ML20005A896
| ML20005A896 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/10/1981 |
| From: | Baxter T SACRAMENTO MUNICIPAL UTILITY DISTRICT, SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | Buck J, Kohl C, Rosenthal A NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP) |
| References | |
| NUDOCS 8107020274 | |
| Download: ML20005A896 (7) | |
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monamT t.zaMLem mancia n. NintNstriN June 10, 1981 y,, /w KicMamo C. GaLEN JuoiTM a. sanoLam mosEmf e. monelNs N
Alan S. Rosenthal, Esquire Dr. John H. Buck Chairman Atomic Safety and Licensing Atomic Safety and Licensing Appeal Panel Appeal Panel U.S: Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washingt c.n, D.C.
20555 Washington, D.C.
20555 Christine N. Kohl, Esquire Atomic Safety and Licensing Appeal Panel U.S.
Nuclear Regulatory Commission Washington, D.C.
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In the Matter of i
Sacramento Municipal Utility District (Rancho Seco Nuclear Generating Station)
Docket No. 50-312 Chief Administrative Judge Rosenthal and Admirlistrative Judges Buck and Kohl:
Please find enclosed, for your information, a letter dated March 25, 1981, from Babcock & Wilcox to Sacramento Municipal Utility District
(" Licensee") on the subject " Reactor Coolant Pump Suction Small Break LOCA."
B&W's small-break, loss-of-coolant accident analyses and the resultant operator guidelines were the suoject of testi-mony before the Atomic Safety and Licensing Board in this f@W.s.
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e SHAW, PITTMAN, POTTS & TROWBRIDGE Alan S.
Rosenthal, Esquire Dr. John H.
euck Christine N.
Kohl, Esquire June 10, 1981 Page Two proceeding.
See, in this docket Initial Decision (Permitting Continued Reactor Operatior.), LBP-81-12, 1."
N.R.C.
slip op. at paragraphs 88-95, 9*/- 103 (May 15, 1901).
Licensee had not completed its evaluation of the B&W letter when the Licensing Board issued its Initial Decision.
Licensee has now completed its. review of the B&W letter and determined, pursuant to Licensee's own internal procedures, that the information provided in the_ letter is not reportable to the NRC.
Licensee's witnesses have also reviewed the B&W letter and determined that the information provided in the letter does not warrant any change to their testimony previously given before the Licensing Board.
Never-theless, because the B&W 1etter might be considered to have some bearing on your review of the Licensing Board's decision, I am serving the letter on the Appeal Board, the Licensing Board and the parties.
Respectfully submitted, Thomas A.
Baxter Counsel for Licen TAB:jah Enclosure cc:
per Certificate of Service 1
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Babcock & Wilcox wo ie.t Power Generation Division a ucoermoit company March 25' 1981 Ess N " '
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File 177/T1.2 t.vneneuro. virginia 24505 (804 a4 sus
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Mr 0 G. Raasch Manager Generation Engineering Sacramento Municipal Utility District 6201 S Street Sacramento. California 95813
Reference:
R. W. Ganthner to D. G. Raasch. letter of October 3. 1980 Sub,iect: -Reactor Coolant Pump Suction Small Break LOCA
,c Dear Mr. Raasch-Following the TMI-2 accident, the NRC requested several small break accident scenarios be evaluated in order to develop operator guidelines for these events. These analyses' included scenarios where auxiliary feedwater ( AFW) was assumed not to be available at the stgrt of the event. The assumed worst case small break LOCA (less than 0.01 ft.') for these analyses is located at the reactor coolant pump discharge.
This assumption was made because under normal circumstances a greater degree of HPI penetration into tnc reactor vessel is achieved during a suction line break. The purpose of this le'.ter is to prc<ide some information regarding this worst case assumption.s it is affected in the scenario where HPI is not actuated and AFW is delayed.
A brief summary of this situation was provided to you in the referenced letter.
Specific details are included in this letter.
For pump suction line breaks, under normal circumstances,100% of the HPI ficw enters the reactor vessel. For cump discharge line breaks only about 70%
of the HPI flow enters the vessel.
However, for the scenario where break size is such that HPI is not automatically initiated and AFW flow is delayed. the rate of system inventory loss before AFW actuation becomes important.
During pump discharge breaks a two phase discharge results due to the effect of the reactor vessel internal vent valves.
This reduces the rate of system inventor.';
loss. A pump suction break w~ill result in the loss of lower quality fluid which will deplete system inventory at a higher rate.
Thus at the time of AFW actuation the RCS inventory will be less for the pump suction line break than for the punp discharge line break.
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Babcock a,Wilcox :
Analyses which have been conducted for the pump discharge break condition i
.demonstrate that operator actions to start AFW flow in 20 minutes will result in acceptable conditions. However. for the case where there is a pump suction
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i break of a size such that boil dry of the stemi generators occurs prior to an RCS pressure decrease to the HPI actuation setpoint. the 20 minute delay in AFW actuation has not been demonstrated to result in an acceptable RCS ' inventory condition. That is. the 20 minutes of delay analyzed for in the pump discharoe break case has not been analyzed for in the pumo suction break case.
Thus.
althouoh there is certainly a delay in AFW actuation which will result in acceptable conditions. the actual time has not been identified for the pump suction break condition.
'While the actual analyses for the pump suction break delayed AFW scenariE has not been conducted. there is a significant amcunt of guidance for the operator regarding actuation of HPI and AFW. Following the TMI-2 accident.
unall break LOCA operating quidel.ines-were developed.
These guideline; instruct G-the operator to ensure that AFW is being delivered to the steam generators and if it is not, to restore feedwater as soon as possible. Additionally. the guidelines recuire manual actuation of HPI should the system reach saturated conditions.
These actions provide for mitigation of the delayed AFW scenario.
Additionally, upgrades of the.AFW control system have been implemented which would ensure AFW flow in times on the order of one minute.
Thus althouoh the specific analysis has not been conducted'. there is adequate reason to believe that current procedures and system characteristics make the identification of the specific time delay superfluous. However. it is not clear what significance i
the demonstration of a 20 minute operator response time was to the NRC.
The licensing significance associated with the demonstration of a 20 minute operator response time is best determined by each utility. Additionally. the AFW upgrades are plant dependent and B&W cannot assess to what extent the probability of this event has been diminished. At present. the following positions appear to be possible resolution paths on this issue:
1.
Review the AFW systems and confirm that the small break LOCA with delayed feedwater is a highly unlikely scenario and need.not be considered part of the design basis for the plant.
Thus, while the analyses performed may not have considered the worst break location for demonstrating the minimum allowable operator response time, the probability of this event along with the generation of the operator guidelines provide adequate assurance that this transient can be safely mitigated.
2.
Use the basic position outlined in Item i except report to the Commission the potential change in the previously submitted analyses.
3.
Perform detailed evaluations of the pump suction small break LOCA with a delay in the delivery of AFW and determine the time frame available to the operator to restore either feedwater or HPI.
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Babcock & Wilcox >
It is B&W'.s position that Item 3 is not necessary. We believe that either Items 1 or 2 are viable alternatives for the' ultimate resolution of this issue.
If you have any questions regarding the nature of this concern, please call me (804-384-5111, extension 2420) or R. W. Ganthner (804-384-5111, extension 2751) at our Lynchburg office.
Very truly yours, D. C. Holt Engineering Product Manaaer cc:
R. A. Dieterich J. T. Janis J. H. Johnsten J. J. Mattimoe R. P. Oubre R. J Rodriquez i
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE-THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of
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SACRAMENTO MUNICIPAL UTILITY DISTRICT
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Docket No. 50-312
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(Rancho Seco Nuclesr Generating
)
Station)
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CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing letter to the Atomic Safety and Licensing Appeal Board with attachment were served this 10th day of June, 1981 by deposit in the U.S. mail, first class, postage prepaid, u'pon the following:
Alan S. Rosenthal, Esquire Chairman i
l Atomic Safety and Licensing Appeal Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. John H. Buck Atomic Safety and Licensing Appeal Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Christine N. Kohl, Esquire Atomic Safety and Licensing Appeal Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 i
Elizabeth S. Bowers, Esquire Chairman Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. Richard F. Cole Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission l
Washington, D.C.
20555
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. Mr. Frederick J. Shon Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 David Se Kaplan, Esquire Secretary and General Counsel Sacramento Municipal Utility District LP.O. Box 15830 Sacramento, California 95813 Richard L. Elack, Esquire Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Christopher Ellison, Esquire California Energy Commission 1111 Howe Avenue Sacramento, California 95825 Herbert H. Brown, Esquire Lawrence Coe Lanpher, Esquire Hill, Christopher and Phillips, P.C.
1900 M Street, N.W.
Washington, D.C.
20036 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Y.
w_u Thomas A. Baxter 1
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