ML20074A782

Responds to SER Issue 48, High Energy Line Breaks (HELB) & 830124 Request to Consider Addl Single Failure within Sys Used to Mitigate Event.Postulated worst-case Scenarios Listed.Occurrence of Events Unlikely.Issue Closed

ML20074A782
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	05/11/1983
From:	James Smith LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SNRC-887, NUDOCS 8305160070
Download: ML20074A782 (4)

Text

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LONG ISLAND LIGHTING COMPANY

• dE%EP l SHOREHAM NUCLEAR POWER STATION k%~ m, . . .m _ u P.O. BOX 618, NORTH COUNTRY ROAD e WADING RIVER, N.Y.11792 Direct Dial Number May 11, 1983 SNRC-887 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SER Issue No. 48 - High Energy Line Breaks Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Reference:

(1) Letter SNRC-786 dated 11/8/82 (2) Letter NRC (A. Schwencer) to LILCO

(.M . S. Pollock) dated 1/24/83

Dear Mr. Denton:

In response to SER Issue No. 48, "High Energy Line Breaks" (HELB),

LILCO had submitted the reference (1) letter forwarding a report entitled "High Energy Line Break / Control System Failure Analysis".

This report represented a comprehensive study, including a walk-down of plant areas, that was conducted (1) to identify non-safety control systems and components that may be affected by postulated pipe breaks and then (2) to conservatively determine the state of the reactor as a result of the simultaneous failure of all affected non-safety control systems. It was concluded that all conditions resulting from the postulated pipe break events are bounded by the accident analyses contained in Chapter 15 of the FSAR, and are therefore capable of being mitigated either automatically or by operator action.

In the reference (2) letter, the staff advised that their review of the above noted report cannot be fully completed until LILCO provides additional information on the effects of humidity, pressure and temperature on the operability of these non-safety control systems.

These effects have been addressed in formulating the conclusions reached in the HELB report, although a brief clarification may be beneficial. As stated in Section 4.1 " Analysis Methodology" 0305160070 830511 PDR ADOCK 05000322 E pm p/

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May 11, 1983 SNRC-887 Page 2 two general methods were used to analyze the pipe break zones utilized in the study. For small confined zones, it was assumed that any HELB would incapacitate all non-safety control components within the zone. This assumption was made even though specific components may not be affected by the jet impingement or pipe whip resulting from a specific break. Using this conservative "sarificial approach", it becomes apparent that the environmental effects on these components are directly enveloped within the scope of the report.

In large, more open zones, only the components within the range of the high energy lines were assumed to fail simultaneously with the pipe break. This is consistent with the goals of the study, to determine whether the result of FSAR Chapter 15 accident analyses are exceeded. FSAR Chapter 15 analyses primarily address short term effects where limiting values generally occur very rapidly after event initiation. Assuming a reactor scram, automatic actions would quickly take place to mitigate the immediate effects of the event. Environmental effects on components in these large spaces would tend to develop relatively slowly in comparison to the dynamic effects on the components which would lead to more rapid automatic and operator initiated mitigative actions.

In addition, the staff requested, in the reference (2) letter, that the HELB study consider an additional single failure within the systems used to mitigate the event. In response, two examples of postulated worst-case scenarios were evaluated for the Shoreham plant. These two scenarios are identified below:

CASE I

, a) HELB occurs in Turbine Building b) Loss of feedwater heating occurs, causing reactor power increase to 117% of rated.

c) Turbine generator trip occurs coincident with peak reactor power d) Scram occurs as a result of turbine generator trip.

Loss of offsite power also occurs.

e) HPCI fails (Single failure) f) RCIC operates g) Reactor water level is restored by RCIC.

CASE II a) Steps a through d are the same as CASE I e) Loss of turbine bypass to condenser (single failure) f) HPCI operates g) Reactor water level is restored by HPCI The occurrence of these events is extremely unlikely. This l

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May 11, 1983' SNRC-887 Page 3 conclusion is based on consideration of the probability that a combination of the worst case conditions occurs concurrently:

The worst case pipe segment breaks on the most important line; HELB can affect all controllers in an area and cause

' failures in worst case modes; l -

' Breaks occur at worst case locations (in reality, many of these locations have low calculated stress levels and thus are unlikely to fracture);

Both turbine trip and reactor high power-level trip occur at appropriate (i.e. worst cases) times; Additional single failure occurs Regardless, these two cases were analyzed quantitatively using conservative Chapter 15 analysis models for the two analyses and the results indicate that the short term part of the event with bypass (turbine trip at the thermal power monitor set-point power) is enveloped by the FSAR Chapter 15 Accident Analysis. In this case, the peak fuel cladding temperature is less than 900 F as compared to 2200 F limit. The second event which imposed a failure of the turbine bypass system on the initial scenarios was estimated to reach a peak cladding tem-perature of about 1200 F, again well within the FSAR Chapter 15 Accident limits. This further confirms the conclusions outlined in the reference (1) letter.

It should be noted that the long term plant cooldown of these two events with various system failures,such as HPCI inoperative, are addressed in the Emergency Procedure Guidelines developed for these types of concerns.

The submittal of this information should be sufficient to close SER issue No. 48.

l Should you have any further questions, please contact this office.

l l

Very truly yours, V

. L. Smith Manager, Special Projects Shoreham Nuclear Power Station RWG:bc l cc: J. Higgins

! All Parties Listed in Attachment 1

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.m . . o ATTACH.*tEMT 1 Lawrence Drenner, Esq. .

Herbert H. Brown, Esq.

Administrative Judge Lawrence Coe Lanpher, Esq. .

. Atomic Safety and Licensing Karla J. Letsche, Esq.

Board Panel Kirkpatrick, Lockhart, Hill U.S. Nuclear Regulatory Commission Christopher & Phillips Washington, D.C. 20555 8th Floor 1900 M Street, N.W.

Washington, D.C. 20036 Dr. Peter A. Morris Administrative Judge

. Atomic Safety and Licensin9 Mr. Marc W. Goldsmith

- Board Panel Energy Research Group U.S. Nuclear Regulatory Commission 4001 Totten Pond Road Washington, D.C. 20555 Waltham, Massachusetts 02154 Dr. James 11. Carpenter MHD Technical Associates Administrative Judge 1723 Hamilton Avenue Atomic Safety and Licensing Suite K Board Panel San Jose, California 95125 U.S. Nuclcar Regulatory Commission .

Washington, D.C. 20555 -

Stephen B. Latham, Esq.

I Tuomoy, Latham & Shea l Daniel P. Brown, Esq. 33 West Second Street Attorney P.O. Box 393 Atomic Safety and Licensing Riverhead, New York 11901 l Board Panel U.S. Nuclear Regulatory Commission ,

Washington, D.C. 20555 j Ralph Shapiro, Esq.

Cammer and Shapiro, P.C.

9 East 40th Street Bernard M. Bordenick, Esq. New York, New York 10016

David A. Repka, Esq.

U.S. Nuclear ilegulatory Commission Washington, D.C. 20555

Mattheu J. Kelly, Esq.

i State of New York James Dougherty Department of Public Service 3045 Porter Street Three Empire State Plaza Washington, D.C. 20003 Albany, ::ew York 12223 1

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