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. DeYeurg, Assistant Director for Light Water .teactors, "'

.:Ui3T FC:i ADDITIC.ML I:iFLPETI3 ('.;-3) FM T! RIE * ILE ISU...] ..;0LI.'?JTATIO.i, U''IT 2.

Flant riana: Three Mile Island .;uciaar Station, Unit 2Jcclet io.: 50-323Stilestene ..o.: 24-04Licansing Stage: CL|1$35 Supolier: Eatcock 3 ..'ilcox.ircnitect Cncineer: Surns 5 'loeOcntainuent Tg e: Dry.lcs;onsi:,le Cranch a Project Manager: L..A , 3 ranch ::o.2, :i. SilverAequested Ccapletion Jate;deview Status: Awaiting Inferration

The enclosed request for additional infomation (Q-3) for the Three ' tileIsland iluelear Station, Unit 2, has been prepared by the Contain. mentSystems Branen after having reviewed the apcropriate sections of theFinal Safety Analysis .?eport (FS.u), as amended, up to and includin-/-nendment 33.

The following is the status of issues that reain to be resolved:

1. The applicant has changed the reacter cavity desi;n to eliminatemovable obstructions to vent flor, the anglicant procoses to installa fixed neutron snield structure over the reacter cavity. !!e will

require additional analysis and infor ation resarding the :.odelineof tne neutron snield structure in order to cocolete our review andconfir-'atory reactor cavity aralysis. The informtion needed is,

included in tne attached request for additional infor-ation.

2. The applicant has not responded to a previous recuest for additional.infor ation (question 21.50) regarding tne nain steam line treakaccident.

3. The acclicant has not ccm.itted to wrfom the contairsent inte; ratedleakaqe rate test at the maxinum calculated containment pressura of55.7 psig as required by Aapendix J to 10 CFR Part 50 We vill requira

the applicant to cm: ply with Appendix J.

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3. Ti.e apolicant das advised us t:ut a rajor revision to Table 5.2-15"Ccatair ant Isolation Valvcc," uill ;e sutritted in a later mena-.ent. TNeefore, car cor.clasions in tne Ca fety Evaluatico '.ecortreariinn t1e contain.. ant isolation syste are subiect tc chany.

6. T% anolicant has not provided an accootable respence to cur incairyinto t':0 proccsad u.2 of tne centain ent curge system durin onrral00 era tion . :ur pc:ition his been stated to the a:clicant in ivestien04.16. . c. ever, the a ?licant has not a.'e mtalv di;cusseI Tc s 9

presant syste desi~n satisfie, cr ches net s tisfy Cracch T civic 21Position C33 S-4. The restense appears to cugg::t that unli .ite?Ourge sj: ten operation will be neces: art'. !? this is the case, t:.aapplicant aas c.ct discussed the plart desien :eculiarities t.,at willnecessitate unlir.ited purge syste, operation.

Orir,ini mned Dnobert L. Tede5C

?.obert L. Tadesco, e ssistant ''irtetorfor ?lant Systems

Division of ..jstrs Saf at;-

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cc: d. l'eine an5. hanauer.i. Coyda. McdonaldK. Kniel'i. LainasH. SilverS. VargaJ. SnacakerJ . ".udrickF. 21twila:. il;mn

Distribution Copies: Docket File; iiRR Reading File; PS Reading File; CSB Reading, File _

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REOUEST FOR ADDITIONAL INFOR:L' TION(CONTAI'.7ENT SYSTEMS)

THREE MILE ISLAND '.UCLEAR STATION UNIT 2LOCKET NO.- 50-320

042.0 CONTAIN!ENT SYSTEMS BRANCH

042.17 Provide the following information for the reactor cavity subcompartnent(6.2.1)

analysis:

1. Provide sufficiently detailed plan and section drawings for several

views showing the general arrange =ent o f the reactor cavity struc-

tures, components, piping, and other major obstructions such as the

proposed neutron shield. These drawings should identify all subcom-

partment nodes and flow paths.

2. Provide and justify the values of the vent loss coefficients and/or

friction factors used to calculate flow from the top of the reactor

cavity and around the proposed neutron shield.

3. Page S3-42-3d of Supplement 3 to the FSAR indicates that the neutron

shield is designed to withstand the differential pressure that cay

develop across it. Provide an analysis of the differential pressure

across the neutron shield, and compare the results to the design

capability.

4. Previde tha resultant loadings on the reactor cavity structures,

reactor vessel and vessel supports, and compare them to design values.

We note that information pertaining to the shield plugs, which no longer

will be used in the TMI 2 design, has not been re=oved from the FSAR. Since

"we do not agree with the analytical model presented in the FSAR to analyze

their renoval under postulated accident conditions and since they will not

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be used, any information related to the shield plugs should be deleted

from the FSAR.

042.18 The response to 042.7 regarding the =ain steam line break accident is(6.2.1)

ince:plete. Provide the following information:

1. Identify the equipment and cceponents relied on to limit the mass

and energy released to the containment following a main steam line

break. Specify the design criteria for this equipment and compon-

ents.

2. For each case analyzed above, identify all suurces of mass and energy

and the time periods during which each source is added mass and energy

to the containment.

3. Provide a tabulation of the results of the above analyses, including

the =aximum containment pressure and temperature, and tinc(s) of

occurrance, and specify the active heat removal equipment assumed to

be operable..

4. Graphically show the containment pressure and temperature responses

for the cases analyzed.

042.19 No response to Question 21.50, which also is concerned with the sain steam_

line break accident, has been provided. Provide a responre to this question.

042.20 In response to Question 042.10 it is stated that the containment atmosphere

(6.2.4)is monitored for the hydrogen content by means of a local sampling station.

No justification is given to show that the operator-will have enough tire

to analyze the sample and take action before the concentration of hydrogen

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approaches the 4% limit. Justify that the proposed sampliag technique is

an acceptable way to onitor the hydrogen concentration within the contain-

cent following an accident.

042.21 The response to 042.11 is unacceptable since it is not clear what is meant(6.2.4)

by the state =ent that another recombiner unit will be stored locally.

Specify whether this hydrogen recombiner unit will be stored on site. If

you intend to share reco=biners between nuclear power plant sites, discuss

your plans for transporting the shared reco=biner.

042.22 Describe the instrumentation that will be provided to monitor the perf or nace

(6.2.4)of the hydrogen reccroiner to assure that it is performing its intended

function. It is our position that such instrumentation sheuld be provided

and that readout and alarm capability should be provided in the control

room.

.

042.23 The response to Question 042.15 is unacceptable. Provide justification by(6.2.4)

reference to statements in Appendix J that certain containment isola tion

valvce (listed in Table 6.2.15) need not be Type C tested.

042.24 Provide the analyses identified in Ites 3.5 of 3 ranch Technical Position

(6.2.4)CSB 6-4, "Contain=ent Purging During Nor=al Plant Operations" to justify

the containment purge system design.

042.25 The response to 042.16 is unacceptable. Discuss your plans for providing(6.2.4)

a purge syste= that cceplys with 3 ranch Technical Position CS3 6-4 or

co=mit to limiting purge system operation to 90 hours per year.

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042.26 Section 6.2.1.4.2 indicates that the containment integrated leakage rate

(6.2.6)test will be perforced at a pressure of 51.4 psig. Ecwever, in Supplement

2 of the FSAR (Page S2-138) the maximum calculated pressure is reported

to be 55.7 psig. It is our positien that the ILRT should be conducted in

cocpliance with Appendix J at the calculated pressure of 55.7 psig.

Discuss your plans for ccmplying with Appendix J.

042.27 Discuss the capability of the containment spray pumps to function reliably

(6.2.1)following the emptying of the chemical additive tanks. Describe any tests

that have been conducted to verify this capability.

042.28 The information presented in the response to 042.1 regarding the minimun(6.2.1)

contain=ent pressure analysis for the ECCS evaluation is unacceptable.

Provide a eccparison between the Three M.11e Island, Unit 2 containment

parameters and those presented in the B&W topical report. Justify the

applicability of 3AW-10103.

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