ML19329E214
| ML19329E214 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 12/11/1975 |
| From: | Bauman R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Boyd R Office of Nuclear Reactor Regulation |
| References | |
| 2000, NUDOCS 8006110628 | |
| Download: ML19329E214 (5) | |
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o.noras ome..: 2 2 w..e uncn.g.n Avenue Jackson. Micnigan 492Cn. Area Code S17 788-0550 December 11, 1975 y --
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Director of Nuclear Reaction Regulation Attention:
Mr. Roger Boyd, Acting Director 7.S p.L 4-l'l 4
Division of Reactor Licensing U. S. Nuclear Regulatory Co= mission l;,
i.f Washington, D. C.
20555
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1 MIDLAND PROJEcr
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DOCKET NUMBERS
-329, 50-330 REACTOR VESSEL S SYSTEM FILE: B5.4.14 SERIAL: 2000 i
The following infor=ation responds to Mr. A. Schwencer's November 14, 1975 letter regarding the design of the reactor vessel support system for Midland Units 1 and 2: shows the model used to calculate the reactor vessel support system loads resulting from a loss-of-coolant-accident. These loads were then combined with normal operating loads and seismic loads to determine the resulting total stresses on the reactor vessel support system.
l As noted in Attachment 1, the blevdown jet forces at the location of the break were considered in the design of the reactor vessel support system.
The other two effects, transient differential pressures in the annular region between the vessel and the shield vall and transient differential pressures across the core barrel within the reactor vessel, vere not considered.
l R. C. Baunan
[j Project Engineer C:
1 M.
RCB/fv
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I REiCIOR. VESSEL SUPPORT EVALUATION FOR LOCA LOADINGS D
. ~ ~ Mhalf-loop model of the reactor coolant system was utilized to evaluate the behavior of the reactor vessel support system for postul'ated LOCA
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' conditions A su= mary of this mathematical model and the LOCA forcing functions considered is presented below.
EALF-LOOP MODEL 5
The half-loop model represents one half of the complete reactor coolant system. To account for those components which are located on the line of symmetry, half of the full section properties, support spring rates, and masses are utilized in the model. In this way a complete half system model was established allowing for the fact that in effect only one half of those components located on the line of sy-etry would be taken into account in the model. Figures 1 and 2 show a plan view and one "m ple of an elevation drawing of the half-loop model.
For the ther=al, deadweight, and seism).c loading cases the behavior of the unziodeled loop can be obtained through sym=etry considerations. The appropriate boundary conditions at the line of sy=netry for the modeling of the ther=al, deadweight and seismic (in the x and y directions only) loading cases were:
Ax Ay Az e
e o
x y
g (free).
(free)
(fixed)
(fixed)
(fixed)
(free)
For the z-direction earthquake the appropric.te boundary conditions along the line of sy etry were:
Ax Ay Az e
e e.,
x (fixed)
(fixed) (free)
(free) (f?ee)
(fixed)
For LOCA loading cases, however, a break could occur in either loop causing unsy etric loadirg. Therefore, for LOCA cases a special treatment i
of the model was necessary. For LOCA evaluation the co=ponents located on
-ll the line of sym=etry were given their full cross-sectional properties; that
!l is, twice the values they were given for the ther=al, deadweight, and seismic analyses, thereby modeling these components as complete. The effects of the I
unmodeled loop of the reactor coolant system was transmitted to the half-loop
[
model by specifying a special stiffness matrix as a boundary condition to the reactor vessel at the hot and cold leg nozzle elevation.
LOCA TURCING FUNCTIONS To evaluate the postulated LOCA condition, the applied thrust due to a pipe rupture was applied to the half-loop model in the form of a time-history e
thrust function. The LOCA thrust force for the hot and cold leg guillotines j
and splits (located at the reactor vessel nozzles) was analyzed using the t
FLASH co=puter code and the relationship, thrust = pressure x area. An instantaneous break opening time was assumed for guillotine breaks, and for splits a constant thrust (i.e., initial pressure x area of split) was
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_ utilized with a 10 millisecond rise time from zero load. Additionally, vertical components of cavity pressure resulting from pipe breaks I
inside the. primary shield wall were simultaneously applied to the
!I half-loop model.
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. ~ NRC DISTRIBUTION FOR PART 50 DOCKET MATERIAL
- (TEMPORARY FORM)
CONTROL NO: / 39 2 3 FILE:
FROM: gonsumers Power Co DATE OF DOC DATE REC'D LTR TWX RPT OTHER Jackson, Michigan 12-11-75 12-15-75 XX n c un,,-,2,,
- TO:
ORIG CC OTHER SENTNRC PDR XX UI'
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XXX 1
' O 330 DESCRIPTION:
ENCLOSU RES:
Ltr re our 11-14-75 ltr.....trans the follow:
Info concerning calculation of reactor vessell support system loads resulting
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- a from a LOCA.....w/addl info pertinent to 1 i.
itt.. " >-
the subject.... (1 cy enc 1 rec'c) i PLANT NAME:
Midland 1 & 2 FOR ACTION /INFC RMATION 12-18-75 ehf BUTLER (L)
SCHWENCER (L) ZIEMANN (L)
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