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The following is a summary of information received from the State of Oregon via phone call: On August 28, 2019, at 1102 PDT, the Oregon Department of Health and Radiation Protection received a report from the R.S. Davis Scrap Yard, that a moisture density gauge was found in a 55 gallon drum that was crimped at the top so the gauge will not fall out. The gauge is a CPN MC-3 Porta Probe with a 10 mCi Cs-137 source and a 50 mCi Am/Be 241 source. The scrap yard employed a contractor to take radiation readings and perform a swipe test of the gauge. The gauge housing appears to be damaged but the source housing seems intact. The contractor was unable to inspect the bottom of the gauge inside the drum.

The following was measured: A reading of 13 mR/hr on contact at the side A reading of 200 mR/hr on contact at the bottom A reading of 3 mR/hr one foot from the side Swipe Test results were less than background The State of Oregon Radiation Inspector believes the gauge is damaged because normal radiation readings for a similar gauge at one meter would be 0.4 mR/hr. The State of Oregon will be conducting an investigation of this event. The gauge is currently under the control of the scrap yard. Oregon Event Reference Number: 19-0037. THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf

EN Revision Text: TECHNICAL SPECIFICATION REQUIRED SHUTDOWN - LOSS OF 480 VOLTAGE EMERGENCY BUS On 8-12-18 at 0158 EDT, Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. This resulted in a Loss of Safety Function due to the 2-2 Emergency Diesel Generator (EDG) being Inoperable coincident with the Residual Heat Release Valve (2SVS-HCV104). A Technical Specification shutdown is required per LCO 3.0.3. The Licensee also stated they were in an unanalyzed condition due to the EDG and Residual Heat Release Valve being inoperable at the same time. The Licensee is shutting down to Mode 5 (Cold Shutdown). The Licensee is notifying the Resident Inspector. The Licensee will be making a Press Release about the unplanned shutdown.

• * * UPDATE ON 08/16/2018 AT 1424 EDT FROM BLASE BARTKO TO KEN MOTT * * *

On 8-12-18 at 0158 (EDT) Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. Per operational guidance, this was determined to be a Loss of Safety Function due to the Unit 2 Emergency Diesel Generator (EDG) being INOPERABLE coincident with the Residual Heat Release Valve (2SVS-HCV104) 10 CFR 50.72(b)(3)(v)(B) and (D). This was also reported as an Unanalyzed Condition 10 CFR 50.72(b)(3)(ii)(b). No Press Release was performed for this event. The NRC Resident Inspector was notified. At 0410 (EDT) a Technical Specification Shutdown was commenced 10 CFR 50.72(b)(2)(i). At 2011 (EDT) the 480 Volt 2P Emergency Bus was restored and energized. Further evaluation of the event has determined that this event was not an Unanalyzed Condition and did not result in a Loss of Safety Function. The classifications of Unanalyzed Condition and Loss of Safety Function are being retracted. The accuracy of the existing guidance relative to Safety Function has been entered in the Corrective Action Program and interim actions have been taken to provide accurate guidance. Notified R1DO (Young) via email.

EN Revision Text: PART 21 - INTERIM REPORT NOTIFICATION The following was received via phone call and email: This report provides notification and interim information concerning an evaluation being performed by AAF Flanders for an unapproved design change in a High Efficiency Particulate Air Filter. An evaluation is underway for filters that underwent a non-approved design change. AAF Flanders has determined that an evaluation cannot be completed within the 60 day period. Discovery of the potential deviation was May 2, 2018. The information required for the 60-Day Interim Report Notification �21.21(a)(2) was provided. We anticipate that the evaluation will be completed by Sept 15, 2018. AAF Flanders is evaluating a potential nonconforming condition associated with filters (model number 0-007-C-42-03-NU-11-13-GG FU5) supplied to Prairie Island Nuclear Generating Plant (PINGP) / Xcel Energy. AAF Flanders notified Prairie Island Nuclear Plant of this potential defect.

• * * UPDATE ON 9/14/2108 AT 1129 EDT FROM SHAWN WINDLEY TO ANDREW WAUGH * * *

The following information was received via email: A notification was submitted to the Commission with the subject matter of, 'Unapproved Design Change in a High Efficiency Particulate Air Filter.' At this time, the evaluation is pending third party qualification testing of the product. Information obtained from the qualification will be used in the determination of a defect. AAF Flanders had anticipated this process to have been completed by Sept 15, 2018 but because it is still on-going, we request an extension until October 31, 2018 to submit a final report to the Commission. The subject filters (model number 0-007-C-42-03-NU-11-13-GG FU5) supplied to Prairie Island Nuclear Generating Plant (PINGP)/ Xcel Energy had not been installed. AAF Flanders has recalled the subject filters and currently have them segregated and stored at our facility. They pose no threat public safety. Notified R3DO (Hanna) and Part 21/50.55 Reactors Group (email).

• * * UPDATE ON 10/31/18 AT 1545 EDT FROM SHAWN WINDLEY TO HOWIE CROUCH * * *

The following information was excerpted from information received via email: AAF Flanders had anticipated the qualification process for said filters to have been completed by Sept 15, 2018. AAF requested a second extension by October 31, 2018 to submit a final report to the Commission, however; at this time our qualification is still pending third party approval. AAF Flanders is requesting another extension on the basis of an incomplete qualification of these filters. We anticipate qualification and a completed report on or before 12/15/2018. Notified R3DO (Stoedter) and Part 21/50.55 Reactors Group (email).

• * * RETRACTION ON 08/20/19 AT 1140 EDT FROM SHAWN WINDLEY TO JEFFREY WHITED * * *

The following information was received via e-mail: AAF Flanders' evaluation of the condition identified in the interim report has been completed. The condition reported has been determined to be not reportable in accordance with 10 CFR 21. The subject filters (model number 0-007-C-42-03-NU-11-13-GG FU5) have passed qualification testing in accordance with the American Society of Mechanical Engineers (ASME) AG-1, Section FC with the new urethane and are fully qualified and listed on the Qualified Products List. The subject filters (model number 0-007-C-42-03-NU-11-13-GG FU5) supplied to Prairie Island Nuclear Generating Plant (PINGP)/Xcel Energy prior to this qualification test had not been installed and were subsequently returned to AAF Flanders. PINGP/Xcel Energy has been notified of the qualification status of the filter model. Notified R3DO (Riemer) and Part 21/50.55 Reactors Group (email).

I. EVENT DESCRIPTION: High enriched scrap fuel material is processed in BWXT NOG-Lynchburg's Uranium Recovery facility to reclaim as much of the uranium as possible. The material is dissolved in acid and transferred to a series of horizontal columns where the acid is neutralized. The solution may be transferred to a set of accountability weigh columns for measurement prior to entering the uranium extraction process. The solution is subsequently transferred to a series of horizontal feed columns. Process water is used to periodically flush the horizontal columns during cleanup for materials accountability. On June 9, 2016, a BWXT Nuclear Criticality Safety (NCS) engineer was notified that a bluish tint had been observed in the favorable geometry process water connection to the horizontal columns. By procedure, a blue dye is added to the acid to aid in its identification in the event of a spill. Further evaluation determined that the favorable geometry process water line was directly connected to the horizontal column system and the presence of the blue dye indicated a potential backflow of uranium bearing solution into the water line. The favorable geometry water line is under constant water pressure. The valves controlling the water flow are normally closed. There is also a check valve in the line to prevent backflow. The line is supplied from a favorable geometry header on the mezzanine above. The header supplies water to other processes in Uranium Recovery, including an unfavorable geometry hot water heater. The Integrated Safety Analysis (ISA) was reviewed and an accident sequence for this potential backflow could not be identified. On June 9, 2016 at 1330 (EDT) it was the determined the accident sequence was unanalyzed and not properly documented in the ISA. Although IROFS (Items Relied on for Safety) listed for other accident sequences were applicable to the backflow scenario, the performance requirements of 10 CFR 70.61 were not maintained. As documented in the ISA, criticality was not 'highly unlikely.' II. EVALUATION OF THE EVENT: Backflow into the favorable geometry water line can only be achieved by a forced flow to overcome the water pressure in the line. The only source for this pressurized flow is during the transfer from the accountability weigh columns to the horizontal column system. The solution is transferred using an air diaphragm pump. A trained and qualified operator opens the valve to initiate the transfer of solution from the accountability weigh columns to the horizontal columns (an uncredited control). The transfer of solution to the horizontal columns is monitored by a trained and qualified operator (an uncredited control). The water heater is substantially upstream of the supply line to the horizontal columns. The direction of flow of the process water in the supply header is away from unfavorable geometry hot water heater. The process water header is a favorable geometry (a credited IROFS). An operator checks the process water pressure on a daily basis (a credited IROFS). If the above existing IROFS and uncredited controls were considered in an ISA accident sequence, the likelihood of a criticality could be demonstrated to be highly unlikely. However, these uncredited controls are not designated as IROFS. Although the as-found condition presented no safety concern, the scenarios as documented in the ISA did not demonstrate that the performance requirements of 10 CFR 70.61 were maintained. There was no immediate risk of a criticality or threat to the safety of workers or the public as a result of this event. Ill. NOTIFICATION REQUIREMENTS: BWXT is making this 24 hour report in accordance with 10 CFR 70, Appendix A, (b)(1), 'Any event or condition that results in the facility being in a state that was not analyzed, was improperly analyzed, or is different from that analyzed in the Integrated Safety Analysis, and which results in failure to meet the performance requirements of 70.61.' IV. STATUS OF CORRECTIVE ACTIONS: A section of the piping from the process water supply header was removed to physically isolate the process water supply from the horizontal columns. Criticality is no longer credible. The hot water heater is the only unfavorable geometry connected to the process water system in the Uranium Recovery facility. The hot water heater was assayed with a gamma survey instrument in several locations along the bottom and up the sides. No counts above background were detected. In addition, multiple liquid samples were taken from the bottom of the water heater and a cartridge filter housing prior to the hot water heater. All samples were well counted and determined to be below the Minimum Detectable Activity (MDA). An investigation of the root causes of this event is ongoing. Corrective actions will be determined as a result of the investigation. The licensee has notified the NRC Resident Inspector and Region II personnel.

• * * UPDATE AT 0944 EDT ON 06/09/17 FROM ROBERT JOHNSON TO S. SANDIN * * *

I. EVENT DESCRIPTION: On June 10th, 2016 BWXT NOG-Lynchburg notified the NRC of an improperly analyzed condition involving the potential backflow of uranium bearing solution from a fissile solution processing system into the Uranium Recovery process water system. This notification was recorded as Event Notification Number 51998. One of the corrective actions in response to Apparent Violation 70-27/2016-004-01 was to conduct an Extent of Cause review of 'tasks involving ancillary systems (e.g., process water, nitric acid, HF acid, compressed air, steam, etc.) in our Uranium Recovery and Specialty Fuel Facility to verify these systems have documented accident scenarios as needed to meet the requirements of 10 CFR 70.' During the review of a waste processing system in the Specialty Fuel Facility (SFF), an additional potential backflow scenario was identified for a waste processing system. The system consists of a series of four favorable geometry columns. Waste solution is transferred to the columns using a less than or equal to 2.5 liter pump. The transfer of solution to the columns is monitored by a trained and qualified operator. Waste solution transferred to the columns is limited to a concentration of 5 grams 235U/liter as a Routine Operating Limit. A process water line is directly connected to the column system. The water is used to further dilute the concentration of the waste solution to a level that is acceptable for discharge into the hot waste drain (less than 0.04 grams 235U/liter). The column system is equipped with vent lines that overflow to the floor. The pump is capable of over-pressurizing the columns and possibly forcing waste solution into the favorable geometry process water system if the overflow vent lines were to fail. An unfavorable geometry hot water heater is located a significant distance downstream of the connection to the process water system. The Integrated Safety Analysis (ISA) was reviewed and an accident sequence for potential backflow of solution from the waste columns into the process water system could not be identified. II. EVALUATION OF THE EVENT: Administrative IROFS were identified in other accident sequences of the ISA that are applicable to the waste column backflow scenario. These IROFS include the operator control of solution concentration initially transferred to the waste processing system, and the daily check of the process water pressure which limits backflow into the system. Although these IROFS were available and reliable, additional IROFS are needed to meet the performance requirements of 10 CFR 70.61. Neither the operator's monitoring of the solution transfer or the overflow through the column vent lines are credited as IROFS in the ISA. They are considered uncredited safety controls. If either of the uncredited safety controls and the currently existing IROFS could be considered in an accident sequence, the likelihood of a criticality could be demonstrated to be highly unlikely. However, only IROFS documented in the ISA can be credited as preventing a criticality accident. Although the as-found condition presented no safety concern, the scenarios as documented in the ISA did not demonstrate the performance requirements of 10 CFR 70.61 were maintained. On June 8, 2017 at 10:00 am it was the determined the potential backflow of solution from the waste processing system was improperly analyzed and not documented in the ISA. As documented in the ISA, criticality was not highly unlikely. The SFF waste processing system was not in operation at the time of discovery. There was no immediate risk of a criticality or threat to the safety of workers or the public as a result of this event. Ill. NOTIFICATION REQUIREMENT: BWXT is making this 24 hour report to update Event Notification Number 51998 in accordance with 10 CFR 70, Appendix A, (b)(1) - Any event or condition that results in the facility being in a state that was not analyzed, was improperly analyzed, or is different from that analyzed in the Integrated Safety Analysis, and which results in failure to meet the performance requirements of 70.61. IV. STATUS OF CORRECTIVE ACTIONS: The process water line connected to the SFF waste processing system was locked out and posted as 'Out of Service.' Additional corrective actions are to be determined. The Extent of Cause review for AV 70-27/2016-004-01 is complete and the results are being finalized. The licensee informed the NRC Resident Inspector. Notified R2DO (Suggs), NMSS Events Notification, and Fuels Group by email.

The following was received via FAX: On May 18, 2016, Nuclear Logistics INC. (NLI) determined that a contactor failure that occurred at the Shearon Harris plant had failed due to an auxiliary contact chatter present on the seal-in circuit for the coil voltage. The auxiliary contact chatter was caused by the loss of the shading coils. NLI will be submitting a full report on the issue to the NRC within 60 days. The contactor that failed was a Size 4 Eaton Freedom Series with a special coil for degraded voltage condition. Reference Number: P21-05192016 NLI reported that the following plants have these types of contactor's: Region 2: Oconee, Turkey Point, Shearon Harris and North Anna Region 4: Columbia and Waterford

• * * UPDATE AT 1832 EDT ON 06/20/16 FROM TRACY BOLT TO DANIEL MILLS * * *

The following is excerpted from the licensee submission: The specific part which fails to comply or contains a defect: The Contactor that failed in service is a Size 4 Eaton Freedom Series with an AZZ/NLI special coil for meeting specific degraded voltage conditions. Extent of condition: Size 3, 4 and size 5 Eaton Freedom Series contactors or starters with an NLI special degraded voltage coil that have been supplied by AZZ/NLI since December 2010. NLI procured the commercial grade contactors and installed the special coil that was required to achieve the specific degraded voltage condition. The units were qualified, dedicated and supplied for safety related applications. The contactors were commercially procured from Eaton, the Original Equipment Manufacturer (OEM). For contactors/starters utilized in continuous duty applications. the OEM shading coils on the contactor core ... have the potential to become loose and fall off. If the shading coils are not in the intended location on the core, there is the potential for excessive chatter to occur on the normally open auxiliary contacts that are closed when the contactor is energized. The safety function of the contactor is to reliably supply uninterrupted power (no contact chatter) to a load on demand. For special degraded voltage applications, the NLI supplied contactor is equipped with an NLI special coil that replaces the OEM coil. When the contactor/starters that have the special coil installed are utilized in a continuous duty operation (continuously energized greater than 60 minutes) the special coil reaches a higher temperature than the original manufacturer's coil. The increased heat is potentially causing degradation of the acrylic resin that is utilized by the manufacturer to hold the OEM shading coils onto the OEM core. After the acrylic resin is no longer providing a secure hold on the shading coils, the shading coils can then become loose from the iron core. Name and address of the individual or individuals informing the Commission. Tracy Bolt, Director of Quality Assurance Nuclear Logistics, Inc. 7410 Pebble Drive Ft. Worth, TX 76118 AZZ/NLI Part 21 Report No: P21-05192016, Rev. 0 Plants potentially impacted include Oconee, Shearon Harris, Columbia, Turkey Point, North Anna, and Waterford. Notified R4DO (Rollins), R2DO (Musser) and Part 21/50.55 Reactors group (via email).

At approximately 1900 EDT on March 28, 2016, an unplanned loss of retail power to the McGuire Meteorological tower occurred. This loss of power impacted communication between the Meteorological tower and the control room, which resulted in a major loss of emergency assessment capability. Compensatory measures existed within the site's emergency planning procedures to obtain meteorological data from the National Weather Service. Meteorological information could have been provided via the emergency notification system to the NRC operations center. Power was restored at 2245 EDT on March 28, 2016. The NRC Resident Inspector has been notified. There was no impact to the public.

• * * RETRACTION AT 1423 EDT ON 4/19/2016 FROM SCOTT SLIETER TO MARK ABRAMOVITZ * * *

This notification is being made to retract event EN 51828 that was reported March 29, 2016. Based on further investigation, Emergency Plan Implementing Procedures provide acceptable alternative methods to perform emergency assessment that are in addition to the control room indications from the meteorological tower. Furthermore, it was determined that the meteorological tower data was valid and available in the control room during the applicable period. It was therefore determined that no actual or potential major loss of emergency assessment capability existed per 10CFR50.72(b)(3)(xiii). This is consistent with NUREG 1022, rev 3, supplement 1 and NEI 13-01, rev 0. The NRC Resident Inspector has been notified of the retraction. Notified the R2DO (McCoy).

The following information was provided via email and telephone. Fort Calhoun Station is currently completing a scheduled refueling outage. On June 5, 2015 at 1330 during performance of surveillance testing on the auxiliary feed water system, (Hydraulic Control Valve) HCV-1107A, Steam Generator RC-2A Auxiliary Feedwater Inlet Valve, did not open when given an open signal. HCV-1107A has been declared inoperable. HCV-1107A is required to open to meet the decay heat removal safety function for Steam Generator A. Fort Calhoun Station is in Mode 3 (Reactor Coolant System temperature is greater than 515 degrees Fahrenheit and not critical). With HCV-1107A inoperable and unable to feed the A steam generator both auxiliary feedwater trains are considered inoperable. HCV-1107A is inside the Containment Building. Fort Calhoun Station Technical Specifications 2.5(1)D. requires: With both AFW trains inoperable, then initiate actions to restore one AFW train to OPERABLE status immediately. Technical Specification (TS) 2.0.1 and all TS actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status. Fort Calhoun Station is evaluating the best approach to repairing HCV-1107A. The Resident Inspector has been notified

• * * UPDATE PROVIDED BY CHARLIES SMITH TO RICHARD SMITH AT 2300 EDT ON 06/05/2015 * * *

Fort Calhoun Station has determined plant cooldown required to perform repairs. Plant cooldown in progress. The licensee will notify the NRC Resident Inspector. Notified R4DO (Whitten)