ML20008E678
| ML20008E678 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 02/27/1981 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| TASK-***, TASK-RR LAC-7387, NUDOCS 8103090425 | |
| Download: ML20008E678 (2) | |
Text
D DA/RYLAND COOPERAT/VE eo Box air. 2 sis EAST AV SOUTH + LA CROSSE. WifCONSIN S401 608) 788 4 000 February 27, 1981 In reply, please refer to LAC-7387 DOCKET NO. 50-409 U. S. Nuclear Regulatory Commission NN ATTN:
Mr. Darrell G.
Eisenhut, Director Y
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Office of Nuclear Reactor Regulation
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Division of Operating Reactors
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Washington, D. C.
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SUBJECT:
DAIRYLAND POWER COOPERATIVE
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J LA CROSSE BOILING WATER REACTOR (LACBWR) b y/'
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PROVISIONAL OPERATING LICENSE NO. DPR-45 SYSTEMATIC EVALUATION PROGPAM (SEP)~ TOPIC REVdj
Reference:
(1)
NRC Letter, Eisenhut to SEP Licensees, dated January 14, 1981.
Gentlemen:
In response to your request (Reference 1) for DPC assistance in preparation of safety evaluation reports for the Systematic Eval-untion Program, we have attached a schedule for topics we will review and submit for NRC evaluation.
Our maintaining of this schedule is contingent upon timely receipt of acceptable SER's on topics for other facilities.
If there are any questions regarding this letter, please contact us.
Very truly yours, DAIRYLAND POWER COOPERATIVE A44t[b M
Frank Linder, General Manager FL:JDP:af Attachment cc:
J. Keppler, Reg. Dir., NRC-DRO III NRC Resident Inspectors N
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0 SEP TOPICS FOR LICENSEE REVIEW - LACBWR l
Topics To Be Reviewed By May 1,1981:
II.l.A Exclusion Area & Control II.l.B Population Distribution l
II.1.C Potential Hazards Due to Trcnsportation, etc., facilities II.2.C Atmospheric Transport & Diffusion Characteristics for Accident Analysis II.3.A Hydrologic Description II.3.B Flooding Potential & Protection Requirements II.3.B.1 Capability of Operating Plants to Cope with Design Basis Flooding Conditions i
Topics to Be Reviewed by Julv 1, 1981:
II.3.C Safety Related Water Supply (Ultimate Heat Sink) i II.4.B Proximity of Capable Tectonic Structures in Plant Vicinity i
II.4.C Historical Seismicity within 200 Miles of Plant
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III.2 Wind and Tornado Loadings l
III.4.B Turbine Missiles t
III.4.D Site Proximity Missiles (Including Aircraft)
III.5.B Pipe Break Outside Containment III.7.D Containment Structural Integrity Tests VI.7.C.2 Failure Mode Analysis ECCS VI.10.A Testing of Reactor Trip Syste- & Engineered Safety Features, including Response Time Testing i
XV-1 Decrease in feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow & Inadvertent Opening of a Steam Generator Relief or Safety Valve XV-3 Loss of External Load, Turbine Trip, Loss of Condenser Vacuum, Closure of MSIV & Steam Pressure Regulating Failure XV-4 Loss of Non-Emergency AC Power to Station Auxiliaries i
XV-5 Loss of Normal Feedwater Flow j
XV-7 Reactor Coolant Pump Rotor Seizure & Reactor Coolant Pump i
Shaft Break r
XV-8 Control Rod Misoperation XV-9 Startup of Inactive Loop or Recire. Loop at Incorrect Temp, etc.
XV-14 Inadvertent Operation of ECCS XV-15 Inadvertent Opening of a BWR Safety Relief Valve XV-10 Radiological Consequence of Failure of Small Lines Containing Primary Coolant Outside Containment XV-19 Loss of Coolant Accidents - Spectrum of Breaks XV-20 Radiological Consequences of Fuel Damaging Accidents e
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