ML19254E087
| ML19254E087 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 09/20/1979 |
| From: | Siegel B Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17266A109 | List: |
| References | |
| NUDOCS 7910310072 | |
| Download: ML19254E087 (5) | |
Text
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UNITED ST TES OF A" ERICA NUCLEARRESULATG?.YCC!'MISSIO:d l
BEFORE tie ATC!"C SACETY A!;D LICENSI!:S API'EAL BOARD In the '4atter of
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FLOPIDA PCUER & LIGHT COMPANY
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(S'. Lucie Nuclear Power Plant
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Dacket Mc. 50-339 Uiit No. 2)
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Affidavit of Byron L. Siegel I, Byron L. Siegel, bein,g duly sworn to depose and state:
I am a Nuclear Reactor Engineer in the Office of Nuclear Reactor 1.
Regulation.
I have prepared a statement of professional qualifications which 2.
is attached to this affidavit.
I have prepared testimony in the captioned proce'eding addressing 3.
the Appeal Board's Question numbered B2 as stated in ALAB-537.
This question requested the staff and applicant to analyze events that would occur between the loss of all AC power and the violation of either the fuel design limits or the design conditions _ of the reactor coolant pressure boundary and in particular the differing responses of the NRC staff and applicant to Question B.l(b) of the Appeal Boards March 10, 1978 Order pertaining to the most limiting safety related failure.
I hereby certify that the above statements are true and correct to the best of my knowledge and belief.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0YMISSION BEFORE THE ATOMIC SA ETY A?!] LICE." SING APPEAL BOARD In the Matter of
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FLORIDA POWER & LIGHT COMPANY
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(St. Lucie Nuclear Power Plant
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Docket No. 50-389 Unit No. 2)
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NRC STAFF TESTIMONY OF BYRON L. SIEGEL INTRODUCTICN
'This t stimony addresses Appeal Board Question 92 relevant to the analysis of events that would occur between the total loss of all AC power and the violation of either the fuel design limits or the design conditions of the reactor coolant pressure boundary and in particular the differing responses of the NRC staff and the applicant to Question B.l(b) of the Appeal Boards March 10, 1978 Order pertaining to the most limiting safety related failure.1/
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- Cuestien B2 In line with the above discussion, the testimony is to analyze events that would occur beuteen the " loss of all AC pcwer" and the violation of either the fuel design limits or the design ccnditions of the reactor ccolant pressure boundary (or any portien thereof).
In particular, the parcies shculd, if possible, reconcile their differing responses to question B.l(b) of our March 10,1978 order, 21/ or, if not, point up precisely wnere the disagreements lie.
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I References fn 24 reproduced below: 1 Applicant suggests that the first safety related failure encountered would be excessive core heatino due to the loss of water frcm the ccndens:.:e s;;r::p tar, an;.a: -v
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- -: r the loss of AC power (Flugger Af fidavit of "arcn 31, 1975, p. 3).
The staff's judgment is that the first failure would be that of a primary pump seal at about one hcur af ter the loss of AC power ---
resulting in a small loss of coolant accident.
(Fitzpatrickiaricavit of June 12, 1978, p. 11),
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M I have reviewed the testimony of Frederick George Flugger provided in response to Question 82 of ALAB-537 contained in the Applicant's June 22, 1979 submittal.
The Applicant's testimony addresses failwe of a reactor ccolant cump seal thus apcarently reconciling the difference between its March 31, 1978 submittal and the NRC Staff's June 12, 1978 submittal.
The content of this testimony, related to the failure of the reactor coolant pump seals, is consistent with the information pro..ded by F. Fehlau (Technical Administrator for Byron Jackson Pumps, manufacturer of the reactor coolant pumps for the St. Lucie t;uclear Pcwer Plant, Unit 2) at a meeting held on May 16, 1979 between the USNRC, staff and the applicant, Florida Power and Light Ccmpany.
The testimony of Frederick George Flugger concluded that in the event of a staticn blackout, which results in a loss of coclin; water ficw to the cartridge seal assembly of the reactcr coolant pump, an a;:preciable leakage path through the seal assembly to the reactor containment building that could result in a significant loss of primary reactor coolant does not exist. The bases for this conclusion were:
1.
All seal components are captured within the seal cartridge assembly and held tocother by hydraulic and spring forces thereby minimizing the leakage paths.
2.
Each, of the four seals that comprise the seal assembly is; designed to providp sealing against full system pressure.
3.
All the components that ccmprise the seal cartridge assembly, except for the elastomeric U-cups and 0-rings, are made of materials that are unaffected by the elevated temperatures resulting from a loss of coolant to the seals.
4.
Confined 0-rings made of the elastomeric material used on the U-cuos and 0-rings have been used on flanged joints of a reactor coolant pump hot test loop where they have been subjected to temperatures of 550 F for in excess of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
The 0-rings maintained their sealing capability although hardening and permanent set cf the 0-rings, as expected, occurred.
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? ES:2d :'. my ravi2.: cf the infor.atica previded, T agree that the above reisen-provide strong basis for acc2ptance of the cenclusion that a significant loss of reactor coolant through the seal cartridge will not occur.
However, since na test data on the seal' design under expected reactor temperatures and pressures folicwing a station blackout and specifically on the elastomeric seals in the geometry utilized in the seal assembly design is available, we have required that the Applicant perform a confirmatory test on at least one of the four seal assemblies that compromise the seal cartrige under expected blackout con-ditions of temperatura, pressure, and time to provide the additional verifi-cation necessary to determine the adequacy of the reactor coolant pung seal design.
It is my position that the information provided by the applicant in combination with results frcm the confirmatory tc3t, which shcw that the loss of coolant through the reactor ccolant pump seals during the duration of station blackout is not sufficient to adversely affect natural circulation, provide adequate assurance that the ability to ecol the reactor core will be maintainec and that fuel and reactor coolant pressure boundary limits will not be exceeded.
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.i STATP'ef;T OF PROFESSIOCL QUALIFICATIGNS CF EVRON L. SIEGEL I ca a Nuclear Peacter Engineer in the Office of t:aclear Peactor Regulacion.
I have recer.ly been assigned to the Bulletins and Orders Task Force following the Three Aile Island accident.
I am responsible for conducting safety reviews and evaluations for light water reactor emergency core cooling, reactor coolant, and various auxiliary systems assigned to me during the review of nuclear reactor license applications.
I received a Mechanical Engineering Degree in 1955 frcm City Collece of t.ew York.
In 1956 I was a petroleum heater design engineer at Foster Wheeler Corporaticn in Na; Ycrk City, New York.
From 1956 to 1973 I was a Nuclear Engineer in the Nuclear Systems Divisien at the National Aeronaatics and Space Administration (formerly National Advisory Committee for Aercnauticc), Lewis Research Center, Cleveland, Ohio.
My assignments included:
~3 years at the Plumbrcok Reactor Facility during construction phase.
' Experimental and developmental heat transfer research related to nuclear propuision and space power applications.
' Project Manager on contracts to fabricate capsules to test fuel element design for space power applications.
' Design of experiments to be tested at the Plumbrook Reactor Facility.
- Responsibility fer analysis of test data and evaluatien of post-irradiation examinations.
During this time I authored or coauthored approximately 12 reports.
In May 1973 I accepted employment with the Atomic Energy Ccmmission (now the Nuclear Regulatory Commission) in the Reactor Fuels Section, where I reviewed reactor fuel element and assembly designs, thermal performance analyses and operating experience.
From May, 1977 to May 1979 I was a member of the Reactor Systen s Branch where I was responsible for reviewing safety systems on light
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Task Force where I have been performing revicws of licensee and vencor suppi ku infomation to support decisions recarding plant operations.
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