05000331/LER-2016-003, Regarding Main Steam Isolation Valve Leakage Exceeded Technical Specification Limit

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Regarding Main Steam Isolation Valve Leakage Exceeded Technical Specification Limit
ML16342C569
Person / Time
Site: Duane Arnold NextEra Energy icon.png
Issue date: 12/06/2016
From: Vehec T
NextEra Energy Duane Arnold
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NG-16-0225 LER 16-003-00
Download: ML16342C569 (5)
v • d • e


LER-2016-003, Regarding Main Steam Isolation Valve Leakage Exceeded Technical Specification Limit
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(i)
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(ix)(A)
10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.73(a)(2)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3312016003R00 - NRC Website
v • d • e

text

December 6, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Duane Arnold Energy Center Docket 50-331 Renewed Op. License No. DPR-49 Licensee Event Report 2016-003 NG-16-0225 10 CFR 50.73 Please find attached the subject report submitted in accordance with 10 CFR 50.73. This letter makes no new commitments or changes to any existing

commitments

~~~

~,/F. A. Vehec

/

Vice President, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC cc:

Administrator, Region Ill, USNRC Project Manager, DAEC, USNRC Resident Inspector, DAEC, USNRC NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324

NRC FORM366 (02-2014)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0113112017

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......! LICENSEE EVENT REPORT (LER)

(See Page 2 for required number of digits/characters for each block)

1. FACILITY NAME Duane Arnold Energy Center
4. TITLE Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.

Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget Washington, DC 20503. If a means used to impose an information collection does not display a currenHy valid OMB control number, the NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.

2. DOCKET NUMBER
3. PAGE 05000-331 1 OF 4 Main Steam Isolation Valve Leakage Exceeded Technical Specification Limit
5. EVENT DATE
6. LER NUMBER
7. REPORT DATE
8. OTHER FACILITIES INVOLVED F AGILITY NAME YEAR N/A DOCKET NUMBER N/A YEAR MONTH DAY YEAR I SEQUENTIAL I REV MONTH NUMBER NO.

DAY FACILITY NAME DOCKET NUMBER 10 18 2015 2016 -

003

- 00 12 06 2016 N/A N/A
9. OPERA TING MODE
11. THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

D 20.2201 (b}

D 20.2203(a)(3)(i)

D 50.73(a)(2)(i)(C)

D 50.73(a)(2)(vii) 5 D 20.2201 <d>

D 20.2203<a><3><H>

D 50. 73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2203<a><1>

D 20.2203<a><4>

D 50. 73(a)(2)(ii)(B)

D

50. 73(a)(2)(viii)(B)

D 20.2203<a><2>

D 50.36(c)(1 )(i)(A)

D 5o. 13(a><2><m>

D 50.73(a)(2)(ix)(A)

10. POWER LEVEL D 20.2203<a><2><H>

D 50.36(c)(1 )(ii)(A)

D 50.73(a)(2)(iv)(A)

D

50. 73(a)(2)(x)

D 20.2203<a><2><rn>

D 5o.3s<c><2>

D 50: 73(a)(2)(v)(A)

D

13. 11 <a><4>

0%

D 20.2203(a)(2)(iv)

D 5o.4s(a)(3)(ii)

D 50.73(a)(2)(v)(B)

D

13. 11 <a><5>

D 20.2203(a)(2)(v)

D

50. 73(a)(2)(i)(A)

D 50.73(a)(2)(v)(C)

D OTHER D 20.2203(a)(2)(vi)

(gJ 50. 73(a)(2)(i)(B)

D 50.73(a)(Z)(v)(D)

Specify in Abstract below or in I.

Description of Event

On October 18, 2016, with the unit shutdown for a planned refueling outage (Mode 5, Refueling, 0% power), an evaluation of data from the scheduled Main Steam Line Isolation Valve (MSIV) (System Code SB) and Main Steam Line Drain valve penetration Local Leak Rate Testing (LLRT) determined the 'as found' maximum pathway leakage for the 'B' Inboard MSIV, CV-4415, and the Outboard Main Steam Line Drain valve, M0-4424, in excess of the Technical Specification (TS) 3.6.1.3 leakage limit of s 100 scfh for a single MSIV and s 200 scfh for combined pathway leakage.

II.

Assessment of Safety Consequences

Reportability Evaluation CV-4415 'as found' leakage was measured at 253.8 scfh. TS Surveillance (SR) 3.6.1.3.9 requires verification of leakage rate through each MSIV is s 100 scfh and that the combined maximum pathway leakage rate for all four main steam lines is :5 200 scfh when tested at~ 24 psig. Failure to meet this SR requires entry into LCO 3.6.1.3.9, Condition D, restore leakage to within limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or 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 />. The as found results exceeded the TS SR limit for individual and combined MSIV leakage and is reportable as a condition prohibited by TS and requires a Licensed Event Report (LER) to be submitted under 10CFR50.73(a)(2)(i)(B).

M0-4424 'as found' leakage was measured at 710 scfh. As stated in Updated Final Safety Analysis Report (UFSAR) Section 6.7.1, the total main steam pathway, which includes the four main steam lines and the inboard MSIV Drain Line, is limited to 200 scfh (i.e., the main steam line drains are included in the total leakage limit for all MSIVs). Therefore, the Main Steam Line Drains valve LLRT failures are included in this LER.

Safety Significance

The Main Steam system conducts steam from the reactor vessel through the primary containment to the main turbine. In performing this function, it serves as part of the primary boundary to prevent radioactive release to the surrounding environment.

The MSIV Leakage Treatment Path is designed to mitigate the release of fission products following a Loss of Coolant Accident (LOCA). This is accomplished by directing any leakage past closed MSIVs and closed Main Steam drain line isolation valves to the main condenser via the Main Steam drain line. The volume and surface area of the condenser provides holdup time and plate-out surface for fission products.

Pathways to other systems connected to the Main Steam system are isolated to ensure leakage is processed through this path.

The NRC has previously assessed the consequences of BWR main steam line releases. In a paper published in 2002 (PRAB-02-01, "Assessment of BWR Main Steam Line Release Consequences"), the NRC determined an MSIV leakage rate that would be considered risk-significant and would warrant further investigation. This assessment is conservative since benefits of holdup time, plate-out, and gravitational settling are ignored. Given a greater than 20 factor of conservatism, the report concludes there should be no regulatory concern if the leakage past the best sealing valve in any single main steam line is less than 10,000 scfh. This result is reflected in the NRC's Inspection Manual Chapter 0609, Appendix H where a simple probabilistic risk framework is provided for use in identifying potentially significant findings related to containment integrity.

The as-found leakage rate in both the 'B' Main Steam line and in the Main Steam Line Drain line is well below the 10,000 scfh criterion. Therefore, this event was of low safety significance and had no impact on public health or safety.

This event did not result in a SSFF.

Ill.

Cause of Event

A Root Cause Evaluation (RCE) was completed. The RCE determined that the cause for the 'B' MSIV failure was failure to perform periodic internal inspections of the MSIVs and the cause for the outboard MSL Drain valve was a non-optimal valve design for steam line drain application.

IV.

Corrective Actions

Immediate Corrective Action

CV-4415 internals were re-machined to restore its leakage to below TS limits.

The MSL Drain penetration pathway was isolated in accordance with TS 3.6.1.3.

Corrective Actions for Cause of Event REV NO.

00 4

3. PAGE OF The frequency of internal inspections of MS IVs will be increased to once every 10 refueling cycles.

Actions are planned to replace the MSL Drain valves with a new valve design.

V.

Additional Information

Previous Similar Occurrences:

A review of NextEra Energy Duane Arnold LERs from the previous 10 years found no other instances of events related to Main Steam Lines failing "as found" LLRT testing.

EllS System and Component Codes:

SB - Main Steam System

Reporting Requirements

This activity is being reported pursuant to the requirements of 10CFR50.73(a)(2)(i)(B). 4

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