IR 05000461/2021011
| ML21231A200 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 08/20/2021 |
| From: | Richard Skokowski Engineering Branch 3 |
| To: | Rhoades D Exelon Generation Co |
| References | |
| IR 2021011 | |
| Download: ML21231A200 (14) | |
Text
SUBJECT:
CLINTON POWER STATION - TRIENNIAL INSPECTION OF EVALUATION OF CHANGES, TESTS AND EXPERIMENTS BASELINE INSPECTION REPORT 05000461/2021011
Dear Mr. Rhoades:
On June 30, 2021, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Clinton Power Station. On July 6, 2021, the NRC inspectors discussed the results of this inspection with Ms. N. Plumey, Plant Manager 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; and was determined to be Severity Level IV. 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 the 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 III; the Director, Office of Enforcement; and the NRC Resident Inspector at Clinton Power Station.
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 III; and the NRC Resident Inspector at Clinton Power Station.
August 20, 2021 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, Richard A. Skokowski, Chief Engineering Branch 3 Division of Reactor Safety Docket No. 05000461 License No. NPF-62
Enclosure:
As stated
Inspection Report
Docket Number:
05000461 License Number:
NPF-62 Report Number:
05000461/2021011 Enterprise Identifier:
I-2021-011-0008 Licensee:
Exelon Generation Company, LLC Facility:
Clinton Power Station Location:
Clinton, IL Inspection Dates:
February 01, 2021 to February 05, 2021 Inspectors:
A. Dahbur, Senior Reactor Inspector G. O'Dwyer, Reactor Engineer E. Fernandez, Reactor Inspector Approved By:
Richard A. Skokowski, Chief Engineering Branch 3 Division of Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a triennial inspection of evaluation of changes, tests and experiments baseline inspection at Clinton Power 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 Provide Written Evaluation Describing the Basis for the Change to the Secondary Containment Definition Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Barrier Integrity Green Severity Level IV NCV 05000461/2021011-01 Open
[P.3] -
Resolution 71111.17T The inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, for the licensees failure to provide a written evaluation describing the basis for determining that the change to the secondary containment completed on March 16, 2020, did not require a license amendment. Specifically, the licensee revised the definition for the secondary containment boundary to include the Fuel Building Railroad Airlock (FBRA) without ensuring that the building meets all the Seismic Category I requirements. This change involved utilizing the FBRA and outer door as the secondary containment boundary when the inner door is open and no adverse weather conditions exist. In the event of a severe weather or radioactive release occurrence, the licensee credited dedicated personnel to close the FBRA inner door. This operator action was necessary because the licensee determined that the FBRA meets all requirements of Seismic Category I structures and the secondary containment except for protection from tornadoes. However, no written evaluation was provided describing adequate basis for determining that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety.
Additional Tracking Items
None.
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), inspectors were directed to begin telework. In addition, regional baseline inspections were evaluated to determine if all or a 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
71111.17T - 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 January 01, 2018.
(1)50.59 Evaluation CL-2018-E-006, Bypass SJAE 2nd Stage Low Steam Flow Switches to Keep 1st Stage SJAE Suction Valve 1 CA002A/B from Tripping Closed on Low Flow.
(2)50.59 Evaluation CL-2018-E-007, Abandon PASS Reactor Coolant Sample Line.
(3)50.59 Evaluation CL-2018-E-008, Power Recovery Via Data Validation and Reconciliation (DVR) Methodology.
(4)50.59 Evaluation CL-2018-E-013, Defeat Turbine Thrust Bearing Wear Detector Trips.
(5)50.59 Evaluation CL-2019-E-007, Use of GNF Armor and IronClad Lead Test Assemblies in Clinton Cycle 20 through 22.
(6)50.59 Evaluation CL-2019-E-031, Revise Secondary Containment Design Basis to Credit Fuel Building Railroad Airlock.
- (7) Screening CL-2018-S-004, Replace 1B21R623A & B with two YOKOGAWA DX1006N.
- (8) Screening CL-2018-S-016, Replace TRCM335 with YOKOGAWA DX1006N.
- (9) Screening CL-2018-S-027, Technical Specifications Basis Change to SR 3.3.8.2.1.
- (10) Screening CL-2018-S-043, Remove Main Condenser Tube and Install Mechanical Pin and Collar Type Tubing Sheet Plug.
- (11) Screening CL-2019-S-006, Surveillance Test Interval Extension Calculation Revisions and As-Found Tolerance Changes.
- (13) Screening CL-2019-S-020, Implementation of TDRFP Silent Trip Logic into Ovation.
- (14) Screening CL-2018-S-023, Emergency Containment Venting, Purging, and Vacuum Relief.
- (15) Screening CL-2019-S-028, Digital FWLCS Power Ascension Testing.
- (16) Screening CL-2018-S-034, Replace RFP Bentley Nevada Vibration Monitoring System.
- (17) Screening CL-2018-S-020, Defeating MDRFP Interlock.
- (18) Screening CL-2018-S-006, Bypass SJAE Suction Valve 1CA002A/B Close Trip.
- (19) Screening CL-2019-S-030, ASME Code Reconciliation for Revised Allowable Stress Valves.
INSPECTION RESULTS
Failure to Provide Written Evaluation Describing the Basis for the Change to the Secondary Containment Definition Cornerstone Significance/Severity Cross-Cutting Aspect Report Section Barrier Integrity Green Severity Level IV NCV 05000461/2021011-01 Open
[P.3] -
Resolution 71111.17T The inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, for the licensees failure to provide a written evaluation describing the basis for determining that the change to the secondary containment completed on March 16, 2020, did not require a license amendment. Specifically, the licensee revised the definition for the secondary containment boundary to include the Fuel Building Railroad Airlock (FBRA) without ensuring that the building meets all the Seismic Category I requirements. This change involved utilizing the FBRA and outer door as the secondary containment boundary when the inner door is open and no adverse weather conditions exist.
In the event of a severe weather or radioactive release occurrence, the licensee credited dedicated personnel to close the FBRA inner door. This operator action was necessary because the licensee determined that the FBRA meets all requirements of Seismic Category I structures and the secondary containment except for protection from tornadoes. However, no written evaluation was provided describing adequate basis for determining that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety.
Description:
The FBRA inner door, Equipment ID Number (EIN) 1HC72G/1SD1-30 currently forms part of the secondary containment boundary. Each time this door is opened to move equipment in and out of the fuel building, the secondary containment is declared inoperable, and Technical Specification Action Statement 3.6.4.1.A is entered. This Action Statement requires the secondary containment to be restored within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, which imposes a burden on plant operations, especially during ISFSI campaigns where the large dry casks are moved into and out of the fuel building. Therefore, it was desired to revise the definition of the secondary containment boundary to include the FBRA. On December 18, 2014, the licensee completed a 50.59 evaluation CL-2014-E-033 and determined that the change did not require a prior NRC approval. The inspectors reviewed the 50.59 evaluation and identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, for the licensees failure to provide a written evaluation describing the basis for determining that the change to the secondary containment completed on December 18, 2014, did not require a license amendment. Specifically, the inspectors determined that the licensee failed to provide basis for the change, which eliminated the tornado wind and tornado missile loading condition from the FB railroad airlock (the enclosure walls and roof), and associated outer door (1SD1-31)
Seismic Category I requirements did not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety.
On March 16, 2020, the licensee completed another 50.59 evaluation CL-2019-E-031 and revised the definition of the secondary containment boundary at the FBRA to ensure that both secondary containment capability and tornado protection are maintained when required.
Specifically, the change involved utilizing the FBRA and outer door as the secondary containment boundary when the inner door is open and no adverse weather conditions exist.
Because of the inspectors challenge to CL-2014-E-033 for not addressing tornado wind and missile protection for the FBRA, this 50.59 Evaluation CL-2019-E-031 included specific, situational control of the FBRA inner door to ensure tornado protection when required.
Manual actions associated with the operation of the FBRA inner door were incorporated into procedure to ensure secondary containment integrity in the event of a severe weather or radioactive release occurrence, the licensee credited dedicated personnel to close the FBRA inner door. This operators manual action was necessary because the licensee determined that the FBRA meets all requirements of Seismic Category I structures and the secondary containment except for protection from tornadoes.
The licensee concluded that this change did not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR. This was justified by the following six items:
1. The FBRA meets all requirements of Seismic Category I structures and the secondary
containment except for protection from tornadoes.
2. The Standby Gas Treatment System (SGTS) has been demonstrated to perform its
required design functions with the additional volume and leakage provided by the FBRA.
3. The design of the secondary containment remains consistent with the Regulatory
Requirements associated with the postulated occurrence of tornadoes. The design function of the secondary containment is to capture leakage from the primary containment during a LOCA to mitigate the dose consequences. While the combination of a simultaneous initiation of a LOCA and a tornado is considered not credible, the occurrence of a tornado is considered credible during the post-LOCA long-term cooling phase (Regulatory Guide 1.117).
4. Consistent with #3 above, manual actions associated with the operation of the inner
door are being incorporated into the design to ensure the secondary containment integrity. Specifically, this is accomplished by requiring that the inner doors are shut whenever a radioactive release is in progress or when severe weather is imminent.
This will ensure the FBRA inner doors are closed prior to a tornado during the post-LOCA long-term cooling phase. The licensee concluded that these actions met the guidance specified in Section 4.3.2, Example #4 of NEI 96-07, for the new operator manual actions.
5. The SSCs that are provided with tornado protection which are co-located within the
secondary containment structures are unaffected. Ongoing protection is provided to the spent fuel pool, ECCS, main steam lines and isolation valves, and required support systems.
6. No SSCs that are required for a response to a tornado are protected by the FBRA
structure.
The licensee also concluded that this change did not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR. This was based on that the impact on accident consequences of utilizing the FBRA and outer door as the secondary containment boundary has already been evaluated and approved per License Amendment No. 210, dated August 17, 2016. The 50.59 Evaluation CL-2019-E-031 indicated that license amendment increased the secondary containment drawdown time from 12 minutes to 19 minutes due to the additional volume from the FBRA and the heat load in the secondary containment from a loaded spent fuel cask.
The inspectors reviewed Licensee Amendment No. 210 and noticed that in section titled Related Regulatory Findings, the NRC staff recognized that on November 6, 2015, the NRC issued a finding to Clinton for failure to obtain a license amendment prior to making modifications to secondary containment for engineering change attempted to modify the established boundary of the secondary containment to include the FB railroad bay airlock.
The NRC staff questioned whether this amendment request was intended to support expansion of secondary containment to include the FB Railroad Bay Airlock, as the initial request did not address the FB Railroad Bay Airlock. The amendment also stated that in the June 2, 2016, supplement, the licensee indicated that a typical drawdown time test result was about 28 seconds and additional testing in 2014 with the fuel building railroad bay airlock volume included was 37 seconds, although the design basis configuration is with the railroad bay airlock inner door closed and this additional volume excluded. However, the licensee indicated that no additional change in the drawdown Surveillance Requirement procedure acceptance criterion is planned and that the FB Railroad Bay Airlock will not be incorporated into secondary containment. The inspectors reviewed the licensee amendment and its associated documents and determined that the FB Railroad Bay Airlock was not part of the NRC staff reviewed change.
The inspectors reviewed Clinton 10 CFR 50.59 Evaluation CL-2019-E-031 and determined that when the licensee made the change to revise the definition for the secondary containment boundary to include the Fuel Building Railroad Airlock, the licensee failed to provide a written evaluation describing the basis for determining that the change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. Specifically, this change involved utilizing the FBRA and outer door as the secondary containment boundary when the inner door is open and no adverse weather conditions exist. In the event of a severe weather or radioactive release occurrence, the licensee credited dedicated personnel to close the FBRA inner door to maintain the functional capability and design/licensing requirement of the secondary containment. The licensee relied on these operators manual actions because the licensee determined that the FBRA meets all requirements of Seismic Category I structures and the secondary containment except for protection from tornado wind and tornado missile. However, there was no evaluation of the manual actions, which is a change in its own right as to whether the use of these actions was acceptable and not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety.
Furthermore, the guidance in Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, Section 4.3, in that concurrent changes that were not interdependent were inappropriately linked and, as a result, were not evaluated separately and independently of each other. Specifically, the concurrent changes that were not interdependent were:
Change to the facility to credit the use FBRA and outer door as the secondary containment boundary when the inner door is open and no adverse weather conditions exist. Per NEI 96-07, Section 4.3.2, this change involved a change in design requirements for tornadoes and should has been treated as potentially affecting the likelihood of malfunction of an SSC and would result in a more than minimal increase in the likelihood of malfunction of the FBRA and outer door in performing the design function as secondary containment boundary. While the FBRA is seismically qualified for the Operating Basis Earthquake and Safe Shutdown Earthquake, the FBRA and outer door are not designed for the effects of tornado wind loads or tornado. (USAR Section 6.2.3.1 states, in part, that The secondary containment structures is of Seismic Category I design USAR Table 3.2-1 Note (c), states that all Seismic Category I structures are designed for the effects of CPS natural phenomena such as tornado, wind loads, external missiles, floods, etc.,
except the Containment Gas Control Boundary (CGCB).)
Change to procedures to add manual actions to close the FBRA inner door whenever a radioactive release is in progress or when severe weather is imminent to isolate the FBRA prior to a tornado during the post-LOCA long-term cooling phase was a change to operate existing equipment (e.g., the FBRA inner door) that was already credited for performing design function as secondary containment boundary and it was not an interdependent procedure change but rather is a compensatory action for addressing nonconforming condition that the FBRA and outer door do not meet the requirement to withstand the effects of tornado wind loads or tornado missiles.
NEI 96-07, Revision 1, Section 4.3, Evaluation Process, stated, in part, that each element of a proposed activity must undergo a 10 CFR 50.59 evaluation, except in instances where linking elements of an activity is appropriate, in which case the linked elements can be evaluated together. A test for linking elements of proposed changes is interdependence. It is appropriate for discrete elements to be evaluated together if
- (1) they are interdependent as in the case where a modification to a system or component necessitates additional changes to other systems or procedures; or
- (2) they are performed collectively to address a design or operational issue. If concurrent changes are being made that are not linked, each must be evaluated separately and independently of each other. Furthermore, in this case the licensees proposed new operator action to close the inner door in the event of a severe weather or radioactive release occurrence as a means of determining that the tornado missile nonconforming condition would not result in a more than minimal increase. However, the licensees 50.59 evaluation failed to consider that the tornado missile nonconforming condition still exists regardless of the new operator action. To accept the tornado missile nonconforming condition as-is, the licensee must apply NEI 96-07, Section 4.3.2, which states, Further, departures from the design, fabrication, construction, testing and performance standards as outlined in the General Design Criteria (Appendix A to Part 50) are not compatible with a no more than minimal increase standard. In addition, NEI 96-07, Section 4.4 guidance for addressing nonconforming conditions states, If an interim compensatory action is taken to address the condition and involves a temporary procedure or facility change, 10 CFR 50.59 should be applied to the temporary change. The intent is to determine whether the temporary change/compensatory action itself (not the degraded condition) impacts other aspects of the facility or procedures described in the UFSAR. If the licensee corrective action is either to accept the condition as-is resulting in something different than its as-designed condition, or to change the facility or procedures, 10 CFR 50.59 should be applied to the corrective action. Moreover, the inspectors did not find examples where NRR staff has reviewed and approved manual actions to meet Seismic Category I (USAR 3.8.4)/Secondary Containment (USAR 6.2.3) requirements.
Corrective Actions: Review any site impact from the 50.59 evaluation and make changes as needed. Develop alternative proposal for addressing FBRA operational concerns.
Corrective Action References: AR 04434012
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to provide a written evaluation describing the basis for determining that the change to the secondary containment definition in the UFSAR, which was completed on March 16, 2020, did not require a license amendment was contrary to 10 CFR 50.59(d)(1) and was a performance deficiency. Specifically, the licensee made a change to the secondary containment pursuant to 10 CFR 50.59(c) and eliminated the tornado wind and tornado missile loading condition from the FB Railroad Airlock (the enclosure walls and roof) and associated outer door (1SD1-31) Seismic Category I requirements when the inner door was open and no adverse weather conditions exist, and did not provide a written evaluation describing the basis for determining that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. The FBRA did not meet Seismic Category I requirements to withstand the effects of tornado wind loads or tornado missiles. Instead, the licensee credited dedicated personnel to close the inner door in the event of a severe weather or radioactive release occur.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine if the changes to secondary containment would have required NRC prior approval.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process (SDP) because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP and is used to inform the severity of the violation.
In this case, the inspectors used IMC 0609, Attachment 4, Initial Characterization of Findings, issued December 13, 2019, and Appendix A, The Significance Determination Process for Findings at Power, issued November 30, 2020, to evaluate the technical issue.
The finding was screened against the Barrier Integrity cornerstone and determined to be of very low safety significance (Green) because the finding only represent a degradation of the radiological barrier function for the SGTS, and it not represent a degradation of the function of the control room against smoke or toxic atmosphere.
Cross-Cutting Aspect: P.3 - Resolution: The organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee failed to adequately correct the 50.59 violation previously identified in Inspection Report 2015003.
Enforcement:
The Reactor Oversite Programs (ROP) SDP 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: The Severity Level is consistent with Example 6.1.d.2 of the NRC Enforcement Policy.
Violation: Title 10 CFR 50.59, Changes, Tests and Experiments, (d)(1) states, in part, that a licensee shall maintain records of changes in the facility made pursuant to paragraph
- (c) of this section. These records must include a written evaluation that provides the bases for the determination that the change does not require a license amendment pursuant to 10 CFR 50.59(c)(2).
UFSAR Section 6.2.3.1 states, in part, that The secondary containment structures is of Seismic Category I design USAR Table 3.2-1 Note (c), states that all Seismic Category I structures are designed for the effects of CPS natural phenomena such as tornado, wind loads, external missiles, floods, etc., except the Containment Gas Control Boundary (CGCB).
Contrary to the above, as of March 16, 2020, the licensees records for the change to the secondary containment boundary definition in the UFSAR failed to include a written evaluation that provides the bases for the determination that the change does not require a license amendment pursuant to 10 CFR 50.59(c)(2).
Specifically, the licensee revised the definition for the secondary containment boundary to include the FBRA without ensuring that the building meets all the Seismic Category I requirements.
In accordance with the Enforcement Policy, (Section 6.1.d.2), the violation was classified as a Severity Level IV violation because the underlying technical issue was of very low risk significance. Because this finding was of very low safety significance, was not repetitive or willful, and was entered into the licensees CAP as AR 04434012, this violation is being treated as an NCV.
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 July 6, 2021, the inspectors presented the triennial inspection of evaluation of changes, tests and experiments baseline inspection results to Ms. N. Plumey, Plant Manager and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
50.59 Deficiencies for EC 619869 Bypassing Flow Switches
07/27/2018
Inadequate 50.59 Screening for Steam Dryer Stop Holes
08/01/2018
Corrective Action
Documents
Demin Access Plug PBI Greater than 90 days with No 50.59
Review
07/18/2018
Corrective Action
Documents
Resulting from
Inspection
Potential Issue from 50.59 Inspection Exit
07/06/2021
CL 2018-E-006
Bypass SJAE 2nd Stage Low Steam Flow Switches
CL-2018-E-008
Power Recovery Via Data Validation & Reconciliation (DVR)
Methodology
CL-2018-E-013
Thrust Bearing Wear Detector Automatic Trip Defeat
CL-2018-S-003
Fuel Building Crane Rail Evaluation/UFSAR
Change #2018-002
CL-2018-S-006
Bypass SJAE Suction Valve 1CA002A/B Close Trip
CL-2018-S-016
Change TRCM335
CL-2018-S-020
Defeating MDRFP Interlock
CL-2018-S-023
Emergency Containment Venting, Purging, and Vacuum
Relief
CL-2018-S-027
Technical Specifications Basis Change to SR 3.3.8.2.1
CL-2018-S-034
Replace RFP Bently Nevada Vibration Monitoring System
000
CL-2018-S-035
Modify SJAE AOV Suction Valve 1CA002B with Redundant
Solenoids
CL-2018-S-043
Remove Main Condenser Tube and Install Mechanical Pin
and Collar Type Tubing Sheet Plug
CL-2019-E-007
Use of GNF ARMOR and Ironclad Lead Test Assemblies in
Clinton Cycles 20 through 22
CL-2019-E-031
Revised Secondary Containment Design Basis to Credit Fuel
Building Railroad Airlock
CL-2019-S-006
Surveillance Test Interval Extension Calculation Revisions
and As-Found Tolerance Changes
Engineering
Evaluations
CL-2019-S-018
Temporary Instrument Tubing for Plant Chiller OW002CA
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
CL-2019-S-019
Clarification of DG Load for TS Basis SR 3.8.1.18
CL-2019-S-020
Implementation of TDRFP Silent Trip Logic into Ovation
CL-2019-S-028
Digital FWLCS Power Ascension Testing
CL-2019-S-030
ASME Code Reconciliation for Revised Allowable Stress
Values
CL-2020-S-001
Engineering Review of Scaffold 1445 in Place Greater than
days
CL-2020-S-012
Plant Barrier Impairment in Place Greater than 90 days
CL-2020-S-030
Installation of Mechanical Gag on Valve 1TD600 in Support
of Maintenance
Miscellaneous
20-05
UFSAR Change for MCR Habitability Hazardous Chemical
Survey Analysis 2019
06/02/2020