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REACTOR PRESSURE VESSEL
SUPPORT MODIFICATION
FOR
MIDLAND NUCLEAR POWER PLANT,
!
; MIDLAND, MICHIGAN
PRELIMINARY REPORT NO. 1
JULY 1980
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CONSUMERS POWER COMPANY
I JACKSON, MICHIGAN
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REACTOR FRESSURE VESSEL
SUPPORT MODIFICATION
ICR
MIDLAND NUCLEAR POWER FLANT
TABLE OF CONTE'iTS
Pace1.0 DITRODUCTION 1
2.0 DESCRI>? ION OF TEE EXISTEIO REAC"0R VISSEL SUPPORT
DESIGN AND 1CRITERIA
3.0 DESCRIFTICN OF THE SUFFORT SYSTDI WDIFICATION 2
h .0 FRELDfETARY DESIGN 4
5.0 CUFREvr STATUS OF CONSTRUC"'IO:t 5
6.0 SCEEDULE roR MODIFICATICIT '40RK 6
T .0 REFERE''CES 6
"'A3LES
1 Effects of Schedule Revision I
FIGURES
1 Position and Nunbering of Studs in Unit 1 8
2 Reactor Vessel Elevation 9
3 Anchor Stud Installation Detail 10
h Fedestal Detail 11
5 Lateral Sup; ort Concept 12
6 Lateral Support Plan 13.
T Bracket Detail lh
8 Detensioning Chart 15
9 Uork Schedule, Rev F 161
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10 Work Schedule, Rev 0 lI
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REACTOR FRESSURE VISSEL
SUPPORT MODIFICATION
FOR
MIDLAND UUCLEAR F0WEl PLAIIT
1.0 EITEOEUCTION
Unit 1 of Consumers Power Company's Midland Plant has
experienced failure ofthree reactor vessel (RV) anchor studs.
Figure 1 shows the location of thethree failures. The first two
failures were in the upper threaded sectionof the studs and the
third failure was in the lover threaded section (seeFigure 3) .
The anchor studs are ASTM A 35h Grade 3D, 2-1/2 inches in dis
=eter and 7 feet,4 inches long. There are a total of 96 anchor
studs per R7 located in twoconcentric rings on each side of the RV
skirt as shown in Figure 1.
Investigation of the failed RV anchor studs was perfor=ed by
Teledyne EngineeringServices (TES) of '4altham, Massachusetts , and
was reviewed by Aptee of northernCalifornia, 3echtel, and Consumers
Power Cc=peny (see References 1, 2, and 3) .Based upon the result
of the investigation, it was concluded that Unit 1 RV studsshould
be detensioned. In their current condition, the studs cannot take
the=oments and the uplift transmitted by the RV. However,
=odification of theRV support syste= by providing additional
lateral supports at a higher elevation(see Figure 2), along with
existing studs stressed to a reduced preload level,can withstand
all the loads transmitted by the RV to the support system.
Latersections in this report vill further discuss the proposed
=odification.
Based on the infomation contained in the Teledyne Engineering
Service Reports(References 1, 2, and 3), the Unit 2 RV studs are
adequate as originally designedwithout modifications. However,
pending final confimatory analyses of the Unit1 =odification
adequacy, Consumers Power Co= pay intends to install
identical=odifications to Unit 2.
2.0 DESCRIPTION OF THE EXISTE!G REACTCE VESSEL SUPPCRT DESIGN
AND CRI"'ZRIA
The RV support system is shewn in Figures 2, 3, and E. The
support syste= isco= prised of the anchor studs, nuts, sole plate,
anchor plate, support ;edestal,shear pins , and shear lugs. "'he
applicable codes were the AISC Code, 63 or 69edition, for steel and
the ACI 318-63 or 71 Code for concrete. '"he cylindricalskirt and
emlar flange supporting the RV are =ounted on 5-1/2 inch
thickannular sole plate seg=ents. The skirt flange botto: is the
interface pointbetween the nuclear stes: supply syste= vendor and
the architect / engineer. Thesole plates rest on concrete ledges
which for= a part of the primary shield vall.The concrete has a
28-day strength of 5,000 psi. The RV is connected to thesole plate
with the 96 previously described anchor studs . These studs
extendthraugh the bottom flange of the RV skirt, sole plates , and
the anchor platewhich is e= bedded in concrete.
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' , '..,Midland Flont Units 1 and 2RF'l Support Mcdification
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The lateral and torsional loads are resisted by shear pins and
shear lugs andare then transmitted to the primary shield vall. The
sole plates , anchorplates , and shear lugs are =ade of ASTM A 36
=aterial. The shear pins aremade of ASTM A 354 Grade 3D
=aterial.
The loads contributed by the RV to the fcundation are obtained
from the reactorcoolant syste= analysis performed by Babcock &
Wilcox. The design require =entimposed by B&W on the holddevn
syste= vas that the studs be tensioned to a finalstress of_55 ksi.
Adding up all the losses such as elastic shortening, creep,the:=al,
etc, the initial prestress level cane to 75 ksi. This tensionire
require-ment provides considerable rotational stiffness to the RV
at the skirt flange. Aspring rate curve showing the relationship
between rotational stiffness androtation =c=ent based on 55 ksi
final prelcad has been obtained by Bechtel andwas used by 3&W
to perfer= its reactor coolant syste= enalysis.
The RV support system is classified as a Seismic Category I
structure and isdesigned for all credible conditions of loadings ,
including ncr=al loads, loadsresulting from a loss-of-coolant
accident, the:=al loads , and seismic loads . Theapplicable lead
ec=binatiens can be found in Subsection 3.S.6.3 of the MidlandFSAR.
The governing load condition is:
D+L+R+T +HA + E'Awhere
D = dead leadsL = live loadsR = local force or pressure caused
by rupture cf any one pipeT. = total the:=sl effects which =sy
occur during a design accident other^
than those considered by HgH, = force on the structure due to
ther=al expansion of pipesaE' = safe shutdown earthquake lead
-F = yield stress of material7
The maxi =u= allevable stress in bending and tension is 0 9 F.,
and for shear itis 0 5 F . All loads were generated by 3&W on
the basis that the anchor studs
7will have a minimu= value of 55 ksi prestress.
3.0 DISCRIpTION OF THE SUpp0RT SYS"'EM XCDIFICATION
3.1 DESCRIPTION OF THE LDDIFICATICN CONCEIT
The brackets that support the cavity annular shield plug at the
top of the RVvill be used as lateral restraints to resist the
overturning =c=ent that producesthe tensile forces in the anchor
studs. As shown in Figure 5, this lateral supportvill enhance the
supporting syste= by reducing displace =ent of the vessel
withoutrequiring the anchor studs to be prestressed in excess of 6
ksi.
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'* *Midland Plant Units 1 and 2RFV Support Modification
As can be noted frem Figures 5 and 6, there are 12 steel
brackets attached to~
the reactor. cavity vall above the no::le penetrations. The
distance between' the brackets and the vessel varies between 1-1/h
inches and 6-1/k inches asshown in Figure 7. If the gaps between
the brackets and the vessel are shi=medin the hot condition, the
brackets vill provide the required upper lateralsupports.
Preli=inary calculations have indicated that after stiffening
thebrackets as shown in Figure 7, they will be capable of taking
the new loadsthat they vill be subjected to in their new function
as upper lateral supportsto the vessel. Thesa same calculations
dec:onstrated that the vall vill becapable of resisting the new
leads transmitted via the brackets. '
- During the cold shutdevn condition, the upper lateral support
vill separatefrom the RV due to contraction of the RV. In this
condition, preliminarycalculations indicate that"the displacement
of RV under a SSE event is s=allenough such that no contact with
the upper lateral support vill be =ade, andthe. re=aining anchor
studs are capable of resisting the seismic loads , Eevever,in the
transient stage, ie, the time it takes the R?/ to cool down fro
operatingcondition to cold shutdown condition, an SSE event could
cause contact between.the REV and upper lateral supports. This
situation is cur ently being investi-gated both frem the point of _
view of RFV integrity and the design of the supportsystem, namely,
the anchor studs and upper lateral supports, to assure that
plantoperating and cold shutdevn conditions envelope all loading
conditions.
3.2 DESIGN CRITIRIA.
The support syste=s are categorized as a Seismic Category I
st:ucture. Thedesign basis vill be same as that used for the design
of other cc=ponent supportssuch as Steas Generator Supports and
Pressurizer Supports .
For the upper lateral support, the applicable codes are:
a. Reinforced Concrete: ACI 318-71 "3uilding C-de Requi.ements
for ReinforcedConc ret e"
b. Structural Steel: AISC-1970 " Code of Standard Practice for
Steel Buildingsand Bridges"
~c. Welding: ' AWS Dl.1-72 including Revisions and Addenda up to
and includingJuly 197h ." Structural Welding code"
'iThe loading conditions and allevable stresses are as stated in
Section 3.8.6 ofthe Midland FSAR.
The allowable stresses for the anchor-studs were based on the
results of TISReport TR-3887-2 (Reference 2l with the following
acceptance criteria beingrecon = ended for the Unit 1 studs.
3 .2.1 The final tension stress level on the studs vill not
exceed 6 ksi on the- tensile stress area.
3.2.2 Short-term loadings are pe =itted if the stress does not
exceed either' h3 ksi or one-half the levest detensioning stress on
any stud which is consideredto contribute to load-carrying
capability in the new design. _ The detensioningload can be
increased above the load required for nut rotation. In this
case,
,
the increased load can be used to dete:=ine the allovable
short-ters loadings. t
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,:. ", .,Midland Plant Units 1 and 2RPV Support Modification
h .0 PREL3!I'TARY DESIG'T
h .1 UPPER LATBAL SUFFORT AND S*UDS
S&W has perfomed preli=inary analysis using the upper
lateral support alongwith zero pretension loaded anchor bolts. The
load cases analyzed were safeshutdown earthquake (SSI) and what
3&W identifies to be the vorst case loss-of-coolant accident
(LOCA), a hot leg guillotine at the RV. The analysis was donewith
the upper lateral support in full contact with the reactor pressure
vessel.The loads trans=itted fro = the RV to the support syste= at
the RV skirt flangeand the upper lateral support are given belev
for illustrative purposes.
AT RV SKIRT FLA: IGE LF/Er v *g~ hear ' vertical corizontal
torsionals(ki s) (kins) ( ft-ki rs ) (ft-kirs)
SSE llh 233 1h7 1,6h6LOCA 1,003 3,3h7 3 ,529 1,113
AT UPPER LAT BAL SUPPORT LE/EL(RADIAL LOADS)
Wall Individual Support(kirs) (kirs )
SSE 166 55LOCA 3,377 1,126
Primarily, calculations have indicated that after stiffening the
brackets, asshown in Figure 7, the brackas will te capable of
taking the leads tabulatedabove. "'he sa=e calculations have
de=onstrated that the vall vill be capableof resisting the new
loads transmitted via the brackets and the stresses in there=aining
anchor studs will be less than the specified allevables.
The upper lateral supports significantly lever the forces and
noments that theanchor bolts vculd be required to carry with the
exception of loads caused byvertical and torsional notion of the
RV. S&W concludes that the addition ofthe upper lateral
supports is an effective =eans of re=cving lead frc= theanchor
belts .
h.2 EFFETS 0 I RElCTOR VESSEL A''D I' ITER:IALS FROM TIE ADDITIO
i CF THE UPPBLATIRAL SUPPCETS
S&W has co=pleted a stress evaluation of the reactor
pressure vessel based on thesupport scheme and load cases =entiened
in Section h.1. It was concluded thatthe vessel stresses have an
adequate safety nargin for the faulted condition.
Displace =ent of the reactor pressure vessel is reduced for the
faulted conditionby the addition of the supports. Displace =ents of
the control rod drive =echanis=and fuel assenbly upper and lever
grid were monitored along with forces and =omentson key reactor
internal connections.
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Midland Plant Units 1 and 2Rpy Support Modification
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h.3 REACTOR COOLA:!T SYSTDI .
. The study of' the reactor coolant syste= is li=ited to the
effects of the reactorpressure vessel on the attached pri=ary
piping for the faulted conditions inSection h .l. Stresses in the
piping at the reactor pressure vessel nosnle connec-tions are
considerably lover with the addition of the upper lateral support.
It'is expected that the addition of the supports vill cause changes
in the seismic. response spectra of the smaller attach =ent piping.
It is also possible thatthe redistribution of nc=ents in the
pri=ar/ piping could cause the change ofbreak locations for a LOCA.
It is expected that the break location effects villbe limited to
one change in split orientation or possibly a~ change from a
longi-tudinal split to a guillotine.
h .h FUfJRE FLAUNE ANALYSIS
The evaluation of the upper lateral support desien is
preli=inar/ in nature anda full analysis of the reactor pressure
vessel, reactor pressure vessel inter::als ,and reactor coolant
syste= is planned to determine the effects of the change
in-design.
5.0 cup 3EE STATUS OF CONST"JC"'ICN
51 DEPENSIONING OF STUDS
The anchor studs of Unit 1 are being detensioned in a series of
passes with areduction of pretension by one-third of the original
value on the first pass.The first stage detensioning pass had been
c0=pleted. The second phase of
,
detencioning vill censist of reducing the current pretension
value by one-thirdfor fcur studs inside the reactor vessel skirt
and checks vill be =ade for any
signs of a preload increase on any of these eight studs over the
values deter-=ined u the end of the first pass. If it is found that
no significant transferof load is occurring, the third pass vill
consist of detensioning all studs totheir final value of 6 ksi.
However, if significant load transfer is evident,the third pass
vill consist of another one-third reduction of preload prior toa
final detensioning pass. The frequency distribution of lift-off
tensilestresses in the detensioned studs are shewn in Figure 8.
52 STATUS OF 3 RACKETS,
Unit 1
The e=bedments 'to which the brackets vould be attached to are
already e= bedded in'
the pri=ar/ shield vall ecccrete. The brackets , as originally
designed for sup-porting the shield plugs, have been fabricated;
hcvever, they are not yet velded'
' to the e= bed =ents .
Unit 2
:The e= bed:ents ' are already embedded in the primar/ shield
wall concrete. TheI -brackets as originally designed for supporting
the shield plugs have already
been velded to the e= bed ents. However, the final concrete ;
cur (which follows'velding) at the base of the bracket has not yet
been done.
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* ' Midland Plant Units 1 and 2RFV Support Modification
6.0 SCHEDULE FOR F0DIFICATION '40FK-
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A preliminar/ schedule, Revision F, has already-been presented
to the !aC'(see Figure 9) This schedule has recently been updated
to Revision 0 (see.Figure 10). Table 1 indicates the changes fres
Revision F to Revision 0 forseveral key activities. The latest
possible date of the NRC's concurrenceto the proposed nodification
work, so that construction activities can be,initiated in time to
avoid impact on system turnover and scheduled fuel load,has been
revised to December 1980 rather than June 1981 (as indica,ted
inRevision F) . We request completion of the :mC review by the
dates indicatedin the revised schedule. The next report to the URC
vill be submitted ir
October 1980.'
7.0 REFERENCES
1. Teledyne Engineering Services Report TR-3887-1, Revision l',
Investigationof Preservice Failure of Midland RFV Anchor Studs, May
15, 1980
2. Teledyne Engineering Services Report, TR-3887-2, Revision 1,
Acceptabilityfor Service of Midland FFI Anchor Studs, May 20 ,
1980
3. .Teledyne Engineering Services Report , TR-3887-1, Addendus
1, Investigationof Preservice Failure of Midland RFV Anchor Studs,
June 6,1980
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Midland Plcnt Unita 1 and 2RPV Support Modification
.
TABLE 1
EFFECTS OF SCHEDULE REVISION
DateDescription Rev. F Rev. O
Teledyne ReportPhase 1 only 5/80 5/80 (completed)Addendum 6/80
6/80 (completed)
Start of NRC review - 6/80 6/80
B&W - Analysis completed 9/80 10/80
Bechtel - Civil evalua- 9/80 10/80tion completed
Start of NRC confirma- 9/80~
10/80tion of release
Early NRC release and 10/80 11/80decision
Latest possible NRC 6/81 12/80release
Construction start 11/80 12/80i
System checkout start 6/81 10/81
System turnover date 2/82 12/81
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