05000263/FIN-2010002-01
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Finding | |
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| Title | Reactor Building Crane Design and Licensing Basis Issues |
| Description | The inspectors reviewed the following licensing documents for the reactor building crane: NRC letter to Northern States Power (NSP), Safety Evaluation by the Office of Nuclear Reactor Regulation [NRR] Supporting Approval of Crane Modification and Use of 70 Ton Spent Fuel Shipping Cask IF-300, dated May 19, 1977; NSP letter to NRC, Response to Request for Additional Information, dated February 28, 1977; and USAR, Section 10.2, page 4 of 24, and Section 12.2 page 28 of 49, and page 29 of 49, Revision 23 and Revision 25.The NRC letter to NSP, Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Approval of Crane Modification and Use of 70 Ton Spent Fuel Shipping Cask IF-300, dated May 19, 1977, established the reactor building overhead crane capacity as a maximum of 85 tons and the crane seismic analysis did not analyze for a maximum 85 ton lifted load concurrent with a seismic event based on extremely low probability of both events occurring simultaneously. The licensee subsequently changed the reactor building crane capacity from 85 tons in USAR, Section 10.2, page 4 of 24,and Section 12.2, page 28 of 49, and page 29 of 49, Revision 23 to a crane capacity of105 tons in USAR Section 10.2, page 4 of 24 and Section 12.2, page 28 of 49 and page 29 of 49, Revision 25.The inspectors noted that the licensee did not perform a written 10 CFR 50.59evaluation to assess the following: 1) whether the change of increasing design loads on the crane and the crane support structure required a license amendment and, 2) probabilistic analysis with consideration for a new maximum crane lifted load of105 tons that evaluates whether or not a lifted load must be considered during a seismic event for the design of the reactor building crane and crane support structure. The inspectors reviewed Calculation Nos. CA 76 138, Structural Requalification for New 85 Ton Crane, Revision 0; CA-05-103, Reactor Building Superstructure Seismic Response Analysis with 105 Ton Crane, Revision 0A; and CA-05-107,Structural Seismic Qualification Reactor Building Crane Upgrade for ISFSI, Revision 0B. The inspectors were concerned that the reactor building crane and reactor building crane support structure had been evaluated using friction in a linear elastic analysis to reduce seismic load effects applied to the reactor building crane and crane support structure. The licensee used much smaller seismic loads limited by the friction force and this resulted in a significant load reduction for qualification of the reactor building crane and reactor building crane support structure. In addition, the non-linear effects of friction have not been addressed in the aforementioned calculations. The licensee was unable to provide evidence that the NRR staff had approved friction in a linear elastic analysis as a method of evaluation for this application. The use of friction to reduce seismic load effects on the reactor building crane and reactor building crane support structure was not discussed in the USAR. The inspectors reviewed Calculation No. CA-05-101, Evaluation of Reactor Steel Superstructure for 105 Ton Reactor Building Crane, Revision 3A. The inspectors were concerned with the following: 1) The minimum yield strength for the American Society of Testing and Materials (ASTM) A1 trolley and the bridge rail has not been established in accordance with the American Institute of Steel Construction (AISC) code and,2) The restraint mechanism in the longitudinal direction of the trolley rail and bridge rail connection was based on the use of friction resistance between the bottom of the rail and the supporting beam to resist the sliding of the rail during a design and licensing basis event. In response to the concern, the licensee initiated corrective action program documents CAP 01214808, RB Crane Seismic Calc may not be Consistent W/License Basis, dated January 22, 2010 and CAP 01222530, Crane Heavy Lift Inspection URI:10 CFR 50.59 for Crane Upgrade, dated March 13, 2010. Near the end of the inspection period, the licensee provided the inspectors additional information relevant to the design basis and licensing basis of the reactor building crane and reactor building crane support structure which will require additional review. Therefore, this issue is considered an unresolved item (URI 05000263/2010002-01, Reactor Building Crane Design and Licensing Basis Issues) pending additional inspector review to determine design and licensing basis requirements |
| Site: | Monticello  |
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| Report | IR 05000263/2010002 Section 1R20 |
| Date counted | Mar 31, 2010 (2010Q1) |
| Type: | URI: |
| cornerstone | Mitigating Systems |
| Identified by: | NRC identified |
| Inspection Procedure: | IP 71111.2 |
| Inspectors (proximate) | K Riemer M Mitchell S Thomas J Bozga P Voss A Scarbearyk Riemers Thomas L Haeg S Bakhsh J Bozga M Learn |
| INPO aspect | |
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Finding - Monticello - IR 05000263/2010002 | |||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||
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Finding List (Monticello) @ 2010Q1
Self-Identified List (Monticello)
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