ML26030A239
| ML26030A239 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 01/30/2026 |
| From: | James Holloway Dominion Energy Virginia, Virginia Electric & Power Co (VEPCO) |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 25-273 | |
| Download: ML26030A239 (0) | |
Text
Dominion Energy Virginia 5000 Dominion Boulevard, Glen Allen, VA 23060 DominionEnergy.com 10 CFR 50.90 September 9, 2021 U. S. Nuclear Regulatory Commission Serial No.:
25-273 ATTN: Document Control Desk NRA/JHH:
R6 Washington, DC 20555-0001 Docket Nos.:
50-338 50-339 License Nos.: NPF-4 NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
NORTH ANNA POWER STATION UNITS 1 AND 2 APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO REMOVE POWER RANGE NEUTRON FLUX RATE - HIGH NEGATIVE RATE TRIP FUNCTION Pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, Virginia Electric and Power Company (Dominion Energy Virginia) is submitting a request for an amendment to the Technical Specifications (TS) for North Anna Power Station (NAPS), Units 1 and 2.
This proposed amendment would revise the TSs to remove the Power Range Neutron Flux Rate - High Negative Rate Trip Function from TS 3.3.1, Reactor Trip System Instrumentation. The proposed change is consistent with the Nuclear Regulatory Commission (NRC) approved methodology provided in Westinghouse Topical Report WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event, dated October 23, 1989.
The enclosure provides a description and assessment of the proposed changes. provides the existing NAPS Units 1 and 2 TS pages marked to show the proposed changes. Attachment 2 provides revised (clean) TS changes. Attachment 3 provides existing TS Bases pages marked to show the proposed changes for information only. The Facility Safety Review Committee has reviewed and concurred with the determinations herein.
~
Dominion
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Energy January 30, 2026
Serial No.: 25-273 Docket Nos.: 50-338/50-339 Page 2 of 3 The proposed amendment does not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The basis for this determination is included in the Enclosure. Dominion Energy Virginia has also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite, or any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change. In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Virginia Official.
Dominion Energy Virginia requests that the amendment be approved by August 1, 2026, to allow for implementation during the Fall 2026 NAPS Unit 2 Refueling Outage. The TS change will be implemented for each Unit during its first refueling following receipt of NRC approval. This is planned for Spring 2027 for NAPS Unit 1 and Fall 2026 for NAPS Unit 2.
Should you have any questions or require additional information, please contact Julie Hough at 804-273-3586.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on Sincerely, James E. Holloway Vice President - Nuclear Engineering & Fleet Support
Enclosure:
Description and Assessment of Proposed Change Attachments:
- 1. Proposed Technical Specification Changes (Mark-Ups)
- 2. Revised Technical Specification Pages
- 3. Proposed Technical Specification Bases Changes (Mark-Ups) For Information Only January 30, 2026
Serial No.: 25-273 Docket Nos.: 50-338/50-339 Page 3 of 3 Commitments made by this letter: None cc: Regional Administrator, Region II U. S. Nuclear Regulatory Commission Mr. G. E. Miller Senior Project Manager - North Anna Power Station U. S. Nuclear Regulatory Commission NRC Senior Resident Inspector North Anna Power Station Old Dominion Electric Cooperative R-North-Anna-Correspondence@odec.com State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street, Suite 730 Richmond, Virginia 23219
Serial No.: 25-273 Docket Nos.: 50-338/339 ENCLOSURE Description and Assessment of Proposed Change Virginia Electric and Power Company (Dominion Energy Virginia)
North Anna Power Station Unit 1 and Unit 2
Serial No.: 25-273 Docket Nos.: 50-338/50-339, Page 1 of 9 DESCRIPTION AND ASSESSMENT 1.0
SUMMARY
DESCRIPTION The proposed license amendment would delete the Technical Specification (TS) requirement for the Power Range Neutron Flux Rate - High Negative Rate Trip function.
The Power Range Neutron Flux Rate - High Negative Rate Trip function, as specified in Table 3.3.1-1, Reactor Trip System Instrumentation, of the TS as Function 3.b, Power Range Neutron Flux Rate - High Negative Rate, would be deleted. The proposed change for the removal of the Power Range Neutron Flux Rate - High Negative Rate Trip function is aligned with the Nuclear Regulatory Commission (NRC) approved methodology described in the Westinghouse Topical Report WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event.
2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The Reactor Trip System (RTS) initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and the Reactor Coolant System (RCS) pressure boundary during anticipated operational occurrences (AOOs), and to assist the Engineered Safety Features (ESF) Systems in mitigating accidents. The protection and monitoring systems have been designed to assure safe operation of the reactor. This is achieved by specifying limiting safety system settings in terms of parameters directly monitored by the RTS, as well as specifying Limiting Conditions for Operation (LCOs) on other reactor system parameters and equipment performance.
The RTS consists of sensors that monitor the various plant parameters and are connected with analog and digital circuitry. The analog circuitry consists of two to four redundant channels, and the digital circuitry consists of two redundant logic trains that receive inputs from the analog channels to complete the logic necessary to automatically open the reactor trip breakers. For the Power Range Neutron Flux Rate
- High Negative Rate Trip function, the circuit trips the reactor when a sudden abnormal decrease in nuclear power occurs in two out of the four power range channels.
Serial No.: 25-273 Docket Nos.: 50-338/50-339, Page 2 of 9 2.2 Current Technical Specifications Requirements The RTS Trip setpoints are established in Table 3.3.1-1 of TS 3.3.1, Reactor Trip System Instrumentation, and are directly monitored by the RTS. The LCO for TS 3.3.1 requires that the RTS instrumentation for each Function of Table 3.3.1-1 be OPERABLE in the identified MODE of applicability. The LCO for the Power Range Neutron Flux Rate - High Negative Rate Trip function requires all four channels to be operable in MODES 1 or 2, when there is the potential for a multiple rod drop accident to occur. The Power Range Neutron Flux Rate - High Negative Rate Trip function channels are not required to be operable in MODES 3, 4, 5, or 6, as the reactor core is not critical and Departure from Nucleate Boiling (DNB) is not a concern. The Power Range Neutron Flux Rate - High Negative Rate Trip function is described as Function 3.b of Table 3.3.1-1 of the North Anna Power Station (NAPS) TSs.
2.3 Reason for the Proposed Change This proposed change removes an unnecessary trip function, preventing automatic reactor trips in response to control rod drop events and avoiding unnecessary reactor transients. As discussed in Section 3.0, documented analysis contained in WCAP-11394-P-A and NAPS specific analysis show that the design limits for Departure from Nucleate Boiling Ratio (DNBR) continue to be met, independent of the Power Range Neutron Flux Rate - High Negative Rate Trip function.
2.4 Description of Proposed Change The proposed change would delete the TS requirement for the Power Range Neutron Flux Rate - High Negative Rate Trip function, as specified in Table 3.3.1-1, Reactor Trip System Instrumentation, of the TS as Function 3.b, Power Range Neutron Flux Rate -
High Negative Rate.
Currently, TS 3.3.1 requires the Power Range Neutron Flux Rate - High Negative Rate Trip Function to have four (4) channels operable while in MODES 1 and 2. Following application of the proposed change, the Power Range Neutron Flux Rate - High Negative Rate Trip Function would be deleted and the trip function would not be required for operation in any MODE. Deletion of the High Negative Rate trip function from Table 3.3.1-1 would leave the High Positive Rate trip function as the only function for the Power Range Neutron Flux Rate Instrumentation. The proposed change would remove the a listing for the High Positive Rate and relocate the associated description of the function to a single row. This change seeks to simplify Table 3.3.1-1 following the removal of the High
Serial No.: 25-273 Docket Nos.: 50-338/50-339, Page 3 of 9 Negative Rate trip function and does not alter any of the descriptions associated with the High Positive Rate Function. provides mark-ups of the proposed changes to NAPS TS Table 3.3.1-1. provides the revised (clean) TS pages. Attachment 3 provides existing NAPS TS Bases pages marked to show the proposed changes for information only.
3.0 TECHNICAL EVALUATION
3.1 Application of WCAP-11394-P-A The original design basis for the Power Range Neutron Flux Rate - High Negative Rate Trip function was to mitigate the consequences of one or more dropped RCCAs. The intent was that in the event of one or more dropped RCCAs, the reactor trip system would detect the rapidly decreasing neutron flux (i.e. high negative flux rate) due to the dropped RCCA(s) and would trip the reactor, thus ending the transient and assuring that DNBR limits were maintained. With this trip available, the reactor is tripped when a high negative rate of decrease in reactor power occurs per unit time in two out of the four power range nuclear instrumentation channels.
In 1982, Westinghouse developed and submitted WCAP-10297-P-A Dropped Rod Methodology for Negative Flux Rate Trip Plants (Reference 6.2). The topical report concluded that the High Negative Flux Rate Trip Function was only required when a dropped RCCA or RCCA bank exceeded a specific reactivity worth threshold value. Any dropped RCCA or RCCA bank that had a reactivity worth below the threshold value would not require a reactor trip to maintain DNBR limits.
An additional evaluation method, WCAP-11394-P-A, developed later by Westinghouse, determined that sufficient DNB margin exists for Westinghouse plant designs and fuel types without the Power Range Neutron Flux Rate - High Negative Rate Trip function regardless of the reactivity worth of the dropped RCCA or RCCA bank, subject to a plant/cycle-specific analysis. In Reference 6.3, the NRC concluded that the analysis contains an acceptable procedure for analyzing the dropped RCCA event for which no credit is taken for any direct reactor trip due to the dropped RCCA(s) or for automatic power reduction due to the dropped RCCA(s).
Serial No.: 25-273 Docket Nos.: 50-338/50-339, Page 4 of 9 3.2 NAPS Cycle Specific Analysis NAPS currently uses the WCAP-11394-P-A methodology for analysis of the dropped rod event. The analysis of record for this event is a cycle-specific analysis that is performed for every reload core in the cycle-specific Thermal-Hydraulic Evaluation, following the WCAP-11394-P-A methodology. The safety analysis limits applicable to this event bound the reload values for the current operating cycles, confirming the DNB design basis is met.
Since implementation of WCAP-11394-P-A at NAPS, the NAPS Units 1 and 2 safety analyses do not take credit for the Power Range Neutron Flux Rate - High Negative Rate Trip function. Specifically, the dropped RCCA(s) analyses applicable to the current operating cycles do not rely on actuation of the Power Range Neutron Flux Rate - High Negative Rate Trip function to mitigate the consequences of the accident. The methodology described in WCAP-11394-P-A assumes no direct reactor trip or automatic power reduction to mitigate the consequences of the dropped rod(s). The analysis assumptions and verification that the DNB design basis is met is part of the reload safety analysis for each reactor core reload. The current safety analysis limits confirm that there is margin to the DNBR limit: therefore, the Power Range Neutron Flux Rate - High Negative Rate Trip function is not required to maintain existing DNBR limits and may be eliminated.
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36 provides the regulatory requirements for TSs.
Specifically, 10 CFR 50.36(b) states, in part: "The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to§ 50.34."
10 CFR 50.36(c)(2)(i) 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.
10 CFR 50.36(c)(2)(ii), Technical specifications, states, A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:
Serial No.: 25-273 Docket Nos.: 50-338/50-339, Page 5 of 9 Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
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.
Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
10 CFR 50, Appendix A, Criterion 10 - Reactor Design states The reactor core and associated coolant, control, and protections 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.
4.2 Precedent The following precedents were identified. The scope of these precedents is the same as the scope for this License Amendment Request (LAR).
Union Electric Company (now Ameren Missouri) submitted an LAR for Callaway Plant, Unit 1, to remove the Power Range Neutron Flux High Negative Rate Trip from their TS. The Nuclear Regulatory Commission (NRC) approved the request in Reference 6.4.
Tennessee Valley Authority (TVA) submitted an LAR for Sequoyah Nuclear Plant, Units 1 and 2 to remove the Power Range Neutron Flux High Negative Rate Trip from their TS. The NRC approved the request in Reference 6.5.
Wolf Creek Nuclear Operating Corporation submitted an LAR for Wolf Creek Generating Station to remove the Power Range Neutron Flux Rate - High Negative Rate Trip function. The NRC approved the request in Reference 6.6.
Serial No.: 25-273 Docket Nos.: 50-338/50-339, Page 6 of 9 4.3 No Significant Hazards Consideration The proposed license amendment would delete the Technical Specification (TS) requirement for the Power Range Neutron Flux Rate - High Negative Rate Trip function.
The Power Range Neutron Flux Rate - High Negative Rate Trip function, as specified in Table 3.3.1-1, Reactor Trip System Instrumentation, of the TS as Function 3.b, Power Range Neutron Flux Rate - High Negative Rate, would be deleted. The proposed change for the removal of the Power Range Neutron Flux Rate - High Negative Rate Trip function is aligned with the Nuclear Regulatory Commission (NRC) approved methodology described in the Westinghouse Topical Report WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event. Dominion Energy Virginia has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1.
Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
Removing the Power Range Neutron Flux - High Negative Rate Trip Function from the North Anna Power Station (NAPS) TS does not increase the probability or consequences of accidents resulting from dropped control rod events that were previously analyzed utilizing the Nuclear Regulatory Commission (NRC) approved WCAP-11394-P-A methodology. The associated accident analysis does not rely on the Power Range Neutron Flux High Negative Rate Trip Function to safely shut down the facility. Other RTS protection functions unrelated to the aforementioned function are not impacted by the deletion of this function. Additionally, the safety analysis of the plant is unaffected by the change and the radiological releases associated with the analysis are not affected.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2)
Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
Serial No.: 25-273 Docket Nos.: 50-338/50-339, Page 7 of 9 The proposed amendment does not create the possibility of or introduce any new accident scenarios. The proposed amendment does not introduce any new failure mechanisms, malfunctions, or accident initiators not previously considered in the licensing basis for NAPS. The proposed amendment does not challenge the performance of existing safety-related systems or components, nor does it impact the design function of other structures, systems, and components.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3)
Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The methodology presented in NRC-approved WCAP-11394-P-A has demonstrated that no direct trip on negative flux rate is necessary to maintain the Departure from Nucleate Boiling (DNB) design basis for a dropped rod and dropped RCCA events. Dominion Energy Virginia has evaluated a NAPS reference cycle with the methodology from WCAP-11394-P-A applied. The application of the methodology for NAPS achieved satisfactory results commensurate with the conclusions reached in WCAP-11394-P-A. As a result, removal of the Power Range Neutron Flux Rate - High Negative Rate Trip function would not significantly reduce the margin of safety for any of the fission product barriers. The application of the methodology in transient analysis for the referenced cycle has demonstrated that the trip function is not required for maintaining the DNBR for dropped control rod events.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Dominion Energy Virginia concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
Serial No.: 25-273 Docket Nos.: 50-338/50-339, Page 8 of 9 4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) 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.
5.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.
However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6.0 REFERENCES
6.1 Westinghouse Topical Report WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event, dated October 23, 1989.
6.2 WCAP-10297-P-A, Dropped Rod Methodology for Negative Flux Rate Trip Plants, dated June 24, 1983.
6.3 Letter from NRC to Westinghouse Owners Group, Acceptance for Referencing of Licensing Topical Reports WCAP-11394(P) and WCAP-11395(NP), Methodology for the Analysis of the Dropped Rod Event, dated October 23, 1989.
6.4 Letter from NRC to Union Electric Company, Amendment No. 56 to Facility Operating License NPF-30, dated August 23, 1990 (ADAMS Accession No. ML021650565).
Serial No.: 25-273 Docket Nos.: 50-338/50-339, Page 9 of 9 6.5 Letter from NRC dated July 12, 2022, to Tennessee Valley Authority, Issuance of Amendment Nos. 357 and 351 Regarding Revision to Technical Specification Table 3.3.1-1, Reactor Trip System Instrumentation (ADAMS Accession No. ML22165A105).
6.6 Letter from NRC to Wolf Creek Nuclear Operating Corporation dated March 8, 2024, Wolf Creek Generating Station, Unit 1 - Issuance of Amendment No. 240 re: Removal of the Power Range Neutron Flux Rate - High Negative Trip Function from Technical Specifications (EPID L-2023-LLA-0032) (ADAMS Accession No. ML24016A070).
Serial No.: 25-273 Docket Nos.: 50-338/50-339 ATTACHMENT 1 Proposed Technical Specification Changes (Mark-Ups)
Virginia Electric and Power Company (Dominion Energy Virginia)
North Anna Power Station Unit 1 and Unit 2
Serial No.25-273 Docket Nos. 50-338/339, Page 1 of 2 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 1.0 USE AND APPLICATION...................... 1.1-1 1.1 Definitions........................ 1.1-1 1.2 Logical Connectors.................... 1.2-1 1.3 Completion Times..................... 1.3-1 1.4 Frequency
...................... 1.4-1 2.0 SAFETY LIMITS (SLs)...................... 2.0-1 2.1 Sls............................ 2.0-1 2.2 SL Violations....................... 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY..... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY.......... 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS............... 3.1.1-1 3.1.1 SHUTDOWN MARGIN (SDM)................ 3.1.1-1 3.1.2 Core Reactivity................... 3.1.2-1 3.1.3 Moderator Temperature Coefficient (MTC)....... 3.1.3-1 3.1.4 Rod Group Alignment Limits............. 3.1.4-1 3.1.5 Shutdown Bank Insertion Limits
....... 3.1.5-1 3.1.6 Control Bank Insertion Limits..
....... 3.1.6-1 3.1.7 Rod Position Indication.....
....... 3.1.7-1 3.1.8 Primary Grade Water Flow Path Isolation Valves... 3.1.8-1 3.1.9 PHYSICS TESTS Exceptions-MODE 2........... 3.1.9-1 3.2 3.2.1 3.2.2 3.2.3 3.2.4 3.3 3.3.1 3.3.2 3.3.3 3.3.4 3.3.5 3.3.6 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 POWER DISTRIBUTION LIMITS..............
Heat Flux Hot Channel Factor (Fq(Z)).......
Nuclear Enthalpy Rise Hot Channel Factor ( F:H )
AXIAL FLUX DIFFERENCE (AFD)...........
QUADRANT POWER TILT RATIO {QPTR)........
INSTRUMENTATION...................
Reactor Trip System (RTS) Instrumentation.
Engineered Safety Feature Actuation System (ESFAS) Instrumentation.........
Post Accident Monitoring (PAM) Instrumentation Remote Shutdown System.............
Loss of Power (LOP) Emergency Diesel Generator (EDG) Start Instrumentation......
Main Control Room/Emergency Switchgear Room (MCR/ESGR) Envelope Isolation Actuation Instrumentation.............
REACTOR COOLANT SYSTEM (RCS)..........
RCS Pressure, Temperature, and Flow Departure
.. 3.2.1-1 3.2.1-1 3.2.2-1 3.2.3-1 3.2.4-1 3.3.1-1 3.3.1-1
. 3.3.2-1 3.3.3-1 3.3.4-1 3.3.5-1 from Nucleate Boiling (DNB) Limits......
. 3.3.6-1
. 3.4.1-1
. 3.4.1-1
. 3.4.2-1 RCS Minimum Temperature for Criticality......
RCS Pressure and Temperature (P/T) Limits.
RCS Loops-MODES 1 and 2 RCS Loops-MODE 3................
RCS Loops-MODE 4.............
3.4.3-1 3.4.4-1 3.4.5-1 3.4.6-1 North Anna Units 1 and 2 AmeRdmeRts 255/236
Serial No.25-273 Docket Nos. 50-338/339, Page 2 of 2 RTS Instrumentation 3.3.1 Table 3.3.1-1 {page 1 of 5)
Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE
- 2. Power Range Neutron Flux
- a. High
- b. Low
- 3. Power Range Neutron Flux Rate
-a-.. High Positive Rate
- b. MigA Negati ve Rate
- 4. Intermediate Range Neutron Flux
- 5. Source Range Neutron Flux 1, 2 3(a)
- 4(a)
- 5(a) 1, 2 1, 2 2
2 4
4 4
4 2
2 2
B C
D E
E F, G H, I I* J K
SR 3.3.1.14 SR 3.3.1.14 NA NA SR 3.3.1.1 S 110% RTP SR 3.3.1.2 SR 3.3.1.3 SR 3.3.1.7 SR 3.3.1. 11 SR 3.3.1.16 SR 3.3.1.1 S 26% RTP SR 3.3.1.8 SR 3.3.1. 11 SR 3.3.1.16 SR 3.3.1.7 S 5.5% RTP SR 3.3.1.11 with time constant
~ 2 sec
~ ~
6 a. a'6 RTP
~ 3.3. 1. 11 witA time
~ 3.3.1. 16 69R5taRt
~~
SR 3.3.1.1 S 40% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.1 S 1.3 E5 cps SR 3.3.1.8 SR 3.3.1. 11 SR 3.3.1.16 SR 3.3.1.1 S 1.3 E5 cps SR 3.3.1. 7 SR 3.3.1. 11 SR 3.3.1.16 SR 3. 3. 1. 1 NA SR 3.3.1.11 (a) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.
(b) Below the P-1O (Power Range Neutron Flux) interlocks.
(c) Above the P-6 (Intermediate Range Neutron Flux) interlocks.
(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.
(e) With the Rod Control System incapable of rod withdrawal. In this condition, source range Function does not provide reactor trip but does provide indication.
North Anna Units 1 and 2 3.3.1-13 Amendment 2Jl/212
ATTACHMENT 2 Serial No.: 25-273 Docket Nos.: 50-338/50-339 Revised Technical Specification Pages Virginia Electric and Power Company (Dominion Energy Virginia)
North Anna Power Station Unit 1 and Unit 2
Serial No.25-273 Docket Nos. 50-338/339, Page 1 of 2 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 1.0 USE AND APPLICATION...................... 1.1-1 1.1 Definitions........................ 1.1-1 1.2 Logical Connectors.................... 1.2-1 1.3 Completion Times..................... 1.3-1 1.4 Frequency
...................... 1.4-1 2.0 SAFETY LIMITS (SLs)...................... 2.0-1 2.1 Sls............................ 2.0-1 2.2 SL Violations....................... 2.0-1 3.0 3.0 3.1 3.1.1 3.1.2 3.1.3 3.1.4 3.1.5 3.1.6 3.1. 7 3.1.8 3.1. 9 3.2 3.2.1 3.2.2 3.2.3 3.2.4 3.3 3.3.1 3.3.2 3.3.3 3.3.4 3.3.5 3.3.6 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY..... 3.0-1 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY.......... 3.0-4 REACTIVITY CONTROL SYSTEMS............... 3.1.1-1 SHUTDOWN MARGIN (SDM)................ 3.1.1-1 Core Reactivity................... 3.1.2-1 Moderator Temperature Coefficient (MTC)....... 3.1.3-1 Rod Group Alignment Limits............. 3.1.4-1 Shutdown Bank Insertion Limits
....... 3.1.5-1 Control Bank Insertion Limits..
....... 3.1.6-1 Rod Position Indication.....
....... 3.1.7-1 Primary Grade Water Flow Path Isolation Valves... 3.1.8-1 PHYSICS TESTS Exceptions-MODE 2........... 3.1.9-1 POWER DISTRIBUTION LIMITS................ 3.2.1-1 Heat Flux Hot Channel Factor (FQ(Z))......
Nuclear Enthalpy Rise Hot Channel Factor (.J{H)
AXIAL FLUX DIFFERENCE (AFD)............
QUADRANT POWER TILT RATIO {QPTR).........
INSTRUMENTATION....................
Reactor Trip System (RTS) Instrumentation.....
Engineered Safety Feature Actuation System (ESFAS) Instrumentation.........
Post Accident Monitoring (PAM) Instrumentation Remote Shutdown System............
Loss of Power (LOP) Emergency Diesel Generator (EDG) Start Instrumentation......
Main Control Room/Emergency Switchgear Room (MCR/ESGR) Envelope Isolation Actuation Instrumentation.............
REACTOR COOLANT SYSTEM (RCS)..........
RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits......
RCS Minimum Temperature for Criticality......
RCS Pressure and Temperature (P/T) Limits.....
RCS Loops-MODES 1 and 2..............
RCS Loops-MODE 3.................
RCS Loops-MODE 4.................
. 3.2.1-1
. 3.2.2-1
. 3.2.3-1
. 3.2.4-1
. 3.3.1-1
. 3.3.1-1 I
. 3.3.2-1
. 3.3.3-1
. 3.3.4-1
. 3.3.5-1
. 3.3.6-1
. 3.4.1-1
. 3.4.1-1
. 3.4.2-1
. 3.4.3-1
. 3.4.4-1
. 3.4.5-1
. 3.4.6-1 North Anna Units 1 and 2 Amendments
Serial No.25-273 Docket Nos. 50-338/339, Page 2 of 2 RTS Instrumentation Table 3.3.1-1 (page 1 of 5)
Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS
- 1. Manual Reactor Trip 1, 2 2
B SR 3.3.1.14 3(al
- 4(al
- 5(al 2
C SR 3.3. 1. 14
- 2. Power Range Neutron Flux
- a. High 1, 2 4
D SR 3.3.1.1 SR 3.3.1.2 SR 3.3.1.3 SR 3.3.1.7 SR 3.3.1.11 SR 3.3. 1. 16
- b. Low 1 (bl* 2 4
E SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3. 1. 16
- 3. Power Range Neutron Flux Rate High Positive Rate 1, 2 4
E SR 3.3.1.7 SR 3.3.1.11
- 4. Intermediate Range Neutron Flux 1 (bl, 2(cl 2
F, G SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11
- 5. Source Range Neutron Flux 2(dl 2
H, I SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3. 1. 16 3(al
- 4(al
- 5(al 2
I* J SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.11 SR 3.3. 1. 16 3(el
- 4(el
- 5(el 1
K SR 3.3.1.1 SR 3.3.1.11 (a) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.
(b) Below the P-1O (Power Range Neutron Flux) interlocks.
(c) Above the P-6 (Intermediate Range Neutron Flux) interlocks.
(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.
3.3.1 ALLOWABLE VALUE NA NA
~ 110% RTP
~ 26% RTP
~ 5.5% RTP with time constant
.:: 2 sec
~ 40% RTP
~ 1.3 E5 cps
~ 1.3 E5 cps NA (e) With the Rod Control System incapable of rod withdrawal. In this condition, source range Function does not provide reactor trip but does provide indication.
North Anna Units 1 and 2 3.3.1-13 Amendment
ATTACHMENT 3 Serial No.: 25-273 Docket Nos.: 50-338/50-339 Proposed Technical Specification Bases Changes (Mark-Ups}
For Information Only Virginia Electric and Power Company (Dominion Energy Virginia)
North Anna Power Station Unit 1 and Unit 2
Serial No.25-273 Docket Nos. 50-338/339, Page 1 of 4 FOR INFORMATION ONLY TECHNICAL SPECIFICATIONS BASES TABLE OF CONTENTS B 2.1 B 2.1.1 B 2.1.2 B 3.0 B 3.0 B 3.1 B 3.1.1 B 3.1.2 B 3.1.3 B 3.1.4 B 3.1.5 B 3.1.6 B 3.1.7 B 3.1.8 B 3.1.9 B 3.2 B 3.2.1 B 3.2.2 B 3.2.3 B 3.2.4 B 3.3 B 3.3.1 B 3.3.2 B 3.3.3 B 3.3.4 B 3.3.5 B 3.3.6 B 3.4 B 3.4.1 B 3.4.2 B 3.4.3 B 3.4.4 B 3.4.5 B 3.4.6 B 3.4.7 B 3.4.8 B 3.4.9 SAFETY LIMITS (SLs).................. B 2.1.1-1 Reactor Core SLs................. B 2.1.1-1 Reactor Coolant System (RCS) Pressure SL..... B 2.1.2-1 LIMITING CONDITION FOR OPERATION (LCO)
APPLICABILITY...........
....... B 3.0-1 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY....... B 3.0-12 REACTIVITY CONTROL SYSTEMS.....
SHUTDOWN MARGIN (SDM)......
...... B 3.1.1-1
...... B 3.1.1-1 Core Reactivity.........
...... B 3.1.2-1 Moderator Temperature Coefficient Rod Group Alignment Limits...
Shutdown Bank Insertion Limits Control Bank Insertion Limits (MTC)...... B 3.1.3-1
......... B 3.1.4-1 Rod Position Indication.....
Primary Grade Water Flow Path Isolation
...... B 3.1.5-1
...... B 3.1.6-1
...... B 3.1.7-1 Valves.................... B 3.1.8-1 PHYSICS TESTS Exceptions-MODE 2.......... B 3.1.9-1 POWER DISTRIBUTION LIMITS............... B 3.2.1-1 Heat Flux Hot Channel Factor (Fq(Z)).....
B 3.2.1-1 Nuclear Enthalpy Rise Hot Channel Factor (F1H)
B 3.2.2-1 AXIAL FLUX DIFFERENCE (AFD)...........
B 3.2.3-1 QUADRANT POWER TILT RATIO {QPTR)
B 3.2.4-1 INSTRUMENTATION...................
B 3.3.1-1 Reactor Trip System (RTS) Instrumentation.
B 3.3.1-1 Engineered Safety Feature Actuation System (ESFAS) Instrumentation..
B 3.3.2-1 Post Accident Monitoring (PAM)
Instrumentation................ B 3.3.3-1 Remote Shutdown System.............. B 3.3.4-1 Loss of Power (LOP) Emergency Diesel Generator (EDG) Start Instrumentation......
Main Control Room/Emergency Switchgear Room (MCR/ESGR) Envelope Isolation Actuation Instrumentation.............
REACTOR COOLANT SYSTEM (RCS)..........
RCS Pressure, Temperature, and Flow Departure B 3.3.5-1 B 3.3.6-11
... B 3.4.1-1 from Nucleate Boiling (DNB) Limits...... B 3.4.1-1 RCS Minimum Temperature for Criticality...... B 3.4.2-1 RCS Pressure and Temperature (P/T) Limits..... B 3.4.3-1 RCS Loops-MODES 1 and 2.............. B 3.4.4-1 RCS Loops-MODE 3................. B 3.4.5-1 RCS Loops-MODE 4................
B 3.4.6-1 RCS Loops-MODE 5, Loops Filled.......... B 3.4.7-1 RCS Loops-MODE 5, Loops Not Filled........ B 3.4.8-1 Pressurizer.................... B 3.4.9-1 North Anna Units 1 and 2 Revision -J.9-..
Serial No.25-273 Docket Nos. 50-338/339, Page 2 of 4 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY FOR INFORMATION ONLY RTS Instrumentation B 3.3.1
- 3.
Power Range Neutron Flux Rate (continued)
. Power Range Neutron Flux-High Positive Rate (continued)
Flux-High and Low Setpoint trip Functions to ensure that the criteria are met for a rod ejection from the power range.
The LCO requires all four of the Power Range Neutron Flux-High Positive Rate channels to be OPERABLE.
In MODE 1 or 2, when there is a potential to add a large amount of positive reactivity from a rod ejection accident (REA), the Power Range Neutron Flux-High Positive Rate trip must be OPERABLE. In MODE 3, 4, 5, or 6, the Power Range Neutron Flux-High Positive Rate trip Function does not have to be OPERABLE because other RTS trip Functions and administrative controls will provide protection against positive reactivity additions. Also, since only the shutdown banks may be fully withdrawn in MODE 3, 4, or 5, the remaining complement of control bank {partial withdrawal allowed) worth ensures a sufficient degree of SDM in the event of an REA. In MODE 6, no rods are withdrawn and the SDM is increased during refueling operations. The reactor vessel head is also removed or the closure bolts are detensioned preventing any pressure buildup. In addition, the NIS power range detectors cannot detect neutron levels present in this mode.
b, Power Range Neutron ~lux Wigh Negative Rate The Power Range Neutron ~lux Wigh Negative Rate tri~
~unction ensures that ~rotection is ~rovieee for multi~le roe ero~ accieents, At high ~ower levels, a multi~le roe ero~ accieent coule cause local flux
~eaking that woule result in an unconservative local QNBR, QNBR is eefinee as the ratio of the heat flux requires to cause a QNB at a ~articular location in the core to the local heat flux, The QNBR is ineicative of the margin to QNB, No creeit is taken for the o~eration of this ~unction for those roe ero~
accieents in which the local QNBRs will be greater than the limi t.
(continued)
North Anna Units 1 and 2 B 3.3.1-13 Revision 4--
Serial No.25-273 Docket Nos. 50-338/339, Page 3 of 4 FOR INFORMATION ONLY RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY Power RaRge NeutroR Flux Rate (coRtiRued)
- b. Power RaRge NeutroR Flux Migh Negative Rate (coRtiRued)
The LCO re~uires all four Power RaRge NeutroR Flux Migh Negative Rate chaRRels to be OPERABLE,
IR MOQE 1 or 2, wheR there is poteRtial for a multiple rod drop accideRt to occur, the Power RaRge NeutroR Flux Migh Negative Rate trip must be OPERABLE, IR MOQE J, 4, §, or ij, the Power RaRge NeutroR Flux Migh Negative Rate trip FuRctioR does Rot have to be OPERABLE because the core is Rot critical aRd QNB is Rot a coRcerR, Also, siRce oRly the shutdowR baRks may be fully withdrawR iR MOQE J, 4, or§, the remaiRiRg complemeRt of coRtrol baRk (partial withdrawal allowed) worth eRsures a sufficieRt degree of ~QM iR the eveRt of aR REA, IR MOQE ij, RO rods are withdrawR aRd the re~ui red ~QM is iRcreased duriRg refueliRg operatioRs, IR additioR, the NI~ power raRge detectors caRRot detect ReutroR levels preseRt iR this MOQE.
- 4.
Intermediate Range Neutron Flux The Intermediate Range Neutron Flux trip Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition during startup. This trip Function provides redundant protection to the Power Range Neutron Flux-Low Setpoint trip Function. The NIS intermediate range detectors are located external to the reactor vessel and measure neutrons leaking from the core. Note that this Function also provides a signal to prevent automatic and manual rod withdrawal prior to initiating a reactor trip. Limiting further rod withdrawal may terminate the transient and eliminate the need to trip the reactor.
The LCO requires two channels of Intermediate Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function.
(continued)
North Anna Units 1 and 2 B 3.3.1-14 Revision -
Serial No.25-273 Docket Nos. 50-338/339, Page 4 of 4 FOR INFORMATION ONLY BASES ACTIONS E.1 and E.2 (continued)
Overpower ~T; RTS Instrumentation B 3.3.1 Power Range Neutron Flux-High Positive Rate; I
Pressurizer Pressure-High; and SG Water Level-Low Low.
J A known inoperable channel must be placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Placing the channel in the tripped condition results in a partial trip condition requiring only one-out-of-two logic for actuation of the two-out-of-three trips and one-out-of-three logic for actuation of the two-out-of-four trips. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed to place the inoperable channel in the tripped condition is justified in Reference 7.
If the inoperable channel cannot be placed in the trip condition within the specified Completion Time, the unit must be placed in a MODE where these Functions are not required OPERABLE. An additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to place the unit in MODE 3. Six hours is a reasonable time, based on operating experience, to place the unit in MODE 3 from full power in an orderly manner and without challenging unit systems.
The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> while performing routine surveillance testing of the other channels. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> time limit is justified in Reference 7.
F.1 and F.2 Condition F applies to the Intermediate Range Neutron Flux trip when THERMAL POWER is above the P-6 setpoint and below the P-1O setpoint and one channel is inoperable. Above the P-6 setpoint and below the P-1O setpoint, the NIS intermediate range detector performs both monitoring and protection Functions. If THERMAL POWER is greater than the (continued)
North Anna Units 1 and 2 B 3.3.1-39 Revision ~