ML19319D022
| ML19319D022 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/18/1976 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | Rodgers J FLORIDA POWER CORP. |
| References | |
| NUDOCS 8003040946 | |
| Download: ML19319D022 (9) | |
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REGUEST FOR TDDITIOT:AL II,FCR"Js' ION f
Recent ar.elyses have sho.in that reactor pressure vessel supports r.tay ce-surye ted to oreviously undcrestiinatsd lateral losds under the-conditior.s that result fecr, the postulation of design basis ruptures of the reactcc coolant piping at the reactor vessel r Zzies.
It is 4
j therefcre necesrary tc reassess the capa'oility of the reactor coolant i
sysica supports to assure tha t the calculated r.:otion of the reactor vessel under the most severe design basis pipe rupture condition will be within the bounds necessary to assure a high probability that the reactor can te brcught safely to a cold shutdown condition.
o The folicaing information should be included in. your re*,ssessment of j
tne reactor vessel supports and reactor cavity structure.
L, 1.
Prcvide engineering drawings of the reacter support systc, sufficient S
to shv.i the geometry of all principle eler.ents and materials of const.ruction.
j}
2.
Specify the detail design loads used in the er wir.al dcsign analyses i-l' of the reactor suppoets giving magnitude, directior. of appiitation and the basis for cach Icad.
Also provice the calculated p-em ;n n
stress in each printicle ele.nent of. e supprt syster ar.i the 1
correspunding allo..able strencs.
1,,
3.
frcvide the infor.mation rcquested in 2 ab' ve consicering a postulated bruaY at the design'N. sis locatiGM thit "esults 'n the i. cst se.ere i
j --
loadin". coqd1' ion f:.r t!'O rract~' r'0:0, M
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I nci'. de F"
D "D TK'
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2-
. a' sumary of the analytical methods cm;;1cyco and specifically
_ state the effects of-asyrwatric pressure differentials across tha -
. c:> r a barrel in corcbination with all e.<ternal loadings includihg
- synetric cavity pressurization calculated to result from the ccquircd postulatt. LThis_ analysis.sr.caid consider:
(a).l limited displace. rent break areas where-Sppliccble (b) consideration 'of fluid structure i"teraction (c) use of actual tir.e dependent forcing function (d) reactor support' stiffness.
4.
If tha results of the analyses required by 3. above indicates loads 9
icading to inelastic action in the reactor supports or displacements exceeding previous desig'n limits provide an evaluation of the following:
(c)
Inelastic be. 'vior (inciuding strcin Nedening) of the material used in the reat or support design and the-effect on the icad transmitted to the recctor cociant system and the backup structures to i;hich the reactor cacian systcm sup;crts are attached.
5.
Address the pde'uacy of.the recctor coolant systen piping, centrol rod drivM steam generator and purp.su;pcrts, structures surcoundira the reactor coolant system, [ core supert structures, fu21 esser.tlies, other~ reactor internals
.] and CCCS pipig for I:oth the eintic ant'/ac ir. elastic cnaly:es to assure that the recctor can be safely brought to cold shutccwn.
For each : te-inc';de the raethod of
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o,2, dre..;ings of the mdels enalcyed and ccr.pariscas of the calculated to cilo.Mie stresses and strcins or daflections with a basis for i.F e allor:abie values.
Ti.e coc;sr* men t multi-nac.: pressure res;o:ue crai sis should-include f
cr.e to4:0.0 n'; inror:w 1c n:
6.
The results of analyses of the differcatial pressures resulting f rca h" le<j ind cold icg (prmp suction and discharga) reactor scolant systai pipe ruptures within the reictor cavity and pipe penetrations.
7.
Oascribe the nodalization sensitivity stud;/
irfor;ed to deternine the minimum number of volare naaes required to conservatively predict the na).imum pressure within the reac*.ar cavit.
The nodalization sensitivity study shculd inciude consideratien of yatial pressure var:ation; e.g., pressure vaciations circc.?erentially, axially and radially v;ithiri the reactor cavity.
3.
Prcvide a schematic drawing showing the r d lizatien of the reactor caviLy.
Frovide a taculation of the nedai net free volumes cnd in'.e connectir.g fic.; rath creas.
9.
Providt sufficientl, detailcd plan and sectic-dra.nngs for seeeral
. a. u,,
s., t o_ u...,u,
,. 1 c....3 sow... g t y~o c,..... c m.
-.., i %,
...e au,.u.
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.c rcactoe venscl, pipir.t., cnd other a;er oxtracticas, and. cc.t arcas, to 'croit verificcticn of the rnacter cr. in n:ic!i? itio1 and v^nt
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_we ju o M
. A A !A1a L-
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- 13.. Pec9id2 and justify the brea:. tya and area used in each cnalysis.
li.
Provide and justify values of vent loss coefficients and/or friction:
factors used to calculate ficet bet',zeen nodal volumes.
tl hen a loss
' cefficient consists of more than ene component, identify each.
cct;:.or.ent, its value and the ficci area at <;hich the loss coefficient ~
applies.
12.
Discuss the canner in vhich rovsble obstructicns to vent flow (such as insulation, ducting, plugs, and seals) tlere treated.
Provide arm.'ytical justification for the rca.cyal of such itenc to obtain vent Provide justification that vent areas *;ill not be partially r
_a.
or corpletely plugged by oisplaced objects.
13 Provide a table of blov.dov:nmss flow rate and energy release rate as fur.ction of tir:e for the reactor ctvity desian basis accident.
1 14 Graphically show the pressure (psia)
. W!ferential gessure (psi) responses as functions of time for each node.
Discuss the basis for establishing the differential pressures.
15.
Prc zide the peak calculated differential pressure and i.ine of peak pressure for each node, and the desicn :ifferential pressure (s) for the reac'or cavity.
Discuss v:hether the design differential pressure is uniformly applied to the reactor csvity or whether it is spatially varied.
in crdetcto revie... the methods'crployec t: w"pute the asymetricai
. pressure differences across the core su scrt ba+Yel 6: ring the subcooled I
jOr'! ion of I,)le bl{?idOND arealysis, Ebb fo) ~ ' n g i f. cfI3biCU '5 IcQue3 Icd
- 1 10.
- r. coraiete descriptien of the hj9aulic c"c(s') used including the
)
I l
I D
(h P'-
m f
A o r, pl j L.
- 5 development of the e;uations being solved, the :.ssu,:pticr.s
.>.d
~
simplifications used to solve ~ the egntions, the limitations resulting fec:a these asst. ptions and simplifications and the numerical methods used to solve the Tinal set of equations.
17.
In support of the hydrsuiic code (s) used '.rovide ccmparisons with'the code (s) to applicable ey.parirental tests, including the fol l o'.. i ng :
(a). CSE tests B-63 and B-/5 (b). 1.0FT t2st L1-2 (c). Semiscale tests S-02-6 and S-02-8 The models developed should be based cc the assumptions proposed for d.e analysis of a P', R.
12.
Provi e a detailed description of N e r:cdel :rgesed for ycur plant und ir.clude a listing of the input data used r.nd a tir.e zerc edit.
Identify the assunptions used in developina the rrodel, specifically ihe treatment of area, kngth and vohme.
19.
Typically the current generation of hydraulic subcooled blowdown aalyf.s codes solve the one-dmensior.51 const
- tion ecuations.
- .ever, they are used to model the multi-dic.ensional aspects of the reactor system (i.e. the do.:nconc crnulus region).
Provide Jus tit ication Tor t.na use o,, t,e cou,n,s, tc.10.e1 tuiti-d w.ensional s
a regiens, including the equivalent rcpresentatico of the recien as icedelled b;/ the code (s,i.
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