ML19330B488

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Reactor Pressure Vessel Support Mod for Midland Nuclear Power Plant, Preliminary Rept 1
ML19330B488
Person / Time
Site: Midland
Issue date: 07/24/1980
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML19330B433 List:
References
NUDOCS 8008040083
Download: ML19330B488 (19)
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Text

{{#Wiki_filter:_ ..L. ../D J ? REACTOR PRESSURE VESSEL SUPPORT MODIFICATION FOR MIDLAND NUCLEAR POWER PLANT MIDLAND, MICHIGAN PRELIMINARY REPORT NO. 1 JULY 1980 i + 1 CONSUMERS POWER COMPANY I JACKSON, MICHIGAN 800804o c(P3

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REACTOR FRESSURE VESSEL SUPPORT MODIFICATION ICR MIDLAND NUCLEAR POWER FLANT TABLE OF CONTE'iTS Pace 1.0 DITRODUCTION 1 2.0 DESCRI>? ION OF TEE EXISTEIO REAC"0R VISSEL SUPPORT DESIGN AND 1 CRITERIA 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 REFERECES 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 16 l 10 Work Schedule, Rev 0 lI l

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 of three reactor vessel (RV) anchor studs. Figure 1 shows the location of the three failures. The first two failures were in the upper threaded section of the studs and the third failure was in the lover threaded section (see Figure 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 two concentric 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 Engineering Services (TES) of '4altham, Massachusetts, and was reviewed by Aptee of northern California, 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 studs should be detensioned. In their current condition, the studs cannot take the =oments and the uplift transmitted by the RV. However, =odification of the RV 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. Later sections 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 designed without modifications. However, pending final confimatory analyses of the Unit 1 =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= is co= 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 69 edition, for steel and the ACI 318-63 or 71 Code for concrete. '"he cylindrical skirt and emlar flange supporting the RV are =ounted on 5-1/2 inch thick annular sole plate seg=ents. The skirt flange botto: is the interface point between the nuclear stes: supply syste= vendor and the architect / engineer. The sole 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 the sole plate with the 96 previously described anchor studs. These studs extend thraugh the bottom flange of the RV skirt, sole plates, and the anchor plate which is e= bedded in concrete. 1 i

Midland Flont Units 1 and 2 RF'l Support Mcdification j The lateral and torsional loads are resisted by shear pins and shear lugs and are then transmitted to the primary shield vall. The sole plates, anchor plates, and shear lugs are =ade of ASTM A 36 =aterial. The shear pins are made of ASTM A 354 Grade 3D =aterial. The loads contributed by the RV to the fcundation are obtained from the reactor coolant syste= analysis performed by Babcock & Wilcox. The design require =ent imposed by B&W on the holddevn syste= vas that the studs be tensioned to a final stress 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. A spring rate curve showing the relationship between rotational stiffness and rotation =c=ent based on 55 ksi final prelcad has been obtained by Bechtel and was used by 3&W to perfer= its reactor coolant syste= enalysis. The RV support system is classified as a Seismic Category I structure and is designed for all credible conditions of loadings, including ncr=al loads, loads resulting from a loss-of-coolant accident, the:=al loads, and seismic loads. The applicable lead ec=binatiens can be found in Subsection 3.S.6.3 of the Midland FSAR. The governing load condition is: D+L+R+T +HA + E' A where D = dead leads L = live loads R = local force or pressure caused by rupture cf any one pipe T. = total the:=sl effects which =sy occur during a design accident other ^ than those considered by Hg H, = force on the structure due to ther=al expansion of pipes a E' = safe shutdown earthquake lead -F = yield stress of material 7 The maxi =u= allevable stress in bending and tension is 0 9 F., and for shear it is 0 5 F. All loads were generated by 3&W on the basis that the anchor studs 7 will 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 RV vill be used as lateral restraints to resist the overturning =c=ent that produces the tensile forces in the anchor studs. As shown in Figure 5, this lateral support vill enhance the supporting syste= by reducing displace =ent of the vessel without requiring the anchor studs to be prestressed in excess of 6 ksi. 2

- '~ Midland Plant Units 1 and 2 RFV 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 as shown in Figure 7. If the gaps between the brackets and the vessel are shi=med in the hot condition, the brackets vill provide the required upper lateral supports. Preli=inary calculations have indicated that after stiffening the brackets as shown in Figure 7, they will be capable of taking the new loads that they vill be subjected to in their new function as upper lateral supports to the vessel. Thesa same calculations dec:onstrated that the vall vill be capable of resisting the new leads transmitted via the brackets. - During the cold shutdevn condition, the upper lateral support vill separate from the RV due to contraction of the RV. In this condition, preliminary calculations indicate that"the displacement of RV under a SSE event is s=all enough such that no contact with the upper lateral support vill be =ade, and the. 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 operating condition 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 support system, namely, the anchor studs and upper lateral supports, to assure that plant operating 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. The design basis vill be same as that used for the design of other cc=ponent supports such 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 Reinforced Conc ret e" b. Structural Steel: AISC-1970 " Code of Standard Practice for Steel Buildings and Bridges" ~c. Welding: ' AWS Dl.1-72 including Revisions and Addenda up to and including July 197h ." Structural Welding code" 'i The loading conditions and allevable stresses are as stated in Section 3.8.6 of the Midland FSAR. The allowable stresses for the anchor-studs were based on the results of TIS Report TR-3887-2 (Reference 2l with the following acceptance criteria being recon = 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 considered to contribute to load-carrying capability in the new design. _ The detensioning load 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 3

Midland Plant Units 1 and 2 RPV 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 along with zero pretension loaded anchor bolts. The load cases analyzed were safe shutdown 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 done with 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 flange and the upper lateral support are given belev for illustrative purposes. AT RV SKIRT FLA: IGE LF/E r v

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~ hear ' vertical corizontal torsional s (ki s) (kins) ( ft-ki rs ) (ft-kirs) SSE llh 233 1h7 1,6h6 LOCA 1,003 3,3h7 3,529 1,113 AT UPPER LAT BAL SUPPORT LE/EL (RADIAL LOADS) Wall Individual Support (kirs) (kirs ) SSE 166 55 LOCA 3,377 1,126 Primarily, calculations have indicated that after stiffening the brackets, as shown in Figure 7, the brackas will te capable of taking the leads tabulated above. "'he sa=e calculations have de=onstrated that the vall vill be capable of resisting the new loads transmitted via the brackets and the stresses in the re=aining anchor studs will be less than the specified allevables. The upper lateral supports significantly lever the forces and noments that the anchor bolts vculd be required to carry with the exception of loads caused by vertical and torsional notion of the RV. S&W concludes that the addition of the upper lateral supports is an effective =eans of re=cving lead frc= the anchor belts. h.2 EFFETS 0 I RElCTOR VESSEL AD I' ITER:IALS FROM TIE ADDITIO i CF THE UPPB LATIRAL SUPPCETS S&W has co=pleted a stress evaluation of the reactor pressure vessel based on the support scheme and load cases =entiened in Section h.1. It was concluded that the vessel stresses have an adequate safety nargin for the faulted condition. Displace =ent of the reactor pressure vessel is reduced for the faulted condition by 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 =oments on key reactor internal connections. k

Midland Plant Units 1 and 2 Rpy Support Modification h.3 REACTOR COOLA:!T SYSTDI. . The study of' the reactor coolant syste= is li=ited to the effects of the reactor pressure vessel on the attached pri=ary piping for the faulted conditions in Section 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 that the redistribution of nc=ents in the pri=ar/ piping could cause the change of break locations for a LOCA. It is expected that the break location effects vill be 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 and a 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 a reduction 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-third for 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 transfer of load is occurring, the third pass vill consist of detensioning all studs to their 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 to a final detensioning pass. The frequency distribution of lift-off tensile stresses 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. The I

-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. 5

Midland Plant Units 1 and 2 RFV Support Modification 6.0 SCHEDULE FOR F0DIFICATION '40FK 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 for several key activities. The latest possible date of the NRC's concurrence to 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 in Revision F). We request completion of the :mC review by the dates indicated in 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', Investigation of Preservice Failure of Midland RFV Anchor Studs, May 15, 1980 2. Teledyne Engineering Services Report, TR-3887-2, Revision 1, Acceptability for Service of Midland FFI Anchor Studs, May 20, 1980

3..Teledyne Engineering Services Report, TR-3887-1, Addendus 1, Investigation of Preservice Failure of Midland RFV Anchor Studs, June 6,1980 m

Midland Plcnt Unita 1 and 2 RPV Support Modification TABLE 1 EFFECTS OF SCHEDULE REVISION Date Description Rev. F Rev. O Teledyne Report Phase 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 10/80 tion completed Start of NRC confirma-9/80 ~ 10/80 tion of release Early NRC release and 10/80 11/80 decision Latest possible NRC 6/81 12/80 release Construction start 11/80 12/80 i System checkout start 6/81 10/81 System turnover date 2/82 12/81 1 I t i 7

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