ML23226A086
| ML23226A086 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 10/12/2023 |
| From: | Michael Mahoney Plant Licensing Branch II |
| To: | Flippin N Duke Energy Progress |
| Hood T | |
| References | |
| EPID L-2022-LLA-0137 | |
| Download: ML23226A086 (32) | |
Text
October 12, 2023 Ms. Nicole L. Flippin H. B. Robinson Steam Electric Plant Site Vice President Duke Energy Progress, LLC 3581 West Entrance Road, RNPA11 Hartsville, SC 29550
SUBJECT:
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENT NO. 277 REGARDING REVISION OF TECHNICAL SPECIFICATIONS TO ADD HIGH-HIGH STEAM GENERATOR LEVEL FUNCTION TO TABLE 3.3.2-1 AND REMOVE OBSOLETE CONTENT FROM TSs 2.1.1.1 AND 5.6.5.b (EPID L-2022-LLA-0137)
Dear Ms. Flippin:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 277 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson). This amendment would add a Feedwater Isolation on High-High Steam Generator Level function to Table 3.3.2-1 of Robinson Technical Specifications (TS) 3.3.2, Engineered Safety Feature Actuation System (ESFAS)
Instrumentation, and remove obsolete content from TS 2.1.1.1, Reactor Core SLs [Safety Limits], and 5.6.5.b, Core Operating Limits Report (COLR).
A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Michael Mahoney, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-261
Enclosures:
- 1. Amendment No. 277 to DPR-23
- 2. Safety Evaluation cc w/encls: Listserv DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 277 Renewed License No. DPR-23 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duke Energy Progress, LLC (the licensee),
dated September 21, 2022, as supplemented by letter dated February 9 and May 31, 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 3.B. of Renewed Facility Operating License No. DPR-23 is hereby amended to read, in part, as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 277 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: October 12, 2023 David J.
Wrona Digitally signed by David J. Wrona Date: 2023.10.12 09:42:57 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 277 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following page of Renewed Facility Operating License No. DPR-23 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Renewed Facility Operating License No. DPR-23 Remove Insert Page 3 Page 3 Replace the following page of the Appendix A, Technical Specifications with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 2.0-1 2.0-1 3.3-28 3.3-28 5.0-25 5.0-25 5.0-25a 5.0-25a 5.0-26 5.0-26 5.0-27 5.0-27 Renewed Facility Operating License No. DPR-23 Amendment No. 277 D.
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 and equipment calibration or associated with radioactive apparatus or components; E.
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 operation of the facility.
3.
This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR 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:
A.
Maximum Power Level The licensee is authorized to operate the facility at a steady state reactor core power level not in excess of 2339 megawatts thermal.
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 277 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
(1)
For Surveillance Requirements (SRs) that are new in Amendment 176 to Final Operating License DPR-23, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 176. For SRs that existed prior to Amendment 176, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 176.
SLs 2.0 HBRSEP Unit No. 2 2.0-1 Amendment No. 277 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest cold leg temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:
2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.141 for the HTP correlation.
2.1.1.2 The peak fuel centerline temperature shall be maintained < [4901 -
(1.37 x 10-3 x (Burnup, MWD/MTU))] °F.
2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained 2735 psig.
2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2 If SL 2.1.2 is violated:
2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.
ESFAS Instrumentation 3.3.2 HBRSEP Unit No. 2 3.3-28 Amendment No. 277 Table 3.3.2-1 (page 4 of 4)
Engineered Safety Feature Actuation System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE NOMINAL TRIP SETPOINT (1) 5.
Feedwater Isolation a.
Automatic Actuation Logic and Actuation Relays 1,2(f),3(f) 2 trains G
SR 3.3.2.2 SR 3.3.2.3 SR 3.3.2.5 NA NA b.
Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements c.
SG Water Level -
High-High 1,2(f),3(f) 3 per SG D
SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 76.16%
75%
6.
ESFAS Interlocks a.
Pressurizer Pressure Low 1,2,3 3
H SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 2005.11 psig 2000 psig b.
Tavg - Low 1,2,3 1 per loop H
SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 544.50 ºF 543 ºF (1)
A channel is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(f)
Except when all MFIVs, MFRVs, and bypass valves are closed or isolated by a closed manual valve.
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
(continued)
HBRSEP Unit No. 2 5.0-25 Amendment No. 277 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 7.
Axial Flux Difference (AFD) limits for Specification 3.2.3; 8.
Boron Concentration limit for Specification 3.9.1; 9.
Reactor Core Safety Limits Figure for Specification 2.1.1; 10.
Overtemperature T and Overpower T setpoint parameter values for Specification 3.3.1; and 11.
Reactor Coolant System pressure, temperature and flow Departure from Nucleate Boiling (DNB) limits for Specification 3.4.1.
12.
ECCS Accumulators boron concentration limits for Specification 3.5.1.
13.
ECCS Refueling Water Storage Tank boron concentration limits for Specification 3.5.4.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. The approved version shall be identified in the COLR. These methods are those specifically described in the following documents:
1.
Deleted 2.
Deleted 3.
Deleted 4.
Deleted 5.
XN-75-32(A), "Computational Procedure for Evaluating Rod Bow, approved version as specified in the COLR.
6.
Deleted 7.
Deleted 8.
Deleted
Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements (continued)
(continued)
HBRSEP Unit No. 2 5.0-25a Amendment No. 277 9.
Deleted 10.
Deleted 11.
Deleted 12.
Deleted 13.
Deleted
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
(continued)
HBRSEP Unit No. 2 5.0-26 Amendment No. 277 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 14.
Deleted 15.
Deleted 16.
Deleted 17.
Deleted 18.
Deleted 19.
Deleted 20.
EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as specified in the COLR.
21.
Deleted 22.
Deleted 23.
Deleted 24.
EMF-2103(P)(A), Realistic Large Break LOCA Methodology for Pressurized Water Reactors, approved version as specified in the COLR.
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
(continued)
HBRSEP Unit No. 2 5.0-27 Amendment No. 277 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 25.
Deleted 26.
BAW-10240(P)(A), Incorporation of M5 Properties in Framatome ANP Approved Methods, approved version as specified in the COLR.
27.
EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, approved version as specified in the COLR.
28.
DPC-NE-2005-P-A, Thermal-Hydraulic Statistical Core Design Methodology, approved version as specified in the COLR.
29.
DPC-NE-1008-P-A, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, as approved by NRC Safety Evaluation dated May 18, 2017.
30.
DPC-NF-2010-A, Nuclear Physics Methodology for Reload Design, as approved by NRC Safety Evaluation dated May 18, 2017.
31.
DPC-NE-2011-P-A, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors as approved by NRC Safety Evaluation dated May 18, 2017.
32.
DPC-NE-3008-P-A, "Thermal-Hydraulic Models for Transient Analysis," as approved by NRC Safety Evaluation dated April 10, 2018.
33.
DPC-NE-3009-P-A, "FSAR / UFSAR Chapter 15 Transient Analysis Methodology," as approved by NRC Safety Evaluation dated April 10, 2018.
34.
BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code, approved version as specified in the COLR.
c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 277 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DUKE ENERGY PROGRESS, LLC H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261
1.0 INTRODUCTION
By letter dated September 21, 2022, (ML22264A149) as supplemented by letters dated February 9 and May 31, 2023 (ML23040A426, and ML23151A251, respectively), Duke Energy Progress, LLC (Duke Energy, the licensee), submitted a license amendment request (LAR) to Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson or RNP). The LAR proposed a revision to the Robinson Technical Specifications (TS) to add a Feedwater Isolation on High-High Steam Generator Level function to Table 3.3.2-1 of TS 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, and remove obsolete content from TSs 2.1.1.1, Reactor Core SLs [Safety Limits], and 5.6.5.b, Core Operating Limits Report (COLR).
The supplements dated February 9 and May 31, 2023, provided additional information that clarified and corrected the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on November 29, 2022 (87 FR 73342).
2.0 PROPOSED CHANGE
S 2.1 Current TS and Proposed Changes 2.1.1 TS 2.1.1.1 and TS 5.6.5 Changes TS 2.1.1.1 specifies the Safety Limit for the departure from nucleate boiling ratio. TS 2.1.1.1 currently states:
The departure from nucleate boiling ratio (DNBR) shall be maintained 1.141 for the HTP [high thermal performance] correlation and 1.17 for the XNB correlation.
The proposed change to TS 2.1.1.1 would state:
The departure from nucleate boiling ratio (DNBR) shall be maintained 1.141 for the HTP correlation.
TS 5.6.5, Core Operating Limits Report (COLR), contains requirements related to core operating limits established prior to each reload cycle. TS 5.6.5.b provides a listing of analytical methods used to determine the RNP core operating limits. These limits are required to be documented in the COLR. The methods listed below would be deleted, with the text replaced with Deleted. The item numbering will remain the same.
- 2. XN-NF-84-73(P), Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events, approved version as specified in the COLR.
- 3. XN-NF-82-21(A), Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, approved version as specified in the COLR.
- 8. XN-NF-78-44(A), Generic Control Rod Ejection Analysis, approved version as specified in the COLR.
- 9. XN-NF-621(A), XNB Critical Heat Flux Correlation, approved version as specified in the COLR.
- 11. XN-NF-82-06(A), Qualification of Exxon Nuclear Fuel for Extended Burnup, approved version as specified in the COLR.
- 16. ANF-88-054(P), PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2, approved version as specified in the COLR.
- 17. ANF-88-133 (P)(A), Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 Gwd/MTU, approved version as specified in the COLR.
- 18. ANF-89-151(A), ANF-RELAP Methodology for Pressurized Water Reactors:
Analysis of Non-LOCA Chapter 15 Events, approved version as specified in the COLR.
- 19. EMF-92-081(A), Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors, approved version as specified in the COLR.
- 21. XN-NF-85-92(P)(A), Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results, approved version as specified in the COLR.
- 23. EMF-92-116, Generic Mechanical Design Criteria for PWR Fuel Designs, approved version as specified in the COLR.
- 25. EMF-2310(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, approved version as specified in the COLR.
2.1.2 TS 3.3.2 Changes The licensee proposed to change the Robinson, Unit 2 TS Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, to add function 5.c, SG Water Level - High-High under function 5, Feedwater Isolation, with the requirements and conditions as shown below:
Applicable Modes or other Specified Conditions Required Channels Conditions Surveillance Requirements Allowable Value Nominal Trip Setpoint (1) 1,2(f),3(f) 3 per SG D
SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 76.16%
75%
The proposed change would apply the following existing requirements:
Table 3.3.2-1 note (f) for applicable modes or other specified conditions: Except when all MFIVs [main feedwater isolation valves], MFRVs [main feedwater regulation valves], and bypass valves are closed or isolated by a closed manual valve.
Table 3.3.2-1 note (1) for Nominal Trip Setpoint: A channel is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
Condition D:
CONDITION REQUIRED ACTION COMPLETION TIME D.
One channel inoperable.
NOTE ----------
For Function 4.c, a channel may be taken out of the trip condition for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for maintenance.
D.1 Place channel in trip.
OR D.2.1 Be in MODE 3.
AND D.2.2 Be in MODE 4.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> Surveillance Requirements:
Surveillance Frequency SR 3.3.2.1 Perform CHANNEL CHECK.
In accordance with the Surveillance Frequency Control Program SR 3.3.2.4 Perform COT [CHANNEL.
OPERATIONAL TEST].
In accordance with the Surveillance Frequency Control Program SR 3.3.2.7 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program 2.1 Reason for the Proposed Change In Section 2.3 of the LAR, the licensee provided the following reason for the proposed changes:
While recently reevaluating the RNP UFSAR Chapter 15.1.2 increase in feedwater flow transient using Duke Energys in-house methods approved by the NRC in 2018 (Reference 8 [of the LAR]), it was recognized that SG overfill protection is needed to mitigate this event. A SG overfill event could potentially lead to either a steam line break that challenges the containment fission product barrier should primary-to-secondary leakage be present or fuel failure (i.e., the clad fission product barrier) as a result of a steam line break or stuck open SG relief valve. The concerns with SG overfill as stated in GL 89-19 include the following: (1) the increased dead weight and potential seismic loads placed on the main steam line and its support should the main steam line be flooded; (2) the loads placed on the main steam lines as a result of the potential for rapid collapse of steam voids resulting in water hammer; (3) the potential for secondary safety valves sticking open following discharge of water or two-phase flow; (4) the potential inoperability of the main steam line isolation valves (MSIVs), main turbine stop or bypass valves, feedwater turbine valves, or atmospheric dump valves from the effects of water or two phase flow.
The proposed change would revise the RNP TS to reflect removal of analytical methods that will no longer be used to determine the core operating limits, as these methods were replaced upon transitioning to NRC-approved Duke Energy methods.
3.0 REGULATORY EVALUATION
3.1 Description of Robinson Engineered Safety Feature Actuation System In RNP Unit 2, the Engineered Safety Feature Actuation System (ESFAS) instrumentation measures temperatures, pressures, flows, and levels in the Reactor Coolant System (RCS),
Reactor Containment, and Auxiliary Systems, and monitors their operation. Based on the values of selected unit parameters, the ESFAS initiates necessary safety systems, such as the Containment Isolation, Steam Line Isolation, Emergency Feedwater, to protect against violating core design limits and the RCS pressure boundary, and to mitigate accidents.
The operation of the ESFAS involves the actuation control functions, except for the SG level control function. The SG level control function is associated with plant cooldown using the auxiliary feedwater pump and involves remote manual positioning of feedwater flow control valves, in order to maintain proper SG water level. The feedwater system is designed to supply water to the SGs under all operating conditions (during normal power operation and during periods of shutdown or abnormal conditions).
In the LAR, the licensee stated, in part, that:
When the level in any SG exceeds the high-high water level setpoint of 75%
(also referred to at RNP as the high SG level valve interlock), existing SG overfill protection equipment responds by tripping the main turbine, tripping the MFW pumps, and closing the MFRV and bypass valves to the affected SG.
The existing SG overfill protection function is not part of the ESFAS; however, consistent with the ESFAS, the existing SG overfill protection instrumentation utilizes a 2 out of 3 channel initiating logic that is safety related.
An increase in feedwater flow can overfill a SG within approximately 5 minutes of the initiation of the event, and therefore automatic action is needed to terminate the event. The high-high SG level trip will prevent SG overfill and ensure that the acceptance criteria for the UFSAR Chapter 15.1.2 event are met. Therefore, the concerns presented by GL 89-19 remain valid and the high-high SG level trip should not have been removed from the TS as part of the STS conversion. This amendment request restores compliance with GL 89-19.
The UFSAR Section 15.1.2 increase in feedwater flow event has been reevaluated using DPC-NE-3009-P-A, FSAR/UFSAR Chapter 15 Transient Analysis Methodology, approved by the NRC per safety evaluation dated April 10, 2018 (Reference 8 [of LAR]). The transient is initiated by a failed-open MFRV to one SG, increasing MFW flow to that SG.
TS 3.3.2, specifies that the ESFAS instrumentation for each function in Table 3.3.2-1 shall be operable. TS Table 3.3.2-1 provides a list of all ESFAS functions and specifies the following for each function: (1) Applicable Modes or other specified Conditions, (2) Required Channels, (3)
Conditions, (4) Surveillance Requirements, (5) Allowable Value, and (6) Nominal Trip Setpoint.
The licensee proposed to add the SG overfill protection function (function 5.c) into this TS table.
In the LAR, the licensee noted that (1) there are no physical changes to plant required as a result of the proposed change, (2) this proposed change does not classify the SG overfill protection system as ESFAS equipment, and (3) the SG overfill protection function is added to the TS ESFAS table to restore compliance with GL 89-19. The licensee provided detail of the System Design, Operation, Background, and reasons for the proposed change of the ESFAS.
3.2 Regulatory Requirements and Guidance The NRC staff identified the following regulatory requirements and guidance as being applicable to the proposed amendment.
3.2.1 Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical Specifications, states regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The regulation does not specify particular requirements to be included in a plants TSs.
Regulations at 10 CFR 50.36(b) requires 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.
Regulations at 10 CFR 50.36(c)(1)(ii)(A) states, in part, 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 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.
Regulations at 10 CFR 50.36(c)(1), the safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down.
Regulations at 10 CFR 50.36(c)(2)(i), Limiting conditions for operation, states, in part, 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.
Regulations at 10 CFR 50.36(c)(2)(ii) requires establishment of a limiting condition for operation for every item that meets one or more of four criteria. Criterion 3 (10 CFR 50.36(c)(2)(ii)(C))
states, 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.
Regulations at 10 CFR 50.36(c)(3), Surveillance requirements states, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
Regulations at 10 CFR 50.36(c)(5) requires that:
Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in [10 CFR] 50.4.
NRC staff notes that:
In Robinsons Updated Final Safety Analysis Report (UFSAR), Chapter 3, Design of Structures, Components, Equipment, and Systems, the licensee stated, in parts:
The General Design Criteria (GDC) in existence at the time HBR [Robinson] 2 was licensed (July 1970) for operation were contained in Proposed Appendix A to 10CFR50, General Design Criteria for Nuclear Power Plants, published in the Federal Register on July 11, 1967. (Appendix A to 10CFR50, effective in 1971 and subsequently amended, is somewhat different from the proposed 1967 criteria.)
HBR 2 [Robinson, Unit 2] was evaluated with respect to the proposed 1967 GDC and the original FSAR contained a discussion of the criteria as well as a summary of the criteria by groups. Section 3.1.1.2 and 3.1.2 present that discussion without substantive change in order to preserve the original basis for licensing.
3.1.2.12 Instrumentation and Control Systems Criterion: Instrumentation and controls shall be provided as required to monitor and maintain within prescribed operating ranges for the essential reactor facility operating variables. (GDC 12) 3.2.2 Regulatory Guidance Regulatory Guide (RG) 1.105, Revision 4, Setpoints for Safety-Related Instrumentation, dated February 2021, (ADAMS Accession No. ML20330A329) describes a method acceptable to the NRC staff for complying with the NRCs regulations for ensuring that setpoints for safety-related instrumentation are initially within and remain within the TS limits. RG 1.105 Revision 4 endorses National Standards Institute/International Society of Automation (ANSI/ISA) Standard 67.04.01-2018, Setpoints for Nuclear Safety-Related Instrumentation. The NRC staff used this guide to establish the adequacy of the licensees setpoint calculation methodologies and the related plant surveillance procedures. Section 4.5, Combination of uncertainties, of ISA S67.04-2018, states that the methods in Subsection 4.5.1, Square-root-sum-of-squares (SRSS) method for random uncertainties, and in Subsection 4.5.2, Arithmetic method, are acceptable for combining non-random or bias uncertainties. This section also states that the alternate methods, including probabilistic modeling or stochastic modeling, or a combination of SRSS and arithmetic technique may also be used.
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 (ADAMS Accession No. ML051810077), discusses issues that could occur during testing of limiting safety system settings and therefore may have an adverse effect on equipment operability. The RIS also represents an approach that is acceptable to the NRC staff for addressing these issues for use in licensing actions.
NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 5.0 (ADAMS Accession No. ML21259A155), provides example TS LCOs and acceptable remedial actions that meet the requirements in 10 CFR 50.36(c)(2)(i) for a standard Westinghouse plant design.
NRC Generic Letter (GL) 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, insofar as it specifies acceptable ways to reference, in the TS, methodologies used to determine cycle-specific parameter limits.
In GL 89-19, Request for Action Related to Resolution of Unresolved Safety Issue A-47 Safety Implication of Control Systems in LWR [Light Water Reactor] Nuclear Power Plants Pursuant to 10 CFR 50.54(f) (Generic Letter 89-19), the NRC staff informed licensees that, among other things, protection should be provided for certain control system failures. GL 89-19, Enclosure 2, Control System Design and Procedural Modification for Resolution of USI A-47, - Item (2),
Westinghouse-Designed PWR [Pressurized-Water Reactor] Plants, provided recommendations concerning automatic SG overfill protection for a group of plants that includes HBR2. It recommended that:
(1) the overfill protection system is sufficiently separate from the control portion of the MFW [main feedwater] control system so that it is not powered from the same power source, not located in the same cabinet, and not routed so that a fire is likely to affect both systems, and (2) the plant procedures and technical specifications include requirements to periodically verify operability of this system
4.0 TECHNICAL EVALUATION
4.1 Evaluation of TS 2.1.1.1 and 5.6.5 Changes 4.1.1 Reactor Core Some current methodologies have been rendered obsolete by the transition to NRC-approved Duke Energy methodologies. The obsolete methodologies for the XNB correlation DNB limit are no longer used in the design and safety analysis of core reloads. The method proposed for removal from TS 5.6.5.b is method number 9, XN-NF-621(A), XNB Critical Heat Flux Correlation. A corresponding change was proposed for TS 2.1.1.1. Analyses completed using the remaining methodologies will continue to provide assurance that the design and safety analysis of the core reloads remain within the TS core operating limits to ensure safe operation.
Accordingly, the NRC staff confirms the changes to remove obsolete information are administrative and finds them acceptable.
4.1.2 Remove Obsolete Methods The proposed change would revise Robinsons TS to reflect the removal of analytical methods that will no longer be used to determine the core operating limits, as these methods were replaced upon transitioning to NRC-approved Duke Energy methods. Proposed revisions TS 2.1.1.1 and TS 5.6.5.b are included to reflect the removal of analytical methods no longer applicable for the determination of Robinsons core operating limits. By letter dated April 29, 2019 (Amendment 263), the NRC approved the addition of NRC-approved methodology BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code, to the list of analytical methodologies in TS 5.6.5.b (ML18288A139).
In Amendment 263, the NRC also approved the revision of the fuel centerline melt safety limit to that used in the COPERNIC code, which allows Duke Energy the ability to self-perform fuel rod mechanical analyses for Robinson. TS 5.6.5.b lists the methodologies approved for use in the design and safety analysis of core reloads, with the COLR identifying the methods and revisions used each cycle. The NRC has approved the following Duke Energy methodologies for use by Robinson to perform the respective analyses, currently listed as items 28 through 33 in TS 5.6.5.b.
DPC-NE-2005-P, Revision 5, Thermal-Hydraulic Statistical Core Design Methodology, by letter dated March 8, 2016.
DPC-NE-1008-P, Revision 0, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, by letter dated May 18, 2017.
DPC-NF-2010, Revision 3, Nuclear Physics Methodology for Reload Design, by letter dated May 18, 2017.
DPC-NE-2011-P, Revision 2, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors, by letter dated May 18, 2017.
DPC-NE-3008-P, Revision 0, Thermal-Hydraulic Models for Transient Analysis, by letter dated April 10, 2018.
DPC-NE-3009-P, Revision 0, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, by letter dated April 10, 2018.
Analyses completed using the Duke Energy methodologies listed above will continue to provide assurance that the design and safety analysis of the core reloads remain within the TS core operating limits to ensure safe operation.
4.
1.3 NRC Staff Conclusion
of TS 2.1.1.1 and 5.6.5 Proposed Changes The NRC staff reviewed the licensees proposed changes for TS 2.1.1.1 and TS 5.6.5.b. The regulation at 10 CFR 50.36 establishes the regulatory requirements related to the content of TSs. Section 50.36(a)(1) requires an application for an operating license to include proposed TSs. The regulations at 10 CFR Part 50, Appendix A, GDC 10, Reactor design, 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.
The NRC staff considered guidance contained in GL 88-16, insofar as it specifies acceptable ways to reference, in the TS, methodologies used to determine cycle-specific parameter limits.
The NRC staff agrees that this difference does not alter the conclusion that the proposed change is applicable to Robinson. Because the Duke Energy methodologies that remain identified in TS 5.6.5.b are sufficient for the licensee to perform core reload analyses, the NRC staff determined that the changes to TS 2.1.1.1 and TS 5.6.5.b are acceptable.
4.2 Evaluation of TS 3.3.2 Changes The NRC staff reviewed the licensees regulatory and technical analyses in support of the proposed changes, as described in the LAR dated September 21, 2022 (ML22264A149), the response to NRCs request for additional information (RAI) (ML23040A426), and licensees supplement dated May 31, 2023 (ML23151A251) regarding the proposed change to add function 5.c to TS table 3.3.2-1. The NRC staff evaluated the proposed amendment by using guidance of RG 1.105, Revision 4, and ANSI/ISA 67.04.01-2018.
As part of its evaluation, the NRC staff performed an independent confirmatory evaluation to:
Verify the licensees setpoint calculation methodology, using the square root of the sum of the squares (SRSS), as the means of combining normally distributed and independent uncertainty terms and algebraic summation as the means of combining uncertainty terms that are not random, not normally distributed or are dependent, to assure that control and monitoring setpoints are established and maintained in a manner consistent with plant safety function requirements.
Verify the licensees setpoint calculation values are adequate to assure, with a high confidence level, that required protective actions are initiated before the associated plant process parameters exceed their analytical limits.
Verify that the licensees calculations identify an appropriate estimate accounting for normal instrument channel rack calibration and drift between successive surveillance intervals, which address the concepts in 10 CFR 50.36(c)(1)(ii)(A) and RIS 2006-17, which addresses the statement 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.
4.2.1 Background The NRC staff reviewed the LAR dated September 21, 2022, and the licensees RAI response dated February 9, 2023, related to Calculation RNP-I/INST-1070 Revision 14, regarding how the Allowable Value (AV) for the Steam Generator Level High-High instrument channel was computed. The NRC staff found a discrepancy between these two documents. The discrepancy was as shown below:
Document AV Equation Note On Pages LAR dated 9/21/2022 AV SP + GAFT AV: Allowable Value; SP: Nominal Setpoint; and GAFT: Group As-Found Tolerance Page 8 of 15 of the enclosure, or pdf Page 10 of 28 RAI response dated 2/9/2023 AV SP + GAFT Page 88 of 105 of, or pdf Page 108 of 411 On May 2, 2023, a virtual observation public meeting was held between the U.S. Nuclear Regulatory Commission (NRC) and representatives of Duke Energy (Duke, the licensee) to discuss the discrepancy above (refer to the Meeting Summary dated May 2, 2023, ML23129A001). In the meeting, the NRC staff provided information from RIS 2006-17.
Specifically, NRC staff discussed requirements of 10 CFR 50.36, with respect to limiting safety system settings (LSSSs) assessed during periodic testing and calibration of instrumentation and how LSSSs are related to AVs. NRC staff requested the licensee to verify the equation stated in the LAR regarding AV (AV SP + GAFT) in the document dated September 21, 2022, if appropriate, because it does not appear to appropriately address the Guidance in RIS 2006-17 for an increasing setpoint. In this guidance, the AV should be greater than or equal to the sum of SP and GAFT.
In the licensees letter dated May 31, 2023, the licensee stated, in part:
In order to clarify computation of the AV limit as well as the acceptable surveillance measured setpoint range, RNP-I/INST-1070 has been revised.
Revision 16 of RNP-I/INST-1070 is provided in Attachment 1 of this letter and provides the appropriate clarification in Section 8.0. Note that changes made to RNP-I/INST-1070 are described in the revision summary included in RNP-I/INST-1070 and notated with revision numbering and revision bars on affected pages.
In Attachment 1 of this Supplement, RNP-I/INST-1070, Steam Generator Narrow Range Level Loop Uncertainty and Scaling Calculation, Revision 16, and on Page 88 of 105 of Attachment 1 (or pdf Page 103 of 128), the calculation was revised, as shown below:
The NRC staff has reviewed the LAR and the supplement dated May 31, 2023. The LAR contains an Enclosure and a formal calculation that describes in detail how the AV was computed. The NRC staff notes that the formal calculation is consistent with the principles in RIS 2006-17 regarding how As-found tolerance should be developed for this limiting safety system setting, to ensure the instrument channel will remain within the Allowable Value. The licensee then submitted the May 31, 2023, supplement which includes a revised calculation and included a statement to clarify computation of the AV limit as well as the acceptable surveillance measured setpoint range. The NRC staff has determined that the revised setpoint calculation provides a correct description of the appropriate range limit for acceptable performance of the channel during a technical specification surveillance.
4.2.2 NRC Technical Evaluation of Proposed Allowable Value and Nominal Trip Setpoint of Proposed TS Table 3.3.2-1 function 5.c 4.2.2.1 Summary of Licensee Methodology NRC staff reviewed the LAR and RNP-I/INST-1070 calculations (Revision 14) (i.e., Attachments 1-4) and the RNP-I/INST-1070 calculations (Revision 16 on Pages 87 and 88 of Attachment 1),
to confirm that the licensees setpoint calculation methodology, regarding the addition of Feedwater Isolation on Steam Generator Level High-High to TS table 3.3.2-1, includes the following:
The equations used to calculate the setpoint limit (SPlimit), Allowable Value (AV), and Margins are consistent with the guidance in RG 1.105.
In the calculations of the total device uncertainties (transmitters and comparators) and the As-Found Tolerance (AFT) of the Measurement and Test Equipment (M&TE) uncertainties, the random components are combined using the Square-Root-Sum-of-the-Squares (SRSS) plus algebraic approaches to combine the uncertainty terms that are not random.
The NRC staff determined that the licensees setpoint methodology used to determine instrument uncertainties was the SRSS plus algebraic is consistent with the RG 1.105 and Section 4.5, Combination of uncertainties, of Part I ISA S67.04-2018, to assure that control and monitoring setpoints are established and maintained in a manner consistent with plant safety function requirements.
4.2.2.2 Proposed Allowable Value and Nominal Trip Setpoint Evaluations The NRC staff performed an independent confirmatory evaluation of the licensees setpoint calculation values to verify whether these values are adequate to assure, with a high confidence level, that required protective actions are initiated before the associated plant process parameters exceed their analytical limits.
For this safety evaluation, the following terms are used:
Analytical Limit (AL) - Limit of a measure or calculated variable established by the safety analysis to ensure that a safety limit is not exceeded.
Allowable Value (AV) - A limiting value that the trip setpoint may have when tested periodically, beyond which appropriate action shall be taken.
Nominal Trip Setpoint (SP1) - A predetermined value for actuation of the final setpoint device to initiate a protective action.
Trip Margin - an allowance provided between the trip setpoint and the analytical limit to ensure a trip before the analytical limit is reached.
1 In the ANSI/ISA Standard 67.04.01-2018, the nominal trip setpoint: NTSP. In the LAR, calibrated setpoint: SP and calculated setpoint limit: SPlimit.
SP Margin (MarginSP) - An allowance provided between the SP and the AL (Region (A + B) in Figure 1 of ANSI/ISA Standard 67.04.01-2018).
AV Margin (MarginAV) - The margin between the AV and the AL that is observable during TS surveillances where the channel may be determined inoperable (Region C in Figure1 of ANSI/ISA Standard 67.04.01-2018).
In addition, the NRC staff evaluated the proposed AV value (76.16% Span) and finds that it is in conformance with RIS 2006-17 regarding whether the licensee properly used the NRC guidance in establishing AVs and Nominal Trip Setpoint to be applied to proposed function 5.c in TS Table 3.3.2-1. Specifically, the licensee proposed to select the SP value (75% Span in increasing steam generator (SG) level) for plant operations and used as as-found LSSSs to calculate the proposed AV value (76.16% Span). Because this proposed AV value would be a limiting value to protect the AL (97% Span in increasing SG level), the staff determined this is adequate to keep the channel operable in accordance with RIS 2006-17.
4.2.2.3 Summary of the Licensees Proposed Setpoints and Results of the Calculations In the RAI response dated February 9, 2023, Sections 6.6 and 6.7 of Calculations RNP-I/INST-1070 Revision 14 provided the calculations of the transmitter and comparators uncertainties, and the results are summarized in the Table 1 below:
Table 1: Transmitter and Comparator Models Uncertainties Summary Error Contributor (Section
- in Calculation RNP-I/INST-1070)
Transmitter Value (% of Span)
Comparator Value
(% of Span)
(Random)
Reference Accuracy RAxmtr = +/- 0.12 (Random)
(Section 6.6.1)
RAcomp = +/- 0.29 (Section 6.7.1)
Calibration Tolerance CALxmtr = +/- 0.50 (Random)
(Section 6.6.2)
CALcomp = +/- 0.50 (Section 6.7.2)
Drift DRxmtr = +/- 0.23 (Random)
(Sections 6.6.3)
DRcomp = +/- 1.00 (Sections 6.7.3)
Analyzed Drift ADxmtr = +/- 1.044% (Random)
ADxmtrBIAS = - 0.227% (Bias)
(Sections 6.6.3)
N/A Measurement and Test Equipment (M&TE) Effect MTExmtr = +/- 0.50 (Random)
(Section 6.6.4)
MTEcomp = +/- 0.31 (Section 6.7.4)
Temperature Effect norTExmtr = +/- 0.76 (Random)
(Section 6.6.5)
TEcomp = +/- 0.93 (Section 6.7.5)
Normal Static Pressure Effect norSPExmtr = +/- 0.57% (Random)
(Section 6.6.6)
N/A Seismic Effect (SE)
SExmtr: Seismic Effect of transmitter in each loop is a Rosemount 3154ND2R2F1E7.
SExmtr = 1.16% span (Section 6.2)
N/A Normal Total Device Uncertainty norTDUxmtr = +/- 1.50%, (Random) norTDUxmtrbias = - 0.23% (Bias)
(Section 6.6.9)
TDUcomp= +/- 1.61 (Section 6.7.7)
Normal Total Device Uncertainty for Emergency Operating Procedure (EOPs) eopTDUxmtr = +/- 1.87%, (Random) eopTDUxmtrBIAS = - 0.23% (Bias)
(Section 6.6.10)
N/A As Found Tolerance AFTxmtr = +/- 0.74 (Random)
(Section 6.6.12)
AFTcomp = +/- 1.16 (Section 6.7.8)
As Left Tolerance ALTxmtr = +/- 0.50 (Random)
(Section 6.6.13)
ALFcomp = +/- 0.50 (Section 6.7.9)
Total Device Uncertainty and Total Loop Uncertainty Calculations:
In the Sections 6.6 and 6.7 of Calculation RNP-I/INST-1070, the licensee used the following equations to calculate:
(norTDUxmtr) = +/- ((CALxmtr)2 + (ADxmtr)2 + (norTExmtr)2 + (norSPExmtr)2)1/2 + ADxmtrBIAS (TDUcomp) = +/- ((CALcomp + MTEcomp)2 + (RAcomp)2 + (DRcomp)2 + (TEcomp)2)1/2 In Section 7.3.1, Total Loop Uncertainty - High Level Valve Interlock, of the Calculation RNP-I/INST-1070, Revision 14, the licensee computed the total loop uncertainty associated with the comparators that provide the High Steam Generator Level Valve interlock using the following equation:
(TLUcomp) = +/- ((norTDUxmtr)2 + (TDUcomp)2 + (SExmtr)2)1/2 + norPME @ 100% Level +
ADxmtrBIAS
= +/- 2.49% Span, -14.24 % of Span
= + 2.49% Span, -16.73 % of Span The negative TLU associated with this setpoint is 16.73% span, which considers effects of reference accuracy (RA), calibration tolerance (CAL), drift (DR), measurement and test equipment (M&TE) effect (MTE), static pressure effect (SPE), temperature effect (TE), power supply effect (PSE), seismic effect (SE), process measurement effect, and analyzed drift bias.
As Found Tolerance (AFT) of Transmitter and Comparator Calculations:
In the Calculation RNP-I/INST-1070, the AFT is computed using the following equation:
AFTcomp = +/- ((CALcomp)2 + (DRcomp)2 + (MTEcomp)2)1/2 AFTcomp = +/- 1.16 % Span In the LAR, the licensee stated, in part, that In addition, the transmitter/sensor is not included in tolerance calculations for this allowable value (AV) because testing consists of a simulated signal injected in place of the field instrument signal. Therefore, the Group As-Found Tolerance (GAFT) used to calculate the AV consists of only the As-Found Tolerance (AFT) of the comparator.
Therefore, AFTcomp is GAFTcomp As Found Tolerance (AFT) - Transmitter AFT (AFTxmtr)
In Calculation RNP-I/INST-1070, Revision 14, the AFT is computed using the following equation:
AFTxmtr = +/- ((CALxmtr)2 + (DRxmtr)2 + (MTExmtr)2)1/2.
= +/- 0.74% of Span Per Design Input 5.27 of Calculation RNP-I/INST-1070, AD may replace RA, MTE, and DR as a single value when calculating the AFT, therefore, AFTxmtr = +/- ((CALxmtr)2 + (ADxmtr)2)1/2
= +/- 1.16 % of Span In the LAR, the licensee stated that The current AFTxmtr value of +/- 0.74% Span is less than, i.e.,
more conservative than, the above calculated AFTxmtr value of +/-1.16% span. For conservatism, AFTxmtr = +/- 0.74% Span will be retained.
Allowable Value (AV) Calculation:
Per Section 8.0 of Calculation RNP-I/INST-1070 (Revision 14), the licensee calculated the AV associated with SP (calibrated setpoint) by following equation:
AV SP + GAFT AV 75% + 1.16% Span AV 76.16% Span In the Calculation RNP-I/INST-1070 (Revision 16, Page 88 of Attachment 1), the licensee revised the AV equation (as shown in Section 3.1 of this SE):
AV = SP + GAFT AV = 75% + 1.16% Span AV = 76.16% Span In the LAR, the licensee stated, in part, that The trip setpoint corresponds to the nominal value at which a device is set and expected to change state. The allowable value is the maximum region associated about a setpoint that is still considered to be acceptable for the instrument to fulfill its function without risking exceeding the analytical limit.
Then, the licensee set the AV 76.16% of Span and Nominal Trip Setpoint is 75% of Span for the proposed TS table 3.3.2-1 function 5.c.
The NRC staff summarizes the results of these calculations in Table 2 below:
Table 2: Summary of the Calculations Results Items / Terms and Equations
% Of Span SP Calibrated Setpoint Setting in TS RNP Unit 2, Table 3.3.2-1, Function 5 (c) (SPlimit with an additional) margin) 75%
The bias component = 14.24% of span
-16.29%
Negative TLU The random component of this loop. The random uncertainty can be reduced by applying the single side of interest and 2 sigma reduction: (0.8225 x 2.49% = +/- 2.05% Span)
Considers effects of reference accuracy, calibration tolerance, drift, measurement, and test equipment (M&TE) effect, static pressure effect, temperature effect, power supply effect, seismic effect, process measurement effect, and analyzed drift bias.
Negative TLU = (2.05% 14.24% = -16.29%)
Span.
SPlimit SPlimit: Calculated setpoint limit (with no additional margin)
SPlimit AL - TLU Where:
AL = analytical limit (97% Span)
TLU = total loop uncertainty ( 16.29% Span) 80.71%
GAFT Group As-Found Tolerance of the comparator. This GAFT includes effects of calibration tolerance, drift, and M&TE uncertainties combined using the SRSS method.
+/- 1.16%
AV AV: Allowable Value setting for Function 5.c in RNP Unit 2 TS Table 3.3.2-1 76.16%
The NRC staff used the information in the LAR and the responses to RAIs (which are reflected in Tables 1 and 2 of this SE) to establish the relationships between the AL, NTS, and associated AV of proposed function 5.c in Figure 1 below:
+ GAFT Note: Diagram Not To Scale a
b Figure 1: & AV & SP of Functions 5 (c) in RPN Unit 2 TS Table 3.3.2-1 Relationships SPLimit
+ ALT GAFT ALT Based on the information in Tables 1 and 2 and Figure 1 above, with respect to the proposed AV and SP (SP calibrated) of Function 5.c in TS Table 3.3.2-1 of RNP Unit 2, the NRC staff has determined the following:
The AV term is computed based on the comparator GAFT and SP that is consistent with the definition of AV in Section 3.6 of the (ANSI/ISA) Standard 67.04.01-2018. This stated, in part, for those plants that include AV values in their technical specifications, the AV is established as the least conservative value of the as-found setpoint that a channel can have during a periodic technical specification required channel calibration, channel operational test, or trip actuating device operational test requiring verification of the channel trip setpoint, beyond which appropriate action shall be taken as specified in the plant technical specifications.
The proposed AV (76.16% of span) is consistent with the guidance of RG 1.105 Revision 4, that includes Figure 1 of this document. This also consistent with the definition of AV in RIS 2006 - 17, which stated, in part, An AV is a limiting value of an instruments as-found trip setting used during surveillances and Many licensees use an AV as as-found LSSSs. This means that licensees perform periodic surveillances and use the AV to verify that the SL is protected and that the channel is operable. If the AV is exceeded during a surveillance, the instrument is declared inoperable because there is not adequate assurance that the instrument will perform its safety function, and appropriate TS-required action must be taken.
SP inclusive (+/-) of its AFT is within the range of AV. That would assure that the trip signal will be initiated before or when SP reached the AV value, and it is consistent with Note (1) of the Nominal Trip Setpoint. Therefore, AV 76.16% Span will provide reasonable assurance that the AL will not be exceeded. But AV SP +/- GAFT, as statement in Page 8 of 28 Enclosure of the LAR, was an error.
Further, evaluation of the setpoints during TS required surveillances to be within +/- AFT band of the SP will ensure that the channel is functioning as required by 10 CFR 50.36(c)(1)(ii)(A), and as recommended by RIS 2006 -17.
The AV value (maximum 76.16% Span increasing level) is very conservative with respect to the SPLimit (which is AL -TLU). The proposed High-High SG level setpoint (75%) provides margin M to the calculated setpoint limit (80.71%). This indicates that there is additional margin beyond TLU, and the proposed SP (75% increasing) will assure the AL (97% Span increasing) will not be exceeded.
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 margin ratio percentages between the MarginSP and MarginAV margins (Row 11 of Figure 1 of this SE) is adequate. This ratio margin ensures that the trip setpoint has been chosen to assure that a trip or safety actuation will occur significantly before the measured process reaches the AL (maximum of upper level). The proposed SP and AV alues support an automatic protective action that will correct the abnormal situation before a safety limit is exceeded.
The value of CALxmtr, MTExmtr, and ALTxmtr is consistent with the vendor information for the transmitter and comparator in Attachments 4, 5, 6 of the Calculation RNP-I/INST-1070.
The NRC staff reviewed the LAR and its attachments to verify that the proposed control and monitoring setpoint values are established and maintained in a manner consistent with regulatory requirements and guidance. The NRC staff evaluated these values and verified that they are consistent with the plant safety functions credited in the UFSAR to assure that protective actions will be initiated before the associated plant process parameter exceeds its analytical limit.
4.2.3 NRC Evaluation of Restoration of Conformance to GL 89-19 In the LAR, the licensee stated, in part:
The high-high SG level trip will prevent SG overfill and ensure that the acceptance criteria for the UFSAR Chapter 15.1.2 event are met. Therefore, the concerns presented by GL 89-19 remain valid and the high-high SG level trip should not have been removed from the TS as part of the STS [Standard Technical Specifications] conversion. This amendment request restores compliance with GL 89-19.
The RNP TS were converted to improved standard TS (STS) based on NUREG-1431, Standard Technical Specifications Westinghouse Plants, as approved by NRC safety evaluation dated October 24, 1997 (Reference 4).
Furthermore, as stated in the RNP TS Bases, Condition D applies to functions that operate on a 2 out of 3 logic. Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> requires the unit be placed in MODE 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
The NRC staff reviewed the LAR, and NUREG-1431 and finds the following:
The proposed Function 5(c) (i.e., applicable modes or other specified conditions, required channels per SG, conditions, and the surveillance requirements) in TS Table 3.3.2-1 were approved in the previous TS.
The proposed MODES (i.e., Modes 1, 2, and 3) for function similar to Function 5(c) are consistent with NUREG-1431.
The proposed required completion time for function 5.c is more conservative than the corresponding completion time in NUREG-1431. The corresponding Condition D in NUREG-1431 specifies to place the channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, while Robinson specifies within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SRs for Function 5.c are consistent with SRs for corresponding function 5.b in NUREG-1431 (i.e., SR 3.3.2.1 (Perform Channel Check), SR 3.3.2.5 (Perform COT),
and SR 3.3.2.9 (Perform Channel Calibration)) except the SR 3.3.2.10 (ESFAS Response Times). In the LAR, the licensee noted that the SR 3.3.2.10 in NUREG-1431 is not applicable to Robinson, because it was not included in Robinsons previously-approved TS.
Table note (1) for Nominal Trip Setpoint in RNC TS table 3.3.2-1 will provide the requirement for a channel that is operable with an actual trip setpoint value outside its calibration tolerance band around the nominal setpoint provided it is still conservative with respect to the associated AV. This channel will be re-adjusted to within the established calibration band of the nominal trip setpoint. This requirement will ensure the instrument channel that will remain within the SPlimit, provided the channel is remains within the AV and is therefore operable, and it is consistent with the principles in RIS 2006-17.
4.2.4 Summary of NRC Staff Conclusion for TS 3.3.2 Proposed Changes As described above, the licensee used an acceptable SRSS combinatorial method to calculate the proposed settings. The NRC staff finds that this methodology provides reasonable assurance that control and monitoring setpoints are established and maintained in a manner consistent with plant safety function requirements and consistent with RG 1.105 Revision 4.
Furthermore, as described above, the NRC staff independently performed evaluations of calculated margins and margin comparisons to confirm that required protective actions will be initiated before the associated plant process parameter values exceed their analytical limits.
Additionally, the licensees proposed As-Left and As-Found values associated with the setpoint changes were determined in a manner consistent with RIS 2006-17 in establishing the As-Left and As-Found tolerances. The NRC staff finds that the proposed SP and AV values for function 5.c in TS Table 3.3.2-1 would provide sufficient margins to satisfy the requirements of 10 CFR 50.36(c)(1)(ii)(A). Therefore, the NRC staff concludes the licensees proposed changes are acceptable.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the South Carolina State official was notified of the proposed issuance of the amendments on July 7, 2023. The South Carolina State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendments change 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 amendments involve 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 previously issued a proposed finding that the amendment involves no significant hazards consideration published in the Federal Register on November 29, 2022 (87 FR 73342),
and there has been no public comment on such finding. Accordingly, the amendments meet 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 amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: HVu, NRR DRahn, NRR CJackson, NRR KWest, NRR DNold, NRR Date: October 12, 2023
ML23226A086 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DEX/EICB/BC NAME MMahoney RButler MWaters (NCarte for)
DATE 08/14/2023 08/24/2023 07/03/2023 OFFICE NRR/DSS/SNSB/BC NRR/DSS/STSB/BC NRR/DSS/SCPB/BC NAME SKrepel VCusumano BWittick DATE 04/24/2023 08/11/2023 08/12/2023 OFFICE OGC - NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME RWeisman DWrona MMahoney DATE 09/28/2023 10/12/2023 10/12/2023