ML25352A315

License Amendment Request - Proposed Relocation of Reactor Protective System Loss of Load Function from Technical Specifications to Technical Requirements Manual

ML25352A315
Person / Time
Site:	Calvert Cliffs
Issue date:	12/18/2025
From:	Para W, Para W Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
Download: ML25352A315 (0)

Text

200 Energy Way Kennett Square, PA 19348 www.constellation.com 10 CFR 50.90 December 18, 2025 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318

SUBJECT:

License Amendment Request - Proposed Relocation of Reactor Protective System Loss of Load Function from Technical Specifications to Technical Requirements Manual Pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG), proposes a change to the Technical Specifications (TS) of Renewed Facility Operating License Nos. DPR-53 and DPR-69 for Calvert Cliffs Nuclear Power Plant (CCNPP), Units 1 and 2.

The proposed change relocates the requirement for the Reactor Protective System loss of load trip function from CCNPP TS to the Technical Requirements Manual. This trip function is specified in TS Table 3.3.1-1, Reactor Protective System Instrumentation, as function 10, Loss of Load. This license amendment request is consistent with the current licensing basis documented in the CCNPP Updated Final Safety Analysis Report (UFSAR) Chapter 14, Safety Analysis, which does not credit the loss of load trip for any analyses. provides an evaluation of the proposed change. Attachment 2 provides the marked-up TS pages indicating the proposed change. Attachment 3 provides a markup of the affected TS Bases pages. TS Bases changes are provided for information only and will be incorporated in accordance with the TS Bases Control Program upon implementation of the approved amendment.

CEG requests approval of the proposed license amendment by March 2, 2026. The proposed license amendment request, if approved, will be implemented within 14 days of issuance.

The proposed change has been approved by the CCNPP Plant Operations Review Committee in accordance with the requirements of the CEG Quality Assurance Program.

There are no regulatory commitments contained in this submittal.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Docket Nos. 50-317 and 50-318 December 18, 2025 Page 2 (b), CEG is notifying the State of Maryland of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions concerning this submittal, please contact Adam Donell at (267) 533-5156.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 18th day of December 2025.

Respectfully, Wendi Para Sr. Manager - Licensing Constellation Energy Generation, LLC Attachments:

1. Evaluation of Proposed Change

2. Proposed Technical Specification Change (Markup)

3. Proposed Technical Specification Bases Change (Markup) (For Information Only) cc:

USNRC Region I, Regional Administrator USNRC Senior Resident Inspector - Calvert Cliffs Nuclear Power Plant USNRC Project Manager, NRR - Calvert Cliffs Nuclear Power Plant Z. Barthel, State of Maryland Para, Wendi E 2025.12.18 16:54:41

-05'00'

ATTACHMENT 1 License Amendment Request Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318 Evaluation of Proposed Change

Subject:

License Amendment Request - Proposed Relocation of Reactor Protective System Loss of Load Function from Technical Specifications to Technical Requirements Manual 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1

System Description

2.2 Current TS Requirements 2.3 Proposed Change

3.0 TECHNICAL EVALUATION

3.1 Safety Analysis 3.2 Conclusion

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 1 of 11 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG), proposes a change to the Technical Specifications (TS) of Renewed Facility Operating License Nos. DPR-53 and DPR-69 for Calvert Cliffs Nuclear Power Plant (CCNPP), Units 1 and 2.

The proposed change relocates the requirement for the Reactor Protective System loss of load trip function from CCNPP TS to the Technical Requirements Manual. This trip function is specified in TS Table 3.3.1-1, Reactor Protective System Instrumentation, as function 10, Loss of Load. This license amendment request is consistent with the current licensing basis documented in the CCNPP Updated Final Safety Analysis Report (UFSAR) Chapter 14, Safety Analysis, which does not credit the loss of load trip for any analyses.

2.0 DETAILED DESCRIPTION 2.1

System Description

2.1.1 Reactor Protective System (RPS)

The RPS consists of sensors, amplifiers, logic, and other equipment necessary to monitor and assess selected Nuclear Steam Supply System (NSSS) parameters and initiate protective action if any one or a combination of parameters deviates from the preselected operating range.

Protective actions include annunciating alarms, prohibiting control element assembly (CEA) motion, or shutting down (tripping) the reactor. The system functions to protect the core and Reactor Coolant System (RCS) pressure boundary from experiencing excessive temperature and pressure transients and is designed so that no single fault or failure will negate protective action. The RPS consists of four trip paths operating through the coincidence logic matrices to maintain power to, or remove it from, the control element drive mechanisms (CEDMs). Four independent measurement channels normally monitor each plant parameter which can initiate a reactor trip. There are ten individual trip units per RPS channel. Each trip unit receives the processed signals from its respective measurement channel and provides output (reactor trip) signals in the form of relay contacts that make up the coincidence logic matrices. Trip action is automatic when the process variable exceeds the trip setpoint and automatically resets when the condition clears. The channel trips are combined by the logic matrices which ensure that when any two channels call for a reactor trip due to the same trip unit, the trip path relays actuate to interrupt the AC supply to the CEDM power supplies, which will cause all CEAs to drop into the core (i.e., a reactor trip).

2.1.2 RPS Loss of Load Function Each Units turbine is equipped with an automatic stop and emergency trip system which trips the turbine stop and control valves to a closed position for various conditions, including turbine overspeed, low bearing oil pressure, low vacuum, or thrust bearing failure. Upon occurrence of a turbine trip from any of the above causes, and when above a fixed reactor power level (currently 15% rated thermal power (RTP)), a signal is supplied to the RPS to automatically trip the reactor. This is accomplished by the loss of load trip units, which are one of the ten trip units in each of the four RPS channels. The RPS loss of load trip is considered anticipatory since it precedes the high pressurizer pressure trip for some events. This trip provides turbine protection, reduces the severity of the ensuing transient, and helps avoid the lifting of the main

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 2 of 11 steam safety valves during the ensuing transient, thus extending the service life of these valves.

Its functional capability enhances the overall plant equipment service life and reliability. The loss of load trip is automatically bypassed when below 15% RTP as sensed by a Linear Range Nuclear Instrumentation (LRNI) bistable that closes relay contacts in the trip input circuitry for the loss of load trip unit such that the trip unit will not see a turbine trip signal. Conversely, the loss of load trip is automatically enabled above 15% RTP when the LRNI bistable contacts open so that the loss of load trip unit will experience an input signal change upon a turbine trip.

2.2 Current TS Requirements Current CCNPP TS 3.3.1, Reactor Protective System (RPS) Instrumentation-Operating, requires that the RPS loss of load function instrumentation remains operable whenever thermal power is greater than or equal to 15% RTP.

TS 3.3.1 Condition A has the following required actions:

If the RPS loss of load instrumentation is inoperable above 15% RTP, then the loss of load trip unit will be placed in bypass or trip within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or an amount of time determined by the Risk Informed Completion Time (RICT) Program, then the trip unit must be restored to operable status or placed in trip.

If the TS 3.3.1 Condition A actions are not completed, then power must be reduced to <15%

RTP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per TS 3.3.1 Action F.1. Surveillance requirements (SRs) 3.3.1.6 (instrument channel functional test) and 3.3.1.7 (automatic bypass removal functional test) apply to the RPS loss of load instrumentation. Other TS 3.3.1 SRs refer to the RPS loss of load function as not being applicable.

2.3 Proposed Change The proposed change would relocate the RPS loss of load function (function 10) in TS 3.3.1 Table 3.3.1-1 to the Technical Requirements Manual (TRM). TS 3.3.1 Conditions F and G will be updated to reflect the relocation of the RPS loss of load function from TS. SRs 3.3.1.1, 3.3.1.4, and 3.3.1.6 will be updated to reflect the relocation of the RPS loss of load function from TS.

Although SR 3.3.1.7 currently applies to the RPS loss of load function, no change is necessary since it also applies to RPS functions that are not relocated and does not refer to the RPS loss of load function. TS Bases (TSB) will be revised to reflect relocation of the RPS loss of load function from TS to the TRM.

This change is proposed primarily to allow modification of the RPS loss of load automatic bypass setpoint, currently 15% RTP per notes in TS 3.3.1 Table 3.3.1-1, by CEG without prior NRC approval. The current bypass setpoint leaves minimal margin from the 10-12% RTP plant startup procedure guidance to synchronize the generator to the grid. Further, when maintenance activities call for operations low in the power range without the turbine generator online to prevent a reactor trip, the 15% RTP setting is unnecessarily restrictive compared to the available turbine bypass system capacity.

CEG plans to maintain the RPS loss of load trip as described in UFSAR Section 7.2.3.8, Loss of Load. CEG plans to add the RPS loss of load trip function, including surveillances for testing of trip functionality, to the CCNPP TRM based on its equipment protective purpose and as an anticipatory trip to the RPS trips credited in safety analyses. The TRM is controlled via CEG processes and is described in UFSAR Chapter 15. If the proposed change is approved, any

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 3 of 11 updates to the RPS loss of load functionality or bypass setpoint would remain subject to 10 CFR 50.59 review and applicable CEG processes. Following approval of the proposed change, CEG plans to increase the CCNPP Unit 1 RPS loss of load bypass setpoint prior to the Unit 1 startup from the Spring 2026 refueling outage. The requested NRC approval date and planned implementation period support increasing the bypass setpoint prior to this CCNPP Unit 1 startup, pending completion of engineering analyses and documentation in accordance with CEG modification processes. CEG plans to increase the RPS loss of load bypass setpoint for CCNPP Unit 2 in Spring 2026 after the Unit 1 modification is complete.

The proposed change is shown in the TS markup in Attachment 2 and TSB markup (for information only) in Attachment 3. Note that the Attachment 3 TSB markup includes a correction to the information on page B 3.3.1-15, which currently states that the linear range NI Level 1 bistable controls the RPS loss of load bypass. This page markup corrects the TSB information to state that the linear range NI Level 2 bistable controls the RPS loss of load trip bypass. This error was identified during the evaluation of the proposed change and is a correction to the TSB, not a change to the plant.

3.0 TECHNICAL EVALUATION

3.1 Safety Analysis CEG has conducted a review of the current CCNPP licensing basis including applicable sections in the TSB and UFSAR.

The current TSB for TS 3.3.1 make several relevant statements regarding the RPS loss of load trip:

1. Most of the analyzed accidents and transients can be detected by one or more RPS Functions. The accident analysis contained in Reference 1 [the UFSAR], Chapter 14 takes credit for most RPS trip Functions. Some Functions not specifically credited in the accident analysis are part of the Nuclear Regulatory Commission (NRC)-approved licensing basis for the plant. These Functions may provide protection for conditions that do not require dynamic transient analysis to demonstrate Function performance. Other Functions, such as the Loss of Load trip, are purely equipment protective, and their use minimizes the potential for equipment damage.

2. The trip setpoints used in the bistable trip units are based on the analytical limits stated in Reference 1 [the UFSAR], Chapter 14, except for the APD [Axial Power Distribution] and Loss of Load Functions, which are not credited in safety analyses.

3. The Loss of Load trip causes a trip when operating above 15% of RTP. This trip provides turbine protection, reduces the severity of the ensuing transient, and helps avoid the lifting of the main steam safety valves during the ensuing transient, thus extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip. Its functional capability is required to enhance overall plant equipment service life and reliability.

The UFSAR makes several relevant statements regarding the RPS loss of load trip in Section 7.2, Reactor Protective System, and Section 14.5, Loss of Load Event:

1. Section 7.2.3.8: The loss-of-load trip is an equipment protective trip and is not required for reactor protection. A loss-of-load trip above a preset power level is initiated by actuation of the turbine trip system. This trip is anticipatory in nature as it precedes the high-pressure trip.

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 4 of 11

2. Section 14.5.1: The most limiting Loss of Load event for primary system overpressure is a turbine trip without a concurrent reactor trip or an inadvertent closure of the turbine stop valves at HFP. A turbine trip would result in the closure of the turbine stop valves.

3. Section 14.5.2: A Loss of Load event can result in an approach to the DNBR [Departure from Nucleate Boiling Ratio] and LHGR [Linear Heat Generation Rate] SAFDLs

[Specified Acceptable Fuel Design Limits] and the RCS Pressure Upset Limit. The action of the TM/LP [Thermal Margin/Low Pressure], the Variable High Power, or the High Pressurizer Pressure Trip will prevent exceeding these limitsThe most limiting criteria for the Loss of Load event are the RCS and Secondary Pressure Upset Limit of 110% of design. Normally the non-safety grade turbine trip would initiate a reactor trip and lessen the peak pressure. In analyzing this event, no credit is allowed for this trip.

CEG completed a review of all UFSAR Chapter 14 described events and analyses. The CEG review concluded the RPS loss of load trip is not credited in any safety analyses.

NUREG-0933 (Reference 6.3), Section 3 (New Generic Issues), Issue 90: Technical Specifications for Anticipatory Trips ICSB [Instrumentation and Control Systems Branch] (Rev.

2), provides the historical context for anticipatory RPS trips, their inclusion in TS, as well as considerations with respect to safety significance. Based on the low risk reduction potential and low value/impact score, this issue was given a low priority ranking in August 1984. In NUREG/CR-5382, it was concluded that consideration of a 20-year license renewal period did not change the priority of the issue. Further prioritization in September 1995 resulted in an impact/value ratio change that placed the issue in the DROP category, which indicates that this issue was determined to have a very low risk significance and was removed from further consideration by the NRC.

Combustion Engineering Standard TS first included the RPS loss of load trip in NUREG-0212 Revision 3 (Reference 6.4), with the explanation that a CE design change added the trip to the protective system. The RPS loss of load trip remains in the current CE Improved Standard TS (Reference 6.5). CCNPP TS, since at least the late 1970s, have included the RPS loss of load trip, although it was renamed from loss of turbine to loss of load in 1983 (Reference 6.2). 10 CFR 50.36, Technical specifications, paragraph (c)(2)(ii) states that a TS limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of four specific criteria. The RPS loss of load trip does not meet any of the 10 CFR 50.36(c)(2)(ii) criteria as described below.

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 1 disposition: The RPS loss of load trip is not used for detection and indication in the control room of any 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 2 disposition: The RPS loss of load trip is not an initial condition of a design basis accident or transient analysis.

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 5 of 11 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 3 disposition: No credit is taken for the RPS loss of load trip in the CCNPP accident analysis. Reference 6.1, 58 FR 39132, states that the primary success path for a particular mode of operation does not include backup and diverse equipment. The RPS loss of load trip is an anticipatory trip to the primary RPS trips credited in safety analyses, and therefore is not considered as part of the primary success path related 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.

Criterion 4 disposition: The RPS loss of load instrumentation is not relied upon as a signal to initiate a reactor trip for any events modeled in the scope of the Probabilistic Risk Assessment (PRA) model. Since the function is still planned to be tested and available, there is no net change in risk. The RPS loss of load trip function is not significant to public health and safety in that no credit was taken for this trip in any accident analysis. Accident analyses performed assuming no credit for the RPS loss of load trip function have demonstrated that the trip function is not needed to protect the public health and safety. NUREG-0933 concluded that the inclusion in TS of anticipatory trips, such as the RPS loss of load trip, is of very low risk significance.

CCNPP has operating experience demonstrating the RPS loss of load trip is not significant to public health and safety. In 2010, CCNPP Unit 1 experienced an automatic reactor trip from 100% RTP when the RPS high pressurizer pressure trip actuated due to a complete load rejection after a plant switchyard 500 kV breaker tripped open (Licensee Event Report (LER) 317-2010-003). In 2022, CCNPP Unit 2 experienced an automatic reactor trip from 100% RTP when the RPS high pressurizer pressure trip actuated due to closure of the turbine control and intercept valves when the turbine control system Load Drop Anticipator function setpoint was exceeded (LER 318-2022-001). The RPS loss of load trip was not actuated during either event, and the response of the plant, including actuation of the RPS high pressurizer pressure trip, ensured no safety limits were exceeded.

3.2 Conclusion Reference 6.1, 58 FR 39132, states that LCOs which do not meet any of the 10 CFR 50.36(c)(2)(ii) criteria may be proposed for removal from TS and relocation to licensee-controlled documents. Because the RPS loss of load trip does not meet any of the criteria, relocation of the RPS loss of load trip from TS to the TRM is acceptable. Future testing and control of the trip function and bypass setpoint will be performed in accordance with CEG processes, which will provide adequate assurance of reliability as an equipment protective trip and as an anticipatory trip to the RPS trips credited in safety analyses.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. CCNPP was not licensed to the General Design Criteria (GDCs) listed in 10 CFR Part 50, Appendix A, but was licensed based on the Atomic Energy

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 6 of 11 Commission (AEC) proposed Principal Design Criteria (PDCs) published on July 10, 1967. The relevant principal design criteria for CCNPP are described in Appendix 1C of the CCNPP UFSAR. CEG has determined that the proposed change does not require any exemptions or relief from the following current applicable regulations and regulatory requirements, which were reviewed in making this determination:

10 CFR 50.36, Technical Specifications 10 CFR 50.36(c) provides that TS will include Limiting Conditions for Operation (LCOs) which are "the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met."

The proposed change relocates control of the RPS loss of load trip function from technical specifications to the TRM. The RPS loss of load trip function does not meet any of the 10 CFR 50.36(c)(2)(ii) criteria. If the proposed change is approved, testing and control of the trip function will be performed in accordance with CEG processes, which will provide adequate assurance of reliability as an equipment protective trip and as an anticipatory trip to the TS required RPS trips credited in safety analyses.

The proposed change does not affect CCNPP's compliance with the intent of 10 CFR 50.36.

10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants 10 CFR 50.65 requires that when performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to structures, systems, and components that a risk-informed evaluation process has shown to be significant to public health and safety. The maintenance activities associated with this change will be assessed and the increased risk will be managed in accordance with 10 CFR 50.65 (a)(4).

The proposed change does not affect CCNPP's compliance with the intent of 10 CFR 50.65.

NUREG-0800 Branch Technical Position 7-9, Guidance on Requirements for Reactor Protection System Anticipatory Trips, Revision 6 NUREG-0800 Branch Technical Position 7-9 provides IEEE requirements for RPS anticipatory trips. The proposed change does not alter the design of the RPS loss of load trip, and therefore does not affect compliance with NUREG-0800 Branch Technical Position 7-9 requirements.

NUREG-0635, Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Combustion Engineering Designed Operating Plants, and NUREG-0737, Clarification of TMI Action Plan Requirements NUREG-0737 specified requirements for Babcock and Wilcox (B&W) and Westinghouse designed reactors to have an anticipatory reactor trip on turbine trip (items II.K.2.10 and II.K.3.12). Since Calvert Cliffs is Combustion Engineering (CE) designed reactor, these

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 7 of 11 requirements do not directly apply. NUREG-0635, an input to NUREG-0737, notes that all operating CE plants (at the time of NUREG-0635 issuance) have an anticipatory reactor trip on turbine trip (with the exception of Arkansas Nuclear One Unit 2, which also has no Power-Operated Relief Valves). The proposed change does not remove the RPS loss of load function. If the proposed change is approved, testing and control of the trip function will be performed in accordance with CEG processes, which will provide adequate assurance of reliability as an equipment protective trip and as an anticipatory trip to the TS required RPS trips credited in safety analyses.

The proposed change does not affect compliance with NUREG-0635 or NUREG-0737 requirements.

UFSAR Appendix 1C Criterion 6, Reactor Core Design (Similar to GDC 10, Reactor Design)

UFSAR Appendix 1C Criterion 6 requires that the reactor core be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power.

The proposed change does not alter the plant design. Plant safety analyses do not credit the RPS loss of load trip. The proposed change does not remove the RPS loss of load function. If the proposed change is approved, testing and control of the trip function will be performed in accordance with CEG processes, which will provide adequate assurance of reliability as an equipment protective trip and as an anticipatory trip to the TS required RPS trips credited in safety analyses.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criterion 6.

UFSAR Appendix 1C Criterion 9, Reactor Coolant Pressure Boundary (Similar to GDC 14, Reactor coolant pressure boundary)

UFSAR Appendix 1C Criterion 9 requires that the reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime.

The proposed change does not alter the plant design. The RPS loss of load trip is not credited in any CCNPP safety analyses, including those with the potential to impact the reactor coolant pressure boundary. The RPS loss of load trip function will be maintained and controlled in accordance with CEG processes.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criterion 9.

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 8 of 11 UFSAR Appendix 1C Criterion 19, Protection Systems Reliability, and Criterion 20, Protection Systems Redundancy and Independence (Similar to GDC 21, Protection system reliability and testability)

UFSAR Appendix 1C Criterion 19 requires that protection systems be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed.

UFSAR Appendix 1C Criterion 20 requires that redundancy and independence designed into protection systems be sufficient to assure that no single failure or removal from service of any component or channel of a system will result in loss of the protection function. The redundancy provided shall include, as a minimum, two channels of protection for each protection function to be served. Different principles shall be used where necessary to achieve true independence of redundant instrumentation components.

The proposed change does not alter the plant design. Reliability and redundancy in the RPS will not be adversely affected since the RPS loss of load trip function as described in the UFSAR will continue to be tested and controlled in accordance with CEG processes.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criteria 19 and 20.

UFSAR Appendix 1C Criterion 33, Reactor Coolant Pressure Boundary Capability (Similar to GDC 15, Reactor coolant system design)

UFSAR Appendix 1C Criterion 33 requires that the reactor coolant pressure boundary shall be capable of accommodating without rupture, and with only limited allowance for energy absorption through plastic deformation, the static and dynamic loads imposed on any boundary component as a result of any inadvertent and sudden release of energy to the coolant. As a design reference, this sudden release shall be taken as that which would result from a sudden reactivity insertion such as rod ejection (unless prevented by positive mechanical means), rod dropout, or cold water addition.

The proposed change does not alter the plant design. The RPS loss of load trip is not credited in any CCNPP safety analyses and will be maintained and controlled in accordance with CEG processes.

The proposed change does not affect CCNPP's compliance with the intent of UFSAR Appendix 1C Criterion 33.

4.2 Precedent Precedents 4.2.1 and 4.2.2 are applicable to this submittal in that the U.S. NRC previously granted other licensees approval to remove from TS functions that were part of the reactor protective system (called reactor trip system in some TS) but that were not credited in safety analyses. Precedent 4.2.3 is applicable to this submittal in that the U.S. NRC previously approved removal from TS of the Limiting Safety System Setting (LSSS) for the RPS loss of

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 9 of 11 load function based, in part, on the fact that the RPS loss of load function was not credited in the safety analyses.

1. Letter from Samson S. Lee (U.S. NRC) to Mr. Cleveland Reasoner (Wolf Creek Nuclear Operating Corporation), Wolf Creek Generating Station, Unit 1 - Issuance of Amendment No. 240 RE: Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip Function from Technical Specifications (EPID L-2023-LLA-0032),

dated March 8, 2024 (ADAMS Accession No. ML24016A070).

2. Letter from Perry H. Buckberg (U.S. NRC) to Mr. James Barstow (Tennessee Valley Authority), Sequoyah Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos. 357 and 351 Regarding Revision to Technical Specification Table 3.3.1-1, Reactor Trip System Instrumentation (EPID L-2021-LLA-0200), dated July 12, 2022 (ADAMS Accession No. ML22165A105).

3. Letter from David H. Jaffe (U.S. NRC) to Mr. A. E. Lundvall, Jr. (Baltimore Gas & Electric Company), Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, May 5, 1983. (ADAMS Accession Number ML010430247).

4.3 No Significant Hazards Consideration Constellation Energy Generation, LLC (CEG), proposes a change to the Technical Specifications (TS) of Renewed Facility Operating License Nos. DPR-53 and DPR-69 for Calvert Cliffs Nuclear Power Plant (CCNPP), Units 1 and 2.

The proposed change relocates the requirement for the Reactor Protective System loss of load trip function from CCNPP TS to the Technical Requirements Manual. This trip function is specified in TS Table 3.3.1-1, Reactor Protective System Instrumentation, as function 10, Loss of Load. This license amendment request is consistent with the current licensing basis documented in the CCNPP Updated Final Safety Analysis Report (UFSAR) Chapter 14, Safety Analysis, which does not credit the loss of load trip for any analyses.

CEG has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1.

Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The Reactor Protective System (RPS) loss of load trip function is not credited in the CCNPP safety analyses for any accidents previously evaluated. The RPS loss of load trip function is an anticipatory trip to other RPS trips credited that are credited in the safety analyses but are not affected by the proposed change. The proposed change does not remove the RPS loss of load trip function, and no physical changes or modifications to the RPS are included in the proposed change. If the proposed change is approved, testing and control of the trip function will be performed in accordance with CEG processes, which will provide adequate assurance of reliability as an equipment protective trip and as an anticipatory trip to the TS required RPS trips credited in safety analyses.

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 10 of 11 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated. The proposed change does not challenge the performance or integrity of any safety-related systems or components. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of relocating the RPS loss of load trip function from TS to the Technical Requirements Manual.

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 change involve a significant reduction in a margin of safety?

Response: No.

The RPS loss of load trip function is not credited in the CCNPP safety analyses. The proposed change does not remove the RPS loss of load trip function, and no physical changes or modifications to the RPS are included in the proposed change. Continued testing and control of the trip function in accordance with CEG processes will provide adequate assurance of reliability as an equipment protective trip and as an anticipatory trip to the TS required RPS trips credited in safety analyses.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, CEG concludes that the proposed amendment presents no 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.

4.4 Conclusions 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

CEG 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, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or

License Amendment Request Proposed Relocation of Reactor Protective System Loss of Load Function from TS to TRM Page 11 of 11 significant increase in the amounts of any effluent 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, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1 58 FR 39132, 10 CFR 50 Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, July 22, 1993.

6.2 Letter from David H. Jaffe (U.S. NRC) to Mr. A. E. Lundvall, Jr. (Baltimore Gas & Electric Company), Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, May 5, 1983. (ADAMS Accession Number ML010430247) 6.3 NUREG-0933, Resolution of Generic Safety Issues, Main Report with Supplements 1-35, 2021.

6.4 Letter from Stephen H. Hanauer (U.S. NRC) to Victor Stello, Jr. (U.S. NRC), Revision 3 to Combustion Engineering Standard Technical Specifications (CE-STS), dated March 1, 1982. (ADAMS Accession Number ML20049K081) 6.5 NUREG-1432, Revision 5, Volume 1, Standard Technical Specifications - Combustion Engineering Plants: Specifications, September 2021, and NUREG-1432, Revision 5, Volume 2, Standard Technical Specifications - Combustion Engineering Plants: Bases, September 2021. (ADAMS Accession Numbers ML21258A421 and ML21258A424)

ATTACHMENT 2 License Amendment Request Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318 Proposed Technical Specification Change (Markup)

Technical Specification Pages 3.3.1-5 3.3.1-7 3.3.1-11

RPS Instrumentation-Operating 3.3.1 CALVERT CLIFFS - UNIT 1 3.3.1-5 Amendment No. 314 CALVERT CLIFFS - UNIT 2 Amendment No. 292 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F.

Required Action and associated Completion Time not met for Axial Power Distribution-High and Loss of Load Trip Functions.

F.1 Reduce THERMAL POWER to < 15% RTP.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> G.

Required Action and associated Completion Time not met except for Axial Power Distribution-High and Loss of Load Trip Functions.

G.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> SURVEILLANCE REQUIREMENTS

---

NOTE --------------------------------

Refer to Table 3.3.1-1 to determine which Surveillance Requirement shall be performed for each RPS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform a CHANNEL CHECK of each RPS instrument channel except Loss of Load.

In accordance with the Surveillance Frequency Control Program

RPS Instrumentation-Operating 3.3.1 CALVERT CLIFFS - UNIT 1 3.3.1-7 Amendment No. 314 CALVERT CLIFFS - UNIT 2 Amendment No. 292 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.4 Perform a CHANNEL FUNCTIONAL TEST of each RPS instrument channel except Loss of Load and Rate of Change of Power-High.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.5

---

NOTE ----------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform a CHANNEL CALIBRATION on excore power range channels.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.6 Perform a CHANNEL FUNCTIONAL TEST of each Rate of Change of Power-High and Loss of Load instrument channel.

Once within 7 days prior to each reactor startup SR 3.3.1.7 Perform a CHANNEL FUNCTIONAL TEST on each automatic bypass removal feature.

In accordance with the Surveillance Frequency Control Program

RPS Instrumentation-Operating 3.3.1 CALVERT CLIFFS - UNIT 1 3.3.1-11 Amendment No. 300 CALVERT CLIFFS - UNIT 2 Amendment No. 277 Table 3.3.1-1 (page 3 of 3)

Reactor Protective System Instrumentation FUNCTION MODES SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 9b. Asymmetric Steam Generator Transient (ASGT)(b) 1, 2 SR 3.3.1.1 SR 3.3.1.4 SR 3.3.1.7 SR 3.3.1.8 SR 3.3.1.9 135 psid

10. Loss of Load(d) 1(e)

SR 3.3.1.6 SR 3.3.1.7 NA (a)

Bistable trip unit may be bypassed when NUCLEAR INSTRUMENT POWER is < 1E-4% RTP or > 12% RTP. Bypass shall be automatically removed when NUCLEAR INSTRUMENT POWER is 1E-4% RTP and

< 12% RTP.

(b)

Bistable trip unit may be bypassed when NUCLEAR INSTRUMENT POWER is < 1E-4%. Bypass shall be automatically removed when NUCLEAR INSTRUMENT POWER is 1E-4% RTP. During testing pursuant to LCO 3.4.16, trips may be bypassed below 5% RTP.

(c)

Bistable trip unit may be bypassed when steam generator pressure is < 785 psia. Bypass shall be automatically removed when steam generator pressure is 785 psia.

(d)

Bistable trip unit may be bypassed when NUCLEAR INSTRUMENT POWER is < 15% RTP. Bypass shall be automatically removed when NUCLEAR INSTRUMENT POWER is 15% RTP.

(e)

Trip is only applicable in MODE 1, NUCLEAR INSTRUMENT POWER 15%

RTP.

(f)

CHANNEL CHECK only applies to Wide Range Logarithmic Neutron Flux Monitor.

ATTACHMENT 3 License Amendment Request Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318 Proposed Technical Specification Bases Change (Markup) (For Information Only)

Technical Specification Bases Pages B 3.3.1-6 B 3.3.1-14 B 3.3.1-15 B 3.3.1-21 B 3.3.1-22 B 3.3.1-26 B 3.3.1-29 B 3.3.1-31 B 3.3.1-34

RPS Instrumentation-Operating B 3.3.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.3.1-6 Revision 12 single input contact opening can provide multiple contact outputs to the matrix logic, as well as trip indication and annunciation.

Trip Functions employing auxiliary trip units include the Loss of Load trip and the APD trip.

The APD trip, described above, is a complex function in which the actual trip comparison is performed within the APD calculator. Therefore the APD trip unit employs a contact input from the APD calculator.

All RPS trips, with the exception of the Loss of Load trip, generate a pretrip alarm as the trip setpoint is approached.

The trip setpoints used in the bistable trip units are based on the analytical limits stated in Reference 1, Chapter 14, except for the APD and Loss of Load Functions, which are not credited in safety analyses. The selection of these trip setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account in the respective analytical limits. To allow for calibration tolerances, instrumentation uncertainties, instrument channel drift, and severe environment errors (for those RPS channels that must function in harsh environments, as defined by Reference 2, 10 CFR 50.49) RPS trip setpoints are conservatively adjusted with respect to the analytical limits. In the case of the TM/LP trip, there is also an additional adjustment for cold leg temperature differences.

A detailed description of the methodology used to calculate the trip setpoints, including their explicit uncertainties, is provided in Reference 4. The nominal trip setpoint entered into the bistable is more conservative than that specified by the Allowable Value. A channel is inoperable if its actual setpoint is not within its required Allowable Value.

Setpoints in accordance with the Allowable Value will ensure that SLs of Chapter 2.0 are not violated during AOOs and the consequences of DBAs will be acceptable, providing the plant is operated from within the LCOs at the onset of the AOO or DBA and the equipment functions as designed.

The Loss of Load trip is not a TS function (Reference 11).

RPS Instrumentation-Operating B 3.3.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.3.1-14 Revision 60 maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip signal. In addition, CEA group sequencing in accordance with LCO 3.1.7 is assumed. Finally, the maximum insertion of CEA banks that can occur during any AOO prior to a Power Level-High trip is assumed.

b. ASGT The ASGT provides protection for those AOOs associated with secondary system malfunctions that result in asymmetric primary coolant temperatures.

The most limiting event is closure of a single main steam isolation valve (MSIV). Asymmetric Steam Generator Transient is provided by comparing the secondary pressure in both steam generators in the TM/LP trip calculator. If the pressure in either exceeds that in the other by the trip setpoint, a TM/LP trip will result.

10. Loss of Load The Loss of Load trip causes a trip when operating above 15% of RTP. This trip provides turbine protection, reduces the severity of the ensuing transient, and helps avoid the lifting of the main steam safety valves during the ensuing transient, thus extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip. Its functional capability is required to enhance overall plant equipment service life and reliability.

Operating Bypasses The operating bypasses are addressed in footnotes to Table 3.3.1-1. They are not otherwise addressed as specific table entries.

The automatic bypass removal features must function as a backup to manual actions for all trips credited in safety analyses to ensure the trip Functions are not operationally bypassed when the safety analysis assumes the Functions are not bypassed. The RPS operating bypasses are:

a preset power level.

The Loss of Load trip is not a TS function (Reference 11) but is described to provide background information because it is an RPS function.

RPS Instrumentation-Operating B 3.3.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.3.1-15 Revision 31 Zero power mode bypass (ZPMB) removal of the TM/LP, ASGT, and reactor coolant low flow trips when NUCLEAR INSTRUMENT POWER is < 1E-4% RTP. This bypass is manually enabled below the specified setpoint to permit low power testing. The wide range NI Level 1 bistable in the wide range drawer provides a signal to auxiliary logic, which then permits manual bypassing below the setpoint and removes the bypass above the setpoint.

Power rate of change bypass removal The Rate of Change of Power-High trip is automatically bypassed at < 1E-4% RTP, as sensed by the wide range NI Level 1 bistable, and at

> 12% RTP by the linear range NI Level 1 bistable, mounted in their respective NI drawers (Reference 5). Automatic bypass removal is also effected by these bistables when conditions are no longer satisfied. The automatic bypass removal feature ensures that the Rate of Change of Power-High trip is enabled when reactor power is between 1E-4% and 12% RTP.

Loss of Load and APD-High trip bypass removal The Loss of Load and APD-High trips are automatically bypassed when at

< 15% RTP as sensed by the linear range NI Level 1 bistable.

The bypass is automatically removed by this bistable above the setpoint. This same bistable is used to bypass the Rate of Change of Power-High trip.

Steam Generator Pressure-Low trip bypass removal. The Steam Generator Pressure-Low trip is manually enabled below the pretrip setpoint. The permissive signal is removed, and the bypass automatically removed, when the Steam Generator Pressure-Low trip is above the pretrip setpoint.

The RPS instrumentation satisfies 10 CFR 50.36(c)(2)(ii),

Criterion 3.

LCO The LCO requires all instrumentation performing an RPS Function to be OPERABLE. Failure of any required portion of the instrument channel renders the affected channel(s) inoperable and reduces the reliability of the affected Functions. The specific criteria for determining channel OPERABILITY differ slightly between Functions. These Loss of Load trip bypass removal - The Loss of Load trip is not a TS function (Reference 11) but the bypass is described to provide background information because it is an RPS function. The Loss of Load trip is automatically bypassed below a preset power level as sensed by the linear range NI Level 2 bistable. The bypass is automatically removed by this bistable above the setpoint.

is

RPS Instrumentation-Operating B 3.3.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.3.1-21 Revision 60 allowance to compensate for the time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the SLs.

This trip may be manually bypassed when NUCLEAR INSTRUMENT POWER falls below 1E-4% RTP. This operating bypass is part of the ZPMB circuitry, which also bypasses the Reactor Coolant Flow-Low trip and provides a T power block signal to the Q power select logic (Reference 5). The ZPMB allows low power physics testing at reduced RCS temperatures and pressures. It also allows heatup and cooldown with shutdown CEAs withdrawn.

b.

ASGT This LCO requires four instrument channels of ASGT to be OPERABLE in MODEs 1 and 2.

The Allowable Value is high enough to avoid trips caused by normal operation and minor transients, but ensures DNBR protection in the event of DBAs. The difference between the Allowable Value and the analysis setpoint allows for instrument uncertainty.

The trip may be manually bypassed when NUCLEAR INSTRUMENT POWER falls below 1E-4% RTP as part of the ZPMB circuitry operating bypass. The Steam Generator Pressure Difference is subject to the ZPMB, since it is an input to the TM/LP trip and is not required for protection at low power levels (Reference 5).

10.

Loss of Load The LCO requires four Loss of Load instrument channels to be OPERABLE in MODE 1, NUCLEAR INSTRUMENT POWER 15% RTP.

RPS Instrumentation-Operating B 3.3.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.3.1-22 Revision 60 The Loss of Load trip is automatically bypassed when NUCLEAR INSTRUMENT POWER falls below 15%, as measured by NIs, to allow loading the turbine.

Bypasses The LCO on automatic bypass removal features requires that the automatic bypass removal feature of all four operating bypass channels be OPERABLE for each RPS Function with an operating bypass in the MODEs addressed in the specific LCO for each Function. All four automatic bypass removal features must be OPERABLE to ensure that none of the four RPS instrument channels are inadvertently bypassed.

The LCO applies to the automatic bypass removal feature only. If the bypass channel is failed so as to prevent entering a bypass condition, operation may continue.

APPLICABILITY This LCO is applicable in accordance with Table 3.3.3-1.

Most RPS trip functions are required to be OPERABLE in MODEs 1 and 2 because the reactor is critical in these MODEs. The trips are designed to take the reactor subcritical, maintaining the SLs during AOOs and assisting the ESFAS in providing acceptable consequences during accidents. Exceptions are addressed in footnotes to the table. Exceptions to this APPLICABILITY are:

The APD-High and Loss-of-Load trips are only applicable in MODE 1, NUCLEAR INSTRUMENT POWER 15% RTP because they are automatically bypassed at < 15% RTP, as measured by NIs, where they are no longer needed.

The Rate of Change of Power-High trip, RPS logic, RTCBs, and manual trip are also required in MODEs 3, 4, and 5, with the RTCBs closed, to provide protection for boron dilution and CEA withdrawal events. The Rate of Change of Power-High trip in these lower MODEs is addressed in LCO 3.3.2. The RPS logic in MODEs 1, 2, 3, 4, and 5 is addressed in LCO 3.3.3.

Most trip functions are not required to be OPERABLE in MODEs 3, 4, and 5. In MODEs 3, 4, and 5, the emphasis is placed on return to power events. The reactor is protected in these MODEs by ensuring adequate SHUTDOWN MARGIN (SDM).

is it is it is

RPS Instrumentation-Operating B 3.3.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.3.1-26 Revision 60 E.1, E.2.1, and E.2.2 Condition E applies to two inoperable automatic bypass removal features. If the automatic bypass removal features cannot be restored to OPERABLE status, the associated RPS channel may be considered OPERABLE only if the bypasses are not in effect. Otherwise, the affected RPS channels must be declared inoperable, as in Condition B, and the bypasses either removed or the automatic bypass removal features repaired. Also, Required Action E.2.2 provides for the restoration of the one affected RPS channel to OPERABLE status within the rules of Completion Time specified under Condition B. Completion Times are consistent with Condition B.

F.1 Condition F is entered when the Required Action and associated Completion Time of Condition A, B, C, D, or E are not met for the APD-High trip and Loss-of-Load trip Functions.

If the Required Actions associated with these Conditions cannot be completed within the required Completion Times, the reactor must be brought to a MODE in which the Required Actions do not apply. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to reduce THERMAL POWER to < 15% RTP is reasonable, based on operating experience, to decrease power to < 15%

RTP from full power conditions in an orderly manner and without challenging plant systems.

G.1 Condition G is entered when the Required Action and associated Completion Time of Condition A, B, C, D, or E are not met except for the APD-High trip and Loss-of-Load trip Functions.

If the Required Actions associated with these Conditions cannot be completed within the required Completion Times, the reactor must be brought to a MODE in which the Required Actions do not apply. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to be in MODE 3 is reasonable, based on operating experience, for reaching the required MODE from full power

RPS Instrumentation-Operating B 3.3.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.3.1-29 Revision 60 Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.4 A CHANNEL FUNCTIONAL TEST is performed on each RPS instrument channel, except Loss of Load and Rate of Change of Power, to ensure the entire channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions.

In addition to reference voltage power supply tests, the RPS CHANNEL FUNCTIONAL TEST consists of three overlapping tests as described in Reference 1, Section 7.2. These tests verify that the RPS is capable of performing its intended function, from bistable input through the RTCBs. They include:

Bistable Tests The bistable setpoint must be found to trip within the Allowable Values specified in the LCO and left set consistent with the assumptions of Reference 4. As-found values must also be recorded and reviewed for consistency with the assumptions of the frequency extension analysis.

The requirements for this review are outlined in Reference 8.

A test signal is substituted as the input in one instrument channel at a time to verify that the bistable trip unit trips within the specified tolerance around the setpoint.

This is done with the affected RPS channel bistable trip unit bypassed. Any setpoint adjustment shall be consistent with the assumptions of Reference 4.

Matrix Logic Tests Matrix logic tests are addressed in LCO 3.3.3. This test is performed one matrix at a time. It verifies that a coincidence in the two instrument channels for each Function

RPS Instrumentation-Operating B 3.3.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.3.1-31 Revision 60 control room indications are continuously monitored by the operators.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.6 A CHANNEL FUNCTIONAL TEST on the Loss of Load, and Rate of Change of Power channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function if required. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions. The Loss of Load sensor cannot be tested during reactor operation without causing reactor trip. The Power Rate of Change-High trip Function is required during startup operation and is bypassed when shut down or > 12% RTP.

SR 3.3.1.7 Surveillance Requirement 3.3.1.7 is a CHANNEL FUNCTIONAL TEST similar to SR 3.3.1.4, except SR 3.3.1.7 is applicable only to Functions with automatic bypass removal features. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specification tests at least once per refueling interval with applicable extensions. Proper operation of operating bypasses are critical during plant startup because the bypasses must be in place to allow startup operation and must be removed at the appropriate points during power ascent to enable certain reactor trips. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. Once the operating bypasses are removed, the bypasses must not fail in such a way that the associated trip Function gets

RPS Instrumentation-Operating B 3.3.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.3.1-34 Revision 73

4. CCNPP Setpoint File

5. Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated June 6, 1995, License Amendment Request: Extension of Instrument Surveillance Intervals

6. Combustion Engineering Topical Report CEN-327, RPS/ESFAS Extended Test Interval Evaluation dated June 2, 1986, including Supplement 1, March 3, 1989

7. Letter from Mr. D. G. McDonald (NRC) to Mr. R. E. Denton (BGE), dated October 19, 1995, Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit No. 1 (TAC No. M92479) and Unit No. 2 (TAC No. M92480)

8. Calvert Cliffs Procedure EN-4-104, Surveillance Testing

9. Combustion Engineering Owners Group Topical Report CE NPSD 1167-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, July 3, 2000

10. Attachment 1 to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing

11. TBD (Letter issuing amendments to remove RPS Loss of Load trip Function from TS)