ML19276F569
| ML19276F569 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 03/19/1979 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Wolford A LONG ISLAND LIGHTING CO. |
| References | |
| NUDOCS 7904060175 | |
| Download: ML19276F569 (4) | |
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NUCLEAR REGULATORY COMMISslON WASHINGTON, D.C 20555 y.
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MAR,191979 Docket No.: 50-322 Long Island Lighting Company ATTN: Mr. Andrew W. Wofford Vice President 175 East Old Country Road Hicksville, New York 11801 Gentlemen:
SUBJEC": REQUEST FOR ADDITIONAL INFORMATION - SH0REHAM NUCLEAR POWER STATION In order to complete our review of the Shoreham application, we require adequate responses to the enclosed requests for this additional infor-mation.
If you have any questioas on this matter, please contact us.
Sincerely, L
' Steven A. Varga, Chi
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Light Water Reactors tranch No. 4 Division of Project Management
Enclosure:
Request for Additional Information cc: See next page 79040GQ/7f
MAR 191979 Long Island Lighting Company s
Howard L. Blau Bl au and C'ohn, P.C.
380 North Broadway Jericho, New York 11753 Jeffrey Cohen, Esq.
Deputy Commissioner and Counsel New York State Energy Office Agency Bui.lding 2 Empire State Plaza Albany, New York 12223 Energy Research Group, Inc.
400-1 Totten Pond Road Waltham, Mass. 02154 Irving Like, Esq.
Reilly, Like and Schnieder 200 West Main Street Babylong, New York 11702 J. P. Novarro Project Manager Shoreham Nuclear Power Station P. O. Box 618 Wading River, New York 11792 W. Taylor Reveley, III, Esq.
Hunton & Williams P. 0. Box 1535 Richmond, VA 23212 Ralph Shapiro, Esq.
Cammer a Shapiro No. 9 East 40tn Street New York, New York iO0l6 Edward J. Walsh, Esq.
General Attorney long Island Lighting Company 250 Old Country Road g
Mineola, New York 11501
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k 212-1 212.0 REACTOR SYSTEMS BRANCH 212.103 (RSP) The analyses of abnormal operational transients in the FSAR (15.0) utilize the REDY computer code described in General Electric Company Topical Report NED0-10802, " Analytical Evaluations for the General Electric Boiling Water Reactor." Recent con-firmatory tests conducted at Peach Bottom Unit 2, however, revealed that in certain cases, the results predicted by REDY are nonconservative. Therefore, we require the applicant to recalculate the minimum critical power ratio for the limiting transient (generator load rejection without bypass) using General Electric Company's new computer code ODYN. We will use the results of this calculation to verify the acceptability of the minimum critical power ratio.
212.104 (RSP)
In analyzing anticipated operational transients, the applicant (15.0) has taken credit for plant operating equipment which has not been shown to be reliable as required by General Design Criterion 29. The staff has discussed the application of this equipment generically with General Electric.
In these discussions General Electric has stated that the most limiting transient that takes credit for this equipment is the excess feedwater event.
Further, General Electric has stated that the only plant operating equipment that plays a significant role in mitigating this event is the turoine bypass system and the Level 8 high water level trip (closes turbine stop valves). We will allow the use of the turbine bypass and Level 8 high water level trip systems in mitigating transients except for the turbine trip and generator load rejection without bypass transients which are currently minimum critical power ratio-limiting.
To assure an acceptable level of performance, it is the staff's position that this equipment be identified in the plant Technical Specifications with regard to availability, set points, and surveillance testing.
The applicant must submit his plan for implementing this requirement along with any system modifications that may be required to fulfill this requirement.
212.105 (RSP) We consider ATWS to be an unresolved safety issue. However, (15.0) we have described the type of plant modifications which, if provided, would reduce ATWS risk to an acceptable level.
Volume 3 of NUREG-0460 which describes the rationale for specifying these plant modifications is being reviewed by the Advisory Comittee on Reactor Safeguards. The Regulatory Requirements Review Comittee has completed its review and concurred with c cr approach described in Volume 3 of NUREG 'J"50 insofar as it applies to Shoreham.
We plan to issue requasts for the industry to supply generic analyses of ATWS mitigation capability and anticipate presenting to the Comission in May 1979 our recomendations for its actions to resolve the ATWS
212-2 212.105 (RSP) concern.
Shoreham would be required to implement plant (15.0) modifications in conformance with the Commission's final resolution on this issue.
We require that the applicant agree to implement modifications on a schedular basis in conformance with the Commission's final resolution of this issue.
In the event that Shoreham starts operation before necessary plant modifications are implemented, we require some interim actions be taken by the applicant in order to reduce, further, the risk from ATWS events. The applicant is required to:
(1) Shoreham must have an acceptable recirculation pump trip to assure that the short-term consequences of ATWS events do not result in excessive primary system overpressurization. The criteria for an acceptable recirculation pump trip design are specified in Appendix C of Volume 3 of NUREG-0460. The recirculation pump trip designs for Zimmer, Monticello, and Hatch (modified) have been found acceptable.
(2) Develop emergency procedures to train operators to recognize an ATWS event, including consideration of scram indicators, rod position indicators, flux monitors, vessel level and pressure indicators, relief valve and isolation valve indicators, and containment temperature, pressure, and radiation indicators.
(3) Train operators to take actions in the event of an ATWS including consideration of immediate manual scramming of the reactor by using the manual scram buttons followed by changing rod scram switches to the scram position, stripping the feeder breakers on the reactor protection system power distribution buses, opening the scram discharge volume drain valve, prompt actuation of the standby liquid control system, and prompt placement of the residual heat removal system in the pool cooling mode to reduce the severity of the containment conditions.
Early operator action as described above, in conjunction with a recirculation pump trip, would provide significant protection for some ATWS events, namely those which occur (1) as a result of common mode failure in the electrical portion of the scram system and some portions of the drive system, and (2) at low power levels where the existing standby liquid control system capability is sufficient to limit the pool temperature rise to an acceptable level.