ML19326C209
| ML19326C209 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/14/1978 |
| From: | David Williams ARKANSAS POWER & LIGHT CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| 1-038-10, 1-38-10, NUDOCS 8004210697 | |
| Download: ML19326C209 (19) | |
Text
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' DISTRT110 TION AFTER ISSUANCI G. OPEllATING LIGSE i
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NRC OISTRIBUTION pon PART 50 DCCKET MATERIAL FRCM: Ark. Power & Light Co.
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,70: Mr. R. W. Reid Little Rock, Ark.
72203 3-14-78 D. H. Williams car 3,agasvac PMop INPuTPonM NUMesa CP CoPfs3 AsCalVED TLaTran C y nszso I EfonicinAL funcL,nassaiso
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IP ancLosuna Info on fracture toughness & potential for lamellar tearing of steam generator & react or coolant pump support material...
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ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 UTTLE ROCK, ARKANSAS 72203 (501)371-4000 March 14, 1978 K yk.;U U ik e a,7 P ' T 7," " [q i)Ul i f""
LL LC.L !. U L Ubi l-038-10 Director of Nuclear Reactor Regulation
/g A'ITN:
Mr. R. W. Reid, Chief P, r
,, r-V Operating Reactor Branch #4 U. S. Nuclear Regulatory Commission Yt Washington, D. C.
20555
!MR2 41978" [~3
Subject:
Arkansas Power G Light Company
% % b w ress e
Arkansas Nuclear One-Unit 1 \\,
M*
Docket No. 50-313 N'k[Q License No. DPR-51 f
Fracture Toughness and Potential for Lamellar Tearing of Steam Generator and Reactor Coolant Pump Support Materials
' (File:
1510)
Gentlemen:
Your letter of September 14, 1977, requested that we provide information to seven (7) items concerning support materials of steam generators and reactor coolant pumps at Arkansas Nuclear One-Unit 1 (ANO-1). Your letter also requested our evaluation of the fracture toughness of these support materials but this request was negated by telecon with your Mr. R. Snaider on October 26, 1977.
Please find enclosed the requested information concerning our steam genera-tor and reactor co'ol' ant pump support materials. Our responses are provided in two parts. The first part contains that information pertaining to the reactor coolant pump supports and the other supports near the top and bottom of the steam generator. The second part contains information as it relates to the skirt support at the bottom of the steam generator.
It should be noted in the responses to Item 2 that the design loads and com-binations for the supports do not strictly adhere to those indicated in your Item 2 as normal, upset, emergency and faulted. However, the design loads used do cover the most critical load case and comply with the intent of Regulatory Guide 1.48.
If we may provide further assistance, please advise.
Very truly yours,
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Daniel H. Williams
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Manager, Licensing DHW:DGM:nf l
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3 PART I ANO - UNIT 1 RESPONSES TO ITEMS OF CONCERN ON FRACTURE TOUGHNESS AND POTENTIAL FOR LAMELLAR TEARING OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORT MATERIALS This report contains responses to the NRC questions concerning the fracture toughness and potential for lamellar tearing of the materials used in the steam generator and reactor coolant pump supports for ANO - Unit 1.
ITEM 1:
Provide engineering drawings of the steam generator and reactor coolant pump supports sufficient to show the geometry of all principal elements.
Provide a listing of materials of con-struction.
RESPONSE
The following drawings show the details of the supports under consideration:
Drawing 6600-C-184, Rev. 6
- Coolant Pump Seismic Supports Drawing 6600-C-186, Rev. 11 - Steam Generator & Coolant Pump Supports Drawing 6600-C-187, Rev. 4
- Equip. Foundations & Details Drawing 6600-C-ll3, Rev. 6
- Liner Plate, Floor Plan and Details, Sh. 2 - - _ _ _ _
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The supports were all fabricated from ASTM A-36 material with the exception of bolts, rods and the 3-1/2" liner plate which was used to anchor the steam generator lower support.
The bolts and rods conf orm to ASTM-A-490.
The 3-1/2" anchor plate conforms ta ASTM-A-516, Grade 60 (Dwg. C-ll3, Section C).
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-ITEM 2:
Specify the detailed design loads used in the analysis and design of the supports.
For each loading condition (normal, upset, emergency and faulted), provide the calculated maximum stress in each principal element of the support system and the corresponding allowable stresses.
RESPONSE
Since the Arkansas Nuclear One Unit 1 design criteria were established prior to the issuance of Regulatory Guide 1.48, the design loads and combinations do not strictly adhere to those indicated in Regulatory Guide 1.48 as normal, upset, 5
emergency and faulted.
However, the design loads use.d do cover the most critical load case and comply with the in-tent of the Guide.
Table I tabulates the maximum design stresses and the corre.s-ponding allowable stresses for the steam generator upper supports, lower supports, the reactor cooling pump lateral suppressor supports and the RCP vertical hanger supports.
For the upper steam generator supports, the most critical loading case was the design basis seismic event and the loading from a hot leg slot type failure (LOCA) condition.
The force from this loading combination was applied in any direction so as to obtain the maximum stresses in the girders and embeds.
The force used'in the design and sizing the various members was larger than the final loads calculated *
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Actual stresses are lower than the cor-responding allowable maximum design stresses (see Table I).
The lower steam generator suoports were designed for two loading combinations which were considered the most critical condition.
Load case 1 was for a cold plant condition (shut-down) which consisted of dead load and design basis earth-quake.
Due to the gap between the steam generator and the upper lateral supports in the cold condition, the overturning moment induced by the seismic force produced the maximum tensile stress in the anchor bolts.
Load case 2 was for the hot plant conditions (normal operation) which consisted of dead load, thermal load, design basis earthquake and the loading from a hot leg slot type failure (LOCA) con-dition.
This combination produced the maximum shear stress in the anchor bolts.
The reactor coolant pump lateral suppressor supports provide restraint for the design basis earthquake load.
The supports were designed for 200 kips which is the capacity of the sup-pressor during fast movement.
The postulated design basis earthquake force is 142 kips which is less than the capacity of the suppressors.
The actual stresses are as shown in Table 1.
The vertical load from the pump and motor is distributed be-tween the piping system and the pump vertical hanger supports.
The load taken by the support is 50 kips each as given by the NSSS vendor.
The stresses are given in Table 1.
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p Nomenclature for Table 1 are as follows:
f,
- design axial stress.
f
- design bending stress, b
f
- design bending stress about weak axix.
bx f
- design bending stress about strong axis.
by f
- design through thickness tensile stress.
t f
- design shearing stress.
y F
- horizontal force.
H f
2 2
+ (f )
S
- maximum shearing stress; equal to
+
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y s
f" S
- maximum principal stress; equal to 7
s
+
S n
F
- all wable bending stress.
b F
- allowable shearing stress.
y F
- the minimum yield stress from mill test.
y C
- compressive force.
D
- dead load E
- design basis earthquake load.
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H
- force on structure due to thermal (operating condition).
M
- overturning moment.
t R
- pipe rupture force.
T
- tensile force.
V
- shearing force.
t
- horizontal thickness of T - joint.
I 1
l t
- vertical thickness of T - joint.
2
(+] sign - axial load in tension.
[-] sign - axial load in compression.
S.G.
R.C.P.
- reactor coolant pump.,.-Ohm 4-6
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't TAP:I 1 S_TP_EO_C_ES T% STEA*t_f.r_T_A. T S1??t%T & FDC*C? COO! AYr P2'P FUPMm?
6AX. PW ".:4 S"* ' T N.3 Al
- J.18 FLE STF Ci/ES f
f, f.
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- g. m surroats so leAnl::c CCsoIT10u 3
(F3I)
(FSI)
(F*I)
(F37)
(?3I)
(KS T)
(K3!)
(Cif) 5, = 16.6 18.25 32.85 s = 18.25 Cirder Erled 15.2
+29.5
= E' + H a
h 1000, 5850 = 6850 5, = 31.4 5 = 32.8%
8 15.3
'g cirier f"Y a
16.6
-9 3 0 902 32.85 18.25 29 37 1.00.
Note #2 j
E' + 5 g
f = 2.58 3
= 1000 = $850 = 6850 tx
' l Case A.
Cold Plar.t S = 51.41 S = 66.25 3 = 99.16 66.25 119 25 D + P.
- E' + H g
19 05
+95.5 3 = 119.c5 Allowale stresse.
a a
for A490 Anchor g = 0 + 0 + 1374 + 0 = 1374,k T
N notes x = 0 = 0 + 53?00 + 0 = 53200 h
M Case P.
Fat Piar.t S = 63 93 S* = 66 E5 F = 0 + 333 + 795 + 385 = 2013" 57.30
+56.71 66.25 119 25
~
D + R + E' + H 8 = W.29 S,= 119.25 d
n M
M = 0 + 19070 + 15070 + 0 =
33140 e
I 9.44 cantilever f =
k.69
-1.0) 0.62 2h 20.00 20.0S 1.00 k
k.05 g,, gy Bm f
=
tX Note 02 8
49 Esbed*
3 20
-19.3 13.00 32,40 eh 10.83
-5 54 0.81 22.00 20.00 19.40 1.b o
e' Nh Bracket Bea:n E' = 142" Note 42 Embed 3.47
-3.95 18.00 32.40 3
Vertical Parder 12.75 4.75 34.92 19.ko 100"/Iwp
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50 Each Support Notes:
1.
Allowable stresses are based on the minian.m yield strergth 4.
Allowable axial atress conforms to AISC specification from alti test reports. See Table 3 section 1.5 with a factor of 1.5 U.N.O.
2.
For nocenclature, see P.S.
5.
A11cvable shearing stress = 0.5 F.
y 3 Cochined stresses as noted are in accordance with ATEC 6.
Allowable bendir4, principal and tensile stress = 0 9 F*.
sixth edition specification section 1.6 squatlocs.
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ITEM 3:
Describe how all heavy section intersecting member weldments were designed to minimize restraint and lamellar tearing.
Specify the actual section thicknesses in the structure and provide deta.ls of typical joint designs.
State the maximum design stress used for the throughthickness direction of plates and elements of rolled shapes.
RESPONSE
A.
Upper Steam Generator Support The upper steam generator support system consists of two built-up welded girders with their web in the horizontal plane.
The girders are connected to the embeds,in the secondary shield wall by means of A490 high strength bolts.
Lateral movement of the steam generator is re-stricted by four impact stubs welded to the two main girders and at 45 degrees to the girder axis.
Details are shown on Drawing C-186.
To reduce through-thickness stresses, several methods were used.
First ' he tension member was extended through the cross member as in the case of the secondary shield wall embeds.
(See Section A and Detail 1 & 2.)
The second method used was to add reinforcement fillet welds to full penetration welds.
(See Detail 4 and Note 7.)
The thir6 method used was to use fillet welds only as was the case for the back plate in the embeds.
(See Section J.).
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The maximum design through-thickness stresses and the thickness of the elements associated with the T joints are tabulated in Table,2.
The typical joint designs are shown on Dwg. C-186, Detail 4 and Section J.
B.
Lower Steam Generator Supports A 3-1/2 inch liner plate was used to transfer the tension force from the anchor bolts to the embedded anchor plates.
The welds between the thickened liner plate and the 4-1/2 inch threaded sleeves are single bevel butt weld with a 1/2 inch fillet weld as a reinforcement.
(See Dwg. C-113, Section C.)
C.
Reactor Coolant Pump Supports Seismic lateral support for the reactor coolant pumps is provided by hydraulic shock suppressors connected to cantilever brackets attached to the secondary shield walls.
(See Dwg. C-184.)
The anchorage to the embeds is welded with double bevel full penetration butt weld to 2 inch plates.
(See Details 1 and 2.)
The cantilever brackets are attached to the 2 inch plate with either a single or double bevel full penetration butt weld.
(See Details 4, 5, 6 & 7.)
The through thickness stresses are shown in Table 2.
Due to the low stress levels, no special procedures were used other than visual inspection.
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- The. vertical load from the reactor coolant pump does not induce-through thickness stress in the vertical hanger support.
The' actual stre sses are shown on Table 1.
For details see Dwg. C-186, Section H.
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4 TABLE 2 T 500CH THTCKNESS TE*!STLE STRESS VAX. EIDENT TYPE OF STFISSES T-JOIhT DESCRIPTI0tt THICVIESS t
t ION CMESIM SM c
y NE t (IN) t (IN)
T C
V (ESI)
(KSI) g p
Girder Embed (Sect. A & B) 3 3
T 12.80 36.5 e4 Section C 2
2 or 4 C
V 36.5 a
p:8 Section D 2.5 3
T ua E.
2.50 36.5 3
DYu
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hes Section E 3
3.5 C
V 36.5
.a g
I d
Section F 2.5 35 V
36.5 m
H V
38.8 Section G O.643 0.5 f
m R
Section H 0.643 05 V
38.8 0
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Steam Generator Lower Support tug. C-187 35 4.5 T
16.08 48 5 Typical Section Detait 1 2
1.5 T
2.31 36.
f 4
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N' Detail 2 2
1.5 T
V k.6 36 2
-y ] '
lg 10.59 40.1
@M Details 4 & 5 2
1.25 T
80 o3 a
.f -
yh
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8M Detail 6 2
0.688 T
3 71 39 9 Detail 7 2
0.688 C
V 39 9 Note: Response to Item 3
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B ITEM 4:
Specify the minimum operating temperature for the supports and describe the extent to which material temperatures have been measured at various points on the supports during the operation of the plant.
RESPONSE
' No physiEa1 'easurements have been taken on the supports but it is estimated n
that the minimum operating temperature is > 500F.
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p-ITEM 5:
Specify all the materials used in the supports and the extent to which mill certificate data are available.
Describe any supplemental requirements such as melting practice, toughness tests and through-thickness tests specified.
Provide the results of. all tests that may better define the properties of the materials used.
RESPONSE
Materials used in the supports have been briefly described in the response to Item #1.
They are tabulated in Table 3 in further detail, showing the type of steel, thickness
-and the corresponding mill test reports.
For-A36 steel, supplemental requirements such as melting practice, toughness test and through-thickness test were rot specified.
Only those tests which wore required by the ASTM standard, such as chemical analysis, yield strength, censile strength and elongation, were included in the mill tast reports.
For A516 steel, Charpy V notch toughness and ultrasonic inspec-tion were specified in addition to those standard tests.
These two tests were conducted by U.S. Steel Corp., the steel manu-
.facturer, in.accordance with ASTM Standards A-300 and A-435, respectively.
The test results were deemed satisfactory,.
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TA9'E 3 FATEftIAL 19ED FtB STTXt i"J'F-ATC9 ELT* OPT & DE0*754 COctAf:7 TV'? FU'T0FT
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ACTUAL MIM Ttr~17 FESC3f?TICA FATI tIAL EIEEC EfICE MILL TECT EZPCRTS ATTACH:E.7 UElla S7alJGTA TE'4SI12 STED C7d SPECIFICATION DE GIZE (HIAT!;0.)
FrER
(*SI)
(KSI) reed Stiffener &
A36 1"
8T311400 7
L6.2
'75 9 ccannection Splice L &
A36 1 1/k" 801k7/,70 7
39 7 69.5 1
Side L of Girder T &
802303310 6
k2.0 71.0 B flarge p
2kW27 1
41.6 66.3 g
Connection Splice t A36 1-1/2" 801*te ein 2
42.2 71.8 3
Girder Web anJ Flange A36 2"
802F0l*23 2
48.0 77.0 h
& Stirrecer 2. of 450 602505200 4
3).9 74.2
",k Stuo 802107189 6
40.0 73 0 801Lo3770 6
41.0 65.0 i
517313))
10 49 0 7k.0 h
450 Etub Flarge Af 2-1/2" 802605150 9
39.5 73 0 z
2 U
Erwd 3, Tie Bar, A36 3*
51721k62 2
41 9 73.0 Stirrener E, and 831E08950 3
k3.0 72.8 k50 Stub Contact t 802208090 h
39 3 69 0 Sc2 03150 7
45.5 7k.6 802IWA70 8
k2.8 76.4 80n05t.ko 8
36.5 73 9 y
E01ro$r/0 11
- 41. 3 77 4 S0301S?O 11 42.1 76.0 t
k50 Stub Web A36 3-1/2" 6023)%70 10 kb.5 73.4 Girder Web Mear Embed A36 4"
801B0179) 3 41.0 70.5 SJ1E03770 9
42.7 73 9
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Liner L A516 3-1/2" 802896470 10 44.5 71.5 S k.
f-d ii Anchor Bolt A490 2-1/1" d 6k876 13 13?.'
150.5 o
EM of Vert. Henger A36 14'E 725056 5
38.8 59 1 Support Pracket Support
.A36 1k826 kk772 Ik k0.1 61.5 14's 87 fi93183 15 41.2 64.2 86
'# h Er.bei &
A36 1 1/2" iis "
1-3/4" Noc Available
- 36
- 58
(* g 2"
^
Cantilever Support A36 6C13 51532 18 41.2 62.5 3/4" 190511 16 43 5 70.1 1
1-1/4" E29373 19 40 3 69 3 if*
220007 17 50.7 79 7 i
p ITEM 6:
Describe the welding procedures and any special welding process requirements that were specified to minimize residual stress, weld and heat affected zone cracking and lamellar tearing of 2
the base metal.
RESPONSE
The welding procedures used for the steam generator supports and the reactor coolant pump supports were manual shield metal arc, manual flux core arc, semi-automatic submerged arc and automatic submerged arc, all in accordance with AWS Dl.0-66.
In addition to the controls specified in the welding procedure, such as preheat and interpass temperatures, weld balancing was used.
Intermediate stress relieving was specified in accor-dance with AWS D1.0-66.
Weld repair procedures for cracks which developed during fabrication consisted of chasing out the crack, gouging out the base material, buttering, rewelding, intermediate stress relieving and completing the welding using the applicable pro-cedure for the joint geometry.
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ITEM 7:
Describe all inspections and non-destructive tests that were performed on the supports during their fabrication and in-stallation, as well as any additional inspections that were performed during the life of the facility.
RESPONSE
Non-destructive tests were not specified originally except for the welds be' tween the threaded sleeves and the thickened liner plate as shown on Dwg. C-ll3, Section C.
Liquid penetrant examination was specified for these joints.
Two cracks and linear indications were revealed during fabrication.
The indications were removed by grinding and rewelded.,Reexamin-ation indicated the repair was satisfactory.
During the fabrication of the upper steam generator supports it was noticed by visual inspection that there was incomplete fusion in the root pass.
Subsequent UT examination revealed cracks in both the base gietal and weld metal for several joints.
' The cracks were repaired by using weld repair procedures and controls a.s, dis-cussed in the resnonse to Item 6.
Nondestructive examination consisted of U.T. to assure complete removal of, the cracks or inclusions. UT examination was also done after the' joint was rewelded. Magnetic particle inspection '
,was done on all joints.after all welding was completed.,
-?
A full time resident shop inspector was assigned to witness all welding NDT and heat; treatment.
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s PART II RESPONSES TO ITEMS OF CONCERN ON.
FRACTURE TOUGHNESS AND POTENTIAL FOR LAMELLAR TEARING OF STEAM GENERATOR AND REACTOR C001 ANT PUMP SUPPORT MATERIALS ITEM 1.
' Drawings:
No.135050E - Assembly and Detail of Support Skirt No. 139754E - List of Material / Steam Generator No.135030E - Shell and Tubesheet Attachment Assembly ITEM 2:
Design Emergency Faulted Load (FY Kip; Mr Kig in.)
1995; 84, 756 2155; 119, 076'~
2155; 406,356 Max. Stress (Kip /in')
- Neg.; *Neg.
.24; 1.13 6.5; 30.4 2
Allowable (Kip /in )
.SSm or 11.3
.5(Sy) or 18
.5(1.2Sy) or 21.6:
.5(1.5Sm) or 17.5(1.5Sy) or 27
.5(1.8Sy) or 32.4 -
- Indicates that the intersecting members are in a comprehensive state.
ITEM'3:
No major loading on the steam generator skirt to show cause for concern about Lamellar tearing. See Item No. I for structure details. See No. 2 for design stress.
ITEM 4:
No physical measurements have been taken on the supports but it is estimated that the minimum operating temperature is > 50 F.
0 ITEM 5:
The materials are listed ih the List of Materials Drawing, a.
b.
The following Mark Nos. have Mill Certification information available -
MK-96 and MK-97.
This support material was ordered in accordance with the requirements of c.
the ASME Code,Section II.
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. ITEM 6:
a.
Weld No.
Process Preheat (Min.)
61 Submerged Arc 4 Manual Arc 200 F 64 Submerged Arc 200 F 0
65 Flux Core 200 F 0
70 4 71 Flux Core 200 F 0
b.
These welds have received a full Section III stress relief.
ITEM 7:
The Non-destructive Examinations are shown on the Drawing Nos. 135050E and 135030E.
The preceeding tabulation provides the information requested by your letter.
Note that our effort has been concentrated on the attachment of the gusset plates (MK-98) to the base plate (MK-97), as these are the only areas that are subjected to loads which might cause Lamellar tearing. The intersecting member of the gusset plate (F1K-98) to the base plate (MK-97) is analyzed in 4
Section 31 of the contract OTSG' Stress Report. This analysis is.used as the basis for the preceeding tabulation of stresses at the intersection of the pieces.
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