ML20127E252

Responds to Jm Griesmeyer Re Review of Probabilistic Safety Study.Review Summary Including Comments of Substance as Well as Editorial Comments Encl

ML20127E252
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	03/24/1982
From:	Lipinski W ARGONNE NATIONAL LABORATORY
To:	Okrent D Advisory Committee on Reactor Safeguards
Shared Package
ML20127E176	List: ML20127E173 ML20127E184 ML20127E252 ML20127E267
References
FOIA-85-44 ACRS-GENERAL, NUDOCS 8506240493
Download: ML20127E252 (4)

Text

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. Walhe C. Lipir, ski "S

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. ARCONNE NATIONAL LABORATORY 97005odCAssAtax,Anqpw.bos604W - IdqAarc712/W2,4639 l l March 24, 1982 l Dr. David Okrent Advisory Comittee on Reac' tor Safeguards

.U. S. Nuclear Regulatory Comission Washington, DC 20555 t

Subject:

Review of the Zion Probabilistic Safety Study

Reference:

Letter, J. M. Griesmeyer, ACRS, to W. C. Lipinski, ANL, same subject, dated March 10, 1982-i

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F In response to Mr. Griasmeyer's letter of March 10, 1982, I perfcrmed a 2 . review of the ZION PRA. I was not able to review the entire report within the

,T limited available time. The attached review sumary includes comments of

substance as well as editorial coment.s which would improve the readability of j the PRA report.

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Sincerely, 4- - \

ryg g g e i -

l Walter C. Lipinski Reactor Analysis and Safety Division WCLlat

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J. M. Griesmeyer, ACRS j . .

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4 Dr. David Okrant February 22, 1982 i

2. Containment response is analyzed for 21 ae'try states to

the containment event tree leading.to 11 exist states, each associated

. with specific release' category. Exit to entry stat;es are coupled

~ by the. containment matrix'(Table II.5-1, Page II.5-10, Vol. 1). .

Each of the entries in this matrix represents a conditional probabi-lity that a given entry state will result in an exit state correspond-

. ing to a release category. It is shown that the event behavior and the containment integrity are affected by ,

1. RCS presscre at the time of RPV failure.

2. 5 2 venting from RCS prior to RPV failure.

3. Timing of fual melting.

4. Operation of containment sprays and fans.

" ~7ailure of Item 4 or containment bypass (failure of two (2) .

RER isolation valves) . lead' to a dire ~ct release.' Integrated core and containment analysis is performed by use cf MARCH code. As

3. recall from another Class 9 Accident discussion at an ACRS meeting, time scale by MARCH analysis may have sizeable variability. Because of that one does know how significant the structural heat sinks will- .

o be~.in'* mitigating the response.

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E ~ Primary containment' ultimate capacity was determined to.be 4

l'49' psia. It was calcolated' by sargent & Lundy (Appendix 474.1) .

The analysis was supposed to cover: 1) containment structure, 2)_ penetrations, 3) rate of loads, 4) uncertainty bounds, 5) failure mechanisms. Only the first twc5 items are dealt with,in some detail.

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. The Reactor Cont =4=nt i.s in the shape ,of 3'6" wall cylinder with (use Fig. 2), 2'8" thick shallow domed roof based on 9' flat ,

foundation slab. The cylindrical portion is prestressed by a-post-tensioning system consisting of horizontal (hoop) and vertic:al . .

tendons. . The dome has a three-way post-tensioning system.. Founda w s tion slab is conventionally reinforced.- The inside of the entire

.. containment is lined with 1/4" welded steel plate. A cylindricab i reactor pit with a wall thickness of 16 fe't e and an internal diameter i of 21 feet is located at the center of the base mata . c- c .

Sargent a Lundy addressed in detail only the containment build- -

{

. ing internal pressure capability and inubcontracted CB&I'to derive the structural capability of the equipment hatch. .The analysis used materials properties derived from the mill tests and from the 90-day concrete cylinder tests. As the failure mode, it strain in hoo tendon is assumed, and.not.the open flow area of 10 to 15 in2, p as specified by the PRA contracter.

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5 Dr. David okrent , ,3 - February 22, 1982 o ..

Analysis consists of hand calculations and of an axisynsastric finite element computer calculation. It is strictly deterministic and the' conclusion that confidence level of 954 is ass.ociated with the calculated containment internal pressure capability is not supported (presumably based on knowledge of materials property statistics, see Item 4 below for further cea: ment) .

.The hand calculations is one of the type

. Icy A l P= R  !

1

..". where e is the yield stress, A is'the corresponding material I rebar) cross section are'a and R is the radius of the l (tendon, cylinder. This assumes that all concrete'has cracked and that the liner has not lost its leaktightness. The calculated result is 134.4 psig, a reasonable number. .

Finite element model is quite complete. It allows for gradual

, concrete cracking through the wall. Included in the model are:

i mat, cylinder and dome. Account is given for non-linear behavior P -

of steel *~(tendons and rebar) ~ and of concrete (by cracking) .' Se results confirm internal pressure at 14 hoop tendon strain obtained

[ '

by hand calculation. This calculated result and the observed mode

of ductile failure is favorable compared to 1

14 scale model test i of a thin-walled post-tensioned concrete containment. In addition i to this failure mode, transverse shear stresses are examined at i various locati.ons. Of these I believe the location of cylinder i attachment'to foundation mat would have to be examined in greater ,

detail for potential leak path formation. This is because at 14 hoop strain in the cylinder, the radial displacement in the cylinder is of the order of 3 to 7 inches greater than that at m the mat. Also, thermal effects are written off as insignificant, -~ ~ 7 j however, gross thermal expansion differences- (between the cylinder

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wall form and the mat) leakage path.ma,y ,cause severe transverse shear an1 potentially,,'l.

containment wall arourid the penetrations (including the equip-ment hat'ch) is significantly thicker than in the cylinder, hence .

stronger for the failure mode postulated. CB&I analysis of the

i. equipment hatch is well organized and provides credible number of l . 134 psig for ultimate buckling capability of the hatch dome. The effect of the projected ultimate internal pressure on the hatch boundary is not considered in the equipment hatch analysis. Poten-l tial for leak path at the hatch boundary has to be evaluated with l realistic boundary conditions.

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Dr. David okrent ,

February 22, 1982 ,

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i. Containment structural capability analysis does not address .-

the systsems' and structures attached to the containment. wall. Three -

. to sevpn inch radial. displacement of the containment wall will definitely cause some distress at the wall attachment of these.

. While 149 psia appear to be a reasonable number for the cylinder, containment bypass . potential as a consequence of some penetration -

breakaway prior to this pressure '.has not been included in the study.

! 3. Reactor pressure vessel faliure dynamics are addressed in l Section II.5-44. The treatment appears to be reasonable and the i f ailure of an instrument nozzle in the bottom head is the likely l first. , ,

4. Seismic fragility of various . safety. related structures and equipment (Section II.7-2) was determined by structural Mechanics l Associates (Section 7.9.2). .The approach taken was to identify the design basis acceleration and the factor of safety for each item ,

e, valuated.- This safety factor was then decomposed into various l contributing ele:nents, . each of which was analyzed for its variabi-lities and uncertainties. These variabilities and uncertainties were then combined to get the variability and uncertainty of the safety factor and 'of the acceleration capability for the item. -

This approach ' appears to be rational and result:.ng families of fr' agility curves should be credible. I believe the question of internal pressure ultimate capacity should have been subjected

.to a similar treatment, leading to family of curves such as Fig. II.7-2, rather than an single curve, such as Fig. II.5-5,

. which has no justified confidence level associated with it.

. 1 e . . s e Very truly yours,

. _e

'WM ons 2ndans

ces Senior Vice President -

cc: Dr. J. Greisneyer, ACRS ,

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