IR 05000260/2024090

NRC Inspection Report 05000260/2024090 and Preliminary White Finding and Apparent Violation - 1

ML24255A027
Person / Time
Site:	Browns Ferry
Issue date:	09/17/2024
From:	Mark Franke Division Reactor Projects II
To:	Jim Barstow Tennessee Valley Authority
References
EA-24-075 IR 2024090
Download: ML24255A027 (16)

Text

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - NRC INSPECTION REPORT 05000260/2024090 AND PRELIMINARY WHITE FINDING AND APPARENT VIOLATION

Dear Jim Barstow:

The enclosed report documents a finding with an associated apparent violation that the U.S.

Nuclear Regulatory Commission (NRC) has preliminarily determined to be of low-to-moderate safety significance (i.e., a White finding). The finding involved an NRC-identified apparent violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, associated with the licensee's failure to identify and correct a condition adverse to quality affecting the integrity of the Unit 2 high pressure coolant injection (HPCI) steam exhaust inner rupture disc. This resulted in the inoperability of the Browns Ferry Unit 2 HPCI system on March 19, 2024. We assessed the significance of the finding using the significance determination process (SDP) and the best available information. The enclosed report includes a detailed risk evaluation with the basis of our preliminary significance determination. We are considering escalated enforcement for the apparent violation consistent with our Enforcement Policy, which can be found at http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html. Because we have not made a final determination, no notice of violation is being issued at this time. Please be aware that further NRC review may prompt us to modify the number and characterization of the apparent violations.

We intend to issue our final significance determination and enforcement decision, in writing, within 90 days from the date of this letter. The NRCs SDP is designed to encourage an open dialogue between your staff and the NRC; however, neither the dialogue nor the written information you provide should affect the timeliness of our final determination.

Before we make a final decision on this matter, we are providing you with an opportunity to (1)

request a Regulatory Conference where you can present to the NRC your perspective on the facts and assumptions the NRC used to arrive at the finding and assess its significance, or (2)

submit your position on the finding to the NRC in writing. If you request a Regulatory Conference, it should be held within 40 days of the receipt of this letter, and we encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effective. If a Regulatory Conference is held, it will be September 17, 2024 open for public observation. If you decide to submit only a written response, such submittal should be sent to the NRC within 40 days of your receipt of this letter.

If you choose to send a response, please include your perspective of the significance of the finding along with the related facts and assumptions used to reach your determination.

Additionally, your response should be clearly marked as "Response to Apparent Violation; (EA-24-075)" and should include for the apparent violation: (1) the reason for the apparent violation or, if contested, the basis for disputing the apparent violation; (2) the corrective steps that have been taken and the results achieved; (3) the corrective steps that will be taken; and (4) the date when full compliance will be achieved. Your response should be submitted under oath or affirmation and may reference or include previously docketed correspondence if the correspondence adequately addresses the required response. Additionally, your response should be sent to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Center, Washington, DC 20555-0001 with a copy to Mr. Louis McKown, Chief, Reactor Projects Branch 5, U.S. Nuclear Regulatory Commission, Region II, within 40 days of the date of this letter. If an adequate response is not received within the time specified or an extension of time has not been granted by the NRC, the NRC will proceed with its enforcement decision or schedule a Regulatory Conference.

If you decline to request a Regulatory Conference or to submit a written response, you relinquish your right to appeal the final SDP determination, in that by not doing either, you fail to meet the appeal requirements stated in the Prerequisite and Limitation sections of Attachment 2 of NRC Inspection Manual Chapter 0609.

Please contact Louis J. McKown at 404-997-4545, and in writing, within ten (10) calendar days from the issue date of this letter to notify the NRC of your intentions. If we have not heard from you within ten calendar days, we will continue with our significance determination and enforcement decision. The final resolution of this matter will be conveyed in separate correspondence.

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, Mark E. Franke, Director Division of Reactor Projects Docket No. 05000260 License No. DPR-52

Enclosure:

As stated

Inspection Report

Docket Number: 05000260

License Number: DPR-52

Report Number: 05000260/2024090

Licensee: Tennessee Valley Authority

Facility: Browns Ferry Nuclear Plant

Location: Athens, Alabama

Inspection Dates: August 29, 2024 to October 31, 2024

Inspectors: S. Billups, Resident Inspector K. Pfeil, Resident Inspector A. Rosebrook, Senior Reactor Analyst T. Steadham, Senior Resident Inspector

Approved By: Mark E. Franke, Director Division of Reactor Projects

Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a NRC inspection at Browns Ferry Nuclear Plant, 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

Browns Ferry Unit 2 HPCI Rupture Disc Failure Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Preliminary White [H.9] - Training 71152A Systems AV 05000260/2024090-01 Open EA-24-075 The inspectors identified a Preliminary White finding and associated Apparent Violation (AV)of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly identify and correct a degraded condition with the Unit 2 high pressure coolant injection (HPCI) turbine exhaust inner rupture disc after pressurizing the HPCI steam exhaust line. During surveillance testing on March 19, 2024, this resulted in the unplanned isolation of HPCI and the discovery of the loss of the associated safety function.

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.

OTHER ACTIVITIES - BASELINE

71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)

The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:

(1) CR 1742531, high pressure coolant injection steam exhaust pressure rose to 160 psig

INSPECTION RESULTS

Browns Ferry Unit 2 HPCI Rupture Disc Failure Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Preliminary White [H.9] - Training 71152A Systems AV 05000260/2024090-01 Open EA-24-075 The inspectors identified a Preliminary White finding and associated AV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly identify and correct a degraded condition with the Unit 2 HPCI turbine exhaust inner rupture disc after pressurizing the HPCI steam exhaust line. During surveillance testing on March 19, 2024, this resulted in the unplanned isolation of HPCI and the discovery of the loss of the associated safety function.

Description:

On December 14, 2021, the licensee removed the Unit 2 HPCI system from service for a pre-planned system maintenance outage. Later that day, the licensee replaced 2-RPD-073-0729, Unit 2 HPCI turbine exhaust inner rupture disc, under work order 121811897. This rupture disc, a Fike model CPV-C LL BT, was procured under purchase order 6909657 and received onsite on November 22, 2021. Included with the rupture discs was the Fike certificate of conformance which included both the stamped burst pressure, 166.25 psig, and burst tolerance. It should be noted that rupture discs will not burst at a specific pressure but at some value in a tolerance range about their stamped burst pressure based upon their quality of manufacture. Because these rupture discs were procured to meet the testing requirements of ASME Sect. III, Subsection NB-7600, each lot of rupture discs had a burst tolerance of +/- 5% of the stamped burst pressure. Therefore, the specific rupture disc installed for 2-RPD-073-0729 on December 14, 2021, would be expected to burst anywhere between 157.94 psig and 174.56 psig at 378 ºF.

On December 15, 2021, operators tagged open 2-FCV-73-16, Unit 2 HPCI steam supply valve, to perform preventative maintenance on the valve. When the valve was opened, the pressure in the HPCI turbine steam exhaust line rose to approximately 160 psig. Operators documented the issue as CR 1742531 and screened the issue as not potentially affecting operability on December 17, 2021. Following maintenance, operators restored HPCI to operable on December 18, 2021.

In reviewing the effects of this pressurization, engineering documented the following discussion in CR 1742531 on December 23, 2021:

Per ASME B31.1 the test pressure is 1.5 times system design pressure. The CR states 160 psi is the maximum system pressure seen during this event with a design pressure of 150 psi. Operating at 1.07 times the design pressure is well below the test pressure and is well below the overpressure protection setpoint of the rupture disc at 175 psi. This event did not challenge the HPCI piping related to this condition report.

Based on this engineering input, the licensee determined that this issue did not represent a condition adverse to quality and closed the CR on December 28, 2021, with no actions taken. However, as described below, the licensee was unaware that this pressure transient damaged the rupture disc because the engineering input failed to adequately evaluate the effects of the transient. The licensee continued to perform routine quarterly surveillance flow tests on Unit 2 HPCI without any apparent impact. During these routine tests, steam line exhaust pressure is normally approximately 45 psig.

On December 22, 2023, Unit 2 HPCI successfully completed a quarterly surveillance flow test. The next time that the pump was operated was for the subsequent quarterly flow test on March 19, 2024. Approximately 37 minutes into that test, the inboard HPCI turbine steam exhaust rupture disc (e.g., the rupture disc) failed causing an automatic isolation of steam to the pump turbine. This failure was documented as CR 1917980.

The licensee sent the rupture disc to a third-party vendor for failure evaluation who concluded that the membrane failed due to cyclic fatigue. Upon disassembly, the rupture disc membrane had a crumpled appearance, a term commonly referred to as turtle-backing. The crumpling action was caused by repeated cyclic pressurization and vacuum conditions in the exhaust line causing the membrane to expand and contract without adequate structural support for the membrane. The vendor measured a gap of approximately 0.7 inches between the membrane and the vacuum support.

Rupture discs of the type used at Browns Ferry in HPCI exhaust lines have a vacuum support intended to provide structural rigidity for the relatively thin rupture membrane to prevent premature failure due to cyclic fatigue. The vacuum support is intended to be snug against the membrane to provide that support. With the observed gap, the membrane flexed every time a vacuum formed in the steam exhaust line. Every time the HPCI pump was secured, such as after each routine surveillance test, the steam exhaust line would briefly form a vacuum as the steam condensed within the exhaust line which caused the membrane to flex. This flexing further deformed the membrane which formed stress lines in the approximately 0.01 thick membrane. This condition continued to weaken the membrane and lower the burst pressure each time the HPCI pump ran.

Following the failure on March 19, 2024, the licensee consulted with the rupture disc manufacturer, Fike Corporation, for insight. Based on manufacturing documents, the rupture disc was originally manufactured with a 3.125 bulge (e.g. the cap) but the third-party vendor measured the cap at 3.84. Fike determined that the only physical explanation for the growth in the height of the cap was rupture disc exposure to either a high temperature approaching 900 ºF, which was not credible, or a pressure greater than the published rupture disc operating ratio of 70%, in this case, 116.38 psig.

Operating ratio is a common industry parameter that represents the maximum pressure that a rupture disc can experience without damage. Exceeding the operating ratio will permanently deform the rupture disc membrane and will result in both reducing the actual burst pressure and increasing the risk of premature failure. Consequently, the manufacturer recommends replacing the rupture disc if the operating ratio is exceeded. Further, Fike provided that once turtle-backing started, failure could be expected to occur within approximately 10 to 12 cycles. At the time of failure on December 22, 2023, the inner rupture disc had already experienced 11 cycles since the event on December 15, 2021.

In their equipment failure evaluation for CR 1917980, the licensee examined HPCI steam exhaust pressures since December 16, 2021. This was the date that work order 121811897 was closed in the work control system. Based upon this review, the licensee determined that the rupture disc was never exposed to a pressure above 45 psig since being installed. Hence, the licensee concluded that the rupture disc was defective upon receipt and submitted a 10CFR50 Part 21 notification on July 25, 2024, (ENS event number 57245).

During the inspectors review of the issue, the inspectors discovered CR 1742531. This examination of the steam line exhaust pressures found that the rupture disc was exposed to a maximum of 161.4 psig for 54 seconds and exceeded the operating ratio for about 12.5 minutes. Based upon the inspectors discovery, the licensee determined that the event on December 15, 2021, damaged the rupture disc membrane which directly led to the March 19, 2024, failure. The licensee retracted their Part 21 notification on August 22, 2024. The licensee upgraded the significance level of CR 1917980 to a Level 1 and initiated a root cause investigation team on August 27, 2024, to perform a more comprehensive review of the events and circumstances surrounding the rupture disc failure.

The inspectors concluded that the evaluation of CR 1742531, associated with the December 15, 2021, pressurization event, applied engineering judgement that was not applicable to rupture discs. The licensee's discussion failed to acknowledge that the observed pressure was within the certified burst tolerance for the installed rupture disc, rendering moot the comparison to the 174 psig design specified rupture pressure. The inspectors discovered that the rupture disc exceeded both the operating ratio, a published parameter for the rupture disc selected by the licensee, and lower bound of burst pressure, provided by the manufacturer at receipt.

The inspectors determined that a knowledgeable engineer experienced in the application of rupture discs would be aware of the impact of exceeding both the operating ratio and the lower bound of burst pressure when developing their technical judgement. Therefore, the inspectors considered the licensees assessment of the impact of the December 15, 2021, pressurization event of the Browns Ferry Unit 2 HPCI inner rupture disc using engineering judgement to be a preventable error.

Corrective Actions: The licensee replaced both the inner and outer HPCI rupture discs on all three Units and initiated a root cause evaluation to evaluate the issue.

Corrective Action References: CRs 1742531, 1917980, 1920437, 1926391, 1926396, 1935939, 1936298, 1952648, 1952845.

Performance Assessment:

Performance Deficiency: The licensees failure to promptly identify and correct a degraded condition with the Unit 2 HPCI turbine exhaust inner rupture disc after pressurizing the HPCI steam exhaust line was a performance deficiency reasonably within their ability to foresee and prevent. Specifically, the licensee failed to identify that pressurizing the line beyond the rupture disc operating ratio, and as high as the lower tolerance of burst pressure, damaged the rupture disc membrane. This pressurization permanently deformed the membrane, reduced its burst pressure, and shortened its lifespan with each successive demand on the system. Consequently, the licensee failed to promptly correct the degraded condition as evidenced by the rupture disc failure on March 19, 2024.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the condition would have prevented the Unit 2 HPCI system from performing its intended safety function.

Significance: The inspectors assessed the significance of the finding using a Detailed Risk Evaluation. The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The affected cornerstone was Mitigating Systems, as determined by IMC 0609, Attachment 4, Initial Characterization of Findings. The inspectors screened the performance deficiency using Exhibit 2 of Appendix A and determined a detailed risk evaluation was required because the degraded condition represented a loss of the PRA function of a single train TS for greater than its TS allowed outage time

RISK ANALYSIS/CONSIDERATIONS A regional Senior Reactor Analyst (SRA) performed a risk assessment in accordance with NRC Inspection Manual Chapter 0609 Appendix A. The SRA used SAPHIRE 8 version 8.2.10 and the Browns Ferry Unit 2 SPAR model version 8.82 dated August 14, 2023. The Browns Ferry SPAR models do not contain sequences for fires or internal flooding. Since the licensee has an NFPA 805 fire PRA and modeled internal flooding and seismic in their CAFTA plant risk model, the licensees results for fire, internal flooding and seismic were considered to be best available information.

ASSUMPTIONS 1. The rupture disc failure was due to cyclic fatigue.

2. A cycle is the rupture disc seeing positive pressure while running and a vacuum after the HPCI turbine is shutdown.

3. An unsupported rupture disc will crumple when it goes thru a vacuum cycle

4. When the disc crumples it shows a turtleback pattern on the disc as seen in the lab

pictures.

5. The crumple lines are stress concentrators and once the disc crumples initially, vender data predicts it will fail within 10-12 additional cycles.

6. The Unit 2 Rupture disc failed on its 12th cycle since installation.

7. The Unit 2 Rupture disc was found to have an excessive clearance between the disc and the vacuum support allowing the disc to flex each cycle and crumple likely in the first few cycles, and the disc failed as predicted.

8. There are 4 pressure sensors between the inner and outer rupture discs which send a signal to isolate the steam admission valves making HPCI Inoperable.

9. HPCI failure is modeled as a Failure to Start since the failure occurred in the first hour of operation. Demand based not run time based.

10.Once the cycle is completed no additional stresses are applied until the HPCI turbine is started. Thus, the condition was present the entire standby period and T is used for exposure time vice T/2.

11.Per the Risk Assessment of Operational Events (RASP) Manual, repair time is also added. The rupture disc was replaced, and post maintenance test was completed on March 23, 2024, while HPCI was not officially declared operable for a few additional days. The SRA considered the repair time to be complete days since that was when the condition caused by the performance deficiency (PD) was repaired.

12.The SRA considered the Unit 2 SPAR model to be the best available model for Internal Events, Tornado/Hurricane/High Winds. However, the SPAR model does not have any Fire sequences, limited flooding sequences, and more consistent Seismic sequence results. Therefore, the licenses CAFTA model is considered best available information for Fire, Flooding, and Seismic.

13.In 2022, a White Notice of Violation (NOV) for a Unit 1 HPCI issue used the licensees model as best available information for Fire, Flooding, and Seismic and this approach was approved by the SERP.

14.The SRA reviewed the licensees results and compared them with the 2022 Unit 1 results and ensured all major differences were understood and had adequate technical basis.

15.Emergency High Pressure Makeup pumps are included in the SPAR model in the appropriate fault tress.

16.FLEX Credit is applied. However, FLEX will have minimal mitigation for most of the dominant accident sequences due to limited time to core damage.

17.Because the steam lines automatically isolate on a rupture disc failure, any recovery or reset of the isolation after the pressure downstream of the rupture discs decayed to below the pressure switch setpoint would not have been sustainable. This is because the pressure switches would have continued to actuate and isolate the steam line if the system was restarted because the rupture disc membrane had failed.

18.Exposure Time: December 22, 2023, through March 23, 2024.

Because the failure was cyclic in nature, the last time that the rupture disc was subjected to cyclic stresses would have been when the turbine was shut down after the surveillance run on December 22, 2023. The membrane would have flexed backwards when a vacuum formed in the steam line. The next time the turbine would have operated would have resulted in another cyclic flex of the membrane thereby weakening the membrane such that it would have ruptured prematurely. This is demonstrably shown when the membrane failed approximately 40 minutes into the surveillance run on March 19, 2024. After repairs were completed, and tags removed, the system was available on March 23, 2024. The operability run was completed two days later which was successful. Therefore, the end date of the exposure time was when the system was restored to service on March 23, 2024.

NRC SPAR MODEL RESULTS The SRA modeled the failure as a failure of the HPCI Steam Admission valve to open, since when the inner rupture disc fails the pressure sensors between the inner and outer disc with send a signal to isolate the HPCI Steam supply. The SRA set HCI-MOV-CC-F016, HCPI Steam Supply Valve F016 Fails to Open, to True. The SRA also applied FLEX credit and set exposure time to 92 days as discussed above. The dominant accident sequence is a Stuck OPEN Primary Relief valve, failure of all high pressure injection sources, and operators fail to manually depressurize the plant. Change in CDF was 1.92 E-6.

LICENSEES PRA EVALUATION RESULTS The incremental conditional core damage probability (ICCDP) delta and incremental conditional large early release probability (ICLERP) delta results in Table 2-1 imply an NRC Regulatory Oversight Program Significance Determination Process (SDP) color of White for BFN Unit 2. Table 2-1: Unit 2 HPCI Unavailable Results.

UNCERTAINTY ANALYSIS The NRCs SPAR model for Browns Ferry does not Include Internal Fire or Internal Flooding sequences. As a result, the licensees CAFTA model was used and considered best available information for Internal Fire, Internal Flooding and Seismic sequence results. These results were qualitatively compared with information from risk information provided in Licensee Amendment requests (such as the NFPA-805 LAR) and SPAR model data from other NFPA-805 BWR plants.

For internal event results of the licensees CAFTA model for Unit 2, the CAFTA model results for Unit 1 in 2022 were adjusted for exposure time. The NRC SPAR model versions 8.82 and 8.61 were all compared and considered.

CAFTA U2: 2.17 E-6 CAFTA U1: 3.8 E-6 SPAR Version 8.61: 3.14E-6 SPAR Version 8.82: 1.92 E-6

UNCERTAINTY PLOT

SENSITIVITY EVALUATIONS Because the licensee data is being used as best available information, it is not feasible to do a sensitivity analysis as internal fire and internal flooding are 2 of the 3 dominant initiators. However, for internal events, the following sensitivities were done:

1. FLEX credit vs. no FLEX credit: There was minimal difference in risk when FLEX credit was removed or applied, 1.92 E-6 vs 1.92 E-6. Total risk is between 6.25 E-6 and 6.48 E-6.

2. Contributions from external events: As demonstrated by Table 2-1 above, fire is the dominant risk contributor for a HPCI failure.

3. Potential risk contribution from large early release frequency (LERF): The licensee CAFTA model was considered the best available information for LERF and was calculated to be 3.95 E-7 which corresponds to a Low-to-moderate Safety Significance finding as well.

CONCLUSION Preliminary characterization of this issue is low-to-moderate safety significance (WHITE) due to change in core damage frequency being approximately 6.5 e-6 and change in LERF being approximately 3.95 E-7.

Cross-Cutting Aspect: H.9 - Training: The organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. The engineer who provided the engineering discussion on the effects of the over pressurization event was not sufficiently knowledgeable in the design concepts associated with rupture discs, such as operating ratio, to make an engineering judgement on the impact on the rupture disc. The engineer compared the maximum pressure seen to the design rupture disc burst pressure instead of the published operating ratio for the disc.

Enforcement:

Violation: 10 CFR 50, Appendix B, Criterion XVI requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.

TVA-NQA-PLN89-A, Nuclear Quality Assurance Plan (NQAP), Revision 39, Section 10.2.2 states, in part, that TVA Nuclear and onsite non-nuclear service organizations performing quality-related activities at nuclear facilities shall promptly identify and resolve conditions adverse to quality.

NPG-SPP-22.300, Corrective Action Program, Revision 23, Section 5 defines that a condition adverse to quality is a condition associated with a structure, system, component or program that is in-scope of the NQAP. The Unit 2 HPCI inner rupture disc, 2-RPD-073-0729, is a safety-related component and is in the scope of the NQAP.

Browns Ferry Nuclear Plant, Unit 2 TS LCO 3.5.1, requires, in part, that the HPCI system shall be operable while in Mode 1, 2, and 3, except when the reactor steam dome pressure is less than or equal to 150 psig. If the HPCI system is inoperable in an applicable Mode, TS 3.5.1, Condition C, requires restoring the system to operable status within 14 days or in accordance with the Risk Informed Completion Time Program. If the required actions for Condition C are not met within the established completion time, Condition G requires to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to reduce reactor steam dome pressure to less than or equal to 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Browns Ferry Nuclear Plant, Unit 2 TS LCO 3.0.4, requires, in part, that when an LCO is not met, entry into a mode or other specified condition in the applicability shall only be made when the associated actions to be entered permit continued operation in the mode or other specified condition in the applicability for an unlimited period of time.

Contrary to the above, on December 17, 2021, the licensee failed to establish measures to assure that a condition adverse to quality associated with 2-RPD-073-0729, the Unit 2 HPCI pump turbine exhaust inner rupture disc, was promptly identified and corrected. Specifically, the licensee failed to identify and correct the damage done to the 2-RPD-073-0729 rupture disc as a result of a steam exhaust line pressurization event on December 15, 2021. This damage rendered the Unit 2 HPCI system incapable of performing its intended safety function as reflected by the rupture disc failure on March 19, 2024, during a routine surveillance test.

Because the licensee did not recognize the inoperability of the HPCI system due to the pressurization event, they failed to declare the HPCI system inoperable and complete the required actions in TS 3.5.1, Conditions C and G, within the established completion times, and the required actions of TS LCO 3.0.4.

Enforcement Action: This violation is being treated as an apparent violation pending a final significance (enforcement) determination.

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

• On September 9, 2024, the inspectors presented the NRC inspection results to Daniel Komm, Browns Ferry Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71152A Corrective Action BFPER 980479

Documents

71152A Corrective Action CRs 1742531,

Documents 1917980,

20437,

26391,

26396,

1935939,

1936298,

1952648,

1952845

71152A Drawings 2-47E812-1 Flow Diagram Unit 2 High Pressure Coolant Injection SystemRev. 48

71152A Drawings D3563-1 Fike Rupture Disc Model 16 CPV-C LL Rev. B

though D

71152A Engineering 020306-Procurement Engineering Evaluation to Add Dedication Plan 12/13/2016

Evaluations CDL059JM0 Approval

71152A Engineering 111332 Review of Third Party Vendor Commercial Grade Dedication 09/07/2021

Evaluations Plan for Rupture Discs per TVA PO 6909657

71152A Engineering 9700058229M0 Procurement Engineering Evaluation for Commercial Grade 09/01/1998

Evaluations Dedication of Fike HPCI Rupture Disc

71152A Engineering CGT078W Procurement Engineering Data Sheet for High Pressure 12/13/2016

Evaluations Coolant Injection Rupture Disc

71152A Miscellaneous CGD-CPV-C-BT Commercial Grade Item Dedication Technical Evaluation and 0 through 1

Test Plan

71152A Miscellaneous L-2024-005 Laboratory Analysis of Browns Ferry Nuclear Station High April 2024

Pressure Coolant Injection Rupture Disc

71152A Miscellaneous PQ003884 Vendor Quote for TVA Request for Quote no. 1740103 06/22/2021

71152A Miscellaneous Purchase Orders

90652, 568538,

238263,

288622,

24454,

Inspection Type Designation Description or Title Revision or

Procedure Date

6909657

71152A Miscellaneous RFQ 1076669 Request for Quote for Fike rupture disc for purchase

requisition 1175715

71152A Miscellaneous Selected December

Dataware plots of 15, 2021

Unit 2 HPCI

steam exhaust

pressure

71152A Miscellaneous Selected main December

control room 14, 2021,

operator logs through

December

18, 2021

71152A Procedures NEDP-8.0 Evaluations for Procurement of Materials, Items, and Rev. 7

Services

71152A Shipping Records 6909657 Shipping and Receipt Documentation for Fike Rupture Discs 11/22/2021

for Purchase Order 6909657

71152A Work Orders 118663321,

118803207,

118803211,

21604907,

21664096,

21811897,

21811901

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