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REACTOR PRESSURE VESSEL

SUPPORT MODIFICATION

FOR

MIDLAND NUCLEAR POWER PLANT,

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; MIDLAND, MICHIGAN

PRELIMINARY REPORT NO. 1

JULY 1980

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CONSUMERS POWER COMPANY

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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 j

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,

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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

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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

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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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