ML20039E832
| ML20039E832 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 12/22/1981 |
| 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, 10CFR50.55E, 1S361, LS361, NUDOCS 8201110510 | |
| Download: ML20039E832 (150) | |
Text
.
Consumm j
Power James W Cook O
Vice President - Projects, Engineering and Construction General offices: 1945 West Parnell Road, Jackson, MI 49201 = (517) 788-0453 g
i Q
December 22, 1981
_neCEM9 9 9;
JAN 81992>. 72 Mr J G Keppler, Regional Director i$h'2'Nu h Office of Inspection and Enforcement 9A Tu "" //
US Nuclear Regulatory Commission
'i
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Region III S /[T[W 799 Roosevelt Ro3d Glen Ellyn, IL 60137 MIDLAND PROJECT -
MIDLAND DOCKET NOS 50-329, 50-330 UNIT NO 1, REACTOR VESSEL BROKEN ANCHOR BOLT -
FILE 0.4.9.35 SERIAL 15361 REFERENCES 1.
CONSUMERS POWER LETTERS TO J G KEPPLER, SAME SUBJECT a.
SERIAL 15035 DATED NOVEMBER 23, 1981 b.
SERIAL 14625 DATED DECEMBER 1, 1981 ENCLOSURE Report entitled, " Reactor Pressure Vessel Support Modification for Midland Nuclear Power Plant, Midland, Michigan, Report No 3", Revision 1, dated December 1981.
Reference 1.a was an Interim 50.55(e) report transmitting the updated technical report describing the reactor support modification, the schedule for accomplishment of that modification and the description of the analytical techniques being used, and Reference 1.b was the Final 50.55(e) report.
The two enclosures to Reference 1.a, " Reactor Pressure Vessel Support Modification for Midland Nuclear Power Plant, Midland, Michigan, Report No 3" and " Letter Report - Teledyne Engineering Services (TES) Project 5355:
Expanded Criteria for Acceptability for Service of Midland Unit 1 RV Anchor Stress", were discussed in detail with the NRC Staff in Bethesda on December 2 and 3, 1981.
During that meeting, the staff requested certain clarifications I N y J J
and supportive material.
I
/4.
I The enclosed report to this letter which is Revision I to the previous Report No 3 has been revised to incorporate Staff comments and in addition includes o,e S$
p l some minor text correction of an editorial nature. The revisions to the enclosed report are indicated in the right-hand margin.
This letter and its enclosed report is intended to 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 oc1181-0965a141 8201110510 811222 PDR ADOCK 05000329 S
SERIAL 15361 2
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reactor vessel support system. Upon completion of this task, the final designs and analytical results will be reported in the FSAR.
JWC/BFH/mo CC Diredio$.of Uffice of5Inspectioni& Enforc ment l(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 Attorney General w/o JHarbour, ASLB w/o DSHood, NRC (2)
FJKelley, Esq, Attorney General w/o t/HMarshall kI)Paton, Esq w/o MSinclair w/o GTTaylor, Esq, Ass't Attorney General w/o 9
oc1181-0965a141
r.
g.
.Y REACTOR PRESSURE VESSEL SUPPORT MODIFICATION FOR MIDLAND NUCLEAR POWER PLANT REPORT NO. 3, REVISION 1 DECEMBER 1981 CONSUMERS POWER COMPANY JACKSON, MICHIGAN
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mil 181-0953a141 9.
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O e
REACTOR PRESSURE VESSEL SUPPORT MODIFICATION FOR MIDLAND NUCLEAR POWER PLANT TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
1 l
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 l
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 l
3.3.9 Tolerances 21 I
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 THERMAL LOADS 24 mi1181-0953a141 i
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o 4.3 PRELIMINARY VS FINAL LOADS 24 5.0 ANALYSIS AND DESIGN OF THE SUPPORT SYSTEM 29 5.1 UPPER LATERAL SUPPORTS 29 5.1.1 Upper Lateral Supports Brackets - ifaxiuum Design Vs 29 Allowable Stresses and Displacements 5.1.2 Embedments - Maximum Design Vs Allowable Stresses 30 5.2 ANCHOR STUDS 31 3.2.1 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 HEAT TRANSFER AND THERMAL ANALYSIS 37 f
6.1 INTRODUCTION
37 6.2 TEMPERATURE AND PRESSURE CONDITIONS OF REACTOR PRESSURE VESSEL 37 NEAR UPPER LATERAL SUPPORTS
(
6.3 UPPER LATERAL SUPPORTS-VESSEL INTERFACE 39 6.4 PRESSURE DEFLECTION OF VESSEL 40 6.5 POTENTIAL GAP CHANGES DURING OPERATION 41 i
6.6 CREEP, THERMAL HACHETING, AND ELASTIC SHAKE DOWN 44 7.0 REACTOR PRESSURE VESSEL SURFACE PREPARATION 47 8.0 DENTENSIONING AND TENSIONING OF THE ANCHOR STUDS 48 J
8.1 DETENSIONING PROCEDURE 48 8.2 CREEP RECOVERY 48 8.3 RETENSIONING PROCEDURE 49 9.0 REACTOR PRESSURE VESSEL INSULATION MODIFICATION 50 10.0 GAP AND TEMPERATURE MEASUREMENTS AND GAP SETTING 51 10.1 MEASUREMENTS DURING HOT FUNCTIONAL TESTING 51 e
10.2 MEASUREMENT PROCEDURE 51 mi1181-0953a141 ii
. ~......
10.3 CORRELATION SETWEEN 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
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11.1.2.2 Internal walls structure 57 11.1.2.3 NSSS supports 58 11.1.2.4 Stiffness of upper lateral supports 59
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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 C5 FROM FINAL ANALYSIS 13.0 CONSTRUCTION STATUS AND SCHEDULE 86
14.0 CONCLUSION
87
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15.0 REFERENCES
89 mi1181-0953a141 iii
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.
D, Justification of the Ductility Ratio for Use in the Design of the ULS.
E.
Methodology for the Computation of the Mathematically Equivalent ULS
<d Spring Rates.
F.
Teledyne Engineering Services Letter (W E Cooper to H W Slager) dated December 11, 1981, Reaffirmation of Letter Report TR-5255-1.
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 1
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
)
l 5.2 Finite Element Model of the Reactor Pressure Vessel Skirt 35 5.3 Position and Numbering of Anchor Studs in Units 1 and 2 36 11.1 RV Isolated Model - Reactor Internals and SSS 71
)
I mil 181-0953a141 iv
-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 11.5 RV. Isolated Model - Elevation View C-C Cold Leg 75 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 f
11.8 Reactor Internals and Service Support Structure 78 l
11.9 Reactor Coolant System Boundaries 79 l
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 l
11.13 Bracket Resistance Versus Displacement Curve, Cold Leg Direction 83 l
l 11.14 Rotational Spring Constants at the RV Base 84 l
i 1
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l mi1181-0953a141 v
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I
1.0 INTRODUCTION
Unit 1 of Consumers Power Company's Midland Plant experienced failure of three reactor vessel anchor studs several weeks af ter being tensioned to a nominal value of 92 ksi 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).
l According to the investigation, the failure was due to stress corrosion crack propagation to a point where brittle fracture took place.
Modifying the reactor vessel supporting system to include the addition of the upper lateral supports (ULS) above the reactor vessel nozzles, t
along with stressing the anchor studs to a reduced preload level, will provide the necessary support for the reactor vessel (RV). Two reports J
(See References 6 and 7) were transmitted to the NRC in July and December 1980. The first report covered the initial design criteria of i
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the new support system including the allowable stresses. The second report covered preliminary design loads and methods of analysis. Other interim reports and responses to NRC questions have been provided and are listed in the transmittal letter for this report.
mil 181-0953a141
-. l
2 It has been determined for engineering reasons, that the gap size
.between the RV and the ULS should be increased from the nominal 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 4
continue to provide the necessary lateral support for the RV from the design basis loss-of-coolant accident (LOCA).
i This Report Number 3 provides the required details for both the design and the analytical methods used, and thus satisfies the commitments l
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 i
i previous reports, this report takes precedence and reflects the product 4
j of studies which have beer performed to both enhance the modified RV
]
support system and to assure the Company that the final design i
j adequately meets all safety requirements.
i 4
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1-4 mi1181-0953a141
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POSITION OF FAILED STUDS IN UNIT 1 NORTH 360*f0*
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o0 000 0 0 O
O O
O i
0 0
O 1
9 0
FAILED O
O O
O o
O O
O dHEACTORM O O 90' 270o._O OL.-FAILED SKIRT l
O O-1
@' O
-FAILED O )O O
O O o O
O 0
O 0
0o O
J O
0 0 0000 m
I 180' FIGURE (1.1)
4
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 l
the c'esign 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 i
remaining four are oriented in the East-West direction. The brackets are welded to embedments in the wall as shown in Figures 2.3 and 2.4, for the radially and the East-West oriented brackets, respectively.
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 were not normalized. The impact properties of the material
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indicate that it is acceptable for use as a material purchased as impact specified 4 TM 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 j
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 machined flat for an area of 8 x 13.5 inches with a
+
surface roughness of 250 or better. With the RV at a temperature of mi1181-0953a141 2
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 be 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 kai to only 5 ksi as discussed in Sectior 8.
The 5 ksi prestress value reduces to 1.5 l
ksi during the normal operating conditions as a result of increased anchor stud temperatures and other losses. The anchor studs alone will
<[E resist overturning moments and uplift forces from all loads on the l
l reactor except those from the design basis LOCA. The ULS will l
partially function to resist the design basis LOCA overturning moments I
on the studs by limiting the RV displacement.
Shear forces and i
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 l
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.
1 1
l mil 181-0953a141 j
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REACTOR PRESSURE VESSEL ELEVATION 4
(W'
- 7'-0" d
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EL 627'-8" N
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3'-2-118"\\
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l 15'-7-%" OD
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31'-5% "
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I R = 6'-0" l
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FIGURE (2.1)
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REACTOR PRESSURE VESSEL UPPER LATERAL SUPPORT PLAN kUPPER 030' 0*
LATERAL SUPPORT (tyP)
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$ "N s
34' \\
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- 270*
l 90' j..
hd RV EDGE
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CONCRETE
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SHIELD PLUG
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PRIMARY SHIELD WALL p
1 180' 4
m FIGURE (2.2)
REACTOR PRESSURE VESSEL UPPER LATERAL SUPPORT BRACKET DETAIL (Typical 8) i f
s'-6" I:
FACE OF RPV
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T.O.S. EL 632'-3" l
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4'-1%" $ ASTM A-540 BOLTS
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i CARBON STEEL STAINLESS STEEL PLAN AT EL 632'-3" i
ELEVATION
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a FIGURE (2.3) t l
UPPER LATERAL SUPPORT BRACKET DETAIL (Typical 4)
FACE OF PRIMARY SHIM ES SHIELD WALL
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3'-7-9/16" N4 3'-7-9/16" I
FACF OF
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ASTM A 516
_ ASTM A 240 XM-19 L.
4 CARBON STEEL STAINLESS STEEL ELEVATION PLAN AT EL 632'-3" 1
FIGURE (2.4)
REACTOR PRESSURE VESSEL UPPER LATERAL SUPPORT BRACKET EMBEDMENT DETAIL N
m-
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i 1 %" 1'-0" T
BAR 1%" x %"
in"
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1'-6" p,,
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SECTION A-A f
M M B-B ELEVATION 1
l FIGURE (2.5)
U e
1
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s REACTOR PRESSURE VESSEL PLAN 360* 0*
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LATERAL t. 2
. LATERAL E 1 REACTOR SKIRT ANCHOR BOLT i
$ o 'o p
SHEAR PIN N
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SOLE E o
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o; 90*
l 270*
5
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180*
EL 603'-1" FIGURE (2.6)
REACTOR PRESSURE VESSEL ANCHOR STUD DETAIL SEMI-FIN HEAVY k, REACTOR HEXAGONAL JAMB NUT SKIRT R = 7'-4 % "
HEAVY HEXAGONAL NUT PLAIN WASHER l
10"'
(hardened)
WASHER 1" THICK 2-518 " 0 ID x 5" OD a
n
- a JE 3" LATERAL t ASTM A 36
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1'-5 % "
PROJ WITH E i
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k 1'-2%" THD j E i h Tl !!A\\N/
' ASTM A 36 p.
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-2" LATERAL E 5%"
4 SOLE E
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7'-4" 5'-1% "
t
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~ 6 1%"x 1%" x 2'I "
2%" A ANCHOR STUD ASTM A 354, GRADE BD SHEAR LUG
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ASTM A 34 E3%" x 17" ASTM A 36
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b 1
5 i HEAVY HEXAGONAL NUT
' SEMI-FIN HEAVY FIGURE (2.7)
HEXAGONAL JAMB NUT
REACTOR PRESSURE VESSEL SHEAR-PIN DETAIL DRILL 1%" $ HOLE AND REAM TO 2.005" $ IN SOLE E 2"
4 e R = 7'-4%"-*
2.015 g REACTOR
(+0 SKIRT
-0.01) 6 1116 "
5" o
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6-1116 "
!)is 63, p45' CHAMFER 602 -31
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a 3/4" 1/8"
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l ASTM A 354, GRADE BD DRILL 1" $ HOLE FOR REAMER PILOT BAR l
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l FIGURE (2.8) l l
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-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
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and LOCA are combined.
3.2.2 The postulated LOCA shall be assumed under 100% power operating condition.
3.2.3 The effects of jet impingement shall not render the RV support system inoperable.during the postulated LOCA design basis event.
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3.2.4 During the normal, upset, faulted, and te 9 conditions with the exception of LOCA, as stated in 3.2.2 abs.*, the following conditions must be met; a.
Reactor coolant system (RCS) temperature variations resulting in RV radial and vertical expansions will not result in forces being placed on the RV by the ULS.
b.
RCS temperature variations resulting in RV radial and vertical expansions will not result in forces on the RV support system causing the system to be impaired or damaged to the degree it cannot perform its safety design function, mil 181-0953a141
15
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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.
4 j
3.1. 6 The ULS shall be designed such that temperature variations 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-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 supports shall not exceed 200*F during all operational modes.
i
~
3.2.8 The RV support system shall be designed so that a continuous i
40 year total radiation dosage will not result in unacceptable degradition of the support material.
3.2.9 The RV support system shall be designed assuming forced cavity air flow. Forced air flow shall be ensured or appropriate operating restrictions shall be imposed in the event of a loss
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of forced flow.
4 3.2.10 The temperature differences that may exist at different locations on the RV during all normal operating conditions shall 4
l be considered in establirhing the proper gap size at individual ULS brackets in order i.
.tisfy the functional criteria set l<d(
4 forth above.
5 1
mil 181-0953a141
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 1 and 2.
i 3.3.2 Codes and Regulations 3.3.2.1 Rv Support system 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.
1 1.)
American Concrete Institute Building Code
~
Requirements for Reinforced Concrete (ACI 318-71).
i 2.)
American Institute of Steel Construction
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Specification for the design, fabrication and j
erection of structural steel for buildings - 1969 2
l Edition with Supplements 1, 2 and 3.
3.3.2.2 NSSS The design of the NSSS shall conform with, but not be i
limited to the applicable codes and specifications listed in FSAR Table 5-2.1.
In smamary, tue RV and RV skirt shall conform with, but not limited to, the following code:
1.)
American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, 1968.
mi1181-0953a141
17 3.3.2.3 The following ASTM material specifications:
l A516 Grade 70 plate material
- l E7018 Filler metal for shielded metal arc welding A240 Type XM-19 stainless steel A354 Grade BD bolts and A325 high strength bolts **
- Note: See the discussion in Section 2.0 on the A516 material.
- Note: The anchor stud material for Unit 1 are subject to the conditions of the Teledyne reports (References 1 thru 5) l l
since they do not satisfy the requirements for the A354 l
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' c) is 5,000 psi and the reinforcing steel is ASTM A615 Grade 60.
l l
mi1181-0953a141
18 3.3.5.2 The structural steel materials are as follows:
Plates A516, Grade 70*
Filler metal for Welding E7018 Stainless steel 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 p.ovided for section 3.3.2.3.
3.3.6 Design Loads The RV support system will be designed to take loads from:
a.
Permanent weights (DL) b.
Stud prestressing (P )
Operating thermal loads (T,)
c.
d.
Operating basis earthquake loads (OBE) e.
Safe shutdown earthquake loads (SSE) f.
Loads from design basis (LOCA) 3.3.7 Load Combinations and Allowable Stresses i
3.3.7.1 Upper Lateral Supports DL + T,***
AISC specification allowable a.
b.
DL + T, + OBE**
AISC specification allowable
<(f mil 181-0953a141 e
19 DL + T, + SSE**
AISC specification allowable c.
d.
DL + T, + SSE** + LOCA AISC specification allowable x 1.5*
- 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. See Appendix D for the justification of the ductility ratio for use.
Since the RV will not be in contact with the i
~
ULS, the seismic loads will consist of l
permanent weight (DL) inertia loads.
J' Thermal effects on the ULS only serve to reduce both the allowable stress and the Young's i
modulus for the steel.
3.3.7.2 Anchor Studs a.
P
$ 6 ksi( }
b.
DL + T, + P 5 6 ksi and it 1.5 ksi(2) t c.
DL + To + (SSE or P )
$ 0.5 Proof test load (3) g d.
DL + T, + SSE2 + LOCA2' 5 0.7 Proof test load (3) mil 181-0953a141
o 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 t 0.7 P as t
described in Reference 9 and Appendix F.
3.3.7.3 NSSS The acceptance criteria for the NSSS and in particular the RV skirt is covered in the 1968 ASME Code Section III " Nuclear Vessel".
The code does require that, "where compressive stresses occur.
. the critical buckling stress shall be taken into account". When the stress reports are revised for the final loadings from Section 11.0, this requirement will be satisfied.
3.3.8 Parameters to be Considered in the Design The following parameters shall be considered in the design.
3.3.8.1 Irradiation effects:
a.
Gamma radiation heating.
b.
Embrittlement of the structural steel, filler material for welding, and anchor studs mi1181-0953a141
21 3.3.8.2 Temperature variation and heat transfer:
a.
The temperature gradient in the brackets and the i
temperature ac the bracket-concrete interface.
l b.
Effect of temperature variation on the studs l
l pretension load.
3.3.9 Tolerances 3.3.9.1 Construction tolerance in setting the gap of +1/64 inch is allowed.
3.3.9.2 construction tolerance in prestressing the studs of
+0.5 ksi is allowed.
I i
i j
mi1181-0953a141 i
I 22 4.0 GENERATION CF 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 developed.
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 walls structure has been used to predict the
<((
design basis LOCA loadings as a function of variable gap size between a
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 ar-considered:
4 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.
mi1181-0953a141
\\
23 i
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 the ULS, 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 phased phenomena of both asymmetric cavity pressure across the RV shell, and to pressure differentials inside the RV.
These forces are described in more detail in Section 11.0.
4.1.3 Design Basis LOCA Loads and Displacements I
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 si,Ne.
mi1181-0953a141
+
I
24 Also determined was the deflection in the most critical ULS
<[(
bracket. The critical ULS bracket is defined as the bracket subjecced to the largest axial compressive deformations which therefore have the potential to exceed the ductility limits imposed by the criteria set forth in Section 3.3.6.
Figure 4.3 illustrates the critical ULS bracket, 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 The preliminary loadings given in the report have been determined to allow design of the modified RV support system to proceed in an~ orderly manner. Models and forcing functions are similar and/or identical to l
those used to produce final loadings.
l mil 181-0953a141
'I 25 Deadweight and thermal loads are being revised to reflect refinements in the internal walls structure model. 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 loadinga, the major reason for the more detailed analyses is to determine the effects of LOCA on the reactor internals.
mi1181-0953a141
26
- 6. 0..
FORCE IN BUMPERS VS. RADIAL GAP 50 HOT LEG GUILLOTINE COLD LEG GUILLOTINE ----
G m
4.0
~
s e
/
c s
O TOTAL FORCE IN ULS s
d 3.0.
\\
h
\\.
s s
j _ _ 1886 KIPS 2.0.
s s
1886 KIPS
'N 7
s 1.0, FORCE IN MOST
\\g CRITICAL BUMPER
\\g \\\\
0.0 0.0 0.1 0.2 0.3 RADIAL CAP (INCHES) i FIGURE k.1 Force In Bumpers VS Radial Gap i
27 i
l RV DISPLACEMENT AND ANCHOR MOMENT VERSUS RADIAL GAP H0T LEG GUILLOTIfE
- 0. 5' '
g COLO LEG GUILLOTINE ----
~
~ [
90 5=
04-A a
80 s
-m
<d
's' #k 70 e
o
~
M G 0.3.
9.o
~
E c-s-
/
p gm 50 zo
/
p yn -
W*
a as g g 0.2.
40
$8
,j,'
EG
, /'
az
- 30 og G as 0.1-
.20
'Eg 10 0.0 0.0 0.1 0.2
- 0.3 RADIAL GAP (INCHES)'
FIGURE h.2 RV Displacement And Base Anchor Moment Versus Radial Gap
28 17 g
17 73 K
H 6, g Q[
k p
0 I
0 43 47 43 f'
I
'g\\
l dl j
j1 1
'w I CRITI M _\\
CRITICAL BUMPER 0
BUMPER 6 18.5 t
30" f-18.5 0
l i
i j
SUPPORT CONDITIONS FOR SUPPORT CONDITIONS FOR HLG at RV CLG at RV DEFLECTION IN MOST CRITICAL BUMPER VS. RADIAL GAP 0.05-HOT LEG GUILLOTINE
^
g COLD LEG GUILLOTINE ----
s N
/
2 0.04"
\\
M
/
\\
v5
/
\\
5 0.03
/
\\
P
/
\\
G'
\\
y 0.02 I
N g
I N
N 0.01<
\\
\\
0._0_
0.0 0.1 0.2 0.3
. RADIAL GAP (INCHES)
FIGURE h.3 DEFLECTION IN MOST CRITICAL SCGER VERSUS RADIAL GAP Deflection in the most critical bumper is given for the two LOCA cases. Displacements are measured in the axial direction. Any deflection due to bending was ignored. For a CLG (Gap > 0 inches), the critical bu=per was considered active after contact with the 730 bu=per was obtained.
l
i j
29 5.0 ANALYSIS AND DESIGN OF THE SUPPORT SYSTEM 1
The reinforcement provided for the upper lateral support brackets were l <((
l designed based on the preliminary loads described in Section 4.0.
The stresses in the anchor studs were analyzed to show that they are within the allowable limits specified in Section 3.3.6.
5.1 UPPER LATERAL SUPPORT 5.1.1 Upper Lateral Support Brackets - Maximum Design Vs Allowable "I
Stresses and Displacements The stresses at the sections shown in Figure 5.1 were governed by the following load combination, DL + SSEi + LOCA The resulting stresses at Sections 1, 4 and 5 are given in the table below:
Maximum Design Stress Allowable Stresses Or Section Or Interaction Value*
Interaction Value*
1**
9.32 ksi 33.48 ksi 4
0.963 1.000 6
1.000 1.000
- Notes: The interaction of axial compression and bending is according to the AISC specification, Section 1.6.1.
Critical loading on Section 1 is from cavity pressure
<$(
- only, t
Inertia loads due to ULS self weight, and weights supported by the ULS.
1.11181-0953a141
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 11.11 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 maximam allowed displacement. Refer to Section 11.1.2.4 for the discussion on the proper use of Figures 11.11 through 11.13.
<({
5.1.2 Embedments - Maximum Design Vs Allowable Stresses For the most critical load combination (DL + SSE + LOCA), the
,q{
< tresses 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 plate k
b.
Bending stress in 32.4 32.4 embedment plate c.
Tensile stress in 30.4 32.4 anchor bar d.
Bearina stress between 3.6 5.95 anchor block & concrete e.
Bending stress in 23.2 32.4 anchor block j
l l
l mi1181-0953a141 l
W 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 j
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 tue concrete pedestal are modeled i
using linear springs which can be axially loaded only. The stiffness of these springs in tensica 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 rigid links, thus representing the boundary edge effect of che 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 mi1181-0953a141
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 allowable stresses. These stresses were calculated by combining the sttesses from SSE and LOCA loads by the square-root-of-the-sum-of-the-squares 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.
The Level D, or Faulted, allowable stresses for the reactor vessel anchor studs will be 0.7 times the minimum measured test load
<((
resulting from the 1980 detensioning effort (ie, 0.7 x 75 ksi = 52.5 ksi) but if higher allowables are desired, these allowables for individual studs will be based on 0.7 times a "new" test load where sufficient controls have been applied to assure the accuracy of mil 181-0953a141
3 33 the load measured. The criteria is discussed further in Appendix F and Reference 9.
Maximum Design Allowable Stress
- Stud Number Strese (ksi)
(ksi)
Outside Diameter 39 33.2 52.5 38 37.5 52.5 37 44.8 52.5 36
BROKEN-------------
35
BROKEN-------------
34 49.2 52.5 33 44.6 52.5 32 43.0 52.5 31 41.9 52.5 Inside Diameter 39 28.8 52.5 38 30.0 52.5 (g
37 29.7 52.5 36 29.1 52.5 35 30.3 52.5 34 33.5 52.5 33 36.4 52.5 32 37.5 52.5 31 37.4 52.5 For stud number location, see Figure 5.3.
- Note: The.tiowable stresses are obtained from Reference 9 and a;pendices C and F.
5.2.2.2 For Unit 2 The allowable stresses will be determined in conjunction with Reference 9 and Appendices A and F at the time of detensioning.
mi1181-0953a141
i i
34 i
i l
3'-6" = 42" l
@ @E 9"- 8 % ". 8%". 8".
8"
's 1
O 0
l g
u.
l l
i e
t i
I i
i I
p.
p- "'# s
/ -
g
/-
4 j
BRACKET ANALYSIS SECTIONS 1
1 i.
FIGURE 5.1 4
t 4
l
REACTOR PRESSURE VESSEL FINITE ELEMENT MODEL OF SKIRT I
l 90*
[
16" 0
24 SPACED @ 3.75*
- l15" 180a o.
e 15"
~
~
[
_ (( [ [ ~_(([ l[ (( (( [ [
8"
_i!7%"
DEVELOPED VIEW OF RPV SKIRT
- a. 7 (Typical for Other Quadrants)
RPV SKIRT
':EIL-R = 88.75" FINITE ELEMENTS OF 270 RPV SKIRT FLANGE FIGURE (5.2)
O
9 8
36 N
A
@g @@@@@@@@ @@@ OuTS l
gg g
G@
LED 1
INSIDE 10 h
G REACTOR SKIRT 11 i
FAILED Q
-FAILED g
@g@
i INSIDE
- e se@@e*@e*
OUTSIDE e @ g g
i POSITION AND NUMBERING OF STUDS IN UNITS 1 AND 2 FIGURE (5.3)
I 37 6.0 HEAT TRANSFER AND THERMAL ANALYSIS 6.1 IUTRODUCTION_
An investigation into the relative motion 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 l
i investigation included heat transfer analysis to determine the temperature distribution in the ULS, of the concrete behind the ULS and l
the RV shell. In order to benchmark the investigation, the same techniques were usad to predict the conditions of hot-functional testing l
(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 l
l modified during HFT, will be used to finalize the gap required between the upper lateral supports and the RV.
<[f I
l.
6.2 TEMPERATURE AND PRESSURE COND M ONS OF REACTOR PRESSURE VESSEL NEAR l(
UPPER LATERAL SUPPORTS 1
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), 550*F (8% power), 575*F (15% power), and
<(s[
554*F (100% power). This range covers all anticipated normal operating conditions other than heatup and cooldown which will have temperatures
~
i less than 532*F and two upset transients:
- 1) reactor trip (loss of i
l 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 mi1181-0953a141 l
l 38 and then a return to 550*F in 1 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 psi. 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.
Tne outlet nozzles, or hot legs, below the restraints carry the coolant fluid which is at a higher temperature. An analysis has 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 temperatares for the l <({
hot and cold leg are 607'F and 554*F, respectively. The outside surface l
of the RV shell at 34.5 inches from the hot leg nozzle centerline is I
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.
I' 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, _alculations were made with the aid of formulae developed in Reference 10.
If the RV shell extern ' 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 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br />. Since the entire transient is less than 5 minutes, it can be concluded that reactor inlet i
temperatures higher than 575*F need not be considered.
mil 181-0953a141
39 l
6.3 UPPER LATERAL SUPPORTS VESSEL INTERFACE The parameters involved in modeling the RV interface with the ULS are l
l very difficult to precisely predict. Considerations are width of gap (A ), conditions of surfaces, and possible air flow conditions. Because G
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 continuous metal 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.
l l
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 case was also run as a base comparison case.
1 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.
1 mi1181-0953a141
i i
40 i
1
- Note: Based on AT = 570-70 = 500*F, a = 7.45x10~8, R = 96.3125 inches.
6.4 PRESSURE DEFLECTION OF VESSEL The deflection of a cylinder at the outer surface subjected to internal pressure is:
2 2Pa b y = E(b*-a )
Z l
where:
l A = The radial displacement due to i (= P for Pressure, or TH for Thermal) in component j (= RV or ULS) l a = Outer surface radius b = Inner surface radius a = The coefficient of linear therr.al expansion then:
b = 96.3125 in.,
j a = 84 in.,
E = 27.8x10s psi at 550*F.
which reduces to:
P ARV = P(2.202x10 5) in.
l thus:
P ARV = 0.05 in., f9r P = 2250 psid I
I P
ARV = 0.055 in., for P = 2500 psid l
P ARV = 0.0605 in., for P = 2750 psid mil 181-0953a141
41 The closure of the RV hemispherical dome head does affect 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
A
= (7.38x10 s)(96.3125)(505) = 0.359 in.
R where:
AT = 575-70=505'F a = 7.38x10~8 at 575'F R = 96.3125 in.
P Assuming a design pressure condition, A
= 0.055 inch, the closure RV effect(y,addsanadditional0.009 inch.
O
- O V*
V RV R
A
= 0.359 + 0.055
- 0.009 = 0.423 inch at 15% power (maximum)
RV 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.
A
= (7.308x10~8)(96.3125)(462) = 0.325 in.
R mi1181-0953a141
o 42 i
l P
ARV = 0.048 in.
i The deflection due to closure is approximately 0.008 inches. Hence, the l
RV displacement is, ARV = 0.325 + 0.048 + 0.008 = 0.381 inch at 0% power 1
i For the case when the NSSS is at 100% power the vessel is at 554*F and l
2,200 psid is assumed, hence A
= (7.346x10~8)(96.3125)(484) = 0.342 in.
R P
ARV = 0.048 in.
= 0.008 in.
y I
i 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.)
i mi1181-0953a141 l
43 l
Distance From Free Edge of Temperature ('F)
Bracket Metal Node (inches) 0.1 gap 0.03125 gap continuity 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 i
858 153.0 184.6 290.9 859 150.8 179.0 273.4 l
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 l
613 -> 638 0.08 in.
stainless (A240 X M-19) l 638 -> 686 5.92 in.
stainless (A240 X M-19) l 686 -> 861 15.5 in.
carbon steel (A516 GR70)
Node 861 represents radial location where bottom of the ULS enters concrete t
l Location Metal Temperature (*F) 0.1 gap 0.03125 gap 0.0 gap A"*#*8* " **
181.3 252.8 464.2 638 Average nodes 686 167.7 220.6 389.9 l
Average remainder of nodes excluding 152.1 182.4 284 l
959 Thermal growth of the ULS for a 0.1 inc'h gap is'as follows:
i (3 = (8.3x10 s)(111.3)(0.08) + (8.3x10~8)(97.7)(5.92) l l
+ (6.03x10 8)(82.1)(15.5)
= (7.4x10~8) + (4.80x10~3) + (7.67x10~3) = 0.0124 in.
3 Similarly, the thermal growth of the ULS for a 0.03125 inch gap is 0.0182 inches and for r ro inch gap is 0.0385 inches.
mi1181-0953a141
44 6.6 CREEP, THERMAL RACHETING, AND ELASTIC SHAKE DOWN 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 i
near the heat treatment range (i.e., T operation is 575'F while the temperature at heat treat was 1,100 to 1,150*F).
Therefore, additional stress relaxation would not be expected to occur, unless creep effect I
l occurs.
I ASME Section III stress criteria recluires that the highest temperature 1
l 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 SA-508 CL 2 is 700 F.
The reported S,value is 26,700
(
j 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.
stress allowables do not diverge from 1/4o,until a temperature of
)
i 800'F.
Therefore, it can be concluded that the vessel is not operating I
in the creep range.
l Thermal racneting 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, mi1181-0953a141
45 maximum general membrane stress yield strength
- Maximum general membrane stress = 2500(84)= 17.5 ksi 12 x = 17.5 ksi/26 = 0.673 y, _ maximum allowable range of thermal stress yield strength
- y' min = 4(1 - X) = 1.308 o
= 1.308(26) = 34 ksi H range
- Note: For additional precautions against thermal rachet, the following endurance limit should be used:
2xS at 108 cycles for SA-508 CL 2; 2(13) = 26 ksi.
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 eariier, 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:
a = E a AT _ 27x10 (7.5x10*8)(43) = 12.4 ksi 3
0.7 0.7 or a
= 2(12.4) = 24.8 ksi < 34 ksi TH range 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 mi1181-0953a141
n 46 l
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 are below the yield strength of the material (42 ksi),
l there would be a negligible, if any, strain cycling.
l 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 l
unlikely.
i I
l l
l j
mi1181-0953a141 i
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 (1.125) inches wide and the top edge of the flat is located at el 631'6-1/2" ( 1/16") with its bottom edge 8 (!1/16) 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 guarantce a flat smooth surface, and a full area of contact between the 5 x 12-inch stain 1ers 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 mach.ned off the RV was checked before it was removed and found to be with: n the acceptable wall thickness limits.
B&W Construction Company designen and built the tools required for the machining, and the machining of both Midland Units 1 and 2 is now complete.
mi1181-0953a141
48 8.0 DETENSIONING AND TENSIONING OF THE ANCHOR STUDS l
8.1 DETENSIONING PROCEDUPE In detensioning Unit 1, a scatter in the lift-off load values war 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 lif t-off values observed in l
Unit 1.
The criteria and procedure to be used to detension the Unit 2 studs is described in detail in Appendix A.
lk l
8.2 CREEP RECOVERY l
The reactor pressure vessel anchor studs in Unit 2 were tensioned to 92 i
l 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-recovery or delayed elasticity. This creep recovery reaches a limiting value leaving an irrecoverable strain or l(
permanent set.
l This creep recovery, if not accounted for, will increase the tension in I
the studs if they retensioned shortly after being detensioned. For this reason, retensioning will not commence immediately after detensioning.
I Furthermore, depending on the time duration between detensioning and retensioning, the magnitude of the tension load will-be adjusted to l
account for the increase due to creep recovery of the concrete.
In addition, after tensioning and after sufficient time has elapsed, such that_almost full creep recovery has taken place, the tension in the
~
mi1181-0953a141 j._
.Q 4-
.49 studs will be checked using an ultrasonic extensometer device as described in Appendix A and adjusted if required.
8.3 RETENSIONING PROCEDURE The retensionir.g procedure for use on the Units 1 and 2 RV anchor studs is described in Appendix A.
mil 181-0953a141
l 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 penetrat).ons will be fitted with seals to reduce heat losses.
4 i
1 m
I mil 181-0953a141-
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 supports a.
and the RV.
b.
The change in the RV surface temperature, the temperature of the upper lateral support 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 Letween 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.
l mi1181-0953a141
52 l
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
<f horizontal directions are individually summed by vector 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 l
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 l
Thermal growth of concrete wall from 70*F
.046
.046
.046 l
l Effect of Closure
_ 008
.009
.008 l
Total
.347 in
.389 in
.364 in
- Note: A positive sign on the displacement indicates a gap closing motion, and a negative sign indicates a gap opening motion.
< l mil 181-0953a141
53 The following assumptions were made in the above-table. The overall wall temperature during normal operational conditions is ~ a. 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. <[] 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 c.<lculated 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 or (15/32" + 1/64" - 0") mi1181-0953a141
54 11.0 ANALYSIS TO DETERMINE FINAL SUPPORT LOADS I 11.1 GENERATION OF SUPPORT LOADS 11.1.1 Technical Basis The methodology used to generate the design loads for the modified Nuclear Steam Supply System (NSSS) supports will l utilize the same analytical techniques and computer codes as B&W l Topical report, BAW-10131, Reactor Or.olant System Structural L Loading Analysis (Reference 11) and the B&W's Owners Group Report entitled, Effects of Asymmetric LOCA Loadings, BAW 1621 l l B&W 177-FA, (Reference 8) which has been submitted to the NRC for review in July 1980. l 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 walls structure, and boundary conditions at the reactor ( coolant pumps and steam generators specific to the Midland l l Plant. The ceismic forcing functions are Midland specific, l l 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 l for structural integrity under the modified support design. mi1181-0953a141 l
55 1 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. 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 such that pretension of the anchor studs is not exceeded. 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 elastic properties, will be used to establish " 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. 11.1.2.1 NSSS Model Because of the complexity of the RV loading conditions and the number of attachments to the vessel, a detailed isolated model of this component will be mi1181-0953a141
1 1 a 1 1 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 l l isolated model is shown in Figures 11.1 through 11.7. l 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 kv finite element techniques, the structural characteristics of components will be used to generate the equivalent finite-beam element and lumped mass representations. i l The criteria for developing the equivalent structural representation is that component stiffness and frequency must be retained, i l 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 mil 181-0953a141
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 i 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 concrete basemat. The boundary conditions at that point are fixed such that no relative rotation or translation is allowed. This internal wall structure mi1181-0953a141
i 58 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 l cold leg piping at the pury; 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 i piping at the pumps and steam generators consist of t 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. l 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 mi1181-0953a141
59 j~ anchor: studs,, Jealized concrete flexibility, and overall flexibility of the RV pedestal from the RV skirt support to the center of the basemat. In the seismic analysis, these stiffnesses are linear since f the anchor stud prestress is not exceeded. J The LOCA enalysis reflects the nonlinearity of the RV base support during " liftoff" in a series of 1 equivalent nonlinear springs connecting the base of the RV skirt to the concrete pedestal. 1 The ULS are gapped such that they are not active i during a seismic event. During a LOCA, the gap between the ULS and RV would close such that the ULS 4 l becomes an active support. ULS structural properties i are incorporated into equivalent nonlinear springs which reflect the appropriate gap along with the i inelastic properties of the support. Localized i concrete and RV flexibility is included in series with j the ULS springs. The ULS equivalent beams are shown in Figure 11.7 as they connect the RV with the primary a shield wall. i 11.1.2.4 Stiffness of Upper Lateral Supports Lateral translation resistance versus displacement curves for the upper lateral restraints in three 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 cold leg 779 x 10 lb/in hot leg (North-South) 215 x 10 lb/in core flood nozzle (East-West) 723 x 10 lb/in See Appendix E for further discussion on the method- <t ology used to compute the spring rates. mil 181-0953a141
61 11.1.2.5 Stiffness Of The Support At The Base Of The Reactor Pressure Vessel i The moment versus rotation curve for the reactor i pressure vessel base, (rotational spring constants K8 t l and K8 ), is shown in Figure 11.14. The curve was l I developed by a finite element analysis that assumed the nominal prestressing load of 20 kips per stud (corresponding to a 5 kai prestress). A dead weight I 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 lif ted 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 l tabulated below: Direction Spring Rate 10 torsional (K ) 1197 x 10 in-lb/ rad 6 lateral (K or K ) 578 x 10 in-lb/ rad g 6 vertical (K, before lift off) 22') x 10 lb/in 6 vertical (K, after lift off) 171 x 10 lb/in y I mil 181-0953a141 i l
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 g z torsional stiffness K The lateral forces and torsional moments 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 the 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 will be considered in detail; a guillotine break at the hot leg outlet of the RV and a guillotine break 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 l( larger loop model of 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 1
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 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 bcsemat. 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 t 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 The vertical respor.se of the reactor 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 this response phenomena. The nonlinear structural mi1181-0953a141
-64 i i 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 histories on the reactor vessel in the system dynamic analysis. The core bounce LOCA forcing functions are the result of the worst case double end guillotine pipe breaks at the RV nozzle. Thermal Hydraulics and 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 coJel. mi1181-0953a141
1 65 l 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 varyi.ng 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 b'reak. 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, i
I mil 181-0953a141 l
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-dimensional, finite segment systems ccasisting 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 loadings. 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 ar. 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; structural analyses including static elastic, plastic, and creep, dynamic, and dynamic plastic analyses, and large deflection and stability analyses; and one-i dimensional fluid flow analyses. Data Reduction Codes 1. FTRAN - A computer code used for Fourier analysis of forcing functions to determine the frequency l content of the forcing function. 2. S1235 - A post-processor program used to tabulate l forces, moments, displacements, and rotations in a specification format. 3. INTFCE - A program used to convert pressure-loading data to force-loading data acceptable for i use by the structural analysis codes. 4. LOPL - A post processor program used to provide time history tabulations and plots of spring forces and resulting loads and displacements. mi1181-0953a141
68 11.1.5 Seismic Analysis Utilizing the geometric and structural properties cf 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 elcment or bar in the model is assigned a damping value based on the location and type of component the element represents. Strain energy camping 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 individual earthquake will be combined by the SRSS method as described in the response to 1 Regulatory Guide 1.92 in Section 3A of the FSAR. The resulting l member loads and displacements will be combined by taking the i SRSS of all five earthquake excitations (three translational and two rotational). Figure 11.10 shows the flow diagram for the seismic analysis. RV Lower Support Ioads 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 skirt. These centerline loads are resolved into support loads l mi1181-0953a141 i
7- - O Q 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-JOINT 50 FIGURE 11.1 FORCES (KIPS) MOMENTE (PT-KIPS) LOAD CASE FX FY FZ MX MY MZ 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 98.8 119.6 3806.2 120.6 4334.1 DISPLACEMENTS RV PROFILE AT ULS JOINT 166 FIGURES 11.1, 11.4 DISPLACEMENTS (INCHES) X Y Z SSE (Combined) .02445 .00148 .02156 OBE (Combined) .01284 .00074 .01124 i WALL PROFILE AT ULS EL. 631'5-1/2" JOINT 170 FIGURE 11.4 DISPLACEMENTS (INCHES) X Y Z SSE (Combined) .03261 .00144 .02854 1 OBE (Combined) .01891 .00072 .01711 _U_LS Loads There is no interaction between the ULS and the RV during a seismic event. mi1181-0953a141
f 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. mil 181-0953a141
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--EfB / EID- -E!D \\ 20 cc. 6zo*- fo CiUiH) st. eu'.o-l '@ y 5 ggy Y W> \\l ,/ o stessarueAt nar -giD \\ / o aa s s.ioiar %_/ O Jot 4T NuM6 M REACT 02 VESSEL uauser xuaesa S et c,os'. o 4 et.so;o" SUMPER y>,,,,,,,,_,, ELEVAT/0Al g a m.is e --Eis! y, 4 a. ssi-s, --5E EiD- @ et. ser;o-dG~~~T -- RV 150Lh,TED w. rat rean. i
- 1,S $ W[-
~~J~7 "~ q MODEL VERTICAL %%LL
- ]' =p""l (REACTOR \\JE6SEL
) -.;. _. !.2f. i5 ',ro*oorz/es SUMPER ELE %TIOM ..i ~ ~ ' ~ ~ 5. FIGURE 11.7 r (tocq . ami, g iI22DI7 C i o s
o 78 a I CONTROL R0D DRIVE SERVICE / STRUCTURE i I?1 ts /
- 7 CONTROL ROD GUIDE TUBE
,' u (COLUMN WELDt1ENT) PLENUM 7 / Ql .) PLENUM ASSEMBLY / 3 UPPER / I I CORE SUPPORT GRID u CYLINDER [ FUEL l . 1P ASSEMBLY T s CORE m. 9 l THERMAL LOWER I SH.IEL.D GRID i s I I LOWER GRIO f SUPPORT FORGING _ ~ f jju~ GUIDE LUG DISTRI UTOR N SKIRT SUPPORT . \\j '( i FIGURE 11.8 Reactor Internals and Service Support Structure
79 [K) c' i [K) i 1 i [K 3 v k y [K) w [K] d 1 _1 t [K3 j (K} seismic only 4 [K] = Stiffness matrix FIGURE 11 9 Reactor Coolant Syster Boundaries i
i 80 i SEISMIC / LINEAR LOCA DEVELOPMENT OF HYOROMASS. HYDROE MASS MATRIX 1 NONLINEAR LOCA STALUM DEVELOPMENT OF 3 GE0M STIFFNESS & ANSYS MODULE FLEXIBILITY MATRICES STALUM FREQUENCIES & i LUMP MODE SHAPES MODULE STALUM LUMP EQUIVALENT STATIC FORCES MODULE 4 i STALUM RESULTANT LOADS i RSTA (DETERMINED MODULE STATICALLY) FIGU3E 11.10 Utilization of Computer Programs
O e 81 BRACKET RESISTANCE VS DISPLACEMENT CURVE HOT LEG (North-South) DIRECTION d AT GAP BETWEEN 0.090" TO 0.125" 6,36 - - - _ _ _ _ _ c _ _ _ _ _,,,,_ _,,,,_,,,, l MAX ALLOWABLE LOCATION VARIES 5,000 d WITH GAP l4 DEFLECTION FOR K l DUCTILITY RATIO K = 10. ^ 4,000 a b .6 3,618* l l 3,000" I l l a 2,000* l l l l l l l 1,000" l l l 1 l 1 0 e i 0.00001 0.10000 0.20000 0.2662 0.02662 0.23134 Displacement (in) I 6 = GAP f ) cOs 43 cOs 17 FIGURE 11.11 i 4 0
82 BRACKET RESISTANCE VS blSPLACEMENT CURVE DECAY HEAT REMOVAL CORE FLOOD (East West) DIRECTION a AT GAP BETWEEN 0.090" TO 0.125" 6.344" c__________ O 5,000 H P EFL T ON F R U ~- DUCTILITY RATIO l = 10. d 4,000* a b 3,772" l l l ~~ 3,000* l l l 3 l l l 2,000" l g 1,000* I I I I I I O i i i 0.01000 0.10000 0.20000 0.254e0 0.02548 0.21756 l Displacement (in) l I 6 = GAP ( ) cOs 47 cOs 18.5 FIGURE 11.12 l
BRACKET RESISTANCE VS DISPLACEMENT CURVE 85 UPP R COLD LEG AT GAP BETWEEN 0.090" TO O.125" MAX ALLOWABLE DEFLECTION FOR DUCTILITY RATIO = 10. (Pd. Ad) d K 6,019 l 5,000" (Pc, Ac) I l K I 4,000 8 I I a 1 3,000 l l JP K e l l i 2,000* l l l l l l 1,000 l l l K (Pb, A b) l l g a 0.1$000 0.20000 AMAX 0.02000 Aa A1 62 A3 0.03000 / AT POINT A a " W (COS 43 ~COS 11.5 AT POINT B l" ^ I" 0.04734 i ~ COS 48,5 COS43 A b "O +0i P = A P, a b AT POINT C (1886-A P,) AP2" 5 P, = A P2+AP, AP2 A2= X 0.01893 4125 A,=Ab+0 2 AT POINT D A3 = 0.02720( ) 5 d = 6019* A d"O P c 3 A MAX = A, + (A, + A ) x 10 2 RANGE FROM 0.34862" TO 0.3833" FIGURE 11.13
l l Sh 1 1 Rotational Spring Constants slope = 121.0 X 10'Oin-!birad KO & KO at the Base of the RPV x z 7 M X 10 (in -Ib) 35 - O'y io slope = 102 X 10 in-Ibirad (b) / NOTE: These spring constant curves arc / applicable for Unit 1.* * / They are based on 5 ksi 25 - (c) f nal stud prestress level and j a total number of 93 anchor studs. j Dead Stud
- M
= 20.16 LO Weight Prestress O' l Equivalent preload per stud: 27.9k + 20k = 47.9k 15 - l If the final stud prestress level varies from 5 ksi, simply change (a) l M linearly in proportion to the total prelo4d to determine the Lg l new lift off moment and Point O'. Draw lines parallel to Points (b) and l (c) from Point O' to complete the new spring constant curves. I (a) Before anchor studs lift off 5-l (b) After anche: studs lift off (upper bound) slope = 574 X 10'Oin-Ibirad I ? X 10-4 (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
.1 l 85 i i 12.0 : CEC!!NG SYSTEMS AND SUPPORTS FOR THE RESULTS FROM FINAL ANALYSIS All attached systems, components,'and component supports will be evaluated for the results of the aforcimentioned analyses. The current forecast date for completion is anticipated in the spring of 1983. mil 181-0953a141 .c
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 ksi and the 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. 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. mi1181-0953a141
87 s 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 anch,r 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 i j analyses described in this report, the Company will submit them as mi1181-0953a141
88 amend;.ents of this report to tha NRC. The final analytical results and design details will be incorporated, as necessary, by amendment into the FSAR. mil 181-0953al41
89 15. References i
- 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, Inv?stigation 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, Continuel Investigation of the Failure of Midland Unit 1 RV Ancher 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 I Loadings", Phase II Analysis, July 1980. i mi1181-0953a141 i i
90 9. " Letter deport - Teledyne Engineering Services (TES) Project 5355: Expanded Criteria for Acceptability for Service of Midland Unit 1 RV Anchor St ess" W.E. Cooper letter to H.W. Slager, dated October 6, 1981.
- 0.
Thermal Stress Techniques, The Franklin Institute Research i Laboratories, American Elsevier Publishing Company Inc., 1965. 11. Babcock & Wilcox Topical Repart, BAW-10131, Reactor Coolant System Structural Loading Analysis, November 1976. l l 1 i l mil 181-0953a141
a APPENDIX A: Procedure for Detensioning and Retensioning the Reactor Building Reactor Pressure Vessel Anchor Studs i mi1181-0953a141
\\ 4 APPENDIX A ~ PROCEDURE FOR DETENSIONING AND RETENSIONING REACTOR BUILDING 4 REACTOR PRESSURE VESSEL ANCHOR STUDS 1.0 SCOPE This nonendix ' Provides a procedure for detensioning and retensioning the reactor pressure vessel (RPV) anchor studs 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 2 will be used for analyzing Unit 1 anchors which have already been detensioned. j 1.1 The Procedure includes the following: }. a. Ib detension and retension 96, anchor studs for the Unit 2 RPV and to detension and retension the 93 remaining anchor studs for I the Unit 1 RPV. 1 i b. Working with a bolting technology consultant who will supervise the use of an ultrasonic extensometer to I monitor deformation in the anchor studs 2.0 CUALITY STANDARDS I The work shall be performed in accordance with i requirements of a quality assurance program approved i by Midland Project Quality Assurance Department (MPQAD). 30 Intentionally Left Blank 4 h.0 Intentionally Left Blank A-1 l 4 .1
I 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 MATERIALS Replacement of nuts and washers, if required, shall conform to the purchase specifications. A A-2 I 1
l l 6.2 LCADING FRAMES Loading frame material shall be as noted in t Appendix 1, Page 1-6. 6.3 HYDRAULIC RAMS i 6.3.1 Rams for Detensioning l The two hydraulic rams to be used fo'r ~~ the detensioning frame shall be a solid plunger type with 2-inch atnimum stroke and a capacity of 100 tons each, such as Model RC-100-H-5.7 by Duff Norton. The base diameter shall be a maximum of i 7 inches and a maximum closed height of 8 inches. The rams shall be a matched pair calibrated and certified, 3 traceable to the National Bureau of Standards (see Section d.1 for calibration procedure). i 6.3.2 Rams for Retensioning i i The hydraulic ram for the retensioning frame shall be a hollow-core type with 2-inch minimum stroke and a capacity of 20 tons such as Model RCH 202 by ) Enerpac (see Section 8.2 for calibration procedure). The minimum - { internal diameter of the core shall be j 1-1/16 inch. The maximum base diameter shall be 6 inches and the maximum closed height shall be 8 inches. l - 6.4 HYDRAULIC SYSTEM ACCESSORIES l 6.4.1 Pressure gages shall be test system gages, 8 to 10 inches in diameter, 1 measuring 0 to 10,000 psi and shall be l f graduated in 50 psi maximum increments with 4 25 psi increments preferred The j 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 l rapid losses of pressure when the equipment is assembled. The main lock-i 1 A-3
off valve shall be a manual, 3-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 diameter ASTM A 354 bolts 7 feet, 4 inches long. 6.6 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. d. Detension Unit 2 anchor studs ) e. Take unstressed extensometer readings on Unit 2 anchor studs j f. Check calibration of detensioning equipment g. Retension Uni
- 2 anchor studs and take extensometer ret. dings b.
Check calibration of retensioning equipment I. Detension Unit 1 anchor studs A-4
i ] i ] j. Recheck calibration of retensioning equipment i ~~ k. Intentionally left blank 1. Check calibration of detensioning r equipment-1 m. Retension Unit 1 anchor studs and take extensometer readings ~~ l 4 n. Recheck calibration of retensioning equipment j j 8.0 CALIB2ATION i. 8.1 RAMS AND PRESSURE GAGES i ) Rams and pressure gages are to be calibrated as described below. Calibration or 1 recalibration shall be done as shown in the j sequence of work or at maximum 30-day intervala. i a. Calibrate the pressure gages in the pressure range of 500 to 10,000 psi. j _ Ensure that the pressure indicated by the gage is within +0.5% of the true i pressure in the 2,000 to 10,000 psi i pressure range b. Mark each ran and pressure gage set so i they are easily identifiable as a set. j These sets must be calibrated and used in the field as a set. Pressure gages shall not be switched between sets. If )i ~ a gage requires repair or replacement, the set must be recalibrated unless a 1 gage with a calibration curve matching i the first is available. (Notify i. project engineering before proceeding j with recalibration.) c. Before calibration, exercise rams three j strokes 0 to 0.9 of full extension at ] 30%,- 50%, and 80% of ram capacity. d. As a minimum, ram pressure calibration i data points shall be taken at the ] following increments: I 1 A5
Pressure Range Pressure Increment (psi) (esi) O to 3,000 500 3,0GO to 10,000 1,000 or maximum capacity of the ram 1 Three sets of load versus pressure readings shall be taken for each ram j pressure gage combination. If the measured loads at a particular pressure level in the three sets of data deviate by more than +1%, additional sets of data shall be taken until consistency is attained. e. Rams are to be calibrated at an extension of 1 inch and must be calibrated in the active mode with the ram actuated by the pump and forcing load on the test machine i f. All calibration measurements are to be 1 traceable to the National Bureau of j 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 certification of calibration and any other reference standards used in the calibration shall be submitted to project engineering and MPQAD with J the ram and pressure gage calibration data. f g. Recalibration shall be performed in a similar manner as described in Sections 8.1, Items c, d, e, and f. If the ram pressure gage combination recalibration readings are found to deviate more than +1% from the measured load, the project enginc.ar shall be notified immediately. 8.2 RETENSIONING EQUIPMEIR The retensioning ram and hydraulic pressure gage assembly shall be calibrated with a universal i testing machine or other calibrated standard. The system shall be calibrated before tensioning A-6
i 4 Unit 2 anchor bolts and again before tensioning Unit 1 anchor bolts. The system calibration shall also be checked after completing Unit 1 tensioning. The calibration can be rechecked using the frame shown in j Appendix 1, Page 1-9. 8.3 CALIBRATION OE' EXTENSOMETER I 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 engineering and MPQAD fcr review. Records shall be maintained indicating all pressure 1 readings against st=ad=ed pressure gages, load I readings against standard loads, and extensometer readings against loads applied by a calibrated stapdard. 9.0 UNIT 2 DETENSIONING 9.1 PREPARATION 4 Prior to taking any readings or setting up detensioning equipment, all threads and stud j ends shall be cleaned to facilitate removal of the nuts. The studs ends shall be inspected j by Raymond Bolting Services for conditions i i which could affect extenseneter readings. Methods of cleaning, acceptability of j 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, j they shall be removed by procedures acceptable to the RPV manufacturer. l 9.2 INITIAL EXTENSOMETER READING 3 i I When preparations have been completed, two complete sets of extensometer readings shall be taken on the studs in their present, o j tensioned state. These readings shall be l taken according to the extensometer manufacturer's instructions. One complete set i 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
tolerance to be determined by Raymond Bolting Services), then the readings shall be repeated until agreement is reached. These readings i shall be recorded against the Teledyne stud numbering system. (See Appendix 2 for the numbering systea.) 9.3 LIFT-OFF READINGS When the initial extensometer readings are " ~ complete, the existing preload forces in the Unit 2 anchor studs can be measured as follows: The detensioning frame shall be set up as shown in Appendix 1, Pages 1-1 through 1-4, starting with stud 37 (Teledyne numbering system). The cross beam and ran support blocks shall be installed first (see schematic in Appendia 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 instructiors. The stud coupler shall then be placed ove:- the transducer and connected to the stud. During this operation, care shall be taken so the transducer is not dislodged or the cable from the transducer is not damaged. When the cross beam and stud coupler are installed, the hydraulic ram support blocks can be placed on the RPV flange. The blocks shall be level and, if necessary, shall be 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 placed directly under the cross beam centerline. The rams shall be installed so the hydraulic hosas are free of sharp kinks and do not rub against sharp corners. The 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 process. When this procedure is complets, the upper nut on the stud coupler shall be brought to a fingertight condition against the crossarm. The displacement gages shall then be installed as shown in Appendix 1., (Page 1-7). The gage support shall be firmly attached to the RPV by A-8
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 determined 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 int hes past che 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, L 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 and the extensometer shall then, pressure gages 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 i lift-off occurred at levels as low as 216 kips. Any stud for which lift-off occurs below J 300 kips must be proof-loaded to 300 kips or j two-times the maximum 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. I.ength displacement gsgo 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 i lift-off readings on the first six studs shall be forwarded to project engineering within one 4 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 been completed as described in Section 9.3, detensioning may begin. The detensioning frame is to be used as described in i Section 9.3, including the nut socket ring. The studa shall be deten=ioned 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 1 previously recorded lift-off pressure. The socket ring is provided to turn the nut when i lift-off pressure is reached. The nut shall l A-10 i
then be retracted approximately 1/4 inch and l 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 chall be repeated until all Unit 2 studs are detensioned. 9.5 EXTENSCMETER READINGS When the Unit 2 studs are detensioned, a complete set of extensometer readings shall be taken as follows: a. Verify that all nuts are loose by 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 an described in Section 8.1. Records shall be maintained as described in Section 8.4. 10.0 RE'1TNSIONING UNIT 2 ANCHOR STUDS 10.1 PREPARATION The anticipated deflection in the studs based upon the stress valge of 5 ksi in the stud tensile stress area of 4 in and which is equivalent to a load of 20 kips is given by: A-11
=P Lt L2 + E A A 3 where L3 = 68 inches (the stressed length of tM i unthreaded body of the bolt) L2 = 13.25 inches (the stressed length of the threaded bolt) P = 5.6 hai x 4.00 square inches = 20 kips A1 = 4.9 square inches A2 = 4.0 square inches E = 29 x 10 ksi This deflection equals 0.0119 inch for the standard stud. The anticipated readings to be obserred 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 0.00 0.0000 88.0000 4 1.00 0.0053 88.0025 a 17 k.25, 0.022k' 88.0104 18 4.50 0.0230' 88.0111 20 5.00 0.026h 88.0123 1
- The talculated stretch in the coupling stud and i
coupler are included. The required gage pressure for the retensioner to develop the specified load of 20 kips shall also be determined from the ras and pressure gage cali-bration data. ) o 9 A-12
10.2 RETEISIONING The studs shall be retensioned in the same order of i that shown in Appendix 2, Pages 2-1 through 2 h. J j The retensioning frame shall be installed as shown in Appendix 1, Page 1-8. First, the extensometer transducer shall be installed according to the 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 net to dislodge the transducer or to damage the cable from the j transducer to the readout unit. After connectdag 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 4 are perpendicular to the pull red. Care must also be taken to ensure that the frame does not rest on adj.acent washers or on the RPV skirt fillets. When alignment is acceptable, the top nut of the pull red shall be brought !J to a snugtight condition. When this is done, ji the ram extension shall be approximately 1 1 inch. Upon completion of the setup, hoses 4 shall be checked to ensure that pressure gages i are visible and no kinks exist. The displacement gage shall be installed securely j. to the RPV and as shown in Appendix 1, Page 1-8. Retensioning may then begin as 1 follows: ) a. Record the readings on the displacement gage and the extensometer l l b. Gradually pressurize the system until ) it reaches the pressure equivalent to the specified loading of 20 kips per l stud on Unit 1. For Unit 2, the retensioning shall con:mence 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 40 days after detensioning, the specified loading shall be 18 kips. If more than 45 days, the specified loading shall be 20 kips. w 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 iS% of the readings taken in Item e above which correspond to the specified load- } ing. Record the displacement gage and extensometer reading. If the values are not acceptable, repeat the retensioning of the particular stud. b i 4 \\ 4 A-lh
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 Appendix 2, pages 2-1 through 2-4. 1 First, a complete set of extensometer readihgs shall 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 3 3 (AEC) where P = the calculated actual load existing in g the stud AE = the measured stud extension (difference 3 between remeasured length and initial length at zero load) aEC = the measured stud extension obtained with the 20 kip load in 10.2c If P4, equals 20 kips +10L 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 P shall be submitted to Project Engineer-A ing. Project Engineering will 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 will also be given. These data vill then be used in the following procedure. j The retensioning frame shall be set up as described in Section 10.2. Initial readings in the extensometer and displactnent gages shall be recorded. By dialing in the stud reference length, the initial extensometer extension reading should equal 4E 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
i 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 slope at P, and the change ~, A in length gage reading from P to P = 20 kips 3 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 i 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 on the I! nit 2 RDV and a.minie of 30 days after all of the studs have been found acceptable and the jam nuts installed on additional set of two extensometer readings on each Unit 2 anchor stud shall be taken. These readings shall be reported l 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 EXTENSOMETER 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. i 11.3 DETENSIONING When the initial extensometer readings are taken, the jam nut can be removed from the A-16
stud and the stud can be detensioned. This shall be done in the order shown in Appendix 2. Detensioning is to be done using the retensioning frame without the ~ displacement gages or extensometer installed. 12.0 RETENSICNING UNIT 1 ANCHOR STLTS 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 commencing 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 1 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 engineering before the start of retensioning. These proof test requirements will be determined upon 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 l
1 l 4 l complete set of extensometer, 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 for studs which have a turned down shank with a gross area of 4.0 square inches. These studs have a centerpoint for machining. Displacement Load Stress cage Reading UT Length j (kips) (ksi) {in.) (in.) 0 0 0.0000 88.0000 l 4 1 0.0057 88.0029 20 5 0.0285 88.0144 a 12.6 RECALIBRATION After completing Unit i retensioning, the ram and pressure gage calibration shall be 2 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. l 13.0 DOCUMENTATION On completion of the retensioning, a report shall be submitted to the project engineer listing the' I pressure, displacement gage, and extensometer readings ~occuring at: a. Unit 2 lift-off b. Unit 2 detensioning (each step) and proof-I testing j c. Unit 2 retensioning i A-ld i
d. Unit i detensioning and proof-testing e. 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 previous ~~ detensioning operation in Unit 1 and to certify that this tensioning operation ascertains that the anchor studs are at the specified load level. >= Y A-19 l
( tj o a e e 9 h APPENDIXES a e f J 4 l }
APPE! DIX 1 DETENSIONING SYSTEM = 1 d HYDRAULIC R AM \\ PUMP 3.WAY NEEDLE VALVE PRESSURE GAGE VALVE ~._ a 4 HYDRAUL IC RAM g / a w PUMP 3.WAY NEEDLE VALVE PRESSURE G AGE VALVE l l l ( l Al-1 i i
t 't iw i 4 e RETENSIONING SYSTEM i 4 I I, 6 I J i 1 HOL LOW-CORE Y _gis 1 HYDRAULIC RAM \\ ~ ]> PUMP 3-WA Y NEEDLE VALVE PRESSURE G AGE VALVE i, k 4 i 1 l I I i i. j i, 4 J 5 Al-2
Jui. n-,m.6 m aA _4h.a..O4A --.Am M 4 = b ,%h. -*-+=**de h r_ RV SKIRT FLANGE OUTSIDE DIAMETER f ___ g -- h ~[ --- - \\ g / 5 l { l l \\h i) Q ) \\ \\ ,J ps u ~ X.T ( ).. sed,exe. ,Vsx,.T w OUTSIDE DIAMETER /Mr'p4 i7 wN [ \\ /' / \\, tw J)%/)Myy= / x--y- -- - p- -,_ _ r wasnER ouTsiDE DiAMETEa caoss BE^u av sx INSIDE DIAMETER RV STUD COUPLING STUD n VIEW A - E: DETENSIONING DEVICE PLAN VIEW 1
r r,, I / 9 $/& MN$ / '/ // /N [/ / / ' ' ' " ' ^ ' ' ' " ' " " ' ' " " ' // / CROSS BEAM p / / /'/ Couetina / kl,/ STUD 77 / / /p / k / /#////// ' //////// / E_ 'N h / /;f x/ Courtino nut s . t,- c TRANSDUCER ) / f> v}y O O / l l CV To T'l - SuePORT BLOCK N / \\ / RV NUT & STUD VIEW A DETENSIONING DEVICE Al-h
RV SHELL RV SKIRT I I I y/ n-9/ / // / / If kI CROSS BEAM (CUT AT CENTER LINE) / / / = / / / / / c / COUPLING STUD / i -J / / / / COUPLER NUT /, ,/ / RAM /, / SUPPORT BLOCK ry v, RV STUD & NUT J L I VIEW B i DETENSIONING DEVICE n-5
0 e e D e e THIS SHZET INTENTIONALLY LE"" BLANK. e ..e O i 4 Al-6
-O e RETENSIONING FRAME l 9" - 4.%" HOLE FOR 1" DIAMETER 6 a a 8.UN-28 THREAD l g B-B TAP ENTIRE DEPTH I l 3-B B nr-I h l (b l l l / / MATERIAL TO BE / ASTM A 194, Gr 2H l 4-%" I 3-%" / HEXAGONAL 3-7/S" ACROSS 3 6 x ' 11/8" DlAMETER ' "'s / / PLAN HOLE A " o A h 3/8" HOLE 3-%" DEEP FOR 2 %" DIAMETER ~ !ll 1" 4-UNC-28 TH READS TAP 2 %" DEEP o 10 %" A-A COUPLING NUT T MATERI AL TO BE 1" DIAMETER 8-UN 2A THREADS ~_ ASTM A 490 A 354 l" ALL PLATE MATERIAL ANY GRADE TO BE ASTA A 36 ELEVATION _r_ RAM SUPPORT FRAME COUPLING STUD Al-T
o DETENSIONING DEVICE ASSEMBLY SEQUENCE 1) 2) I 1 (qp oSo oei i -BLOCK BEAM AB0VE STUD -INSERT & BLOCK STUD COUPLER lNSERT RAM SUPPORT BLOCKS - ATTACH UT TRANSOUCER -lNSERT NUT SOCKET RING 3) 4) 'I 3qr MFM g m O O m ED rT ATTACH COUPLER UNBLOCK BEAM, REST ON RAMS lNSERT RAMS MOUNT DISPLACEMEN',' f':.~JICATORS & LEVEL BEAM, ATTACH TOP UNIT IO
- 5) FOR INSIDE STUD RING LENGTH GAGE -
k N RPV SKIRT , LEVEL GAGE g g M LENGTH GAGE ___a -PLACE il ON TOP UNIT & FASTEN WITH JAM UNIT MOUNT OlSPLACEMENT IN0lCATOR MOUNT DISPLACEMENT INDICATOR OFF ll ON TOP OFF COUPLER Al-8
i i m-1" d STUD MATERIAL ll l i i i i l i I I l l _ HOLLOW CORE l l 20 TON RAM 1 I I I I I I I l' 'l I i i COUPLING STUD I COUPLING NUT ULTRANSONIC TRANSDUCER ^ RAM SUPPORT FRAME \\ \\ RV STUD & NUT RETENSIONING DEVICE P Al-9
1 RECALIBRATION FRAME 1 %" DIAMETER SOCKET FOR ATTACHMENT i / TO TEST MACHINE ) I I I l i i I "I i i l I L_J l i 1 1" DIAMETER ASTM 2& RAIN A 325 BOLT GAGES / BOLT 20-TON HOLLOW CORE l l l ~ BAR 2-%" x 6" WIDE l l l [_l l ASTM A 36 MATERIAL f I f l I l U NOTE: CALIBRATE FRAME AND STRAIN GAGES IN TESTING MACHINE BEFORE USING TO CALIBRATE 20. TON R AM. i 1 SCALE 3"=1' 0" Al-10
l Appendix 2 TABLE 1 l DEEISIONIllG AND RETENSIGNING SEQUENCE REACTOR VESSEL ANGOR STUDS Bolt Number ) Sequence B&W Teledyne 1 Ol in 37 in l 2 02 in 13 in l 3 03 in 01 in 4 04 in 25 in i. 5 01 out 37 out 6 02 out 13 out 7 03 out 01 out l 8 04 out 25 out i 9 05 out 43 out i 10 06 out-19 out j 11 07 out 07 out 12 08 out 31 out 13 05 in 43 in i 1 14 06 in 19 in 1 l 15 07 in 07 in l 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 1 21 09 out 40 out 22 10 out 16 out 23 11 out 04 out 24 12 out 28 out j 25 13 out 46 out 26 14 out 22 out 27 15 out 10 out i 28 16 out 34 out 1 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 l
Bolt Number Sequence B&W Teledvne 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 ~ i 49 25 in 41 in 50 26 in 17 in 51 27 in 05 in 52 28 in 29 in l 53 25 out 41 out 54 26 out 17 out 4 55 27 out 05 out 56 28 out 29 out i 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 1 63 31 1n 11 in } 64 32 in 35 in 65 33 in 39 in j 66 34 in 15 in 1 67 35 in 03 in i 68 36 in 27 in 69-33 out 39 out 70 34 out 15 out i 71 35 out 03 out {' 73 37 out 45 out 72 36 out 27 out 74 38 out 21 out 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 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 i 89 45 out 48 out 90 46 out 24 out 91 47 out 12 out 92 48 out 36 out i A2-2
I l 3elt ma-5 er Sequence B&W Teledvne 93 45 in 48 in 94 46 in 24 in 95 47 in 12 in 96 48 in 36 in l l i l D A2-3
e N A @@@ D@ @ g@@@@@,h ourSiDE b@ FAILED INSIDE 4o @ 11 @ G REACTOR SKlRT @g FAILED h@@ FAILED g g iNSiDE g@@@@@@@@@@@, OvT IDE POSITION AND NUMBERING OF STUDS IN UNITS 1 AND 2 A2-4
p APPEhTIX B: Procedure for Gap and femperature Measurement at the Reactor Pressure Vessel Upper Lateral Support mi1181-0953a141
{ APPENDIX B PROCEDURE FOR GAP AND TEMPERATURE HEASUREHENT l AT THE REACTOR PRESSURE YESSE1 i I UPPER 117ERAL SUPPORT l q 10 SCCPT Ihis appendix provides for the installation of equipment and the implementation of procedures necessary to measure the following while 1 the nuclear steam supply srstas (NSSS) undergoes hot 1 ~ 1 l functional testing (HFT) (see Appendix 1, Page 1-2 for nomenclature): i 1 The change in gap between the reactor vessel gpper lateral support (ULS) and the reactor 2ressure vessel (RPY) l h. The corresponding change in ULS l RPV surface and concrete surface temperature i near the embedment. I j Ihis appendix does not specify the exact design of the electrical system (transducer, thermocouples, I wiring, readout system, and other accessories) 4 required to obtain and record the data. Ihe intent l of the appendix is to provide a systes complete ) with all accessories sufficient to obtain the required data. 1.1 ITEMS INCLUDED 1 111 gupply, calibration, installation, and operation of thermocouples, distance-reading ievices, and resdout units. gapply of all associated viring and other miscellaneous iteas required 2or obtaining seasurements i 1 1.2 Supply, installation, and removal of temporary shia pack replacement material J.1.3 Collection of data from all instruments covered by this appendix. 4 ) B-1
1 1.4 Eemoval of all instruments and associated equipeent and wiring upon completion of HFT 12 Intentionally left blank 4 i j C.0 Intentionally Lef t Blank 3.O QEALIII.,ST A It D A Egg, 111 procedures, calibrations, and sessurements shall be in accordance with.the requirements of a quality assur-ance program approved by the Midland Project Quality Assurance Department (MPQAD). u.0 REFERENCED COD ES L MD STA ND& EQS, leerican Na tional Standards Institute (ANSI) MC96 1 laetica n Society for Testing and Materials (ASTM) B29-79 Standard Specification for Pig Lead i 10 SUBMITTALS Ihe instrument supplier shall provide a complete descrip tion of all instruments, including method of operation, calibration, eff ect of taaparature, changes, wiring, accessorias, and power required in its proposal for approval by project engineering. 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
f,. 0 EQ3EIX2.,,CO N DIT I O N S f,.1 SENERAL York covered by this appendix is to be done during HFT for the Consumers Power Company Midland Plant Units 1 and 2 in Midla nd, Michigan. Hot functional tests are scheduled as follows: Unit 1 May 1 to June 26, 1983 Uni t 2 January 26 to Barch 22, 1983 Instruments for the gap and temperature neasurement 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 1007.. Ihe recorders or readout 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 mariana temperatures are 50 and 120F, respectively. 20 CENER A L P EO UTR EEEEIS, 21 REFINITIONS A) Sas11 distanca transducers are g means of converting the physical movement of an object into an electrical signal. Eor purpose of this specification, the distances to be seasured are less than 1 inch. Ihe instruments are to;be capable of resolving a novement of 0.001 inch. Ivo types of transducers are mentioned in this appendix: the eddy-current noncon tac ting type and the contacting irpe. Ihe eddy-currant noncontacting type works by producing magnetic fields which induce eddy currents in the B-3
adjacent ta rge t as terial. Changes in distance from the transducer to the s target result in impedasca changes in the active coil of the transducer, 1 ghich can be seasured and converted to d distance sensurement. Ihe Contacting type transdurer has a aoveable spindle within a fixed coil assenbir. Uben the spindle is displaced, a voltage change is produced, which can be seasured and converted to a distance leasurement. h) The noncontacting 1ransducer gan be used remotely from the moving vessel by attaching a tar 7et of the same laterial as the vessel, st a fixed distance f rom the vessel. 4 g) Electronic 1:e point is a means nf eliminating errors in theraccouple readings cansed by changes in ambient 2 temperature and internal thermal { voltages generated in the readout instrument by referencing the therzocouple leads to a temperature gf OC or 32F. 12 CENERAL 1s stated in Articla 1 0, this document sets criteria for obtaining equipment and its calibra tion, installing the equipment, and establishing operation procedures to obtain the gap measurement hetween the ULS and the i RPY and to obtain the RPY and ULS temperature j at selected points iuring HFr. Io measure the 4 cap, temporary instrumentation consisting of small distance transducers installed on each 4 ULS, their wiring, and readout units will be required. Io obtain the RPY and ULS [ tempera tures during HF?, a series of thermocouples installed on each ULS, their viring, and readout devices will be requirad. Erovisions will be made so that, if a portion ] of the systen or one readout device f ails j jfhe amount of da ta lost will not invalidate j the test. 73 Intentionally left blank. Bh t l
I 24 1?.ACKET PREPARATION The ULS, shall be complete with all goverplates, stiffeners, and nachined faceplates in place and complete. Ihe shia material shall be raplaced by chemical lead 3/4-inch thick sith additional stainless stee: l shins sufficient to bring the machined faceplate to 15/32 inch from the RPY at 4 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 of the bracket shall be tightened to a snugtight condition. Ihe gap between the RPT and the f aceplate shall then be within+1/64 inch and -0.0 inch of the specified gapD Ihe j temperature of the RPY and ULS shall be recorded when the C as arc installed and 7 adjusted. 15 EHIELDING PREPARATION The permanent concrete shield plug cover and the i removable steel shield plug boxes, complete with filler material, will be completed and installed before the HFT. { l.6 IRER H0 COUPLES 4 i Iive theraccouples are required at each ULS I for a total of 60 per unit. gne thermocouple shall be located on the RPY surface adjacant j lo the machined contact patch. Ihe second l shall be placed on the ULS web near the ] e nd pla te. Ihe third shall be placed aidway along the ULS on the web. Iha fourth shall b. placed on the embedaent surf ace near the ULS i web. Ihe fifth shall be placed on the i concrete surface adjacent to the embedsent ( see Appendix 1, Pgs.1-1 and 1-3, for location). Thermocouples shall measure temperr.tures from 50 to 600F. 4 4 j 17 p,IST ANCE-MEASURING DETICES i Iach ULS requires two distanca-measuring transducers (for a total of 24 per unit) of 4 the conta:t or noncontacting type. Iht B-5
~ transducers shall be a ttached in the location s ho w n la A p p e n d i x 1, Pages 1-1 and 1-h or 1-5 28 EEADOUT UNITS Eeadout units aar be strip type recorders and/ or digital readout units. Eor the temperature readings, a single readout unit aar read as many as 12 thermocouples (12 channels). Eor the distance seasuraments, a maximum of three I UL5 (six aeasurements) sar be takan 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 viring vill not cause loss of all data from any group of three adjacent ULS. Instrument suppliers aar propose data processors which will automatically record all output from the readout units gisultaneouslr on paper. Relaxation 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 specification is temporary.
Hiring details and routing shall be determined in conjunction with the Consumers Power Company startup and testing group. Kiring in the reactor cavity shall be subject to the specified temperature and humidity (see Section 6.2). Irpes of wiring must be Conpatible with the instruments being used and the test conditions, and shall be reviewed br the instrument supplier. Eources and loca tions of tempora ry power shall be designated by the Consumers Power Company startup and testing group. 2 10 ZEACTOS PRESSURE YESSEL INSULATION 4 Rofore starting the HFT and after installing all instruments, all RPY insulation supplied by the Mirror Insula tion Unit of Diamond Power shall be in place. Ihis shall include any insulation required to seal the opening through which the ULS projects. Eortions of this insulation aar have to be rearved later to remove the instrumentation and viring. t Supplying, insta llin g, and removing the Bs6 i i
insulation is the rasponsibility of the NSSS supplier. 1 11 EE1ATED MEASURE 3ENTS Io relate the temperatures and distance measurements taken under this specifica tion lo operating conditions, temperature seasurements of the reactor coolant systes shall he recorded at reactor inlets and outlets. Ihese data shall be recorded simultaneously. with the theraocouple and distance readings. Additional reactor coolant systen temperature measurements may be recorded between the thermocouple / distance readings. 1 2 12 CA1IBRATION I 111 instruments to be used in accordance with this appendLc shall be calibrated for use in the temperature range which will exist l<{ 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 sorrected for temperature effects. gslibration 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 the target material is required in the calibration procedure. All calibrations aust be traceable to the Nat.ional Bureau of Standards (see Section 9.2). 10 11T ER T R is 11 RERMANENT HATERIA15 Zermanent asterials are to be installed by others as specified in the design drawings or by the NSSS and insulation vendors. Ihe exception is the shin pack between the endplate and the RPY isceplate (see A ppendix 1, Pages 1-3,1-4 and 1-5). 12 EHIH PACK 1 3/4 inch thick portion of the permanent shia stack material ( ASTE A-240, stainless steel), shall be replaced for the duration of the h7T vith chemical lead meeting the requirements of B-T l
I ASTM 3 29-79. Ihe initial sira and shape of tne lead shall be as shown in Appendix 1, P a ge 1-6. In the test shia pack, the lead shim shall be i located between two 1/8 inch thick stainless steel shins. Ldditional stainless steel shims of various thicknesses will be added to bring the gap to the predetermined value (see Section 7.u). 111 stainless steel shias used j in the test shin pack shall have the same configura-tion as i; hose used for final construction. 13 IHERH0 COUPLES Iheraocouples (60 plus 6 sparas per unit) shall be Cement-on Thersocouples, Style III, ANSI Designation K Chronel-Alueel 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. Ihermoccuple and extension wire insula tion shall be fused teflon, tapa-i teflon-impregna ted glass, or equivalent i guitsble for continuous use to 600F. Cement i to attach thermocouples to ULS, embedaent and concrete shall be two-part thermocoat gopper oxide cement. Gement to attach thermocouples to the RPY surface shall be Omega CC sodium silicate cement or equivalent. Ihe brand name products discussed in this paragraph are supplied by Omega Engineering, Inc. Ihe j instrument supplier may proposa alterna tive systems for review and approval by project engineering. i 14 RISTANCE-MEASURING SYSTEMS i ( 17 stems shall include transducers, cables, power supplies, sending units (if required), readout units, and all necessary accessories. i 1 4.1 21 stance-seasuring devices (24 and 2 spares r77 unit) shall be eddy-current, n oncon tat; tang irpe or contacting type transducers. Ihe devicas shall have a range from 0 to 1 inch. Ihe transducer j vill be rated for service in a temperature range of 50 to 600F. B-8
1.u.2 Zower supplias chall be coa;atihie with the transducer and 1he available temporary power. Ihe electrical power for no more than six trausducers shall be supplied from the same gover circuit. 1 4.3 gables shall be supplied with sufficient length to reach from the ends of the ULS to the readout a nd da ta acquisition area._ 1 4.4 Ihe distance-measuring system shall be calibrated throughcot the given temperature range. Ihe eddy-current systen, if used, shall be calibrated using a thrgat of the same material as the RPY isee Section 9.2.2 for the I calibration procedure). Ihe instrument supplier may propose i alternative or nodified devices or materials if it ieems then more i suitable. j 15 READOUT UNITS i leadout units shall be one of, or a j combination of, the following:. strip type j chart recorders, digital readout indica tor, sultipoint recorders, data loggers, or j alterna tive Eroposals by the instrument i suppliers. Lheir range shall be from 0 to 650F for tempera ture, and 0 to 1 inch for 1 linear displacement. Iber shall be capable of i being read to the nearest 1.0F for ) tempera ture, and to 0.001 inch when measuring distance. 1 hen multichannel units are usad to reduce the guaber of readout instruments required, input data from ne more than three i U1S 1 hall be read or recorded on one j instrument. At least four separate systess shall be used to record the data (see Section 7.8 ). j i Ihe readout systen shall be compatible with the thermocouples, transducers, power supplies, Ziring, and operating conditions present during the HTT. Ihe instrument supplier shall provide complete details on its proposed systen, including all instrunents and accessories Iequired and all equipment and power to be supplied by others. B-9 m ~
nossible data collection systems include: a) S trip type chart recorders with one to-three gha nnels using such accessories as the Da taplex 10 lutomatic Signal Scanner, expander /aarker and pen lift, electronic ice point, tap 11fiers, and two-wire transmitters as supplied by Caega Engineering, Inc h) Digital readout'indi:stors in combination with multipoint gelectors, amplifiers, and transmitters may be used; glectronic ice point and other accessories are to be supplied, if necessary, ty Omega Engineering, Inc. A tranducer indica tor sisilar to Model 1002-0010 with eight-channel output can be gsad for displaying 4 contacting type transducer data. Ihese are as supplied by Trans-Tek, Inc. I Ihe eddy-current noncontacting transducer is pset of g seasurement system which includes a digital readout. Ihe instrument supplier could provide multiple readout units for a l' combination of as many as eight 1ransducers. Euch a system is supplied by Kasan Sciance Corporation. g) A recording device, such as Digistrip by Kaye Instruments, may be considared f or use provided tha t the supplier can demonstra te the recorder's reliability to the satisfaction of project engineering. 1) Alternative proposals shall be submitted for project gagineering approval. Ihe instrument supplier must provide complete data on devices to be i used, as well as all required accessories and viring. Ihe supplier shall provide technical da ta necessa ry to assess sensitivity, range, and accuracy of all instruaants as well as their suitability for use in this application. Ihe supplier shall also describe all requirements for the proper operation of its equipment (see Section 5.0). B-10 a a
o.0 LisTiti ATTON A ND ?! STING 11 EEQUENCE OF WORK lork shall be performed in the following sequence: 1 A) Calibrate (in the laboratory) distance-t sensuring devices and thermocouples as specified in Section 9.2 h) Install distance-seasuring devices and thermocouples g) Test installed devices for proper operation d) Install power supplies, readout unit giring, and associated iccessories g) Test and calibrate the measuring system in place 1) Complete RPY insulation g) Hot functional test Unit 2 1 January 1983) h) Hot functional test Unit 1 iMar 1983) A) Remove testing equipment 22 SALIBRATION 221 Ibermocouplas shall be calibra ted.such that the output voltages at standard temperatures in the gange of.50 to 600F are within ANSI MC96 1 arror limits f or the thermocouple type supplied. A minimum of three thermocouples shall be enacked by heating the thermocouples through the rance of 50 to 600F, 111owing then to cool to 50F, and plotting the output voltage against the actual temperatures (indica ted by a calibrated standard) at 25F intervals 1hroughout the range. Ihis heating and cooling sequance shall be repeated a second time to show repeatability. Ihe three therno:ouples used in the calibration shall be of the same type and from the same Eroduction run as those to be supplied. B-ll s
1 ~ 1 p,,alibration procedures and certification of standards shall be submitted to project engineering for approval, as described in Section 5.0. All standards must be traceable to the National Boreau of Standards. liternative es11bration methods based on recognized stsadards or codes aar be prepared and used With prior project gngineering a pproval. Ealibration da ta shall be submitted, as described in Section 5.0, to project engineering for review and approval before installing gar thermocouples. 4 1 2.2 Iransducers shall be calibrated as fo11oes. i Ihe transducar shall be attached to a section of 1-1/2-inch thick plate in i the same manner in which it will be } attached to the ELS during the HFT. 1 ], 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 cod 1 shall be installed in a furnace or oven j capable of bringing the unit to a uniform temperature and holding it a t 50F and at intervals of 51 to 500F. Ihe temperature shall be held at each data point for 2 minutes. Ihe transducer shall be connected to a readout unit of the same type to be used in the actual test. Ihe rod.from j the ta rget or the transducer shall j extend out of the furnace in such a sanner that it can be deflected by a calibrated micrometer or dial gage through the design range of the transducer isee Appendix A, page 8 for j suggested arrangement of test a ppa ra tu s ). Elots shall be made of I actual deflection versus indicated i deflection for increments a' minisua 0 025 inch and a aaxiana 0 05 inch. Ihree sets of readings shall be takan a t each temperature. Ihe readings at each point are not acceptable if ther vary by more than 0.005 i ~ h. Nonconforming transducers shall be adjussed or repaired and recalibrated or replaced with a calibrated i transducer. At temperatures beyond the normal operating range of the transducer, tne actual deflection may vary considerably from the measured deflection. This B-12
variation must be consistant and predictable it each point of seasurement. Calibration procedures shall be submitted to project engineering for review and approval before goamencing calibration. 111 neasurements will be traceable to the Nations 1 Bureau of Standards. Ealibration results are to be submitted to project engineering for review sad approval, as specified in Section 5 0, before installing any instruments. 23 INSTALLATION 131 Iransducees shall be installed with /d fty sheet metal brackets screwed to the ULS sideplate. Ihe bracket shall hold the transducer by a clamp or other means i suited to the transducer design. In either case, the attachment shall be positive, have essentially rero deformation, and shall not relax with time or heat. Ihe transducer contact rods or (if used) ta rgets shall be inserted through the 1/2 inch diameter holes in the ULS bearing plates. i R,efer to appendix 1, Pages 14 and 1->, for transducer location Eetails. Gare shall be taken to ensure that the transducer is installed so that it may detect HPV movement through the range of 0 to 0.5 inch. 2 3.2 Ihermocouples shall be installed with cement suitable for approximately 600F. Suggested cements are Thermocoat and Omega CC hign-te=perature cement by Omega Engineering Inc. Refer to Appendix 1, Page 1-3 for thermo-couple locations. 1 3.3 Eiring shall be installed as recommended by the instrument supplier. Eiring in the reactor cavity which is subject to heat shall be grotected or coated as recommended by instrument suppliers. Llthough tha viring is B-13
temporar7, it chall be neatly tied to supports, pcotected froa ibrasion, and installed so it will not fail during the test period. All connections shall be protected consistent with good electrical prar tice to ensur's continuous system operation during the test. 2 3.4 Eeadout instruments shall be located in the area designated bT the Consumers Power Compant startup and testing group. Eauipment shall be neatly installed. 2 3.5 Ihe location of all instrumen'ts shall r be documented according to the instrument sarial number. Eiring diagrams shall be made which indicate the instrument number, gover supp1r circuit, accessories, and readout unit and which trace the physical gath of the string. Ihe ULS shall be designa ted as shown in A ppendir 1, page 1. 2.a IESTING i is instruments are installed,' ther shall be checked for proper operation. Iransducers and thermocouples 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 ve.~tfied using calibra ted standards traceable to the la tional Bureau of Standards. Ehen instruments are installed sad wiring completed, all circuits, instruments, and readout devices shall be checked for proper operation as described above. Calibrated standards shall be used to verify proper system functioning. Ihe above described activitpes shall be performed according cogineer@ing. to proc, ures reviewed and approved by project Complete documentation shall be maintained on all tests, repairs, godifica tions, and other B -lh
l corrective actions required to verify system operstien. 25 gCT 7UNCTIONAL TEST, UNITS 1 AND 2 1 5.1 Ihe HFT shall be carried out based on information provided by Censumers Power Company startup and testing group. 1 5.2 Ihe saquence of temperature versus time during the HFT shall be as ghown in Appendix 1, Sheet 9. 2 5.3 leaperature and distance data shall be recorded as follows: Eeadings shall be taken at the beginning, middle, and end of each temparature hold and at the sidpoint of til transitions as the temperature rises and falls (refer to A ppendix _l, Page 1-9). A minimum of one set of readings shall ha tsken each 12 hours. 2 5.4 111 equipment failures shall be reported to the project angineer. Ihe failure shall be corrected if possible. Eecause each HTT goes through two cycles of heating and cooling, til instranent failures shall be corrected between the first and second cycles if possible, without unacceptable delay of the HFT. 2 5.5 1 complete set of data shall be transmitted to the project engineer a fter ahe first RCS heating and cooling cycle is complete. Ihis shall include the temperature of the RCS coolant at the inlets and outlets. Ihe location of the RCS, oolant tempera ture sensors shall be ;;ven end the $s ta shall be referenced to the date and time of day. Ihese tempertteres shall be reccrdai simultaneously with the theraccouple B-15
and iirtance reedings raquired by this p 0C9 dure. c.f.6 Ehen the Unit 2 HFT is completed, the test will ha reviewed by project engineering. It that time, all f ailures and secuence of data recording shall be reviewed. If it is believed that the systes should be modified in any way to grovide reliable data, these modifi:s tions shall be made before Itarting the Unit 1 HFT. If it is believed that more or less data are required, the test procedure Khall be modified to a cconsoda te these requirements. All procedural modifications shall be revieved and approved-by project engineering. 1 1 5.7 Eroject engineering shall be informed of the status of the installation and j testing a t all times. Eroject engineering shall be advised before the testing is scheduled to begin. 1 representative of project engineering shall be present at the start and for the duration of the HFT. 10.0 V0PK 10 FC110E 10.1 INCINEERIIG lhen project engineering receives the HTT data, the temperatures will be compared to those predicted by analysis for operating conditions. Ihese predicted tempera tures aar vary at different locations on the EPY. Ihe distances measured during HTT will then be j proportioned to operating temperatures. Ihis vill verify the specified gap in existence at operating leaperature, and will confirm the gap required at cold shutdown. Ehen the cold shutdown gap is verified, the informa tion will be added to the applicable drawing and the shins will be installed to bring the ULS faceplate to the required gap. { 10 2 GLEAN UP 3 hen HFT is completa, all transducers and wiring shall be removed from the reactor gavity. lay temporary attachments f or supports shall also be removed. Ihe area B -16
shall be left in a clean condition as de arsined by Consuaars Power Company startup and testing. 10 3 2THER CONSIDERATIONS ~ Sap and temperature sessurement during Unit 1 BFT shall be carried out in a similar sanner as Unit 2. Eroject engineering vill review both the data collected during Unit 2 HFT and the test operation. Changes in procedures for setting gaps and changes in equipeent will be made, if necessary. 10.4 IQUIPMENT REMOVAL All equipment and viring used for the gap and tempera tura seasurement shall be removed from the containment building at a time to be designated by Consumers Power Company startup and testing. t i l B -17
APPElDIX 1 C UPPER 360' 0' LATERAL SUPPORT ? (TYP) 12 \\ 17' ,,r / 3 ~ j / \\ [ l N \\ l / \\ \\ 3 'o t' 1 13' 90' 270* R = 12*-0" i +, 3 Rv EDaE t 4 4 \\ 47 i / \\ <-*~ , s / \\ jy j/ / ~~ s ~. CONCRETE 4 SHIELD PLUG \\ _/ TYP 7 6 PRIMARY SHIELD 180* UPPER LATERAL SUPPORT PLAN AND NUMBERING SYSTEM UNIT 1 SHOWN, UNIT 2 OPPOSITE HAND B1-1
REACTOR INSULATICN EL633* 5%~ ,, 1, 1 1 1, 1 1, REACTCR i 8 PRESSURE I A F-WALL REMOVA8LE l SHIELD l EL 632*-3"N>f f / //1 / // / s l F sg 1 m-- rr--g-5%* es I g iI UPPER LATERAL 88 g ii SUPPORT (ULS) ll l )) \\ j S U ULS BUMPER ] l a, \\ PLATE ' \\* . 4, - l 9 a l .....a 8" 8 ,','a.* ese l ~~ l p'[ \\ p - . ;. ' 5' ULS BEARING 'N ld SHIELD . = i ':. PLATE L_ Ptuo
- g
- =~
gi SECTION A-A NOMENCLATURE B1-2
1 l 4" r f f f f f f f f f f INSULATION p r \\ THERMOCOUPLE 2 2 (ON WEB) r-s s g_ / l ' / / / / // / // f ggly I e _ _ _ _ THERMOCOUPLE _,__.____._,7____l,___, ll j Il [+h' \\ ll r---h ~ ~ l 2-4 (AT JUNCTION II OF WEB AND EMBEDMENT) Il 1 THERMOCOUPLE l 11
- I ci ll It i 7...
. _.' 4 \\ 21 "C T l ~4 f
- THERMOCOUPLE
'Y ,l/ : l j '.1 2 5 ON CONCRETE 'i,h /. "' k-ADJACENT TO l p EMBEDMENT 1 \\ \\ N - 1 =. '_ p N l f THERMOCOUPLE l 2-3 ON ULS f WEB ../ l ,,, 1._ l$-
- c..*:
3 L_ INSULATION
- l ..
P SECTION A-A THERMOCOUPLE LOCATIONS SHOWN FOR ULS NO. 2 1 (5 UNITS PER ULS) B1-3
i I r-1 l l4 TOS EL 632*-3" 1 %" E (TYP) FACE OF RPV i N / s i N , r- - y g,r - - j \\ I n n ii r--- li 11 l II l II 18 'l ,l u,M u u u CONTACTING TYPE ,l l l TRANSDUCER SUPPORTED ?n / 74X-M+:-
- [
l ON THE ULS _, _e, l 1 1 -+ +- I ? I SHIMS l l ROD SPRING LOADED b-TO FOLLOW RPV MOVEMENTS SECTION A CONTACTING TYPE TRANSDUCER (2 UNITS PER ULS) B1 14
l I y_ I 4-----FACE OF RPV TOS EL 632' 3" 1 %" ((TYP) g N / i N I _ q' r ,,----,r---_,- g 11 il l It I ll is is li c- - - l- \\ is si si ii i c- [ l si ll ii ii i g u u u uq"
- :-i-l NONCONTACTING TYPE I
l TRANSDUCER SUPPORTED 5h ', " =f ~~ -WMl ON THE ULS wl l L, h".,. f" a l 1 + + l SHIMS l l 4" x 4" TARGET AND ROD b-SPRING LOADED TO FOLLOW RPV MOVEMENTS I SECTION A-A NONCONTACTING TYPE TRANSDUCER (2 UNITS PER ULS) B1-5
27 3/4" For Non Radial Shia Pack (See Det. 2. Drwg. C-367(Q)) ' 24 3/4" For Typical Shim Pack (See Det. 1, Drvg. C-367(Q)) 1/8" 3" 1/8" 1" 1"- k l 2 1/4" / \\ l} R= 3/4" R= 3/4" Stainless Steel (Typ) (Typ) Plate Each Face 9 3/4" (ASTM A-240XM-19) .b-Lead Shim 29 Pig Lead) 23 3/4" For Non Radial Shim Pack 20 3/4" For Typical Shim Pack LEAD SHIM to For Test Shim Pack Ym
i e t 64.*e.e. =r. % t l. s * ,.a.
- d8 T
"r_ m.., ..i m, i q 8 #
- ,*O*,0
- Data Acquisition i
Area .=. / / *n..m m / ,\\ d ._4 . f ..t. // tr e.i. ~, r .m o + . i a m.. ,..'m,f. . I._ .~. . pas-.ma ~ y.+v ~ .v A i a g *** ** cu 3 sa j l w :sw J
- 4,-
....; Q ! 4.. 5-f" *., . r fu. s ',,,. y :,,: g-t';,$.,.\\,.f' ...i ~ - "g,.. (u s..,s/. /[ _ % ,\\ f %.h-[: .) "1'*4... 3.,, - . > + f rJ 6, s. _... s 5_ ~ t s u r., /, .s ,s. f \\l.. ' _1 _L
- . LL L
w. d: I si ./ 'f - =;'- - C. .,/./ : ) ( (g w f' . w,, i ..uve. Iq. 3 7 *w. 8'W t T W* \\ T. / / e;
- g aa'ne
%*ts g-f a.sas.*-'s* tf 34 l s i aw.. .g = s=~'-* m.,,.;.. , N'
- ? - f. 'l
,=, h i -s s, i l f'" p.e. c.o,.m.e - I s a o.- s ,l g b g . - = = 'a a <, t = .' 'A.,_ \\ ,;. K/ j ' l '5 NN N, N. .e: [*# %. y R n w -- t[/ 4., 6 .-s i I =_ i
==**g T ; an,*sg, .es +r s..
- v..-
- 1. r
.r, e. a.r J mtw c:.. a s s. c-DATA ACQUISITION AREA LOCATION (UNIT 1 SH0 tin, UNIT 2 OPPOSITE HAND) B1-7
FURNACE SPRING-LOADED ROD TO FOLLOW MICROMETER l l 3" x 5" x 5" THERMOCOUPLE STEEL BLOCK WIRING A READOUT MICROMETER m
== s TRANSDUCER %" HOLE l/ l ALL OTHER PL THERMOCOUPLE [ w READOUT 4 i A 510 A 240 XM 19 m CARBON STEEL STAINLESS STEEL SUGGESTED CAllBRATION SETUP l l B1-8 I
l Pressurize ' ) ]-Pnsture 2000 a. j 1750 l I I 600 1500." I 5 500 1250 3 400 1000 = a. 300 .RC 750 5 Temperature'f-a !! 200 500 ~ I 250 100 w 0 2 4 6 8 10 12 14 2 4 6 8 10 12 14 Time, Days Time, leeks HOT FUN 0T10NA1. TESTING RC TEMPERATURE AND PRESSURIZER PRESSURE l B1-9
r-APPEh')IX C: Unit 1 Anchor Stud Lift-Off Data mil 181-0953a141 1
l TABLE C1 l UNIT 1 REACTOR VESSEL ANCHOR STUDS i Lift-Off Data Stud Number (2)' Date Hydraulic Bolt Stress Pressure to Nearest ksi Sequence B&W Teledyne (psig) 1 l l 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 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 s. 18 10 in 16 in 5-16 11,700 79 19 11 in 04 in 5-19 13,700 93 20 12 in 28 in 5-19 12,400 84 mil 181-0953a141
TABLE C1 (Continued) 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 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 36 20 in 26 in 5-27 9,600 65* 37 17 out 38 out 5-28 12,500 85 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 42 22 out 20 out 5-30 12,300 83 43 23 out 08 out 6-17 12,300 83 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 6-18 10,900 74* 47 23 in 08 in 6-19 12,300 83 48 24 in 32 in 6-19 12,400 84 mi1181-0953a141
l TABLE C1 (Continued) 49 25 in '41 in 6-20 12,200 83 50 26 in 17 in 6-20 11,800 80 1 51 27 in 05 in 6-20 13,000 88 l 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 l 57 29 out 47 out 6-25 10,200. 69 58 30 out 23 out 6-25 12,200 83 59 31 out 11 out 6-26 12,200 83 l l 60 32 out 35 out BR0 KEN I 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 l 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 1 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
1e TABLE C1 (Continued) 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 88 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 l 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. 4
- )
.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. l mi1181-0053a141
n @@@ @@@g @@@@@@@/@ D OurSIDe FAILED INSIDE g w@ 11 0 G REACTOR SKIRT @g FAILED FAILED g@@ NSIDE b g @ @@ @@ @@ @ OUTSIDE POSITION AND NUMBERING OF STUDS IN UNITS 1 AND 2 FIGUPS C-1
t l APPENDIX D: Justification of the Ductility Ratio for Use in the Design of the ULS N mi1181-0953a141
n The following-provides the justification for the use of an allowable ductility ratio of 10 in the design of the upper lateral support (ULS) brackets: 1. The least radii of gyration (r,g,,) at Sections 1 through 6 (See Figure 5.1 f' rom Section 5.0 of this report) are given in the table below. Section 1 2 3 4 5 6 r. (in.) 4.85 4.41 4.48 4.57 5.59 4.97 min. The effective buckling length of the brackets, assuming that the end welded to the steel embedment is clamped and the end in contact with the RV can, displace without rotation, is 1.2 times the length of the bracket (42 in.); see' Reference D1. Using the least r. from the table above, a min. J - conservative (upper bound) effective slenderness ratio (KL/r) for the brackets can'oe calculated as follows, namely KL/r = (1.2 x 42)/4.41 = 11.4. 2. Reference'D2 specifies that the slenderness ratio of a stub-column must not exceed 20. In otherwords, to insure that the column will yield, undergo plastic deformation and go into strain hardening; its slenderness ratio should be equal to or less than 20. 3. The column strength curves established by the structural stability research~ council for rolled and welded built-up sections (Reference D3) indicates that the strength of columns where A5 15 will be controlled by yielding rather than buckling, where A is defined as, A = fE. The parameter A for the ULS brackets is equal to 0.13 which is less than 0.15. l ~~ mi1181-0953a141 D-1 a
=. w 4. Norrir and others (Reference D4) stated that a compression member will develop its full ultimate compressive strength without buckling if its ' effective slenderness ratio does not exceed 15. 5. One can establish based on the aforementioned arguments that the brackets .can develop their ultimate compressive strength without buckling. In this case, the allowable ductility can be as high as , where cy and est are the strains at the initiation of yielding and at the onset of strain I hardening, respectively. For structural steels with a yield stress of 50 ksi or less, a lower bound value for **' is equal to 10, see Reference DS. cy 6. In addition to what has been outlined above, the AISC nuclear specifica-225 tion (Reference D6) allows a ductility ratio in compression of 'A2 with a maximum of 10. Using A = 0.13, for the brackets, the allowable ductility I calculated from the AISC formula is more than 10; in this case, an allowable ductility of 10 should be used. t 7. Furthermore, the ASCE Committee on impactive and impulsive loads allows a ductility ratio of 140,000/oy(fb)2 with a n x vun of 10, see Reference D7. Using cy = 38 ksi and -L = 11.4 c r ae .ackets, the allowable K ductility ratio calculated from the ASCf formula 9xceeds 10, in such a case the maximum value of 10 should be used. 8. In order to achieve the established allowable ductility ratio withrut premature local buckling of plate elements, the AISC specification (Reference 6)doesnotallowthewidth/thicknessratio"toexceed76/ycy' .forunstiffenedplateelementsand238/./3y'forstiffenedplateelements. The width / thickness ratios for the plate elements of the ULS brackets satisfy these requirements. mi1181-0953a141 D-2
J i 9. The presence' of bending in the brackets -is considered in establishing ' the - -axial compression which will intiate yielding. Furthermore, the allowable . ductility ratio in bending is 12.5 (See References D6 and D7). Hence..the presence of bending will not cause the ' allowable ductility to be less than 10. j. l i l l l l. l .mi1181-0953a141 0,-3 e -y v
REFERENCES D1. Structural Stability Research Council, Guide to Stability Design Criteria for Metal Structures, Third Edition (1976), Pg. 74. D2. Structural Stability Research Council, Guide to Stability Design Criteria for Metal Structures, Third Edition (1976) - Technical Memorandum No. 3: Stub-Column Test Procedure, Pg. 561. D3. Structural Stability Research Council, Guide to Stability Design Criteria for Metal Structures, Third Edition (1976), Pg. 68. D4. Structural Design for Dynamic Loads, by Norris, Hansen, Holley, Biggs, Namyet and Minami (1559), Pg. 13-15. DS. Plastic Design in Steel, ASCE Manual No. 41 (1971); Chapter 5 - Verification of Plastic Theory, Pg. 42. D6. AISC Specification for the Design, Fabrication and Erection of Steel Safety-Related Structures for Nuclear Facilities, issued for trial use and comments on July 17, 1981. D7. Report of the ASCE Committee on Impactive and Impulsive Loads, Vol. V of the Proceedings of the Second ASCE Conference on Civil Engineering and Nuclear Power, September 15-17, 1981; Pg. 2-112. mi1181-0953a141 D-4
-r-o APPENDIX E: Methodology for the Computation of the Mathematically Equivalent ULS Spring Rates l mi1181-0953a141
The following provides the methodology to develop the spring rates for the RV upper = lateral support (ULS) brackets: 1. Capacity of the ULS Brackets Under Axial Compression and Bending: Under combined axial compression and bending the axial capacity "Py" is determined from the AISC interaction formulas given in Section 2.4, Reference E1. Since the bracket has a variable cross-sectional properties along its length and the bending moment varies due to the variation of the axial load eccentricities, the axial capacity "Py" was calculated at six different sections (See Figure 5.1 from Section 5.0 of this report). The capacity "Py" used is the minimum obtained. In order to insure that the axial capacity "Py" will be reached without premature local buckling of the bracket's plate elements; the width-thickness ratios of these elements satisfy the requirements of Section 2.7 of the AISC specification (Reference EI). 2. Lead-Displacement Relationship of Individual Brackets: Load-Displacement relationship of individual brackets are shown in Figure E-1. The axial capacity "Py" is calculated as described in Item 1 above. The corresponding displacement Ay is calculated as follows: 5 OY* i= 1 Where E = Young's Modulus of Elasticity, Ai = Average Cross-Sectional Area of Segment i, and Li = Length of Segment i The bracket was divided into 5 segments as shown in Figure 5.1 of this report. mi1181-0953a141 E-1 l
I The maximum displacement Amax is. equal to 10 x Ay; using an allowable ductility ratio of 10. 3. Spring Constant Curves'(Load-Displacement - Relationship) for ULS Brackets Combination in a Specified Direction: To determine the load-displacement relationship for the ULS brackets combination in the three directions given in Section 11.1.2.4 of this report and illustrated in Figures 11.11 through 11.13; the structural model shown in Figure E-2 was used. For a given direction of motion and from geometry, the first brackets to come in contact with the RV are determined. A structural analysis is then carried out using the model shown in Figure E-2 with the brackets in contact only supporting the RV. From this analysis, the load-displacement relationship of the system is determined. The load and the corresponding displacement of the system to cause contact with additional brackets are then calculated. This load and the corresponding displacement represent a point on the load-displacement curve where stiffness changes. At this point, the new bracket or brackets to come in contact with the RV is added to the model and a new load-displacement relationship is determined. The process is repeated until all the brackets in contact with the vessel yield or the displacement exceed 10 times the displacement of the bracket which yielded first. When a bracket reaches its yield capacity Py, the support provided by the bracket is removed from the model. A load equal to Py representing the capacity of the yielded bracket, however,'is applied on the vessel opposite to,the direction of motion. The load at which one or more brackets yield and the corresponding displacement represent a point on the load-displacement curve where stiffness changes. mi1181-0953a141 E-2
7_.- .9 /.- References ~ E l.' - AISC Specification for the Design,- Fabrication and Erection o'f Structural Steel for Buildings - 1969 Edition with Supplements 1, 2 and 3. P mi1181-0953a141 E-3
Load Py I I I e l I l 3 I I l I i i t Ay Amax Displacement Figure E Load Displacement - Relationshin of Individual Brackets / ~.L, i ! / Primary Shield Wall s i \\ ES i jj Brackets 'C I O
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y ot od t Represented by Rigid Beam Elements Figure E Model to Establish Surine Constant Curves for ULS Brackets E-4
.= .,7 ..l (, N -~ he - i t - l i APPENDIX F: Teledyne Engineering' Services Letter (W E Cooper to H W Slager) Dated [. = December 11, 1981,. Reaffirmation of Letter Report TR-5255-1. i. l I, r l l L l l i l e i I t i i mi1181-0953a141: t
'..,.a e ' 9 '#TELEDYNE ENGINEERING SERVICES 130 SECOND AVENUE WALTHAM, MASSACHUSETTS C2254 1617) 890 3350 TWX(71C) 324-75C8 December 11, 1981 5355-2 Mr. Harvey W. Slager Consumers Power Company 1945 W. Parnall Road P. O. Box CP 10-4672-Q Jackson, Michigan 49201
Subject:
Reaffirmation of Letter Report TR-5255-1, Expanded Criteria for Acceptability for Service for Midland Unit 1 RPV Anchor Studs
Dear Mr. Slager:
TES TR-5355-1 expanded the criteria for acceptability of the subject studs to include Service Levels B, C and D limits in addition to the Service Level A criteria provided by TES TR-3887-2, Rev.1. At a December 2,1981 meeting, NRC questioned the expanded criteria and provided the basis for their concern. TES has evaluated all available information and reaffirms the reconinendation of TR-5355-1. The background may be summarized as follows: 1. The RPV is constrained from rotation during a system faulted condi-tion by studs which restrain a flanged RPV support skirt against a pedestal upper ring plate. 2. The studs in question are 2-1/2" in diameter and 7'-4" long and are tensioned between the upper surf ace of the skirt flange and a lower ring plate near the bottom of the studs. 3. Lateral shear is carried by separate structural elements, so the stud carries no significant lateral shear. 4. The stud is tensioned by a stud tensioner, not by torquing, so the stud carries no significant torsional shear. 5. Three of the 96 studs f ailed by stress corrosion cracking after in-stallation as a consequence of the combined e-fects of stud ma-terials, atmosphere, and prestressing. 6. Certain design revisions were adopted which inclJded decreasing the 'i prestress to a small (less than 6 ksi) value to avc id long-term stress l corrosion cracking and limiting the service stress to a value propor-tional to the stress to which the studs were loadtd subsequent to the period of initial high prestress. The minimum *,alue of this stress ~ the av'e$gF?% measured during the 1980 detensioning was 75 '.si and value was approximately 85 ksi. 4 7
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DEC1 881 ENGINEERS AND METALLURGISTF P* v*y ...M er.. .a 6 ' k, ** t
.e WTELEDYNE Consumers Power Company ENGNEERING SERMCES December 11, 1981 + Page Two 5355-2 7. Because of the previous stress corrosion cracking and because the fracture toughness of such studs is relatable to susceptibility to such cracking one must assume that the stress carrying capability of a stud is limited to that which has been demonstrated. Therefore, the effective yield strength, effective ultimate tensile strength and failure load strength of the stud must be considered to be no higher than that measured, even though that value is much smaller than the minimum specified yield strength. 8. The basic allowable stress, Service Level A, should be taken as one-half of the effective ultimate tensile strength. Increase factors may be applied to the Service Level A values for Service Levels B, C and D. 9. Although satisfaction of ASME Code rules is not required for Midland component supports, ASME Boiler and Pressure Vessel Code, Section III, Subsection NF rules are to be applied for guidance. 4
- 10. Regulatory Guide 1.124 is applicable to the Midland component sup-ports.
The above listing is intended to sumarize the background in a manner which is consistent with past discussions and which is acceptable to all parties. The present issue is that of establishing the increase factors applied' to the Service Level A values to determine the allowable value for Service Level D. Although there arc differences in the increase factors proposed for Service Levels B and C, tne differences are slight and the lower of the two values is acceptable. There are also partially compensating dif-ferences in the quantities to be multiplied by the increase factors. The differences may be sumarized as follows: Service Level TES NRC A 0.500 S 0.500 S T N 8 0.500 S 0.575 S T N C 0.667 S 0.625 S T y 0 0.700 S 0.625 S T N wnere: ST = The lowest value measured on any stud during the 1980 detensioning, 75 ksi, or a higher value if con-firmed by a specific test on the specific stud at a later time. SN = The value measured on a specific stud, whether in 1980 or at a later time. F-2
'RTELEDYNE Consumers Power Company ENGINEERING SERVICES December 11, 1981 Page Three 5355-2 l Specific values for Service Level D for the more highly stressed studs are as follows: Outer Stud 1980 Value TES NRC Calculated 38 85 52.5 53.1 37.5 37 75 52.5 46.7 44.8 34 85 52.5 53.1 49.2 33 92 52.5 57.5 44.6 32 83 52.5 51.9 43.0 31 94 52.5 58.7 41.9 The tabulated calculated values are those now calculated, and all are sufficiently low as to satisfy either the TES or NRC criteria. However the margin for stud 37 is but 1.04 against the NRC allowable as compared to 1.17 against the TES allowable. Similar values for stud 34 are 1.08 and 1.07, respectively. Margins for all other studs are above 1.18 against either criteria. Subsequent calculations may result in increased values, and the minimum margin against the NRC criteria is but 1.04 as compared to 1.07 for the TES criteria. These numbers are misleading, however, because the TES procedure would permit increasing the margin for stud 34 to 1.17, the value applicable to stud 37, by retesting to 82.4 ksi, a value smaller than that measured in 1980. Any increase in the margin with the NRC criterion would require retesting to values higher than those measured in 1980 with significantly increased probability of stud fracture. The apparently small difference between the TES and NRC criteria really repre-sents a major difference in level of risk involved in satisfactory comple-tion of construction. It is this concern that causes us to try to justify the previous TES position. The difference between the TES and NRC criteria arises because Subsection NF provides for more than one desis n procedure for component supports. TES has applied the load rating pror edure, a procedure which is completely defined by Subsection NF and Regulatory Guide 1.124. NRC has applied the design by analysis procedure, and has introduced as part of that procedure an incomplete and unpublished action of the ASME Committee. Even if that action were complete and published, however, it would only be applicable to the design by analysis procedure and would not be a requirement when the load rating procedure was applied. Furthermore, we will demonstrate that the present direction of the Cornnittee completing thtt action is ac - ceptance of an increase f actor equal to that used in the TES criteria. Justification for use of the load rating procedure is provided in TR-5355-1 so will not be repeated here other than reemphasizing that that procedure permits a sampling procedure, whereas each and every Midland stud was tested. It is also noted that there is no Subsection NF prohibition against applying the load rating procedure to component supports which include threaded parts. F-3 1
^ P TELEDYNE Consumers Power Company ENGNEERING SERVICES December 11, 1981 Page Four 5355-2 The present Subsection NF may be considered by NRC to be incomplete with respect to Levels C and D Service Limits for bolts when the desion by analysis procedure is used. TES considers the present rules to be complete and consistent with the increase factors applied by the TES criteria. ASME Main Committee Action 81-73 at the January 1981 meeting revised the rules for bolts when the design by analysis procedure is followed. Publi-cation has been withheld pending completion of the NF-3000 rewrite. That action applied the same increase factor for Service Level D as for Service Level C,1.25 which results in the previously stated limit of 0.625 S
- N However, the action taken qualified the Level D factor as an interim value pending revision of Appendix F.
(The Working Group on Component Supports had previously approved the action without an increase factor but with words referencing Appendix F.) Appendix F is being revised and the re-quired words are included in F-1335.1 of Draft 11 dated August 1981. This reads: 3 "The average tensile stress computed on the basis of the avail-able tensile stress area (independent of any initial tightening force), shall not exceed the smaller of 0.7 S IS. When high u y strength bolts or threaded parts having an ultimate strength greater than 100 ksi (689 MP ) at operating temperature, are a used in component applications, the maximum value of the stress at the periphery of the bolt cross section resulting from direct tension plus bending and excluding stress concentrations shall not exceed S. The bolt load shall be the sum of the external u load and any bolt tension resulting from prying action produced by deformation of the connected parts." It is our understanding, based upon participation on the Task Group, that the.. is concern over permitting S in some applications, but we have not heard any concern about the 0.7 S ycriterion. u Based on this discussion, TES is of the opinion that the criteria recom-mended by TR-5355-1 is applicable to the studs in question and is in complete conformance with Code and Regulatory Guide requirements. We recomend that this proposal be applied rather than the alternative pro-posal of NRC. Very truly yours, TELEDYNE ENGINEERING SERVICES p William E. Cooper Consulting Engineer WEC/lh F-4 .-}}