IR 05000220/2019003
| ML19312A053 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/06/2019 |
| From: | Erin Carfang Reactor Projects Branch 1 |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| Carfang E | |
| References | |
| IR 2019003, ISFSI IR 2019001 | |
| Download: ML19312A053 (23) | |
Text
November 6, 2019
SUBJECT:
NINE MILE POINT NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000220/2019003 AND 05000410/2019003 AND INDEPENDENT SPENT FUEL STORAGE INSTALLATION INSPECTION REPORT 07201036/2019001
Dear Mr. Hanson:
On September 30, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Nine Mile Point Nuclear Station. On October 23, 2019, the NRC inspectors discussed the results of this inspection with Mr. Peter Orphanos, Site Vice President, 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. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violation or significance or severity of the violation 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 Nine Mile Point.
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 Nine Mile Point. 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 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Erin E. Carfang, Chief Reactor Projects Branch 1 Division of Reactor Projects
Docket Nos. 05000220, 05000410,
and 07201036 License Nos. DPR-63 and NPF-69
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000220, 05000410, and 07201036
License Numbers:
Report Numbers:
05000220/2019003, 05000410/2019003, and 07201036/2019001
Enterprise Identifier: I-2019-003-0036
I-2019-001-0074
Licensee:
Exelon Generation Company, LLC
Facility:
Nine Mile Point Nuclear Station
Location:
Oswego, NY
Inspection Dates:
July 1, 2019 to September 30, 2019
Inspectors:
G. Stock, Senior Resident Inspector
J. Dolecki, Resident Inspector
B. Sienel, Resident Inspector
E. Burket, Senior Reactor Inspector
O. Masnyk Bailey, Health Physicist
J. Nicholson, Senior Health Physicist
R. Rolph, Health Physicist
J. Steward, Acting Senior Resident Inspector
Approved By:
Erin E. Carfang, Chief
Reactor Projects Branch 1
Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Nine Mile Point Nuclear Station 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 Obtain Station ALARA Committee Approval for a Dose Estimate Change Which Resulted in Collective Dose Greater Than 5 Person-Rem and Exceeding the Estimate by More Than 50 Percent Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000220/2019003-01 Open/Closed
[H.4] -
Teamwork 71124.02 The inspectors identified a finding and Green NCV of Technical Specification 6.4.1,
"Administrative Controls, Procedures," for Exelon's failure to follow its procedure to maintain doses as low as reasonably achievable (ALARA) during the most recent Unit 1 refueling outage. Specifically, Exelon did not follow procedure RP-AA-401, "ALARA Planning and Controls," and did not obtain Station ALARA Committee review when raising the dose estimate for drywell main steam isolation valve activities. This resulted in the dose estimate changing from a Category 1 to a Category 2 work activity which required approval by the Station ALARA Committee in accordance with Exelon procedure RR-AA-401, ALARA Program, step 4.2.7(3). The site had additional exposure estimate increases which resulted in the accumulated dose exceeding 5 person-rem and exceeding the original estimate by greater than 50 percent.
Additional Tracking Items
None.
PLANT STATUS
Unit 1 began the inspection period at 100 percent power. On August 4, 2019, Unit 1 experienced a trip of the reactor recirculation motor-generator 11 resulting in power decreasing to approximately 87 percent. Later that day, operators raised power using the four remaining reactor recirculation motor-generators and reached 100 percent. On August 6, 2019, following repairs to the reactor recirculation motor-generator 11, operators performed a planned power reduction to approximately 85 percent to place the component in service. Later that day, Unit 1 was restored to 100 percent power. Unit 1 remained at or near 100 percent power for the remainder of the inspection period.
Unit 2 began the inspection period at 100 percent power. On September 20, 2019, operators performed a planned power reduction to approximately 70 percent to conduct quarterly turbine valve reactor protection system testing, control rod sequence exchange, and scram time testing. On September 22, 2019, Unit 2 was restored to 100 percent power. Unit 2 remained at or near rated thermal power for the remainder of the 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 performed plant status activities described in IMC 2515 Appendix D, Plant Status, and conducted routine reviews using IP 71152, Problem Identification and Resolution. 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.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Summer Readiness Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated summer readiness of alternating current offsite and alternate power systems on August 14, 2019.
External Flooding Sample (IP Section 03.04) (1 Sample)
- (1) The inspectors evaluated readiness to cope with external flooding on September 30, 2019.
71111.04Q - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 1 emergency cooling system 11 on July 12, 2019
- (2) Unit 2 Division III diesel generator on August 22, 2019
- (3) Unit 2 high pressure core spray system on September 4, 2019
- (4) Unit 1 emergency service water system 11 on September 12, 2019
- (5) Unit 1 emergency diesel generator 103 on September 24, 2019
71111.05Q - Fire Protection
Quarterly Inspection (IP Section 03.01) (8 Samples)
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Unit 1 reactor building 237', emergency escape hatch room, fire area 1, on June 14, 2019
- (2) Unit 1 reactor building 281', spent fuel pool cooling pump room, fire area 1, on June 14, 2019
- (3) Unit 1 reactor building 237', north instrument room, fire area 1, on June 14, 2019
- (4) Unit 1 reactor building 340', refuel floor, fire area 1, on June 14, 2019
- (5) Unit 2 reactor building 215', 'A' reactor water cleanup pump room, fire area 34, on June 14, 2019
- (6) Unit 2 reactor building 215', 'B' reactor water cleanup pump room, fire area 35, on June 14, 2019
- (7) Unit 1 261' emergency diesel generator 103 room, fire area 19, on September 24, 2019
- (8) Unit 1 261' power board 103 room, fire area 24, on September 24, 2019
71111.07T - Heat Sink Performance Triennial Review (IP Section 02.02)
The inspectors evaluated heat exchanger/sink performance on the following:
- (1) Unit 1 reactor building closed loop 11 cooling heat exchanger
- (2) Unit 1 ultimate heat sink
- (3) Unit 2 'A' residual heat removal heat exchanger
- (4) Unit 2 'B' low pressure core spray room cooler
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 1 operator performance during response to an oil leak on the feedwater pump 13 clutch actuator package on July 18, 2019.
- (2) The inspectors observed and evaluated a Unit 2 downpower to 70 percent for a rod pattern adjustment and scram time testing on September 20, 2019.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)
- (1) The inspectors observed and evaluated a Unit 1 simulator evaluation which included the inadvertent opening and failure to close of an electromatic relief valve, a high-power failure to scram, and a steam leak into primary containment on August 20, 2019.
- (2) The inspectors observed and evaluated a Unit 2 simulator evaluation which included reactor power oscillations, a failure to scram, and an un-isolable steam line break in the turbine building on September 5, 2019.
71111.12 - Maintenance Effectiveness
Routine Maintenance Effectiveness Inspection (IP Section 02.01) (2 Samples)
The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:
- (1) Maintenance Rule Periodic Assessment - October 1, 2015 to September 30, 2017; completed on July 12, 2019
- (2) Unit 2 instrument air system on August 2, 2019
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (6 Samples)
The inspectors evaluated the risk assessments for the following planned and emergent work activities:
- (1) Unit 1 emergent work associated with feedwater pump 13 actuator package oil leak on July 18, 2019
- (2) Unit 1 emergent work associated with reactor recirculation pump 11 trip on August 6, 2019
- (3) Unit 2 elevated risk condition during a severe thunderstorm warning followed by planned maintenance on the Division III diesel generator on August 21, 2019
- (4) Unit 2 elevated risk condition during planned Division III diesel fuel injector replacements on September 4, 2019
- (5) Unit 1 elevated risk during a 345-kilovolt Line 8 outage to support a grid repair on September 20, 2019
- (6) Unit 1 elevated risk during planned emergency diesel generator 102 maintenance on September 25, 2019
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 02.02) (6 Samples)
The inspectors evaluated the following operability determinations and functionality assessments:
- (1) Unit 2 service water system and supported systems following chemical evacuation treatment on July 16, 2019
- (2) Unit 1 intermediate range monitors following trend in downscale indications on July 16, 2019
- (3) Unit 1 high pressure coolant injection system during a 115-kilovolt grid disturbance on July 17, 2019
- (4) Unit 1 containment spray raw water intertie valve 93-72 failure to fully open on July 25, 2019
- (5) Unit 2 residual heat removal system during replacement of relief valve 2RHS*RV57A on August 27, 2019
- (6) Unit 1 spent fuel pool Boraflex rack without one cell blocker installed on September 23, 2019
71111.19 - Post-Maintenance Testing
Post-Maintenance Test Sample (IP Section 03.01) (7 Samples)
The inspectors evaluated the following post-maintenance tests:
- (1) Unit 1 containment spray raw water pump strainer 93-06 maintenance on August 6, 2019
- (2) Unit 2 Division III diesel generator crankcase pressure switch replacement on August 21, 2019
- (3) Unit 1 conductivity monitor replacement on September 3, 2019
- (4) Unit 2 Division III diesel generator fuel injector replacements on September 6, 2019
- (5) Unit 1 emergency service water pump 12 replacement on September 12, 2019
- (6) Unit 1 emergency condenser vent to torus block valve preventive maintenance on September 17, 2019
- (7) Unit 1 emergency diesel generator 102 four-year preventive maintenance on September 27, 2019
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance tests:
Inservice Testing (IP Section 03.01)
- (1) N2-OSP-RHS-Q@005, RHR System Loop B Pump and Valve Operability Test, System Integrity Test and ASME XI Pressure Test, on July 3, 2019
- (2) N1-ST-Q6C, Containment Spray System Loop 112 Quarterly Operability Test, on July 31, 2019
- (3) N1-ST-Q6B, Containment Spray System Loop 121 Quarterly Operability Test, on August 7, 2019
FLEX Testing (IP Section 03.02) (1 Sample)
- (1) S-PM-001, Flex 3419MX Water Pump Test, on August 14, 2019
71114.06 - Drill Evaluation
Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated the conduct of a routine Exelon emergency preparedness drill on September 5, 2019.
Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)
The inspectors evaluated:
- (1) A Unit 1 simulator training evaluation which included the inadvertent opening and failure to close of an electromatic relief valve, a high-power failure to scram, a steam leak outside primary containment, and ultimately the declaration of a General Emergency on August 20,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Contamination and Radioactive Material Control (IP Section 02.03) (1 Sample)
The inspectors evaluated:
- (1) Licensee processes for monitoring and controlling contamination and radioactive material by observing equipment and material leaving the radiologically controlled area, radiation protection technicians surveying equipment, and reviewing release procedures.
71124.02 - Occupational ALARA Planning and Controls
Radiological Work Planning (IP Section 02.01) (1 Sample)
The inspectors evaluated the licensees radiological work planning.
- (1) The inspectors reviewed the following activities:
RWP NM-1-19-00506 Drywell Scaffold and ALARA Plan
RWP NM-1-19-00509 Drywell Main Steam Isolation Valve Activities and ALARA Plan
RWP NM-1-19-00518 Drywell In Service Inspection Activities and ALARA plan
RWP NM-1-19-00601 Reactor Building Reactor Water Clean Up System Maintenance Activities and ALARA Plan
RWP NM-1-19-00901 Reactor Disassembly/Reassembly and ALARA Plan
Implementation of ALARA and Radiological Work Controls (IP Section 02.03) (1 Sample)
The inspectors reviewed as low as reasonably achievable practices and radiological work controls.
- (1) The inspectors reviewed the following activities:
RWP NM-1-19-00506 Drywell Scaffold and ALARA Plan Work in Progress Reviews and Post Job Review
RWP NM-1-19-00509 Drywell Main Steam Isolation Valve Activities and ALARA Plan Work in Progress Reviews and Post Job Review
RWP NM-1-19-00518 Drywell In Service Inspection Activities and ALARA plan Work in Progress Reviews and Post Job Review
RWP NM-1-19-00601 Reactor Building Reactor Water Clean Up System Maintenance Activities and ALARA Plan Work in Progress Reviews and Post Job Review
RWP NM-1-19-00901 Reactor Disassembly/Reassembly and ALARA Plan Work in Progress Reviews and Post Job Review
71124.04 - Occupational Dose Assessment
External Dosimetry (IP Section 02.02) (1 Sample)
- (1) The inspectors evaluated the external dosimetry program implementation.
Special Dosimetric Situations (IP Section 02.04) (1 Sample)
The inspectors evaluated the following special dosimetric situation:
- (1) The licensees implementation of requirements to manage radiation protection of declared pregnant workers for four workers, application of NRC-approved external dosimetry methods, and neutron dose assessments.
71124.07 - Radiological Environmental Monitoring Program
Site Inspection (IP Section 02.01) (1 Sample)
The inspectors evaluated the radiological environmental monitoring program implementation.
- (1) The inspectors reviewed the following:
Walkdowns, Calibrations, and Maintenance Record Review
The inspectors observed air sample collection at R1, R2, R3, R4, R5 and observed the thermoluminescent dosimeters (TLDs) at the same locations.
Environmental Sample Collections and Preparation Observation
The inspectors observed the collection of two milk samples and the preparation for shipment.
Licensee Actions in Response to Missed Sample, Inoperable Sampler, Lost TLD or Anomalous Measurement
The inspectors reviewed issue reports for power losses at a couple air monitor stations and a couple of lost environmental TLDs.
Sampling Program for the Potential of Licensed Material Entering Groundwater
The inspectors observed the condition of several monitoring wells to ensure maintenance is being maintained.
Groundwater Protection Initiative (GPI) Implementation (IP Section 02.02) (1 Sample)
- (1) The inspectors reviewed the Groundwater Protection Program. All tritium (H-3) sample results are less than 500 pCi/l.
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04)===
- (1) Unit 1, July 1, 2018 - June 30, 2019
- (2) Unit 2, July 1, 2018 - June 30, 2019
MS06: Emergency AC Power Systems (IP Section 02.05) (2 Samples)
- (1) Unit 1, July 1, 2018 - June 30, 2019
- (2) Unit 2, July 1, 2018 - June 30, 2019
MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)
- (1) Unit 1, July 1, 2018 - June 30, 2019
- (2) Unit 2, July 1, 2018 - June 30, 2019
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (2 Samples)
- (1) Unit 1, September 2018-August 2019
- (2) Unit 2, September 2018-August 2019
PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample.
(IP Section 02.16) (2 Samples)
- (1) Unit 1, January 1, 2018-June 30, 2019
- (2) Unit 2, January 1, 2018-June 30, 2019
71152 - Problem Identification and Resolution
Annual Follow-up of Selected Issues (IP Section 02.03) (1 Sample)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) IR 04234861, Evaluation of Local Leak Rate Type B and C Test Failures Associated with Isolation Valves During the 2019 Unit 1 Refueling Outage
71153 - Followup of Events and Notices of Enforcement Discretion Event Followup (IP Section 03.01)
- (1) Unit 1 recirculation pump 11 trip due to degraded tachometer generator on August 4,
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
60855.1 - Operation of an Independent Spent Fuel Storage Installation at Operating Plants Operation of an Independent Spent Fuel Storage Installation at Operating Plants
- (1) The inspectors evaluated Nine Mile Point's independent spent fuel storage installation cask loadings on July 8-11, 2019. Specifically, the inspectors observed the following activities:
Fuel selection
Heavy load movement of the transfer cask and dry shielded canister
Decontamination of the transfer cask
Closure welding and non-destructive weld evaluations
Canister drying and helium backfill evolutions
Helium leak testing
Radiological field surveys
INSPECTION RESULTS
Failure to Obtain Station ALARA Committee Approval for a Dose Estimate Change Which Resulted in Collective Dose Greater Than 5 Person-Rem and Exceeding the Estimate by More Than 50 Percent Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety
Green NCV 05000220/2019003-01 Open/Closed
[H.4] -
Teamwork 71124.02 The inspectors identified a finding and Green NCV of Technical Specification 6.4.1, "Administrative Controls, Procedures," for Exelon's failure to follow its procedure to maintain doses As Low As Reasonably Achievable (ALARA) during the most recent Unit 1 refueling outage. Specifically, Exelon did not follow procedure RP-AA-401, "ALARA Planning and Controls," and did not obtain Station ALARA Committee review when raising the dose estimate for drywell main steam isolation valve activities. This resulted in the dose estimate changing from a Category 1 to a Category 2 work activity, which required approval by the Station ALARA Committee in accordance with Exelon procedure RR-AA-401, ALARA Program, step 4.2.7(3). The site had additional exposure estimate increases, which resulted in the accumulated dose exceeding 5 person-Rem and exceeding the original estimate by greater than 50 percent.
Description:
Prior to the most recent refueling outage, Nine Mile Point Unit 1 had been experiencing a rising trend in unidentified leakage. It was determined during the outage that the cause was steam leakage from the main steam isolation valve IV-01-01 packing. A radiation work permit and ALARA plan were written to repair the valve. A work-in-progress (WIP) review was performed at 80 percent of the dose estimate and identified that an increase in the estimate was required due to the increase in work scope. The valve internals required repair by adding a weld overlay, and machining had to be performed to reassemble the valve. Subsequently, the dose estimate was revised from 2.900 person-rem to 3.526 person-rem. This exceeded the 3 person-rem level and raised the task from a Category 1 to a Category 2 per procedure RR-AA-401, step 4.1.3(6). Step 4.2.7(3) of the procedure states, "If a WIP Review results in a change in dose estimate from ALARA Category 1 to Category 2, then SAC [Station ALARA Committee] approval is required." This approval was not obtained when the dose estimate was revised to 3.526 person-rem, and as a result the total actual dose received from the activity was 7.651 person-rem which was greater than 5 person-rem and greater than 50 percent of the dose estimate.
Corrective Actions: Exelons immediate corrective actions included reviewing the lessons learned in the post-job review following the work and entering the issue into its corrective action program.
Corrective Action References: IR 04275201
Performance Assessment:
Performance Deficiency: The inspectors determined that Exelon failed to implement their ALARA procedure to obtain Station ALARA Committee approval when changing the dose estimate for drywell main steam isolation valve activities as required by RR-AA-401, step 4.2.7(3). This was a performance deficiency that was reasonably within Exelons ability to foresee and correct and should have been prevented.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program & Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors determined the performance deficiency was more than minor because it was associated with the Program & Process attribute of the Occupational Radiation Safety cornerstone.
Specifically, it adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Additionally, this finding is similar to IMC 0612, Appendix E, Example 6.i. Specifically, the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent.
Significance: The inspectors assessed the significance of the finding using Appendix C, Occupational Radiation Safety SDP. The inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency did not result in Exelons current 3-year rolling average collective dose exceeding 240 person-rem.
Cross-Cutting Aspect: H.4 - Teamwork: Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. The work process includes the identification and management of risk commensurate with the work and the need for coordination with different groups or job activities. Exelon's ALARA WIP review process and increase in the dose estimate was not reviewed and approved by Station ALARA Committee and caused the accumulated dose to be greater than 50 percent of the revised estimate for the task.
Enforcement:
Violation: Technical Specification 6.4.1, requires, in part, that written procedures be established, implemented, and maintained as required by NRC Regulatory Guide 1.33, Appendix A, November 3, 1972,Section I.5(b), "Factors to be taken into account in preparing the detailed work procedures, including the necessity for minimizing radiation exposure to workmen." Exelon procedure RR-AA-406, "ALARA Planning and Controls," requires Station ALARA Committee approval for work activities with dose estimates greater than or equal to 3 person-rem. Contrary to the above, Exelon failed to follow the procedural requirement to obtain Station ALARA Committee approval for a dose estimate revision greater than 3 person-rem.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Observation: Unit 1 Local Leak Rate Test Failures 71152 A series of primary containment valves and reactor coolant system pressure isolation valves did not meet Exelons local leak rate testing acceptance criteria during the Unit 1 refueling outage in the spring of 2019. Exelon generated issue reports to identify and resolve the issues. IR 04234861, as well as IR 04237553, captured several of the local leak rate test failures.
The inspectors reviewed the issue reports, the results of the local leak rate tests, and associated corrective actions to verify the site was in compliance, adequately identified issues, and initiated corrective actions. Depending on the severity of the leak rate, Exelon took corrective actions to make changes to preventative maintenance frequency, performed engineering evaluations to remove conservatism in the permitted leak late, or repaired the valve.
The inboard steam supply emergency condenser isolation valves, IV 39-09R and IV 39-10R, failed the local leak rate test, as detailed in IRs 04230591 and 04232618, respectively.
Exelon subsequently repaired the condition, which included having to split the valve stem nut, perform post-maintenance testing, and return the valves to operable status. Inspectors challenged the duration of the IV 39-09R and IV 39-10R inoperabilty after identifying that a past operability evaluation was not in the respective IRs and, therefore, a review of reportability was not performed. Exelon performed a past operability evaluation and determined the valves did not have an extended period of inoperability because IV 39-09R and 39-10R passed the most recent surveillance tests on March 14, 2019, and January 30, 2019, respectively. The inspectors determined there was reasonable assurance that these valves were capable of fulfilling their intended functions while the plant was operating.
The inspectors independently evaluated the issues of concern in accordance with the guidance in IMC 0612, Appendix B, "Issue Screening," and Appendix E, "Examples of Minor Violations." As a result, no violations or performance deficiencies were identified.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On October 23, 2019, the inspectors presented the integrated inspection results to Mr. Peter Orphanos, Site Vice President, and other members of the licensee staff.
On July 11, 2019, the inspectors presented the Independent Spent Fuel Storage Installation inspection results to Mr. Todd Tierney, Plant Manager, and other members of the licensee staff.
On August 8, 2019, the inspectors presented the Triennial Heat Sink Performance inspection results to Mr. Peter Orphanos, Site Vice President, and other members of the licensee staff.
On September 19, 2019, the inspectors presented the Radiological Environmental Monitoring Program inspection results to Mr. Matthew Bush, Maintenance Director, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
04056415
04056415
264985
Procedures
N1-IPM-110-001
Sample Conductivity Calibration
00500
N1-SOP-33A.3
Major 115 KV Grid Disturbances
201
N2-MSP-GEN-
V001
Revetment Ditch Structure Inspection
00501
N2-OP-70
Station Electric Feed and 115kv Switchyard
2400
N2-SOP-70
Major Grid Disturbances
00400
OP-NM-108-107-
1002
Off-Site Power Operations and Interface
004
WC-NM-8003-
1012
Nine Mile Point Nuclear Station Units 1 and 2 Nuclear Plant
Interface Requirements (NPIRs)
Work Orders
C92770203
71111.04Q Corrective Action
Documents
04107700
236901
258410
258431
Drawings
C-18017-C,
Sheet 1
Emergency Cooling System P&I Diagram
56.01
C-18022-C,
Sheet 1
Service Water, Reactor and Turbine Buildings P&I Diagram
Procedures
N1-OP-13
Emergency Cooling System
04300
N1-OP-18
Service Water System
04400
N1-OP-45
04800
N2-OP-100B
HPCS Diesel Generator
2700
N2-OP-100B-
LINEUPS
HPCS Diesel Generator - LINEUPS
200
N2-OP-33
High Pressure Core Spray System
01700
N2-OP-33-
HPCS System - LINEUPS
00100
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
LINEUPS
71111.05Q Procedures
N1-PFP-0101
Unit 1 Pre-Fire Plans
00500
Corrective Action
Documents
4182503
Miscellaneous
673606
Nine Mile Point Unit 2 Forebay Silt Mapping Report
2/28/2018
N2-TTP-RHS-
4Y003
Residual Heat Removal System Heat Exchanger
Performance Monitoring Test
Performed
2/12/2016
Procedures
GL 89-13 Program Implementation Instructional Guide
71111.11Q Procedures
N2-OSP-RMC-
@001
Control Rod Drive Scram Insertion Time Testing
2500
N2-SOP-30
Control Rod Drive Failures
00500
NF-AB-720-F-1
Control Rod Sequence Review and Approval Sheet
Corrective Action
Documents
04195955
04196529
04196538
200048
214686
223198
226384
Miscellaneous
CFR 50.65
(a)(3)
Maintenance Rule Periodic Assessment
October 1,
2015 to
September 30,
2017
Work Orders
C93536085
C93536250
C93687564
C93695810
C93696635
C93697980
Corrective Action
Documents
265049
265299
265764
269525
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Drawings
B-18023-C,
Sheet 1
Shaft Driven Reactor Feedwater Pump #13 Gear and
Clutch Oil P&I Diagram
Procedures
N1-ARP-F2
Control Room Panel F2
200
N1-OP-1
Nuclear Steam Supply System
075T1
N1-OP-16
Feedwater System Booster Pump to Reactor
06700
N1-OP-43B
Normal Power Operations
2500
N1-SOP-1.3
Recirc Pump Trip at Power
00400
N2-OP-102
Meteorological Monitoring
2600
OP-NM-108-117
Protected Equipment Program at Nine Mile Point
00500
On-Line Work Control Process
Work Orders
C93710586
C93716460
Corrective Action
Documents
255747
262272
262316
263602
263911
264027
264036
264279
264283
264399
264985
265111
266278
271674
274661
278008
Drawings
PID-31A
Piping and Instrumentation Diagram Residual Heat
Removal System
Procedures
N1-OP-16
Feedwater System Booster Pump to Reactor
06700
N1-OP-33A
115 KV System
03300
N1-OP-38B
2100
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
N1-SOP-33A.3
Major 115 KV Grid Disturbances
201
N2-OP-11
Service Water System
01400
N2-PM-@11
SWP Mullusk Biocide Treatment (System Alignment and
Valve Control)
00700
Work Orders
C93576635
C93676790
Drawings
C-18012-C,
Sheet 1
Reactor Containment Spray Raw Water System P&I
Diagram
C-18012-C,
Sheet 2
Reactor Containment Spray Raw Water System P&I
Diagram
Procedures
N1-ST-M4A
Emergency Diesel Generator 102 and PB [power board]
2 Operability Test
2500
N1-ST-Q13
Emergency Service Water Pump and Check Valve
Operability Test
2100
N1-ST-Q5
Primary Containment Isolation Valves Operability Test
03400
N2-OP-100B
HPCS Diesel Generator
2700
N2-OSP-EGS-
M@002
Diesel Generator and Diesel Air Start Valve Operability Test
- Division III
2300
S-EPM-GEN-067
Limitorque MOV Actuator PM
00800
Work Orders
C92535400
C93587063
C93635628
C93667492
C93667768
C93677587
C93703618
C93717108
C93717187
C93721801
Procedures
CC-NM-118-101
Beyond Design Basis Administrative Controls
00400
N1-ST-Q6B
Containment Spray System Loop 121 Quarterly Operability
Test
01600
N1-ST-Q6C
Containment Spray System Loop 112 Quarterly Operability
01800
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Test
N2-OSP-RHS-
Q@005
RHR System Loop B Pump and Valve Operability Test,
System Integrity Test and ASME XI Pressure Test
01300
S-PM-001
Flex 3419MX Water Pump Test
200
Work Orders
C93667166
C93670946
Procedures
Emergency Classification and Protective Action
Recommendations
EP-CE-114-100-
F-05
Part 1 Notification Fact Sheet
G
Corrective Action
Documents
249395
254827
265123
267023
Procedures
NISP-RP-002
Radiation and Contamination Surveys
NISP-RP-004
Radiological Posting and Labeling
NISP-RP-005
Access Controls for High Radiation Areas
Radiation Work
Permits (RWPs)
NM-1-19-00506
Dry Well Scaffold
NM-1-19-00509
Drywell MSIV Activities
NM-1-19-00518
Drywell ISI Activities
NM-1-19-00601
Reactor Building RWCU System Maintenance Activities
NM-1-19-00901
Reactor Disassembly/Reassembly
Corrective Action
Documents
249516
265156
Procedures
Occupational ALARA Planning and Controls
Corrective Action
Documents
248516
Procedures
Exposure Control and Authorization
Dosimetry Issue, Usage, and Control
Operation of the Whole Body Counter (WBC) using APEX-
INVIVO
External Dose Assessments from Contamination
Prenatal Radiation Exposure
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Resulting from
Inspection
280602
Groundwater Monitoring Wells
September 19,
2019
280617
R1 Air Station
September 19,
2019
280625
Eberline Radiation Monitor
September 19,
2019
Procedures
Radioactive Effluent and Environmental Monitoring
Programs
Radiological Environmental Monitoring Program
CY-NM-170-101
Environmental Air Monitoring Sample Collection
CY-NM-170-102
Radiological Sample Collection, Processing and Shipment,
Land Use Census and Quality Control (Vendor Procedure)
Radiological Groundwater Protection Program
S-ENVSP-4.3
Environmental Air Monitoring Station Inspection and
Maintenance
00601
71151
Procedures
Mitigating Systems Performance Index (MSPI) Monitoring
and Margin Evaluation
Regulatory Assessment Performance Indicator Guideline
Corrective Action
Documents
230591
230937
231860
231867
232255
232357
232473
232618
232897
234861
235453
235965
236035
237039
Drawings
C-18006-C
Drywell & Torus Isolation & Blocking Valves P & I Diagram
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
C-18017-C
Emergency Cooling System P & I Diagram
Miscellaneous
NRC Information Notice 2010-03
Failure of Motor-Operated Valves Due to Degraded Stem
Lubricant
February 3,
2010
Procedures
N1-MMP-068-
251
Maintenance of Containment Vacuum Relief Valves
00701
N1-ST-TYC-001
MSIV Type C Leak Rate Tests
200
N1-ST-TYC-005
Emergency Condenser Type C Leak Rate Tests
00100
N1-ST-TYC-008
Torus Vacuum Relief Valve Type C Leak Rate Tests
00000
N1-ST-TYC-026
Loop 11 and 12 Steam Line Drain 1st and 2nd Isolation
Valves Type C Leak Rate Tests
00000
N1-TSP-201-550
Local Leak Rate Test Summary (Type B and C Tests) For
10CFR50 Appendix J and Secondary Containment Bypass
Leakage
01400
Work Orders
C93321759
C93620060
C93628030
C93701790
C93702441
C93702637
C93703870