ML19220B962
| ML19220B962 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/01/1977 |
| From: | Silver H Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19220B963 | List: |
| References | |
| TASK-TF, TASK-TMR NUDOCS 7904280012 | |
| Download: ML19220B962 (5) | |
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4 UNITED STATES k
NUCLEAR REGU' ATORY COMMISSION f(I'#I j
WASHINGTON. D. C. 20555
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y R 011977 DOCKET NO.
50-320 APPLICANT: Metropolitan Edison Company FACILITY:
Three Mile Island, Unit 2
SUMMARY
OF MEETING ON THREE MILE ISLAND UNIT 2 Representatives of the applicant and his contractors met with members of the staff on February 18, 1977 to discuss the design of the systems and equipment used to mitigate the consequences of a steam line break accident on TMI-2. Other items were also discussed, as noted below.
Steam Line Break After briefly tracing the chronology of concerns about the steam line break accident, the applicart described changes already made in the mitigating systems, primarily aimed at improving the capability of resisting single failure. With regard to Issue 1 of NUREG-0138, the applicant contended that the application of the staff position therein' to TMI-2 imposes specific failures which is not required by GDC-2.
The applicant stated that given a seismic event which causes a steam line break, any one or more of the following additional failures or occurrences would preclude unacceptable consequences due to that accident.
1.
Loss of offsite power, which would trip the condensate andbooster pumps and the reactor coolant pumps, thereby limiting core cooli ng.
2.
All control rods fall on reactor trip, i.e., no stuck rod 3.
A main feed line is broken Formal analyses have not been completed to verify that these events mi tigate the accident consequences, and quantitative probabilities of these occurrences were not presented.
In response to staff inquiry, the applicant again indicated that the criterion in their present analysis for protecting the core of pre-venting a return to criticality is used because analyses are not available to show the extent of core damage resulting frcm a return to power under conditions following a steam line break accident.
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MAR 011977 2
The staff further explained the pos' tion in NUREG-0138 as folicws.
For seismic events causing a steam line break, the accident consecuences must be mitigated only by seismic Category I components, after assuming single f ailure in any seismic Category I ccmponent.
For random breaks in a steam line, accident consecuences must be nitigated by safety grade equipment, but non-safety grade equipment may be used as a backup for single failure in safety grade equipment.
After a caucus, the staff stated its position on TMI-2 as follcws:
1.
The present systems and equipment will be acceotable if it is shown by analysis that the unmitigated consequences of the steam line break are acceptable, using the present assumotions, or 2.
The mitigating eouipment must conform with the cosition taken in Issue 1 of NUREG-0138 and restated above, and approcriate analvsis performed to show acceptable consecuences.
3.
A documented program identifying and scheduling the acclicant's intended actions must be submitted to succort further licensino actions.
If it acoears these actions cannot be completed before the expected fuel loading date, probability arguments involving other costulated mitigating occurrences could be considered in conjunction with this program to further justify continued licensing actions.
Steam Line Break Analysis The staff again noted that its calculated peak containment temoerature is approximately 50*F higher than the acolicant's and is above the design temperature. With regard to activating instrument action, the acolicant stated that only the reactor coolant pressure transmitter is recuired to function after this temperature peak is reached.
A thermal analysis has been perfor ed which shows that the temoerature within the transmitter housing remains below the design temoerature.
This analysis will be submitted. The instruments necessary for main-taining safe shutdown and for long term monitoring are being defined and their resistance to the short term tem::erature soike wi11 be determined. The staff will provide its calculated temperature nrofile to the applicant.
The ouestions on the steam line break analysis raised in our letter of February 1, 1977 were discussed and clarified.
Electrical Items The nossibility of a " hot short" defeating feed latch c30 ability to isolate feedwater flow was discussed.
Construction within cabinets 84 184
MAR v i 1977 acceared satisfactory and the acclicant acreed to furnish additional information on cabling fire resistance, insulation, separation, and tray loacing.
The staff indicated that a formal cuestion will be transmitted re-garding testing of output relays in the SFAS.
Hydrogen Line Break The apolicant presented and discussed hand calculations showing the capability of installed systems to mitigate the consequences of a break in the hydrogen line in t.}e Auxiliar-Buildina.
p, H. 54 ver, Project Manager Lig't/aterReactors Branch No. 4 Divtsion of Project Management 8 A ~ 18Li
Metropolitan Edison Ccmpany George F. Trowbridge, Esq.
Shaw, Pittman, Potts & Trowbridge 1800 M Street, N. W.
Washington, D. C.
20036 Chauncey R. Kepford, Esq.
Chairman York Committee for a Safe Environment 2586 Broad Street York, Pennsylvania 17404 Mr. Richard W. Heward Project Manager GPU Service Corporation 260 Cherry Hill Road Parsippany, New Jersey 07054 Mr. Thomas M. Crimmins, Jr.
Safety and Licensing Manager GPU Service Corporation 260 Cherry Hill Road Parsippany, New Jersey 07054 Metropolitan Edison Company ATTN: Mr. R. C. Arnold Vice President P. 0. Box 542 Reading, Pennsylvania 19603 84~1EG
Attendance List
.'iRC H. Silver W. Jensen R. Tedesco G. Lainas J. Shapaker F. Eltawila D. Ross D. Fischer T. Novak S. Israel J. Watt F. Ashe V. Benaroya P. C. Hearn GPUSC D. Heward E. Wallace L. Lanese B&W L. Pletke B. Gray D. LaBelle R. Lovell J. Agar Met. Ed.
C. Smyth Burns & Ree R. Schlosser A. Dam 84~167