ML19329A929
| ML19329A929 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/21/1976 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | Roe L TOLEDO EDISON CO. |
| References | |
| NUDOCS 8001150834 | |
| Download: ML19329A929 (8) | |
Text
.
g I ml g
Jietri M on e
4 4 f~9~}
~
Docket File ~
NRC PDR n
(
G5 Local PDR
.v JL L
..L 3
LWR 4 File
. ri,,
R. C. DeYoung F. J. Williams Docket Ho. 50-346 R. Boyd
.-======
R. Heineman S. Vargn R. Tedesco Ioledo Edison Cowpany R. Maccary AITi:
Mr. Lowell E. doe R. R. Butler Vice President, Facilities L. Engle Developraent M. Rushbrook Edison Plaza ACRS (16) 300 aadisuu Avenue ELD Toledo, Gaio 43652 IE (3)
J. Miller Centlemen:
bec:
J. R. Buchanan, NSIC T. B. Ahernathy, TIC Cn..oveaoer 21, 1975, we informed you or a potential satety question wnica nas been raised regarding toe design of reactor pressure vessel support systems, ne requested taat you review tne design bases for tne reactor vessel support system tor yotr facility to determine vnetner tne transient loads described in the enclosure to our letter were appropriately taaen into account in tne design.
Your reply or Lecemaer 19. 1975 indicates tnat the transient differential pressures in the annular region oetween the reactor vessel and tne cavity anield wall and scross tne core carrel were not considered in tae aupport design.
Io our letter of.iovemoer 21, 1975, we attacaed a preliminary liatin;:
of potential requests for additional information snould we later determine, on ene oasis of your initial review, that a reassesscent of the vessel support design is required. we have now made auca a determination that reassessnent of the vessel support design is required.
As you are prossoly aware, we have oeen discussing witn tae Pwd sendors and various arenitocc/ engineer firms tne generic aspects of tnis prooien.
5aould you contemplate utilizing organizations otner enan your FA vocaor tar calculation of tae internal loads, we suggest you contact us for tae oenefit of a orief review of our generic discussions to date. we will continue tnese generic discussion $with tne vendors and arenitect/ engineers, but such discussions are not intended to pace your evaluation of this
\\ concern r to eli:sinate tne possioility tnat we say have additional questions re.;arding your evaluation af ter sub:sittal. Wnile the emphasis givea ia tais letter deals wien the reactor vessel cavity, for your infor-
.sation and guidance our generic review m1y consider ocaer areas in ene nuclear steam supply system and fortner evaluation may be required.
errica p evenams >
oave >
Fenn AsC 31s (Rev. 7 33) AscM 0240 W u. e. eovenmusar resanne orricas teve.sae see 8001150 h p' M
- W e
w r.
foledo Laison Company.,gy ; _ g Please inform us, within 30 days af ter receipt of tais letter, of your acnedule for providing ua with your evaluation of tne adequacy ot the pressure vessel supports unen these loads are calcalated and taken into account in a :nanner wnich you determine oest represents enese phenomena.
tour evaluation anould include the answers to ene enclosed request tor additional information.
Thir request for generic information was approved by GAO olanket clearance number 3-130225 (R0072). Inis clearance expires July 31, 1977.
Sincerely, 0.4-132' ? 2U7 5 c ;aer Walter R. Butler, Caier Light Water Reactors Branen 4 Division of Project.danagement
Enclosure:
Request for Acditional Inforsation mo u
ec: See page 3 vo n -
Ob-
-]
4 I
i k
RETYPED PER OELD (SEE PREVIOUS YELLOW FOR CONCUR.RENCES)
FI
.,,,c.,
.. _k "..._
d_,,,,
5/18/76 5/p/76 GFess WRButle
.v a- - = >
5/- /76 Feem M )l3 (s47.s-$3) M s24s k v. ei Sevemmasms? PegasT1ase ergr eye. p ges
MAY 2 57; 3-Toledo Edison Company ect Mr. Donald B. Hauser, Esq.
The Cleveland Electric Illuminating Company P. O. Sox 5000 Cleveland, Ohio-44101 Gerald Charnoff Shaw, Pittman, Fotts and Trowbridge 910 17th Street, N. W.
gn D
Washington, D. C.
20036 D
O C' Leslie Henry, Esq.
3")
~
Fuller, Seney, Henry and Hodge D
- L{
- {
d 300 Madison Avenue O
Toledo, Ohio. 43604 V
~orrec s >
eusseases >
Days >
Feen ABC415 (Rev. 9 53) AEQt 0240 W u. a. eowemmassmv passerme orricas sep4.eae..o l
Enclosre l' m
i.
9 l
OU g
~
\\
T R@n m
a h [} b
-NY; f
Y REQUEST FOR ADDITIONAL INFORMATION Recent analyses have shown that reactor p'ressure vessc' supports may be ' subjected to previously underestimated lateral leads under the conditions that result from the postulation of design basis ruptures of 1
the reactor. coolant piping at the reactor vessel nozzles.
It is therefore necessary to reassess the capability of the reactor coolant system supports to assure that tne talculated motion 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 y
can be brought safely to a cold shutdown condition.
P The following information should be included in your reassessment of the reactor vessel supports and reactor cavity structure.
1 1.
-Provide engineering drawings of the reactor support system sufficient il to show the geometry of all principle elements and materials of.
s I'
construction.
J 1
2.
Specify the detail design locds used in the original design analyses t
t -
j'.
of.the reactor. supports giving magnitude, direction of application i
o and the basis-for each load. Also provide-the calculated maximun
. stress.in each principle element of the support system and the corresponding allowable stresses.
3.
Provide the.information requested _in 2 above considering a postulated
. l'
. break at the design basis location that results in the most severe
.I-
. loading ~ condition for the reactar pressure vessel supports.
Include
)
(
[
L
,m
- 2'-
a summary-of the anal'ytical methods employed and specifically state the effects of asymmetric pressure differentials across the core barrel in combination with all external loadings including asymmetric cavity' pressurization calculated to result from the
- required postulate.
This analysis should consider:
(a) limited' displacement break areas where applicable (b) consideration of fluid structure interaction (c) use of actual' time dependent forcing function (d) reactor support stiffness.
4.
If the results of the analyses required by 3 above indicates loads leading to inelastic action in the reactor supports or displacements exceeding previous design limits provide an evaluation of the following:
i (a) Inelastic behavior (including strain hardening) of the material L
used in the reactor support design and the effect on the load L
transmitted to'the reactor coolant system and the backup-structures to'which'the reactor ccolant system supports are attached.
j 5.
. Address the adequacy of the reactor coolant system piping, control
}
rod drives, steam' generator and pump supports, structures surrounding the reactor coolant system, [ core upport structures, fuel assemblies, L
[;
fother reactor internals
....] and ECCS pipin~g for both the ' elastic and/or inelastic analyses to assure-that the reactor can be safely l
brought ~to cold shutdown. -For each item include the method of
[
Li u
4 m. __
b' m
m -
s t
3-analysis, the structural and hydraulic computer codes employed, drawings of the models employed and comparisons of the calculated to ' allowable stresses and strains or deflections with a basis for the allowable values.
-The compartment multi-node pressure response analysis should include
' the following information:
6.
The results.of analyses of the differential pressures resulting from hot leg and cold leg (pump suction and discharge) reactor coolant system pipe ruptures within the reactor cavity and pipe
- penetrations.
7.
Describe the nodalization sensitivity study performed to determine the ninimum number of volume nodes required to conservatively predict the maximum pressure within the reactor cavity. The nodalization sensitivity study should include consideration of soef:ial pressure variation; e.g., pressure variations circumferentially, axially.and radially within the reactor cavity.
8.
Provide a schematic drawing showing the ncdalization of the ceactor cavity.
Provide a tabulation of the ncdal net free volumes and interconnecting flow path areas.
9.
Provide sufficiently detailed plan and section drawings for several
. views showing the arrangement of the reactor cavity structure,
~
reactor' vessel, piping, and other major obstructions, and vent areas, to permit verification of the reactor cavity nodalization and vent locations.
e
_g_
10.
Prov'ide' and justify the break type and area used in each analysis.
11.
Provide and justify values of vent loss coefficients andf or friction
. factors used to calculate flow between nadal volumes. When a loss coefficient consists of more than one component, identify each component, its value and the flow area at which the loss coefficient applies.
12.
Discuss the manner in which movable obstructions to vent flow
.(such as insulation, ducting, plugs, and seals) were treated.
Provide analytical justification for the removal of such items to obtain vent area.
Provide justification that vent areas will not be partially or completely plugged by displaced objects.
13.
Provide a table of blowdownmass flow rate and energy release rate as a function of time for the reactor cavity design basis accident.
- 14. Graphically show the pressure (psia) and differential pressure (psi) responses as functions of time for each node.
Discuss the basis for establishi.ng the differential pressures.
15.
Provide the peak calculated differential pressure and time of peak pressure'for each node, and the design differential pressure (s) for the reactor cavity.
Discuss whether the design differential pressure is uniformly applied to the reactor cavity or whether it is spatially vari ed.-
In order to review the methods. employed to ccmpute the asymmetrical pressure differences across the. core support barrel during the subcooled portion cf the blowdown analysis, the following information is requested:
16.'
A complete description of the. hydraulic code (s) used including the
^
..A.,-
f **
O
.s
. -development of the equations being solved, the assumptions and simplifications used to solve the equations, the limitations
.resulting from these assumptions and simplifications and the numerical me'Eods used to solve:the final set of equations.
17.
In support of;the hydraulic code (s) used provide comparisons with the code (s) to applicable experimental tests, including the following:
(a). CSE tests B-63 and B-75 (b). LOFT test L1-P.
(c). Semiscale tests S-02-6 and S-02-8 The models developed should be based on the assumptions proposed for the analysis of a PWR.
18.
Provide a: detailed description of the model proposed for your piant and include a listing of the input data used and a time zero edit.
Identify the assumptions used in developing the model, specifically the treatment of area, length and volume.
- 19. Typically the current generation of hydraulic subcooled blowdown analysis codes solve the one-dimensional conservation equations.
However, they are used to model the multi-dimensional 'spects of a
the reactor system (i.e. the downccmer annulus region).
Provide justification for the use of the c" ode (s) to medel multi-dimensional regions, including the equivalent representation of the region as modelled by the code (s).
.m
= eh w
f a
W 7
a Toledo Edison Company ;
L Oi Please inform us within 39 days af ter receipt of this letter,' our senedule for prodiding u ur evaluation of the adequacy of the pressure vessel supports wnen taese loads are calculated and taken into account in a manner cien you dotermine oest represents taese pnenomens.
Your
\\ evaluation Id include the answers to the ei; -uod request for additional.Inforaation.
- dWA Inis request for generic infor:sation was approved by GAO olanaec clearance nummer 3-Idu223 f.RG072). Inis clearance expires July 31, 1977.
Sincerely, Ivalter R. Butler, Chief Light Water Reactors 3ranca '+
' Division of Project danasement t.nc los u re :
AJGuest for Additional Intormation cc:
Gee page J A t\\e.u,
\\~
I-L'eW
_ D M-4 GJ LWR 4 e......
- Englw
[bbs WRButler 3/76 5/ /76 5/ u/76 5
1 Pore AsC.)te (Rev. 9.$3) AIcM 0240
- u. s. eova.mmany reintime orrecas sen.eae.see i