ML17250B058
| ML17250B058 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/15/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17250B057 | List: |
| References | |
| NUDOCS 8911200191 | |
| Download: ML17250B058 (4) | |
Text
Enclosure 1
SUPPLEMENTAL SAFETY EVALUATION BY THF. OFFICE OF NUCLEAR REACTOR REGULATION FOR SYSTEMATIC EVALUATION PROGRAM/STRUCTURAL UPGRADE PROGRAM R.
E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
1.0 INTRODUCTION
On March 24, 1987, the NRC issued to RG&E a Safety Evaluation (SE) on the Structural Upgrade Program (SUP).
In that report the NRC staff concluded that subject to the plant modifications which RG&E had committed to install, the implementation of the Structural Upgrade Program would provide reasonable assurance that Ginna Station could safely shut down following specified environmental events.
The staff also stated that these conclusions were subject to satisfactory resolution of nine open items.
On May 24,
- 1987, RG&E responded with the necessary information requested by the staff, as well as its comments on the staff's earlier review findings.
The staff met with RG&E on December 14, 1988 to discuss the outstanding items.
The meeting resulted in the staff request for additional information which RG&E submitted to the staff for its review on January 25, 1989.
The staff conducted an audit of the calcu1ations at RG&F engineering office on August 3, 1989, and visited the plant to inspect various SEP modifications on August 4, 1989.
The audit and plant visit. resulted in RG&E's submitta1 for further analysis and confirmatory information on August 31, 1989.
2.0 EVALUATION The following provides our evaluation and resolution of the open items as well as the technical basis for closure of R.
F.. Ginna Structural Upgrade Programs.
With respect to the issue of using actual thermal loads in load combinations for areas of the plant known to have high operating temperatures (e.g.
concrete surrounding the reactor),
the licensee performed a worst-case condition analysis and the results were found to be acceptable.
Furthermore, the licensee has stated based on recorded data that the concrete temperature at reactor vessel supports (hot spots) ranged from 80'F to 110'F which is less than the 150'F maximum allowable temperature in concrete, therefore, the issue is considered resolved.
Regarding the straight line wind load distribution used in the design, the
'.icensee has demonstrated that for Ginna structures use of stepped wind loads per ANSI A58.1-1982 wou1d result only in minor total wind load variation and would not affect the results of the overall structural analysis.
At the audit, the staff reviewed the licensee's drawings and calculation for wind and tornado loads and confirmed that the use of straight line distribution was acceptable.
This issue is resolved.
Regarding the issue of adequately accounting For any loads imparted by the siding or decking to the steel frame, the staff verified that the licensee conservatively assumed its calculations of wind and tornado loads that the siding and decking wi 11 remain intact and thus, transfer the full magnitude of these loads to the structure.
This conservative approach is acceotab~e and the issue is resolved.
89l1200191 891115 PDR ADDCK 05000244 p
PDC On the issue related to design adequacy of roof decking failure due to snow loading, the licenseee stated that all structural roofs of Ginna Station except that of the diesel generator building have load carrying capacities greater than their applied loads.
The diesel generator building roof will be up-graded in the Ginna's Structural Upgrade Program to assure no failure due to snow loads.
At the audit of August 3, 1989, the staff reviewed RGKF' pertinent internal structural documents and ca1culations and inspected the ongoing diesel generator building upgrade work and concluded that the licensee's approach is acceptable for resolving the issue.
lilith respect to the issue of potential buckling of the roof decks used in Ginna structures, the licensee stated that the metal roof decking used at the Ginna is a "multiple stiffened element" per the de<inition of the Cold-Formed Steel Design Manual of the American Iron and Steel Institute Standards.
Because of the shape and the width-to-thickness ratios of the compression zones of the deck, the licensee claims that the full bending capacity of the shape can be developed without buckling.
The staff accepts the licensee's iustification and the issue is considered resolved.
Previously, the licensee committed to examine the east wall of the control building and portion f the diesel generator building for tornado winds and missiles.
The east wall of the control building has been modified and the diesel generated building is being modified to withstand wind and tornado loads including missiles.
The design criteria for these modifications were found acceptable by the staff.
During the August 4, 1989 plant visit, the staff inspected areas affected by these modifications.
The staff concludes that the modifications are acceptable.
On the issue of assuring that previous conclusions reached regarding seismic capability of Ginna structures remair valid considering seismic loads in combination with other loads, the licensee oerformed additional analyses including a complete seismic "analysis o~. an area of the plant that was 'udged to be most critical for overall structural stability of the plant.
This "slice methodology" consisted of analyzing the common wall between the turbine and intermediate buildings.
The results of the analysis were presented to the staff at the December 14, 1988 meeting and confirmed the licensee position that the structures are adequately designed for a combination of seismic and other loads.
The staff did examine some of the calculations at the August 3, 1989 audit meeting and found that the design documentation was in order and acceptable.
The staff concludes that this issue is resolved.
On the issue of eva'1uation of the effects of masonry wall failure on main steam and feed water lines and associated valves an Robert C. Mercredy 2
November 15, 1989 Our Supplemental Safety Evaluation is provided in Enclosure 1.
We consider our efforts on TAC No.
54364 complete.
We understand that full implementation of SUP modifications is targeted for September 1990.
Sincerely,
/s/
Allen Johnson, Project Manager Project Directorate I-3 Division of Reactors Projects I/II Office of Nuclea~ Reactor Regulation
Enclosure:
Supplemental Safety Evaluation cc:
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