L-20-119, Report of Facility Changes, Tests, and Experiments

Report of Facility Changes, Tests, and Experiments

ML20125A315
Person / Time
Site:	Davis Besse
Issue date:	05/04/2020
From:	Bezilla M Energy Harbor Nuclear Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-20-119
Download: ML20125A315 (6)

Text

.a energy Energy Harbor Nuclear Corp.

Davis-Besse Nuclear Power Station

~ harbor 5501 N. State Route 2 Oak Harbor, Ohio 43449 Mark B. Bezilla 419-321-7676 Site Vice President, Davis-Besse Nuclear May 4, 2020 L-20-119 10 CFR 50.59(d)(2)

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 Report of Facility Changes, Tests, and Experiments In accordance with 10 CFR 50.59(d)(2), Energy Harbor Nuclear Corp. hereby submits the Report of Facility Changes, Tests, and Experiments for the Davis-Besse Nuclear Power Station, Unit No. 1. The attached report covers the period of May 12, 2018 through May 4, 2020.

There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A.

sµ/4.~

Lentz, Manager, Nuclear Licensing and Regulatory Affairs, at (330) 315-6810.

Mark B. Bezilla /

Attachment:

Davis-Besse Nuclear Power Station, Unit No. 1 Report of Facility Changes, Tests, and Experiments cc: NRC Region Ill Administrator NRC Resident Inspector NRC Project Manager Utility Radiological Safety Board

Attachment L-20-119 Davis-Besse Nuclear Power Station, Unit No. 1 Report of Facility Changes, Tests, and Experiments Page 1 of 5

Title:

Component Cooling Water System Procedure Activity

Description:

This activity involved identifying a slight increase trend in oil leakage from the inboard bearing of the component cooling water (CCW) 1 pump. Operational oil leakage from the inboard sleeve bearing adaptor of the CCW 1 pump increased to an amount that requires oil addition during the 30-day mission time. With the recent leakage trend, this condition has the potential to affect pump performance during its required mission time if actions are not taken to ensure oil level is maintained at an adequate level.

A similar oil leakage trend was identified that was associated with the inboard bearing of the CCW 3 pump. CCW 2 pump is known to have bearing oil leaks. These issues have been documented in the corrective action program. All three CCW pumps are the same model and have the same bearing configurations.

The CCW pump leakage from all three pumps, inboard and outboard bearings, was evaluated in a Prompt Operability Determination (POD). The POD requires compensatory actions to support CCW pump operability for its mission time with the identified leakage.

The CCW system procedure was updated to implement the following compensatory actions.

1. During normal operations, at least once per shift, operators shall verify the oil levels for all six CCW pump bearing oil sight glasses are above the halfway point of the operating band identified in the CCW pump lubrication data sheets. If, at any time, oil level of any of the CCW pump bearings is identified to be at or below the halfway point of the operating band marked on the sight glass, then oil shall be added to refill that bearing's reservoir to the high mark on the sight glass. Oil addition shall be completed within one shift of identification.

2. During post-accident conditions, Operations shall monitor pump bearing oil levels at 24-hours intervals and request maintenance to add oil as needed to maintain oil level at or above the halfway point of the operating band identified in the CCW pump lubrication data sheets.

Attachment L-20-119 Page 2 of 5 Summary of Evaluation:

All design basis functions of the CCW pumps were met during, and remain met after, implementation of this activity. No accident initiators are altered. The likelihood of equipment malfunction is not increased. The consequences of accidents or malfunction of systems, structures, or components are unchanged. No new or different accidents or malfunctions of systems, structures or components are created. No fission product barrier limits are changed or challenged. No new methods of analysis have been used in the safety analyses performed to assess the change. In conclusion, the proposed revision of the procedure does not meet any of the 10 CFR 50.59(c)(2) criteria; therefore, a license amendment is not required.

Title:

Control Rod Survivability Analysis During a Loss-of-Coolant Accident (LOCA)

Activity

Description:

This activity involved evaluation and implementation of a revision to the site calculation for control rod survivability during a LOCA. Specifically, this involved the evaluation of the potential for control rod melt with absorber expulsion during a LOCA. The calculation revision added a detailed evaluation of the potential for control rod melt during a small break loss-of-coolant accident (SBLOCA). An update to the Updated Final Safety Analysis Report (UFSAR) was completed along wi~h the calculation revision.

During the initial licensing of the Davis-Besse Nuclear Power Station (DBNPS), the NRC staff requested a description of the analytical methods and uncertainties associated with the calculations performed to evaluate the potential for control rod melt. The original analysis described in the UFSAR determined temperatures throughout the control rod structures by modeling conductivity, gap conduction, convection, and thermal radiation heat transfer as well as internal heat generation. The new analysis described in the revised calculation used the same basic methodology as the original but revised multiple elements to address identified non-conservatisms in the original methodology. The revised analysis used the RELAP5/MOD2-B&W code.

Summary of Evaluation:

The use of an alternate method rather than the method originally described in the UFSAR that was used in establishing the safety analyses and establishing the design bases meets the definition of an adverse change that must be evaluated under 10 CFR 50.59(c)(2)(viii). The screening determined that the proposed activity involved only a change to a method of evaluation, and the criteria of 10 CFR 50.59(c)(2)(i-vii) are not applicable.

Attachment L-20-119 Page 3 of 5 Nuclear Energy Institute (NEI) 96-07, Revision 1, Guidelines for 10 CFR 50.59 Evaluations, identifies that the use of a methodology revision that is documented as providing results that are essentially the same as, or more conservative than, the previous revision of the same methodology is not considered a departure from a method of evaluation described in the UFSAR. This evaluation demonstrated that the revised analysis methodology for evaluating control rod survivability during a LOCA provides more conservative results than the methodology described in the original DBNPS Final Safety Analysis Report.

Title:

UFSAR Section 18 - Leak Chase Monitoring Program changes necessary following commencement of the Period of Extended Operation.

Activity

Description:

The Leak Chase Monitoring Program is a condition monitoring program, consisting of observations and activities to detect leakage from the spent fuel pool, the fuel transfer pit, and the cask pit liners due to age-related degradation. The Leak Chase Monitoring Program includes periodic monitoring of the spent fuel pool, the fuel transfer pit, and the cask pit liners leak chase system.

Loss of intended function of the spent fuel pool, the fuel transfer pit, and the cask pit liners, that is gross leakage, would not occur at values at or just above the original selected acceptance criteria of 15 milliliters per minute (ml/min). Gross leakage is leakage in excess of makeup capacity. The UFSAR provides a value of makeup from the decay heat removal system to the spent fuel pool from the borated water storage tank by pumped or gravity-fill methods on the order of 70 gallons per minute. The magnitude of leakage being discussed herein is 25 ml/min through the leak chase monitoring lines. Assuming all 21 leak chases were leaking at this acceptance criteria rate, it would only result in a leak rate of 200 gallons per day or 0.14 gallons per minute. The increase in leak rate from the original acceptance criteria of 15 ml/min to the new acceptance criteria being evaluated of 25 ml/min is only 80 gallons per day or 0.06 gallons per minute. This loss is minimal when compared with evaporative losses that are still bounded by the makeup rate on the order of 70 gallons per minute.

The DBNPS UFSAR was revised to change the Leak Chase Monitoring Program criteria for documentation in the corrective action program of measurement of leakage from any leak chase monitoring line from 15 ml/min to 25 ml/min.

Attachment L-20-119 Page 4 of 5 Summary of Evaluation:

The 10 CFR 50.59 Evaluation determined that all design bases functions are met. No accident initiators are altered. The likelihood of equipment malfunction is not increased.

The consequences of accidents or malfunctions of systems, structures, or components are unchanged. No new or different accidents or malfunctions of systems, structures, or components are created. No fission product barrier limits are changed or altered. No methods of evaluation have been changed. Therefore, the proposed revision of the procedure does not meet the criteria of 10 CFR 50.59(c)(2) and prior approval is not required to implement this change.

Title:

Extension of Steam Turbine Valve Test Intervals Activity Description :

This activity implemented the extension of the steam turbine valve test intervals for the stop valves, control valves, and combined intermediate valves. The test interval was extended from quarterly to a maximum of six months. The test interval extension is based on a calculation from MPR Associates Inc., which determined the effect on the turbine missile generation probability. Valve testing is performed using periodic tests (main turbine stop valve test, main turbine control valve test, and the main turbine combined intermediate valve test). This activity only extends the test interval and no other changes were made to the periodic valve test procedures. The UFSAR section for turbine steam flow control was updated to indicate valves are tested closed within a 6-month interval.

Summary of Evaluation:

The 10 CFR 50.59 Evaluation concluded the design function of the overspeed protection system (OPS) was adversely affected by changing the turbine valve testing interval.

Changing to a greater than quarterly test frequency increases the probability of an overspeed event and therefore decreases the reliability of the OPS. Per NEI 96-07, a change that decreases the reliability of a function that could initiate an accident is considered an adverse effect to a design function.

The turbine missile generation probability was calculated as a function of the valve test interval. An analysis was completed which determined DBNPS will remain below the NRC approved value of 4x10-5 per year.

There are no new malfunctions or accidents introduced by this change in test frequency.

In addition, there are no consequences of previously evaluated events affected. All design basis limits for fission products are unaffected because there is no change to any

Attachment L-20-119 Page 5 of 5 assumed failures of mitigating systems. The turbine missile generation description in the UFSAR does not describe a methodology, therefore, the activity does not result in a departure from a method of evaluation described in the UFSAR.