05000263/LER-2010-001, Re Missed Safety Relief Valve Lift Test Surveillance Interval
| ML101690375 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 06/18/2010 |
| From: | O'Connor T Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-10-039 LER 10-001-00 | |
| Download: ML101690375 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2632010001R00 - NRC Website | |
text
June 18,2010 Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 L-MT-10-039 10 CFR 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed License No. DPR-22 LER 201 0-001, "Missed Safety Relief Valve Lift Test Surveillance Interval" The Licensee Event Report (LER) for this occurrence is attached.
Summary of Commitments itments and no revisions to existing commitments.
clear Generating Plant Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC 2807 West County Road 75 e Monticello, Minnesota 55362-9637 Telephone: 763-295-5151
- Fax: 763-295-1454
LICENSEE EVENT REPORT (LER)
While performing a review of operating experience station personnel identified that the plant may have exceeded the American Society of Mechanical Engineers (ASME) Code required testing interval for five of eight of the Safety Relief Valves (SRVs) [RV]. The apparent noncompliance with the ASME Code was validated by the engineering staff on 4/20/2010 and the Shift Manager subsequently declared a missed surveillance and entered SR 3.0.3. A risk assessment was performed as required by SR 3.0.3, which determined that the risk significance was minimal.
The cause of the event was an interpretation of the test interval requirements that resulted in the plant utilizing an install to test interval vice a test to test interval for new and refurbished Corrective actions taken or planned are: a one-time relief request to go beyond the 5 year testing interval has been submitted, revise the Corporate Directive regarding the testing of SRVs and to develop guidance regarding the process to use and evaluate code
LICENSEE EVENT REPORT (LER)
TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
Event Description
While performing a review of operating experience station personnel identified that the plant may have exceeded the American Society of Mechanical Engineers (ASME) Code required testing interval for five of eight of the Safety Relief Valves (SRVs) [RV]. The apparent noncompliance with the ASME Code was validated by the engineering staff on 4/20/2010 and the Shift Manager subsequently declared a missed surveillance and entered SR 3.0.3. A risk assessment was performed as required by SR 3.0.3, which determined that the risk significance was minimal.
Upon review of the SRV test history, it was determined that five of eight SRVs had exceeded the allowable five year test interval based on a test to test interval. Up to this point, the station had based their testing interval on an installation to test interval, which did not count the time a refurbished valve was stored in a warehouse prior to installation. In March 2005, a Corporate Directive had been revised to document how the station would address ASME OM Code interpretation 01-1 8, related to SRV test intervals. The directive provided a distinction between new and refurbished valves and any valves which were in-service. A brand new valve or a refurbished valve was not considered an in-service component and therefore was not subject to the requirements of the ASME code until it was installed.
Event Analysis
The event is reportable to the NRC under 10 CFR 50.73(a)(2)(i)(B) - Operation or Condition Prohibited by Technical Specifications. There was no required 50.72 notification for this event; however a Licensee Event Report is required.
Safety Significance
There were no nuclear, radiological or industrial safety significant consequences related to The Monticello risk assessment group reviewed the event for risk impact. The failure probabilities for the five SRVs were adjusted to correspond to a ten year test interval. The Monticello PRA model was then re-quantified to determine the impact on the Core Damage Frequency (CDF) and the Large Early Release Frequency (LERF). The method of calculation contained the following conservative assumptions:
e Although the surveillance test is designed to test only the setpoint of the SRVs, this risk assessment conservatively elevated all possible SRV failures that could lead to failure (9-2007)
U.S. NUCLEAR REGULATORY COMMlSSlOP LICENSEE EVENT REPORT (LER)
TEXT (If more space is required, use additional copies of NRC Form 366AX17) of the SRV to open, regardless of pressure setpoint. Failure of the valve to open from a remote electro-pneumatic signal (Automatic Depressurization System [BF], Low Level Setpoint or manual signal) was impacted in this assessment.
0 This assessment assumed the surveillance interval to be ten years versus the nominal interval of five years. The actual interval will be shorter in every case.
0 The common cause failure events impacting all eight of the SRVs were adjusted similarly to the individual five SRV failure rates.
In conclusion, the risk impact of extending the surveillance interval to ten years is very low.
The impact on CDF and LERF was calculated to be very low (< I
.O E-1 Olyr and < I.0 E-I llyr, respectively). The Incremental Core Damage Probability (ICDP) and Incremental Large Early Release Probability (ILERP) limits of I
.O E-06 and 1.0 E-07 respectively are not exceeded, uith several orders of magnitude of margin.
Cause
The cause of the event was an improper interpretation of this testing code requirement.
Corrective Action
The following corrective actions have been taken or are being tracked under ARO1228141:
1 Submitted a one time request for relief from the 5 year test interval.
- 2. Revise the Corporate Directive 5.5, lnservice Testing, to align the technical position for SRV test intervals with ASME OM Code Interpretation 01-1 8.
- 3. Develop formal fleet guidance related to using and evaluating code interpretations.
Failed Component Identification None
Previous Similar Events
There were no previous similar events identified by the investigation.
VRC FORM 366A (9-2007)