ML19224C373
| ML19224C373 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 05/18/1979 |
| From: | Trammell C, Trammell C Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| FOIA-79-98 NUDOCS 7907020236 | |
| Download: ML19224C373 (5) | |
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Meeting Summary f r Trojan N clear Plant May 18, 1979 Docket Files NRC POR Mr. H. H. Phillips Portland General Electric Company Local PDR
'i 121 S.W. Salmon Street 1
ORB 1 Reading l
NRR Reading Portland, Oregon 97204
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H. Denton E. Case j
Warren Hastings, Esquire Y. Stello r' Counsel far Portland General Electric Company D. Eisenhut i
121 S.W. Salmon Street B. Grimes R. Vollmer Portland, Oregon 97204 A. Schwencer D. Ziemann Mr. J. L. Frewing, Manager
'l Generation Licensing and Analysis P. Check I
G. Lainas Portland General Electric Company D. Davis 121 S.W. Salmon Street B. Grimes Portland, Oregon 97204 T. Ippolito R. Reid Columbia County Courthouse V. Noonan Law Library, Circuit Court Roca G. Knighton St. Helens, Oregon 97501 D. Brinkman Project Manager Director, Oregon Department of Energy OELD Labor and Industries Building, Room 111 OIaE (3)
Salem, Oregon 97310 C. Parrish ACRS (16)
Richard M. Sandvik, Esquire NRC Participants Counsel for Oregon Energy Facility J. Buchanan Siting Counsel and Oregon Department TERA of Energy Licensee 500 Pacific Building Short Service List 520 S.W. Yachill A. Thadani Portland, Oregon 97204 P Mathews f
( C Ro p Michael Malmrose U. S. Nuclear Regulatory Commission Trojan Nuclear Plant P. O. Box 0 Rainier, Oregon 970a3 790702 0p_3 6 Q)smBN
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UNITED STATES e
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- l NUCLEAR REGULATORY COMMISSION j
WASHINGTON. D. C. 20555
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May 18, 1979 I'ocket No. 50-344 LICENSEE: Portland General Electric Company FACILITY: Trojan Nuclear Plant
SUBJECT:
SUMMARY
OF MEETING HELD ON MAY 12, 1979 TO DISCUSS THE AUXILIARY FEEDWATER SYSTEM AT TROJAN NUCLEAR PLANT On May 12, 1979, the NRC staff met with representatives of Portland General Electric Company for a detailed discussion of the auxiliary feedwater system at Trojan Nuclear riant. Attendees are listed in attachment (1).
The meeting was a part of a larger scope task to consider what, if any, changes are required for non-B&W plants in light of the Three Mile Island Unit 2 experience. The material discussed is listed in attachment (2).
The information gathered at this mee ing will be included in a NRC staff report to the Commission.
It will include information generic to g plants and specific information on Trojan. The report will include any recomended changes the staff believes are needed and present conclusions as to whether or not continued operation is justified (all CE and g plants).
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Charles Tramell Operating Reactors Branch #1 Division of Operating Reactors Attachments:
As stated cc: See next page
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s LIST OF ATTENDEES PORTLAND GENERAL ELECTRIC COMPANY MEETING May 12, 1979
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PGE NRC H. Schmidt W. LeFave G. Zimerman C. Trammell R. Barkhurst S. Asselin (Sardia)
S. Hoag J. Hickman (Sandia) b.b h
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'A3 part of its on-g:ing review of the Three Mile Island Unit 2 accid:-t,
, the staff finds that it needs additional information regarding tne tailiary fe26:ater systems (AFL'S).
This information as outlined talew, is rc;uired to evaluate AFWS reliability for Combustion Engineering (CE) and Wes-inghouse (W' designed pressurized water reactors. The requested information is in addition to that requested in the IE Bulletins, and should be brought to the maeting scheduled with the staff on Mcy 8 thru May 12, 1979.
Written system description (as built) including:
l List of Support Systems for Auxiliary Feed System Operation (Both Electric j
and Steam)'
Water Supplies for AFWS (primary and backup)
Current operating precedures and test and maintenance requirements including:
All LCO's for AFWS, main FJ system and related support systems.
Listing of operater actions (local and/or control recm) and timing require-ments for such actions.
Procedures for reinitiating main feedwater ficw.
As Suilt P& ids with symbol keys including condensate and steam side Ledgible Equipment layouts drawings including:
Isemetrics, if available Identification of inhibits preventing accessibility to AFJS components and related electrical equipment Relevant control systems description including:
Schematic or logic control diagrams Listing of actuation signals / logic and control MSIS logic for isolating' AFWS, if installed electric pcwer dependences
.All " readouts" available in control room for AF45 operation AC & DC Power One line diagrams (normal and emergency pcwer supplies)
Divisional designation e.g., Train A, Train B, requirements on all AFWS components and support systems
. List of normal valve states and loss-of-actuation pcwer failure position Operating Experience, including Number of main feedwater interruptions per year experienced to date for each unit Number of demands on AFWS per year to date (test and actual) for. each unit Surrmary of AFWS malfunctions, problems, failures Provide Available reliability analyses Steam Ganerator dry-cut times (assuming less of all fe56;ater ficw, with 100%
initial pcwer, with Reactor trip, no line breaks)
System design bases including:
Seismic and envircnmental qualification Code and Quality, QA l
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r7:We s.ritts.* resp r.iss t; tr.e folicwing si: Of questic.a.s by t/D a l
D scribe backup systems available (to auxiliary fe:d.:ater) for ;reviding feedwater to steam generators. Discuss actions and time re;uiras to make these systems available. Are proceduras available?
If so, pr: vide.
Frovide the follcuing procedures:
loss of offsite pcwer loss of feedwater LOCA (small and large) i Steam Line Sreak Provide folicwing information for p0RV's:
Number capaciti setpoints (open and close) manufacturer and model i
indications of position record cf periods isolated (isolation valve shut) challenges during life of. plant (from plant records) including perfomance of valve, cause of challenge.
i experience of two-phase or succooled discharge of.FORVs and' safety valves with description of valve perforr.ance Provide indicaticns of pCRV isolation valve in the control room.
Provide the follcwinc information on ECCS:
initiation setpoints system description pump performance characteristics (head curves)
Provide reactor protection system trip setpoints.
Provide infomation en charging pumps, how they rel. ate to ECCS including:
number flow vs. pressure pcwer sources and backup water sources seismic qualification List all challenges (and cause) to ECCS as indicated on plant records. -
List 'nd discuss all instances during which your plant has undergone natura a
circulation.
Describe all automatic and manual features which canstop the reactor coolant pumps.
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