calculating potential fuel failure. Since no credit was assumed for the heat flux reduction due to the revised CEA drop times, the change in CEA drop time would have no impact on the calculation of potential fuel failure.
The licensee evaluated the effect of the proposed CEA drop time changes on a representative seized rotor case based on a conservative response time delay of 0.5 seconds and a holding coil delay time of 0.6 seconds, which would equate to a total delay time of 1.1 seconds. The revised rod drop time (holding coil delay of 0.8 seconds) and an assumed response time of 0.4 seconds resulted in a total response time of 1.2 seconds. This increase in total response time of 0.1 seconds would have a very small impact on the representative seized rotor case, since it would result in an increase of approximately 0.1 seconds for the estimated time to minimum DNBR.
For the peak primary pressure consideration, the LOL/turbine trip events were the limiting events (discussed in Section 3.1.7.3 of this SE) and demonstrate that for a 0.1-second increase in upper gripper coil delay time, the increase in primary pressure would be insignificant.
In summary, the seized rotor transient analyses are not impacted by the revised CEA drop times. Therefore, the NRC staff concludes that the existing analysis of the seized rotor event remains valid.
3.1.7.3 Reanalysis of the SAR Section 15.1.7 SAR Section 15.1.7 analyzes the LOL concurrent with a loss of feedwater event, which would bound the LOL, turbine trip and the loss of condenser vacuum event. The licensee previously performed, and the NRC approved, the analysis of the LOL event for the 2002 ANO-2 power uprate and the 2008 updates to account for the NGF WSSV-T CHF correlation. In support of the revised CEA drop times, the licensee reanalyzed the event starting from the NGF analyses and revised only the CEA drop time discussed in Section 3.1.7 above. The reanalysis used the same methods and computer codes discussed in SAR Section 15.1.7.4.2, Cycle 15 at 3026 MWt [MegaWatt Thermal] Loss of External Load and/or Turbine Trip, and was based on the same initial conditions in SAR Table 15.1.7-6.
For the LOL reanalysis, the licensee first baselined the existing analysis to the most recent code version of the current NRC-approved CENTS2 model listed in the core operating limits report, and then reanalyzed the LOL event based on the revised CEA drop times to determine the impact of the drop time change. The results showed that the reanalysis continued to be within the limits of peak primary pressure of 2750 pounds per square inch absolute (psia), the secondary peak pressure of 1210 psia, and the peak pressurizer water volume of 1200 cubic feet.
The LOL event was reanalyzed by using NRC-approved methods and computer codes, and the same initial conditions for the SAR limiting case, and the results met the pressure limits and pressurizer water volume limit. Therefore, the NRC staff concludes that the LOL reanalysis is acceptable for the revised CEA drop times.
For the main steam safety valves out-of-service cases, the existing analysis credits LSGL trip for event mitigation. The analysis assumes a trip response time of 1.3 seconds for the LSGL.
The modeled trip response time of 1.3 second is greater than the actual response time of 0.78 seconds by 0.52 seconds, which is sufficient to offset the 0.2-second increase in the CEA drop times. For a subset of the main steam safety valves out-of-service cases, the limiting condition would occur well before the reactor trip occurred. For these cases, the increase in the upper gripper coils decay time would have no impact on the limiting conditions. Therefore, the NRC staff concludes that the existing analysis of this event remains valid.
2 As described in ANO-2 SAR Section 15.1.0.6.10, CENTS is an interactive, high fidelity computer code for simulation of the Nuclear Steam Supply System (NSSS) components. It calculates the transient behavior of a pressurized-water reactor for normal and abnormal conditions, including accidents. CENTS also determines the core power and heat transfer throughout the NSSS, and computes the thermal and hydraulic behavior of the reactor coolant in the primary and secondary systems.
3.1.8 SAR Non-Chapter 15 Analyses 3.1.8.1 SAR Section 5.2.2 - Overpressure Protection SAR Section 5.2.2 discusses the overpressure protection of the RCS and SGs, and indicates that the peak RCS and SG pressures used in the overpressure protection report are based on the analysis of the limiting overpressure event, which is the LOL/turbine trip event documented in SAR Section 15.1.7. As discussed in SE Section 3.1.7.3 above, the reanalysis of the LOL event shows that analysis of the event with the proposed revised CEA drop time continues to meet the safety limits of the RCS peak pressure, SG peak pressure, and the pressurizer surge following the reactor trip. Therefore, the NRC staff concludes that the current overpressure protection report described in SAR Section 5.2.2 remains valid.
3.1.8.2 SAR Section 6.2.1 - Containment Functional Design SAR Section 6.2.1 discusses the containment functional design and indicates that the containment overpressure analysis is based on the inputs of the analysis of the mass and energy releases following a LOCA or main steam line break (MSLB). The licensee confirmed on page 16 of the Enclosure to the LAR, that the LOCA mass and energy release analysis does not credit core reactivity due to the CEA insertion.
For the MSLB, the licensee indicated that the mass and energy releases analysis would have a negligible impact from small changes in reactivity vs. time, because the primary driver for the event is the blowdown of the SG. The effects of the RCS and core transients are less important, and the small change in reactivity vs. time would affect only the first few seconds of the event. SAR Table 6.2-9C, Maximum Containment Pressure and Temperature Results (MSLB), shows that for the MSLB cases, the peak temperature occurs after 48 seconds, and the peak pressure after 140 seconds. The licensee confirmed that the time of reactor trip in the mass and energy release analysis varies case-by-case from approximately 2 to 5 seconds. The temperature and pressure peaks would thereby occur well after the reactor trip, and therefore, the increased CEA drop time would have a negligible effect on the analysis.
Based on the above evaluation, the NRC staff concludes that the containment analysis that was based on the inputs of LOCA and MSLB mass and energy releases would not be impacted by the CEA drop time change and remains valid.
3.1.8.3 Departure from Nucleate Boiling (DNB) Propagation The licensee previously performed, and NRC approved, a DNB propagation for the extended power uprate and subsequently updated to account for the NGF. The analysis shows that DNB propagation would not occur.
The limiting event for DNB propagation concerns was the seized rotor event. As discussed in Section 3.1.7.2.b of this SE, the analysis of the seized rotor event is not impacted by the revised CEA drop times. Therefore, the NRC staff concludes that the DNB propagation analysis and the associated results of the maximum strain and time in DNB performed for the NGF remain valid.
3.1.8.4 Fuel Cladding Burst The licensee previously performed, and NRC approved, a fuel cladding burst analysis for the extended power uprate and subsequently updated to account for NGF. The analysis demonstrated that cladding burst would not occur.
The fuel cladding burst analysis was based on the inputs from the analysis of the pre-trip MSLB and rod ejection events. The analysis of both events credited the CPCs VOPT for event mitigation. As discussed in Section 3.1.4 of this SE, because of the allocation of response time margin of the CPCs VOPT, the transient results of both events that were provided as inputs to the clad burst analysis would not be impacted. Therefore, the NRC staff concludes that the fuel cladding burst analysis performed for the NGF remains valid.
3.1.8.5 Fuel Cladding Strain The licensee previously performed, and NRC approved, a fuel cladding strain analysis for the extended power uprate and the NGF licensing applications. The NRC-approved cladding strain analyses used the transient results for the analysis of the hot zero power CEA withdrawal and the rod ejection events as inputs. The analysis of both hot zero power CEA withdrawal and rod ejection events both credit the CPCs VOPT for event mitigation. As described in Section 3.1.4 of this SE, because of the allocation of response time margin of the CPCs VOPT, the transient results that were provided as inputs to the cladding strain analysis would not be impacted by the CEA drop time increase. Therefore, the NRC staff determined that the fuel cladding strain analysis performed for NGF remains valid.
3.2 Summary of Technical Evaluation The licensee has proposed to modify ANO-2 TS LCO 3.1.3.4 to increase the maximum arithmetic average of all CEA drop times from 3.7 seconds to 3.9 seconds and increase the maximum individual CEA drop times from 3.2 seconds to 3.4 seconds. The NRC staff finds that most of the existing SAR Chapter 15 analyses and relevant non-Chapter 15 analyses would not be affected by the CEA drop time increase and remain valid. For the reanalysis of three events that would be affected by the CEA drop time increase, the licensee used a methodology previously-approved by the NRC staff. The results of the licensees reanalyses, reflecting the proposed increased CEA drop times, meet the requirements of GDC 10 related to the fuel integrity, GDC 15 related to the reactor coolant pressure boundary limits, and 10 CFR 50.46 related to emergency core cooling system performance acceptance criteria. Additionally, the inclusion of the proposed CEA drop times in the TSs meets the 10 CFR 50.36 requirements specified in Criterion (2) regarding initial conditions used in the SAR Chapter 15 analysis and relevant non-Chapter 15 analysis, and Criterion (3) regarding the systems, structures and components credited in the SAR Chapter 15 analysis and relevant non-Chapter 15 analyses for consequences mitigation. Therefore, the NRC staff concludes that the proposed revision to TS LCO 3.1.3.4 is acceptable.
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION The NRC proposed to find that the requested amendment involves no significant hazards consideration in its Federal Register notice of February 25, 2020 (85 FR 10732). The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
The licensee presented an evaluation of the issue of no significant hazards consideration in its December 18, 2019, application, which is quoted below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change increases the TS 3.1.3.4 average and individual CEA drop time limits. The CEA drop time is required to be verified prior to Modes 1 or 2 of plant operations. The probability of an accident previously evaluated remains unchanged since the CEAs will continue to insert into the core as a result of an accident or transient condition, and CEA drop time does not in itself initiate an accident.
The proposed change to the CEA drop time requirements have been evaluated for impact on the accident analyses. The accident analyses assumptions remain valid and, therefore, accident analysis results remain within applicable regulatory acceptance criteria.
Based on the above, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or involve a change in the methods governing plant operation. The ANO-2 Safety Analysis Report (SAR) has previously evaluated conditions where a reactor trip fails to occur upon a valid signal or where CEAs fail to insert following a reactor trip. The proposed change to CEA drop times does not affect the SAR assumptions respective to these failure modes. In addition, CEA drop times are not associated with accident initiators.
Therefore, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The increase in CEA drop times as proposed has been determined to have no significant impact on the accident analyses described in the SAR
and, therefore, the proposed change does not result a significant reduction on the existing margins of safety for the fuel, the fuel cladding, the reactor coolant system boundary, or the containment building. The change in CEA drop time does not impact the fuel rod design or mechanical design analysis. The slightly slower drop time would produce a smaller impact on the fuel assembly and lower stresses on the CEAs.
The accident analysis consequences are slightly more adverse, but all remain within the regulatory acceptance limits.
Therefore, this change does not involve a significant reduction in a margin of safety.
Based on the above evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Arkansas State official was notified of the proposed issuance of the amendment on March 19, 2020. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes requirements with respect to installation or use of facility components 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 February 25, 2020 (85 FR 10732), and there has been no public comment on such finding. Further, the Commission has made a final determination that no significant hazards consideration is involved for the proposed amendment as discussed in Section 4.0 of this safety evaluation. 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.
7.0 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: S. Sun Date: April 8, 2020
ML20087L803 *by e-mail OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA* NRR/DSS/SNSB/BC(A)* NRR/DSS/STSB/BC*
NAME TWengert PBlechman JBorromeo VCusumano DATE 3/31/2020 3/31/2020 3/26/2020 4/1/2020 OFFICE OGC* - NLO w/comments NRR/DORL/LPL4/BC* NRR/DORL/LPL4/PM*
NAME KGamin JDixon-Herrity TWengert DATE 4/6/2020 4/8/2020 4/8/2020