ML19220B496
| ML19220B496 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/20/1975 |
| From: | Silver H Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904260392 | |
| Download: ML19220B496 (6) | |
Text
.
UNITEo STATES NUCLEAR REGULATORY COMMISSION W AS HIN G TO N, O. C. 2 0 S S S November 20, 1975 Docket No.
50 -320 Applicant:
Metropolitan Edison Ccapany Facility:
Three Mile Island Unit 2 Sir.ARY OF MEETING ON CPEN ITED Representatives of the applicant and his contractors =et with members of the NRC s:aff on October 30, 1975 to discuss cpen items in the staf f review.
As an:icipated, =uch of the discussion related to items defined by the Containment Sys tems 3 ranch, the stea line break accident being chief among these. Specific ite=s are identified below, and any discussion or agreene=ts are surrarized.
Centainment Subeccoartment Analysis 1.
Letdown Ccoler Ceccartment Analvsis
~~he staf f reques ted the applicant to:
a.
Identify the nodes in Figure 42.1-1
- b. Provide L/A in Table 42.1.4
- c. Cceple te Table 42.1-4 to shew all the flew paths,
i.e.,
there are no ficw paths frc Nodes 6 and 7 to other nodes.
The applicant agreed to furnish this infor ation.
2.
5:ean Generator Ccepartment Tae staff requested the applicant :o:
- a. Cceple:e Table 42.1-4 to show the pressure in each of the 32 nedes, or provide a f a ily of curves to show the pressure as a function of time:
- b. Provide the design pressure of node 1 to node 9 as a minicu=.
c.
Identify :he break node.
- d. ?rovide L/A in Table 42.1-3.
Tae applicant agreed to furnish this information.
- ' f $N [
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.s 44 790426 o37;L
. 3.
Reactor Cavity Analysis The staff rrquested the applicant to:
Shcw the shield plugs or shielc plug caterials. Crawing previdec 1.
by letter fren R. C. Arnold to A. Gia=busso, July 11, 1975 do not show this.
b.
Justify that 8.55 square feet cold leg break has icwer sass and energy blowdewn race than the 5.;~ square feet hot leg break.
c.
Provide detail discussion of the shield plug dyna =ic :odel.
d.
Provide schematic drawings to shew the nedslization cf reactor cavity, Provide nodalization sensitivity etudy.
e.
f.
?:svide a cable :o snew the j unction data.
f.
Provide a table to shew:
1.
maximun calculated profile pressure and uniform pressure.
2.
design profile pressure and unifor= pressure.
The applicant indicated he would furnish this inf 7r=ation.
Ite=s b.
through g. will be in A=end en 34, due Oc cber 31, 1975.
Shield 'a'all Pipe Penetration 4
a.
The staff requested that the applican: provide the analysis of a break in this area.
The applican: stated this will be part of A=end=ent 34 5.
Stea= Generator Suecor: Skin:
The cass and energy release rate as a function of ti=e for the 5.2 a.
f2 break should be provided.
The applicant indica:ed he will furnish this.
6.
General Discussion ensued regarding the application of the results of the s ubce:-
partment transient analysis. The applicant noted tha: all structures had been designed using single node analysis and criteria, that they S.
..,. a
~ will not recheck these structures for :he results of the current subec=partment analysis, and that they would resiat vigorously any physical changes required as a result of this analysis. The st.ff indicated tha: the purpose of the analysis is to identify transient conditions, and that if proble=s are identified, further ef f orts to resolve these would be required - either analytical or physical, or both.
(In response to the applicant's request, the Structural Engineering Branch subsequently verified that if pressures identified in the analysis exceed design prersures, it will indeed be necessary to verify that the structures are capab7e of withstanding these higher local loads. This was then ree:phasized to the cpplicant by telephone.)
Centainment Heat Removal Svste=
1.
3&W has recen:1y performed a new spray syste analysis indicating SHT, STT and 3WST do not draw down together as they were previously pre-dicted folicwing a LOCA. This results in e=ptying of the SHT and STT up to 22 sinutes before the 3WST is depleted. As a consequence, air will be drawn in through the empty tanks and cavitation will be forced in the spray pu=ps.
The applicant stated this is not a proble: in this plant, and that the response to Q 310.4, schedule for Dece:ber 17, will confir: this.
Contain=ent Isolation Svste 1.
Table 6.2-15 to provide the cicsure ti=es for all the isolation valves should be ccepleted.
The applicant will supply this infor ation.
ECCS 3ackpressure 1.
Applicant refers the ECCS 3ackpressure analysis to 3&W's Topical Report 3AW-10103 which is under review. The staff agreed to consider reviewing that docunent only for this subject for TMI-2, rather than await a full generic : view.
In:2 grated Leak Tes:
1.
The staff requested that the containment integrated leak test be perfor:ed 2: 35.7 psig, the =aximum pressure calculated in the analysis in respense to Q 3.4.
The applicant sta*ed tha: the design basis pressure s 31.4 psig, and that the Q 3.4 analysis is essentially a sensitiv1:7 study.
The s:aff reiterated its pcsiticc, but in response to the applicant's request, agreed :o verify the pcsition.
(This was subsecuently done.
The staff will require testing at the higher pressure in accordance with Appendix J.)
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. 2.
In respctse to the staff's request to perfor: c Type C tes: on all isolatica valves listed in FSA3 table 6.2-15 (Q 43.15), the applicant indicated tha:
their interpretation of Appendix J does not require such testing for closed seismic category 1 lines not under contain=ent pressure. No agree-
=ent was reached on this ite=.
3.
Interpretation of Appendix J was also cited as the applicant's reason for not venting and draining some lines for Type A tests. No agree =ent was reached on this ites.
Clarification was requested between section 6.2.1.4.4.1 which does not require local testing of electrical penetrations, and Tech Spec Section 4.4.1.2.1.f, which does.
5 The applicant agreed to add the Appendix J require =ent to test equipment hatch and fuel transfer tube seals at least every rwo years (Tech Spec Section 4.4.1.2.5.5).
o.
The applicant ' greed to = edify Tech Spec Section a.4.1.1.3.c to require the test duration of Type C tests to be at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless specifically approved by NRC.
7.
With regard to Tech Spec Section 4.e.1.1.1.5, Appendix J requires Type 3 & C tests to be perfor=ed at Pc.
3.
In Tech Spec Section 4.4.1.2.3, ".067." should be 60%.
9.
With regard to Tech Spec Section 4.4.1.2.5.c, Appendix J requires testing at each opening and every 6 months at Pa.
S te a: Line Break The applicant stated that the LOCA is the present design basis for the contain-
=ent and :ha: :he prior analysis of the stea line break in the FSAR does not assu=e a single f ailure.
In responding to Q 042.7, which requires consideration of specific single failures, analysis shows : hat in the case of a feedwater valve f ailing open, centainment pressute exceeds design pressure approxi=ately 6 =inutes after the accident. Although the reactor returns to a low power level, it was stated that significant fuel failure will not res ult. Other single failures are not as tevere from the stand point of containment response. The applicant objects to consideration of single f ailures in the stea= line break accident noting previously accepted design bases and the status of plant construction.
Considerable dicussion identified the following points :
Opera:or acticn within 6 =inutes could ter=inate the contain=ent over-pressure proble an ' would be considered by NRC.
The addition of automatic tripping of the condensate booster pu=p,
~
snutting down the feedwater syste=, could likewise avoid containment ove rp res s ure, end would be considered by NRC.
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. Probabilistic argu=ents wculd he censidered by NRC, if co=plete and app rop riate.
Analyses =ust include consequences of the accident the=selves, such as probable failure of the ecolant pe=ps in the contain=ent at=csphere after a stea line break.
The NSSS supplier has been aware of NRC concerns and approaches regarding stea= line break accidents.
Af ter further discussicn, the staff indicated it would require analyses shcving the plan is saf e f rc= the point of view both of the contain=ent and the core in the event of a stea line break inside contain=ent. The follcwing would have to be consicered:
1.
a s p e c tru= c f b re aks.
~
2.
- ost reactive red stuck out.
3.
single ac:ive failures.
4 all acc. dent co ns equenc es.
5.
offsite pcwer both on and of f.
Probabilistic argu=ents and cpetator action would be considered as a;prepriate.
Contain=en: Purze
- n response :o :he applican:'s request for addi:ional infor=atien en the bases for :he positien,
..;e s:af f indicated :he basis :or the require:ent to li=it our;ing to 1~ per year for large openings was probabilir-tc. The prebabili:y of exceeding ?ar: 1 0 levels in the event of an acciden, ue :o ficw thrcugn the Open purge va ves before : hey cicse is reduced by.0 - if they are open nc : ore chan that frac:ica o f :he ti=e.
The staff feels :hese valves shottid ce closed as =uch as possible withi the require =ents of plant operation and
=aintenance. An 3" purge opening will produce acceptable dose resul:s.
The applican: noted : hat the valves are designed to close wi:hin 3 seccnds, but that no analysis had been =ade for Part 100.
Principal purpcses of the purge is te=perature and hu=idity control, and since the need for entry into contain=ent cannot be accurately predicted, they object Oc the 1% li=itation en purge : =e.
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. The staff noted that a cc=plete response to question 42.16 will be required.
satisfactory analysis in accordance with Section 3.5 of the position will also be necessary for the system proposed, along with detailed justificarica for openings larger than 8".
The applicant discussed the tising of system ele =ents and physi:al phenocena in a LOCA, and was cautioned on the difficulty of acceptance of such an approach. Other possible approaches were discussed briefly.
4
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Harley i ver, Project Manager Light @er Reactcrs Branch 2-2 Division of Reactor Licensing
Enclosure:
List of Attendees e
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