ML19210C776

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Summary of 79627 Meeting of ACRS Subcommittee on Floating Nuclear Plant,In Washington,Dc Re Review of Proposed Magnesium Oxide Core Ladle Design for Floating Nuclear Power Plant
ML19210C776
Person / Time
Issue date: 08/25/1979
From:
Advisory Committee on Reactor Safeguards
To:
Advisory Committee on Reactor Safeguards
References
ACRS-1651, NUDOCS 7911200091
Download: ML19210C776 (33)


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MINUTES CF THE ACRS SUBCOMMITTEE ON g

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'T. hv THE FLOATING NUCLEAR PLANT (h.

WASHINGTON, DC

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5 JUNE 27, 1979 s

Th( A;RS Subcommittee on the Floating Nuclear Plant met with representatives of the NRC Staff and Offshore Power Systems (OPS) in Washin; ton, D. C. on June 27, 1979, to continue its review of the Offshore Power Systems applica-tica for a manuf acturing license for the Floating Nuclear Plant. The specific topic of the meeting wa., the review of the proposed magnesium oxide core ladle design. A notice of the meeting appeared in the Federal Register on Jur.e 12, 1979 (Attachment A).

A copy of the detailed presentation schedule is attached (Attachment B). A list of attendees at the Subcommittee Meeting is attached (Attachment C). A list of documents provided to the Subcommittee for this meeting is attached (Attachment D). There were no public statements either written or cral. The entire meeting was open to members of the public.

MEETING WITH THE NRC STAFF AND OFFSHCRE POWER ? STEMS (OPEN SESSION';

1.0 Subcc=ittee Chair an's Opening Remarks Dr. Siess, Acting Subcommittee Chairman, introduced the members of the Sub-committee and noted the purpose of the meeting was to review the proposed core ladle design. He pointed out that the meeting was being conducted in accordance with the provisions of the Federal Advisory Committee Act and tne Government in the Sunshine Act and that Mr. Gary Quittschreiber was the Designated Federal Employee for the meeting. He statec that no requests for oral statements nor written statements from members of the public had been received with regard to this meeting.

2.0 Introductory Remarks 2.1 Status of the FNP Application Mr. Blair Haga, OPS, briefly discussed the FNP manufacturing license applica-tion status noting tne folicwing points:

o Application was made in early 1973.

)3bb b

Atomic Scfety and Licensing Board hearings are essentially complete.

o The ACRS must complete its review before the final hearings can be completed.

7011200 f

FNP Meeting June 27, 1979 o The OPS core lad'e report has been amended to incorporate the NRC Staff's comments.

o OPS has answered all the NRC Staff's questions with regard to the ECCS; however, there are still some questions with regard to the McGuire appli-cation which involve a complete review of the ECCS/UHl.

Mr. Haga noted that OPS has formed a task group working closely with Westing-house to evaluate the Three Mile Island Accident. He noted that it would be several years before an FNP was manufactured and that any TMI-related items on the FNP could be handled during the manufacturing period.

Mr. Haga stated tha,t.the FNP core ladle is a direct result of the liquid path-way study reviews with the ACRS. He noted that the Final Environmental State-ment, Part 3, imposed the requirement for the core ladle on the FNP to mitigate environmental consequences. Since this is an environmental issue, they do not intend to do an exhaustive tett and design effort for tne ladle, but will apply the best available technology, which is the approach identified in the NE?A.

Special emphasis is being given to ensure that the core ladle will not compro-mise the existing public health and safety requirements.

2.2 Status of the NRC Staff Review Mr. Ralph Sirkel, NRC Staff, discussed the status of the NRC Staff's review, noting that they have finished the environmental review and have issued Final Environmental Statement, Part 3, which requires the concrete beneath the reac-tor vessel be replaced with a pad constructed of magnesium oxide. He noted that OPS submitted Tepical Report No. 36A59, FNP Core Ladle Design and Safety Evaluation. The NRC Staff is new evaluating Topical Report No. 36AS9.

He noted that the NRC Staff has not reviewed the structural aspects of the core lau.. design due to shcrtage of manpower resources.

Mr. Birkel noted that the NRC Staff has about 10 items to finish in its review and that if adequate priority is given, they shculd finish the review and issue SER Supplement No. 3 some time during the summer of 1979. The safety aspects of the review were covered in a 1975 SER. The present NRC Staff core ladle review will be performed to ensure that the changes being made to accommodate 1366 02I

FNP Meeting June 27, 1979 environmental requirements do not have any deleterious safety implications that would change or void any of their earlier conclusions. He added that the NRC Staff feels that the results of the liquid pathways study was a key-stone to the FES, Part 3, which reflects environmental conditions which the NRC Staff is obliged to address.

Mr. Birkel suggested that in order to effectively benefit from the Subcommittee Meeting, that the ACRS issue a letter to Mr. Gossick on the proposed core ladle design, similar to letters issued on the liquid pathways study.

The Subcommittee discussed the difficulty in separating environmental from safety concerns. Dr. Siess noted that the FES discusses a number of benefits from delaying a melt-through and that some of these benefits relate to safety.

Mr. Haupt, NRC Staff, noted that when the question of whether it was appropr#iate to include the Class 9 accident in the FNP Envircnmental Report was brought before the Comnission, they supported the NRC Staff; hcwever, no order was ever issued, possibly because OPS agreed to incorporate a core ladle, making the order unnecessary. Haupt indicated the core ladie is an environment'l matter simply because it was covered in the environmental report. Mr. Birkel noted that even without the core ladie the 10 CFR 100 calculated doses are acceptable for the FNP.

2.3 NRC Staff's Evaluation of the CPS FNP Core Ladie Mr. Andrew Marchese, NRC Staff, discussed the background of the NRC require-ment for the core ladle and also provided a brief status report cn their evaluation of the ladle. He noted that CPS, in Topical R port No. 36A59, confirmed and concurred with the NRC Staff's conclus, ion in FES III, Appendix E, that magnesium oxide appears to be the most promising candidate for the refractory material of the core ladle and that there was reasonable assurance that a del'ay of melt-through of molten core debris on the order of two days was feasible. Marchese stated that based on their review, so far, together with the Applicants commitments, the Staff is of the opinion that OPS has met the FES III requirement in the area of delaying core melt-through.

The NRC Staffs' *best estimate time for the proposed design for delaying a full core melt-through is 5 to 8 days.

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FNP Meeting June'27, 1979 Mr. Marchese discussed several NRC related R&D informational needs concerning the interac. ions of the molten core debris materials with the refractory sacrificial materials (See Attachments 1 and 2). RES is considering the NRR research request on this item for ccmpletion over the next three years.

Mr. Etherington noted several items discussed earlier which he thought needed further consideration. These included:

o Thermal shock of magnesite o Use of magnesite in ladle lining o Bonding of brick without mortar Applicability of steel-making experier.ce to core-melt o

a Water intrusion into the magnesite Expansionofbricksatthejoints o

o Lack of ventilation of the concrete i

3.0 Technical Presentations by Offshore Power Systems 3.1 Summary of the Core Ladle Design Requirements Dr. Oee Walker, OPS, discussed the CPS interpretation of the FES III functional requirements for the core ladle design. OPS has transformed the FES III re-quirements into the folicwing set of functional requirements:

1.

Maximize debris retention time within available space constraints a) Minimum delay time - about 2 days b) Minimum refractory thickness - 4 feet 2.

Ladle materials shall not form a large volume of gas which can sparge through melt.

3.

Ladle volume shall be sufficient for core materials and 25". Icwer vessel steel.

4.

Ladle design shall not compromise the integrity of the containment boundary or platform.

5.

15 mrem /hr radiation limit retained for adjacent compartments outside containment (during plant operation).

Dr. Walker discussed the method they used to calculate the core melt debris volume estimates for sizing the ladle. Details of these calculations are 1366 923

FNP Feeting June 27, 1979 provided in Attachments 3 and 4.

He also discussed the properties of candidate materials for the core ladle, including magnesium oxide (See attach-ment 5).

3.2 Description of Ladle Design Mr. Clint Dotsen, OPS, provided a description of the ladle and discussed the location of the ladle within the FNP (See Attachments 6 & 7).

Mr. Dotsen presented a h inch to 1 foot scale model of the portion of the FNP including the ccmponents around the reactor and the ladle. A scale model of the entire FNP, which is about 35% complete, is available in Jacksonville, Florida for inspection.

Mr. Cotson defined the following design requirements they have established for construction of the ladle.

1.

Incorporate a ladle into the existing FNP design with minimum alterations.

2.

Alterations to the reactor cavity shall not compromise safety requirements including:

a) Structural integrity of the platform shall be maintained for all operating and design basis conditions prior to a postulated core melt accident.

b) Water-tight redundancy shall be maintained between the basin and reactor cavity.

c) Radiation shielding requirements shall ce maintained.

3.

The platform structure shall withstand loading conditions for the duration of core melt debris retention.

4.

The reactor cavity structure shall not beccme the weakest link of the containment pressure boundary as a result of the addition of the ladle.

5.

The ladle configuration and material shall not compromise other safety requireemnts.

6.

The ladle shall be as thick as practicable within the various constraints but shall not be less than four feet in any direction.

7.

The ladle cool volume shall be sufficient to contain the molten core debris during continuous basin motions (1/2 ).

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FNP Meeting June Ei, 1979 8.

The ladle shall be designed and analyzed to remain functional for operating-basis environmental conditions. For more severe condi-tions, the plant can be shut down for inspection of the ladie.

3.3 Boundary Constraints Mr. Dotson discussed the boundary constraints for the ladle. He noted that the space beneath the reacter vessel available for placement of refractory material for retarding the debris is bounded by the containment boundary pressure bulkheads, the platform primary strength bulkheads, and the incore instrument cables (See Attachments 8 & 9).

3.4 Configuration and Arrangement Mr. Dotson discussed.the configuration and arrangement of the ladle. The size of the ladle was selected to utilize the available space within tne various design constraints and still maintain a thickness of not less than four feet in any direction, with a pool volume sufficient to accommodate molten core debris of approximately 920 cubic feet. Attachments 10 and 11 provide the details of the configuration of the ladle and the arrangement of the magnesium oxide bricks.

Mr. Dotson discussed the comparison of a configuration of an existing furnace design and the proposed ladle. The arch design and interlocking brick were used in both.

Mr. Dotson noted the FNP ladie brick will be installed infide a steel box (See 2). The steel box will provide waterproofing and a gap between the ladle and surrounding concrete for venting gases reler. sed from the concrete.

Dotson indicated that experience in the steel industry shows that locking of the brick from thermal expansion should minimize or eliminate the possibility of the bricks coming loose or floating out, especially considering the tongue-and-groove arrangement.

OPS has suggested that a 14-foot brick wall be erected directly above the ladle and around the entire perimeter of the lower cavity. This brick 1366 a25

FNP Meeting June 27, 1979 wall will shield the concrete from the thermal radiation produced from the melt debris ar.d will limit the primary platform steel temperature such that the structure can withstand the expected loading conditions for the duration of the debris retention. Attachment 13 shows the proposed location of the vertical brick wall.

3.5 Platform Motion and t.oading Conditions In response to cuestioning from Dr. Siess concerning the design basis for the volume of the ladle, Dotson noted that it would take a little more tnan 1/2 degree tilt at the design basis debris volume to spill over the inner edge of the ladle. A 1/2 degree tilt is a design basis to occur no more than 10'. of the tic.e at the worst site under normal wind and wave conditions.

It would take a 2 degree tilt, a 100-year condition, to get to the upper lip.

Mr. Dotson discussed the loading conditions for the reactor cavity structure and core ladle during normal plant operation and design basis events (See 4).. He also discussed the loading conditions for the reactor cavity structure and core ladle subsequent to a postulated reactor vessel melt-througn (See Attachment 15).

The Subcommittee discussed the possibility of large amounts of moisture entering the steel box surrounding the brick and the possible effects on the perfor-mance of the ladle. The Applicant does not consider this to be a significant problem.

3.6 Melt Penetrating Calculations and Gas Generation Estimates Dr. Henry Stumpf, CPS, discussed their estimate of magnesium oxide erosion due to a molten pool. They assumed a fixed fraction of decay heat directed into the magnesium oxide bed in both the horizontal and vertical directions from 10% to 100%. Other assumptions were uniform heat flux in the horizontal and vertical dir :ctions and no conduction into the magnesium oxide aneac of the melt front. The heat source used assumed:

o Entire core melt o 20% of decay heat lost through volatiles o Entire heat from :irconium-water reaction was used.

1366 n26

FNP Meeting June 27, 1979 Erosion depths are shown on Attachment 16, showing 6 days penetration time for the eight-foot thickness assuming 100% of the heat being directed into the melt.

In response to questions concerning erosion rates, Mr. Arthur Chait, Harbisen-Walker, noted that the bottoms of ladles, which had been contact with liquid metals, typically erode in a uniform fashion. He indicated that he had not experienced rapid erosion, as might be expected with a core melt. Mr. Chait said that with uniform erosion he would expect wear to include a ccmbination of slagging and melting.

In response to a question frca Dr. Siess concerning the expected ratio of heat into the melt versus radiation upward, Stumpf expected abcut 80% of the heat to be directed to the melt. He expected very little heat to the surrounding walls after one or two hours, once the wall temperature reaches a temperature near that of the pool.

Mr. Stumpf discussed outgassine of the basaltic concrete beneath the ladle and on the sides. He indicated that you would probably begin to outgas the con-crete long befcre the melt front reaches the concrete. The gas frcm the con-crete would be vented out the side of the ladle so it would not bubble thrcugh the molten pool. Stumpf noted that if you outgas all of the concrete under the ladle, it would increase containment pressure roughly 631 psi. He added that outgassing the walls surrounding the pool and above the pool would add another 9 or 10 psi to the centainment pressure.

Mr. Bob Brusoloff, CPS, indicated they have not calculated the containment temperature as a result of a core melt accident.

Mr. Haga said this was not being addressed since the ladle is a requirement of the liquid pathway study. The Subccimiittee expressed some concern that delaying the melt-through may increase the chance of earlier containment rupture and higher airborne releases.

Mr. Stumpf ciscussed their calculation of the thermal effects of radiation frcm the pool to the side walls. For this calcualtion, OPS assumed that the heat radiation from the pool was equal to the total decay heat being generated 1366 027

FNP Meeting June 27, 1979 at that time. As the walls heat up they will radiate heat back to the pool so that the net transfer is small once the walls reach equilibrium pool surface temperature. Attachment 17 shows expected wall temperatures of brick and concrete. Stumpf noted that since the melting point of basaltic concre+,e is around 26500 R, and since the expected temperature due to upward radiation is above this temperature, that ull surfaces in the cavity will have to be lined.

Mr. Marchese noted that the staff is evaluating the advantages and disadvantages of pumping water into the core melt debris. He noted that an advantage is that it would remove heat but the disadvantages are steam explo-siens and sparging of radioactivity. He added, that the decision has not been made whether ts require the operator to tur off ECCS pumps once it is known the core has melted.

It was noted that since the sump water contains,

the activity sparged by the containment spray system, pumping the sump cnto the debris would likely add to the air release and to the liquid release; however, there is also the possibility that it would cool the core and prevent burn-through.

In respcnse to questions from Mr. Ctherington concerning the ability of the structures to support the vessel following a core melt, Haga said they are only concerned that the vessel is supported for the specified length of time until burn-through of the platform occurs.

Mr. Stumpf discussed containment pressure build-ug due to outgassing of the concrete (See Attachment 13). He estimated tr.e entire decomposure of 2C00 cu. ft. of basaltic concrete around the vessel and 3000 cu. ft. on the sur-rounding wa.ls would add 16 psi to the containment pressure.

3.7 Ocse Estimates for Regicns Surrounding the Ladle Dr. Walker discussed the dose estimates during normal operation for the lower cavity walls and occupied spaces belcw the cavity. He noted that the design dcse rate limit is the same as it was before the ladle was incorporated in the design. The calculated dose rate beneath the ladle with the current design is 6 mrem /hr. He added, that there was a margin of about two feet of Mg0 which could be removed and still meet the 15 mrem /hr limit.

1M6 328

FNP Meeting ine 27, 1979 3.8 Related Experience with Refractory Materials in the Steel Making Industry Mr. Chait discussed some of Mr. Etherington's earlier comments. He noted that magnesite is not normally used in ladies in the steel industry. He added that magnesite is very sensitive to thermal shock; however, a single dump would only cause thermal shock to the initial surface. The FNP ladle has chemically bended brick which will take the initial thermal shock.

Mr. Chait discussed his experience involving inspection of ladies and furnaces following use, noting that even thcugh the bricks crack, they do not float uo on top of the melted metal or f all out wnen the ladle is turned upsue down.

Mr. Noble Seaberne, CPS, discussed experience from steel-making furnaces and a

operaticns and related this experience to the FMP ladle. Seaborne discussed comparable sizes, drop heights, and temperatures in the steel-making industry with core-melt debris in the FNP.

4.0 Miscellaneous Discussion The Succommittee discussed scme of the Sandia comments on Tcpical Report 36A59 which were presented in a letter dated May 23, 1979.

In response to c:mments on the Sandia letter. OPS and Harbison '.ialker re:resentatives noted the following:

o The decay heat frcm complete oxidation of the Zirconium is assumed o The decay heat frcm chremium oxidation was not considered. OPS agreed to look into this, o Examination of samples fr0m industrial pr:casses which are at lower temperatures than a core melt, shows that chemical attack does play a signi-ficant role in the failure of magnesium oxide bricks.

o There is a lot of experience that would indicate that flotation of the brick will not be a problem.

o The ratio of the total weight of Mg0 to UO2 is about 100 to 1.

The ratio of the total weight of Mg0 to molten steel is about 750 tons to 200 tons.

Experience has shcwn that refractories that have slagged and been altered o

were demagcd during cooldown; therefore, inspecting cooled bricks for damage may not be representative of the' conditions cccurring during a core-melt accident.

1366 029

FNP Meeting June 27, 1979 5.0 Conclusions / Remarks Mr. Haga discussed the importance to CPS of obtaining a manufacturing license for the FNP. He noted that they are preparing proposals to be submitted this summer to utilities which have expressed a very strong interest in building an FNP. He did not think that any utilities would make any decision on the FNP until such time as OPS has a manufacturing license.

Dr.Siess discussed Mr. Birkel's request to have the ACRS perform a review of the preliminary ladle design and to write a~ letter to Mr. Gossick ccamenting on the design before the NRC Staff finishes its review. Dr. Siess said he would discuss this request with the ACRS at the July 1979 ACRS Meeting.

If the ACRS agrees, OPS'and the NRC Staff would likely come to the August 1979 ACRS Meeting to discuss the preliminary design of the core ladle and any i

other liquid pathway study related items.

The meeting was adjcurned at 3:25 pm.

    • r2.***********,,

For additional details, a complete transcript of the meeting is available in the Nuclear Regulatory Commission Public Occument Rocm, 1717 H Street, N.W.

Washington, D. C. 20555, or frcm Ace-Federal Reporters, Inc., 444 North Capital Street, N.

W., Washington, D. C.

1366 030

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33~11 idktified m the initial session have be considered dunn; the meeting and to

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bc sdequately covered and whether formulate a report and fu=c s. :sra.

4 the p iect is ready for review cy tYe recommendutions to the full committee.

r full Cdnmittee.

At the cun.lusion of me F.aceutive

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g Session. the subcom-uttee will hear e

A rink.tuociete Admins nterfer Errernal Furthkinformstion re;urdinf coics to be diseassed whether the =eettng presentsuuns by and noid discussions ini o l -imes rue m.rm us 4 if" jt,/,tw,.

has been c teviled or reschiduled the with represe ttutives of the NRC Staff.

l eaa:>.o 3ces m es.u -

1 Chairman's tulinq on requpsis for the Cffshon Power Systems, et :! and their j

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opportunity tA; resent oraf statements cutisuit:nts, perunent to das r: view.

and the time spotted ther'efer'e can be The Subcom:uttee may then csucus to j

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obtamed by a prkpoid 'elephone call to determine wheder de matters t

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been adequately cove cd.

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'.f Sr the Sube=rn=::tes to hold One or more f

Safeguar:3.S te:mm;; n on c!csed sessions for de purpose of P:ticleg:c:J Effects and Site Cated:[

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lohn C. Hof,e, information. I have determined !n The ACRS ubc:==itte's en Advisog 'Cammittee. Atene;em et Cfiest.

ae:Ordance with Subsecuan IC(d) =f Public !.sw 9 -4:3. thst. should such -

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sessions be equired.it is necessary to wdl hold cn c}en meeting:n June 27.

smyc ceos tsnme c!cse dese sesmons to protect 15 3 In Room 1c46.1717 H St N.W.

propnetary information (5 U.S.C.

Washington. C'C. 3333 rd dis =:ss changes in the.tRC rese..:h pt: gram Acvisery Ccmraitte.e en Reacter 53:b(c'(4)).

budget and sev$ral oth:f=st:sts in the Safeguares, Sub:cmmittee en the Furder in!:rm::!:n n;sti. g t:;ics a:vas of raiolch:sl effec's and si:e Ficat:ng Nuclear P! ant; 2.!eeting to be d;scussed. whether the mes:ing as bun ganceM u nschedu!ad the ev:Ination.Neu: of dis meetig was The ACES Sub::==1ttee en the airman s @ en nwn:s the published un Mat 24.1373 (44 FR 017-').

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= apes:um oxide bncis. Nonce :f this can be.,cund in d::umen:s en f.!e :nd c:nsdran:s. and Stsif. Pers:ss desi.ng

=ceting was pubi'shed on May 24.1373 avsdable fut ;ubuc ins;e :::s at the to =aia oral s:s:sn-n:s sh:uld neufy (44 FR 0,017-)~

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Street. N.W, W:s:: inst:n. CC 0333 :nd in adva 'e ss ~sdi:sbie so '~'N.a'.....de Cctober 4.1973. (43 FR 43:31. eral :-

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kept. and questions =:y be asked emy x; cg;;g,)

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a Cated:Jeea m LOC a.=. urdil conclus;ca cf business.

censultants, and S:sff. Pe sens destnng The Subce:Emittee mskmeet in to make orsi statements shculd not2fy John C. Hoyle.

F.te:utivs Sessica.with cny fits the Cesipated Feders! Emplevee as f::

Ad"*JrY C3mmute' AtencWr: ear C$cen censultant.9who may be ;- esatt. to in advance as pesc:!:sb!c so that 7t Dama r.:

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.. ge c:nciunon of the Exe'$uuve Wednerdey.fune,,,,..,srs Proceder:1 Cekn Sn J;n. ce Sube:mmitice will hes:

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tens ';y sad hc!d dwaisions 8:00 s.m. until be cenclusten of June 5.:21.

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men: *o tais review wnh f.\\

business.

As directed.T he C...er Cecyiq Motion to r I:ene Pr:chfin~[s.ss r ;re: n:sttves of the NRC stait.:nd The Subcommittee may rr.cet in d

on Apel.1977. the United S tes in.1:d s;e:Lers frem outstue NRC-\\

Executive Session, wqh any of its i 9 Sub cetmittee may then ::uca;s to consultanta who may ne ;resen:. to Posts! prwce submitted a testhenial est.-mme wncther the ma explore and exchan:;e their prehmmary filing,m this cocket en May 31.1:73. A b

"P'*i "* 8di"8 "3"c *n'ch should pr.eminary exummsuon af the Se'. < ice s D'((

l0D' filies reve i that it :entains tec3=csily 1366 031 nn I

I A

PRESCCATION SCHEOULE FLCATING NUC*.FM PLVC SUSC2'?.:~ TEE idee *DG q

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JUNE 27, 1979 g

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La P R F.5 C C A * ! C J APPRCX; MAT T?.*A.1.*.

'T.'..t..'l*

MEE DG WI*9 CFP5HCRE PCER SYSTEMS AND ~~dE NRC STAFF (CPEN SESSION) 1.0 Scheercittee Chair =ans Opening Recarks S:45 a=

~

2.0 Introduccory Recarks 2.1 Applicants 5t=: ary Of Fun :icnal Oesign 15 =in 8:50 a Requirecents and M.aterials Selectica 2.2 NRC Staff Discussi:n cf Review Schedule 5 min 9:20 a=

and Major Pr:'clecs in acview, if any 3.0 Technical Presentations by Applicant with NRC Staff P.sspcase W

3.1 Cescripti:n of Ltdle Casign 3.1.1 Cesign C:r -=ints 10 min 9:30 e-3.1.2 Cenfiguration 10 min 9:50 a.

3.1.3 Ladle Support 10 min 10:10 a 3.1.4 Platfer= Struccural Censideratien 10 min 10:30 a:-

Coffee Break 10 min 10 : 50 a--

3.1.5 Platfe:m Moti:n and Seismic cesign criteria 10 min 11:00 a-3.1.5 Miscellane:us Considerati:r.s 10 min 11:20 e.

gju ^ F 3.2 t

Mel: Penetratien Calculati:ns and Gas Generati:n 20 =in 11:40 a-Estimates hecJ.3 Cose Esticates f:r Regien 3eneath Ladle 10 =in 12:20 r Break for I.unch 12:40 - 1:40 r 3.4 Ciscussi:n of Related Experience with Refract::-(

45 min 1:40 :.

$ erpt.t'G Materials in Steel Making Industry p

~

4.0 Caucus 3:10 I:-

- Cenclusier.s/Recarks

- Discuss Future Meeti:gs Schedule Adjeurnment 3:20 p Note:

(1)

A -taxi..um of 30 =inutes will be ell:wed for receivi.q oral statements fr:m =ecners of the public if recuested.

(2)

The speakers sheeld limit their prepared presentationshh ?)2 to the time allowed. .an allewance, amounti.g to ICC'i of the presentation :i:2, has been made for questioni.g _, -w. c.x- .,I I w e

9 ACRS FNP SUBCCMMITTEE MEETING JUNE 27, 1979 WASHINGTCN, DC ATTENCEES LIST r.: 5 NRC C. Stess, Acting Chairman R. Sirkel

r. '.n w en, Memeer C. Haupt

..watnis, Me ter A. Marchese H. Etneringten, Member M. Silverberg

1. Catten, Ccnscitant G. Caittschreiber, Staff"
  • :esignated Federal Employee El Harbisen - Walker Refractories B. W494 A. Chait D. dalker C. Cotson

'tu pf H. s 'i. Sea::cene K. Bruscicff E. Capo Sandia Labcratcry D. Powers 1366 '33 smr.- c

DOCUMENTS PROVIDED TO THE SUBCCP14ITTEE FCR THIS MEETING 1. Topical Repcrt No. 26A59, dated April 1979, Offshore Pcwer Systems FNP Core Ladle Design and Safety Evaluaticn 2. Sandia Report, dated May 23, 1979, Review of Informaticn Needs for Cesign of a MgG Core Retenti:n Oevice. 3. View graphs shcwn at the meeting are provided as Attachments 1 through 13. A complete set of all handouts are provided in the meeting transcript and in the ACRS Office file for this meetieg. Ob$ $k b b / 1366 034 ATTACNMENT D

8, fiODI 00E IBI'6 m1 ERIE liflEREfl0l6 Wlill IfflEf0W SACRIFICIM.MIERIMS R R. D INfDIETIfNN. tEEIS MlE #0 EXIDff 0F W[(ELT IEN.TMTION INTO llE SERIFICIN. MTERIN., INClJJDING IDEIMTION liff0 CM0(S IEnfB1 BRIO (S OF llE SERIFICIl m1ERIO EXIIRT OF llEIEL Sil0CK CMCKING #fD/0R SPALLATION OF TIE SACRIFICIR mlERIO OlMfilY #0 GMOSITION OF VNDRS #0 CASE 0l6 Pf011X. S IElfASW lT01 TIE T ~ SERIFICIN. MIERIN_; IIIGI-llMMlUIE TllEIMFilYSICAL PROPERflES, ESPECIALLY llE lMST PELTING POINT ElJIECTIC llMMlU[ OF llE SERIFICIAL mlERIAL AFIER EXfDSilE TO CDE KLT IEBRIS2 8'.g ~d

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C^ ,s PROPERTIES OF CANDIDATE LADLE MATERIALS l bblCh BILITY Li I T llG S' IFIC q{ F h kkIf t/s C) 3 o (can ) (cat /cn ) iiATERIAL t ALUMIlluM 0X1DE 2037 fl.0 0.272 256 2939 i 6kAPillTE 2760 1.9 0.l58 2218 HAGilESillM 0XIDE 2852 3.5 0.313 112 8 11168 SILICA 1728 2.32 0.266 31 1065 Til0RillH 0XIDE 2800 9.95 0.07 72 2till8 IIIAltlUM CARIllDE 3076 11. 8 0.207 283 3976 URAlllUM OXIDE 2815 11.0 0.07 71 2675 ZlRCONIUM 0XIDE 2760 5.7 0.155 169 3'116 D p 9 I x s

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