IR 05000317/2020003
| ML20318A160 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/13/2020 |
| From: | Matt Young NRC/RGN-I/DRP/PB5 |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| Young M | |
| References | |
| IR 2020003 | |
| Download: ML20318A160 (24) | |
Text
November 13, 2020
SUBJECT:
CALVERT CLIFFS NUCLEAR POWER PLANT, UNITS 1 AND 2 -
INTEGRATED INSPECTION REPORT 05000317/2020003 AND 05000318/2020003
Dear Mr. Hanson:
On September 30, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Calvert Cliffs Nuclear Power Plant, Units 1 and 2. On October 20, 2020, the NRC inspectors discussed the results of this inspection with Mr. Thomas Haaf and other members of your staff. The results of this inspection are documented in the enclosed report.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. One Severity Level IV violation without an associated finding is documented in this report. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at Calvert Cliffs Nuclear Power Plant, Units 1 and 2.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; and the NRC Resident Inspector at Calvert Cliffs Nuclear Power Plant, Units 1 and 2. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, X /RA/
Signed by: Matthew R. Young
Matt R. Young, Chief Reactor Projects Branch 5 Division of Reactor Projects
Docket Nos. 05000317 and 05000318 License Nos. DPR-53 and DPR-69
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000317 and 05000318
License Numbers:
Report Numbers:
05000317/2020003 and 05000318/2020003
Enterprise Identifier: I-2020-003-0071
Licensee:
Exelon Generation Company, LLC
Facility:
Calvert Cliffs Nuclear Power Plant, Units 1 and 2
Location:
Lusby, MD
Inspection Dates:
July 1, 2020 to September 30, 2020
Inspectors:
H. Anagnostopoulos, Senior Health Physicist
E. Bousquet, Resident Inspector
J. Brand, Reactor Inspector
R. Clagg, Senior Resident Inspector
J. DeBoer, Reactor Inspector
K. Mangan, Senior Reactor Inspector
J. Rady, Emergency Preparedness Inspector
C. Roettgen, Senior Resident Inspector
A. Turilin, Reactor Inspector
Approved By:
Matt R. Young, Chief
Reactor Projects Branch 5
Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Calvert Cliffs Nuclear Power Plant, Units and 2, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors.
Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Implement a Procedure to Mitigate Jellyfish Intrusion Event Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000317,05000318/2020003-01 Open/Closed
[P.2] -
Evaluation 71111.11Q The inspectors documented a self-revealing Green finding and associated non-cited violation of Calvert Cliffs Nuclear Power Plant Unit 2 Technical Specification 5.4.1.a, for the licensees failure to implement written procedures as required by Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, Appendix A, Section 5,
Procedures for Abnormal, Offnormal, or Alarm Conditions. Specifically, the licensee failed to implement procedure OP-CA-119, Jellyfish Intrusion Mitigation Strategy, Revision 0, following indications of high concentrations of jellyfish in the intake bay.
Failure to Obtain Prior NRC Approval for Change to a Design Basis Limit for a Fission Product Barrier Cornerstone Significance Cross-Cutting Aspect Report Section Not Applicable NCV 05000317,05000318/2020003-02 Open/Closed Not Applicable 71111.17T The inspectors identified a Severity Level IV non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.59(c)(1) because the licensee failed to obtain NRC approval prior to changing a design basis limit for a fission product barrier.
Additional Tracking Items
None.
PLANT STATUS
Unit 1 operated at or near rated thermal power for the entire inspection period.
Unit 2 operated at or near rated thermal power for the entire inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
Starting on March 20, 2020, in response to the National Emergency declared by the President of the United States on the public health risks of the coronavirus (COVID-19), resident and regional inspectors were directed to begin telework and to remotely access licensee information using available technology. During this time the resident inspectors performed periodic site visits each week, increasing the amount of time on site as local COVID-19 conditions permitted. As part of their onsite activities, resident inspectors conducted plant status activities as described in IMC 2515, Appendix D; observed risk significant activities; and completed on site portions of IPs. In addition, resident and regional baseline inspections were evaluated to determine if all or portion of the objectives and requirements stated in the IP could be performed remotely. If the inspections could be performed remotely, they were conducted per the applicable IP. In some cases, portions of an IP were completed remotely and on site. The inspections documented below met the objectives and requirements for completion of the IP.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Impending Severe Weather Sample (IP Section 03.02) (2 Samples)
The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending adverse weather conditions for:
- (1) Units 1 and 2, forecasted severe thunderstorms and high winds, July 28, 2020
- (2) Units 1 and 2, forecasted impacts from Tropical Storm Isaias, August 3, 2020
External Flooding Sample (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated readiness to cope with external flooding for protected area storm drains and Unit 1, 1A emergency diesel generator building on September 14, 2020.
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 2, 22 high pressure safety injection pump train during 21 high pressure safety injection pump train out of service for maintenance, July 28, 2020
- (2) Unit 2, 21 containment spray train during 22 containment spray train out of service for maintenance, August 13, 2020
- (3) Unit 1, 11 shutdown cooling heat exchanger during 12 shutdown cooling heat exchanger out of service for maintenance, September 8, 2020
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Unit 2, 21 and 22 emergency core cooling systems, fire areas 1 and 2, July 28, 2020
- (2) Units 1 and 2, horizontal chase and 69' electrical room, fire areas 35-38, July 28, 2020
- (3) Unit 2, service water pump room, fire area 15, August 13, 2020
- (4) Unit 1, 1A emergency diesel generator building, fire area EDG1A, September 14, 2020 (5)0C (station blackout) diesel generator building, fire area EDG0C, September 14, 2020
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the fire brigade performance during an unannounced drill on August 27, 2020.
71111.06 - Flood Protection Measures
Inspection Activities - Internal Flooding (IP Section 03.01) (1 Sample)
The inspectors evaluated internal flooding mitigation protections in the:
- (1) Unit 1, service water pump room, September 23, 2020
Cable Degradation (IP Section 03.02) (1 Sample)
The inspectors evaluated cable submergence protection in:
- (1) Manholes 1MH21, 1MH24, 1HH25, 1HH26, for cables associated with the 1A emergency diesel generator and 0C diesel generator, August 17, 2020
71111.07T - Heat Sink Performance Triennial Review (IP Section 03.02)
(1)12 component cooling water heat exchanger, cooled by service water, Section 02.02.b (2)1B emergency diesel generator jacket water cooler, closed loop cooling, Section 02.02.c (3)2B emergency diesel generator lube oil cooler, closed loop cooling, Section 02.02.c (4)0C station blackout emergency diesel generator room cooler, air cooled, Section 02.02c
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (2 Samples)
- (1) The inspectors observed and evaluated Unit 2 licensed operator performance in the main control room in response to a decrease in circulating water flow due to a jellyfish intrusion event and subsequent unit downpower on July 13, 2020.
- (2) The inspectors observed and evaluated Units 1 and 2 licensed operators performance in the main control room during periods of increased risk for U-4000-12 service transformer maintenance in conjunction with conducting surveillance testing on September 28, 2020.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated a training event involving loss of feedwater, reactor trip, loss of safety related buses, loss of reactor coolant system inventory, and degraded containment with fuel damage indication resulting in a General Emergency declaration on August 11, 2020.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (3 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Unit 2, AR04356491, power reduction for circulating water debris intrusion, September 22, 2020
- (2) Unit 1, AR04370224, 11 charging pump has excessive leakage, September 25, 2020
- (3) Unit 2, AR04369586, 26B traveling screen will not rotate in hand, September 28, 2020
Quality Control (IP Section 03.02) (1 Sample)
The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:
- (1) Unit 1, WO C93716946, 1A emergency diesel generator 13, 24-month inspection, September 30, 2020
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (3 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
- (1) Unit 1, Yellow risk condition due to 13 high pressure safety injection pump train out of service for maintenance, July 9, 2020
- (2) Unit 2, Yellow risk condition due to 21 high pressure safety injection pump train out of service for maintenance, July 28, 2020
- (3) Unit 2, Yellow risk condition due to 22B service water heat exchanger cleaning train out of service for maintenance, August 14, 2020
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Unit 1, AR04358331, slight biological growth in the 1B emergency diesel generator day tank, July 29, 2020
- (2) Unit 1, AR04361745, 12 high pressure safety injection pump coupling alignment issues on vibration data, August 12, 2020
- (3) Unit 2, AR04362114, elevated leak rate on 21 service water pump outboard oil reservoir, August 13, 2020
- (4) Units 1 and 2, AR04362934, the entire sub-surface drain system is blocked, August 14, 2020
- (5) Unit 2, AR04369012, active leak on 22 low pressure safety injection pump outboard pump seal, September 14, 2020
71111.17T - Evaluations of Changes, Tests, and Experiments Sample Selection (IP Section 02.01)
The inspectors reviewed the following evaluations, screenings, and/or applicability determinations for 10 CFR 50.59 from [June 15, 2020 to June 19, 2020].
Safety Evaluations
- (1) SE00561, 1A Emergency Diesel Generator Governor System Upgrade (ECP-16-
===000742), Revision 0
- (2) SE00563, Unit 2, Engineered Safety Features Actuation System Replacement (ECP-16-000760), Revision 0
- (3) SE00564, Unit 1, Engineered Safety Features Actuation System Replacement (ECP-17-000052), Revision 0
- (4) SE00565, Modify Small Break Loss of Coolant Accident Analysis (ECP-18-000503),
Revision 0
- (5) SE00567, Electrical Distribution Reliability Improvement Project 500KV/13.8 KV (ECP-17-00029, 000335, 000334), Revision 0
- (6) SE00569, Revise the Modified Barnett Correlation Limit used in Steam Line Break Methodology (ECP-19-000380), Revision 0
- (7) SE00571, Containment Air Cooler Replace 42/L Relays & Add Surge Suppression (ECP-19-000214), Revision 0
- (8) SE00573, Implementation of CE-STAR Topical WCAP-16011-P-A at Calvert Cliffs (ECP-20-000050), Revision 0
Screens (Code of Federal Regulations 50.59 Screened-out Evaluations)
- (9) ECP-14-000184, Auxiliary Feedwater Pump Room Steam Traps and Drains Upgrades, Revision 2
- (10) ECP-14-000864, ESR-12-000896, Replace Control Room Ventilation Radiation Monitor, Revision 0
- (11) ECP-16-000588, ESR-16-000067 ESR (0000) - IR 02453451 FCR 84-0149 Installed a 20 Second Time Delay on the Auxiliary Feedwater Actuating System Start Circuit. A Plus or Minus 2 Second Tolerance (18-22 seconds) was Specified in the Acceptance Criteria, Revision 0
- (12) ECP-17-000261, Temporary Change to Replace 2FT5211/2FT5212 with a Different Model, Revision 0
- (13) ECP-17-000334, Electrical Distribution Reliability Improvement Project - Unit 2 Tie-In, Revision 0
- (14) ECP-17-000470, Revise RM-2000 Data Base to Allow a Fix Flow Value for Units 1 and 2 Wide Range Noble Gas Monitor Systems, Revision 0
- (15) ECP-17-000473, ESR-17-000067 ESR (0000) - Install a Fly-Back Diode Around ALR Relay Coil in 1B, 2A, and 2B Emergency Diesel Generator Control Circuits, to Prevent Inductive Kick on the 125 Volts Direct Current Bus, Revision 0
- (16) ECP-17-000548, ESR-17-000378 ESR (0000) - Provide a T Mod to Remove the Fire Detection System from the Unit 1 27' and 45 Switchgear Rooms, Revision 0
- (17) ECP-17-000579, Use of American Society of Mechanical Engineers Code Higher Allowable (see AR02688235-07), Revision 0
- (18) ECP-17-000597, Evaluate Saltwater System Piping Stress Due to Replacement of 1-HVSW-112 (MR90 Temporary Change), Revision 0
- (19) ECP-17-000600, ESR-17-000352 Modify Piping at 1/2 HVSI-510 Test Connection to Allow for Permanent Installation of Local Pressure Indicator for Monitoring MOV-652 Leakage at Power, Revision 0
- (20) ECP-17-000734, Unit 2 Component Cooling System Containment Isolation Valve and Service Water System Turbine Building Isolation Valve Redundant Solenoid Valve Modification, Revision 0
- (21) ECP-18-000121, Lost Parts Evaluation (Flashlight in Containment), Revision 0
- (22) ECP-18-000160, Provide Justification to Allow Unit Startup and Run with Disengaged Core Shroud Tie Rod - Unit 1, Revision 0
- (23) ECP-18-000339, ESR-17-000692 ESR (0000) - Fairbanks Morse LS Circuit Seal-in Contact Modification, Revision 0
- (24) ECP-18-000534, Design Analysis for Operation of Calvert Cliffs Units 1 and 2 with 4 or More Functional Reactor Vessel Internals Core Shroud Tie Rods, Revision 0
- (25) ECP-18-000610, Perform Tech Eval of PCTCC TA-230 - Substitute Fixed Value for Main Vent Stack Flow on Units 1 and 2 Wide Range Noble Gas Monitor, Revision 0
- (26) ECP-18-000617, Replace Units 1 and 2 Main Vent Stack Flow Instruments, Revision 0
- (27) ECP-18-000631, Perform an Evaluation to Determine if the New Isolation Valves in the Emergency Diesel Generator Common Air Start Header Should be Changed from Normally Open to Normally Closed (LOCKED CLOSED), Revision 0
- (28) ECP-19-000049, Revise Instrument Loop Uncertainty Calculation for Surveillance Interval Changes, Revision 0
- (29) ECP-19-000239, ESR-19-000275 ESR (0000) - Create New Voltage Drop Calculations for Unit 1 120VAC Vital Panels, Revision 0
- (30) ECP-19-000540, ESR-19-000511 ESR (0000) - Op Eval for 2CV5150 Which Did Not Fully Stroke During STP O-065N-2, Revision 0
71111.19 - Post-Maintenance Testing
Post-Maintenance Test Sample (IP Section 03.01)===
The inspectors evaluated the following post maintenance test activities to verify system operability and functionality:
- (1) Unit 2, WO C93721101, lubricate and inspect 2MOV4143, 21 refueling water tank discharge valve, July 28, 2020
- (2) Unit 2, WO C93289350, overhaul 2HVSI-324, 22 containment spray pump discharge valve, August 13, 2020
- (3) Unit 1, WO C93716946, 1A emergency diesel generator maintenance, 24-month inspection of Société Alsacienne de Constructions Mécaniques diesel generators, emergency diesel generator 13, August 25, 2020
- (4) Unit 1, WO C93720302, replace 1PCV3830, 12 shutdown cooling heat exchanger component cooling outlet valve, September 8, 2020
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance tests:
Surveillance Tests (other) (IP Section 03.01)
- (1) Units 1 and 2, STP-M-260-0, "Seismic Instrumentation Channel Check," Revision 008, August 11, 2020
- (2) Units 1 and 2, WO C 93719405, "Function Tests of Reactor Coolant System Make-Up Pump," September 15, 2020
Inservice Testing (IP Section 03.01) (1 Sample)
- (1) Unit 1, STP-O-005A, "AFW System Quarterly Surveillance Test," Revision 8, July 20, 2020
71114.02 - Alert and Notification System Testing
Inspection Review (IP Section 02.01-02.04) (1 Sample)
- (1) The inspectors evaluated the licensee's maintenance and testing of the alert and notification system on August 17-20, 2020, for the period of August 2018 through July 2020.
71114.03 - Emergency Response Organization Staffing and Augmentation System
Inspection Review (IP Section 02.01-02.02) (1 Sample)
- (1) The inspectors evaluated the readiness of the licensee's emergency preparedness organization on August 17-20, 2020.
71114.04 - Emergency Action Level and Emergency Plan Changes
Inspection Review (IP Section 02.01-02.03) (1 Sample)
- (1) The inspectors evaluated the following submitted Emergency Action Level and Emergency Plan changes onsite on August 17-20, 2020.
- Evaluation No.: 19-83, EP-AA-1011, Exelon Nuclear Radiological Emergency Plan Annex for Calvert Cliffs Station, Revision 4
- Evaluation No.: 19-85, EP-AA-1011, Exelon Nuclear Radiological Emergency Plan Annex for Calvert Cliffs Station, Revision 5
- Evaluation No.: 19-90, EP-AA-1011, Addendum 1, Calvert Cliffs Nuclear Power Plant On-Shift Staffing Technical Basis, Revision 1
- Evaluation No.: 20-34, EP-AA-1011, Addendum 3, Calvert Cliffs Nuclear Power Plant Emergency Action Levels, Revision 6
This evaluation does not constitute NRC approval.
71114.05 - Maintenance of Emergency Preparedness
Inspection Review (IP Section 02.01 - 02.11) (1 Sample)
- (1) The inspectors evaluated the maintenance of the emergency preparedness program on August 17-20, 2020, for the period of August 2018 through July 2020.
71114.06 - Drill Evaluation
Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated the conduct of an emergency preparedness drill involving a loss of saltwater cooling and a loss of coolant accident resulting in a General Emergency declaration on September 15, 2020.
Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)
The inspectors evaluated:
- (1) The inspectors observed and evaluated the conduct of a simulator training event involving a loss of feedwater, reactor trip, loss of safety related buses, loss of reactor coolant system inventory, and degraded containment with fuel damage indication resulting in a General Emergency declaration on August 18,
RADIATION SAFETY
71124.03 - In-Plant Airborne Radioactivity Control and Mitigation
Permanent Ventilation Systems (IP Section 03.01) (1 Sample)
The inspectors evaluated the configuration of the following permanently installed ventilation systems:
- (1) Unit 1, containment penetration room Unit 2, emergency core cooling system pump room
Temporary Ventilation Systems (IP Section 03.02) (1 Sample)
The inspectors evaluated the configuration of the following temporary ventilation systems:
- (1) All systems associated with steam generator primary side inspection
Use of Respiratory Protection Devices (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated the licensees use of respiratory protection devices.
Self-Contained Breathing Apparatus for Emergency Use (IP Section 03.04) (1 Sample)
- (1) The inspectors evaluated the licensees use and maintenance of self-contained breathing apparatuses.
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
EP01: Drill/Exercise Performance (IP Section 02.12)===
- (1) July 1, 2019 - June 30, 2020
EP02: ERO Drill Participation (IP Section 02.13) (1 Sample)
- (1) July 1, 2019 - June 30, 2020
EP03: Alert & Notification System Reliability (IP Section 02.14) (1 Sample)
- (1) July 1, 2019 - June 30, 2020
MS06: Emergency AC Power Systems (IP Section 02.05) (2 Samples)
- (1) Unit 1, July 1, 2019 - June 30, 2020
- (2) Unit 2, July 1, 2019 - June 30, 2020
MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)
- (1) Unit 1, July 1, 2019 - June 30, 2020
- (2) Unit 2, July 1, 2019 - June 30, 2020
MS08: Heat Removal Systems (IP Section 02.07) (2 Samples)
- (1) Unit 1, July 1, 2019 - June 30, 2020
- (2) Unit 2, July 1, 2019 - June 30, 2020
MS09: Residual Heat Removal Systems (IP Section 02.08) (2 Samples)
- (1) Unit 1, July 1, 2019 - June 30, 2020
- (2) Unit 2, July 1, 2019 - June 30, 2020
MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)
- (1) Unit 1, July 1, 2019 - June 30, 2020
- (2) Unit 2, July 1, 2019 - June 30,
INSPECTION RESULTS
Failure to Implement a Procedure to Mitigate Jellyfish Intrusion Event Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems
Green NCV 05000317,05000318/2020003-01 Open/Closed
[P.2] -
Evaluation 71111.11Q The inspectors documented a self-revealing Green finding and associated non-cited violation of Calvert Cliffs Nuclear Power Plant Unit 2 Technical Specification 5.4.1.a, for the licensees failure to implement written procedures as required by Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, Appendix A, Section 5, Procedures for Abnormal, Offnormal, or Alarm Conditions. Specifically, the licensee failed to implement procedure OP-CA-119, Jellyfish Intrusion Mitigation Strategy, Revision 0, following indications of high concentrations of jellyfish in the intake bay.
Description:
On July 13, 2020, the licensee identified high differential pressure across Unit 2 traveling screens 26A and 26B, which provide suction screening for the 26 circulating water pump, and a resulting decrease in main condenser vacuum. In addition, the licensee identified low saltwater cooling flow on the 21A/B service water heat exchanger, which is provided by the 23 saltwater pump and shares an intake bay with 26 circulating water pump.
The licensee took immediate actions to lower reactor power to 94 percent to stabilize main condenser vacuum and attempted to restore saltwater cooling to the 21A/B service water heat exchanger by starting the 21 saltwater pump. Saltwater flow remained below the minimum required value and the A trains of the safety-related saltwater, service water, containment spray, containment cooling, and the emergency core cooling subsystems, in addition to the 2A emergency diesel generator, were declared inoperable and entry into the applicable technical specification limiting conditions for operation were made.
Initial investigation by the licensee identified a large intrusion of jellyfish onto the 26A and 26B traveling screens with carryover into the suction of the 23 saltwater pump as the direct cause of the high differential pressure on the traveling screens and the subsequent decreases in main condenser vacuum and saltwater cooling flow to the 21A/B service water heat exchanger. The licensee subsequently took actions to clean the 26A and 26B traveling screens to restore main condenser vacuum and allow return of Unit 2 to 100 percent reactor power. Additionally, the 21 A/B service water heat exchanger was cleaned to restore saltwater cooling flow above the minimum required level and support the restoration of the A train subsystems listed above to operable status.
The inspectors reviewed licensee procedure, OP-CA-119, Jellyfish Intrusion Mitigation Strategy, Revision 0, and noted Attachment 1 provided a checklist designed to implement a mitigation strategy when high concentrations of jellyfish are occurring. Attachment 1 provided site considerations such as installing additional spray pumps below the water line to break up jellyfish concentrations, running all or selected traveling screens in hand mode, stationing operators and firehoses at the waterfront to monitor traveling screen conditions and break up jellyfish streams, and increasing jellyfish monitoring.
The inspectors reviewed the licensees main control room narrative logs for Units 1 and 2 and noted 10 log entries that documented impact on service water heat exchanger performance or traveling screen differential pressure due to jellyfish beginning on July 2, 2020, and continuing up to the Unit 2 down power on July 13, 2020. The inspectors reviewed several corrective action program documents (AR04354687, AR04355618, AR04356083, AR04356491, AR04348874) and noted the identification of high concentrations of jellyfish in the intake bay and impacts on the sites cooling water systems. The inspectors also noted that the licensee attempted to initiate a bubbler curtain system to mitigate the jellyfish intrusion impact, however, the system was not in service at the time of the event. The inspectors determined that a high concentration of jellyfish was present in the intake bay and concluded that the licensee failed to implement procedure OP-CA-119.
Corrective Actions: The licensees corrective actions included staging fire hoses to disperse jellyfish off traveling screens, briefing turbine building operators on flushing saltwater strainers, and providing daily environmental conditions for the plan of the day.
Corrective Action References: AR04356491, AR04371117
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to implement Exelon procedure OP-CA-119, Jellyfish Intrusion Mitigation Strategy, Revision 0, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors reviewed Inspection Manual Chapter 0612, Appendix B, Issue Screening, issued on December 12, 2019, and determined the issue is more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected its cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to implement procedure OP-CA-119, Jellyfish Intrusion Mitigation Strategy, Revision 0, resulted in the inoperability of the Unit 2 A train of the saltwater subsystem and as a result the inoperability of the A train of the service water, containment spray, containment cooling, and the emergency core cooling subsystems, in addition to the 2A emergency diesel generator.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, issued on December 20, 2019, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings at Power issued on December 13, 2019, the inspectors determined that this finding is of very low safety significance (Green).
Using Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 1 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings at Power, the inspectors screened this finding for safety significance and determined that a detailed risk evaluation was required. Specifically, the degraded condition resulted in the partial loss of the saltwater subsystem due to the degradation of flowrates below minimum design requirements for the 21A and 21B service water heat exchangers.
A Region I senior reactor analyst completed the detailed risk evaluation and estimated the increase in core damage frequency associated with the performance deficiency to be 2E-9/year or of very low safety significance (Green). To perform the detailed risk evaluation, the senior reactor analyst used the Systems Analysis Programs for Hands-On Integrated Reliability Evaluation, Revision 8.2.2, and Standardized Plant Analysis Risk Model, version 8.64 for Calvert Cliffs Unit 2. The senior reactor analyst set basic events SRW-HTX-PE-21A and SRW-HTX-PE-21B to TRUE. This reflected the degradation of the heat exchangers due to an environmental condition. The exposure time of the condition was estimated at a nominal 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />. The analyst performed the risk assessment using a conservative exposure time assumption of one day.
The dominant core damage sequence involved a loss of the 4 kilovolt Bus 24, with failure to isolate a stuck open power operated relief valve, with failure of shutdown cooling and failure of containment cooling. In accordance with Inspection Manual Chapter 0609 guidance, a review of the external event risk or effect on large early release frequency was not performed given the increase in internal event core damage frequency was below the 1E-7/year threshold and would not be expected to change the risk estimation.
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. The inspectors reviewed Inspection Manual Chapter 0310, Aspects Within Cross Cutting Areas, issued on February 25, 2019, and determined that this finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because the licensee failed to evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to evaluate the identified high concentration of jellyfish and ensure that resolutions addressed causes and extent of condition.
Enforcement:
Violation: Calvert Cliffs Nuclear Power Plant Unit 2 Technical Specification 5.4.1.a, requires, in part, written procedures shall be implemented covering Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, Appendix A, Section 5, Procedures for Abnormal, Offnormal, or Alarm Conditions. Contrary to the above, from May 1, 2020 to July 13, 2020, the licensee failed to implement a written procedure covering Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, Appendix A, Section 5, Procedures for Abnormal, Offnormal, or Alarm Conditions.
Specifically, the licensee failed to implement OP-CA-119, Jellyfish Intrusion Mitigation Strategy, Revision 0, following indications of high concentrations of jellyfish in the intake bay.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Obtain Prior NRC Approval for Change to a Design Basis Limit for a Fission Product Barrier Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000317,05000318/2020003-02 Open/Closed
Not Applicable 71111.17T The inspectors identified a Severity Level IV non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.59(c)(1) because the licensee failed to obtain NRC approval prior to changing a design basis limit for a fission product barrier.
Description:
The inspectors reviewed the licensees 50.59 Safety Evaluation SE569, Revise the Modified Barnett Limit Used In Steam Line Break Methodology, Revision 0, to determine if the 50.59 safety evaluation correctly concluded that the change to the 95/95 departure from nucleate boiling ratio Barnett Correlation value could be made in accordance with the requirements of 10 CFR 50.59 without a license amendment. The inspectors noted that the change was needed after the licensees fuel vendor identified that the 95/95 departure from nucleate boiling ratio design basis limit for the Barnett Correlation value, previously approved by the NRC for the Calvert Cliff Units 1 and 2, was incorrect.
The inspectors noted that the licensee was informed by their fuel vendor that the 95/95 departure from nucleate boiling ratio design basis limit for the Barnett Correlation value needed to be changed as a result of new fuel experimental data. The fuel vendor indicated that the previous value of 1.135 should be increased to account for non-conservative behavior and recommended the value be changed to an equation that expresses the value based on core exit pressure. The licensee entered the issue into their corrective action program as AR04182536. As part of the modification engineering change package, ECP-19-000380, developed to implement the change to the new 95/95 Barnett Correlation value, the licensee completed SE569 on February 6, 2020, to evaluate whether this change to the 95/95 departure from nucleate boiling ratio design basis limit required a license amendment from the NRC. SE569 documented the licensees conclusions that an amendment was not required and ECP-19-000380 was implemented.
The inspectors reviewed SE569 and noted that the NRC approved the 95/95 departure from nucleate boiling design basis limit for the Barnett Correlation value of 1.135 in Calvert Cliffs Unit 1 License Amendment 297 and Calvert Cliffs Unit 2 License Amendment 273, ADAMS Accession No. ML110390224. The inspectors noted, that in Section 5 of the NRC Safety Evaluation for License Amendments 297 and 273, the NRC staff documented their consideration of the use of the previously approved 95/95 departure from nucleate boiling design basis limit for the Barnett Correlation value.
5.1 Anticipated Operational Occurrences - The licensee is implementing the method described by AREVA NP licensing topical report EMF-2310(P)(A), Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," to analyze AOOs [anticipated operational occurrences]. The NRC staff reviewed the methodology, the CCNPP-specific implementation of the methodology, and the modeling assumptions and results of specific postulated transients, as described in the following sections.
5.1.1 Methods Implementation - Topical Report EMF-2310 (P)(A) pertains to non-LOCA accident and transient analyses that are part of the CCNPP licensing basis.
. The specific acceptance criteria applicable to CCNPP are identified in Chapter 14 of the CCNPP Updated Final Safety Analysis Report (UFSAR).
The inspectors reviewed Topical Report EMF-2310(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, May 2003, and noted that it described four correlations that could be used and that the licensee selected the modified Barnett Correlation for analyzing fuel performance following a main steam line break. The inspectors determined the methodology involved using the correlation to calculate how many, if any, fuel rods exceed the approved 95/95 departure from nucleate boiling ratio Barnett Correlation value and would therefore be assumed to fail. The inspectors noted that the 95/95 departure from nucleate boiling ratio Barnett Correlation value of 1.135 was used as stated in the following sections of the Calvert Cliff Updated Final Safety Analysis Report.
14.1.4.4 Fuel Performance Models and Acceptance Criteria, B. Postulated Accidents, 1. Site Boundary Dose
The pre-trip SLB event assumes that all fuel rods with a DNBR less than the NRC-approved safety limit experience DNB. The safety limit DNBR gives a 95% probability, at a 95%
confidence level, that the hottest fuel rod will not experience DNB. For dose calculations, all rods experiencing DNB are assumed to fail. For the return to power SLB [steam line break]
analysis, DNB is assumed when the rod experiences a DNBR less than 1.135 calculated using the limit associated with the modified Barnett CHF correlation.
14.14.4 CORE AND SYSTEM PERFORMANCE - 14.14.4.3 Results
The post-trip SLB results show that the minimum DNBR is above the modified Barnett SAFDL (Specified Acceptable Fuel Design Limits) of 1.158 including a 2% mixed core penalty and the peak LHGR remains below the SAFDL of the more limiting of 21 kW/ft (per Reference 15)or the cycle specific limit.
The inspectors noted, consistent with Nuclear Energy Institute NEI 96-07, Guidelines for 10 CFR 50.59 Evaluation, Revision 1, Section 4.3.7, which was endorsed for use by NRC Regulatory Guide 1.187, that the 95/95 departure from nucleate boiling ratio value of 1.135 (and 1.158 for mixed cores) is a design basis limit for a fission product barrier (fuel cladding).
Specifically, the inspectors determined the limit is fundamental to the barriers integrity, the limit is expressed numerically, and the limit is identified in the updated final safety analysis report.
The inspectors reviewed SE569 to determine the basis for the licensees conclusion that a license amendment from the NRC was not required. In particular, the inspectors focused on the adequacy of the licensees determination that the change met 10 CFR 50.59(c)(2)(vii),specifically, that it did not result in a design basis limit for a fission product barrier as described in the updated final safety analysis report being exceeded or altered. In addition, the inspectors focused on the adequacy of the licensees determination that the change met 10 CFR 50.59(c)(2)(viii), specifically, that it did not result in a departure from a method of evaluation described in the updated final safety analysis report used in establishing the design bases or in the safety analyses.
The inspectors determined that the licensee identified the change altered a design basis limit for a fission product, but concluded, based on NEI 96-07, Section 4.3.7, noted below, that this change should be evaluated under Criterion (viii) and that Criterion (vii) was not applicable.
NEI 96-07:
Altering a design basis limit for a fission product barrier is not a routine activity, but it can occur. An example of this would be changing the DNBR value from the value corresponding to the 95/95 criterion for a given DNB correlation, perhaps as a result of a new fuel design being implemented. (A new correlation or a new value for the "95/95 DNB criterion" with the same fuel type would be evaluated under criterion (c)(2)(viii) of the rule.). These are infrequent activities affecting key elements of the defense-in-depth philosophy. As such, no distinction has been made between a conservative and non-conservative change in these limits.
The inspectors concluded that under Criterion (vii) the design basis limit 95/95 departure from nucleate boiling ratio Barnett Correlation value for the fuel would be considered altered and would require a license amendment. The inspectors noted that the NEI 96-07 indicated the evaluation of this type of change to the design basis limit may be evaluated using the requirements of Criterion (viii) in order to determine if a license amendment is required; but disagreed with the licensees conclusion that the criteria should be marked not applicable.
The inspectors concluded that the guidance in Criterion (vii) describes the use of Criterion (viii) requirements and, therefore, is part of the evaluation. The inspectors review of the licensees response to Criterion (viii) found that the licensee concluded their new departure from nucleate boiling ratio correlation value, represented by an equation rather than a specific numerical value, was developed using an NRC-approved methodology, and therefore responded No to Criterion (viii), Does the change result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses? Specifically, the licensee documented their conclusion that the NRC previously approved the methodology for determining the 95/95 departure from nucleate boiling ratio Barnett Correlation value as described in an NRC safety evaluation report accompanying License Amendment 71, for the HB Robinson Steam Electric Plant, dated July 23, 1982, ML020520304. The licensee concluded that they could make the change without NRC approval based on NEI 96-07, Section 4.3.8.
Use of a new NRC-approved methodology (e.g., new or upgraded computer code) to reduce uncertainty, provide more precise results, or other reason, provided such use is
- (a) based on sound engineering practice,
- (b) appropriate for the intended application, and
- (c) within the limitations of the applicable SER.
The inspectors consulted with NRC staff from the Office of Nuclear Reactor Regulations Nuclear Methods and Fuels Branch and determined that the HB Robinson amendment and associated safety evaluation report did not approve a methodology for determining the 95/95 departure from nucleate boiling ratio Barnett Correlation limit. The HB Robinson license amendment approved the 95/95 departure from nucleate boiling ratio Barnett Correlation limit value of 1.135. The accompanying safety evaluation report described the basis for the NRC staffs acceptance of the value of 1.135 but did not describe approval of a methodology to calculate this design basis limit. The inspectors reviewed multiple recent and older approvals of 95/95 departure from nucleate boiling ratio limits which had similar wording as that in the HB Robinson safety evaluation report. In each case, the NRC staff provided similar justification as to the reason why the value was approved. In no case was there a provision included that allowed a licensee to independently establish the 95/95 departure from nucleate boiling ratio correlation value (design basis limit). Rather, only the 95/95 correlation value has been approved following the NRCs review of test data submitted by the fuel vendor where correlations are proposed by the vendor. The inspectors concluded the licensee used a methodology not previously approved by the NRC in this instance.
The inspectors also noted in the licensees SE569 response to Criterion (viii) that the new 95/95 departure from nucleate boiling ratio Barnett Correlation limit value was an element of a methodology approved by the NRC in Topical Report EMF-2310(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, May 2003, ADAMS Accession No. ML032460863.
The inspectors reviewed EMF-2310 and noted that it describes a methodology used to determine the core response for certain design basis events and is listed in the Calvert Cliffs Units 1 and 2, Technical Specifications as the method used in the core operating limits report. The inspectors did not find that the selection of the 95/95 departure from nucleate boiling ratio design basis value is an element of the EMF-2310 methodology. NEI 96-07 states, that inputs to the methodology are considered elements of the methodology if the method of evaluation includes a methodology describing how to select the value of an input parameter to yield adequately conservative results. The inspectors noted that EMF-2310 does not describe how to select the value for the design basis limit, the limit was approved by the NRC. Additionally, the design basis limit has no impact on whether the results from the methodology are conservative. The results of the methodology are compared to the safety limit to ensure that departure from nucleate boiling does not occur. Specifically, Topical Report EMF-2310, Section 5.4.1 states:
The potential for fuel failure from DNB is assessed by comparing the calculated MDNBR to the applicable departure from nucleate boiling ratio (DNBR) safety limit.
The inspectors noted that the licensee documented in their SE569 response to Criterion (viii)that this change was determined to result in a conservative change to the methodology because the difference between the new 95/95 departure from nucleate boiling ratio Barnett Correlation limit value and fuel channel characteristics for departure from nucleate boiling ratio determined by using EMF-2310 was reduced when compared to the previously NRC approved value of 1.135. Therefore, based on NEI 96-07, Section 4.3.8.1, noted below, the change did not require prior NRC approval.
Conservative vs. Non-Conservative Results Gaining margin by changing one or more elements of a method of evaluation is considered to be a non-conservative change and thus a departure from a method of evaluation for purposes of 10 CFR 50.59. Such departures require prior NRC approval of the revised method.
Analytical results obtained by changing any element of a method are "conservative" relative to the previous results, if they are closer to design bases limits or safety analyses limits (e.g., applicable acceptance guidelines).
The inspectors determined that the NEI guidance in regard to conservative or non-conservative changes is not applicable to changes in design basis limits. The NEI guidance specifically states that the evaluation of conservative vs. non-conservative is performed by comparing the new value, as a result of the change, to the design bases limits or safety analyses limits. Additionally, the inspectors noted that NEI 96-07, noted below, related to changing design basis limits specifically states that changes to design basis limits are neither conservative or non-conservative. This guidance reflects the NRC Statement of Considerations issued with the 10 CFR 50.59 Rule, Federal Registration, Volume 64, Number 191, October 4, 1999, page 53597.
These are infrequent activities affecting key elements of the defense-in-depth philosophy. As such, no distinction has been made between a conservative and non-conservative change in these limits.
The inspectors concluded that the correct response to Criterion (viii) for this change was yes, and that a license amendment from the NRC was therefore required prior to implementing this change.
Corrective Actions: The licensee entered this condition into their corrective action program.
Corrective Action References: AR04351464
Performance Assessment:
The inspectors determined this violation was associated with a minor performance deficiency.
Enforcement:
The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.
Severity: In accordance with NRC Enforcement Manual, Section 2.1.3, Enforcement of 10 CFR 50.59 and Related FSAR Violations, Section D.5.a, the inspectors determined this violation is categorized as Severity Level IV because the change required prior Commission review and approval, and the licensee failed to obtain Commission approval. The consequence of the activity or change evaluated by the significance determination process is of very low safety significance (i.e., Green), and the NRC most likely would have approved the change.
Violation: 10 CFR 50.59(c)(1) states, in part, that a licensee may make changes in the facility or procedures as described in the updated final safety analysis report without obtaining a license amendment pursuant to Section 50.90 only if the change does not meet any of the criteria in paragraph (c)(2) of this section. 10 CFR 50.59(c)(2)(viii) states, in part, that a licensee shall obtain a license amendment pursuant to Section. 50.90 prior to implementing a proposed change if the change would result in a departure from a method of evaluation described in the updated final safety analysis report used in establishing the design bases or in the safety analyses.
Contrary to the above, as of February 6, 2020, the licensee made a change in the facility or procedures as described in the updated final safety analysis report without obtaining a license amendment pursuant to Section. 50.90, and that change resulted in a departure from a method of evaluation described in the updated final safety analysis report used in establishing the design bases or in the safety analyses. Specifically, following the identification that the value described in the updated final safety analysis report, for the 95/95 departure from nucleate boiling ratio criterion for the design basis limit for fuel was incorrect, the licensee incorporated a new value but did not submit a license amendment request to obtain approval for the new value.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On October 20, 2020, the inspectors presented the integrated inspection results to Mr. Thomas Haaf and other members of the licensee staff.
- On October 20, 2020, the inspectors presented the evaluations of changes, tests, and experiments inspection results to Mr. Thomas Haaf and other members of the licensee staff.
- On August 20, 2020, the inspectors presented the Emergency Preparedness Program Inspection inspection results to Mr. Thomas Haaf and other members of the licensee staff.
- On August 6, 2020, the inspectors presented the Triennial Heat Sink Inspection inspection results to Mr. Thomas Haaf and other members of the licensee staff.
- On June 19, 2020, the inspectors presented the evaluations of changes, tests, and experiments inspection results to Mr. Thomas Haaf and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
71111.07T Engineering
Evaluations
IR 04163543 0C DG Control Room Air Handler Unit
Condenser (a)(1) determination
10/02/2018
71111.07T Miscellaneous
System Health Report Salt Water Service Water
System Health Report Emergency Diesel Generators
System Health Report Safety Related Ventilation
System Health Report Component Cooling Water
08/01/2020
C93299055-080
Component Cooler Eddy Current Inspection Report
11/14/2018
71111.17T Corrective Action
Documents
6/19/20
71111.17T Miscellaneous
EMF-2310 (P)(A)
SPP Chapter 15 Non-LOCA Methodology for Pressurized
Water Reactors
71111.17T Miscellaneous
- H.B. Robinson Steam Electric Plant, Unit No 2 License
Amendment to Facility Operating License (Amendment No.
71),
July 23,
1982
71111.17T Miscellaneous
NRC Safety
Evaluation
Final Safety Evaluation for Framatone AMP Appendix A to
Topical Report EMF-92-153 (NP)(A), HTTP: Departure from
Nucleate Boiling Correlation for High Thermal Performance
Fuel
January 6,
2006
71111.17T Miscellaneous
License Amendment for Calvert Cliffs Nuclear Power Plant,
Unit Nos 1 and 2 - Amendment Re: Transition from
Westinghouse Nuclear Fuel to Areva Nuclear
February 18,
2011
71111.17T Miscellaneous
SE 069 50.59
Safety Evaluation
Revising the Modified Barnett Limit used in Steam Line
Break Methodology
71111.17T Miscellaneous
SE069 50.59
Screen
Revise UFSAR 14.1, 14.4 and 14.14 to Change Modified
Barnett DNBR Limit
71111.17T Self-Assessments
Self-Assessment, 10 CFR 50.59 Review
7/20/17
Corrective Action
Documents
04347195
Miscellaneous
Design Report
Calvert Cliffs Nuclear Power Plant Alert and Notification
System Design Report
Revision 2
Corrective Action
Documents
04176383
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
Exelon Nuclear Standardized Radiological Emergency Plan
Revision 33
Miscellaneous
Exelon Nuclear Radiological Emergency Plan Annex for
Calvert Cliffs Station
Revision 5
Miscellaneous
Addendum 1
Calvert Cliffs Nuclear Power Plant On-Shift Staffing
Technical Basis
Revision 2
Procedures
CFR 50.54(q) Change Evaluation
Revision 10
Corrective Action
Documents
04162575
Miscellaneous
EP-AA-121-F-15
Calvert Cliffs Equipment Matrix
Revision 2