ML20036C976

Independent Spent Fuel Storage Installation - Report of Changes, Tests, and Experiments -10 CFR 50.59 and 10 CFR 72.48

ML20036C976
Person / Time
Site:	Calvert Cliffs
Issue date:	01/31/2020
From:	Flaherty M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
Download: ML20036C976 (18)

Text

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,._. Exelon Generation January 31, 2020 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555 Mark D. Flaherty Site Vice President Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, MD 20657 410 495 5200 Office 443-534-5475 Mobile www.exeloncorp.com mark.flaherty@exeloncorp.com 10 CFR 50.59 10 CFR 72.48 Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos.. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318 Calvert Cliffs Nuclear Power Plant Independent Spent Fuel Storage Installation, License No. SNM-2505 NRC Docket No. 72-8

Subject:

Report of Changes, Tests, and Experiments -10 CFR 50.59 and 10 CFR 72.48 In accordance with 10 CFR 50.59(d){2) and 10 CFR 72.48(d)(2), a report of changes, tests and experiments is provided as Attachment (1). The attachment contains brief descriptions of changes, tests, and experiments approved under the provisions of 10 CFR 50.59 and 10 CFR 72.48 between January 1, 2018 and December 31, 2019. There were no 72.48 Evaluations performed during this time period.

There are no regulatory commitments contained in this correspondence.

Should you have questions regarding this matter, please contact Mr. Larry D. Smith at (410) 495-5219.

Respectfully,

~~~

Mark D. Flaherty Site Vice President MDF/KLG/lmd 7E47

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Document Control Desk January 31, 2020 Page 2

Attachment:

(1)

Calvert Cliffs Nuclear Power Plant Report of Changes, Tests, and Experiments

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

cc:

NRC Project Manager, Calvert Cliffs NRC Regional Administrator, Region I NRC Resident Inspector, Calvert Cliffs D. A. Tancabel, MD-DNR M. Dapas, NMSS

ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

Calvert Cliffs Nuclear Power Plant January 31, 2020

Document Id SE00563 Subject Summary ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Unit 2 Engineered Safety Features Actuation System Replacement (ESFAS)(ECP-16-000760)

Proposed Activity:

Date Issued 7/18/2018 The proposed activity replaces the existing ESFAS sensor cabinets and actuation train logic and relay cabinets. Key components of the replacement ESFAS include bistable input modules, sensor maintenance modules, discrete input modules, logic solving modules, relay output modules, logic and actuation modules, and logic and blocking modules.

Additional components of the replacement system include the chassis, required power supplies, and actuation relays.

Reason for Activity:

The proposed activity replaces the existing ES FAS cabinets in order to address obsolescence and reliability issues, including single point vulnerabilities {SPVs) that can cause spurious actuation of equipment.

Effect of Activity:

The replacement ESFAS sensor cabinets and actuation train logic and relay cabinets correspond to the existing cabinets in function and in cabinet location. The existing power feeds to the ESFAS cabinets from the 120VAC vital buses are being retained. The replacement ESFAS utilizes the existing input signals and uses the same setpoints and coincidence logic to actuate the same field equipment. The replacement ESFAS also includes features, as at present, to allow blocking of certain functions; bypassing, resetting, and testing of channels and devices; and maintenance ofthe system. Therefore, there is no change in overall ESFAS functionality. There are minor changes to the existing control room annunciators which warn of ESFAS problems and to the local indications and the features used for local testing.

==

Conclusions:==

However, the method of performing the ESFAS logic functions is slightly different than that used in the existing system.

The following aspects of the replacement ESFAS were identified as potentially adverse in the associated 50.59 screening:

1

Document Id SE00563 ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR S0.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Implementation of first-of-a-kind ES FAS hardware in a U.S. nuclear plant Date Issued 7/18/2018 Aggregation of separate channels and ESFAS functions on a single module {Bl Ms, DI Ms, LSMs, and ROMs)

Use of triple modular redundancy in the logic channels Use of a clock signal with module-to-module serial communication in the logic panels Other aspects of the modification, including v~rious system enhancements and minor changes, were "screened out" in the associated 50.59 screening. The 50.59 evaluation determined the number of components susceptible to a spurious actuation which could initiate an accident was reduced.

Other aspects of the modification, including various system enhancements and mi!1or changes, were "screened out" in the associated 50.59 screening. The 50.59 evaluation determined the number of components susceptible to a spurious actuation which could initiate an accident was reduced.

Since there is no change in overall ESFAS functionality, the replacement ESFAS will continue to mitigate the radiological consequences of accidents and malfunctions in the same manner and with the same results as the existing ESFAS.

Therefore, the proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR or in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR. Since existing safety analyses remain bounding, the proposed activity does not result in a design basis limit for a fission product barrier, as described in the UFSAR, being exceeded or altered.

The existing defense-in-depth of the plant instrumentation and control systems is maintained. Required ESFAS functions will be accomplished if a single failure in the system were to occur. Compliance with required standards and the quality processes and design features incorporated into the replacement equipment provide assurance that a common cause failure that could cause multiple failures, such that ESFAS functions would not be accomplished, is not considered credible.

Therefore, the proposed activity does not create the possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR. Further the evaluation determined that changes to the 2

Document Id SE00563 ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Date Issued 7/18/2018 calculations and evaluations involved changes to input parameters thus the replacement ESFAS does not involve an adverse change to an element of a UFSAR described methodology, or the use of an alternative methodology, that is used in establishing the design bases. Therefore, it does not result in a departure from a method of evaluation.

3

Document Id SE00565 Subject Summary ATIACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR S0.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Modify Small Break LOCA Analysis (ECP-18-000503)

Proposed Activity:

Revision Date Issued 0000 10/8/2018 This activity replaces the UFSAR Chapter 14.17, Loss of Coolant Event, Small Break LOCA portion only, with a new Analysis of Record (AOR). The new AOR establishes a new reference Peak Clad Temperature (PCT) which provides a baseline for assessment of change or errors in accordance with 10 CFR 50.46. The new analysis; a) Resolves previously reported errors, and b) Implements Small Break LOCA, Supplement 1 methodology (EMF-2328(P)(A), Revision 0, Supplement 1, Revision 0).

Reason for Activity:

Errors in Calvert Cliffs' Small Break LOCA analysis have been reported as required by 10CFR50.46. Two of the errors resulted in more than 50F change from the Analysis of Record Peak Clad Temperature (PCT). 10 CFR 50.46, Acceptance Criteria for emergency core cooling systems for light-water nuclear power reactors, requires a schedule for reanalysis if the change or error is more than 50F from the calculated Peak Clad Temperature (PCT). Calvert Cliffs schedule for reanalysis stated that AREVA was updating the Small Break LOCA topical report, and Calvert Cliffs Small Break LOCA could be reanalyzed following NRC approval of the report. The updated topical report has been approved by the NRC (ML17082A173), and this activity reanalyzes Small Break LOCA using the NRC approved updated topical report supplement.

Effect of Activity:

Plant Operation -1\\b Impact The revised Small Break LOCA analysis has no impact on plant operation. Emergency Operating Procedure 05 (EOP-05)

Technical Basis and OP-CA-I 02-106 Operator Response Time Program at Calvert Cliffs (Action TCA SIMIO) require securing RCPs within 4 minutes after Subcooling is less than 20 °F. The Supplement 1 Small Break LOCA analysis validates the time of 4 minutes.

4

Document Id SE00565 ATTACHMENT {1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR S0.59{d){2) and 10 CFR 72.48{d){2)]

Doc Type 50.59 Design Basis Rev Status Approved Revision Date Issued 0000 10/8/2018 UFSAR 14.17, Small Break LOCA portion is changed. The Unit 1 and Unit 2 COLR lists of approved methodologies require modification to add NRC approved Supplement 1, Revision 0. [The Technical Specification 5.6.5.b does not require change because it does not list revision numbers or supplements]. Editorial changes to procedures and bases that reference the current Small Break LOCA analysis are required.

The new AOR establishes a new PCT that provides the reference for assessment of future errors in or changes to the Small Break LOCA Evaluation Model. The new analysis conforms to the Technical Specification Appendix C License Conditions. The Technical Specifications Bases require revision to include credit of LPSI in Small Break LOCA analysis.

Safety Analyses described in the U FSAR UFSAR 14.17, Small Break LOCA portion is changed. The supplemented methodology resolves previously reported errors and incorporates NRC recommendations provided in the Safety Evaluation Report approving the fuel transition from Westinghouse to AREVA.

==

Conclusions:==

This activity requires a 10 CFR 50.59 Safety Evaluation because safety analyses are reanalyzed, resulting in new results for comparison to 10 CFR 50.46 acceptance criteria. The revised Small Break LOCA analysis corrects previously identified errors and resets the Analysis of Record Peak Clad Temperature. The new Supplement 1 methodology was previously approved by the NRC for this application and has been applied to a CE designed reactor. The revised Small Break LOCA analysis credit for Low Pressure Safety Injection, previously credited in Large Break LOCA analysis only. No new operational or system requirements are imposed. The reanalysis demonstrates that the 10 CFR 50.46 acceptance criteria 1 thru 4 (peak clad temperature, maximum cladding oxidation, maximum hydrogen generation (core wide oxidation), and coolable geometry) are met. Therefore, submittal of a LAR is not required.

5

Document Id SE00564 Subject Summary ATIACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Unit 1 Engineered Safety Features Actuation System Replacement (ESFAS) (ECP-17-000052)

Proposed Activity:

Date Issued 1/22/2019 The proposed activity replaces the existing Engineered Safety Features Actuation System (ESFAS) sensor channel cabinets (1C91, 1C92, 1C93, and 1C94), actuation train logic cabinets (1C67L and 1C68L), and actuation train relay cabinets (1C67 and IC68).

Reason for Activity The proposed activity replaces the existing ESFAS cabinets in order to address obsolescence and reliability issues, including single point vulnerabilities (SPVs) that can cause spurious actuation of equipment.

Effect of Activity:

The replacement sensor channel cabinets and actuation train logic and relay cabinets correspond to the existing cabinets in function and in cabinet location. The existing power feeds to the ESFAS cabinets from the 120Vac vital buses are being retained. The replacement ESFAS utilizes the existing input signals and uses the same setpoints and coincidence logic to actuate the same field equipment. The replacement ESFAS also includes features, as at present, to allow blocking of certain functions; bypassing, resetting, and testing of channels and devices; and maintenance of the system. Therefore, there is no change in overall ESFAS functionality. There are minor changes to the existing control room annunciators which warn of ESFAS problems and to the local indications and the features used for local testing. The overall power usage of the new system is less than the existing system.

The SPVs are being reduced; however, some SPVs will remain after implementation of the new system. No new SPVs are being created.

6

Document Id SE00564 ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR S0.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59

==

Conclusions:==

Rev Status Approved Revision*

Date Issued 0000 1/22/2019 The method of performing the ESFAS logic functions is slightly different than that used in the existing system. The following aspects of the replacement ESFAS were identified as potentially adverse in the associated 50.59 screening:

Implementation of first-of-a-kind ESFAS hardware in a U.S. nuclear plant Aggregation of separate channels and ESFAS functions on a single module (Bl Ms, DI Ms, LSMs, and RO Ms)

Use of triple modular redundancy in the logic channels Use of a clock signal with module-to-module serial communication in the logic panels Other aspects ofthe modification, including various system enhancements and minor changes, were "screened out" in the associated 50.59 screening. The 50.59 evaluation determined the number of components susceptible to a spurious actuation which could initiate an accident was reduced.

Since there is no change in overall ESFAS functionality, the replacement ESFAS will continue to mitigate the radiological consequences of accidents and malfunctions in the same manner and with the same results as the existing ESFAS.

Therefore, the proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR or in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR. Since existing safety analyses remain bounding, the proposed activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

The existing defense-in-depth of the plant instrumentation and control systems is maintained. Required ESFAS functions will be accomplished if a single failure in the system were to occur. Compliance with required standards and the quality processes and design features incorporated into the replacement equipment provide assurance that a common cause failure that could cause multiple failures, such that ESFAS functions would not be accomplished, is not considered credible.

Therefore, the proposed activity does not create the possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR. Further the evaluation determined that changes to the calculations and evaluations involved changes to input parameters thus the replacement ESFAS does not involve an 7

Document Id SE00564 ATIACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision Date Issued 0000 1/22/2019 adverse change to an element of a UFSAR described methodology, or the use of an alternative methodology, that is used in establishing the design bases. Therefore, it does not result in a departure from a method of evaluation.

8

Document Id SE00567 Subject Summary ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Date Issued 7/08/2019 Electrical Distribution Reliability Improvement Project (EDRIP) 500kv/13.8kV (ECP-17-00029, 000335, 000334)

Proposed Activity:

The Calvert Cliffs Electrical Distribution Reliability Improvement Project (EDRIP) installs a new plant service transformer (OXP-13000-3), which will be fed from a new breaker bay tapped off of the existing 500 kV Red and Black buses in the new, extended Switchyard Bay.

Reason for Activity:

Calvert Cliffs has had three dual unit trips in the past 6 years that have challenged operations and the site. In addition, equipment failures have occurred in the site electrical system that have complicated operator responses due to the effects of the loss of power.

Calvert Cliffs' electrical distribution system consists of two plant service transformers P-13000-1 and P-13000-2, each of which are capable of supplying both units' total plant auxiliary load. The service transformers are fed from a single 500 kV switchyard with two independent 500 kV buses (Red and Black). The Southern Maryland Electric Cooperative (SMECO) line serves as a third independent offsite source and can be manually aligned to supply the essential loads required to maintain the plant in a safe shutdown condition. Technical Specifications require two independent offsite sources to be operable; loss of one independent offsite source for a particular unit requires entry into a 72-hour Limited Condition of Operation (LCO) 3.8.1.a and 3.8.1.c.

The overall intent of this modification is to improve the reliability of the Calvert Cliffs' electrical distribution system. Primary objectives are to:

1.

Prevent a Unit trip on the loss of a plant service transformer.

2.

Avert entry into a dual Unit LCO when a plant service transformer is out of service.

3.

Provide the ability to restore off-site power to the site as quickly as possible.

4.

Afford a means to perform transformer maintenance during an on-line work window completely isolated 9

Document Id SE00567 ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR S0.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved from on-site and off-site sources.

Effect of Activity:

Revision Date Issued 0000 7/08/2019 The existing Unit 1Service Transformer (P-13000-1} is directly connected to the Black Bus and the existing Unit 2 Service Transformer (P-13000-2} is directly connected to the Red Bus. This modification will install a Hybrid Breaker (OBKR 552-31} between Black Bus and P-13000-1 high voltage side and OBKR552-51 between Red Bus and P-13000-2 high voltage side which are capable of operating from Control Room using new EDCS or the transmission/distribution Baltimore Gas and Electric (BGE} Switchyard Control House. Panels 1PNL1C225 &

2PNL2C225 are designed to be installed in Switchyard Control House and connected to the EDCS via fiber optics for plant operator to manipulate the Hybrid Breakers via HM ls installed in the Control Room (due fo space limitation a Hybrid Breaker design with integrated disconnect switches was selected}.

==

Conclusions:==

Addition of the 5 breakers, two current limiters trip initiator (low voltage trip circuit} and EDCS do not create any new functions or direct/indirect interfaces with important to safety components that are not included in the system as currently configured. The system does not have any interaction with any important to safety components and will not change the environment of any important to safety SSC. This modification does not change the function or performance requirements for the system as described in the UFSAR. The additio*n of new equipment does not increase any plant operating parameters that would result in increased challenge to important to safety components. The Power Separation at 4160V lE power system is not changed as a result of this modification.

The proposed engineering change activity evaluated under this 50.59 Evaluation was prepared to nuclear digital equipment lifecycle design process standards. Additionally, the proposed engineering change activity did not replace any non-digital system or component. A failure mode evaluation assessment was performed under this activity to assure the 10

Document Id SE00567 ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Date Issued 7/08/2019 digital control and relay functions provide equivalent or better reliability to similar, existing 13.SkV power distribution equipment presently installed. The 50.59 Evaluation determined submittal of a LAR was not required.

11

Document Id SE00571 Subject Summary ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR S0.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Containment Air Cooler Replace 42/L Relays & Add Surge Suppression (ECP-19-000214)

Proposed Activity:

Date Issued 9/18/2019 The proposed Activity modifies control logic for the Containment Air Coolers (CAC) to include a nominal one second time delay for a start demanded at low speed. A general-purpose control relay (42/L) is replaced in each CAC slow speed control logic circuit with a time delay relay to accomplish that delay. The process computer input associated with a CAC low speed start demand is moved from relay 42/L to a parallel auxiliary relay 42X/L due to fewer available contacts on the new time delay relay for 42/L. The new time delay associated with relay 42/L also impacts the logic associated with the CAC Auto Start Failure alarm in the Main Control Room.

Surge suppressors are being installed across the starter contactor coils associate with both the low and fast speed operation of the CACs.

Reason for Activity:

The CAC motor starters have a history of failing when shifting fan motor speed during quarterly Safety Injection Actuation System (SIAS) testing. CAC failures are highly undesirable since these components are relied upon to provide a safety function of cooling containment during a Design Basis Accident, and a CAC failure places the plant in a 7-day LCO. Additionally, there is a potential for starter failure to cause damage to CAC motors, requiring emergent replacement in Containment.

The cause of the failures has been attributed to timing overlaps of the control relay contacts in each fan motor start circuitry which momentarily actuates the fast and slow speed contactors at the same time. The proposed activity will install a time delay relay in the slow speed start circuit and will install surge suppressors across each of the slow and fast speed contactor coils for each CAC. The time delay relay will provide a robust barrier to prevent energizing both the slow and fast speed contactors simultaneously on receipt of a SIAS. The surge suppressors reduce the contactor 12

Document Id SE00571 ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS (10 CFR S0.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Date Issued 9/18/2019 coil dropout voltage to within the rated voltage for the relay contacts. The proposed Activity will eliminate contact arcing and contactor cycling to prevent failures and increase the reliability of the existing electrical interlocks in the CAC control logic.

Effect of Activity:

There is no change in the design bases or plant operation as described in the UFSAR other than that associated with the addition of a time delay in the slow speed start circuit for the CACs.

The time delays introduced by the proposed Activity effectively extend the duration from a CAC start demand, by control switch action, an automatic start signal from SIAS with offsite power available, or an automatic start signal from SIAS and the Loss of Coolant Incident {LOCI) Sequencer with offsite power unavailable, to a CAC motor start (contactor energization) in slow speed by one second.

The delay in the CAC impacts the containment integrity analysis for both a Loss of Coolant Accident (LOCA) and a Main Steam Line Break (MSLB). Currently, these analyses include a 0.9 second CAC actuation signal delay in the equipment response time delays of 35 seconds for a LOCA and 10 seconds for a MSLB for the CAC to start and develop full flow on slow speed. The containment integrity analysis is revised under the proposed Activity to include a 1.2 second delay for the new time delay relay to account for the setting tolerance (+/-0.2 seconds) during relay calibration. The total response time for the CACs increases from 35.9 seconds to 37.1 seconds for a LOCA and from 10.9 seconds to 12.1 seconds for a MSLB. The critical containment parameters affected by the delay are the peak pressure, the time to peak pressure, the peak temperature, and the time to peak temperature.

The new time delay associated with relay 42/L will delay the loading of the CACs on their respective diesel generator by one second following a motor start demanded by the LOCI Sequencer. The duration from the closure of the diesel generator output breaker to the slow speed contactor energizing is extended from 15 seconds to 16 seconds.

The new time delay associated with relay 42/L effectively reduces the LOCI Sequencer Step No. 3 loading interval for 13

Document Id SE00571 ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Date Issued 9/18/2019 the CACs from 5 seconds to 4 seconds for diesel generator voltage and frequency recovery prior to the LOCI Sequencer Step No. 4 loads being energized by the diesel generator. Change Notice ECP-19-000214-CN-032 will revise Calculation E-92-046 to reflect the shorter sequencer load interval for diesel generator recovery.

The ES FAS Response Time testing program and associated procedures is impacted by the additional delay from a CAC start demand to reach normal motor running current in slow speed. Currently this CAC equipment response time is limited to 10 seconds in the acceptance criteria for ESFAS Response Time testing.

The new time delay relay is interlocked with an existing time delay relay 2/SIA8 (2/SIB8) in the circuit associated with the CAC Auto Start Failure alarm in the Main Control Room. The new time delay relay associated with relay 42/L increases the duration from a CAC start demand on slow speed to recognizing that relay 42/L successfully energizes and changes state to one second from essentially zero seconds. The time delay associated with relay 2/SIA8 (2/SIB8) is extended from one second to two seconds to preclude nuisance alarms, otherwise there is no operational impact on the CACs due to this change in the alarm circuit.

==

Conclusions:==

The proposed activity does not introduce the possibility of a change in the frequency of an accident because the CACs are not an initiator of any accident. Any credible failure mode for the new relay, surge suppressors, and associated circuit connections will not initiate an event previously evaluated in the UFSAR.

The impact of the one second time delay due to this proposed activity has been evaluated in the containment integrity analyses and the changes in peak vapor temperature and peak pressure are insignificant for both the limiting LOCA and MSLB events analyzed. The calculated values for peak temperature and peak pressure remain within the maximum values currently reflected in the UFSAR.

An assessment determined that any delay in the start of the CACs after the LOCI sequencer Step 3 demand at 15 14

Document Id SE00571 ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2)]

Doc Type 50.59 Rev Status Approved Revision 0000 Date Issued 9/18/2019 seconds from diesel generator output breaker closure would not significantly impact diesel generator voltage and frequency transient performance or the ability to accept all ESFAS loads in a manner that would prevent it from performing its intended design function. Therefore, there is no potential for the new time delay to impact entire trains of Engineered Safety Feature (ESF) equipment. The affected containment response analyses either considered or evaluated the loss of a CAC train or a loss of a single emergency power train and associated ESF equipment so the limiting failure associated with the proposed change would remain bounded.

The introduction of a time delay in the slow speed start circuit for the CACs has an impact on containment peak temperature and pressure for the LOCA and MSLB containment response. However, a sensitivity case run was performed for the LOCA containment response with a 12 second CAC actuation which showed containment pressure remained less than the design pressure of 50 psig. Therefore, the proposed activity does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

A more current version of GOTHIC software is used in Revision I to Calculation CA07725 to support the analysis of the impact of an additional one second delay in CAC response time. This calculation revision performs benchmarking runs against the currently calculated values for peak pressure, time to peak pressure, peak vapor temperature, and time to peak vapor temperature in the current calculation for both the limiting LOCA and MSLB events. The results of the benchmarking indicate that the difference is insignificant such that the results are essentially the same and within typical rounding methods. Therefore, since the newer version of GOTHIC (Version 8.3(QA)) generates results "essentially the same" as the version of GOTHIC (Version 7.2a(QA)) in the current revision of the calculation, the proposed activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Based on the conclusions from 50.59 Evaluation SE00571, the proposed activity can be implemented without obtaining a license amendment change.

15