05000306/LER-1917-002, Regarding Reactor Coolant System Shutdown Communication Live Vent Through Wall Defect
| ML17346A079 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/11/2017 |
| From: | Northard S Northern States Power Company, Minnesota, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-17-049 LER 17-002-00 | |
| Download: ML17346A079 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 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 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 3061917002R00 - NRC Website | |
text
1717 Wakonade Drive Welch, MN 55089 800.895.4999 xcelenergy.com December 11, 2017 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Unit 2 Docket No. 50-306 Renewed Facility Operating License No. DPR-60 Xcel Energy RESPONSIBLE BY NATUR L-PI-17-049 10 CFR 50.73 LER 50-306/2017-002-00, Reactor Coolant System Shutdown Communication Line Vent Through Wall Defect Northern States Power Company, a Minnesota Corporation (NSPM), doing business as Xcel Energy, hereby submits Licensee Event Report (LER) 50-306/2017-002-00, Reactor Coolant System Shutdown Communication Line Vent Through Wall Defect.
If there is any question or if any additional information is needed, please contact Leonard Sueper, at 612-330-6917.
Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.
Scott Northard Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company-Minnesota Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC
ENCLOSURE Licensee Event Report 50-306/2017-002-00 3 pages follow
NRC FORM 366 (04-2017)
NRC FORM 366 (04-2017)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020 the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
LICENSEE EVENT REPORT (LER)
(See Page 2 for required number of digits/characters for each block)
(See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
- 1. FACILITY NAME Prairie Island Nuclear Generating Plant
- 2. DOCKET NUMBER 05000306
- 3. PAGE 1 OF 3
- 4. TITLE Reactor Coolant System Shutdown Communication Live Vent Through Wall Defect
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05000 10 16 2017 2017
- - 002
- - 00 12 11 2017 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE 5
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x)
- 10. POWER LEVEL 0
20.2203(a)(2)(ii) 50.36(c)(1)(ii)(A) 50.73(a)(2)(v)(A) 73.71(a)(4) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C) 73.77(a)(1) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(D) 73.77(a)(2)(i) 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(vii) 73.77(a)(2)(ii) 50.73(a)(2)(i)(C)
OTHER Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT TELEPHONE NUMBER (Include Area Code)
Leonard Sueper, Senior Regulatory Engineer (612) 330-6917 CAUSE SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX
CAUSE
SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX B
AB VTV V085 Y
N/A N/A N/A N/A N/A
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED SUBMISSION DATE MONTH DAY YEAR YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
NO ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On October 16, 2017, with Unit 2 shutdown for a refueling outage, investigation into a boric acid indication identified a through wall leak at the socket weld that joins a 3/4 inch line to Loop A Reactor Coolant System (RCS)[AB]
shutdown communication line valve 2RC-8-37 )[VTV]. The leak was isolated by closed valves that would have limited primary coolant leakage to within the capacity of the charging system when the reactor coolant system was pressurized. The quantity of dry boric acid at the location was small (estimated at 1/2 teaspoon in volume). This failure constituted a welding or material defect in the primary coolant system that was not found acceptable under ASME Section Xl and an event or condition prohibited by Technical Specifications.
The cause of the leakage was determined to be stress corrosion cracking. Valve 2RC-8-37 was replaced. In addition, Prairie Island Nuclear Generating Plant intends to perform phased array ultrasonic inspections of socket welds on similar Class 1 piping containing stagnant water during future refueling outages.
DESCRIPTION OF EVENT
On October 16, 2017, with Prairie Island Nuclear Generating Plant (PINGP) Unit 2 shutdown for refueling outage 2R30, an indication of leakage was found on the pipe socket weld upstream of valve 2RC-8-37 [VTV] during boric acid corrosion examinations. Valve 2RC-8-37 functions as a vent path on the normally isolated 3/4 inch Loop A shutdown communication line [AB]. Subsequent non-destructive examination confirmed a pressure boundary leak existed, which was not found acceptable under ASME Section Xl.
EVENT ANALYSIS
This event is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(A), Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded because the weld defect in the primary coolant system was not found acceptable under ASME Section XI. In addition, the pressure boundary leakage exceeded the zero reactor coolant system operational leakage limit specified in Tech Spec 3.4.14 and therefore was reportable per 50.73(a)(2)(i)(B) as a condition prohibited by plant technical specifications.
CAUSE
Laboratory analysis determined the cause of the leak to be stress corrosion cracking. Contributing causes included weld material sensitization and the presence of sulfur and oxygen on internal surfaces. Oxygen is introduced from containment air when the system is in use during outages and is not removed by the normal primary system oxygen scavenging mechanisms because the line is normally isolated / stagnant.
SAFETY SIGNIFICANCE
There was no actual safety consequence associated with this event. The quantity of dry boric acid at the location was small (estimated at 1/2 teaspoon in volume). Even if the weld had experienced a complete circumferential failure, the leak was isolated by closed valves that would have limited primary coolant leakage to within the capacity of the charging system when the reactor coolant system was pressurized (see figure next page).
CORRECTIVE ACTIONS
Valve 2RC-8-37 and the associated weld were replaced and the reactor coolant pressure boundary was restored.
PINGP intends to perform phased array ultrasonic inspections of socket welds on similar Class 1 piping containing stagnant water during future PINGP Unit 1 and 2 refueling outages.
PREVIOUS SIMILAR OCCURRENCES A similar boric acid deposit was previously identified during refueling outage 2R29 in 2015 on valve 2RC-8-37 at the same location. However, visual and dye penetrant testing performed in accordance with plant procedures showed no indication of flaws or an active leak. The origin of the boric acid residue on the valve at the time was incorrectly attributed to a prior leak (e.g. reactor coolant pump seal leak) or maintenance activity.
Page 3 of 3 (04-2017)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 3/31/2020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
- 3. LER NUMBER Prairie Island Nuclear Generating Plant 05000-306 YEAR SEQUENTIAL NUMBER REV NO.
2017
- - 002
- - 00 Figure