ML20038B441

Interim Deficiency Rept Re Reactor Vessel Broken Anchor Bolt,Initially Reported 791012.Two Repts Encl Describe Design Concept,Analytical Techniques & Completion Schedule for Mod of Reactor Vessel Support Sys.Final Rept by 811203

ML20038B441
Person / Time
Site:	Midland
Issue date:	11/23/1981
From:	Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
10CFR-050.55E, 10CFR-50.55E, 15035, NUDOCS 8112080292
Download: ML20038B441 (150)

Text

{{#Wiki_filter:~o. \\ Consumers Power James W Cook Vice President - Projects, Engineering and Construction o.a.r.: omc : 1945 West Parnell Road. Jackson, MI 49201 e (517) 788 o453 79-10 #11 -November 23, 1981 <3 Mr J G Keppler, Regional Director 8 Office of Inspection and Enforcenent l2-h; g 799 Roosevelt Road h- /g8 ,s j US Nuclear Regulatory Commission \\' / NI Region III %4 j/ ^ G) Glen Ellyn, IL 60137 -j N 3 MSLU MIDLAND PROJECT - MIDLAND DOCKET NOS 50-329, 50-330 UNIT NO 1, REACTOR VESSEL BROKEN ANCHOR BOLT - FILE 0.4.9.35 SERIAL 15035 REFERENCES 1. CONSUMERS POWER LETTERS TO J G KEPPLER, SAME SUBJECT a. HOWE-267-79 DATED OCTOBER 12, 1979 b. HOWE-311-79 DATED DECEMBER 14, 1979 c. HOWE-51-80 DATED MARCH 3, 1980 d. HOWE-80-80 DATED APRIL 30, 1980 e. SERIAL 8971 DATED MAY 16, 1980 f. SERIAL 8809 DATED AUGUST 1, 1980 g. SERIAL 9330 DATED JULY 24, 1980 b. SERIAL 9787 DATED DECEMBER 10, 1980 i. SERIAL 11524 DATED MARCH 31, 1981 . ;g : s

j. SERIAL 12051 DATED JULY 17, 1981 t ;k. s.

2. J G KEPPLER LETTER TO S H HOWELL, DOCKET NOS 59-329 AND 50-330 DATED AUGUST 18, 1980 3. R L TEDESCO LETTER TO J W COOK, DOCKET NOS 50-329 AND 50-330 DATED MARCH 6, 1981 4. D S HOOD LETTER TO CONSUMERS POWER DATED JULY 7, 1980 Enclosures 1. Report entitled, " Reactor Pressure Vessel Support Modification for Midland Nuclear Power Plant, Midland, Michigan, Report No 3", dated November 1981. 2. Letter Report - Teledyne Engineering Services (TES) Project 5355: Expanded Criteria for Acceptability for Service of Midland Un't 1 RV Anchor Stress. 811208029281112b DR ADOCK 0500o Q DECd 2 198h oc1181-0965a141 16 M A.A

..g , SERIAL 15035 2 References 1.a through j were Interim 50.55(e) reports, concerning the broken anchor bolts in the Unit No I reactor vessel support skirt. Reference 1.g provided interim technical information concerning-the reactor pressure vessel support modification and the schedule for the accomplishment-of that modification. Reference 1.h provided the description'of the analytical techniques being used that the NRC had requested in Reference 4. to this report supersedes References 1.g and 1.h by providing updated and' current information as to the design of the modified support system, analytical techaiques to be used and the completion schedules. provides a report from Teledyne Engineering Services on expanded acceptance criteria for the anchor stud stress. The two enclosed reports comprise a complete and current package of documentation describing the design concept, the analytical techniques-to be used and the completion schedule for the modification of the teattor vessel support system..The reports are in concurrence with the requirement in Reference 3 to keep the NRC informed of developments and progress made by the Company with regard to this issue. Immediately following NRR's review of the enclosures, it is the Company's intent to meet with NRR staff members on December 3, 1981 to present a summary of these reports and to resolve any concerns they might have and thereby obtain formal recognition that the conditicas and understandings specified in References 2 and 3 have still been satisfied. This letter is intended to be an interim'50.55(e) report transmitting our final technical report on the reactor vessel anchor bolt modification. The-final 50.55(e) report will be submitted on or before December 3, 1981. Upon completion of this task, the final designs and analytical results will.be reported in the FSAR. I JWC/BFH/cl cen81-0965alh1

, SERIAL 15035 3 CC Director of Of fice of Inspection & Enforcement (15) Director, Office of Management, Information and Program Control (1) Atomic Safety and Licensing Appeal Board CBechhoefer, ASLB w/o MMCherry, Esq RJCook, Midland Resident Inspector FPCowan, ASLB w/o RSDecker, ASLB w/o HDenton, NRC (5) SHFreeman, Esq, Ass't Attort y General w/o JHarbour, ASLB w/o DSHood, NRC (2) FJKelley, Esq, Attorney General w/o WHMarshall WDPaton, Esq w/o MSincit.ir w/o GTTaylor, Esq, Ass't Attorney General w/o oc1181- 0965a 141

~ ) '#TELEDYNE ENGINEERING SERVICES 3a3 EEAR M!LL RCAO v.sM. whvwiii5 0a54

sm snma nw a> =mca October 6, 1981 TR-5355-1 TELEDYNE ENGINEERING SERVICES Mr. Harvey W. Slager CONTROLLED Consumers Power Company DOCUMENT 1945 W. Parnall Road TES PROJ NO. dW5 P. O. Box CP 10-4672-Q Jackson, Michigan 49201 DATE

/O A-4/

Subject:

Letter Report - Teledyne Engineering Services (TES) Project 5355: Expanded Criteria for Acceptability for Service of Mid-land Unit 1 RPV Anchor Stress

References:

1. TR-3887-2, Rev.1, Acceptability for Service of Midland RPV Anchor Studs, TES, May 20, 1980 2. ASME Boiler and Pressure Vessel Code, Section III, Sub-section NF and Appendices with Addenda through Summer 1979 3. USNRC Regulatory Guide 1.124, Rev. 1, Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports, January 1978

Dear Mr. Slager:

This letter is prepared in response to your request that we expand the acceptance criteria originally established by Reference 1 to incfude selective proof testing and Level D Service Limits. As in Reference 1, the contents of Reference 2 are applied for guidance in the evaluation even though Code rules are not requirements for Midland component supports. Based upon our investigation, we recomend that the following reformulated criteria be applied to determine the acceptability for, service of Midland Unit 1 RPV anchor studs: 1. The lowest load measured during the detensioning phase, either lif t-off or proof testing, of the studs during 1930 is considered to be the Test Load, TL, as defined by NF-3262 of Reference 2. The nominal tensile stress caused by this load, 75 ksi, is considered to be both the yield stress, S, and the ultimate tensile strength, S, of the 7 u stud. 2. Should subsequent evaluations indicate the need for higher values of TL, S and S for specific studs, it is acceptable to obtain such y u higher values by proof testing to the higher values. ENGINEERS AND METALLURGISTS ]

s \\ TN ENGNEERING SERVICES appendix to TES.TR-5355-1 (Cont'd) Page Two In fact, the present status of these studs is such that the NF-3260 rules for " Design by Load Rating" are most pertinent. In accordance with NF-3261: "The procedure for load rating shall consist of imposing a total load on one or more duplicate full size samoles of a component support equal to or less than the load unde'..ich the component support fails to perform its required func.fon." In this application, each and every one of the RPV sttds has been 20 tested. Service Limits Given the proof test results, and utilizing the proof test stress as though it were the ultimate tensile strength, Reference 1 states as follows: " Based upon the experience to date, the argument could be made that the studs could be reloaded to the present preload stress value for some short period at any time in the future. To be conservative, the nominal ASME Code f actor of safety of 2 will be applied... " Reference 1 goes on to define the test load to which this nominal factor of safety would be applied, as described in the third paragraph under the heading " Material Behavior" in this appendix. The " nominal ASME Code factor of safety of 2" is that which applies to Levels A and B Service Limits in accordance with XVII-2461.1. System evaluations at that time indicated that the resultant allowable stress, expected to be 43 ksi or slightly less, would be sufficient with the modified design. This was still believed to be true when the detensioning was complete and a value of 37.5 ksi resulted. Subsequent evaluations have indicated that the stress level in a number of studs will exceed this value during system Faulted Conditions. Consideration of different allowable stresses for differing Operating Con-ditions was neither implicitly or explicitly excluded in Reference 1. However since the TES recomendation was silent on this matter, it was considered advisable to prepare expanded criteria. The stress which was present prior to detensioning, or the stress present during proof testing, continues to be conservatively considered as the yield strength and ultimate tensile strength of the material. Given this conservative way of defining the controlling material property, or considering the proof test load to be the Test Load used with design by load rating, it is appropriate to modify the f actor of safety as various Operating Conditions are considered. This has been done and the recom-mended Level C and Level 0 Service Limits are in conformance with both the ASME Code and with the Regulatory Guide, as applicable to linear-type component supports. ]

) s e Y ENGNEERNG SERVICES APPENDIX TO TES TR-5355-1 Material Behavior The RPV anchor studs were originally prestressed or proof tested to stress levels at or in excess of 75 ksi. Three of the studs failed by stress corrosion cracking. An investigation was performed and Reference 1 is one of a series of reports on that investigation. It was concluded that the studs should be detensioned, that retensioning should be to a very low value (to ensure metal-to-metal contact but to eliminate additional cracking) and that subsequent service stresses should be limited on the basis of the stress levels measured during detensioning. Specifically, that the Code factors of safety should be applied to the detensioning stress as though that detensioning stress level were the ultimate tensile strength. Alternatively, one could consider the deten-sioning load to be a Test Load on a linear-support designed by the Load Rating method defined by NF-3260. Both approaches are included in Item #1 of the letter. In order to control operations during detensioning and to assure that any low preload studs received attention, Reference 1 included the requirement that the detensioning load used in determining the allowable stress for all , studs ba 'the lowest measured detensioning load on any stud which is considero to contribute to load carrying capability." Subsequeiitly TES was asked to consider the posP bility of additional proof testing of spe-cific studs to justify a higher allowable stress value for those studs. TES is of the opinion that such se?ected testing is preferable to addi-tional proof testing of all studs to raise the lowest value, because any additional testing increases the probability that studs will fail during proof testing and additional reserve capacity during service is more important. This is the reason Item #2 has been included in the letter. Physical Description The RPV is supported by a full diameter skirt which terminates at a flange which rests upon an upper ring plate on the RPV pedestal. The flange is both inside and outside of the skirt, and there are 47 studs on a 48-stud bolt circle of lesser diameter than the skirt and 46 at a greater diameter. The studs are each 23s inches in diameter and 7 feet 4 inches long. The stud is tensioned between the upper surf ace of the skirt flange and a lower ring plate near the bottom of the studs, thereby clamping the lower surface of the skirt flange to the upper ring plate. This discussien is included to indicate that although the studs are threaded and made from forged bars, they are not part of a classic bolted connection, but are linear-type component support parts. Therefore, the concern expessed in Reference 3 concerning the applicability of NF-3231.1, XV11-2110(a) and F-1370(a) to bolts and to bolted connections is not ap-plicable to the RPV anchor studs which because of their length / diameter ratio cannot carry significant shear stresses. J

E 3 o . r WM ENGNEERING SERVICES Consumers Power Company Page Two TR-5355-1 October 6, 1981 3. The studs shall be considered to be threaded parts of linear-type supports. On this basis: Level A and Level B Service Limits f In accordance with Reference 1 (and with NF-3281 and XVII-2460) the average tensile strength, Ftb, computed on the basis of the actual tensile stress area available (independent of any initial tightening force) shall not exceed: S (Ftb)A = (Ftb)B *- Level C Service Limit In accordance with XVII-2110(a),. and consistent with NF-3262.3 and Regulatory Position 6 of Reference 3, the allowable stress may be increased by one-third, giving: S 2S 4 u u (Ftb)C " 3 T " 3 Level D Service Limit In accordance with F-1370(a) of Reference 2, and consistent with Regulatory Position 7 of Reference 3, the allowable stress may be increased by a factor of 1.4, giving: 1.4 S (Ftb)D " 2 u = 0.7 S The appendix to this letter discusses the recormiended limits for Levels C and D. If there are additional questions I would be pleased to respond. Very truly yours, TELEDYNE ENGINEERING SERVICES P William E. Cooper j Consulting Engineer WEC/lh Attachment f 1

q 3 (i A 4 REACTOR PRESSURE VESSEL SUPPORT MODIFICATION FOR MIDLAND NUCLEAR POWER PLANT REPORT NO. 3 NOVEMBER 1981 CONSUMERS POWER COMPANY JACKSON, MICHIGAN i I mi1181-0953a141 J

r3 1 4 8 o REACTOR PRESSURE VESSEL SUPPORT MODIFICATION FOR MIDLAND NUCLEAR POWER PLANT TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

1

2.0 DESCRIPTION

OF THE SUPPORT SYSTEM MODIFICATIONS 4 3.0 FUNCTION AND DESIGN CRITERIA 14 3.1 SAFETY DESIGN FUNCTION 14 3.2 FUNCTION CRITERIA 14 3.3 DESIGN CRITERIA 16 3.3.1 Introduction 16 3.3.2 Codes and Regulations 16 3.3.3 Midland FSAR 17 3.3.4 System Classification 17 3.3.5 Construction Material 17 3.3.6 Design Loads 18 3.3.7 Load Combinations and Allowable Stresses 18 3.3.8 Parameters to be Considered in the Design 20 3.3.9 Tolerances 21 4.0 GENERATION OF PRELIMINARY SUPPORT LOADS 22 4.1 PRELIMINARY LOCA LOADS 22 4.1.1 Design Basis Breaks 22 4.1.2 Analytical Model 23 4.1.3 Design Basis LOCA Loads and Displacements 23 4.2 DEAD LOADS AND THEIL%L LOADS 24 mi1181-0953a141 i i

y 3 o e 4.3 PRELIMINARY VS FINAL LOADS 24 5.0 MALYSIS AND DESIGN OF TE SUPPORT SYSTEM 29 5.1 UPPER LATERAL RESTRAINTS 29 5.1.1 Upper Lateral Restraint Brackets - Maximus Design Vs 29 Allowable Stresses and Displacements 5.1.2 Embedments - Maximus Design Vs Allowable Stresses 30 5.2 ANCHOR STUDS 31 5.2.1 Analytical Model to Determine Stress Distribution 31 5.2.2 Maximum Design Vs Allowable Stresses 32 5.2.2.1 For Unit 1 32 5.2.2.2 For Unit 2 33 6.0 EAT TRANSFER A'ID TERMAL ANALYSIS 37

6.1 INTRODUCTION

37 6.2 TEMPERATURE AND PRESSURE CONDITIONS OF REACTOR PRESSURE VESSEL NEAR UPPER LATERAL RESTRAINTS 6.3 UPPER LATERAL RESTRAINTS-VESSEL INTERFACE 39 6.4 PRESSURE DEFLECTION OF VESSEL 40 6.5 POTENTIAL GAP CHANGES DURING OPERATION 41 6.6 CREEP, THERMAL RACHETING, AND ELASTIC SHAKE DOWN 44 7.0 REACTOR PRESSURE VESSEL SURFACE PREPARATION 47 8.0 DENTENSIONING AND TENSIONING OF TE ANCHOR STUDS 48 8.1 DETENSIONING PROCEDURE 48 8.2 CREEP RECOVERY 48 13 RETENSIONING PROCEDURE 49 9.0 REACTOR PRESSURE VESSEL INSULATION MODIFICATION I 50 i 10.0 GAP AND TEMPERATURE MEASUREMENTS AND GAP SETTING 51 10.1 MEASUREENTS DURING HOT FUNCTIONAL TESTING 51 10.2 MEASUREMENT PROCEDURE 51 mil 181-0953a141 ii f ,,..-.,,._,m.mm,,.,,,c._-....

( O 10.3 CORRELATION BETWEEN MEASURED AND CALCULATED VALUES 51 10.4 SETTING THE GAP 52 11.0 ANALYSIS TO DETERMINE FINAL SUPPORT LOADS 54 11.1 GENERATION OF SUPPORT LOADS 54 11.1.1 Technical Basis 54 11.1.2 Mathematical Model 55 11.1.2.1 NSSS model 55 11.1.2.2 Internal wall structure 57 11.1.2.3 NSSS supports 58 11.1.2.4 Stiffness of upper lateral restraints 59 11.1.2.5 Stiffness of the support at the base 61 of the reactor pressure vessel 11.1.3. Load Cases Analyzed 62 11.1.4 Methods of Analysis 63 11.1.4.1 Seismic forcing functions 63 11.1.4.2 LOCA forcing functions 63 11.1.4.3 Computer codes used for NSSS Analysis 65 11.1.5 Seismic Analysis 68 11.1.6 LOCA Analysis 70 12.0 CHECKING SYSTEMS AND SUPPORTS FOR THE RESULTS 85 FROM FINAL ANALYSIS 13.0 CONSTRUCTION STATUS AND SCHEDULE 86

14.0 CONCLUSION

S 87

15.0 REFERENCES

89 mi1181-0953a141 iii

s APPENDICES A. Detensioning and Retensioning Reactor Building Reactor Vessel Anchor Studs. B. Gap and Temperature Measurement at the Reactor Pressure Vessel Upper Lateral Supports. C. Unit 1 Anchor Stud Lift-Off Data. LIST OF FIGURES 1.1 Positions of Failed Studs in Unit 1 3 2.1 Elevation View of the Reactor Pressure Vessel 6 2.2 Plan View of the Upper Lateral Support 7 2.3 Upper Lateral Support Bracket Detail (Typical 8) 8 2.4 Upper Lateral Support Bracket Detail (Typical 4) 9 2.5 Upper Lateral Support Bracket Embedment Detail 10 2.6 Plan View of the Reactor Pressure Vessel Lower Support 11 2.7 Anchor Stud Detail 12 2.8 Shear Pin Detail 13 4.1 Force in Bumper Versus Radial Gap 26 4.2 RV Displacement and Base Anchor Momement Versus Radial Gap 27 4.3 Deflection in Most Critical Bumper Versus Radial Gap 28 5.1 Bracket Analysis Sections 34 5.2 Finite Element Model of the Reactor Pressure Vessel Skirt 35 5.3 Position end Numbering of Anchor Studs in Units 1 and 2 36 11.1 RV Isolated Model - Reactor Internals and SSS 71 11.2 RV Isolated Model - Plan View 72 11.3 RV Isolated Model - Elevation View A-A Hot Leg 73 11.4 RV Isolated Model - Vertical Wall and Reactor Vessel Bumper 74 Elevation '1.5 RV Isolated Model - Elevation View C-C Cold Leg 75 mil 181-0953a141 iv {

c 4 11.6 RV Isolated Model - Elevation View B-B Cold Leg 76 11.7 RV Isolated LOCA Model - Vertical Wall and Reactor Bumper 77 Elevation 11.8 Reactor Internals and Service Support Structure 78 11.9 Reactor Coolant System Boundaries 79 11.10 Utilization of Computer Programs 80 11.11 Bracket Resistance Versus Displacement Curve Hot Leg Direction 81 11.12 Bracket Resistance Versus Displacement Curve, Core Flood Line 82 Direction 11.13 Bracket Resistar.ce Versus Displacement Curve, Cold Ecg Direction 83 11.14 Rotaticnal Spring Constants at the RV Base 84 1 4 i [ mi1181-0953a141 v

a 1 4 f

1.0 INTRODUCTION

Unit 1 of Consumers Power Company's Midland Plant experienced failure of three reactor vessel anchor studs several weeks after being tensioned to a nominal value of 92 kai in their tension area. Figure 1.1 shows the location of the three failures. The anchor studs were purchased as ASTM A354 Grade BD, 2.5 inches in diameter and 7 feet, 4 inches long. There are a total of 96 anchor studs per reactor vessel in two concentric rings on each side of the reactor vessel skirt. Investigation of the failed reactor vessel anchor studs was performed by Teledyne Engineering Services (TES) (References 1 through 5). According to the investication, the failure was_due to stress corrosion crack propagation to a point where brittle fracture took place. i. Modifying the reactor vessel supporting system to include the addition of the upper lateral supports (ULS) above the reactor vessel notzies, j along with stressing the anchor studs to a reduced preload level, will l provide the necessary support for the reactor vessel (RV). Two reports 1 (See References 6 and 7) were transnitted to the NRC in July and i December 1980. The first report covered the initial design criteria of the new support system including the allowable stresses. The second i report covered preliminary design loads and methods of analysis. Other interim reports and responses to NRC questions have been provided and j -are listed in the transmittal letter for this report. i i mi1181-0953a141 i l

l i l 2 l \\ It has been determined for engineering reasons, it has been determined that the gap size between the RV and the ULS should be increased from l the noeinal 1/32 inches previously reported to the NRC, to a gap size large enough to avoid contact between the RV and ULS during a seismic event and continue to provide the necessary lateral support for the RV from the design basis loss-of-coolant accident (LOCA). This Report Number 3 provides the required details for both the design and the analytical methods used, and thus satisfies the commitments made by the Company to the NRC. This report supersedes both previous reports (See References 6 and 7) by presenting the previous and new material in a single document. Where differences occur in either the design or the analytical methods between this report and the two previous reports, this report takes precedence and reflects the product ' of studies which have been performed to both enhance the modified RV support system and to assure the Company that the final design adequately meets all safety requirements. i l I l mi1181-0953a141 I

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2.0 DESCRIPTION

OF THE SUPPORT SYSTEM MODIFICATIONS The brackets that were originally provided to support the cavity annular shield plugs at the top of the RV have been reinforced to also serve as the ULS, to partially resist the RV overturning moment from the design basis loss of coolant accident (LOCA), thus reducing the stresses in the anchor studs. As can be noted from Figure 2.1, the brackets are located opposite the RV between the head flange and the nozzle belt. There are 12 brackets in each reactor cavity, and they are approximately equally spaced as shown in Figure 2.2. All but four of the brackets are radially oriented with respect to the RV and the remaining four are oriented in the East-West direction. The brackets are welded to embedments in the wall as shown in Figures 2.3 zud 2.4, for the radially acd the East-West oriented brackets, re;pectively. The brackets are made of a material originally purchased.,as ASTM A516 steel however some of the A516 material 1-1/2 and 1-1/4 inch thick plates was not normalized. The impact properties of the material indicate that it is acceptable for use as a material purchased as impact specified ASTM A516. The embeded plates are made of ASTM A36 steel. Details of the embedments are shown in Figure 2.5. The ULS will have stainless steel shim packs permanently mounted at their ends to provide the required gap between the brackets and the RV as shown in Figures 2.3 and 2.4. The contact surface area of the ULS shim pack is 5 x 12 inches and has been machined to a surface roughness of 250. Opposite the ULS shim pack on the RV surface, a corresponding contact surface has been mschined flat for au area of 8 x 13.5 inches with a surface roughness of 250 or better. w e... the RV at a temperature of mi1181-0953a141 r 1

4 5 about 70*F, a gap of 15/32 (0.469) inches will be set between the ULS and the RV. During normal operation at 100% power, this gap will bc 0.121 inches. This gap has been determined to insure that the RV will only contact the ULS in the event of the design basis LOCA at 100% power operation. Further details as to how the gap size was determined are presented in the subsequent sections. The modification to the RV skirt flange support consists of reducing anchor studs prestressing load from the intended 75 ksi to only 5 ksi as discussed in Section 8. The 5 ksi prestress value reduces to 1.5 ksi during the normal operating coaditions as a result of increased m anchor stud temperatures. The anchor studs alone will resist overturning moments and uplift forces from all loads on the reactor except those from the design basis LOCA. The ULS will partially function to resist the design basis LOCA overturning moments on the ~ studs by limiting the RV displacement. Shear forces and torsional moments at the RV skirt flange support are transferred to the concrete pedestal by the shear pins between the RV skirt flange and the sole plate and the shear lugs welded to the bottom surface of the sole plate. Details of the RV skirt flange support are shown in Figures 2.6, 2.7 and 2.8. mil 181-0953a141

l ~ REACTOR PRESSURE VESSEL ELEVATION -..}g"; @*y j _l ) EL 627'-8" t g I I ~ g. x-3'-2-1/8" l 15'-7-% " OD \\ 4 31'-5 % " l i =,pg.,. * / a 9 l R =, _11'-0" h, R = _6' 0" t ( N _. E._-_. I m i FIGURE (2.1)

REACTOR PRESSURE VESSEL ~ UPPER LATERAL SUPPORT PLAN kUPPER 360* 0* LATERAL SUPPORT (WP) _ 17 / ' N 2.jQi "N s, / \\ [ x' 34* \\ '.f \\ l \\ / N N -.c i s 270a 90* .{ RV EDGE \\ / \\ / e ', yj' j / N. / CONCRETE SHIELD PLUG' [ PRIMARY SHIELD I \\ / p 180* s i FIGURE (2.2)

1 1 REACTOR PRESSURE VESSEL l UPPER LATERAL SUPPORT BRACKET DETAIL (Typical 8) l l f 3'-6" I I: FACE OF RPV [- T.O.S. EL 632'-3" l I l 4'-1%" $ ASTM A-540 1 I u l .J~ J-l! - _ _1__T f_ BOLTS 3 m =L+- 1 2" E]I_" 5 I is n a in i 1'-6 % " i n" n" u; { u u u n / I H H il n 6" o i l i f" '1 h j If2 22=22 2 -T ei . i 6" I SHIMS--HE I

-)

U U. BRACKET i l 1'-2" 6" _g t --{ 4: : : - L ASTM A 516, ASTM A 240 XM-19 ~- CARBON STEEL STAINLESS STEEL i PLAN AT EL 632'-3" \\ ELEVATION f I cn l FIGURE (2.3)

REACTOR PRESSURE VESSEL UPPER LATERAL SUPPORT BRACKET DETAIL (Typical 4) FACE OF. PRIMARY SHIMIS SHIELD WALL [ 3'-7-9/16" 3'-7-9116 " >l N4 I FACE OF ,f RPV N l T.O.S. EL 632'-3" I r' 'N i ll l r" _ _3 m = -;N l ll ll __ s E ]=lj-et

Q _, +1'-6%"

j -- t

I

/ i jL _j"_ _S. a_ j u i _i 32" p, g,, ~ l'f ~4 l -__ x { 6 1 [- ASTM A 516 , ASTM A 240 XM-19 j'.5" ~' 4 CARBON STEEL STAINLESS STEEL l ELEVATION PLAN AT EL 632'-3" FIGURE (2.4)

REACTOR PRESSURE VESSEL UPPER LATERAL SUPPORT BRACKET EMBEDMENT DETAIL m-N 1 %" 1'-0" T BAR 1%" x %" 1h" N l (typ) I ( 1 %,, 1%" 1'-6" p { 6"" ~ t ' N._ 6" 10%" A D' 8" g -h-gM j n 10 " X. t h >= d-1,_ k/ 5-B 8" g / ( 21/4" x 25" x 3'-0" g 6"" [ ~ (1 % / s (7/8" SECTION A-A M D SECTION B-B ELEVATION FIGURE (2.5) E

REACTOR PRESSURE VESSEL PLAN 360* O LATERAL E 2 . LATERAL t 1 REACTOR SKIRT ANCHOR BOLT $ o 'o SHEAR PIN + " o. o SOLE E o,o o O O 270 90* \\' 1, oo o o M 180* EL 603'-1" FIGURE (2.6)

e i l REACTOR PRESSURE VESSEL ANCHOR STUD DETAIL SEMFFIN HEAVY i REACTOR k,, SKIRT HEXAGONAL JAMB NUT R = T'-4 % " HEAVY HEXAGONAL NUT

1. 4 PLAIN WASHER 10".

(hardened) WASHER 1" THICK 2-518 " O ID x 5" OD [ [ ,][ 3" LATERAL t ASTM A 36 a n 1'-5 % " 7 PROJ WITH E[ h yl !! kN/ ' ASTM A 36 I/ e-2" LATERAL t 1'-2 % " THD 7 j E E l 3 4 5%" e t SOLE E / / <r 1 ASTM A 36 a 7 '-4 " 5'-1% " 1 -- ? N ~ 6 1%"x 1%" x 2'I " 2%" A ANCHOR STUD SHEAR LUG ASTM A 354, G3ADE BD ~ ASTM A 36 + r i /- E 3%" x 17" ASTM A 36 l c c T t3 i x ee g HEAVY HEXAGONAL NUT ~ 'SEMhFIN HEAVY FIGURE (2.7) HEXAGONAL JAMB NUT

i = i l REACTOR PRESSURE VESSEL i SHEAR-PIN DETAIL DRILL 1%" $ HOLE AND REAM TO 2.005" $ IN SOLE t. e R = 7'-4 % "--- 2.015 ' 2"$ 4 i REACTOR (+0 SKIRT l 4,01) 6 1116 " 5"t o i C _2" $ / l n 1 7 i g l 4Q / 6-1116 " j TOC EL f, I, 602'-3 % " 45* CHAMFER f fg j \\ j _\\ n v j v 3/4" 118 " I 1/16 " 2" h x 0'-61116" = (48-reqd) l \\ ASTM A 354, GRADE BD i I DRILL 1" $ HOLE FOR REAMER PILOT BAR i N FIGURE (2.8)

14 3.0 FUNCTION AND DESIGN CRITERIA 3.1 SAFETY DESIGN FUNCTION The safety design function of the RV support system is to provide support for the RV as specified in the following. 3.2 FUNCTION CRITERIA 3.2.1 The RV support system shall remain functional during a safe shutdown earthquake (SSE), or from a LOCA. The loads from SSE and LOCA combined. 3.2.2 The postulated LOCA shall be assumed under 100% power operating conditica. 3.2.3 The effects of jet impingement shall not render the RV support system inoperable during the postulated LOCA design basis rvents. 3.2.4 During the normal, upset, faulted, and test conditions with the exception of LOCA, as stated in 3.2.2 above, the following conditions must be met; Reactor coolant system (RCS) temperature variations a. resulting in RV radial and vertical expansions will not I result in forces being placed on the RV by the ULS. b. RCS temperature varia_tions resulting in RV radial and vertical expansions will not result in forces on the RV mi1181-0953a141

O 9 1 15 support system causing the system to be impaired or damaged i to the degree it cannot perform its safety design function. 3.2.5 Based on operating condition information, temperature variations resulting in RV and ULS radial expansion will not create a gap between the RV and the ULS small enough to cause contact during an SSE seismic event. t 3.2.6 The ULS shall be designed such that temperature variations t induced in the RV because of the proximity of the ULS and the insulation cutouts for the ULS do not result in RV stresses in. i excess of the RV acceptance criteria stated in the FSAR. 3.2.7 The RV support system shall be designed so that the temperature of the concrete in the local vicinity of the srpports shall not exceed 200*F during.all operational modes. 3.2.8 The RV support system shall be designed so that a continuous 40 year total radiation dosage will not result in unacceptable a degradation of the support material. I 1 3.2.9 The RV support system shall be designed assuming forced cavity air flow. Forced air flow shall be ensured or appropriate t operating restrictions shall be imposed in the event of an interruption, or loss, of forced flow shall be identified. 1 i 3.2.10 The temperature differences that may exist at different f locations on the RV during all normal operating conditions shall be considered in establishing the proper gap size at individual ( ULS in order to satisfy the functional criteria set forth above. 7 j mi1181-0953a141 I _.----,--.-s---

i 16 3.3 DESIGN CRITERIA 3.3.1 Introduction The criteria under this section shall apply in the design of the RV support system for Midland Plant Units I and 2. 3.3.2 Codes and Regulations The design of the RV support system shall conform with, but not be limited to the applicable codes and specifications listed below, except where specifically stated otherwise. 3.3.2.1 American Concrete Institute Building Code Requirements for Reinforced Concrete (ACI 318-71). 3.3.2.2 American Institute of Steel Construction Specification for the design, fabrication and erection of structural steel for buildings - 1969 Edition with Supplements 1, 2 and 3. I i l i l l i l l mi1181-0953a141

17 3.3.2.3 The following ASTM material specifications: A516 Grade 70 plate material

• E7018 Filler metal for shielded metal are welding A240 Type XM-19 stainless steel A354 Grade BD bolts and A325 high strength bolts **

niote: See the discussion in Section 2.0 on the A516 material.

• • Note: The anchor stud material for Unit I are subject to the conditions of the Teledyne reports (References 1 thru 5) since they do not satisfy the requirements for the A354 Grade BD material.

3.3.3 Midland Plant FSAR 3.3.4 System Classification The supporting system is classified as a Seismic Category I structure. 3.3.5 Construction Material 3.3.5.1 The concrete compressive strength (f' ) is 5,000 psi c and the reinforcing steel is ASTM A615 Grade 60. mi1181-0953a141

= 18 3.3.5.2 The structural steel materials are as follows: Plates A516, Grade 70* Filler metal for Welding E7018 Stainless stsel for the A240, Type XM-19 shim block Bolts in the ULS A325 Anchor studs and shear pins A354, Grade BD*

• Note: See the foot notes provided for section 3.3.2.3.

3.3.6 Desian Loads The RV support system will be designed to take loads frce: Permanent weights (DL) a. b. Stud prestressing (Pt) c. Operating thermal loads (To) d. Operating basis earthquake loads (OBE) Safe shutdown earthquake loads (SSE) e. f. Loads from design basis (LOCA) 3.3.7 Load Combinations and Allowable Stresses 3.3.7.1 Upper Lateral Supports DL + T,*** AISC specification allowable a. b. DL + T, + OBE** AISC specification allowable x 1.25 mil 181-0953a141

19 DL + T, + SSE** AISC specification allowable x 1.3 c. DL+T,+}SSEz**+LOCA'AISCspecificationallowablex1.5 d. 2

• Notes: Under LOCA loads, yield strain may be exceeded.

The maximum strain, however, shall not exceed 10 times the strain at the initiation of yielding. Since the RV will not be in contact with the ULS, the seismic loads will consist of permanent weight (DL) inertia loads. Thermal effects on the ULS only serve to reduce both the allowable stress and the Young's s modulus for the steel. 3.3.7.2 Anchor Studs a. P 5 6 ksi( } g i b. DL + T, + P 5 6 ksi and 2 1.5 ksi(2) c. DL + To + (SSE or P ) 5 0.5 Proof test load ( DL+T,+/SSE2 + LOCA ' 5 0.7 Proof test load ( } d. 2 mi1181-0953a141

20 Notes: (1) See Reference 2. (2) Required to mitigate normal operating vibrations (3) The faulted condition allowable stress level for the anchor studs has been increased from 0.5 P to 0.7 P as g g described 12 Reference 9. 3.3.8 Parameters to be Considered in the Design The fo! lowing parameters shall be considered in the design. 3.3.8.1 Irradiation effects: Gamma radiation heating. a. b. Embrittlement of the structural steel, filler material for welding, and anchor studs 3.3.8.2 Temperature variation and heat transfer: The temperature gradient in the brackets and the a. temperature at the bracket-concrete interface. b. Effect of temperature variation on the studs pretension load. mi1181-0953a141

.m._ W i i i - 21 i4 3.3.9 Tolerances l l 3.3.9.1 Construction tolerance in setting the gap of +1/64 inch ] is' allowed. I 3.3.9.2 Construction t.olerance in prestressing the studs of +0 5 ksi is allowed I I 4 i 'i 2 1 i i l i I 4 4 1 i i i mil 181-0953a141 4 4

• y,--=-..c---,mwg, ~. - y-e e. --g ew.yc,

--,.m.--.y,v.-,.w. _%y,--eu4.._,m...-,,-w.-..,.,, y -c - 9 9-n. w ry, g .ww-m-----m.-+, =~--ws, w---- - - - -,--v, e---

22 4.0 GENERATION OF PRELIMINARY SUPPORT LOADS The seismic analyses of the NSSS needed to generate the support loads have been finalized at this time and they are described in Section 11.0. Preliminary analyses of the RV with the modified support system for the design basis LOCA has been performed to allow for the design of the upper lateral supports to proceed. The following subsections describe the process by which the preliminary LOCA loads were dercloped. 4.1 PRELIMINARY LOCA LOADS To expedite the design of the modified reactor support system, a simplified nonlinear computer model of the RV shell, RV internals, and the concrete internal structure wall has been used to predict the desi n F basis LOCA loadings as a function of variable gap size between the ULS and the RV. 4.1.1 Design Basis Breaks Design basis LOCA breaks are assumed to occur at 100 percent power operating conditions. Two design basis breaks are considered: 1) 0.39A* guillotine at the RV outlet nozzle 2) 0.24A* guillotine at the RV inlet nozzle

• Note:

A is equivalent to the internal cross-sectional area of the pipe being considered. mil 181-0953a141

23 4.1.2 Analytical Model The analytical model used to determine the preliminary RV support system design LOCA loads is a simplified version of the model described in Section 11.0. The model consists of two springs and a single degree of freedom. One spring represents the combined spring rate of the RV anchor support, the RV support skirt and the hot leg and cold leg piping. The second spring represents the combined effects of localized wall spring rates, and radial flexibility of the RV shell. The single degree of freedom (SDOF) has a mass representative of the RV shell and RV internals. The mass / spring rates which represent the RV are developed such that the SDOF oscillator frequency closely matches that of the first mode of the RV shell and RV internals. Damping for the SDOF is 7 percent of the critical damping. Forcing functions, representing LOCA presures versus time, acting on the SDOF consist of combined time phesed phenomena of both asymmetric cavity pressure across the RV shell, and to pressure differentisls inside the RV. These forces are described in more detail in Section 11.0. 4.1.3 Design Basis LOCA Loads and Displacements The model described in 4.1.2 is subjected to loadings through gaps ranging between 0.0 inches and 0.3 inches. The resulting force in the ULS and the moment on the RV base anchor are depicted in Figures 4.1 and 4.2 as a function of gap size. mi1181-0953a141

24 Also d ermined was the deflection in the most critical ULS bumpe-The critical ULS is defined as the ULS subjected to the largest axial compressive deformations which therefore have the jvcential to exceed the ductility limits imposed by the criteria set forth in Section 3.3.6. Figure 4.3 illustrates the critical ULS bumpers, and ensuing deflections for both hot and cold leg LOCA's. A vertical force at the RV base of 4,697 kips is considered in the design. 4.2 DEAD LOADS AND THERMAL LOADS The modification of the RV support system does not affect the dead loads and thermal loads on the RV base support. There are no loadings on the ULS due to deadweight or thermal expansion of the NSSS. The previously calculated dead loads and thermal loads on the RV skirt base are given below. Deadweight = - 2595 kips Thermal (8% power) = + 420 kips Thermal (15% power) = + 356 kips Thermal (100% power) = + 330 kips

• Note:

A negative sign indicates a downward applied load, and positive sign indicates an uplift load. 4.3 PRELIMINARY VERSUS FINAL LOADS i The preliminary loadings given in the report have been determined to allow design of the modified RV support system to proceed in an orderly Models and forcing functions are similar and/or identical to manner. those used to produce final loadings, mil 181-0953a141

s. e 25 Deadweight and thermal loads are being revised to reflect refinements in the concrete building models. These results are not expected to vary significantly from those previously calculated. Seismic results as presented in Section 11.0 are final. LOCA displacements and loads are considered adequate for design use. The parameters reflected in the simplified LOCA model are those which are most significant in determining RV support loads. Although more detailed LOCA analyses are currently being performed (See Section 11.0) which will verify the preliminary loadings, the major reason for the more detailed analyses is to determine the effects of LOCA on the reactor internals. i l l l i i i i l mi1181-0953a141 l l L

4 26 6.0., FORCE IN BUMPERS VS. RADIAL GAP 50 HOT LEG GUILLOTINE G COLD LEG GUILLOTINE ---- e 4.0 /--.' ~ s c / a 's c s TOTAL FORCE IN ULS y 3.0. s 's 2 ' s 2.0. 1886 KIPS s s 1886 KIPS N s

1. 0 <.

h FORCE IN MOST \\g \\g CRITICAL BUMPER \\ 0.0 \\ O.0 0.1 0.2 0.3 RADIAL GAP (INCHES) FIGURE h.1 Force In Bumpers VS Radial Gap

27 RV DISPLACEMENT AND A'CHOR MOMENT VERSUS RADIAL GAP H0T LEG GUILLOTINE

0. 5' "

COLD LEG GUILLOTINE ---- -m E 90 G ux m c=, 0. 4. g, a <d s',- - 80 s' g-70 c c ]G 0.7 60 [=a ., ', [- ' ' 50 9 am s s z 95 $ n g- /'f/ 40 @E oG e :; ,s' ~30 gg >c: ao 0.1. ..20 !c 10 0.0 0.0 0.1 0.2 0.3 RADIAL GAP (INCHES) l l i FIGURE k.2 l RV Displacement And Base Anchor Moment Versus Re. dial Gap i

26 17 n 17 73 W N \\lig s =- 0 A '%43 h 470 43 h, 01 L CRITICAL CRITICAL BUMPER BU ER / 18.5 i 30" l j-18.5 l l .I SUPPORT CONDITIONS FOR SUPPORT CONDITIONS FOR HLG at RV CLG at RV DEFLECTION IN MOST CRITICAL BUMPER VS. RADIAL GAP 0.05 HOT LEG GUILLOTINE ^ g COLD LEG GUILLOTINE ---- N / G 0.04- \\ M / \\ 5 / \\ 5 0.03 / \\ e / N b \\ 0.02 J N / N g N \\ 0.01< g \\ 0.0 0.0 0.1 0.2 0.3 RADIAL GAP (INCHES) FIGURE h.3 DEF'ECTICN I '_j40ST CRITICAL Et2GER 'TERSUS RADIAL GAP Deflection in 7.he most critical bumper is given for the two LOCA cases. Displacements are measured in the axial direction. Any deflection due to bending was ignored. For a CLG (Gap > 0 inches), the critical bu=per was considered active after contact with the 730 bumper was obtained.

29 5.0 ANALYSIS AND DESIGN OF THE SUPPORT SYSTEM The reinforcement provided for the upper lateral restraint brackets were designed based on the preliminary loads described in Section 4.0. The stresses in the anchor studs were analyzed to show that they are within the allowable limits specified in Section 3.3.6. 5.1 UPPER LATERAL RESTRAINTS 5.1.1 Upper Lateral Restraint Brackets - Maximum Design Vs Allcwable Stresses and Displacements The stresses at the sections shown in Figure 6.1 were governed by the following load combination, DL+fSSE2t + LOCA2' The, resulting stresses at Sections 1, 4 and 6 are given in the table below: Maximum Design Stress Allowable Stresses Or Section Or Interaction Value* Interaction Value* 1** 9.32 ksi 33.48 ksi 4 0.963 1.000 6 1.000 1.000

• Notes:

The interaction of axial compression and bending is according to the AISC specification, Section 1.6.1. Critical loading on Section 1.1 is from cavity pressure only. t Inertia loads due only to ULS self weight, and weights supported by the ULS. mil 181-0953a141

30 It should be noted that the axial load in the bracket used is the maximum allowed which will result in yielding of the bracket based upon the minimum specified yield stress. From Figures 11.1 through 11.13, it can be noted that the maximum allowed displacement of the bracket is 0.2548 inches (based on a ductility ratio of 10). The preliminary calculations indicate a maximum displacement of 0.045 inches which is considerably less than the maximum allowed displacement. Refer to Section 11.1.2.4 for the discussion on the proper use of Figure 11.1.. 5.1.2 Embedments - Maximum Design Vs Allowable Stresses For the most critical load combination (DL + SSE2 + LOCA ), the 2 stresses in the components of the embedments are given in the table below. In the same table the corresponding allowable stresses are shown. Maximum Design Allowable Stress Stress (ksi) (ksi) a. Bearing stress behind 5.94 5.95 embedment plact b. Bending stress in 32.4 32.4 embedment plate l c. Tensile stress in 30.4 32.4 anchor bar d. Bearing stress between 3.6 5.95 l anchor block & concrete i e. Bending stress in 23.2 32.4 anchor block mil 181-0953a141

y a 31 f. Bearing stress between 2.2 2.975 shear lugs and concrete g. Bending stress in shear 26.4 32.4 lugs h. Shear stress in shear lugs 4.53 18.0

• Note:

It should be noted that the axial load in the bracket used is the maximum which will result in yielding based on the maximum yield stress of the material. 5.2 ANCHOR STUDS 5.2.1 Analytical Model to Determine Stress Distribution The reactor pressure vessel skirt and skirt flange have been modeled with flat rectangular shell elements (5 degrees of freedom per node) using the finite element computer program BSAP* (CE-800). The finite element model is shown in Figure 5.2. The anchor studs and the concrete pedestal are modeled using linear springs which can be axially loaded only. The stiffness of these springs in tension, is equal to the stiffness of the studs and their stiffness in compression, is equal to the stiffness of the pedestal. The loads are applied at the center of the circular top edge of the skirt which is connected to the nodes on the top edge of the skirt by a spoked arrangement of j rigid links, thus representing the boundary edge effect of the RV. The solution for the stud stresses is obtained through iteration by first assuming the position of the neutral axis and then checking the assumption and adjusting it as required until the mil 181-0953a141

32 location of the neutral axis is determined. Due to the non-linear nature of this analysis, this procedure was followed to calculate the stresses in the bolts due to loads from the design basis LOCA and East-West, North-South and vertical SSE earthquake, respectively.

• Note: The description of BSAP along with its validation is provided in Appendix 3C of the FSAR.

5.2.2 Maximum Design Vs Allowable Stresses 5.2.2.1 For Unit,_1 The maximum design stresses due to dead loads, thermal loads, SSE loads, and design basis LOCA loads are given in the table below along with their corresponding allowaole stresses. These stresses were calculated by combining the stresses from SSE and LOCA loads by the square-root-of-the-sum-of-the-square-root-of-the-square method (SRSS). The governing LOCA load case was a break in the cold leg of the NSSS closest to the two broken studs in the outer radius of Unit 1. mil 181-0953a141

4 33 Maximum Design Allowable Stress

• Stud Number Stress (ksi)

(ksi) Outside Diameter 39 33.2 57.4 33 37.5 59.5 37 44.8 52.5 36

---

BROKEN-------------

35

---

BROKEN-------------

34 49.2 59.5 33 44.6 64.4 32 43.0 58.1 31 41.9 65.8 Inside Diameter 39 28.8 55.3 38 30.0 55.3 37 29.7 61.6 36 29,1 55.3 35 30.3 55.3 34 33.5 62.3 33 36.4 63.0 32 37.5 58.8 31 37.4 33.9 For stud number location, see Figure 5.3.

• Note: The allowable stress 2s are obtained from Reference 9 and Appendix C.

5.2.2.2 For Unit 2 The allowable stresses will be determined in conjunction with Reference 9 and Appendix A at the time of detensioning. mil 181-0953a141

34 3 '-6 " = 42 " b @ @6 ,9", 8%",8 % " 8". 8" l S m l 0 M M M N N M M M M M M N N Mm 0 l 3 l e #,. = \\ /* ~ I i m BRACKET ANALYSIS SECTIONS FIGURE 5.1 -,-,--( w,

1 REACTOR PRESSURE VESSEL FINITE ELEMENT MODEL OF l SKIRT l 90* n r-l L j 16 " 0 90* 24 SPACED @ 3.75* [15" 180a / o' 1 ) 15" ~ ~ 8" / DEVELOPED VIEW OF RPV SKIRT , n. (Typical for Other Quadrants) RPV SKlRT fit R = 88.75" 1 FINITE ELEMENTS OF 270 RPV SKIRT FLANGE FIGURE (5.2) O

36 @ @@ @@@@ @@ @ b g g OUTSIDE g LED INSIDE 39 b ( REACTOR SKIRT FAILED -FAILED g 9, 0 @ INSIDE 18 OUT POSITION AND NUMBERING OF STUDS IN UNITS 1 AND 2 FIGURE (5.3)

37 6.0 HEAT TRANSFER AND THERMAL ANALYSIS

6.1 INTRODUCTION

An investigation into the relative action of the RV and the ULS as a result of variation in the NSSS system temperatures and pressures during normal plant operating and upset conditions has been conducted. The investigation included heat transfer analysis to determine the temperature distribution in the ULS, of the concrete behind the l*;S and the RV shell. In order to benchmark the investigation, the same techniques were used to predict the conditions of hot-fuactional testing (HFT). See Section 10.0 for further discussion of the HFT testing program. The calculated values will be correlated against measured values during HFT. The results obtained from the study, confirmed or modified during EFT, will be used to finalize the gap required between the upper lateral restraints and the RV. 6.2 TEMPERATURE AND PRESSURE CONDITIONS OF REACTOR PRESSURE VESSEL NEAR UPPER LATERAL RESTRAINTS The part of the RV shell opposite the ULS is exposed to the inlet, or cold leg, fluid temperature. The predominant temperature values are 532*F (0% power hot standby), 555*F (8% power), 575'F (15% power), and 554*F (100% power). This range covers all anticipated normal operating conditions other than heatup and cooldown which will have temperatures less than 532*F and two upset transients:

1) reactor trip (loss of feedwater), and 2) accidental rod withdrawal. These transients are similar with respect to inlet fluid temperature in that there is a change in temperature from 554*F to 590*F in approximately 30 seconds mil 181-0953a141

38 and then a return to 550*F in I to 3 minutes. (Note: The 590*F temperature is held for a very short time; i.e., approximately 10 seconds.) The reactor pressure is nominally at 2,250 pai. Again there are short time fluctuations where pressure varies over a range of 2,100 psi to 2,600 psi. Of course, during the heatup and cooldown operations the pressure falls below 2,100 psi. The outlet cozzles, or hot legs, below the restraints carry the coolant fluid which is at a higher temperature. An analysis bis indicated that the outside surface of the RV shell is 567'F at a distance of 41.5 inches from the nozzle centerline when the fluid temperatures for the hot and cold leg are 607'F and 555'F, respectively. The outside surface of the RV shell at 34.5 inches from the hot leg nozzle centerline is 575'F. The bottom of the machined contact surface opposite the ULS on the RV is 48.5 inches from the outlet centerline. Therefore, there will be very little if any effect of the hot leg fluid at the ULS location. As stated previously, the basic temperature conditions of the reactor inlet are below 575*F except for certain trip conditions. To eliminate considerations of these short-time temperature excursions, calculations were made with the aid of formulae developed in Reference 10. If the RV sLell external temperature is a uniform 550'F, and the inner surface temperature is increased to 590'F, the time required to raise outside surface temperature to 575'F is 1.53 hours. Since the entire transient is less than 5 minutes, it can be concluded that reactor inlet temperatures higher than 575'F need not be considered. mi1181-0953a141

39 6.3 UPPER LATERAL RESTRAINTS-VESSEL INTERFACE The parameters involved in modeling the RV interface with the ULS are very difficult to prscisely predict. Considerations are width of gap (AG, c nditions of surfaces, and possible air flow conditions. _Because of the nature of the insulation and proximity of the restraint to the vessel, convective effects should be small. Conduction through air and radiant effects are functions of the gap and surface conditions. The conditions affecting radiant effects will change over the plant life due to oxidation of the surfaces. To bound the potential effect of these parameters, three cases were analyzed. Two gapped cases 0.1 inch and 0.03125 inch, allowing conduction through the air as the only heat transfer mode, were analyzed. A third case assumed continucus metal f between the vessel and restraints. The vessel fluid and cavity air conditions were not varied. In all cases, convective effects were included on the ULS. The most severe temperature distribution in the vessel wall was identified and a finite element structural analysis was performed to determine stresses and thermal deflections associated with this condition. An isothermal c te was also run as a base comparison case. The fluid temperature was 570*F and the air temperature was 132*F. The hand-calculated radial thermal deflection is 0.3588 inch *. The computer-calculated deflection is 0.3589 inch *. The calculated radial deflection opposite the restraints due to the most severe thermal gradient is 0.350 inches. Therefore, with respect to the RV, the restraint to the RV interface has the potential for a small error from thermal growth considerations. mi1181-0953aI41 i

40

• Note: EcJed on AT = 570-70 = 500*F, a = 7.45x10 s, R = 96.3125 inches.

6.4 PRESSURE 9EFLECTION OF VESSEL The deflection of a cylinder at the outer surface subjected to internal pressure at the outer surface is: 2 = 2Pa b 9 E(b3-a ) Z where: A = The radial displacement due to i (= P for Pressure, or TH for Thermal) in component j (= RV or ULS) a = Outer surface radius b = Inner surface radius a = The coefficient of linear thermal expansion then: b = 96.3125 in., a = 84 in., E = 27.8x10s psi at 550*F. which reduces to: y = P(2.202x10'5) in. thus: y = 0.05 in., for P = 2250 psid y = 0.055 in., for P = 2500 psid y = 0.0605 in., for P = 2750 psid mi1181-0953a141

41 The closure of the RV hemispherical dome head does effect the RV deflection, however, the effect is small (maximum = 0.009 inches.) 6.5 POTENTIAL GAP CHANGES DURING OPERATION For the case when the RV starts in an untensioned (head bolts) and cold condition (70*F), the maximum thermal growth of the RV will be achieved if there is a prolonged hold at 15% power (575'F). A = AT a R R = (7.38x10 s)(96.3125)(505) = 0.359 in. where: .1T = 575-70=505'F a = 7.38x10 s at 575'F R = 96.3125 in. Assuming a design pressure condition, = 0.055-inch, the closure y effect(y,addsanadditional0.009 inch. ORV

• OR V*

V ARV = 0.359 + 0.055 + 0.009 = 0.423 inch at 15% power (maximum) For the case during either 0% power, or hot functional testing, the vessel is at 532*F and 2,200 to 2,250 psi. The 2,200 psi will be used because minimum deflections are of interest. ARV = (7.308x10~8)(96.3125)(462) = 0.325 in. mi1181-0953a141

42 e y = 0.048 in. The deflection due to closure is approximately 0.008 inches. Hence, the RV displacement is, ARV = 0.325 + 0.048 + 0.008 = 0.381 inch at 0% power For the case when the NSSS is at 100% power the vessel is at 554*F and 2,200 psid is assumed, hence ARV = (7.346x10~8)(96.3125)(484) = 0.342 in, y = 0.048 in. y = 0.008 in. ARV = 0.342 + 0.048 + 0.008 = 0.398 in. 100% power. The thermal growth of the ULS needs to be assessed for different gap conditions. The thermal distribution through the center of the ULS is as follows. (Node points are listed from the outer surface of the RV towards the concrete primary shield wall surface.) mi1181-0953a141 e-

O 43 Distance From Free Edge of Temperature Bracket Metal Node (inches) 0.1 gap (*F) 0.03125 gap ('F) continuity (*F) 613 (0.00) 184.8 260.9 476.4 638 (0.00) 177.7 244.7 451.9 686 (6.00) 157.7 196.4 327.9 734 157.0 194.8 322.4 858 153.0 184.6 290.9 859 150.8 179.0 273.4 860 150.1 177.2 267.7 861 (21.50) 149.8 176.5 265.5 959 149.8 176.5 265.5 Radial distance from RV and material is as follows: Distance Between Node Nodes Material 613 -> 638 0.08 in, stainless (A240 X M-19) 638 -> 686 5.92 in. stainless (A240 X M-19) 686 -> 861 15.5 in. carbon steel (A516 GR70) Node 861 represents radial location where bottom of the ULS enters concrete Location Metal Temperature 0.1 gap (*F) 0.03125 gap (*F) 0.0 gap ('F) 613 Average nodes 638 181.3 252.8 464.2 538 Average nodes 686 167.7 220.6 389.9 Average remainder of nodes excluding 152.1 182.4 284 . 959 Thermal growth of the ULS for a 0.1 inch gap is as follows: ~8)(111.3)(0.08) + (8.3x10[s))(82 1)(15 5) 3 = (8.3x10 8 97.7)(5.92) + (6.03x10 ( 3 =.(7.4x10 s) + (4.80x10~8) + (7.67x10~8) = 0.0124 in. r s SimilarIV, the thermal growth of the ULS for a 0.03125 inch gap is 0.0182 inches and for a zero inch gap is 0.0385 inches. -mi1181-0953a141 es 3

44 6.6 CREEP, THERMAL RACHETING, AND ELASTIC SHAKE DOWN 1 The long-term positional stability of parts is a consideration in the design of a gapped structure. This section is concerned specifically with the stability of the reactor vessel. The vessel was subjected to an ASME code heat treat at the conclusion of all welding. The vessel is in a vertical position and will not be again subjected to temperatures near the heat treatment range (i.e., T Peration is 575'F while the aax temperature at heat treat was 1,100 to 1,150'F). Therefore, additional stress relaxation would not be expected to occur, unless creep effect occurs. ASME Section III stress criteriIa requires that the highest temperature versus stress allowable be within the bounds of creep criteria stated in ASME Section I. The highest reported temperature in the 1968 ASME Section III for SS-508 CL 2 is 700*F. The reported Se value is 26,700 psi which is constant for the full temperature range. This is one-third the minimum ultimate strength of 80 ksi. The material SA-508 CL 2 is not in ASME Section I, but SA-302 GR B is listed and has strength and chemistry characteristics very similar to SA-508 CL 2. ASME Section I stress allowables do not diverge from 1/4a until a temperature of 800*F. Therefere, it can be concluded that the vessel is not operating in the creep range. Thermal racheting is another consideration in the design of a gapped structure. The relevant criteria is given in ASME Section III. Under this thermal condition, cyclic radial gradient thermal stresses occur in an essentially constant pressure stress field. mil 181-0953a141

45 , maximum general membrane stress yield strength

• Maximum general membrane stress = 2500(84)= 17.5 ksi 12 X = 17.5 ksi/26 = 0.673 l

Y,., maximum allowable _ range of thermal stress yield strength

• y',1, = M 1 - X) = 1.308 TH range = 1.308(26) = 34 ksi o
• Note:

For additional precautions against thermal rachet, the following endurance limit should be used: 2xS at 10s cycles for SA-506 CL 2; 2(13) = 26 ksi. a Under conditions where the pressure in the system is relatively constant, and the fluid temperature is between 532 and 590*F. The maximum up-ramp is either 532 - 575'F = 43*F or 554 - 590*F = 36*F. Thus, maximum up-ramp is 43*F. As discussed earlier, the outside diameter of vessel will not reach 590*F, therefore, the maximum down-ramp is 575 - 532*F = 43*F. Assuming a step change in fluid temperature and an infinite film coefficient, the radial gradient thermal stress cannot exceed: 3 ,_ E a AT, 27x10 (7.5x10-s)(43) = 12.4 ksi 0.7 0.7 or TH range = 2(12.4) = 24.8 ksi < 34 ksi a Therefore, thermal racheting is not possible. A third criteria, stated in ASME Section III, which allows primary plus secondary stress range of 3S, should be investigated. This limit ensures elastic shakedown in a few cycles, but it does not prohibit a small incremental growth. The primary and secondary stress range in r mi1181-0953a141

s 4G this area is 48.9 ksi, which is composed of a plus stress intensity of 18.1 ksi and a minus stress intensity of -30.8 ksi. Because both of these stresses a e below the yield strength of the material (42 ksi), there would be a negligible, if any, strain cycling. The maximum additional stress induced in the vessel during the extreme condition of contact with the bumper was 9.36 ksi. The stress allowance for 3S,, and the material yield strength are not exceeded under this condition. Therefore long-term distortion of the vessel is considered unlikely. mi1181-0953a141

+ 47 7.0 REACTOR PRESSURE VESSEL SURFACE PREPARATION Twelve local areas on the RV opposite to the ULS have been machined flat to improve the contact surface between the ULS and the RV. The machined area is 13.5 (i.125) inches wide and the top edge of the flat is located at el 631'6-1/2" (11/16") with its bottom edge is 8 (11/16) inches below the top. The flat areas are within 1/500 of vertical and 1/500 of perpendicular to the RV radius and have a surface finish of 250. The aforementioned dimensions and location guarantee a flat smooth surface, and a full area of contact between the 5 x 12-inch stainless steel pad at the end of the ULS and the RV in the event of the design basis LOCA. The amount of material to be machined off the RV was checked before it was removed and found to be within the acceptable wall thickness limits. B&W Construction Company designed and built the tools required for the machining, and the machining of both Midland Units 1 and 2 is now complete. l l l l mi1181-0953a141

48 i 8.0 DETENSIONING AND TENSIONING OF THE ANCHOR STUDS 8.1 DETENSIONING PROCEDURE In detensioning Unit 1, a scatter in the lift-off load values was observed (See Appendix C). A more rigorous and accurate detensioning procedure will be used for Unit 2 in measuring the lift-off load values. This will aid in explaining the scatter of lift-off values observed in Unit 1. The criteria and procedure to be used to detension the Unit 2 studs is described in detail Appendix A. 8.2 CREEP RECOVERY The reactor pressure vessel anchor studs in Unit 2 were tensioned to 92 ksi during the summer of 1979. When they are detensioned more than two years later, part of the compressive strain of the concrete will be recovered instantaneously. This will be followed by a time-dependent recovery known as creep-reco~very or delayed elasticity. This creep recovery reaches a limiting valve leaving an irrecoverable strain or permanent set. This creep recovery, if not accounted for, will increase the tension in the studs if they retensioned shortly after being detensioned. For this reason, retensioning will not. commence immediately after detensioning. Furthermore, depending on the time duration between detensioning and retensioning, the magnitude of the tension load will be adjusted to account for the increase due to creep recovery of the concrete. In addition, after tensioning and after sufficient time has elapsed, such I { that almost full creep recovery has taken place, the tension in the i mi1181-0953a141

49 studs will be checked using an ultrasonic extensometer device as described in Appendix A and adjusted if required. 8.3 RETENSIONING PROCEDURE The retensioning procedure for use on the Units 1 and 2 RV anchor studs is described in Appendix A. 4 t b I I e i i h mi1181-0953a141 1

50 9.0 REACTOR PRESSURE VESSEL INSULATION MODIFICATION Cutouts will be made in the reflective insulation to accommodate the penetration of the upper lateral restraints. These penetrations will be fitted with seals to reduce heat losses. l r o I mi1181-0953a141

4 51 10.0 GAP AND TEMPERATURE MEASUREMENTS AND GAP SETTING In order to benchmark the assumptions made in the heat transfer and thermal analyses discussed in Section 6 of this report, displacement and temperature measurements will be taken while the NSSS undergoes hot functional testing. 10.1 MEASUREMENTS DURING HOT FUNCTIONAL TESTING The following additional measurements will be taken during hot functional testing: The change in gap between the reactor vessel upper lateral a. restraints and the RV. b. The change in the RV surface temperature, the temperature of the upper lateral restraint and the concrete wall. 10.2 MEASUREMENT PROCEDURE The criteria and procedure is described in detail in Appendix B. 10.3 CORRELATION BETWEEN MEASURED AND CALCULATED VALUES Comparison between the measured and calculated temperatures and displacements will be made. If differences occur, the calculation will be modified to account for this difference. In this case, the calculated temperatures and displacements for operating conditions will be as accurate as possible. mi1181-0953a141

52 10.4 SETTING THE GAP In order to prevent contact between the RV and the ULS during normal operational conditions and in the case of a seismic event, the gap was calculated as follows: The SSE displacements of the RV and ULS in both horizontal directions are individually summed by reactor addition. The maximum seismic gap computed is 0.076 inch (0.043 inch wall displacement and 0.033 inch vessel displacement). The seismic gap was calculated using the conservative approach of adding (using the absolute sum) the displacements of the wall and the vessel thus assuming that they will move out of phase. The gap required to compensate for the thermal growth of the vessel and the wall as well as the effect of the pressure in the vessel and the closure effect is as shown for 0%, 15% and 100% power, in the table below. Thermal Displacements * (inches) Case Power Level __0% 15% 100% Thermal growth of vessel from 70*F .325 .359 .342 Pressure in vessel .048 .055 .048 Thermal growth of bracket from 70*F .012 .012 .012 Thermal growth of concrete wall from 70*F .046 .046 .046 Effect of Closure .008 .009 .008 Total .347 in .389 in .364 in

• Note:

A positive sign on the displacement indicates a gap closing motion, and similarly a negative sign indicates a gap opening motion. mil 181-0953a141

53 The following assumptions were made in the above table. The overall wall temperature during normal operational conditions is 1. 130*F. b. The growth of the ULS is based on a 0.1 inch gap between the ULS and the RV during normal operation. In establishing the gap, a construction tolerance of +1/64 and -0.0 inch was assumed. i From the above information, the gap required between the ULS and the RV at 70*F (approximately the construction temperature) is determined and is necessary to prevent contact between the RV and the ULS at 15% power (most critical condition, maximum thermal growth) during an SSE is calculated as follows: Displacement (in) Thermal growth, effect of pressure, and effect of closure 0.389 Seismic gap 0.076 Construction tolerance 0.016 TOTAL 0.481 (15/32" + 1/64" - 0") or 1 mil 181-0953a141

54 11.0 ANALYSIS TO DETERMINE FINAL SUPPORT LOADS 11.1 GENERATION OF SUPPORT LOADS 11.1.1 Technical Basis-The methodology uced to generate the design loads for the modified Nuclear Steam Supply System (NSSS) supports will utilize the same analytical techniques and computer codes as used in developing the B&W's Owners Group Report entitle'., Effects of Asymmetric LOCA Loadings, BAW 1621 B&W 177-FA, (Reference 8) which has been submitted to the NRC for review in July 1980. i Modifications will be made to the existing mathematical models of the NSSS and its supports to incorporate the upper lateral support spring rates, reactor vessel anchor stud spring rates, internal wall structures, and boundary conditions at the reactor coolant pumps and steam generators specific to the Midland Plant. The seismic forcing functions are Midland specific, however the LOCA forcing functions used to determine the support loadings are based on break areas equal to or larger than those specifically applicable to Midland. The analyses will incorporate state-of-the-art techniques (described herein) which insure that all components supporting, and attached to, the reactor vessel will receive a full review for structural integrity under the modified support design. mil 181-0953a141 ,a

-55 d 11.1.2 Mathematical Model A single mathematical model will serve as the basis for both seismic and LOCA' analyses. Minor modifications allow the model to be used for linear seismic or linear / nonlinear LOCA analyses. i For seismic analysis, the ULS will be gapped such that the RV and the ULS will not contact. The moment on the RV skirt is i such that pretension of the anchor studz is not exce.'ded.

Thus, linear elastic analysis for the support system will be applicable. Stresses are checked against allowables to insure the validity of this assumption. The STALUM computer. code is used to generate results.

For_LOCA analyses, the model will be modified to reflect the gap between the RV and the ULS, the inelastic properties of the ULS and the bilinear spring rate which reflects loads exceeding prestress on the anchor studs. The STALUM code, with linear I elastic properties, will be used to (;tablish " benchmark" LOCA results. The ANSYS code will be used to achieve results reflecting nonlinear and inelastic conditions of the support system. The results will' be compared with linear STALUM analyses to insure reasonability. I 11.1.2.1 NSSS Model 4 Because of the complexity of the RV loading conditions and the number of attachments to the vessel, a 4' detailed isolated model of this component will be e i I J mil 181-0953a141

56 constructed. This model will be a complete representation of the reactor vessel and its appendages (eg, control rod drive mechanisms, service support structure, and reactor internals). It will also include both the hot legs extending to the steam generators and the four cold legs extending to the coolant pumps. Boundary conditions will be imposed at the ends of the pipes where they connect to the components to simulate the remainder of the NSSS. The isolated model is shown in Figures 11.1 through 11.7. The isolated portion of the NSSS will be modeled utilizing finite beam-element and lumped mass representations of each component. Finite element methods are used where necessary to define the structural characteristics of components such as the fuel and plenum assemblies. Once determined by finite element techniques, the structural characteristics of components will be used to generate the equivalent 4 finite-beam element and lumped mass representations. The criteria for developing the equivalent structural representation is that component stiffness and frequency must be retained. The various components that make up the total RV and its internals are identified in Figure 11.8. By comparing Figure 11.8 with the lumped-mass model shown mi1181-0953a141 1

e 57 in Figure 11.1, the correlation between the components and the model elements representing them can be seen. In addition to the structural representation of the components, the NSSS mathematical model incorporates . the effects of fluid coupling between components into the overall structural response of the system. This is accomplished by developing a mass matrix using the height of concentric cylinders, the distance between the cylinders, and various parameters describing the fluid between the cylinders. The mass matrix which is generated is combined with the diagonal mass matrix terms defining component mass distribution to generate a full system mass matrix. 11.1.2.2 Internal Walls Structure The internal walls structural model properties included are the axial area, shear area, moments of inertia, modulus of elasticity, and Poisson's ratio for different elevations in the wall. Lumped masses and mass moments of inertia at different elevations define the mass distribution and mass resistance of the wall structure. The internal wall structure is modeled in the seismic analysis to the center of the 4 concrete basemat. The boundary conditions at that point are fixed such that no relative rotation or translation is allowed. This internal wall structure l mil 181-0953a141 l i i

58 s model is shown in Figure 11.4. For LOCA, the internal walls are modeled to include separately the primary and secondary shield walls along with springs at their base to represent the soil flexibility, this model is shown in Figure 11.7. 11.1.2.3 NSSS Supports For the isolated RV model, the NSSS supports are described as the boundary conditions imposed on the cold leg piping at the pumps and the hot leg piping at the steam generators, the reactor vessel skirt support, and the upper lateral supports near the RV flange. The boundary conditions imposed on the reactor coolant piping at the pumps and steam generators consist of stiffness matrices that represent the characteristics of the structures to which the pipes are attached. They are obtained from a full system model by disconnecting the pipes at the component nozzles and computing a stiffness matrix of the remaining component with its supporting structures and other attached piping. The RV skirt support is modeled in the seismic analysis as a boundary condition at the base of the RV skirt support in the form of a set of springs. The boundary conditions reflect the flexibility of the i mi1181-0953a141

S-59 anchor studs, localized concrete flexibility, and overall flexibility of the RV pedertal from the RV skirt support to the center of the basemat. In the seismic analysis, these stiffnesses are linear since the anchor stud prestress is not exceeded. The LOCA analysis reflects the nonlinearity of the RV base support during " liftoff" in a series of equivalent nonlinear springs connecting the base of the RV skirt to the concrete pedestal. The ULS are gapped such that they are not active during a seismic event. During a LOCA, the gap between the ULS and RV would close such that the ULS becomes an active support. ULS structural properties are incorporated into equivalent nonlinear springs which reflect the appropriate gap along with the inelastic properties of the support. Localized concrete and RV flexibility is included in series with the ULS springs. The ULS equivalent beams are shown in Figure 11.7 as they connect the RV with the primary shield wall. 11.1.2.4 Stiffness of Upper Lateral Restraints Lateral translation resistance versus displacement curves for the upper lateral restraints in three r directions are developed. No movement of the wall is considered in the development of these curves, hence mi1181-0953a141 ~ - _. - - ~

60 the stiffness of the brackets are added in series to the local wall stiffness. The three directions considered are the hot leg direction (North-South), the core flood nozzle direction (East-West) and the cold leg direction that lie midway between the North-South and East-West axis. The resistance curves are developed for a gap in the range of 0.090 to 0.125 inches and represent the stiffness of four ULS, and are given in Figures 11.11 through 11.13. The origin in the curves represents the first contact between the RV and the ULS. For the curves representing the stiffness in the direction of the cold leg, the stiffness of the first bracket to come in contact with the RV is neglected (the deflection, however, was considered). The neglected stiffness is comparatively small since the brackets are relatively flexible in bending about their minor axis. The local stiffness of the primary shield wall, which was determined by the finite element method of analysis is tabulated below: Break / Direction Spring Rate 5 cold leg 779 x 10 lb/in 6 hot leg (North-South) 215 x 10 lb/in 5 core flood nozzle (East-West) 723 x 10 lb/in mi1181-0953a141

61 11.1.2.5 Stiffness Of The Support At The Base Of The Reactor Pressure Vessel The moment versus rotation curve for the reactor pressure vessel base, (rotational spring constants KS g and KS ), is shown in Figure 11.14. The curve was developed by a finite element analysis that assumed the nominal prestressing load of 20 kips per stud (corresponding to a 5 ksi prestress). A dead weight of 27.9 kips per stud has also been factored into the analysis. The curve is bilinear and the flatter portion represents the stiffness after the studs have lifted off. For Unit 1, two slopes are given for the flat portion of the curve representing the upper and lower bound stiffness. The actual slope of the curve depends on the orientation of the moment with respect to the broken studs, and is between these two bounds. For Unit 2, where no studs are broken, the upper bound curve is used. The stiffness in the other four directions are tabulated below: Direction Spring Rate 10 torsional (K ) 1197 x 10 in-lb/ rad 0 lateral (K or K ) 578 x 10 in-lb/ rad x z 0 vertical (K, before lift off) 223 x 10 lb/in 6 vertical (K, after lift off) 171 x 10 lb/in mil 181-0953a141

62 In deriving the above stiffnesses, a finite element analysis was also used. The studs have no contribution to the lateral stiffness K or K and the z torsional stiffness K The lateral forces and torsional momeats are transmitted via the shear pins between the RV skirt flange and the sole plate, and from the sole plate through the shear lugs welded on the bottom surface of the sole plate to the concrete pedestal support. 11.1.3 Load Cases Analyzed The isolated model will be subjected to four' load cases in the process of determining tne design loads on the supports. Two sets of seismic analyses will be performed; one for the OBE and the other for SSE. Two LOCA cases for will be considered in detail; a guillotine at the hot leg outlet of the RV and a guillotine at the cold leg inlet to the RV. Other LOCA load cases will be assessed if they are shown to produce contact between the RV and the ULS. The support system is designed such that the ULS will receive no deadweight or thermal loads from the RV. Deadweight and thermal load on the RV base are analyzed using a larger loop model and the NSSS and supports. These results are currently being modified in a program unrelated to the RV support redesign. Preliminary results are given in Section 4.0. mi1181-0953a141 { ~---

63 11.1.4 Method Of Analysis 11.1.4.1 Seismic Forcing Functions The seismic forcing functions that will be applied to the mathematical model consist of response spectra curves for SSE at damping values from 1% to 5%. Response spectra is supplied for earthquakes in five directions, North-South, East-West,. vertical, rotation I about North-South and rotation about East-West. The rotation is applied as occurring about the geometric center of the RV at the elevation of the basemat. 11.1.4.2 LOCA Forcing Functions LOCA forcing functions are composed of three sets of time histories which are applied simultaneously to individual degrees of freedom. The forcing functions are the result of blowdown into the cavity between the RV and the primary shield wall, and pressure wave propagation inside the RV due to the break in the reactor coolant pressure boundary. Core Bounce i The vertical response of the rr. actor internals and Fuel Assemblies (FA) result in a time varying force composed of the structural response to differential pressures. Core bounce is the terminology given to i this response phenomena. The nonlinear structural mi1181-0953a141

64 response reflecting holddown springs and vertical gaps is calculated in a decoupled analysis. The FA core and reactor internals are simulated with a planar model consisting of beam elements, nonlinear axial springs, and lumped masses. The ANSYS code is used to calculate the vertical reactions of the core, which are then used as applied force time histnries on the reactor vessei in the ayctem dynamic analysn The core bounce LOCA forcing functions are the result of the worst case double end guillotine pipe breaks at the RV nozzle. Thermal Hydraulics y L Dynamic Response The pressure waves through the RV produce several reactions that are not considered in the core bounce forcing functions and which can be applied directly to a dynamic system. For the reactor vessel,.the horizontal pressure gradient results in horizontal forces on the RV, core support cylinder, thermal shield, and the plenum cylinder. The vertical gradient results in vertical forces on the RV. The integration of the pressure-time history defines the time history forces which are applied to discrete mass joints of the mathematical model. I mi1181-0953a141 l

65 The thermal hydraulic loadings applied directly to the linear dynamic model are the result of a hot leg pipe rupture and a cold leg rupture. Asymmetric Cavity Pressures Pipe ruptures which occur in the cavity between the RV and the wall result in differential pressures across the RV in a time varying manner. The differential pressures, when integrated across the area of the RV, produce time varying forces which are applied to discrete mass joints on the RV. The cavity pressure loadings on the RV for these analyses result from mass and energy data for double ended pipe guillotine ruptures equivalent to or larger in area than the actual pipe break. The same differential pressures applied to the RV are also applied to the primary shield wall. 11.1.4.3 Computer Codes Used For NSSS Analysis The three analytical computer programs and the four data reduction codes used in the seismic and/or LOCA analyses for the support design loads are described below. mi1181-0953a141

66 Structural Analysis Codes 1. HYDROE - A computer code used in calculating the hydrodynamic mass coupling of concentric cylinders. 2. STALUM - A computer program for analyzing three-fireosional, finite segment systems consisting of uniform or nonuniform bar/ piping segments, closed-loop arrangements, and supporting elements. STALUM performs both static and dynamic structural analyses undergoing small linear, elastic deformations. The static analysis is based on the matrix displacement method. The static loadings are static mechanical forces, thermal, and/or support displacement loadinga. The dynamic analysis is based on lumped-mass and normal-mode extraction techniques. The dynamic input loadings can be response spectra or time history forcing functions. The essential input to the program consists of the physical properties of the system, the boundary conditions, and/or the loading information; the essential output consists of the resultant joint displacements, rotations, forces, moments at both ends of each segment, and stresses at various locations in each segment. mi1181-0953a141

~ 67 3. ANSYS - The ANSYS general purpose program solves a wide variety of engineering problems more efficiently than most special purpose programs. ANSYS includes capabilities for transient heat transfer analyses including conduction, convection, and radiation; structura1 analyses including static clastic, plastic, and creep, dynamic, and dynamic plastic analyses, and large deflection and stability analyses; and one-dimensional fluid flow analyses. Data Reduction Codes 1. FTRAN - A computer code used for Fourier analysis of forcing functions to determine the frequency content of the forcing function. 2. S1235 - A post processor program used to tabulate forces, moments, displacements, and rotations in a specification format. l 3. INTFCE - A program used to convert pressure-loading data to force-loading data acceptable for use by the structural analysis codes. 4. LOPL - A post-processor program used to provide time history tabulations and plots of spring i i forces and resulting loads and displacements. t i l mil 181-0953a141 l t

68 11.1.5 Seismic Analysis Utilizing the geometric and structural properties of the mathematical model shown in Figures 11.1 thru 11.6; the STALUM code is used to determine the structural frequencies and mode shapes of the isolated NSSS, the internal walls structure and the NSSS supports as a coupled system. Each element or bar in the model is assigned a damping value based on the location and type of component the element represents. Strain energy damping is used to determine a composite damping for each mode. The modal accelerations are applied to the model dynamically to reflect the structural amplification. Equivalent static forces for each mode are determined and applied to each degree of freedom to give resulting modal displacements and member forces. The modal responses for each in'dividual earthquake will be combined by the SRSS method as described in the response to Regulatory Guide 1.92 in Section 3A of the FSAR. The resulting member loads and displacements will be combined by taking the SRSS of all five earthquake excitations (three translational and two rotational). Figure 11.10 shows the flow diagram for the seismic analysis. i j RV Lower Support Loads I l The seismic loads on the RV lower support are taken directly from the seismic analyses and are the forces and moments from the combined five earthquake components at the base of the RV i skirt. These centerline loads are resolved into support loads l mil 181-0953a141 l t l

e 69 for the stress evaluation described in Section 12.3.1. The final seismic loads and displacements are given below. REACTOR VESSEL SUPPORT LOADS SKIRT LOAD AT ANCHOR JOIh7 50 FIGURE 11.1 FORCES (KIPS) MOMENTS (FT-KIPS) LOAD CASE FX FY FZ MX MY MZ 4 SSE X Trans-Z ROT 276.4 4.4 62.7 1332.1 154.1 8105.7 SSE Y Tran 1.4 193.9 8.5 70.3 89.9 31.9 SSE Z Tran-x ROT 71.71 37.7 230.3 7157.3 145.3 1443.8 SSE (Combined) 285.6 197.6 ~ 238.8 7280.6 230.1 8233.3 ~ OBE (Combined) 145.8 98.8 119 3806.2 120.6 4334.1 . /.. DISPLACEMENTS RV PROFILE AT ULS JO NT 166" FIGURES 11.1, 11.4' ~ DISPLACEMENTS '(INCHES) X Y ~ Z ~ SSE (Combined) ~ w .02445 ,.00148 .02156 OBE (Combin'eb) ~ -- . 01284 ,00074 ' .01124 l a WALL PROFILE-AT ULS EL. 631'5-1/2" JOINT 170 FIGURE 11.4 i DISPLACEMENTS (INCHES) e X Y / Z~ SSE (Combined) .03261 .00144 .02854 i OBE (Combined) .01891 '.00072' .01711 ULS Loads There is no interaction between the U'LS and the RV during a seismic event. .a mi1181-0953al41 I .A. / l I f' m-a - - :.4.--- -. ~ - - = - - - - 'S- ~

70 11.1.6 LOCA Analysis The geometric and structural properties along with the nonlinear properties of the ULS and the reactor skirt support are included in the model utilizing the ANSYS computer code. The three sets of LOCA forcing functions are applied simultaneously to individual DOF's to represent the structural loadings to the components during the LOCA event. Displacement and member force responses are determined for each node or joint and element. The resulting displacements and member forces and moments are stored such that time-for-time or peak results are available for any member or joint. RV Lower Support loads The peak forces and moments, regardless of their time of occurrance, will be obtained from the time history LOCA analysis output, and used as the total centerline load imposed by the RV on the support. ULS Loads The total peak horizontal force on the equivalent springs representing the ULS will be given as the maximum load on the support and the primary shield wall. The peak displacement of the total ULS system will also be available as needed. mi1181-0953a141 -=

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8 76 = I I q f l.fi f'.'5,l,la \\ N - i f ~*% .,Yl f I i _fu I .I 64 j 'y' ~ ~ ~ ~ ~ ig g Hst-- g l m e N 3 l~ 9 s E9I 5 ll? / $\\ 1o 4- \\ Z' N 1 9 n < lit & g -,j "m" Yi TW'~ I ~T1i,i ^ l lI I 5 eA I I I ~ >= I ? 6j + 1 l I l g" k og i E w I I v 5e ,I t I 5 g Q$- '1 b-4 ) p 6ff'D lt 4 { / / /. s? V s1 s g s;. l ldf "1 ,j;:;I )6 143-w '~~7 7 2 '4, f l (D fI

• 005

= I-- L / N dt N k ~g l-I I c

I pt /G 22t/ Mle l 9 l I 3 .a arvssoes Y SecMOMY WALT. 4& EL. 66 5*.O* b .get. oso:o-PRIM A R Y p 1 WALT @ et. 64o-o X a \\\\ 4 si. asz'a-Sl* 'Y* (gy ByApp*gf) / st. 6Ss"- /

• l I

-EiB / EB- -M) \\ st.s u -to" GBH> 4 ut. sis'.o-l p Q +z 5 c 5-KEY El. 4/S - 7y r s EiB-o strucrueAL.kuar \\ !u _.g3 y / o asss.miar _ _,./ Q. lowr NuMace EL.fo05 5 yp REACTOE VESSEL C3 uwen " ween at s.,. a-S 4 et. soi-o' SUMPER y ELEVATOAl f' a,merao a,an, se a.,,- c -EfB fl. O

• O y

-EEE EiD- - g et. set.o* WM., RV ISOLATED r.m&..ms; r. d [ [*;** '[-i MODEL VERTICAL %ll ..~~j,.c.d +3=,.

• ~d"fE@

(REACTOR NESSEL SFs53fs.fi FIGURE 11.7 suueeg ettvuiora J.- - - - ~ i r,co.ocvz/ss (LOCM, II??%17 l' n ,._..m...-

T8 i i CONTROL R00 [ DRIVE SERVICE [ STRUCTURE It 8f I ca ^ / \\ }M CONTROL R0D GUIDE TUBE e (COLUMN WELOMENT) PLENUM p COVER d Q '! - PLENUM _-) ASSEMBLY / s 3 s ,'l / UPPER / [, A CORE SUPPORT GRID d CYLINDER FUEL REACTOR s CORE l-LOWER THERMAL GRID l j SH.IEL.D 4 s S A GUIDE LUG LOWER GRID L ) SUPPORT FORGING _ - il i fj FLOW DISTRIBUTOR N SKIRT SUPPORT i FIGURE 11.8 Reactor Internals and Service Support Structure

79 [K] [K] sq i ( K] [g3 w s A M \\ [K] ~ [K] seismic only (K] = Stiffness matrix l FIGURE 11 9 Reactor Coolant System Boundaries l

80 i SEISMIC / LINEAR LOCA DEVELOPMENT OF HYOROMASS HYDROE NASS NATRIX HONLINEAR LOCA STALUM DEVELOPMENT OF GE0M SilFFNESS & ANSYS NODULE FLEXIBILITY NATRICES STALUM FREQUENCIES & LUMP MODE SHAPES NODULE STALUM LUMP EQUIVALENT STATIC FORCES NODULE STALUM RESULTANT LOADS RSTA (CETERMINED MODULE STATICALLY) FIGU3E 11.10 l Utilization of Computer Programs

81 BRACKET RESISTANCE VS DISPLACEMENT CURVE HOT LEG (North-South) DIRECTION AT GAP BETWEEN 0.090" TO 0.125" 6,360* ----____c,,,___________ 6,000 LOCATION VARIES l MAX ALLOWABLE 5,000 A WITH GAP f DEFLECh0N FOR K l DUCTILITY RATIO K ^ = 10. j 4,000 a b l 6 3,618* l 3,000* l l I i s 'E 2,000" l l l l l. l 1,000* l l l I o. 0.00001 0.10000 0.20000 0.2662 0.02662 0.23134 Displacement (in) I 6 = GAP f 1 1 COS 43 COS 17 l FIGURE 11.11 i l l t

82 BRACKET RESISTANCE VS dlSPLACEMENT CURVE DECAY HEAT REMOVAL - CORE FLOOD (East West) DIRECTION 3 g AT GAP BE~ WEEN 0.090" TO 0.125" 6,344" K 6,000 LOCATION VARIES MAX ALLOWABLE 5,000* O l DEFLECTION FOR A WITH GAP U DUCTluTY RATIO K 6 4,000 8 b l = 10. 3,772* g g l j 3,000* l l l l 2,000 K I I l i 1,000* l l l .I I I I o 0.01000 0.10000 0.20000 0.25480 0.02548 0.21756 Displacement (in) 6 = GAP I I I y COS 47 COS 18.5 FIGURE 11.11 l l

BRACKET RESISTANCE VS DISPLACEMENT CURVE 83 UPPER COLD LEG AT GAP BETWEEN 0.090" TO 0.125" MAX ALLOWABLE DEFLECTION FOR DUCTIUTY RATIO g (Pd. Ad) d = 10. 6,019 I g 5,000 l K (Pc,Ac) l 6 c l' l K e 4,000 8 l 1 a k .g 3,000 l l Jpf g K e i I l 2,000 l l K l l l 1.000* l l l - _ },A b) ( / O.1dOOO 0.20000 AMAX 0.01000 0.02000 g, 31 o2 63 0.03000 / AT POINT A a " W (COS 43 ~ COS 11.5 l AT POINT B 1 = GAP ( OS 8,5 )AP,= 0 ~ CO 0.04734 I I A,=A,+Ag P, = A P, AT POINT C 2"1375 (1886-A P,) AP P, = A P2+AP, AP2 A2= X 0.01893 l l A,=A,+A 2 AT POINT D A3 = 0.02720( 0 9 ) 5 d = 6019" A, = A, M P 3 A MAX = A, + (A, + A ) X 10 2 RANGE FROM 0.34862" TO 0.3833" l FIGURE 11.13

8h Rotational Spring Constants sbpe = 121.0 X 10'Oin-Ib/ rad KO, & KO, at the Base of the RPV 7 M X 10 (in -Ib) 35 - O'j io slope = 102 X 10 in-lbn _d p (b) / NOTE: These spring constant curves are / applicable for Unit 1.* * / 25 - ge) They are based on 5 ksi 7 final stud prestress level and a total number of 93 anchor studs. j

• M

= 20.16 Dead Stud LO Weight Prestress C' Equivaient preload per stud: 27.9k + 20k = 47.9k 15 - If the final stud prestress level varies from 5 ksi, simply change (a) l M linearly in proportion to the total proletd to determine the Lg l new lift-off moment and Point 0*. Draw lines parallel to Points (b) ar"1 l (c) from Point O' to complete the new spring constant curves. (a) Before anchor studs lift off 5" l (b) After anchor studs lift off (upper bound) slope = 574 X 10'Oin-lb! rad

X 10-'(rad) -

0.351 1.0 2.0

• Moment when studs lift off
• *The upper bound curve can be used for Unit 2.

FIGURE 11.14

85 12.0 CHECKING SYSTEMS AND SUPPORTS FOR THE RESULTS FROM FINAL ANALYSIS All attached systems, components, and component supports will be evaluated for the results of the aforementioned analyses. The current forecast date for completion is anticipated in the spring of 1983. L f l l i 5 i i r I mi1181-0953a141 .. - - -. -... -. - ~ -..

86 13.0 Construction Status and Schedule The stiffening of the shield plug brackets to form the ULS, and the machining of the flat surfaces on the reactor pressure vessel in both units are complete. Unit I studs were detensioned to a nominal stress of 6 kai and tha lift-off forces measured during detensioning are given in Appendix C. Detensioning, measuring lift-off loads, and retensioning the studs in Unit 2, as described in Section 8 of this report, is scheduled for completion in May 1982. This will be followed by detensioning and retensioning the studs in Unit I to their final prestress level, and this is scheduled for completion on October 1982. 1 The insulation will be modified and installed as described in Section 9 of this report after the cold hydro and before the hot functional testing. I mil 181-0953a141

87 14.0 Conclusion This report has described the analysis and design of the modified reactor vessel support system for the Midland Nuclear Power Station. Particular attention has been devoted to the physical modification required for the upper lateral support, computer modeling and analytical techniques being used. The methods presented herein represent the standard techniques utilized by the NSSS suppliers for primary system analyses and by A/E's in designing Category I structures. The design modification is mandatory for Unit 1 because of the anchor stud failures experienced. Based on the investigations conducted, the Company has decided.to modify the Unit 2 reactor support design to be identical with that of Unit 1. Thus the analyses of the NSSS for both units will be covered by a single analysis. This report provides updated information regarding the design and analytical techniques stemming from engineering evolution in the course of this project. The design of the upper lateral supports has proceeded using preliminary design loads as described in this report. The supports are designed with respect to these preliminary loads using conservative assumptions. The confirmation of the adequacy of the design will be made upon receipt of the final support loads, and the project schedule indicates that this will occur around November of 1982. In the event that further evolutions occur in either the design or analyses described in this report, the Company will submit them as mi1181-0953a141

a 88 amendments of this report to the NRC. The final analytical results and design details will be incorporated, as necessary, by amendment into the i FSAR. i i 1 i 4 .1 3 l f l. l k t t i l l l ( mi1181-0953a141 t t i-._...,.-.,.._._.......__._____.______._...._,....___...

89 15. References 1. Teledyne Engineering Services Report, TR-3887-1, Rev 1, Investigation of Preservice Failure of Midland RV Anchor Studs, May 15, 1980. 2. Teledyne Engineering Services Report, TR-3887-2, Rev 1, Acceptability for Service of Midland RV Anchor Studs, May 20, 1980. 3. Teledyne Engineering Services Report, TR-3887-1, Addendum 1, Investigation of Preservice Failure of Midland RV Anchor Studs, June 6, 1980. 4. Teledyne Engineering Service Report, TR-4599-1, Continued Investigation of the Failure of Midland Unit 1 RV Anchor Studs - Data Report, February 11, 1981. 5. Teledyne Engineering Services Report, TR-4599-2, Continued Investigation of the Failure of Midland Unit 1 RV Anchor Studs - Analysis Report, February 11, 1981. 6. Reactor Pressure Vessel Support Modification for Midland Nuclear Power Plant, Report No 1, July 1980. 7. Reactor Pressure Vessel Support Modification for Midland Nuclear Power Plant, Report No 2, December 1980. 8. BAW 1621 B&W 177-FA Owners Group, " Effects of Asymmetric LOCA Loadings", Phase II Analysis, July 1980. mil 181-0953a141

90 9. " Letter Report - Teledyne Engineering Services (TES) Project 5355: Expanded Criteria for Acceptability for Service of Midland Unit 1 RV Anchor Stress" W.E. Cooper letter to H.W. Slager, dated October 6, 1981. 10. _Tbernal Stress Techniques, The Franklin Institute Research Laboratories, American Elsevier Publishing Company Inc., 1965. mi1181-0953a141

APPENDIX A: Procedure for Detensioning and Retensioning the Reactor Building Reactor Pressure Vessel Anchor Studs f a t i 4 i l I l l I t 1 i l i i [ mil 181-0953a141 i

v .o APPENDIX A ~~~ FROCEDURE MR DETENSIONING AND RETENSIONING REACTOR BUILDING REACTOR PRESSURE VESSEL ANCHOR STUDS 1.0 SCOPE This accendix . provides a procedure for detensioning and retensioning the reactor pressure vessel (RPV) anchor str.Js in Midland Plant Units 1 and 2. A procedure for verifying the preload in the anchor studs is also included. Data collected during the detensioning of Unit 7. will be used for analyzing Unit 1 anchors which have already been detensioned. 1.1 The Procedure includes the following: a. To detension and retension 96 anchor studs for the Unit 2 RPV and to detension and retension the 93 remaining anchor studs for the Unit 1 RPV. b. Working with a bolting technology consultant who will supervise the use of an ultrasonic extensometer to monitor deformation in the anchor studs 2.0 QUALITT STANDARDS The work shall be performed in accordance with requirements of a quality assurance program approved by Midland Project Quality Assurance Department (MPQAD). 3.0 Intentionally Left Blank k.0 Intentionally Left Blank 6 A-1 1

5.0 REFERENCE DRAWINGS The required detensioning frame is shown in Appendix 1, Pages 1-2 through 1-5. The description and details of the required retensioning frame are given in Appendix 1, Page 1-6. 6.0 MATERIALS.AND EQUIPMENT, 6.1 BOLTING MATERIM.S Replacement of nuts and washers, if required, shall conform to the purchase specifications. l I A-2

6.2 LCADING FRAMES Loading frame material shall be as noted in Appendix 1., Page 1-6. 6.3 EYDRAULIC RAMS 6.3.1 Rams for Detensioning The two hydraulic rama to be used fo'r ~~ the detensioning frame shall be a solid plunger type with 2-inch minimum stroke and a capacity of 100 tens each, such as Model RC-100-E-5.7 by Duff Norton. The base diameter shall be a maximum of 7 inches and a maximum closed height o' 8 inches. The rams shall be a matched pair calibrated and certified, traceable to the National Bureau of Standards (see Section 8.1 for calibration procedure). 6.3.2 Rams for Retensioning The hydraulic ran for the retensioning frame shall be a hollow-core type with 2-inch mi nimum stroke and a capacity of 20 tons such as Model RCE 202 by Enerpac (see Section 8.2 for calibration procedure). The minimum internal diameter of the core shall be 1-1/16 inch. The maximum base diameter shall be 6 inches and the maximum closed height shall be 8 inches. ~ 6.4 HYDRAULIC SYSTEM ACCESSCRIES 6.4.1 Pressure gages shall be test system gages, 8 to 10 inches in diameter, measuring 0 to 10,000 psi and shall be graduated in 50 psi maximum increments with 25 psi increments preferred The gage shall be accurate to +0.5% of actual pressure in the 2,000 to 10,000 psi range. The calibration shall be traceable to the National Bureau of Standards. 6.4.2 Hoses, fittings, valves, and pumps shall be compatible with the rams and gages specified. They shall be in good condition as determined by construction engineering and shall have no leaks or rapid losses of pressure when the equipment is assembled. The main lock-A-3

9 off valve shall be a manual,.?-way type for positive load holding, which, when closed, will prevent cylinder movement. The valves used for throttling shall be manual shutoff valves of fine needle, two-way directional type capable of being used for throttling. 6.5 ULTRASONIC EXTENSOMETER The extensometer shall be a Raymond Engineering Inc., Power-Dyne Division, ultrasonic extensometer, with an acceptable transducer as determined by Raymond Engineering Inc., from test results for 2-1/2-inch diame.ter ASTM A 354 bolts 7 feet, 4 inches long. '6 DISPLACEMENT GAGES Displacement gages (length and level) shall have graduations of 0.0001 inch and shall have a minimum extension of 0.5 inch. (1 inch is recommended.) The gages shall be calibrated to a standard traceable to the National Bureau of Standards. 7.0 SEQUENCE OF WORK Work shall be performed in the following sequence: a. Calibrate equipment b. Take initial, as stressed, extensometer readings on Unit 2 anchor studs c. Measure lift-off on Unit 2 anchor studs [ and proof test if necessary. i d. Detension Unit 2 anchor studs Take unstressed extensometer readings e. on Unit 2 anchor studs i f. Check calibration of detensioning equipment g. Retension Unit 2 anchor studs and take extensometer readings l l h. Check calibration of retensioning equipstent I. Detension Unit 1 anchor studs l A-4 l L

j. Recheck calibration of retensioning equipment k. Intentionally left blank ~. 1. Check calibration of detensioning equipment m. Retension Unit 1 anchor studs and take extens meter readings Recheck calibration of retensioning equipment n. 8.0 CALIBRATION 8.1 RAMS AND PRESSURE GAGES Rams and pressure gages are to be calibrated as described below. Calibration or recalibration shall be done as shown in the sequence of work or at maximum 30-day intervals. Calibrate the pressure gages in the a. pressure range of 500 to 10,000 psi. _ Ensure that the pressure indicated by the gage is within +0.5% of the true pressure in the 2,000 to lh 000 psi pressure range b. Mark each ran and pressure gage set so they are easily identifiable as a set. These sets must be calibrated and used in the field as a set. Pressure gages shall not be switched between sets. If a gage requires repair or replacement, the set must be recalibrated unless a gage with a calibration curve matching the first is available. (Notify project engineering before proceeding l with recalibration. ) l c. Before calibration, exercise rams three strokes O to 0.9 of full extension at 30%, 50%, and 80% of ram capacity. i d. As a minimum, ram pressure calibration i data points shall be taken at the following increments: I t l l A-5 -- - _.~._____ _.__._._- _ _. _ _ _ _ _ _. _. _ _ _ _ _ _ _ -. _ _ _ _ _ _ _ _ _ _ _ _ _ _.,.. _ _

4 h 7 m i- ~ Pressure' Range, Pressure Increment (nsi) ' (esi) [ O to 3;000 500 3,000 to 10,000 l',000 or maximum -capacity 'f the ~ o - ram ~ j Three ' sets of load versus piossure readings _shall be taken for~each ram pressure gage combinatien. If the' measured 11oads at-a'particular pressthe > / level in the three setsrof data deviate' by more than 2,1%, additional-sets of ~ data shall be taken_until consistency ' is attained. ~ e. Rams are to be calibrated at un extension of 1 inch and must be. ~ calibrated in the active mode with the ran actuated by the, pump and forcing load on the test mar-hine f. All calibration measurementi are,.to b'e traceable to the-National Bureau of Standards. If a testing machine is, used for calibrations, its calibration shall have been certified within the last year. A copy-of the testing / machine's certificatiori of calibration J_ and any other referenc'e standards used in the calibration shall be submitted to project engineering and MPQAD with I the ram amd. pressure gage calibration s data. l g. Recalibration shall be performed in a ' sf - t ~' t ( similar mannel as--described in Sections 8.1, Items c,'d e, and f..#if" the ram pressure gage com,b'ination recalibrati6ii readings are-found to' deviate more tbsn +1% froe the measured-- ~, load, the project engineer shall be / notified, imreediately. 7 8.2 RETENSIONING EQUIPMSiT ~ The retensioning ram and hydraulic pressure gage assembly shall'be calibrated with a universal testing machine or other calibrated standard. The system shall be calibrated before tensioning l A-6 N i ~

4; ~. Unit 2 anchor bolts and again before tensioning Unit 1 anchor bolts. The system d' calibration shall also be checked after completing Unit 1 tensioning. The calibration '4 can be rechecked using the frame shown in ~ Appen,Hv r, Page'l-9. ? ),. 8.3 CALIBRATION OF EXTENSOMETER l -. ~." The procedure for calibration and use of the extensometer shall be provided by Raymond Bolting Services and submitted for review to project engineering and MPQAD. 8.4 RECORDS All calibration procedures and records shall ~ be prepared and submitted to project 1 engineering and MPQAD for review. Records shall be maintained indicating all pressure readings against s*=nd=rd pressure gages, load ~ readings against standard loads, and -f ~ extensometer readings against loads applied by a calibrated s*=pd=rd. 9.0 UNIT 2 DETENSIONING n 9.1 PREPARATION Prior to taking any readings or setting up detensioning equipment, all threads and stud ends shall be cleaned to facilitate removal of the nuts. The studs ends shall be inspected by Raymond Bolting Services for conditions which could affect extensometer readings. Methods of cleaning, acceptability of ~ f cleaning, and methods of repair shall be determined by field engineering. If there are burrs on the RPV skirt between anchor bolts in "~ the bearing area of the detensioning frame, r! they shall be removed by procedures acceptable to the RPV manufacturer. 9.2 INITIAL EXTENSCMETER READINGS ~ When preparations have been completed, two ~ ~ complete sets of extensometer readings shall 4,_ be taken on the studs in their present. , tensioned state. These readings shall be i f taken according to the extensometer manufacturer's instructions. One complete set ~ , h. of readings shall be taken and recorded; then a second set shall be taken and recorded. If the two lengths are not identical (acceptable A-7 s psw . /

s.,;

a f w .g.. - #,-, -,-*---m ~*-e-' ' ' - - - - * " ' - * " ~ - - - ' ' " ' ' ~ " " ' ' ~ ~ - ~ ' " ' '

2" .g 1 y, y _V ,j' y. 'iolerance to be determined by Raymond Bolting t Services), then the readings shall be repeated I until agreement is reached. These readings shall be recorded against the Teledyne stud numbering system. (See Appendix 2 for the ~ o numbering system.) 9.3 LIFT-OFF READINGS when the initial extensometer readings are

• A ".

complete, the existing preload forces in the 'i ' ~ ~ ' Unit 2 anchor studs can be measured as e follows: f The detensioning frame shall ~ be' set up as shown in Appendir 1, Pages.1-1 through 1-4, starting with stud 37 (Teledyne M numbering system). The cross beam and ram u, .n support blocks shall be installed first (see schematic in Appendix 1., Page 1-7). When this is done, the stud coupler with the transducer and cable inserted can be installed. The ~ ~ transducer shall be attached to the stud end according to the instrument manufacturer's a instructions. The stud coupler shall then be placed over the transducer and connected to the stud. During this operation, care shall '*L be taken so the transducer is not dislodged or the cable from the transducer is not damaged. s, 's When the cross beam and stud coupler are installed, the hydraulic ram support blocks t can be placed on the RPV flange. The blocks shall be level and, if necessary, shall be I modified to avoid overlapping washers or the fillet welds on the RPV skirt base. When the support blocks are level, the rams can be installed. The rams shall be vertical and g placed directly under the cross beam ~ centerline. The rams shall be installed so the hydraulic hoses are free of sharp kinks and do not rub against sharp corners. The N m pressure gages shall be positioned to allow easy reading. The rams shall then be jacked to level the cross beam and checked by using a ' ~ mason's level. A minimum 1-inch extension of the rams is required during the leveling 1 process. When this procedure is complete, the G upper nut on the stud coupler shall be brought to a fingertight condition against the crossarm. I S The displacement gages shall then be installed ' p. as shown in Appendix 1., (Page 1-7 ). The gage v support shall be firmly attached to the RPV by A-8 is t '~'3,, b 4 i, .. -. _. _,, _.,,... -. _. ~,,

a magnetic attachment so the gages are easily readable but cannot be dislodged during the testing procedure. The lift-off procedure shall begin by recording the initial readings of all displacement gages, pressure gages, and the extensometer. Lift-off shall be datermined when 0.002-inch feeler gages can be removed, from between the stud washer and the nut. ~ These feeler gages will be placed in position after passing lift-off on the first stud loading. To accomplish this, the rams shall be pressurized in 100 psi increments until two feeler gages can be easily installed approximately 1/2 to 1 inch under the nut on opposite sides of the stud. The feeler gages shall be within 1/4 inch of the stud and extend under the nut a minimum of 2 inches past the centerline of the stud. During stud loading, care must be taken to keep the crossarm level by keeping the changes in the level gage readings equal. Adjust the ram pressure to level the crossarm if necessary. If the ratio of ram pressures is greater than 1.05 or less than 0.95, depressurize the rams, check the alignment, and reset the rams, if necessary. The rams can then be repressurized. When the feeler gages have been installed, the pressures shall be reduced by a minimum of 500 psi below lift-off to the nearest 500 psi or 1,000 psi reading below apparent lift off. Length gage pressure gages 3 and the extensomet.ar shall then be read and recorded. During the next portion of the test, a plot of the length gage readings versus pressure shall be made as the test progresses. The pressures shall then be increased in 100 psi increments with the readings recorded and plotted at 200 psi intervals. During this time, the feeler gages shall be gently tugged. When the feeler gages pull out from under the nut, the readings of all instruments shall be recorded as corresponding to lift-off. The test shall carry on far enough (another 300 psi minimum) to show a break in the curve pressure gage reading versus length gage reading, to indicate lift-off. The rams shall then be returned to zero pressure. The nut shall not be turned at this time. If any stud is loaded to 360 kips before lift-off occurs, the load shall be reduced to less A-9 {

than 200 kips and project engineering shall be informed. Alternatively, the load on the ram can be reduced to zero and the setup moved to the next stud while awaiting the project engineer's instructions. It is anticipated that lift-off will occur at approximately 320 kips, although Unit 1 lift-off occurred at levels as low as 216 kips. j Any stud for which lift-off occurs below 300 kips must be proof-loaded to 300 kips or two-times the =awimum anticipated stress, but no higher than 344 kips. The proof test value will be given by project engineering before the start of detensioning. This can be done after lift-off is measured. Length displacement gage and extensometer readings shall be recorded at the proof-loading. When studs with centerpoints for machining are encountered, this shall be recorded on the data sheets. This procedure shall be performed on all studs in the sequence shown in Appendix 2, Pages 2-1 through 2-3, before detensioning. The lift-off readings on the first six studs shall be forwarded to project engineering within one working day of recording. Confirmation will then be made that the applied loads as measured by the pressure gages and, as determined from the extensometer readings are within acceptable tolerances. 9.4 DETENSIONING ~ When lift-off readings and proof-tests have i been completed as described in Section 9.3, detensioning may begin. The detensioning frame is to be used as described in Section 9.3, including the nut socket ring. The studs shall be detensioned in the sequence shown in Appendix 2. The detensioning frame shall be installed as previously described, except the displacement gages are not required. When using the detensioning frame, care shall be taken to keep the crossarm level within tolerance and alignment. The studs can be loaded gradually to the previously recorded lift-off pressure. Pressures shall not be allowed to increase more than 100 psi over the previously recorded lift-off pressure. The socket ring is provided to turn the nut when lift-off pressure is reached. The nut shall A-10

then be retracted approximately 1/4 inch and load releasing can begin. The load shall be released in increments determined by the elongation measurement capacity of the extensometer. The nut shall be returned to a snugtight condition at each step of the detensioning and the extensometer dial gage and pressure gage readings shall be taken after the load has been released. A complete set of extensometer readings is to be recorded and retained for each detensioning step. This procedure shall be repeated until all Unit 2 studs are detensionad. 9.5 EX'fTNSOMETER READINGS ) When the Unit 2 studs are detensioned, i complete set of extensometer readings shcIl be taken as follows: Verify that all nuts are 1cose by a. inserting a feeler gage between each nut and washer b. Following the manufacturer's instructions, attach the transducer and obtain an extensometer reading for each stud c. Record this value After this procedure has been completed for all studs, it shall be repeated a second time to verify the readings. 9.6 CHECK CALIBRATION OF DETENSIONING EQUIPMENT Upon completion of detensioning the Unit 2 anchor studs, the calibration of the rams and pressure gages shall be checked as described in Section 8.1. Records shall be maintained as described in Section 8.4. l 10.0 RETENSIONING UNIT 2 ANCHOR STUDS i i 10.1 PREPARATION The anticipated deflection in the studs based upon the stress valge of 5 ksi in the stud tensile stress area of k in and which is equivalent to a load of 20 kips is given by: A-11

= P, Lj + L2 E ~ (^1 ^:j where Lj = 68 inches (the stressed length of the unthreaded body of the bolt) L2 = 13.25 inches (the stressed length of the threaded bolt) P = 5.0 kai x 4.00 square inches = 20 kips A3 = 4.9 square inches A2 = 4.0 square inches E = 29 x 10 kai This deflection equals 0.0119 inch for the s+=ad=rd stud. The anticipated readings to be observed during the retensioning are as follows: Displacement from UT Length Length Gage Reading on RV Reading at Stud End After Load Stress coupler Top

• Corrections (kips)

(ksi)* (in.) (in.) 0 0.40 0.0000 88.0000 4 1.00 0.0053 88.0025 IT 4.25 0.022h 88.0104 18 h.50 0.0230 88.0111 20 5.00 0.0264 88.0123

• The calculated.:tretch in the coupling stud and coupler are included.

The required gage pressure for the retensioner to develop the specified load of 20 kips shall also be determined from the ram and pressure gege cali-bration data. A-12 l

10.2 RETENSIONING The studs shall be retensioned in the same order of that shown in Appendix 2, Pages 2-1 through 2-h. The retensioning frame shall be installed as shown in Appendix 1, Page 1-8. First, the extensometer transducer shall be installed according to thi manufacturer's instructions. The frame, pull rod and coupling shall then be placed over the stud. The coupling and pull rod shall then be connected, taking care not to dislodge the transducer or to damage the cable from the transducer to the readout unit. After connecting the pull rod, caution must be - exercised to ensure that the hollow-core hydraulic cylinder is centered on the pull rod and the frame and that all bearing surfaces are perpendicular to the pull rod. Care must ~ also be taken to ensure that the frame does not rest on adiacent washers or on the RPV skirt fillets. When alignment is acceptable, the top nut of the pull rod shall be brought to a snugtight condition. When this is done, the ram extension shall be approximately 1 inch. Upon completion of the setup, hoses shall be checked to ensure that pressure gages are visible and no kinks exist. The displacement gage shall be installed securely to the RPV and as shown in Appendix 1, Page 1-8. Retensioning may then begin as follows: a. Record the readings on the displacement gage and the extensometer b. Gradually pressurize the system until it reaches the pressure equivalent to the specified loading of 20 kips per stud on Unit 1. For Unit 2, the retensioning shall commence a minimum of 15 days after the detensioning of the last anchor stud. If the Unit 2 studs are to be retensioned 15 to 25 days after detensioning, the speci-fied loading shall be 17 kips. If the Unit 2 studs are to be retensioned 26 to k0 days after detensioning, the specified loading shall be 18 kips. If more than 45 days, the specified loading shall be 20 kips. A-13

c. Record the displacement gage and extensometer readings at that time d. Bring the nut to a snugtight condition and release the load e. Record the displacement gage and extensometer reading again f. To compensate for relaxation, subtract the extensometer reading taken in Item e (above) from that taken in Item c (above). Add this difference to the readings taken in Item c and reload this stud until the displacement gage and extensometer reaches the total value. Bring the nut to a snugtight condition and release the load. g. Recheck the displacement gage and extensometers to ensure that they are within 25% of the readings taken in Item e above which correspond to the specified load-ing. Record the displacement gage and extensoceter reading. If the values are not acceptable, repeat the retensioning of the p rticular stud. A-lh

1 1 h. Repeat this procedure on all studs When retensioning of all bolts has been completed once, the load level shall be checked and adjusted in the following manner. For retensioning, the procedure shall follow in the same order of that given in AppenM r 2, pages 2-1 through 2-4. I First, a complete set of extensometer readihgs ~ mhall be taken. The reference length of the respective stud obtained in 9.5c shall be dialed in to the instrument, and the existing stretch read and recorded. The existing load shall then be obtained as follows: AE P = 20 kips ,X g 3 (AEC) where P = ther calculated actual load existing in g the stud AE = the measured stud extension (difference A between remeasured length and initial length at zero load) aE the measured stud extension obtained with C = the 20 kip load in 10.2c l If Pg, equals 20 kips 110%, the load in the stud is acceptable. If P is less than 18 kips or A greater than 22 kips, PA shall be adjusted as follows: The values of PA shall be submitted to Project Engineer-ing. Project Engineering vill calculate the UT extenso-meter deflection required for PA to reach the specified load. The corresponding adjusted displacement gage reading including the stretch in the coupling stud vill also be given. These data vill then be used in the following procedure. The retensioning frame shall be set up as described in Section 10.2. Initial readings in the extensometer and displacement gages shall be recorded. By dialing in the stud reference length, the initial extensometer extension reading should equal AE. Next, a plot of pressure versus' length displacement shall be made. The ram shall be gradually pressurized. Extensometer readings shall be recorded, and length gage readings shall be recorded and plotted at 500 psi increments A-15

until the ram pressure is within 500 psi of the pressure required at P. (the existing load in the bolt). Subsequently, readings shall be taken and plotted at 100 psi intervals. The tensioning shall continue until the extensometer stretch is that required for P = 20 kips. The plotted curve should then show a change in slepe at P, and the change g in length gage reading from P to P = 20 kips ~ g should equal that calculated by project engineering. If this is not the case, notify project engineering. If the readings are correct, complete the tension adjustment by following the procedure outlined in Section 10.2, Items d, e, f, and g. This procedure shall be repeated until all measured elongations (loads) are within 110% of that obtained at the specified load. This is considered to be equivalent to a stress of 5 kai 10.5 kai. When the actual loads have been found acceptable, the jam nuts shall be placed on the studs and tightened. shortly before installa-tion of the insulation em the Unit 2 R*V and a minime or 30 days after all of the studs have been found acceptable and the jam nuts installed on additional - set of two extenso=eter readings on each Unit 2 anchor stud shall be taken. 'Ihese readings shall be reported to Project Engineering for review. 10.3 CHECK CALIBRATION OF RETENSIONING EQUIPMENT Upon completion of the retensioning of the Unit 2 anchor studs, the calibration of the rams and pressure gages shall be checked as described in Section 8.1 and 8.2. Records shall be maintained as described in Section 8.4. ~ 11.0 DETENSIONING UNIT 1 ANCHOR STUDS 11.1 PREPARATION Anchor studs shall be prepared as described in Section 9.1. 11.2 INITIAL EXTENSOM5TER READINGS One set of extensometer readings shall be taken on the studs as described in Section 9.2. These values shall be recorded and are for project engineering information purposes only. 11.3 DETENSIONING when the initial extensometer readings are taken, the jam nut can be removed from the A-16

g 6 stud and the stud can be detensioned. This shall be done in the order shown in Appamh r 2 *. Detensioning is to be done using the retensioning frame without the displacement gages or extensometer installed. 12.0 RETENSIONING UNIT 1 ANCHOR STUDS 12.1 PREPARATION Preparation for retensioning Unit 1 is as described in Section.10.1. 12.2 EXTENSOMETER READINGS A complete set of extensometer readings shall be taken on the detensioned studs as described in Section 9.5. 12.3 RECALIBRATION Before coastencing Unit 1 anchor stud ram retensioning, the retensioning ram, pressure gage, and pump assembly shall be recalibrated as described in Section 8.2 or by checking the assembly in the strain gage frame shown in Appendix 1, Page 1-9. This frame shall have been previously calibrated by measuring the strain shown in the strain gages versus the applied load applied by a certified testing machine. Records shall be maintained as described in Section 8.4. 12.4 PROOF-TESTING Designated studs in Unit I will require proof-testing to two times the maximum anticipated stress, but no higher than 344 kips. The proof test value and stud numbers will be given by project angineering before the start of retensioning. These proof test requirements will be determined ucon review of the lift-off data from Unit 2. The procedure is as follows. After the studs have been detensioned, two sets of extensometer readings shall be taken as described in Section 9.5. The Unit 2 detensioning frame shall be installed as described in Section 9.3 with the extensometer and dial gages in place. The load shall be brought to the given value in stages as determined by the extensometer capacity, When the load is at the proof-test value, it shall be held for 1 minute and then removed in stages as described in Section 9.4. ~ A A-17

.( complete set of extersometer, pressure, and dial gage readings shall be recorded for this operation. Upon completion of the proof-testing of the Unit 1 anchor studs, the calibration of the rams and pressure gages shall be checked as described in Section 8.1. Records shall be maintained as described in Section 8.4. 12.5 RETENSIONING Retensioning shall be done as described in Section 10.2. In addition to the anticipated readings shown in Section 10.1, the following are expected 1or studs which have a turned down shank with a gross area of 4.0 square inches. These studs have a centerpoint for machin4ng. Displacement Load Stress Cage Reading UT Length (kips) (ksi) (in.) (in.) 0 0 g 0.0000 88.0000 4 1 0.0057 88.0029 20 5 0.0285 88.0144 12.6 RECALIBRATION After completing Unit i retensioning, the ram and pressure gage calibration shall be rechecked as described in Section 8.2 or 12.3. If there is more than 2% variation between the load indicated by the previous calibration and the applied load, the load in the Unit 1 anchor studs shall be rechecked as described in Section 10.2 using the recalibration load versus pressure data. 13.0 DOCUMENTATION on completion of the retensioning, a report shall be submitted to the project engineer listing the pressure, displacement gage, and extensometer readings occuring at: a. Unit 2 lift-off b. Unit 2 detensioning (each step) and proof-resting c. Unit 2 retensioning A-lb

. o d.. Unit 1 detensioning and proof-testing t. Unit 1 retensioning The report shall include a description of the operation, difficulties encountered, and pertinent remarks. This information will be used by the project engineer and the consultant to explain scatter in the previo~us ~ detensioning operation in Unit 1 and to certify that this tensioning operation ascertains that the anchor studs are at the specified load level. a b* A-19 1

= w w g g e 4 e' ,e e 5 2 APPENDIXES a a ? I l

APPENLIX 1 DETENSIONING SYSTEM d 5E"2 HYDRAULIC RAM \\ PUMP 3.WAY NEEDLE VALVE PRESSURE G AGE VALVE = .EEC1 HYDRAULtC RAM / PUMP 3.WAY NEEDLE VALVE PRESSURE GAGE VALVE l Al-1

I RETENSIONING SYSTEM z 4 HOLLOW COR E miC2 HYDRAULIC RAM \\ PUMP 3WAY NEEDLE VALVE PRESSURE G AGE VALVE Al-2

B W E IV R EE 4 NR GT E NA M LA F ID T E R E R GE I D NT g KI S AE S T LM VU F A RO TI L_SI E D

g t\\

T K D k. / R N / J I I S V K ya\\ f N RI S '~ N 'N V / a R ~ n M L A E Q / \\ B y g- / \\ ~ S S I / \\ / /, \\ O /. R E IC i \\ C l /\\ A V I i R E D E W DW T U E I b/ I V NV T M S A E GE G I I D N I E L NN S P k A OA U J B O I L C S M P N A R ET ) E \\ g / R D y- \\ /N 4 \\ j i E T / \\ N ~N \\ / E l/ { M f A I D E R x ID E \\ Y w X,/ TT ? (j S R E T g; O M U j /y y P A O s I 'j / \\ D R j E \\j E H s,\\ D i' S I A S f\\ T W UO D U TS VR / C& j \\

a {5 4 6 0 l l r,, 9 9 $/M 22N$ a +/ <x [ / / / CROSS BEAM /j p/ _ g p -n--- c,,,,,,, 9 h b $) b ///////// /' //////// i i c / COUPLING NUT /qI RAM ULTRASONIC r 1RANSDuceR o / y 7 p o A J rV To - SUPPORT BLOCK \\ N / RV NUT & STUD VIEW A DETENSIONING DEVICE u-k . e - - - - - - -----i----- - -a

o RV SHELL RV SKIRT I I 3o' r y- / / / / / / / h6 hh ~ CROSS BEAM (CUT AT CENTER LINE) / / 4 // / COUPLING STUD / _l,, / / COUPLER NUT f h RAM r / / SUPPORT BLOCK - e v v ' RV STUD & NUT J f I VIEW B DETENSIONING DEVICE n-5

l o . o l TIIIS SHEET INTENTIONALLY LEFT BIAmt. O 4 e S e i i I l l l Al-6 i

t I l RETENSIONING FRAME be - 4.%"-- HOLE FOR 1" DIAMETER i g o 8 UN 28 THREAD I I B-B TAP ENTIRE DEPTH I l 7 8 nr-B l / l / MATERI AL TO BE

ASTM A 194, Gr 2H I

4 %" l 3 %" ! /' HEXAGONAL 57/8" ACROSS I / / FLATS \\ 1 6 / m PLAN 1 W DIMETER HOLE A_L o A / 4 h3/8* \\ HOLE 3-%" DEEP FOR 2 %" DIAMETER l l l 1" 4-UNC-2B TH READS TAP 2 %" DEEP o 10 %" A-A COUPLING NUT T-MATERI AL TO BE 1" DN AETER 8-UN 2A THREADS ~ ASTM A 490 A 354 j., 11 %" or A 540 ALL PLATE MATERIAL ANY GR ADE TO BE ASTM A 36 ELEVATION _L RAM SUPPORT FRAME COUPLING STUD 1 Al-T

DETENS10NING DEVICE ASSEMBLY SEQUENCE 1) 2) w odB n o i i -BLOCK BEAM ABOVE STUD -INSERT & BLOCK STUD COUPLER lNSERT RAM SUPPORT BLOCKS ATTACH UT TRANSOUCER -INSERT NUT SOCKET RING 3) 4) 3r9r RPR a m U O ~= CT rm ATTACH COUPLER -UNBLOCK BEAM, REST ON RAMS lNSERT R AMS MOUNT DISPLACEMENT INDICATORS & LEVEL BEAM, ATTACH TOP UNIT

5) FOR INSIDE STUD RING LENGTH GAGE -

k N RPV SKIRT , LEVEL GAGE g Q M LENGTH GAGE W 'r r1 r1 r1 r1 PLACE t ON TCP UNIT & FASTEN WITH JAM UNIT p MOUNT DISPLACEMENT INDICATOR OFF ( ON TOP OFF COUPLER Al-8 9

e 9 1** d STUD MATERIAL i II I I I I i 1 l l HOLLOW CORE l l 20-TON RAM 1 I I I I I I I si 'I l l COUPLING STUD ULTRANSONIC COUPLING NUT TRANSDUCER f RAM SUPPORT FRAME \\ \\ RV STUD & NUT l t l RETENSIONING DEVICE l Al-9

l i i RECAllBRATION FRAME 1 %" DIAMETER SOCKET FOR ATTACHMENT / TOTEST MACHINE !I ii i "I i i l I L_J l l l 1" DIAMETER GM 24 TRAIN GAGES / BOLT ) A BOLT 20-TON HOLLOW CORE a l l l BAR 2-%" x 6" WIDE l l [_l l l ASTM A 36 MATERIAL f a f l I l 1 i l NOTE: CAllBRATE FRAME AND STRAIN GAGES IN TESTING MACHINE BEFORE USING TO CAllBR ATE 20 TON R AM. SCALE 3"=1' O" Al-10

Appendix 2 TABLE 1 DETENSIONING AND RETENSIONING SEQUENCE REACTOR VESSEL ANCHOR STUDS Bolt Number Sequence B&W Tetedyne ~ 1 01 in 37 in 2 02 in 13 in 3 03 in 01 in 4 04 in 25 in 5 01 out 37 out 6 02 out 13 out 7 03 out 01 out 8 04 out 25 out 9 05 out 43 out 10 06 out-19 out 11 07 out 07 out 12 08 out 31 out 13 05 in 43 in 14 06 in 19 in 15 07 in 07 in 16 08 in 31 in 17 09 in 40 in 18 10 in 16 in 19 11 in 04 in 20 12 in 28 in 21 09 out 40 out 22 10 out 16 out 23 11 out 04 out 24 12 out 28 out 25 13 out 46 out 26 14 out 22 out 27 15 out 10 out 28 16 out 34 out 29 13 in 46 in 30 14 in 22 in 31 15 in 10 in 32 16 in 34 in 33 17 in 38 in 34 18 in 14 in 35 19 in 02 in 36 20 in 26 in 37 17 out 38 out 38 18 out 14 out 39 19 out 02 out 40 20 out 26 out 41 21 out 44 out 42 22 out 20 out A2-1

1 i l Bolt Number Sequence B&W Teledyne 43 23 out 08 out 44 24 out 32 out 45 21 in 44 in 46 22 in 20 in 47 23 in 08 in 48 24 in 32 in ~ 49 25 in 41 in 50 26 in 17 in 51 27 in 05 in 52 28 in 29 in 53 25 out 41 out 54 26 out 17 out 55 27 out 05 out 56 28 out 29 out 57 29 out 47 out 58 30 out 23 out 59 31 out 11 out 60 32 out 35 out 61 29 in 47 in 62 30 in 23 in 63 31 in 11 in 64 32 in 35 in 65 33 in 39 in 66 34 in 15 in 67 35 in 03 in 68 36 in 27 in 69 33 out 39 out 70 34 out 15 out 71 35 out 03 out 72 35 out 27 out 73 37 out 45 out 74 38 out 21 out l 75 39 out 09 out 76 40 out 33 out 77 37 in 45 in 78 38 in 21 in 79 39 in 09 in 80 40 in 33 in 81 41 in 42 in 82 42 in 18 in 83 43 in 06 in l 84 44 in 30 in 85 41 out 42 out 86 42 out 18 out 87 43 out 06 out 88 44 out 30 out 89 45 out 48 out 90 46 out 24 out 91 47 out 12 out 92 48 out 36 out l l A2-2

. i Seit Ni~ker Secuence B&W Teledyne 93 45 in 48 in 94 46 in 24 in 95 47 in 12 in 96 48 in 36 in l l A2-3

N~ - A, '. ;-- @ @@g ~ g @h..g g OUTSIDE - ~- 7e*@ - g o ^'s e s b INSIDE g g b ~ G REACTOR SKIRT FA! LED -FAILED g INSIDE @ g....e@. e g OUTSIDE POSITION AND NUMBERING OF STUDS IN UNITS 1 AND 2 j A2-4

r,. / , ~ ~. 2 ~

- w y

. ~, /r. * ?'~ l t. T J. c ,.. A + W f. p ' ^ " " #g 5* ' APPENDIX B: s v.; -~ /' y-Procedure for Gap and Temperature Measurement at the Reactor Pressure Vessel ~ .,Ipper Lateral Support i / (v 9 r h a ? I 1 i l i-1 l l s 1 i I a i i l 1 mil 181-0953a141

L ', -) ] s s APPENDIX B PROCEDURE FOR e CAP AND TEMPERATURE REASURE3ENT AT THE REACTOR PRESSURE YESSE1 UPPER 1&TERAL SUPPORT 10 SCOPE this appendix provides for the installation of equipment and the implementation of procedures gecassary to measure the following while the nuclear steam supply system (NSSS) undergoes hot functional testing (HFT) (see Appendix 1, Page 1-2 for nom encla ture): 1 The change in gap between the reactor vessel apper lateral support (U1S) and the reactor gressure vessel (RPT) h. The corresponding change in U15 temperature, RPY surface, and concrete surface near the gabedaant this appendix does not specify the exact -design of the electrical system (transducer, thermocouples, wiring, readout system, and other accessories) required to obtain and record the data. Ihe intent of the appendix is to provide a system complete with all accessories sufficient to obtain the required data. 11 ITEMS INC1DDED 111 Euppir, calibration, installation, and operation of thermocouples, distance-reading levices, and readout units. Euppir of all associated wiring and other miscellaneous items required for obtaining asasurements 1 1.2 Eupply, installation, and removal of temporary shin pack replacement material 1 1.3 Collectice of da ta from all instrumen ts covered by this appendix. 4 e B-1 ,,---.--e

1 1.4 Eemoval of all instruments and associated equipnent and wiring upon completion of HFT 12 Intentionally left blank 2.0 Intentionally Left Blank 20 22A,1 TTY STANDARDS 111 procedures, Calibrations, and sensurements shall be in accordance with the requirements of a quality assur-ance progran approved by the Midland Project Quality Assurance Depart =ent (MPQAD). u.0 REFER ENCED COD ES A ND STA NDE RDS leerican National Standards Institute (ANSI) MC96.1 leerica n Society for Testing and Ma terials ( ASIH) B29-79 Standard Specification for Pig Lead 10 SUBMITT415 Ihe instrument supplier shall provide a complete descrip tion of all instruments, including method of operation, calibration, effect of tamparature-changes, viring, accessorias, and power required in its proposal for approval by project angineering. The supplier shall submit calibration procedures to project engineering for approval. Once calibration is completed, calibration data and certification of standards used shall be submitted for approval. B-2

10 EBKTNG CONDITIONS 11 EENERAL y!ork covered by this appendix is to be done during HTT for the Consumers Power Company Hidland Plant Units 1 and 2 in Eidland, Nichigan. Hot functional tests are scheduled as follows: Unit 1 May 1 to June 26, 1983 Unit 2 Januarr 26 to Earch 22, 1983 Instruments for the gap and temperature measurement shall be installed before the HFT. 6.2 CONDITIONS OF SERVICE The anticipated maximum temperature at the RPV surface-is 530F. Although the reactor cavity air temperature vill be lover, the distance measuring instruments, thermocouples and associated viring shall be rated for use in the temperature range of 500 to 600 F and humidity range of 0 to 100%. the recorders or rendout units will be remote from the RPY in an area outside the reactor cavity and will be subject to changes in tempera ture and humidity during the testing period. Ihe expected minimum and aarinus temperatures are 50 and 120F, respective 17 l 20 ENERat PEOUTREEE 21 REFINITIONS A) Small distance transducers are g means of converting the physical novement of an object into an electrical signal. Eor purpose of this specification, the distances to be measured are less than 1 inch. Ihe instruments are to-be capable of resolving a novement of 0 001 inch. Ivo types of transducers are samtioned in this specification: the eddy-current noncontacting type and the contacting 17pe. Ihe eddy-current noncontacting type works by producing sagnetic fields which induce eddy currents in the B-3 1

4 adjacent la rge t as terial. Changes in distance from the transducer to the s target rescit in impedance changes in the active coil of the transducer, 1hich can be nessured and converted to a distance sessurement. Ihe contacting type transdu:er has a noveable spindle within a fired coil assemb17. Eben the spindle is displaced, a voltage change is produced, which can be seasured and converted to a distance leasurement. h) The noncontacting transducer gan be used remotely from the noving vessel by attaching a target of the same gatorial as the vessel, st a fixed distance f rom the vessel. g) Electronic 1:e point is a means 2f eliminating errors in thermocouple readings caused by changes in ambient temperature and internal thermal voltages generated in the readout instrument by referencing the thermocouple leads to a temperature 2f OC or 32F. 12 SE3ERAL 1s stated in Articia 10, this document sets criteria for obtaining equipment and its calibra tion, installing the equipment, and establishing operation procedures to obtain the ga p sensurement hetween the ULS and the RPY and to obtain the RPY and ULS temperature a t selected poin ts iuring HTT. Io measure the gap, temporary instrumentation consisting of sas11 distance transducers installed on each U1S, their viring, and readout units will be required. Io obtain the RPT and DLS temperatures during HTT, a series of thermocouples installed on each ULS, their viring, and readout devices will be requirad. Erovisions will be made so that, if a por, tion of the systes or one readout device failsJj sfhe amount of da ta lost will not invalidate the test. 73 Intentionally left blank. B-h

24 1 RACKET PREPAR ATICS The ULS, shall be complete with all goverplates, stiffeners, and aschined faceplates in place and complete. Ihe shia material shall be replaced by cheeical lead 3/4-inch thick with additional stainless stee* shins sufficient to bring the nachined faceplate to 15/32 inch from the RPT at ambient temperature (see Section 8.2 and Appendix 1, Pgs.1-1,1-3,1-4 ~ and 1-5 for details). Ihe lead shall be installed and the bolts connecting the faceplate to the end af the bracket shall be tightened to a snugtight condition. Ihe gap between the RPT and the faceplate shall then be within41/64 inch acd -0.0 inch of the specified gapD Ihe temperature of the RPY and ULS shall be recorded when the shias are installed and adjusted. Z5 EHIELDING PREPARATION The permanent concrete shield plug cover and the removable steel shield plug boxes, complete with tiller material, vill be completed and installed before the EFT. 16 IHEEE0 COUPLES Iive thermocouples are required at each ULS for a total of 60 per unit. Qne thermocouple i shall be located on the RPY surf ace adjacant l lo the nachined contact patch. Ihe second shall be placed on the ULS web near the endplate. Ihe third shall be placed midway along the ULS on the web. Ihe fourth shall b. placed on the embedaent surface near the ULS web. Ihe fif th shall be placed on the concrete surface adjacent to the embedmont (see Appendix 1, Pgs.1-1 and 1-3, for location). Thermocouples shall measure te=peratures from 50 to l 600F. ~ 1 17 RISTANCE-EEASURING DETICES Iach ULS requires two distance-seasuring transducers (for a total of 24 per unit) of 1 the contact or noncontacting type. Ihe B-5

transducers shall he attached in the location shown in Appendix 1 Pages 1-1 and 1-4 or 1-5 28 EEADOUT UNITS leadout units may be strip type recorders and/ or digital readout units. Ior the temperature readings, a single readout unit any read as many as 12 thermocouples (12 channels). Ior the iistance sensurements, a marians of three ULS (six seasurements) any be taken by one readout unit. If multiple measurements are taken by one readout unit, the unit shall be connected to the instruments so tha t failure of the readout unit or its 11 ring will not cause loss of all data from any group of three adjacent ULS. Instrument suppliers may propose data processers which will i automatically record all output from the readout unfts glaultaneousir on paper. Relaration in requirements for the number of readout units will be considered if the supplier can demonstrate acceptable reliability in its proposed unit. 29 XIRING 111 wiring for the test

• instruments described in this specifica tion is temporary.

Hiring details and routing shall be determined in conjunction with the Consumers Power Company startup god testing group. Eiring in the reactor cavity shall be subject to the specified temperature and humidity (see Section 6 2). Irpes of wiring aust be compatible with the instruments being used and the test gonditions, and shall be reviewed by the instrument supplier. Ecurces and locations of temporser power shall be designa ted by the Consumers Power Company startup and testing group. 2 10 REACTCH PRESSURE YESSEL INSU1ATION lefore starting the HTT and after installing all instruments, all RPY insulation supplied by the Mirror Insulation Unit of Diamond Power shall be in place. Ihis shall include any insulation required to seal the opening j l through which the ULS projects. Eortions of l this insulation aar have to be removed later to remove the instrumenta tion and wiring. l gupp1 ring, installing, and removing the Ba6 l l

insulation is the responsibility of the NSSS supplier. 1 11 EElATED MEASUREMENTS Io relate the temperatures and distance measurements taken under this specifica tion lo operating conditions, temperature seasurements of the reactor coolant systen shall he recorded at reactor inlets and outlets. Ihese data shall be recorded simultaneously with the thermocouple and distance readings. i Additional reactor coolant system temperature sensurements any be recorded between the thermocouple / distance readings. 1 12 GALIBBATION 111 instrumenta to be used in accordance with this specification shall be calibrated for use in the temperature range which will exist during the test. Each instrument supplier must furnish a calibration procedure for approval when instruments are gurchased and shall produce a correction curve or demonstrate the means for the data to be corrected for temperature effects. Ellibration procedures shall account for the temperature conditions and length and type of wire from the instrument to the readout unit. If an eddy-current, distance-seasuring device is used, consideration for lhe target satorial is required in the calibration procedure. All calibrations must be traceable to the National Bureau of Standards (see Section 9.2). 10 EITER T R LS 11 EERMANENT HATERIAlS Zersanent asterials are to be installed by others as specified in the design drawings or by the NSSS gad insulation vendors. Ihe exception is the shin pack between the endplate and the RPY incaplate (see A ppendix 1, Pages 1-3,1-4 and 1-5). 12 1HIH PACK L 3/4 inch thick portion of the permanent shis stack material ( ASTE A-240, stainless steel). shall be replaced for the duration of the h7T with chemical lead meeting the requirements of O

4 e ASTM B 29-79. Ihe initial sira and shape of the lead shall be as shown in Appendix l, P a g e 'l-6.- In the test shin pack, the lead shim shall be located between two 1/8 inch thick stainless steel shias. Additional stainless steel shiss of various thicknesses will be added to bring the gap to the predetermined Islue (see section 7.4). 111 stainless steel shias used in the test shin pack shall have the san:e confire.ra-tion as those ssed for final construction. 13 IHER 50 COUPLES Ibermocouples (60 plus 6 spares per unit) shall be Ceeent-on Thersocouples, Style III, ANSI Designation K Chronel-Alumel or E Shronel-Constantan or equivalent. Extension wire shall be chosen to match thermocouple type and layout to ensure sufficient signal output to sensing, sending (if required), and recording devices. Ihermocouple and extension wire insula tion shall be fused teflon, tape-teflon-impregnated glass, or equivalent guitable f or continuous use to 600F. Cement I to attach thermoccupies to Uls, embedment and concrete shall be two-part thermoccat gepper oxide cement. gesent to attach thermocon;1es to the RPY surface shall be Omaga CC sodium silicate cement or equivalent. Ihe brand name products discussed in this parsgraph are supplied by Omega Engineering, Inc. Ihe instrument supplier may proposa alterna tive systems for review and approval br project engineering. 14 RISTANCE-MEASURING SISTE55 17 stems shall include transducers, cables, power supplies, sending units (if required), readout units, and all necessary accessories. 1 4.1 Ristance-sensuring devices (24 and 2 spares per unit) shall be eddy-current, noncontacting lype or contacting type tra nsd uce rs. Ihe devices shall have a range from 0 to 1 inch. Ihe transducer vill be rated for service in a temperature range of 50 to 600F. B-8

1.n.2 Eover supplias shall be cespa tible with the transducer and the Evailable temporary power. Ihe electrical power for no more than six transducers shall be supplied from the saae gover circuit. 1 4.3 gables shall be supplied with sufficient length to reach from the ends of the UlS to the readout and da ta acquisition area._ g.a.a Ihe distance-seasuring system shall be calibrated throughout the given temperature range. the eddy-current system, if used, shall be calibrated using a target of the same material as the RPY isee Section 9.2 2 for the calibration procedure). Ihe instrument supplier aar propose alternative or modified devices or materials if it 1eems then more suitable. 15 READOUT UNITS leadout units shall be one of, or a combination of, the following. Atrip type chart recorders, digital readout indicator, multipoint recorders, data loggers, or alterna tive groposals by the instrument suppliers. [ heir range shall be from 0 650F for tempera ture, gnd 0 to to 1 inch for linear displacement. Iber shall be capable of being read to the naarest 1.0F for tempera ture, and to 0.001 inch when seasuring distance. Ehen multichannel units are used i reduce the number of readout instruments to j required, input data from no more than three ULS ghall be read or recorded on one instrument. It least four separate systess shall be used to record the data (see Section 7.8 ). Ihe readout systen shall be compatible with the the rm oc ouple s, transducers, power supplies, Eiring, and operating conditions present during the HTT. Ihe instrument supplier shall provide complete details on its proposed systen, in=1uding all instruments and accessories required and all equipment and power to be supplied by others. B-9 3 y.,,,g ._..._m__._ .,,, _. - -... - _,.. ~ -. _. - - -. _ _ _

Zossible data collection systems include: 4 g) Strip tTpe chart recorders with one to-three thannels using such accessories as the Dataplez 10 Lutomatic Signal Scanner, expander / marker and pen lift, electronic ice point, gay 11fiers, and two-wire transmitters as supplied by Caega Engineering, Inc h) Digital readout' indicators in combination with multipoint gelectors, amplifiers, and transmitters may be useds glectronic ice point and other accessories are to be supplied, if necessary, hr bdesa Engineering, Inc. A tranducer indica tor similar to Model 1002-0010 with eight-channel Sutput can be Esed for dispiaring contacting type treasducer data. Ihese are as supplied by Trans-Tek, Inc. Ihe eddy-current noncontacting transducer is psrt of g seasurement system which includes a digital readout. Ihe instrua6a t supplier could provide multiple readout units for a combination of as aanT as eight transducers. Iuch a system is supplied by Kaman Sciance Corporation. s) A recording device, such as Digistrip by Kaye Instruments, aar be considated f or use provided that the supplier can demonstra te the recorder's reliability to the satisfaction of project engineering. 1) Alternative proposals shall be submitted for project gegineering approval. Ihe instrument supplier must provide complete data on devices to be used, as well as all required gecessories and viring. Ihe supplier l shall provide technical data necessary to assess sensitivity, range, and gecuracy of all instrumants as well as their suitability for use in this application. Ihe supplier shall also l describe all requirements for the proper operation of its equipment (see Section 5.0). B-10 i w,,,m. 7 ---.7-- ,r ,,-e

10 LISTA11 ATION A ND TESTINO 11 EEQUENCE OF WORK l lork shall be performed in the following sequence: g) Calibrate (in the laboratory) distance-measuring devices and thermocouples as specified in Section 9.2 1 h) Install distanco-seasuring devices and thermocouples g) Test installed devices for proper operation 1) Install power supplies, readout unit ziring, and associated gecessories g) Test and calibrate the measuring system in place 1) Complete RPY insulation 2) Hot functional test Unit 2 1 January 1983) h) Hot functional test Unit 1 iMay 1983) 1) Remove testing equipment 12 SALIBRATIDI 2 2.1 Iberaccouplas shall be calibrated.such that the output volta ges at standard temperatures in the range of.50 to 600F within ANSI HC96 1 error limits f or are the thersoccuple type supplied. 1 minimum of three thermocouples shall be checked by heating the thermocouples through the range of 53 ta 600F, 111owing then to cool tc $0F, and plotting the output voltage against the actual temperatures (indica ted by a calibrated standard) at 25F intervals throughout the range. Ihis heating and cooling sequence shall be repeated a second time to show repeatability. Ihe three thermocouples used in the calibra tion shall be of the same tTpe and from the same groduction run as those to be supplied. B-11

9,alibration procedures and certification of standards shall be submitted to project engineering for approval, as described in Section 5.0. All standards aust ha traceable to the National Bureau of Standards. 11ternative calibration methods based on recognized stsadards or codes ear be prepared and used eith prior project ingineering a pproval. Ealibration da ta shall be submitted, as described in Section 5.0, to project engineering for reviee and approval before installing gar thermocouples. 1 2.2 Iransducers shall be calibrated as folloes. Ihe transducar shall be attached to a section of 1-1/2-inch thick plate in the saae manner in ehich it eill be attached to the ELS during the HTT. A target or rod of the same type. to be used in the actual test shall also be mounted on the plate. Ihe transducer, its mounting plate, target, or rod shall be installed in a furnace or oven capable of bringing the unit to a eniform temperature and holding it at 50F and at intervals of 53 to 500F. Ihe temperature shall be held at each data point for 2 minutes. Ihe transducer shall be connected to a resdont unit of the same type to be used in the actual test. Ihe rod.from the target or the transducer shall extend out of the furnace in such a manner that it can be deflected by a calibrated microseter or dial gage through the design range of the transducer isee Appendix 1, page 8 for suggested arrangement of test a ppa ra tus). Elots shall be made of actual deflection versus indicated deflection for increments a miniana 0 025 inch and a aariana 0 05 inch. Ibree sets of readings shall be taken a t each temperature. Ihe readings a t each point are not acceptable if ther vary by more than 0.005 inch. Nonconforming transducers shall be ad, justed or repaired and recalibrated or replaced with a calibrated transducer. At temperatures beyond the nor=a3 operating range of the transducer, the actual " deflection suisy vary considerably from the measured deflection. This B-12

Variation aust be consistent and predictable a r each point of measurement. Salibration procedures shall be submitted to project engineering for review and approval before gcamencing calibration. 111 seasurements will be traceable to the National Bureau of Standards. Ealibration results are to be submitted to project engineering for review and approval, as specified in i Section 5 0, before installing any instruments. 13 INSTALLATION 1 3.1 Iransducers shall be installed with /d fey sheet metal brackets screwed to the ULS sideplate. Ihe bracket shall hold the transducer br a clamp or other means suited to the transducar design. In either case, the attachment shall be positive, have essentially zero deformation, and shall not relax with time or heat. Ihe transducer contact rods or (if used) targets shall be inserted through the 1/2 inch diameter holes in the,ULS bearing plates. E,efer to Appendiz 1, Pages 1 h and 1-), for transducer location Retails. C,a re shall be taken to ensure that the transducer is installed so that it may detect RPY novement through the range of 0 to 0.5 inch. 2 3.2 Ihermocouples shall be installed with cement Kuitable for approximately 600F. Suggested cements are Ther::ccoat and Omega CC high-temperature cement by Omega Enginee;ing Inc. Refer to Appendix 1, Pege 1-3 for ther=o-couple locations. 1 3.3 11 ring shall be installed as recommended by the instrument supplier. Hiring in the reactor cavity which is subject to hea t shall be 2rotected or coated as recommended by instrument suppliers. Lithough the wiring is B-13

a temporary, it shall be neatly tied to supports, protected from ibrasion, and installed so it will not fail during the test period. All connections shall be protected consistent with good electrical practice to ensure continuous system operation during the test. 2 3.4 Eeadout. instruments shall be located in the area desitnated by the Consumers Power Company startup and testing group. Eculpment shall be neatly installed. 2 3.5 Ihe location of all instrumen's shall t be documented according to the instrument sarial number. Eiring diagrtas shall be made which indicate the instrument number, gover supply circuit, accessories, and readout unit and which trace tne physical gath of the wiring. Lhe ULS shall be designa ted as shown in Appendix 1, page 1. R.4 ZESTZRC As instruments are installed," they shall be checked for proper operation. Iransducers and thermoccuples installed according to Sections 9.3.1 and 9.3.2 shall be ghecked using a temporary readout unit and circuit to ascertain their 2 roper opera tion by measuring the actual contact temperature and by displacing the transducer through a known deflection. Eeadings shall be verified using calibra ted standards traceable to the lational Bureau of Standards. Zhen instruments are installed and viring completed, all circuits, instruments, and readout devices shall be checkad for proper operation as described above. Galibrated standards shall be used to verify proper system functioning. Ihe above described activit" performed according to precip.es shall be uros reviewed and approved by project engineecing. Complete documentation shall be maintained on all tests, repairs, modifica tions, and other B -lh

corrective actions required to verify system operation. 15 gCT FUNCTIONAL TEST, UNITS T AND 2 1 5.1 Ihe HFT shall be carried out based on information provided by Consumers Power Company startup and testing group. 1 5.2 Its saquence of temperature versus time during the HFT shall be as shown in Appendir 1, Sheet 9. 25.3 leaperature and distance data shall be recorded as follows: Readings shall be taken at the beginning, middle, and end of each temparature hold and at midpoint of til transitions as the the temperature. rises and f alls (refer to A pp en dix 1, Page 1-9). A minimum of one of readings shall be taken each set 12 hours. 25.4 111 equipment feilures shall be reported to the pro 3ect angineer. the failure shall be corrected if poss'ible. lecause each HFT goes through two cycles of beating and cooling, all instrument failcres shall be corrected between the first and second cycles if possible, without unacceptable delar of the HFT. 1 5.5 1 complate sat of data shall be transmitted to the project engineer after ahe first RCS heating and cooling cycle is complete. Ihis chall include the temperature of the RCS coolant at the inlets and outlets. Iha location of the RCS coolant temperature sensors shall be given end the $sta shall be referenced to the da te and time of day. Ihese tear.ers tcras shall be recordai simultaneously with the thersecouple B-15 m m ---, + +-, w m. -m

and iictance readings raquired by this precedure. 2 5.6 1 hen the Unit 2 HFT is completed, the test will ha reviewed by project engineering. It that time, ell failures and senaeace of data recording shall be reviewed. If it is believed that the system should be modified in any way to Erovide reliable data, these modifica tions shall be made before 2 tarting the Unit 1 HIT. If it is believed tha t more or less data are required, the test procedure 2 hall be modified to acconnoda te these requirements. 111 procedural modifications shall be reviewed and approved by project engineering. 1 5.7 2roject engineering shall be informed of the status of the installation and testing at all times. Eroject engineering shall be advised before the tasting is scheduled to begin. A representative of project engineering shall be present at the start and for the duration of the HTT. 13.0 10FK TO FOL10E 10.1 INCINEE2IIC 1 hen project engineering receives the HTT data, the temperatures will be compared to those predicted by analysis for operating conditions. 1hese predicted tempera tures may vary at different locations on the EPY. Ihe distances measured during HTT will then be proportioned to operating temperatures. Ihis will verify the specified gap in existence at opera ting leapera ture, and will confirm the gap required at cold shutdown. 1 hen the cold shutdown cap is verified, the informa tion will be added to the applicabla drawing and the shias will be installed to b rin g the UlS faceplate to the required gap. 10 2 SIEAN UP lhen HTT is complets, all transducers and viring shall be removed from the reactor gavity. lay tenporary attachsents for supports shall also be removed. the area B -16 l ,__ - - - - - - - - - - ~ - - ~~" ~ ~ ~ '

shall be left in a clean condition as determined by Consumars Power Company startup and testing.- 10.3 2THEa c0NsIDERATIONS Eap and temperature sessurement during Unit BFT shall be carried out 1 in a similar manner as Unit 2. Eroject engineering will review both the data collected during Unit the test operation. 2 HFT and Changes in procedures for setting gaps and changes in equipaent will be made, if necessary. 10.4 IQUIPMENT REMOY11 All equipment and viring used for the gap and temperatura seasurement shall be removed from the containment building at a time to be designa ted by Consumers Power Company startup and testing. i .s e B -17

o.. e APPENDIX 1 360' 0' LA AL SUPPORT (TYP) 12 / N 11 !j [ ,.. \\ / N /~ 3 1_ _ a -__= 1_ FT7 1-3^ 3 ,3 y 90* 270' l 8 R = 12*-O" i, \\. /).\\ l ^ \\; - / f o ~_ CCNCRETE 4 [ SHIELD PLUG N 2" / 1 7 6 PRIMARY SHIELD Wall 180' UPPER LATERAL SUPPORT PLAN AND NUMBERING SYSTEM UNIT 1 SHOWN, UNIT 2 OPPOSITE HAND B1-1

o REACTCR INSULATICN EL f'!3' 5%* 7, REACTCR I 8 PRESSURE PRIMARY \\ VESSEL I s SHIELD s (ppy} I- -- WAR

wAu, REMOVABLE 8

] SHIELD l EL 632*-3"% i s / ///s, s s / I r a I m - - g is si 5%" ll UPPER LATERAL 18 e - - '- ll M--]-]* \\ l j lI SUPPORT (ULS) l b' 'd d' U ULS BUMPER / l p --. N etArE e .,A e a x ,e I ,d v l a' l =- - s,# l # a s l n.88 l T.,..".*' ULS SEARING \\ lI SHIELD PLATE y PLUG A

• p a.

SECTION A-A NOMENCLATURE B1-2

e d~ f/ / / # / / / / / / / Y INSULATION 9 n \\ THERMOCOUPLE 8 2 2 (ON WEB) F-5 s A l ) ____,,_/ / / / / / / / # f g9gg i e _____._,,__._,___y I THERMOCOUPLE ll I j [+h l 2 4 (AT JUNCTION Il r-- OF WEB AND EMBEDMENT) iI I l 1I ii et iI Ii p' THERMOCOUPLE w w w, 21 w w j l ~'4, ~ * ' THERMOCOUPLE ' * ' 'T , 3 '- ' : e 2-5 ON CONCRETE 's ' 3 ',2 - ~' ~ l p ADJACENT TO x l p EMBEDMENT C2" : _ c.. M l l ,~,'s 40 \\ ! p'",, THERMOCOUPLE b l 2-3 ON ULS f WEB ../ l l$ g L_ INSULATION SECTION A A THERMOCOUPLE LOCATIONS SHOWN FOR ULS NO. 2 (5 UNITS PER ULS) B1-3

I r-1 I I TOS EL 632'-3" 1 %" E (TYP) 4 FACE OF RPV I N / s i N y m g Tr-- j \\ i I n n ii r--- si ll 11 i l l Is it il 11 i ,,,,,c- .i l u u uA U, %' _. -.-- 1 CONTACTING TYPE \\ .l-l l TRANSDUCER SUPPORTED ?- rime-ON THE ULS ' 'd 2 :! $ ' ~.9 l 1 ~ I / -+ +- r l SHIMS l l l ROD SPRING LOADED L-TO FOLLOW RPV MOVEMENTS SECTION A CONTACTING TYPE TRANSDUCER (2 UNITS PER ULS) B1-4

i l I l r_ 4---FACE OF RPV TOS EL 632*-3" 1 %"((TYP) l \\ / N l i,,.---,,---_- q g Il ll ,1 Il 11 11 I l ll (- - y g l l ll 11 Il 11 I is li li li i. c' U U U U- ,l s l NONCONTACTING TYPE l TRANSDUCER SUPPORTED 32"" ON THE ULS h,,... l w a 1 i l l' + + l shims I I l 4" x 4" TARGET AND ROD SPRING LOADED TO FOLLOW RPV MOVEMENTS SECTION A-A NONCONTACTING TYPE TRANSDUCER (2 UNITS PER ULS) 31-5

) d a 8 e / L 1 g i P "I 9 3 ) B 9 le1 M ec-T eaH S 8 tFX A / S 0 k 1 h4 sc2 m sa-i eEA h l S neM itT d aaS a tlA e SP( L "4 4/ / 1 3 2 9 " ) 4p /y ) 3T k ( c )Q = a ( sR ) P ) 2 7 Q 6 ( l Mm 3 7 Ii Hh 6 C 3 SS Dt g C As v E e r LT D g k w c r r a o 2 D P k F c t 1 m a i P e h D S m c i e e l h e D a S S i ( e d l e a a k (S R c c i a n p P k o y c N T m a i P r r h o o S m F F i l h "4 "4 a S i / / d l 3 3 a a R c 3 0 i 2 2 n p o y N T r r o o F F 4 4 / / 3 3 7 4 r "4 2 2 "2 ) p /y 3T( =R gln

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.ca. y l

Data Acquisition

• **Z '

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-s e., e- , [" ..s.o,. / a~ i - Nw.) -~ -ls v' ~ ,r.sa.. \\, 4 'F t A v ,"jp 3 NN N,A .e::.. .. yl g=- eg g.j e t 1 I.*.g~V,

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3 t ( 3L AhJ C t.. 334*.0* l l DATA ACQUISITION AREA LOCATION (UNIT 1 SHOWN, UNIT 2 OPPOSITE HAND) B1-7

FURNACE SPRING-LOADED ROD TO FOLLOW MICROMETER l l 3" x 5" x 5" THERMOCOUPLE l STEEL BLOCK WIRING g READOUT l MICROMETER

== % m s R_= TRANSDUCER %" HOLE l[ { l ALLOTHER PLA THERMOCOUPLE ~ READOUT / 4 A 516 , A 240 XM 19 CARBON STEEL STAINLESS STEEL l SUGGESTED CALIBRATION SETUP B1-8

2250 \\ Pressurize ' ) f FPussutt p..... a. 2000 I 1750 l i 600 .l 3 1500." eft* 4'i 500 1250,,- m 1000 3 g 400 = a. % 300

RC 750 k

Teeperature ' 5 200 500 ~ [ 250 100 3 -g 0 2 4 5 8 10 12 14 2 4 6 8 10 12 14 Time, Days Time, leeks HOT FUNCTIONAL TESTING RC TEMPERATURE AND PRESSURIZER PRESSURE B1-9

APPENDIX C: Unit 1 Anchor Stud Lift-Off Data / i 4 mil 181-0953a141

r' TABLE C1 a ,,j UNIT 1 REACTOR VESSEL ANCHOR STUDS Lift-Off Data x Stud Number (2) Date Hydraulic Bolt Stress Pressure to Nearest ksi Sequence B&W Teledyne (psig) 1 1 01 in 37 in 4-08 13,000 88 2 02 in 13 in 4-23 11,900 81 3 03 in 01 in 4-25 13,400 91 4 04 in 25 in 5-19** 9,300 63* 5 01 out 37 out 8,000 54* 6 02 out 13 out 12,500 85 7 03 out 01 out 10,800 73* 8 04 out 25 out 5-12 8,400 57 9 05 out 43 out 5-13 12,500 85 ~ 10 06 out 19 out 5-13 12,500 85 11 07 out 07 out 5-13 13,400 91 12 08 out 31 out 5-14 13,800 94 13 05 in 43 in 5-14 12,300 83 3 14 06 in 19 in 5-14 11,500 78 15 07 in 07 in 5-15 12,000 81 16 08 in 31 in 5-15 11,400 77 17 09 in 40 in 5-16 12,300 83 18 10 in 16 in 5-16 11,700 79 19 11 in 04 in 5-19 13,700 93 e 20 12 in 28 in 5-19 12,400 84 mi1181-0953a141

N TABLE C1 (Cwtinued) 21 09 out 40 out 5-20 12,200 83 22 10 out 16 out 5-20 12,500 85 23 11 out 04 out 5-20 13,000 88 24 12 out 28 out 5-21 12,300 83 25 13 out 46 out 5-21 12,800 87 26 14 out 22 out 5-21 11,500 78 27 15 out 10 out 5-21 12,300 83 28 16 out-34 out 5-22 12,600 85 29 13 in 46 in 5-22 11,100 75 30,

14 in 22 in 5-22 12,100 82 c. s 31 '15 'in 10 in 5-23 9,300 63* 32 16 in 34 in 5-23 13,100 89 33 17 in 38 in 5-23 11,600 79 34 18 in 14 in 5-27 9,500 64* 35 19 in -02 in 5-27 13,300 90 .s 36 3 20 in 26 in 5-27 9,600 65* \\ , 37 17 out 38 out 5-28 12,500 85 s 38 18 out 14 out 5-28 12,300 83 39 19 out 02 out 5-29 14,000 95 40 20 out 26 out 5-29 12,100 82 41. 21 out 44 out 5-30 12,200 83 4 42 22 out 20 out 5-30 12,300 83 .,' 43 23 out 08 out 6-17 12,300 83 L g _. 44 24 out 32 out 6-18 12,300 83 45 21 in 44 in 6-18 12,800 87 46 22 in 20 in " s ?.8 10,900 74* 4 i 47 23 in 08 it -79 12,300 83 48 24 in 32 in 6-4) 12,400 84 mil 181-0953a141 s s ~ ~w,n'.- ~

e TABLE C1 (Continued) 49 25 in 41 in 6-20 12,200 83 50 26 in 17 in 6-20 11,800 80 51 27 in 05 in 6-20 13,000 88 52 28 in 29 in 6-23 12,800 87 53 25 out 41 out 6-23 12,500 85 54 26 out 17 out 6-24 12,700 86 55 27 out 05 out 6-24 8,900 60* 56 28 out 29 out 6-25 12,500 85 57 29 out 47 out 6-25 10,200 69 55 30 out-23 out 6-25 12,200 83 59 31 out 11 out 6-26 12,200 83 60 32 out 35 out BR0 KEN 61 29 in 47 in 6-26 11,900 81 62 30 in 23 in 6-27 12,400 84 63-31 in 11 in 6-27 11,800 80 64 32 in 35 in 6-27 11,600 79 65 33 in 39 in 7-02 11,700 79 66 34 in 15 in 7-02 11,700 79 67 35 in 03 in BR0 KEN 68 36 in 27 in 7-03 12,300 83 69 33 out 39 out 7-03 12,100 82 70 34 out 15 out 7-03 12,300 83 71 35 out 03 out 7-07 12,000 81 72 36 out 27 out 7-07 10,300 70* 73 37 out 45 out 7-07 12,600 85 74 38 out 21 out 7-08 12,500 85 75 39 out 09 out 7-08 12,200 83 76 40 out 33 out 7-08 13,600 92 mil 181-0953a141

t, TABLE C1 (C*ntinuid) 77 37 in 45 in 7-09 13,000 88 78 38 in 21 in 7-09 11,500 78 79 39 in 09 in 7-09 12,200 83 80 40 in 33 in 7-10 13,200 90 81 41 in 42 in 7-10 11,800 80 82 42 in 18 in 7-10 12,500 85 83 43 in 06 in 7-11 10,200 69* 84 44 in 30 in 7-11 12,300 83 85 41 out 42 out 7-11 12,200 83 86 42 out 18 out 7-14 10,400 71* 87 43 out 06 out 7-14 11,800 80 83 44 out 30 out 7-14 11,700 79 89 45 out 48 out 7-15 13,100 89 90 46 out 24 out 7-15 10,400 71* 91 47 out 12 out 7-15 11,700 79 92 48 out 36 out BR0 KEN 93 45 in 48 in 7-16 12,500 85 94 46 in 24 in 7-16 11,900 81 95 47 in 12 in 7-16 12,100 82 96 48 in 36 in 7-17 11,700 79 NOTES:

1) Ram area of tensioner = 27.134 sq in, bolt area = 4.00 sq in.

2) Refer to Figure C-1 for the locations of the studs.

• )

Proof loaded to 75 ksi after detensioning.

• • )

Tensioner run up to 14,200 psig/96 ksi on initial attempt without being able to rotate nut. The lift-off data shown is the result of detensioning attempt after 20th in sequence. mil 181-0953a141

y 9% D @ g @e @ge e@,@ O g g OuTSiDE g@ FAH.ED g INSIDE h { REACTOR SKIRT hFAILED @g -FAILED 17 g@ iNSiDE O@ @O OUTSIDE g @ g@@@@g @g@ POSITION AND NUMBERING OF STUDS IN UNITS 1 AND 2 FIGURE C-1 _}}