05000306/LER-2014-001, Re Unanalyzed Condition Due to Removal of Multiple Steam Generator Lateral Support Shims and Bumpers
| ML14062A091 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/03/2014 |
| From: | Davison K Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-PI-14-021 LER 14-001-00 | |
| Download: ML14062A091 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(vii)(B), Common Cause Inoperability |
| 3062014001R00 - NRC Website | |
text
Xcel Energy@
March 3, 2013 US Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Unit 2 Docket 50-306 Renewed License No. DPR-60 L-PI-14-021 10 CFR 50.73 LER 50-306/2014-001-00, Unanalyzed Condition Due to Removal of Multiple Steam Generator Lateral Support Shims and Bumpers Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), encloses Licensee Event Report (LER) 50-306/2014-001-00 for the Unanalyzed Condition Due to Removal of Multiple Steam Generator Lateral Support Shims and Bumpers.
Summary of Commitments This letter contains no new commitments and no changes to existing commitments.
Kevin Davison Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company-Minnesota Enclosure (1) cc:
Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC Department of Commerce, State of Minnesota 1717 Wakonade Drive East
- Welch, Minnesota 55089-9642 Telephone: 651.388.1121
ENCLOSURE1 LICENSEE EVENT REPORT 50-306/2014-001-00 4 Pages Follow
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 COMMISSION
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL I REV 2 OF 4 NUMBER NO 2014- 001 - 00 The Hot Gap clearance checks and adjustments were performed following the Steam Generator (SG) 1 Replacement Project. The Hot Gaps were checked for trending purposes during the course of plant heat-up for thermal growth to ensure sufficient clearances were maintained as to prevent thermal growth binding.
On December 20, 2012, the Engineering Change for Steam Generator Project Reactor Coolant System Piping and Steam Generator Supports was approved. The Engineering documents did not identify precautions or limitations regarding the required ULS shim adjustment during Mode 3.
On June 19, 2013, Work Package (WP) for hot gap measurements was issued, which includes lower lateral support (LLS) one-bumper at a time restriction during Mode 3. There was no ULS Mode 3 restriction or an acceptable method of checking hot gaps as found in the Unit 1 Steam Generator Replacement WPs. Work package was approved by vendor and Xcel Energy.
On December 29, 2013, at 1327 CST, Unit 2 reached Normal Operating Pressure (NOP) I Normal Operating Temperature (NOT) per 2C1.2-M3, Mode 3.
On December 29, 2013, at 1552 CST, SGR project reported@ 1545 that 21SG & 22SG have the LS-6 shim on the lower lateral supports stuck in position. A partial cooldown took place to adjust the shims. Shims were loose @ 535 degrees F and stuck @ 54 7 degrees F.
On December 30, 2013, at 0333 CST, Unit 2 again reached Normal Operating Pressure (NOP) I Normal Operating Temperature (NOT) per 2C1.2-M3, Mode 3.
On December 30, 2013, the field team had difficulty achieving the repeatable measurements of the ULS shim gaps when the shims were in place. Craft, Field Engineers (FEs), and Quality Control (QC) measured the shim gaps separately using feeler gages and arrived at different gap clearances.
On December 30, 2013, a meeting was held by vendor engineering, Foremen, FEs, QC, Xcel Energy, and others. The resolution to the gap clearance differences was to remove the shim packs and create shims to fill the gaps, using C-clamps to hold the shim sections together for consistent measurements. Once the shim packs were installed there would be no further measurements. There was no discussion on restriction to remove one ULS shim pack at a time.
On December 30, 2013, Craft workers were directed by vendor mechanical superintendent to remove all shim packs to plant elevation 755' for detailed measurement and adjustment.
1 EllS System Code - SG (01-204)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET Prairie Island Nuclear Generating Plant Unit 2 05000-306 APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0/31/2017
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL I REV 3 OF 4 NUMBER NO 2014- 001 - 00 On December 30, 2013, at approximately 2100 CST, all ULS shims were removed to plant elevation 755' for gap measurement and adjustment.
On December 31, 2013, at 0630 CST, the dayshift Xcel Project Oversight noted on arrival in
. containment that all the shim packs on SG 21 were removed and only positions 6 and 12 were installed on SG 22.
On December 31, 2013, at 0933 CST, both RCS Loops were declared inoperable and entry into Tech Specs 3.4.5 Condition D commenced, which placed the Unit 2 Reactor Coolant System (RCS) in an unanalyzed condition for seismic and pipe rupture analyses and was reportable under 1 0 Code of Federal Regulations (CFR) 50.72(b)(3)(ii)(B).
EVENT ANALYSIS
The Steam Generator is supported on a structural system consisting of four vertical columns fitted at the top and bottom with a double clevis and pin assembly. The vertical column clevis base plates are bolted to the SG support feet and permit movement in the horizontal plane to accommodate reactor coolant pipe thermal expansion. Horizontal restraint is accomplished at two locations. The lower lateral support is located at the support feet and the upper lateral support is located near the center of gravity below the transition cone. This combination of upper and lower supports and included stops and hydraulic shock suppressors limit and control horizontal movement for pipe rupture and seismic effects.
On December 31, 2013, at 0933 CST, both RCS Loops were declared inoperable and entry made into Tech Specs 3.4.5 Condition D, due to the Unit 2 RCS in an unanalyzed condition for Seismic and pipe rupture analyses. The unanalyzed condition was reportable under 1 OCFR50.73(a)(2)(ii)(B), any event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.
Based on further analysis, it was determined that with both RCS loops being inoperable, this condition was a loss of safety function, reportable under 1 OCFR50. 73(a)(2)(v)(B), any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat.
Additionally, this condition was reportable under 1 OCFR50.73(a)(2)(vii)(B), any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to remove residual heat, and 1 OCFR50. 73(a)(2)(ix)(A)(2), any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to remove residual heat. (01-204)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET Prairie Island Nuclear Generating Plant Unit 2 05000-306
SAFETY SIGNIFICANCE
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0/31/2017
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 6. LER NUMBER
- 3. PAGE YEAR I SEQUENTIAL I REV NUMBER NO 4 OF4 2014- 001 - 00 There were no actual nuclear, radiological, industrial, and/or environmental events. There is a regulatory impact.
A potential consequence of excessive gap may be increased stresses in the piping and support system beyond allowable design basis and subsequent reduction in design margin. Shim gap requirements are established to restrain movement of the Steam Generator in a seismic event, primary line break, or secondary line break and ensure analyzed stresses in the system remain within design basis allowable levels.
CAUSE
The root cause evaluation has not been completed at this time.
CORRECTIVE ACTION
The root cause evaluation has not been completed at this time.
PREVIOUS SIMILAR EVENTS
None.