ML13056A426
| ML13056A426 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/15/2013 |
| From: | Grecheck E Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML13056A426 (13) | |
Text
Om Dominion Nuclear Connecticut, Inc.
Dominion
'5000 Dominion Boulevard, Glen Allen, VA 23060 Web Address: www.dom.com February 15, 2013 U.S. Nuclear Regulatory Commission Serial No.12-770 Attention: Document Control Desk NLOS/WDC RO Washington, DC 20555 Docket No.
50-336 License No.
DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2 ASME SECTION XI INSERVICE INSPECTION PROGRAM RELIEF REQUEST RR-04-14, REACTOR VESSEL SUPPORTS Pursuant to 10 CFR 50.55a(g)(5)(iii), Dominion Nuclear Connecticut, Inc. (DNC) requests relief for Millstone Power Station Unit 2 (MPS2) from certain examination requirements of Section Xl of the 2004 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Specifically, DNC requests relief from the surface and visual examination of the reactor vessel supports required by ASME Section Xl, IWB-2500 and IWF-2500. provides the specific relief request and discussion of the basis for the request. provides the reactor vessel weld designation drawing for information only. Attachment 3 provides the cavity seal ring drawing and photo. The relief request has been reviewed and approved by the station's Facility Safety Review Committee.
If you have any questions regarding this submittal, please contact Wanda Craft at (804) 273-4687.
Sincerely, Eugene S. Grecheck Vice President - Nuclear Engineering and Development Attachments:
- 1.
Relief Request RR-04-14, Reactor Vessel Supports
- 2.
Reactor Vessel Weld Designation, Drawing # 25203-29525
- 3.
Cavity Seal Ring Drawing and Photo Commitments made in this letter: None M 41
Serial No.12-770 Docket No. 50-336 Page 2 of 2 cc:
U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd.
Suite 100 King of Prussia, PA 19406-2713 J. S. Kim Project Manager - Millstone Power Station U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08 C2A Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station
Serial No.12-770 Docket No. 50-336 ATTACHMENT 1 RELIEF REQUEST RR-04-14 REACTOR PRESSURE VESSEL SUPPORTS MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC.
Serial No.12-770 Docket No. 50-336, Page 1 of 5 Relief Request RR-04-14 Proposed Alternative In Accordance with 10 CFR 50.55a(g)(5)(iii)
-Inservice Inspection Impracticality-
- 1. ASME Code Components Affected The affected components are the Millstone Power Station Unit 2 (MPS2) reactor vessel supports, specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section Xl examination categories and item numbers covering examinations of the reactor pressure vessel (RPV) supports. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1, and IWF-2500 and Table IWF-2500-1 of the ASME BPV Code Section Xl.
ASME Code Class:
Code Class 1 Examination Category:
B-K, F-A Item Numbers:
B10.10 Pressure Vessel, Welded Attachment F1.40 Supports other than Piping Supports Component Identification: Reactor Vessel Supports 2-RVS-1, 2-RVS-2, &
2-RVS-3 (Throughout this request, the above examination categories may be referred to as "the subject examinations" and the ASME BPV Code, Section Xl, may be referred to as "the Code.")
Drawing # 25203-29525, the reactor vessel weld designation drawing, is provided for information only.
2. Applicable Code Edition and Addenda
ASME Section XI, 2004 Edition, No Addenda (Reference 1)
3. Applicable Code Requirement
IWB-2500, Table IWB-2500-1, Category B-K requires 100% surface examination of one of the pressure vessel welded attachments from the accessible side once each ten-year interval.
IWF-2500, Table IWF-2500-1, Category F-A requires visual examination (VT-3) of 100% of reactor vessel supports once each ten-year interval.
Serial No.12-770 Docket No. 50-336, Page 2 of 5
4. Reason for Request
In early 2012, MPS performed a review of industry operating experience (OE) and found instances where reactor vessel supports were not correctly identified for examination in the Inservice Inspection (ISI) Program. As a follow-up to this industry OE, a review of the MPS2 ISI Program was performed to determine if the reactor vessel supports were appropriately identified for examination as required by ASME Section Xl. The review determined that these vessel supports were not included in the MPS2 ISI Program. The review was expanded to include the previous 10-year intervals of the MPS2 ISI Program.
No evidence was found that these supports were ever included in the ISI Program. It is concluded these supports have never received the Code required examinations. This condition was entered in the MPS Corrective Action Program.
The reactor pressure vessel supports for MPS2, 2-RVS-1, 2-RVS-2, and 2-RVS-3, consist of three pads welded to the underside of two inlet nozzles and one outlet nozzle, which normally have stress loads in compression. These pads rest on plates which can slide on a set of base plates anchored in the concrete biological shield wall that surrounds the pressure vessel. The support arrangement permits radial thermal growth of the vessel while maintaining it centered and restrained from movement resulting from seismic force. provides the drawing for locations of the RPV supports.
The 2004 Edition of ASME Section Xl is the current Code of Record for the MPS2 ISI Program and requires examination of component supports during the 10-year interval.
ASME Code,Section XI, Examination Category F-A, Item No. F1.40 requires VT-3 visual examination of these supports once every ten years. Also, ASME Code, Section Xl, Examination Category B-K, Item No. B10.10 requires a surface examination of one of the support integral attachment welds once every ten years.
The inspection of the MPS2 reactor vessel supports was attempted during the fall of 2012 refueling outage, 2R21, but was not successful. The inspection was attempted from above by lowering a camera through each of six refuel cavity seal ring manways to provide viewing for inspection.
However, due to the location of the supports in relation to the manways, and the insulation panels that blocked the viewing, the inspection did not provide any meaningful results. Other entry points were evaluated. Access through the piping penetration opening of the bioshield wall was investigated, but it was determined that inspection through this access point would not be successful due to inadequate clearance between the insulated piping and the wall opening precluding the passage of the camera or a boroscope.
Furthermore, due to the configuration of the insulation on this piping, removal can only be performed from the reactor vessel side of the bioshield wall. This would require removal of the permanently welded cavity seal ring to gain access to remove the insulation. Access from below is not possible due to the limited space between the vessel and the bioshield wall that prohibits building of the required scaffolding.
Serial No.12-770 Docket No. 50-336, Page 3 of 5 To access these supports in order to complete the Code required examinations would require extensive maintenance activities for removal of the permanently welded refuel cavity seal ring and then removal of the insulation panels that surround the supports. Attachment 3 provides the drawing and photo of the cavity seal ring.
Additionally, to remove the cavity seal ring would require mechanical removal of two welds that attach the inner side of the ring to the reactor vessel flange and the outer side of the ring to the refuel cavity floor.
This removal activity introduces an opportunity to significantly damage the ring, which would require additional repair or replacement.
In addition to the extensive maintenance support activities to facilitate these examinations, the dose rate at the support locations is estimated at 4 Rem/hour.
The following list contains estimates of personnel dose that would be received, if the subject examinations were pursued:
" Removal and reinstallation of refuel cavity seal ring - 20.6 Rem (based on historical exposure from 2005 permanent cavity seal ring installation)
Removal and reinstallation of insulation panels - 74.4 Rem (based on 24 person-hours at 0.1 Rem/hour and 24 person-hours at 3 Rem/hour)
Paint removal and surface preparation for examinations - 12 Rem (based on three person-hour at 4 Rem/hour)
Perform examinations - 8 Rem (based on two person-hour at 4 Rem/hour)
Health Physics support estimated at 1 Rem Totaling the above items, the accumulated dose of this task is estimated to be 116 Rem. This dose would present an undue burden on inspectors and support staff to perform these inspections.
Millstone Engineering evaluated the implications of failure to perform the required ASME Section Xl examinations on the reactor vessel nozzle support qualification/functionality.
Engineering concluded that while not in compliance with the Code, the MPS2 reactor vessel supports are expected to remain capable of performing their design function.
The supports were last visually inspected by MPS engineering (non-ASME Code) in 1988 following the cleanup of a boric acid leak at the reactor vessel flange. This examination was performed prior to the installation of the permanent refuel cavity seal ring in 2005. No support degradation was noted at that time. There has been no evidence of boric acid leakage that could affect these supports since 1988.
Additionally, since 2003, under vessel visual inspections for leakage that are performed each refueling outage have confirmed no boric acid leakage in the area.
MPS has revised the procedure governing the under vessel inspections with direction to develop a remediation plan if any evidence of boric acid leakage is detected that could potentially impact these supports.
Serial No.12-770 Docket No. 50-336, Page 4 of 5 In addition, MPS2 vessel supports have been designed with ample structural capacity (using less than 40% of capacity of DBE).
The original reactor vessel support design included consideration of normal operating loads plus pipe rupture loads plus the maximum hypothetical earthquake (Reference 2). The reactor vessel support applied loads due to a main reactor coolant loop pipe rupture are typically an order of magnitude larger than seismic loads. Per Millstone FSAR Section 3.A.5, the Leak-Before-Break (LBB) analyses for the reactor coolant system (RCS) main coolant loops, the pressurizer surge line, and unisolable RCS portions of the safety injection and shutdown cooling piping demonstrated the probability of fluid system piping rupture was extremely low. The LBB analyses were reviewed and approved by the NRC per documents 3A-25, 3A-26, and 3A-27, referenced in MPS2 FSAR Section 3.A.5.
The LBB analysis eliminated these pipe rupture cases from consideration and greatly reduced the load applied to the reactor vessel supports.
Seismic stresses and allowable limits used by the Nuclear Steam Supply System vendor in the design of the reactor vessel supports are based on the material used, incorporate the shape factor of the particular section being analyzed, and originally included seismic and rupture loads.
The seismic stresses and allowable limits based on Millstone Engineering calculations for the reactor vessel supports are provided below:
Stress Level, Percentage of Load Combination ks Allowable Limit, ksi Allowabe Lm ksi Allowable Limit Maximum Primary 9.1 53.4 16.8 Membrane Membrane Plus Bending 25.9 66.6 38.9 Significant margin exists in the reactor vessel nozzle supports to accommodate the remaining required design basis loading, even considering possible support degradation (though considered to be unlikely).
Based on the simple design, previous inspection results and significant margin between the support design and the required support load carrying capability, the reactor vessel supports are considered to be functional for the required design basis loading.
Significant margin exists and the reactor vessel supports are considered functional for the required design basis loading.
Inspections were attempted through the manways, and other available entry points were evaluated for access. The design configuration limits access to the reactor vessel supports, which would result in extensive maintenance support activities (removal of cavity seal ring and insulation panels) for completion of the Code required examinations.
Dominion Nuclear Connecticut, Inc. (DNC) considers that performance of these examinations, with the associated accumulated dose, is impractical.
Therefore, DNC requests the proposed relief be granted pursuant to 10 CFR 50.55a(g)(5)(iii).
Serial No.12-770 Docket No. 50-336, Page 5 of 5
5. Proposed Alternative and Basis for Use
None.
6. Duration of Proposed Alternative
This relief is requested for the duration of the fourth 10-year ISI interval at MPS2, which began on April 10, 2010, and is scheduled to end on March 31, 2020.
- 7. References
- 1. ASME Boiler and Pressure Vessel Code, Section Xl, 2004 Edition, No Addenda, American Society of Mechanical Engineers, New York.
- 2. MPS2 FSAR, Table 4.2-2A, Table of Loading Combinations and Primary Stress Limits.
Serial No.12-770 Docket No. 50-336 ATTACHMENT 2 REACTOR VESSEL WELD DESIGNATION Drawing # 25203-29525 MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC.
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Serial No.12-770 Docket No. 50-336 ATTACHMENT 3 CAVITY SEAL RING DRAWING and PHOTO MILLSTONE POWER STATION UNIT 2 DOMINION NUCLEAR CONNECTICUT, INC.
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