05000306/LER-2003-001, Exceeded Tech Spec Completion Time
| ML031410027 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/13/2003 |
| From: | Solymossy J Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-03-050 LER 03-001-00 | |
| Download: ML031410027 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 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)(A), Completion of TS Shutdown |
| 3062003001R00 - NRC Website | |
text
Praire Island Nuclear Generating Plant Commltedto Nuclear Exclence' Operated by Nuclear Management Company, LLC L-PI-03-050 May 13, 2003 10 CFR 50.73 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET 50-306 LICENSE NO. DPR-60 LER 2-03-01: Unit 2-Exceeded Technical Specification Completion Time The Licensee Event Report for this occurrence is attached.
In our initial assessment of this occurrence, Nuclear Management Company (NMC) had reached a conclusion that this issue was not reportable per 10 CFR 50.73 on the basis that the "A" Train of the Cooling Water System was "Operable but Degraded" and capable of performing its intended safety function. However, based on a review of NRC Inspection Manual Part 9900 and Generic Letter 90-05; "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping", and discussions with NRC personnel, NMC has concluded that some ambiguity exists in this guidance regarding Operability of piping with minor leakage. Since these ongoing discussions with the NRC have not reached a final conclusion regarding this issue, NMC has decided to submit the LER per the 60 day requirement of 1 OCFR 50.73. Should the discussions with the NRC conclude that the "A" Train of the Cooling Water System was Operable and should not have been reported, a supplement to this LER will be issued.
In the report, NMC has made no new NRC commitments. Please contact Bob Alexander (651-388-1121) if you have any questions related to this letter.
<g eph M. Solymo Site Vice President, Prairie Island Nuclear Generating Plant CC Regional Administrator, USNRC, Region III Project Manager, Prairie Island Nuclear Generating Plant, USNRC, NRR NRC Resident Inspector - Prairie Island Nuclear Generating Plant Glenn Wilson, State of Minnesota Attachment 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 6-30-2001 (1-2001)
COMMISSION
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3) 05000 306 Prairie Island Nuclear Generating Plant Unit 2 1 OF 5
TITLE (4)
Exceeded Technical Specification Completion Time EVENT DATE 5)
LER NUMBER (6)
REPORT DATE 7)
OTHER FACILITIES INVOLVED 8)
SEQUENTAL REV FACILITY NAME DOCKET NUMBER MO DAY YEAR YEAR NUMBER NO MO DAY YEAR None 03 1 4 03 03 01 00 0
3 FACILITY NAME DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CF (Ct leck all that apply) (11)
MODE (9) 20.2201(b) 1 20.2203(a)(3)(ii) 150.73(a)(2)(ii)(B) 50.73(a)(2)(ix)(A)
POWER 100 20.2201 (d) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x)
LEVEL (10) 1L" 20.2203(a)(1) 50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 73.71 (a)(4) 20.2203(a)(2)(i) 50.36(c)(1)(ii)(A) 50.73(a)(2)(v)(A) 73.71 (a)(5) 20.2203(a)(2)(ii) 50.36(c)(2) 50.73(a)(2)(v)(B)
OTHER 20.2203(a)(2)(iii) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C)
Specify In Abstract below or in NRC 20.2203(a)(2)(iv) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(D)
(if more space is required, use additional copies of (If more space is required, use additional copies of NRC Forrn 366A) (17)
Risk Siqnificance The pinhole leak occurred on the 3/4" line to Relief Valve (RV) on the cooling water side of the heat exchanger. An assessment was completed by Engineering concerning the potential impacts of the leak, which concluded the following.
- 1) RV Function Capability-The RV provides overpressure protection for the tube side of the HX. The small leak on the inlet to the RV would not have affected this function.
- 2) Flow/Heat Removal Capability-It was determined that the maximum flow through an assumed failed RV inlet line would be 109 gallons per minute (gpm). The assessment concluded this flow loss would not jeopardize the needed heat removal capability of the Component Cooling Heat Exchanger for analyzed accident conditions.
- 3) Flooding Impact-With the max flow at 109 gpm it would take 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> for water to reach the Residual Heat Removal 6 room barriers allowing sufficient time for Operator action before flooding of equipment would occur.
- 4) Water Spray Impact-a walk down/review of the area was done to determine if a "water spray' could affect safety-related equipment. No problems were identified.
IMPACT ON SAFETY SYSTEM FUNCTIONAL FAILURE PERFORMANCE INDICATORS Since the cooling water system was available, this event did not involve either a partial or complete loss of a safety system function and is not reportable per 10 CFR 50.73(a)(2)(v).
CORRECTIVE ACTIONS TO PREVENT RECURRENCE
- 1) To address the concems associated with the methods used by the Shift Managers to make prompt operability calls, site administrative procedures will be revised to include a symptom based guide, flow chart or checklist for the Shift Manager to use. In addition, all Shift Managers will receive training on the use of this process following its revision.
- 2) To correct the concerns that OPRs are not considered "High Risk Activity", site administrative procedures will be revised to require that OPRs be evaluated under the "high risk" process.
6 EIIS System dentifier:BP (If more space is required, use additional copies of NRC Form 366A) (17)
- 3) Persons performing, reviewing and approving OPRs will receive training on the use of the revised operability determination process, NRC Inspection Manual Part 9900 and GL 90-05.
PREVIOUS SIMILAR EVENTS
None