ML19248D344
| ML19248D344 | |
| Person / Time | |
|---|---|
| Site: | Zimmer |
| Issue date: | 06/04/1979 |
| From: | Maura F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19248D324 | List: |
| References | |
| NUDOCS 7908150637 | |
| Download: ML19248D344 (22) | |
Text
.
UNITED STATES OF AMERICA NL' CLEAR REGULATORY COM".ISSION BEFORE THE ATOMIC SAFETY ANL LICENSING BORAD In the Matter of
)
)
Docket No. 50-358 CINCINNATI CAS & ELECTRIC COMPANY
)
)
(Wm. H. Zimmer Nuclear Power Plant)
)
)
DIRECT TESTIMONY OF FEDERICO A. MAURA RECARDING CONTENTIONS No. 15 and 16, CONTROL RODS THICKNESS AND SEALS State of Illinois
)
)
ss.
County of DuPage
)
Federico A. Maura, having first been duly sworn hereby states as follows:
I am employed as a Reactor Inspector in the Reactor Operations and Nuclear Support Branch, Region III, Office of Inspection and Enforcement, Nuclear Regulatory Co==ission, Glen Ellyn, Illinois. My educational and professional qualifications are set forth immediately below:
Education B.S.
Electrical Engineering, Virginia Military Institute 105o.
I have rc:eived two certificates from the General Electric Company coverin; Fundamentals of BWR Operation and SWR Technology, plus one certificate of qualification from the USNRC regarding BWR Advanced Technology Course.
In addition I have held
'z9 0 815 063 ].
G%060
o
-2_
Operator and Senior Operator Licenses issued by the Atomic Energy Coc=ission for the operation of BWR's designed and built by the Allis-Chalmers Company.
Experience I joined the USNRC in November, 1970 as a Reactor Inspector.
In this capacity I have perforced inspections of power reactors during the preoperational and startup testing, and operational phases to ascertain conformity with design and other criteria; observed and evaluated the adequacy of licensees' controls and provisions for overall operational safety; management's organizational control, procedures and practices, and their relation to the safety of operations; and the status of com-pliance of licensees with licensee provisions, rules, orders, and regulations of the Coc=ission.
I have assisted in specialized inspections of BWR's during the construction phase because of =y knowledge on the specific subject.
Prior to joining the Commission I worked eleven years for the Allis-Chalmers Cocpany, Nuclear Energy Division in the design, testing, and startup of their 3WR's.
I held the position of Site Manager for the La Crosse Boiling Water Reactor Project.
L'-Tik:f
. From 1957 through 1959 I worked for the Duquesue Light Company at the Shippingport Atomic Power Station as Test Engineer during the preopera-tional testing and startup of that facility and later as Shift Reactor Engineer during the operation of the facility.
The Miami Valley Power Project has raised the following contention:
Contention 15: Control Rods Control rods which cust be easily inserted into and removed from the reactor core have been inadequately manufactured so that they do not meet the size specifications for such control rods.
Prior to installation in the reactor vessel the control rods (See Attachment A) are inspected to ensure no damage occurred during ship-ment and/or handling at the site.
This site inspection was perfor=ed during the period July through October, 1978 by Reactor Controls, Incorporated and consisted of several visual observations as well as measurements as outlined by the checklist used.
(See Attachment 3).
The use of the two envelope gauges and the determination of whether to accept or reject a control rod was controlled by General Electric Co:pany, Document No. 22A4387, Revision 4, " Control Rod Handling and Inspection" bObOUf
s (See Attachment C).
The 0.280-inch envelope gauge is used to deter =ine if the control rod blade thickness, at any one point in the length of the blade, exceeds 0.280-inches.
The gauge looks like a tuning fork
~
approximately 1-inch
..>ide.
The 0.320-inch envelope gauge is similar to the 0.280-inch gauge except it is approximately 1-foot long and is used to determine if bowing exists over a wider area.
During the initial site inspection, conducted by Reactor Controls, Incorporated, of the 137 controls rods 86 did not pass the 0.280-inch thickness envelope gauge. Of those 86 that did not pass, 4 also did not pass the 0.320-inch gauge used to locate undesirable bowing and the four were rejected.
In accordance with the GE Inspection Procedure (22A4387) a 40-pound force clamp was placed against the blade sheath, adjacent to the high area of the re=aining 82 control rods. The purpose of the cla:p was to deter =ine if the local sheath bulge was flexible, and to ensure the absence of foreign matter between the sheath and the poison rods which form the blade. According to the design engineer the use of 40-pounds was socewhat arbitrary since his main interest was determining the flexibility of the bulge.
Seven control rods did not pass the 0.280-inch gauge with the 40 pound force clamp applied, one of which was rejected and not placed in the reactor.
The remaining 6 control rods were accepted by the licensee after the General Electric Inspection Procedure was clarified to indicate that the clamp could be placed over b.3.5){ fj.]
the high point in question, and the surface araa of interest on a con-trol blade was redefined. The clarification was done through the issuance of Field Deviation Disposition Request No. KN-1-286 (See Attachment D).
Because the Reactor Controls, Incorporated and the General Electric Company (FDDR No. KN-1-286) records were not clear as to how the final results (i.e., acceptance of the 6 control rods) were obtained, the NRC inspector requested a reinspection of those 6 control rods. The 6 control rods were reinspected April 10, 1979.
The reinspection was witnessed by NRC inspectors.
The worse case found was for control rod Serial No. A510 which on blade No. 2 has a high spot near the edge of the blade which is approximately 0.300-inch in thickness. With the clamp placed adjacent to the high area the thickness decreased to approxi-mately 0.285-inch, and with the clatp over the high spot the thickness was reduced to 0.280-inch or less.
(Refer to Attachment D for location of high spot).
For comparison it should be noted that each control rod has an upper guide roller on each blade (See Attachment A) with a thick-ness of 0.333-inch.
Attachment E shows a typical core cell. According to the Wm.
H. Zimmer FSAR the gap between two fuel asse:blies in a cell is 0.522-inch which translates into a gap of approx 1=ately 0.120-inch between the fuel asse:bly and a control rod of 0.280-inch thickness.
Control Rod No. ASIO reduces this gap, in the small high spot area, by 69UON
.~ less than 10%.
On the same date, a meeting was held with the manufacturer's design engineer to discuss the use of 40 pound force clamp and determine
~
the actual control rod blade thickness acceptance criteria limits.
During the meeting, the inspector reviewed a test report containing the results of control rod qualification tests performed by r's manufacturer for the pu pose of determining the degree of misalignment, channel deformation (water gap reduction), etc., at which operational performance of the blades would be affected.
During the qualification tests, the control rods were cycled (scrammed, withdrawn, and inserted) for the expa d 20-year design life of the blades, and the wear of the blade sheath and fuel channel was measured.
Based on our review of the inspection records, the reinspection of 6 control rods, the results of the General Electric Company control rod qualification tests, the discussions held with the =anufacturer's design engineer, and the knowledge that the systems will be preoperationally tested prior to fuel loading, it is our conclusion that the control rod blades presently supplied to the Zimmer Station are satisfactory because:
1.
The use of 40-pound force clamp was =ainly to determine the flexibility of the blade sheath's high area and ensure the absence of foreign matter between the sheath and the poison rods which form the blade.
2.
The 0.280-inch gauge was a control placed by the design engineer to ensure he was consulted and could analyze any deviations from the desired des.gn parameter.
3.
The blade thickness acceptance criteria could be increased above the maxi =um thickness measured at the Zic=er Station before the first operational dif ficulty would be experienced.
4 Preoperational testing of the control rod drive syste= will demonstrate whether any problems exist which affect the operational performance of the control rods.
The Miami Valley Power Project has also raised the following contention:
Contention 16: Control Rod Seals Almost all of the seals on the control rods, which when properly set prevent radioactive water from leaking out when the reactor is shut down for =aintenance, do not meet =iaimum specifications for smooth-ness.
Rough seals cannot set properly, making servicing more difficult and unnecessarily endangering workers and the general public by causing leakage of radioactive water.
t:n.
t.sJ O ;; g 9
The seal area (See Attachment F) is a machined surface on the control rod bottom casting velocity limiter.
When a control rod drive unit must be removed for maintenance the control rod is fully withdrawn at which time the seal area on the control rod seats in its guide tube, similar to the sealing done in ones bath tub.
With regard to the condition of the seal surface on the control rods, the licensee initially rejected one control rod because of nicks on its seal surface, and it was not placed in the reactor.
The NRC inspectors looked at the seals on the 6 control blades inspected on April 10, 1979, and found them satisfactory.
Based on the inspector's review of the licensee's inspection record, it is our conclusion that the seals on the 137 installed control rods are satisfactory.
As stated earlier the purpose of the seal is to permit the removal of a control rod drive for maintenance while the reactor is shutdown and depres-surized without having to removed all the fuel and drain the reactor vessel.
Since these are not perfect metal-to-metal seals licensees must be prepared to expect a small a=ount of leakage until the drive is re=oved at which time a blind flange can be installed on the control rod drive housing if needed. This small leakage may create an inconvenience to =aintenance b.#CO fj7
_9_
personnel during removal and subsequent reinstallation of control red drives but, in no case, does it create a safety problem.
s Attachments: Attach =ents A through F
'Yl !< 4)
, // In sL
~
Federico A. Maura Subscribed and sworn to before this.)u1.al day of June me 1979.
3_Si
' ^* b t' (/
Notary PublIc My Commission expires: /-f-[O Y OUOaB
i ATTACH",ENT A 9.7$ in.
['
y HANDLE
/
s.-
/
upper C'Jt0E MLL E8 s
/
4 0
TYPIC A 4 PL ACES L5 in.
o TELDED END PLUG x
1 4'
\\
O
% ol l9 l #
NEUTRON ABSORBER o
%l RODS SHEATH I
hl p
9 I
h Y
g 143 n.
O i
b i
h 1
16 in. BET? EEN BLADE
- BALL 5 s TYPICAli p
- ~:
U O
l p
I 5
V L
hf is l
h~
COUPLING W
/
t # RELEASE HANDLE
' f f
r 2
ij VELOCITY LIMITER gs
/g U
- NOTE-E PLUG
/-
la ABSORSER R0051G5'8 in.
i l
l ABWRBER ROD LOWER CuiCE ELLER COUPLING SOCKET 630009 Figure 2.J-7 Control Rod Page 1 of 1
ATTACHMENT B Reactor Controls, Inc.
CO:Trilo!.. ROD I:I.7PT" IOU Ci!T!1IST SI2L1L #
1.
V E IFY THAT ALL ROLL?3S RUli FREE, 2.
V~RIFY THAT LATCH ITUT IS TACKED.
- 3. - VEIFY THAT LATCH IS FREF TO TRAVEL.875" MI:!II'1EI.
1.4 ~ VEIFY DISTA~.IC ' FRCI: TOP OF ROD TO BOTTCH OF LATCH IS 2.629" ::I:!IiHJM.
- F.
"v
.. T _ v. r v: a-t. s i n_- u _ n.,u~ ~.m. 0.u. _o 0 _ r:n.-_ va 0.m C G a _.'60s e
s r
R0D TO DOTTO:I 0F SOCH.72 IS.803" MINIIRJM.
6.
IllSPDC' S"?AL SURFACES FOR HICKS, SCRATCHES, ECT.
7 VACCuu wTIR' LEIOTH OF 3'sADS.
A
\\r_--,T. r.r. ~n....v.,..,
C.s.._._,3
,a, c -n n
- m.. a a.u. nu. rT3, n-
.7..m..C22/E a_
mm..Io.
. ep.
m 7
v.
a
~
CHICK APPRO?RIATI LIUI IF ACCZPTABL2.
BLAD' d
.230" GAG 7 320 GAGE' 1
2 3
T
,a.,q n, 7.-.,-
--.n C n m ;
.a
- r.... m C =....u -..., 1 r, s
.y. C _.,S Sr e..,.
7_...s C R u-. >.,0 1..
a.~
a_.
.a
- v. 3. u 5 n.--... -- n 0 -
e
,,. _.. C..,, a-..
,a
..w
.-w a
C ON~.??S :
I l
IHSPETID 3:
DNI E:
!.!1 NOT2: 1:UH3"RI:IO OF SLADT3 CC.G OL ROD SEIAL y
' JILL 32 PE SK?Cii.
-r w
i
- r t, c
r2 e
6
- N CONTROL ROD H/0'DLE jt3 f ;%._ Q. ><<0
-v q
Q.- i{
~
Y 0
n/,
j Page 1 of I i
t
' i
^
EIS IDENT:
CONTROL R00, HDLG & INSP REYlSION STATUS SH EET GENER AL h ELECTRIC -'
NUCLEAR ENERGY OlvisiO N MPL - S11/313-3070 SPECIFICAT:0N ORA *:NG Y OTHER TYPE INSIAll AIION SP EC ' # 7 C 27 "
DOCUMENT TITLE CONTROL ROD. PaNCLING ANO PlsoECTION LE GEN D-REVISICNS ARE IDENTIFIED C A SPACE (,', )
R EYl51CH 5 lC___
tem 0
DMF-817 1
Per ECN NE 55523.
Sheets affected:
2 and 3.
Revision identified by a scade, ( Q).
~
g1 w.trei..~
Y 7g j
^W
- Y.
W. ritiv5e'%
APR 10 E5 I rk 2
Per ECN NE 38663. Sheets affected:
2 and 3. C,fvisicns identified with R
g o,q y j g p c q
- 55**O "
JUL 9 E5 vc.s. s. zi...sta pp.c;t::
t-Od uta3 "";
r f '. -AF p M J t
lj; 3
Per ECN NE71043.
Sheets 3 and 4 affected. Revisions identified witn a 5::ade ($$y} gg
\\
f2 1
CH KD EY :
.u - ['7 u
A Long k'.C'c m d
- 190 t
4 G.J S c:ra NN' b -,0
$q' 377 L'
t
})
ECn NE?9:33 -}
l 9
(
. s
.. ~
CH KD BY :
- CN f f
M' ps f
l cEics:pTics er cs:s l FONTS 70'N
-. _.....Lm. ~ ~,-><
- y - -U
....=,..m. J/,
"7
- t -
nn
.m
,y Ti%Mrw3 2 i G5 P "s C.s 1
UN071 Iz e n * = < x... c,
sw s.m
ATTACHMENT C 4
)
G E N E R A L Q1 E L E C T R I.C 22A4387 s. ~o.
2 NUCLEAR ENERGY DIVis!ON nev.
4 1.
SCOPE 1.1 This document specifies the minimum basic installation requirements for a Control Rod into a Reactor Assemoly.
$ 2.
APPLICASLE DOCUMENTS 2.1 General Electric Dccurents. The following listed documents form a part of this specification to tne extent specifieo herein.
2.1.1 Suctorting Documents a.
Control Rod Handling Method
- 767E711 b.
0.250 Inch Envelope Gage
- 131C9192 G1 c.
0.320 Inch / Foot Enveloce Gage - 131C9193 d.
Inspection Clamp
- 166B8514 e.
Del eted f.
Del eted g.
Deleted h.
Del eted 2.1. 2 Surtl emental Cccuments - None 2.2 Codes and Standards
- None 3.
DESCRIPTIG!,
3.1 Instructiccs specified herein delineate the inspection and handling repuire-ments wnica are necessary to ensure that the Control R0d will be properij hardled and ins talled into tne reacter. Tnese requirements form a basis for the CE ins tallation section of the Reactor Asserriy Specification and the Reactor Asse cly for a particular project.
4 REQUIREMENTS 4.1 General 4.1.1 The Control Red (CR) shall be handled in a manner that will protect it from damage during handling into tne reactor vessel.
<J.n._ u: ;,,,.
- 1
ATTACILTC C 4
GENER AL h ELECTRIC 22A43a7 s. ~o.
3 NUCLEAR ENERGY DlVislON 4
A E V.
4.2 Insoection 4 4.2.1 Inspect 0.280 inch blade thickness by passing the 0.2S0 inch envelope gageJ(131C9192Gl) over the area indicated in Figure 2.
The 0.280 inch thick-ness is a point requirement. Therefore it is acceptable to use a deep tnroat micrometer to inspect any points that do not pass the gage.
D 4.2.2 Del eted.
-)4.2.3 Inspect 0.320 inch blade envelope for the full length of the sheath by passing the 0.320 inch envelope gage (131C9193) over the area indicated in Figure 3.
4.2.4 If gages do not pass over the sheath, due to local sheath bulge, then use inspection clamp (166535'.1) adjacent to inspection gages.
If gage still dces not pass over sheath then the control rod is rejecuble.
4.3 Handl ing 4.3.1 Handling equipment and its use is described in Drawing 767E711. This equipment is used to take the control rod out of the shipping container and prepare it for installation.
4.4 Data 4.4.1 After inspection is complete fill out data sheet (Figure 1) and return tu responsible design engineer through the project engineer.
U.3 oC7,3
,. UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSIOS BEFORE THE ATOMIC SAFETY AND LICENSING BOAPS In the Matter of
)
)
CINCINNATI GAS AND ELECTRIC
)
Docket No.30-358 COMPANY, _et _al.
)
(W:. H. Zi==er Nuclear Power
)
Station, Unit No. 1)
)
CERTIFICATE OF SERVICE I hereby certify that copies of the following documents have been served upon the following by deposit in the United States mail in Glen Ellyn, Illinois, this 5th day of June, 1979:
(1) Direct testimony of Tho=as Vandel Regarding the Pressure Testing of Doors; '2) Direct Testi=ony of Thomas Vandel and Harvey Wescott Regarding Contention 14, Cable Tray Welding; and (3) Direct Testimony of Federico A. Maura Regarding Contentions 15 and 16, Control Rod Thickness and Seals.
Charles Bechhoefer, Esq., Chair =an Leah S. Kosik, Esq.
Atomic Safety and Licensing 3454 Ccinell Place Board Panel Cincinnati, Ohio 45220 U.S. Nuclear Regulatory Ce==ission Washington, D.C. 20555 W. Peter Heile, Esq.
Assistant City Solicitor Dr. Frank F. Hooper Room 214, City Hall School of Natural Resources Cincinna ti, Ohio 45220 University of Michigan Ann Arbor, Michigan 48109 Ti=othy S. Hogcn, Jr.,' Chairman Board of Cc==issioners Mr. G.cnn 0. Bright 50 Market Street Atomic Safety and Licensing Clermont County Board Panel Batavia, Ohio 45103 U.S. Nuclear Regulatory Co==ission Washington, D.C. 20555 John D. Woliver, Esq.
Clermont County Co== unity Council Troy B. Conner, Esq.
Box 181 Conner, Moore and Corber Batavia, Ohio 45103 1747 Pennsylvania Avenue, N.W.
Washington, D.C. 20006 G%074
William J. Moran, Esq.
Atocic Safety and Licensing General Counsel Appeal Board (5 copies)
Cincinnati Gas & Electric Company U.S. Nuclear Regula: cry Co==ission P.O. Box 960 Washington, D.C. 20555 Cincinnati, Ohio 45201 Docketing & Service Section (4 copies Atomic Safety and Licensing Office of the Secretary Bocrd Panel U.S. Nuclear Regulatory Cocsission U.S. Nuclear Regulatory Cocmission Washington, D.C. 20555 Washington, D.C. 20555 2
l 4
- //
C.q /z
" it.s4x v u.w Catherine Holahan June 5, 1979
'!}l].ff',*.'[;
ATTACHMENT C GENERAL @ ELECTRIC 22A4387 su ~o.
5 NUCt.E AR ENE RGY DIVISION REV 4 final
)
\\~
/
O e
O O
O o
O O
o O
O O
o o
oo oo o e g
c O
O O
O y
o o
o
{/~ [/{E W f / /o
/ ~4 { IT g
/l f
e c0 0
0 o
o o o 9 9' o \\ /0
/
/ /
/)
/
/.,
/ P s / /,
/
,. _.68 mx h* 280 inspection area 3.06 Max. *
^
(4 blades)
$ Figure 2
\\
C C
O O
[
O C
o C
0 0
O O
O O
O O
O C G f
C O
O O
C O
'l
' l/
l
~ '
s // f//
/,\\
c
/ /
/
/
C C
\\
C C
O i
l C
C O
e e
C y,
l I
O C
f /gi,
\\
// / / C/ // /
/ E/
'/
l......
G
.320 inspecticn a~'
F$
(4 blades)
.) Figure 3 1
e
ATTACHMENT C 22A4387 SH. NO. 4 REV, 4 CO \\ TROL ROD EN Vi_0 3E
^
\\S ECT ON QR S/N ACCEPT REJECT REASON (L'LI8Me) i I
I I
i l
i I
I i
I i
I l
l l
l l
l 1
I l
I l
l i
i Ijhbh,['[
ATTACH'ENT D
-. ~. --
W. b.! 1 ; w;: 8;.E"liin a
_ _ z. _-
FIEl.D DEVIM10N DISPOSITION REQUEST tsuso anoa gc l
m
~-
NUCLEAR ENERGY DIVISIONS L 75/
scoR NO.
KN-1-286 1
SHEET OF Zi==er I pg a,E CT 9/5/78 J DATE ORIGINATED TELECON OR TWX APPROV A1 SY:
P.
P e n t ::
9/5/78
.a...
a.e EQu1PVENT (CEStRIPTION AN0iO A MPu B13-D009: Contro1 Roc owG. SPEC. ETC. NO.
l SHEET NO.
A EV NC.
OWG SPEC. ETC. TITLE 814E935 4
Outlinet Control Rod DESCRIPTION OF DEVIATICN:
IMPACT CLASSIFICATION The control rods were inspected as outlined by 22A-CATEGORY
/
4387, R4 with the following changes:
/d
-4.2.1 Reduce the 0.280 inspection area per Fig A.
CATECORYli O
-4.2.4 For the following control rods, the ir. spec-PRIORITY CLASS!FICATION tion clamp (16638544) was placed directly EMERGENCY O
on the high point / area (0.280) and deep URGENT O
Mj4
/.
throat microneter readings (Table 1)
.a s ROUTINE outlined by Fig. B was used as the basis for acceptance: S/Nst A 5 01 -:.A 515, A 4 4 0, A 4 5 3,
APPgCVED 8 -
and A420.
8O**
2N N DESIGN ENGIN1HER CATE A 0.290 high spot was established on control rod, l S/S: A510 as detailed in Table 1.
O hg c y,N hy
/ch>7<
SITE IMPACT: Engineering concurrence required MAT'l S APPL ENGIMEER prior to WK.39/70.
SUGGESTED DISPOSITION 2 Md Engineering is to evaluate.the inspection procedurt changes and advise.
/
Accept control Rod S/N: A510 as is.
y,7, p ggg PSCJECT Cost Accounting: GE Responsibility.
4 E' J ANAGER5 is,s 3-/ 19 FIE LO '.i AN A G E 9 DISAPPRCVED BY:
,?"
- ,t DATE
%./
-3 SUPPLEMENT INST AECuggg; t'
YE0 NO NUMaga FIN AL CIS*OSITION I
~
0 K "*'
SM M E
$~ S' ASW E -
1N ALL A R.E4 3 C;STRIBUT'CN COC E f&o w 4 jy ;~lG* A Co u T2. 2 t. 2 0 0 / CHANNEL t WTERFe&%
INTERN AL EXTERNAL ud o L.t uo r o cc 08 Pe> J 2 7~o E A/ 0 0 f CM A w~'c.
l / ;= 5, y 5-St}n : IS T.R U[ 50 2. f/A/5~/O-YHE g$ e yp -nS C L Al~17 dS D/%UC$EO A 50 Q 4,2
/
iM C2:'T y
~7~ M /
"7-W I 9 2.2QJg,7 J.
OW PJW W WEO
/
YES O IDENTIFY No E
,r<,
sArETwREuAsiury Review RECutR eo:
vEs O NOR i "E'o 7 oa' ca ER No-STATEMENT b UE M M'
/
h 7' f or f/, U VI E w Vo T g gn17,,
[,
- __, [.,Eq E
C mp3;z bWw mM r"=cF'?'*A *Ei n/r n)
- m....,., m
FDDR EN-1-286 page 2 of 2 ATTACletENT C
~
o
- /
J f
[h
\\
\\
\\
N N
N l\\
\\\\\\\\d
- 74,5 "zrn r A use n:u ~
asssa.
f c ~ c-
.y s. a,,,,
OWN (sc-4,urs)
~b U I N ? A N N / I E A'
/g
~l
\\
(UMW bYS
-~
Af,
5 "I
/
/
l 2
m i
a 9
U
=
ua.n.
aI
~
u
'~
_X Y
N 6
t, b
}
!/A'T E l
1 TABLE 1 CTO=tter l
k
- h..
Reading s'ontrol Rod S/N Bladel H
G X
Y z
ich Point Chance F l
A420 4
29 1
2 1-1/8 0.292 0.017 A440 2
28 3/4 3
\\
0.296 0.018 A453 1
5-3/4 1-5/32 2
1 0.290 0.026 A501 2
l3 5/8 1/8 1/8 0.282 0.008
~
ASIO 2
5 7/y -
h b
1 D.290 S
A515 4
12 9/16 1/8 1/8 0.282 0.006
,qr, t
f
- : ;o
, t.]
p accuracy of i 1/32 G
C'
' " ~ '
!!21
~
(q)lg } ;;[.)
difference between cla= ped-free condition
- location is basis for accept as is
ATTACHMENT E C. 2 $ I in e "
C+
/0000000O)
(OOOOOOOd f' 0 0 0 0,0 0 0,Q i ~O O O O O O O O DOOO0000 00000000 000@O000 00000000 00000000 O
00000 D0000000 O
00000
'_ 00000000 O O, 0000 OOOOOOOO,,y QOOOOOOg s
O g
Yh 1 k 9 02O O'o @,0 0 @ @
~
i h
RODO.D.
o.493 in.
cuo 1sicmss o.034,.
su u m u r O.o.
e.,m.
gggggggg p@@@@@@@@j
,ms,0, q
$ a^o'us j
g 4
A 5
h WATER ROD Typical Core Celi G 3' 3 0 S O
k f
m Acu e r r e
f r
3 e=ma c.
'd)
[u,.,. -
v l
{
I I
/
N js l
l l
l l
l
\\
\\
I E
l L.iise t5
~
r i".a.h 8
s-s e
r p
i e
n m =O e
C. O
- f l
5 -,
p,isc l
9 b.u.~ g:
z' =
a <
. m. j<
-r--
r
~ 4c, c
m I
D i
1 is 9
i
- 2. m L
gg
<I.
3 i
.,, y lh l
'Lt.'.; O.. s.3..*.
(
i.!
W b y
h H'-
Page i et A
>