ML19206B193
| ML19206B193 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/16/1975 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7905080019 | |
| Download: ML19206B193 (17) | |
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5 375 Docket No. 50-320 Voss A. Moore, Assistant Director for Light Water Raactors, Group 2, ?J REQUEST FOR ADDITIONAL INFCRMATION FOR TEZ TdREE MILE ISLA'O NUCLEAR STATION, UNIT 2 Plant Nara: Three Mila Island Nuclear Station, Unit 2 Docket No.:
50-320 Licensfag Stage: CL USSS Supplier: 3abcock & Wilcox Architect Engincar: Surns & Roe Contaic=ent Type: Dry Responsible 3 ranch & Project Manager: LWR 2-2; 3 Washburn Requeatad Complet. ion Date: April 25, 1975 Applicant's Response Data: June 20, 1975 Raview S ta tus : Awaiting Information ne enclosed request for additional infor=ation (Q-2) for the Three Mile Island Nuclear Station, Unit 2, has been prepared by the Contain= ant Systems Branch af ter having reviewed the appropriate sections of the final anfaty analysis report (FSAR), as amended. During our review, we noted that the applicant's responses to =any of our Q-1 questions were inadequate.
Se following significant cou:ssats are based on our review of the FSAR, up to and including amenda.nc 26:
(1) the applicant has not discussed the nodalisation sensitivity studies performed to determine the mini m nu=ber of volu:na nodes required to conservatively predict the aricu:2 pressure in each subcompartment; (2) the applicant has not provided justification for assuming that obstructions to vent flow will be removed to increase vent areas in the subcompartment pressura rssponse analysis; (3)
- he appif emnt has not provided an analysis of the pressure response of the steam ganarator support skirt to a postulated RCS pu=p suction line break; (4) the applicant has not adequately described the calculational method and assumptions used in the main steam line break analysis to establish the maxt:num containmant pressure or the contninment design temperature; and, y.-
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~. u V. A. Moore s p,, 6 1976 (5) the applicant has not provided a miM,n, contain=ent pressure analysis for ECCS studies. We have requested that the applicant do this, and have attached Branch Technical Position CSB 6-1, " Minimum Contain-sent Pressure Model for FVR ECCS Performance Evaluation," which should be forvarded to the applicant.
It should be notad that the maxi =um calculated dif ferential pressures for so:ne of the subecmpartments analyzed are =uch higher than the design pressures; e.g.,
the design diff erential pressure for the reactor cavity is 92.9 pai and the mav% calculated diff erential pressure is 132 pai; the design differential pressure for the steam generator co=part=ent is 23.5 psi; and the maximum calculated differential pressure is 39.44 psi.
Because this is significant, and because other information requested concerning the subcompartment analysis has not been provided, a meeting should be arranged with the applicant to discuss the statua of his sub-compartment analysis, analytical methods and results.
u original signed b71 / <. p e Eobert L TN Robert L. Tedesco, Assistant Director for Contain,uant Safety Division of Technical Review
Enclosures:
As s ta t ed
- cc:
S. Hanauer F. Schroeder A. Cla=busso W. Mcdonald D. 31senhut K. Kniel G. Lainas
- 3. Washburn S. Varga J. Kudrick J. Shspaker D. Shu:n J. Glynn bec: NRR Reading File CS Reading File CS3 Reading File n*
4 -l !
Docket Files -
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m 042.1 The response to questian 03.1 regarding the contain=ent (6.2.1.1) subcompartren an.tly s is is ince:plete. Proside the tellcw-2nal :ei:
ing information for e:ch suicompartment
/
(1) describe the nadalization sensitivity study performed to determine the minimun number of volume nodes required to conservatively predict the maximum pressure within the subec=partment. The nodaliza tica sensiti.-i:7 s tudy should consider spatial pressure variations; i.e.,
circunferential, anial, and radial within the subcompar:-
cent, particularly in the reactor cavity, (2) provide schematic drawings showing the nadalization of each subcompartment, and indicate nadal net free volumes and interconnecting flos path areas;
~
(3) provide suffic en:1v detailed clan and section drawines
- or several views shcwing the genera., arrangement of subcompartment structures, components, piping, and other major obstructions, and from which subcompartment nodes and :.,cw paths can be veritieu,;
(4) justify the design basis break type and area usec in the analysis; (5) prcvide and j us tiff the values of vent loss coefficients and/or friction factors used to calculate ficw betweer nadal volumes.
'.ihen a loss coefficient consists of more than one corponent idantify each cc penent, g i'ce its value and specify shere in the. flaw path :he loss coe::icient appiles; (6) discuss how =cvable abs truc tions to vent ficw (suua as 4,su1a.4,,
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e Include analytical justification for the renc/21 of suc2 iters to ob tain 7ent area.
?rovide assurance that
'c e n t areas will not be partially or corpletelf plugged by displaced objec:s; (7) provide a table of the mass and energy bicwdown rates e,..,..
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042.2 Provide an analysis of the pressure response of the s team
( 6. 2.1.1) genera cr support skirt to a postulated pump suctica line break inside the skirt.
Oa2.3 Provide the f ollevin~n additional informa:ica for,the rea. tar (6.2.1.1) cavi:7 analysis:
(1) provide drawings of the reactor cavity she.;ing desi;n details, and the dimensicas used to calculate nocal volumes, vent areas and path len;;hs.
Shaw the locations of the shield plugs, and provide a drawing sac:.ing the shield plug construction. Specify the shield plu; materials of cons truction; (2) specify and justify the break type and area; (3) discuss the design pro. sions to ensure th2: the sh:21d plugs do not become damagi:.c v.issiles :chen they are blown free of the reactor cavity, C-)
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,a is made that all shield plugs are instantaneous')
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free to provide additi:nal ven; area.
Provide :he resul:s of a reactor cavity analysis bas'd on dy unic, dell ;
of shield plug reverent to obtain the additionni ven; area. Discuss and justify the assu=ptions race in develon-in; the shield plug dynamic =cdel; (3) justify the assumption : hat all piping insulation wil. be ins tantaneously ejected f rc the reactor ca;ity snield pipe annuli and that :he insula:ica surroundin2 the reactor vessel will be comores A.sd to zero th.ckness; and, (6) provide a table comparin; the desi;n differentia: erassures
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calculated differential pressure for the reactor cavity is 132 psi, and the maximu calculated differential pressure for the stea generator compar: rent is ^ 9. -- p s..
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042-3 042.5 The response to qucstica 03.3 is not complete; provida the
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correlatica to calculate the steady-state hea: transfer C o _.c.e.:
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042.0 Fevise the containment pressure analysis in the respense a (o.3.3) quest.4on Os.,
.oy using :ne 3
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(1) Provide the mass and energy release rates as a functica o.r -.< _, a-a
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c (2) Discuss and justify the assumptions made to predict :he mass and energy releases > and arcvide a detailed discussi:n c
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due to condensing heat transrer to tne passive heat sinks and the fan coolers; (c) provide a table of the peak values of containmen accosc.here tenn.erature ana c.ressure :or eacn or.
the above breaks ;
( f) for the case
- ich results in the maximum contain-ment atmosphere :e perature, grapalca.,)
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(h) specify and j ustify the design temperature of the c o n a <
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m C'2.3 The response to question 03.11 is not a d e q ua te.
Provide ac (9.4.13) a..a l"s a-o.#
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c loss-of-coolant accidents be perforced in accordance ~ith the requirements of Sectica 50.'6.
Appendix K, "ZCCS Evaluatian
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assumed in the analysis. Provide the contairm.ent pressure, te=perature and sump temperature response for the =cs: con-servative assumptions.
042.10 Discuss the design and functional capability of the hydrogen (6.2) analyzer sys tem, and provide a pipin; and instruncatation d.a.
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t 3 ranch Technical Position CS3 6-1 MINIMUM CCNTAINMENT PRESSU?2 MDDEI, FOR P'a'R ECCS PERIODIANCE I7ALUATICN A.
BACKGROUND Paragraph I.D.2 of Appendix K. to 10 C7R Part 50 (Ref.1) requires that the contain=ent pressure used to evaluate the perfor=ance capability of a pressurized water reactor (?T4) e=ergency core cooling syste=
(ECCS) not exceed a pressure calculated conservatively for that pur-pose.
It further requires that the calculatien include the effects of operation of all installed pressure-reducing syste=s and processes.
Therefore, the following branch technical position has been developed to provide guidance in the perfor=ance of =ini=u= contal=ent pressure analysis. The approach described below applies only to the ECCS-related contain=ent pressure evaluation rad not to the conen 6 ent functional capability evaluation for postulated design basis accidents.
3.
BRANCH TECHNICAL FOSITION 1.
Inout Infor=ation for Model a.
Initial Centain.ent Internal Conditions The =in1=u= contain=ent gas te=perature, =ini=u= contain=ent pressure, and -awmu= hu=idity that may be encountered under 11=1 ting nor=al operating conditions should be used.
b.
Initial Outside Conta1=ea.: A=bient Conditions A reasonably low a=bient tc=perature external to the contain=ent should be used.
c.
Containment Volu=e The mnW mm net free contain=ent volu:ae should be used. This
-aH u= free volu=e should be deter =ined from the gross contain-
=ent volu=e =inus the volu=es of internal structures such as walls and floors, structural steel, =ajer equip =ent and piping.
~he individual volu=e calculations chould reflect the uncertainty in the component volu=es.
n'
'0 3 L1 2.
Active Heat S inks a.
Sorav and Fan Coolinz Systers The operation of all engineered safety feature centain=ent heat re=cval systa=s operating at mv4 u= heat removal capacity
- 1.e.,
with all centai=ent spray trains operating at -nvinu= flev cen-ditions and all energency fan cooler units operating, should b e assu=e d.
In addition, the =ini=um temperature of the stored water for the spray cooling syste= and the cooling water supplied to the fan coolers, based on technical specification li=its, should be assumed.
Deviations frca the foregoing vill be accepted if it can be shcun that the worst conditions regarding a single active failure, stored water te=perature, and cooling water te=perature have been selected f rcm the standpoint of the overall ECCS =odel.
b.
Centainment S tean Mixine 'Jith Scilled ECCS '4ater The spillage of subecoled ECCS vater into the containment pro-vides an additional heat sink as the subcooled ECCS vater mixes with the eteam in the contain=ent. The effect cf the stess-vater mixing should be considered in the contain=ent pressure calculaticus.
c.
Containnen t S team Mixinz 'Jith 'Jater frem Ice hit The water resulting fren '.ce melting in an ice condenser centain-ment provides an additional heat sink as the subcooled water mixes with the steam while draining frem the ice condenser into the lower contain=ent volume. The effect of the steam-vater
=1xing should be considered in the contaimnt pressure calcu-lations.
3.
Passive Heat S inks a.
Identification The passive heat sinks that should be included in the contai=ent evaluation model should be established by identifying these s tructures and co=ponents within the containnent that could 4 <> R i
Ln
9 influence the pressure response. The kinds of structures and cc=penents that should be included are listed in Table 1.
Data on passive heat sin'<s have been co= piled frc= previous reviews and have been used as a basis for the s1=plified =cdel outlined belev. This =odel is acceptable for =inimu= containment pressure analyses for tenstruction per=it applications, and until such eine (i.e., at the operating license review) that a co=plete identifica:1cn of available heat sinks can be =ade.
This s1=plified approach has also been followed for operating plants by licensees complying with Section 50.46 (a)(2) of 10 CFR Part 50.
For such cases, and for construction permit reviews, where a detailed listing of heat sinks within the centa1==ent of ten cannot be provided, the following procedure may be used to =odel the passive heat sinks within the contaic=ent:
(1) Use the surf ace area and thickness of the primary contain=ent steel shell or steel liner and associated anchors and concrete, as appropriate.
(2)
Esti= ate the exposed surface area of other steel heat sinks in accordance with Figure 1 and assu=e an average thickness of 3/8 inch.
(3) Model the internal concrete structures as a slab with a thickness of 1 foot and exposed surface of 160,000 ft'.
The heat sink ther=ophysical properties that would be acceptable are shown in Table 2.
At the operating license stage, applicants should provide a detailed list of passive heat sinks, with appropriate dimensicas and properties,
b.
iies t Transfer Coefficients The folicwing conservative condensing heat transfer coefficients for heat transfer to the exposed passive heat sinks during the Li
e 4_
blevdcvn and post-bicwdewn phases of the less-of-coolant accident should be used (See Figure 2):
(1) During the bicwdown phase, assu=e a linear 1:. crease in the condensing heat transf er coefficient from h
~
d 8 Btu /hr-ft
- F, at t= 0, to a peak value four times greater than the =ax1=u= calculated condensing heat transfer coefficient at the end of blevdewn, using the Taga=1 correlation (Ref. 2),
0.62 0
h
= 72.5
=ax
- VtP, 2
where h
= =ax1=u= heat transfer coefficient, Stu/hr-ft
- F
=ax Q
= pri=ary ecolant energy, Stu V
= net free containment volune, ft ti=e interval to end of blowdown, sec.
t
=
p (2) During the long-ters post-bicwdown phase of the accident, characterized by Icw turbulence in the centainment at=osphere, assu=e condensing heat transfer coefficients 1.2 ti=es greater than those predicted by the Uchida data (Ref. 3) and given in Table 3.
(3) During the transition phase of the accident, between the end of blowdcwn and the long-ter= post-bicwdevn phr.se, a reasonably conservative exponential transition in the condensing heat transfer coefficient should be assumed (See Figure 2).
The calculated condensing heat transfer coefficients based on the above mathed should be applied to all exposed passive heat sinks, both =etal and concrete, and for both painted and unpainted surf aces.
Heat transfer between adjoining =aterials in passive heat sinks should be based en the assu=ption of no resistance to heat flcw at the material interfaces. An mmle of this is the contain=ent 11=er to concrete interf ace.
n' i <> d i '
Gi
9 C.
REFEPINCES 1.
10 CFR Section 50.46, " Acceptance Criteria for E=ergency Core Cooling Syste=s for Light '4ater Nuclear Power Reactors," and 10 CFR Part 50, Appendix K, "ECCS Evaluation Models."
2.
T. Tags =1, "Interin Report on Saf ety Assessments and Facilities Es tablish=ent Project in Japan f or Period Ending June 1965 (No. 1),"
prepared for the National Reactor Testing Station, Februar/ 23, 1966 (unpublished work).
3.
'd. U chida, A. Oy a=a, and Y. Toga, " Evaluation of Post-Incident Cooling Systems of Light '4ater Pcwer Reactors," Proc. Third Inter-national Conference on the Peaceful Uses of Atomic Energy, Volu=e 13, Session 3.9, United N1tions, Geneva (1964).
Q 7( '
i k> /
l t
TABLE 1 IDE'!TIFICATION OF CCNTAINMENT HEAT SINKS 1.
Centad-ent Building (e.g., liner plate and external concrete walls, floor, and su=p, and liner anchors).
2.
Contain=ent Internal Structures (e.g., internal separation walls and floors, refueling pool and fuel transfer pit walls, and shielding walls).
3.
Supports (e. g., reactor vessel, steas generator, pu=ps, tanks, maj or components, pipe suppor s, and storage racks).
4.
Lhinsulated Syste=s and Cc=ponents (e. g., cold water syste=s, heating ventilation, and air conditiening syste=s, punps, =otors, fan coolers, reco=biners, and tanks).
5.
Miscellaneous Equip =ent (e.g., ladders, gratings, electrical cable trays, and cranes).
n.
Si g i
TA3LE 2 REAT SIS"4 THER"CPHYSICAL PRCPERTIES Specific Ther=al Density Heat Conductivity Mate rial lb /* t3 Stu/lb '?
3tu/hr-ft *F Concrete 145 0.156 0.92 Steel 490 0.12 27.0 7 '.
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it /
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- ?)
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(3tu/hr-ft
- F) 50 2
3 29 20 8
2.3 37 18 9
1.8 46 14 10 1.3 63 10 14 O.8 98 7
17 0.5 140 5
21 0.1 280 4
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