ML20040E108
| ML20040E108 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco, Arkansas Nuclear |
| Issue date: | 01/28/1982 |
| From: | Black R NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Buck J, Kohl C, Rosenthal A NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP) |
| References | |
| ALAB-655, ISSUANCES-SP, NUDOCS 8202030118 | |
| Download: ML20040E108 (2) | |
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er ed January 28, 1982 Alan S. Rosenthal, Esq., Chairman Christine N. Kohl, Esq.
Atomic Safety and Licensing Atomic Safety and Licensing Appeal Board Appeal Board U.S. Nuclear Regulatory Comission U.S. Nuclear Regulatory Comission Washington, D.C.
20555 Washington, D.C.
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s In the Matter of Sacramento Municipal Utility District
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Docket No. 50-313 (SP)
Dear Members of the Appeal Board:
On December 11, 1981, the NRC Staff filed responses in the form of affidavits to the seven items of information requested in your Memorandum and Order dated October 7,1981 (ALAB-655).
Item No. 5 was an affidavit of Walter L. Jensen, Jr. dated November 24, 1981, regarding Staff comments on the March 25, 1981 letter from B&W to SMUD concerning
" Reactor Coolant Pump Suction Small Break LOCA." This same affidavit was referenced by both Staff and Licensee in their respective pleadings with respect to the Licensee's Motion for Reconsideration of the Licensing Board's Partial Initial Decision of December 14, 1981, in the matter of Metropolitan Edison Company (Three Mile gland Nuclear Station, Unit No. 1), Docket No. 50-289 (Restart).-
While considering the filings of the Licensee and the Staff in the TMI-1 proceeding, the Licensing Board requested that Mr. Jensen's affidavit be supplemented.
If See Licensee's Motion for Reconsideration, dated December 14, 1981; IIRf Staff's Answer in Support of Licensee's Motion for Reconsideration, dated January 8, 1982.
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In particular, the Board wanted Mr. Jensen to explain in more detail the bases for his conclusion that the time available for operator action to assure adequate core cooling of the reactor. core would not be significantly different for breaks occurring in the reactor coolant pump suction and discharge piping.
Mr. Jensen, in an affidavit dated January 22, 1982, provided the supplementary information requested by the THI-1 Licensing Board.
Since that affidavit provides clarifying and supplementary information i
to Mr. Jensen's previous affidavit submitted in this proceeding, I am enclosing it for your information and consideration.
j Sincerely, I
Richard L. Black Counsel for NRC Staff cc: Dr. Richard Cole Mr. Frederick J. Shon Elizabeth S. Bowers, Esq., Chairman Mr. Richard D. Castro James S. Reed, Esq.
David S. Kaplan, Esq.
Herbert H. Brown, Esq.
Thomas A. Baxter, Esq.
Senator Allen R. Carter, Chairman Christopher Ellison, Esq.
Mr. Michael R. Eaton Atomic Safety and Licensing Board Panel Atomic Safety and Licensing Appeal Board Panel Docketing and Service Section Distribution:
Black Reis/Lessy Olmstead Christenbury/Scinto Cunningham/Engelhardt 0 ELD Formal Files MPadovan-330 NRC Docket Files /PDR/LPDR g
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COP 91ISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of METROPOLITAN EDISON COMPANY, ET AL (Three Mile Island Nuclear Station Dock
-289 Unit 1) 1 AFFIDAVIT OF WALTON L. JENSEN. JR.
I Walton L. Jensen, Jr., being duly sworn, depose and state that:
1.
I am an employee of the U.S. Nuclear Regulatory Commission (NRC). My present position is Senior Nuclear Engineer, Reactor Systems Branch, Division of Systems Integration within the Office of Nuclear Reactor Regulation. A copy of my professional qualifications is attached.
2.
The purpose of my affidavit is to supplement the information provided in my affivavit dated November 24, 1981. That affidavit was filed in a proceeding in the matter of Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station, Docket No. 50-312 SP. The infonnation in this supplemen-tary affidavit is provided to explain in more detail the basis for the staff's conclusion that the minimum time available for operotor action after a break in the reactor coolant pump suction piping would not be significantly different l
l from that available for operator action after a break in the reactor coolant pump discharge piping.
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The NRC staff has reviewed the information cintained in the letter from
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Babcock and Wilcox to GPU Service Corporation on " Rec'+.or Coolant Pump Suction break LOCA" dated March 25, 1981, which indic. ted that more reactor coolant system water might be lost for small breaks at the reactor coolant pump suction than fcr breaks at the reactor coolant pump discharge in the event that all feedwater were temporarily lost. We conclude that this infor-mation is not significant with regard to the safe operation of TMI-1 and that additional small break LOCA analyses at the Reactor Coolant Pump suction need not be performed. This conclusion was derived from the following considerations:
a.
Regardless of the postulated break location in the cold leg piping, the reactor vessel water level would initially decrease to the same approximate elevation.
b.
The additional loss of primary system inventory during a break in the pump suction piping would be from water in the cold leg piping, c.
In the absence of Emergency Feedwater the operator has a ninimum of 20 minutes to actuate High Pressure Injection (HPI) regardless of break location in the cold leg piping.
d.
Emergency procedures instruct the operator to actuate HPI immediately regardless of break location if a loss of all feedwater has occured.
The discussion below addresses these areas:
In the event of a small break LOCA at TMI-1 the liquid level in the broken pipe would decrease to the elevation of the break by liquid discharge from the break. The break flow would then be steam which would be generated in the core by decay heat. The High Pressure Injection system at TMI-1 has l
sufficicnt capability to replenish the wat;r boiled in the com by the decay heat.
For breaks in the cold leg piping at the reactor coolant pump suction water would be lost from the reactor vessel to the break until the liquid level dropped to the reactor vessel inlet nozzle elevation. At this point only steam would be lost from tne reacter vessel. Liquid discharge would continue from the break until the break was uncovered.
More water would be lost for a break in the reactor coolant pump suction piping than for a break in the discharge piping since the cold leg suction piping is located at a lower elevation than the cold leg discharge piping.
The additional coolant lors however would be limited to the cold leg piping inventory below the reactor vessel inlet nozzle.
The water loss from the reactor vessel which prc vides core cooling is limited by the elevation of the reactor vessel inlet nozzle so that loss of vessel water would be approximately the same regard-less of the break location in the cold leg.
Following the event at TMI-2, B&W performed small break LOCA analyses beyond those which had been presented to the staff as a licensing basis to show compliance with 10 CFR 50.46. These additional analyses were performed for i
the purpose of providing guidance to the operator and are documented in the B&W report titled " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177-FA Plant" dated Pay 7, 1979.
These analyses demonstrated that Emergency feedvater would be requir'd for 2
breats of 0.01 ft and smaller in the reactor coolant pump discharge piping j
to depressurize the reactor system sufficiently to actuate High Pressure Injection.
It was further demonstrated that operator action within 20 minutes to manually actuate HPI would prevent core uncovery. Operators at TMI-1 are instructed to initiate HPI immediately in the event that all feedwater is lost in both the small break LOCA and loss of feedwater proce-dures.
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Pump suction breaks coincident with a loss of all feedwater were not analyzed in the May 7th report. For the reasons discussed above, the amount of reactor vessel water that would be available to cool the core would be approximately the same after 20 minutes for a break at the pump suction as for a break at the pump discharge.
It should be noted that more HPI water would be available to makeup the water boiled by decay heat in the core for the pump suction break than was assumed for the pump discharge break. The pump discharge Ireak analyses in the May 7th report assumed that the break was between the HPI nozzle and the reactor vessel and that 30% of the total HPI flow was lost through the break. For a pump suction break, all of the HPI water would be available to flow to the We therefore conclude that a minimum of 20 minutes would be available core.
to the operator to actuate HPI and prevent core uncovery for breaks in the pump suction as well as at the pump discharge, even if all feedwater is temporary lost, and that the operating procedures at TMI are adequate for either event.
The above statements and opinions are true and correct to the best of my personal knowledge and belief.
/V0 OH WaltonL.'Je[en,Jr.
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Subscribed and sworn to before me this 12#' day of/2muov/ 19 h 0
Onda D, Notary Public lif consnitsion Expires: (L,/u h /90A er F
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WALTON L. JENSEN, JR.
PROFESSIONAL QUALIFICATIONS I am a Senior Nuclear Engineer in the Reactor Systems Branch of the Nuclear Regulatory Commission.
In this position I am responsible for the technical analysis and evaluation of the public health and safety aspects of reactor, systems.
Frca June 1979 to December 1979, I was assigned to the Bulletins and Orders Task Force of the Nuclear Regulatory Commission.
I participated in the preparation of NUREG-0565, " Generic Evaluation of Small Break Loss-of-Coolant Accident Behavior in Babcock & Wilcox Designed 177-FA Operating Plants."
From 1972 to 1976, I was assigned to the Containment Systems Branch of the NRC/AEC, and from 1976 to 1979, I was assigned to the Analysis Branch of the NRC.
In these positions I was responsible for the develnpment,and evaluation of computer programs and techniques to calculate the reactor system and containment system response to postulated loss-of-coolant accidents.
From 1967 to 1972, I was employed by the Babcock and Wilcox Company at Lynchburg, Virginia. There I was lead engineer for the development of loss-of-coolant computer programs and the qualification of these programs by comparison with experimental data.
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From 1963 to 1967, I was employed by the Atomic Energy commission in the Division of Reactor Licensing.
I assisted in the safety reviews of large power reactors, and I led the reviews of several small research reactors.
I received an M.S. degree in Nuclear Engineering at the Catholic University of i
America in 1968 and a 5.5. degree in Nuclear Engineering at Mississippi State University in 1963.
I am a graduate of the Oak Ridge School for Reactor Technology, 1963-1964.
I am a member of the American Nuclear Society.
I am the author of three scientific papers dealing with the response of B&W reactors to Loss-of-Coolant Accidents and have authored one scientific paper dealing with containment analysis.
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