ML20029E602
| ML20029E602 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/11/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20029E599 | List: |
| References | |
| NUDOCS 9405190156 | |
| Download: ML20029E602 (5) | |
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N.....,f WASHINGTON. O,C. 2055M001 SAFETY EVALUATION BY_THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE EFFECTS OF HIGH TEMPERATURE ON DRYWELL SHIELD WALL AND BIOLOGICAL SHIELD WALL GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219 INTRODUCTION In a safety evaluation dated October 24, 1986 (Ref. 1), the staff had indicated that the concrete of the drywell shield wall (DSW) being subjected to high temperatures required further investigation by the GPU Nuclear Corporation (licensee).
By letter dated May 15, 1990 (Ref. 2), the licensec provided responses to several open items. One of the responses addresscc the issue of high temperatures in the DSW concrete by referring to the analysis performed by the licensee's consultant, ABB Impell Corporation.
The staff reviewed the May 15, 1990 response and requested additional information on the DSW analysis, and also on the condition of the biological shield wall (BSW) around the reactor vessel.
The licensee provided the responses to the staff's requests by letters dated July 17, 1991, October 3, 1991, and March 6, 1992 (Refs. 3,4,5).
In September 1992, the licensee provided a detailed structural evaluation report (Ref. 6) of the spent fuel pool (SFP) prepared by the licensee's consultant, ABB Impell Corporation (In lieu of responding to the staff's questions on the earlier Impell report). This report addresses the integrity of the SFP considering the observed cracks in the girders supporting the SFP and additional weight of the consolidated fuel in the storage racks.
The upper portion of the DSW is included in the finite element model of the SFP and its supporting structures.
However, the report does not directly address the issue of high temperatures in the DSW.
The staff requested the information pertinent to the SEP item by two teleconferences.
By letters (and attachments) dated June 30, 1993, and November 19, 1993 (Refs. 7,8), the licensee provided the requested information.
The following evaluation regarding the integrity of the DSW and the BSW is based on the above information provided by the licensee.
[1ALUATION Drywell Shield Wall In Reference 7,'the licensee has stated that the latest thermocouple data indicated the maximum temperature in the drywell at 94 ft. elevation to be
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285'F.
During the staff's site visit in February 1991, the licensee had provWed 9405190156 940511 PDR ADOCK 05000219 l
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5 the 1988 temperature data at various elevations of the drywell. The data showed the maximum temperatures to vary from 204*F at 89 ft elevation to 264*F near the apex of the drywell dome (Approx.105 ft. elevation).
In evaluating the temperature effects in the DSW, the licensee used bulk temperature in the DSW near the dome as 285*F.
Considering the temperatures on the outside of the shield building as 40*F and 110*F, the licensee has estimated the temperatures at the inside face of the DSW to vary from 181*F to 213*F.
For the purpose of analyzing the DSW subjected to T (operating temperature load), and E (Operating Basis Earthquake) or E' ($afe sh,utdown Earthquake), the licensee's consultant utilized the inside face temperature as 200*F.
The staff recognizes that the spot temperatures at some locations on the inside face of the DSW could be higher.
However, for the purpose of estimating th( effects of high temperatures on the thick DSW concrete, reinforcing bar stresses and to determine the extent of high tension areas, the staff believes that the use of 200*F on the inside face of the DSW is reasonable.
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In the analysis performed by the licensee's consultant (Ref. 6), the upper portion of the DSW supporting the SFP girders and interfacing structures was modelled.
The transfer canal forms a long slot in the interfacing DSW and SFP structure, representing a significant structural discontinuity area.
Though the model does not represent the entire perimeter of the DSW, The staff agrees with the licensee's contention that the modelled portion of the DSW represent the worst case condition for evaluating the high temperature effects on the DSW.
The licensee's consultant developed a detailed finite element model of the SFP and the DSW for the analysis using ANSYS computer program. The DSW is modelled using 3-D solid elements, with two elements across the thickness, 20 elements along the circumferential direction, and 19 elements in the vertical direction.
The mesh is finer in the upper cylinder to match the finer mesh of the SFP walls, and coarser in the lower portion of the DSW.
The stress plots provided with Reference 8 indicate three areas of the model having noticeable tensile stresses; (1) near the bottom of the fuel-transfer canal, (2) at the bottom of the DSW model boundaries, and (3) at the junction of the SFP wall and the DSW, and a few feet from the top of the SFP wall. After comparing the nodal stress magnitudes and the depths of the tensile zones, the licensee selected area (1) as the critical area for further evaluation.
Though none of the nodal tensile stresses exceeds 6/f lbs. per square inch (psi) [where f ' is the reduced compressive strength,'(in psi) of the concrete to account 'for the temperature induced reduction], the critical sections are subjected to axial tension and moments in both directions (circumferential and vertical). A comparison of the moment capacities of sections with the maximum applied forces indicated a minimum margin of 10% against reaching the limiting stresses calculated using ACI 349 Code (Ref. 9) criteria.
The maximum tensile stresses in reinforcing steel is found to be 33 kips per square inch (ksi), for the reinforcing steel having the minimum yield strength of 40 ksi.
The staff considers this evaluation by the licensee acceptable.
However, the staff has concerns regarding the concrete cracks on the outside surfaces of the DSW.
The water spills or leakages from the SFP or the refueling cavity (during refueling activities) has the potential to corrode the highly stressed
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1 reinforcing bars, and with time, make the DSW vulnerable to increasing amount of cracking and degradation. Therefore, the staff recommends that the licensee monitor and repair, as necessary, the upper portion of the DSW to maintain its existing condition.
One criterion to initiate crack repairs is when the crack-widths are larger than 0.5 mm. (0.02 in.).
In a commentary to Section 10.6.4 of Reference 9, a crack-widths range of 0.013 in to 0.016 in. is indicated as the limiting crack widths beyond which it is believed that the moisture migration could proceed and potentially initiate corrosion of the reinforcing steel.
For a massive structure, such as the DSW, the staff believes the maximum tolerable crack width of 0.02 in. is reasonable.
In Reference 10, the licensee indicates that it is developing procedures for monitoring the condition of the DSW during each refueling outage. This is acceptable to the staff.
Bioloaical Shield Wall The biological shield wall (BSW) is a composite steel-concrete cylinder surrounding the reactor vessel.
The wall is framed with a vertical grillage of 27 in. deep wide-flange members, and covered with 5/16 in. plate on the outside and 1/4 in plate on the inside.
The areas between the plates is filled with high density concrete to satisfy the shielding requirements.
The BSW serves as a radiation shield and provides lateral support to the reactor vessel and the containment dry well at specified elevations.
In Reference 4, the licensee provided the thermocouple readings taken in year 1988. The temperatures along the height of the BSW vary between 134*F and 200'F (taken above the top of the USW). The staff agrees with the licensee's contention that these type of temperatures would not adversely effect its structural and shielding functions.
During its visit in February 1991, the staff had requested the licensee to inspect the exterior of the BSW during subsequent refueling outages.
In Reference 5, the licensee stated that four small cracks have been observed on the exterior plates:
Two of the cracks are at re-entrant corners around penetrations for a reactor coolant pipe, and the other two are in seam welds.
The licensee contends that (1) these cracks seem to have existed for several years, (2) none of them appear to be propagating, and (3) they could have been caused by the fatigue due to temperature variations.
The licensee reinspected the cracks during the 14R (March 1993) outage and concluded that they were not propagating. As the cracks could be related to fatigue stresses caused by temperature variations may propagate with time, the staff recommends that the cracks and the general condition of the exterior of the BSW should be monitored periodically.
Ir. Reference 10, the licensee indicates that a structural-system engineer has been assigned to the Oyster Creek site who is responsible for ensuring that the structures at the site are monitored and evaluated. The staff believes that this blanket commitment by the licensee, if properly implemented, would ensure the continued functioning of the BSW.
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_4-CONCLUSION Based on the review of the information submitted by the licensee in the last 4 years, the staff concludes that the Drywell Shield Wall (DSW) and the Biological Shield Wall (BSW) at the OCNGS are capable of performing their intended functions.
The staff, however, recommends that the licensee monitor the conditions of the DSW and the BSW at periodic intervals to ensure their continued functions.
Based on the completion of this effort the staff considers SEP Topic III.7-8
" Design Codes, Design Criteria, Load Combinations, and Reactor Cavity Criteria" resolved.
Principal Contributor:
H. Ashar Date:
May 11, 1994
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.. References 1.
Letter from J. Zwolinsky (NRC) to P. Fiedler (GPUN) with a Safety Evaluation Related to Section 4.12 (SEP Topic 111-7.8) of NUREG 0822, i
Design Codes, Design Criteria and Load Combinations, Dated Oct. 29, 1986.
2.
Letter from J. Devine, Jr. (GPUN) to NRC, " Response to Request for Additional Information on SEP Topic III-7.B," Dated Nov. 15, 1990.
3.
Letter from J. Devine, Jr. (GPUN) to NRC,
" Response to Request for Additional Information on SEP Topic III-7.B," Dated July 17, 1991.
4.
Letter from J. Devine, Jr. (GPUN) to NRC,
" Response to Request for Additional Information on SEP Topic III-7.B," Dated October 3,1991.
5.
Letter from J. Barton (GPUN) to NRC,
" Response to Request for Additional Information on SEP Topic 111-7.8," Dated March 6,1992.
6.
Report Prepared by ABB Impell Corporation for GPUN Corporation, "0CNGS Structural Evaluation of the Spent Fuel Pool," Dated June 29, 1992, submitted to NRC in August 1992.
7.
Letter from R. Keaton (GPUN) to NRC, " Response to Request for Additional Information on Drywell Temperature (SEP Topic III-7.8)," Dated June 3,.
1993.
8.
Letter from R. Keaton (GPUN) to NRC, " Response to Request for Additional Information on Drywell Temperature (SEP Topic III-7.B)," Dated November 19, 1993, 9.
Code Requirements for Nuclear Safety Related Concrete Structures and its Commentary, ACI 349-80.
10.
Letter from R. Keaton (GPUN) to NRC, "Drywell Shield Wall Integrity,"
Dated April 19, 1994.
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