ML19352A992
| ML19352A992 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/29/1981 |
| From: | Novarro J LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SNRC-578, NUDOCS 8106020518 | |
| Download: ML19352A992 (85) | |
Text
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LONG ISLAND LIGHTING COMPANY FACO SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY ROAD
- WADING RIVER. N.Y.11792
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May 29, 1981 SNRC-578 s
N 16?
h3 Mr. Harold R.
Denton, Director N
b Office of Nuclear Reactor Regulation D
bt M'
U.S.
Nuclear Regulatory Commission
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Washington, D.C.
20555 y
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Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322
Dear Mr. Denton:
Forwarded herewith are sixty (60) copies of LILCO's responses to the Safety Evaluation Report (SER) Outstanding Issues listed in Attachment 1.
Please note that our responses to Outstanding Issue Numbers 9,
" Environmental Qualification" and #57, "TMI-2 Requirements" have been forwarded to you under separate cover via letters SNRC-576 and SNRC-579 respectively, dated May 29, 1981.
Very tguly yours, O
N OW P.
Noyarro
' Project Manager Shoreham Nuclear Power Station - Unit 1 CC/pd Enclosures cc:
J.
Higgins THIS DOCUMENT CONTAINS P00R QUAUTY PAGES
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ATTACHMENT 1 - SER OUTSTANDING ISSUES Number Issue 20 Appendix G - IV.A.2.a - Nil Ductility Temperature 21 Appendix G - IV.A.2.c - Pressure Temperature Limit 22
- Appendix G - Impact Testing 23 7ppendix G - IV.B - Minimum Upper Shelf Lnergy 35 Containment Isolation 37 Secondary Containment Bypass Leakage 38
- Fracture Prevention of Containment Pressure Boundry.
51
- Fracture Toughness of Steam Line and Feedwater Materials 52 Management Organization
- The information provided in this response supplements the information provided in SNRC-566, dated May 15, 1981.
+..
The information contained on the following pages is provided in response to the staff concerns identified as Shoreham SER Outstanding Issues 20, 21, 22, 23, 38 and 52.
This data provides an adequate basis for resolution of the staff's concerns.
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SHOREHAM REACTOR VESSEL BELTLINE PLATE AND WELD INFORMATION 1.
Available charpy V-notch and drop-weight NDT data are presented in Tables 1, 2, and 4 for Shorehrm beltline plates and welds.
Table 2 gives supplementary transverse charpy results which were determined for one of the Shoraham surveillance plates.
Table 3 shows a typical test certificate for a Shoreham beltline plate.
Table 5 verifies the location of weld materials in beltline weld seams, as well as verifying what data is not available.
2.
The beltline layout is shown in Figure 1.
This gives plate heat numbers and locations, as well as weld seam locations and identifications.
3.
Copper and phosphorous values, to estimate the effects of radiation on toughness, are presented in Table 6, when available.
It can be seen in Table 6 that there are cases for longitudinal weld seams where copper contents are not available.
This has been verified as shown in the letter from the vessel vendor (Table 5).
4.
Estimated starting (unirradiated) RT values are given in ND Table 6.
Theyareestimatedbyusingt$edatainTables1and 4 in accordance with GE procedure Y1006A006 which meets the intent of ASME Code paragraph NB-2300.
This procedure is l
explained in paragraph 5.2.4.3 (Attachment A) of the SUPS-1 FSAR, Rev. 18.
The data base for this procedure is further clarified in response to Zimner (ZPS-1) Q 121.15.
5.
Estimated end-of-life (EOL) RT NDT values (for 1/4 thickness location from the vessel inside diameter, as-built dimensions) are given in Table 6.
These estimations are in accordance with NRC Regulatory Guide 1.99 Rev. 1.
Where Cu content analy ses are not available, maximum RT NDT shift (AkRT NDT) values are conservatively assumed in accordance with Reg. Guide 1.99 Rev. 1.
This results in limiting EOL RT ND. values for the longitudinal weld seams in the beltline. Note that the Shorehan vessel sur-veillance program is designed to represent longitudinal weld seam 1-308 J, although the exact same weld materials were not used.
This will be clarified in more detail in the description of the Shoreham surveillance program in this response.
6.
Charpy V-notch upper shelf toughness was not a requ'i~rement when the Shoreham vessel was manufactured.
Thus, such data is not available for the Shoreham beltline welds, but is available for the plates as shown in Tables 1 and 2.
A very conservative assumption of 65% factor on longitudinal upper shelf can be applied to the results of Table 1 in order to estimate transverse orientation upper shelf.
(Table 2 shows that an 85% factor may be more accurate than the con-servative 65% factor of MTEBS-2).
The factor of 65% would result in a longitudinal requirement of 115 ft-lb, in order to meet the 10CFR 50 Appendix G value of 75 ft-lb, transverse i
upper shelf.
This value is met by all plates in Table 1 except C4806-2, which has an average of 107 ft-lb.
- However, i
since the Cu content of this plate is only 0.15% (Table 6) a reduction of upper shelf of only 20% is conservatively predicted by Reg. Guide 1.99 Rev. 1.
Combining these 2 con-servative factors of 65% and 20% results in an initial longitudinal upper shelf value of only 96 ft-lb, to meet the goal of 50 ft-lb. transverse upper shelf at EOL.
This value of 96 ft-lb., as calculated in the following equation, is exceeded by plate C4806-2.
50 =.65 (L)
(.20) [.65 (L) )
(where L is the longitudinal upper shelf value at start of life)
As seen in Table 4, upper shelf toughness values are not available 7
for Shoreham welds.
However, all charpy results at the test temperature of +10' are in excess of the 75 ft-lb, value for the E8018 weld materials.
Thus, they should not be a concern.
The charpy values for the submerged arc weld materials in l
Table 4 do not meet the 75 ft-lb. level in 4 cases.
- However, l
it is expected that further testing at higher temperatures woald have revealed an upper shelf in excess of 75 ft-lb.
Evidence in this respect is presente6 in Tables 7 through 11 which show weld procedures and upper shelf toughness results for similar materials.
A1. upper shelf (a-100% shear) results in Tables 8 and 11 are in excess of 75 ft-lb.
These welds are considered to be representative of the Shcreham welds since the welding processes (generally tandem wire submerged are for the bulk of the welds), post weld heat treatment, and weld materials are similar.
Particular attention should be given to the LaSalle 1 results, since these welds were made by the same vendor (Combustion Engineering) and with the exact same weld procedure (Table 7) in many cases, as for Shoreham. The material 1P357A/"
l 3958 in Table 8 is in both the LaSalle 1 and Shoreham surveillance programs and was prepared by the weld procedure in Table 7, l
which was used for the Shoreham longitudinal limiting welds.
I Futhor testing of these baseline surveillance specimens gave an upper shelf of 110 ft-lb. as shown in Table 8.
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7.
Drop-weight NDT values for the Shoreham weld materials were not determined by testing.
However, evidence for a con-servative assumption of -50*F is four.d in Table 8, based on the LaSalle 1 results.
All values of NDT are -50'F or lower.
Futher results in this respect are also shown in Table 11 (CBIN welds) and verify NDT valuey of -50*F and lower, except for one case.
This case (lP6484/0156 fer Laquna Verde 2) is considered to be nonrepresentative of Shoreham, because of the relatively low charpy test value (17 ft-lb.) at +10 and O'F for this material.
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j SHOREHAM REACTOR VESSEL NON-BELTLINE INFORMATION 1.
Limiting RT NDT values which eff te.t vessel testing and operation are shown in the FSAR (paragraphs shown in Attachment A).
The estimation rocedures for these RT values are in accordance with CO procedure Y1006A006, NDT and are also explained in the FSAR.
As with the beltline, the data base for this procedure is further clarified in response to Zimmer (ZPS-1) Q121.15.
2.
A sentence has been added to the FSAR (Attachment A, paragraph 5. 2.4. 3) to clarify further that these are limiting values for the vessel.
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CONTRACT NO._3067 VENDOR _ Inkens Steel Company JOS NO. V-70194 003
' HEAT No, C 4882-1 CODE NO._E*" "~1 RATERI AL DESCRIPTION _262-1/4" x 153" x 6-1/2" Lower Intermediate Shell MILL CHEMICAL ANALYSIS nw e
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.56 MECHANICAL TESTS ULTauATE TEST 980.
CAUGE TF.5T TIELD TEN 5tLE ELONO.
REDUCTION l
TEMPERATURE *F STRENGTH KSI f STRENGTH K5J IN 2**%
OF ARE A %
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VALUES
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VALUE1 NDT /
Charpy Pt/Lbs Y. Shear Lat. Exo.
Drom Weichts
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15 0
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-60 1F
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+10 68.0 30 49
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+110 L16.0 90 82
+160
,L30.0 100 86
+160* ;L31.0 100 84 ADDITION AL DATA D.CLLDING MEAT TREATMENT:
(a) 1550 - 1650*r 4 hrs. Water quenched.
(b) 1225*P + 25' 4 hrs.
Air cooled.
(c) 1150*F{25'40 hrs.
Turnace cooled to 600*P.
Tha impacts were taken parallel to the major rolling direction of the plate at the 1/4T level ed notched perpendicular to the plcte surface.
N tonelles were taken in occordance with ASm A-20-68. /
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These tests were witnessed by G.
E. Representative, P. W. Quinlan.
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General Electric Corp.
May 5,1981 175 Curtner Ave.
Sm: Jose, California 95114 ATTENTION: Mr. R. K. Langerbeam
Reference:
General Electric Telecopy dated April 24, 1981, R. K. Longerbeam to J. L. Pyle (attached)
Attached is a copy of Weld Inspection Records and Test Certifications for welding materials used on welds identified in referenced telecopy.
Weld No.
Weld Material 1-3 A
3/16" B4 Wire Heat 20291 7
1092 Flux Lot 3854 1/4" 8018 Electrode HADH 3/16" 8018 Electrode LOEH l
1/4" 8018 Electrode LACH 1-307 B 3/16" B4 Wire Heat 20291 l
1092 Flux Lot 3854 3/16" 8018 Electrode LDEH 1/4" 8018 Electrode LACH 1-307 C 3/16" B4 Wire Heat 20291 1092 Flux Lot 3854 1/4" 8018 Electrode HADH 1/4" 8018 Electrode LACH l
1-308 G 3/16" B4 Wire Heat IP2815 3/16" B4 Wire Heat 21935 3/16" B4 Wire Heat 12008 1092 Flux Lot 3869 1/4" 8018 Electrode LACH 1-308 H 3/16" B4 Wire Heat IP2815 3/16" B4 Wrie Heat 21935 3/16" B4 Wire Heat 12008 1092 Flux Lot 3869 1
1
e..f6 4., f N.'k.K.Longerbeam
' May 5,1981 e
Weld No.
Weld Material
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1-308 J 3/16" B4 Wire Heat IP2815 3/16" B4 Wire Heat 21935 3/16" B4 Wire Heat 12008 1092 Flux Lot 3869 1/4" 8018 Electrode FAGI 1-313 3/16" B4 Wire Heat 90099 3/16" B4 Wire Heat 10120 0091 Flux Lot 3458 1/4" B018 Electrode FAGA 1/4" 801B Electrode CCJA In response to specific questions on referenced telecopy, the following connents apply:
1.
HT No. Lot No., Flux Lot No. by weld.
All of this information is noted above.
2.
Chemistry - Including Cu and P.
Chemistry data is being furnished, but in some cases, Cu and P are not available.
3.
Charpy imoact data and temperature.
Tnis oata is attacned on each test report.
4.
Droo weight test data.
These tests were not performed, therefore, no data is available.
5.
Upper Shelf Energy.
These tests were not performed, therefore, no data is available.
Very truly yours, 1
CONWUSTION ENGINEERING, INC.
l ck L. Pyle J W/ksc cc Mr. R.A. Hillis 9
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+ 91-FASE c.st sec7
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-ro 11y
-/ tow we c.cor
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/-2/3 foor/t.rs:ry szt aser
- sp
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- NR = Ator Byertsel or /aAA. (Me, rima ~ ABLr assaw J p t.t Ky. Guick. l-99 Bes.t)
..m,_
ColmUSTION ENCINEERINC, INC.
[asI[4 L 7 ge,/M4 3
NUCLEAR COMPONENTS DEPARTIE 4/ pyg j
Chattanooga, Tennessee
- A**
N /- ez. G u
/
CONTMCT.NO. :
DETAIL UELDING TROCEDinE DRAUIUC NO.:
No.:
TSAA-2 (A)
Rev.1 WELD NO.:
DATE:
~-
~
REFERENCES:
' M&P 6.1.1.2 (c),
t, (8)
H&P 4.3.8.5(b), SAA-33-27 f
QSAA-11A(3), QMA-llA(1)F4 b
Non-Destrt/etive festinr.:
I P.T.
[
Nf@
1 -
M.T.
N'N\\_&
rn s
-w > s_
_ a WELDING COMDITIONS:
I Elcetrode Type & Size See attached sheet.
FillerMetalType5,3ize y
l
/
Flux Type & Size
[
5 Welding Current & Polarity 7
- Arc Voltage
'Thnvel Speed (in/ min.)
Shield Gas Type & Flow l
Cas Cup Size
[
j' y Gas Cup to Work Distance Other l
- 1 10*/. of Value or Range WELDINC POSITIONS:
Flat t
Prshest:
1s.p
'F.
Hold Dt90t90DutMDOC Until P.W.H.T.
I Interpass:
500
- F.
Post-vcid hest treatment:
1150 *+
25
'F hold one hour / inch thickness of weld.
15 minutes.
I Intermediate P.W.H.T.
1100 *+ __ 50
- F, hold _
r I
j
_T* Sl set-7 t
~
DETAIL WELDING PROCEDURE r
No.:
TSAA-2 (A)
Rev.1 Sheet:
2 of 2 s
r.
i 4
TELDING SEQUENCE TRAVEL AMPS
- VOLTS _*
- j
'Ist Pass - 9.D. 3/16"W Mil.B4 Mod.
(
'st Increment" - 0.D. 3/16"$ Mil.54 Mod.
Single Arc
?
ll) 0.D. to Ib" Level 13 IPM 650 AC 31 Single Arc 3/16"# Mil. B4 Mod.
i k
l2)
I.D. to 1" Level 22 IPM 600/550 AC 31
/
3/16"5 Mil. B4 Mod.
Tandem Arc h
(3)
- Remainder - I.D.
22 IPM 600/550 AC 31 3/16"5 Mil. B4 Mod.
?
Tandem Arc A
(4) 0.D. to 3" Level 2'2 IPM 600/550 AC 31 3/16"5 Mil. B4 Mod.
Tandem Arc g
(5)
' Remainder 0.D.
31 l
3/16"5 Mil. B4 Mod.
Tandem Arc
?
Root or Backweld 1/4"# E-8018 C-3 (Flat only) 325-375 DC-RP 25 6
or 3/16"6 E-8018 C-3 210-260 DC-RP 25 3
- Flux Linde 1092 65 x 200 l
g
,l
=
1 e
l t
F I
9
=
-.m...
_ = = - ~. _ _ _ -
e n
I Ta/lr 8 -
MIL S - Y flN*We L td' /0 92 FIM Su bnrerar ed' Arc Vessef die /d LaAnen-Da 4
\\
j (ia Sr }/e J - G,,,f us hon fneskeeriny P/a >, +
og 5
s.
7),
L + eral.
y or
(*6 Temp Ma E Ft-Expamon Shear af-No.
(*F) cs-r)
(mild 2_tos/sss9
+ so*
97,9o; 63 a4 v4
'acot/Seef t.io
- s 7, 9o, g3 v A-
"n 3ose/sy+7
+ /o
- i2, 4 G, 10 A' A yn 14,9/, 92 2006/3 M7 tio
- 12, 9 t,1 ?
A/A
/
4/A 3054'f.3ltFf
.tfo
- 62,87,92.
NA NA 9'357
)
tto
- 5 yo, St4, y (,
yn yA T
74, 48, G Y tuo '
r 11I,Ito,I09 14,7B,f
'i Y's #Y t
ffGS/9/0/45 -40 t/O*
/04, /of, //G A'A N4 fG5/9f19%-80
+ /O '
//0,77,/2.C WA-80,70, 9 O f'G5(1/o&S3 --(,o O*
88, 9f,9 G, MA 4 o,70, 70 tb D 12 l, l 2.I, I20 NA loo, loo, loo 42i2."
I2.5,l55,133 A/A loo,ioo, tot This wOfiin{fT) 15 /
Mefe / f Skftb w
w l
s w e.- nx pm A' s A/d Aswih bk A
-,-9
---w-
~. _ _
T*./
~
TaS/L B -
MtL8-4 r/eelrod'r, LrA /o9x f/s Sah momed Arc Vess#) &lof " Td u e ls a m w hk J
%, + (u s;.)4 /-GL.4, Lk s
eeHes yeg NDT W
En La.+ ers\\,
- l*
(*F) em W
YIt I ' 4ot 4-(*F)p ECT)
Expaeoien She4-Flux u (mits) le197/5919
+1D
\\
j L M4(,5?)f 'y 54
+1O toI, t08,10.3 5Aorp 5P5s12/ogst
.gn +1b
,,g, gg2, q eq tPSTSSlOtbl ~10
+10 gyy 109,104,184 bbl %3'jb68
+-t O g y, 103, t 5,98 SlF7+/34sy
+,o Si W, 97 1 L-1L j
FAGA l
'Ae foCA TAA-w n
W.
tA
- Ai l
6A 6
'8 RAA'CE CA NA2 KJaic hv (IlM8' Mt-A i
We
---e-r--v-
,w
RELEASED FOR CD&1 CONTRACTS F _ h,'
WELD PROCEDURE SPECIFICATION CONTRACT NO.
DY DATE g
r.y w y, PG sea-he Grooves s Buildup a
General Electric Comeanv P"%f8,v,=,e-WPS 323-2F4FC astwen PnopucT muct r An vrent t fclaats Al PAGE No 1
or 1
ersenerviou Shielded Metal Arc and Submerced oAvg 2-17-69 Arc Welding of ASME P12B Subgrou_o 1 Material nEvisiow wo_ 4 f 9-2 3-70) RAT RE FERENCE SPECIFICATIONS
. PREHEAT REQUIREMENTS:
General WPS 800 Latest Revision General WPS 820 Latest Revision Minimum preheat of 300*F shall be PROCEDURE oUALIFICATION applied uniformly to the full NO-POstTION THICKNESS RANGE adjacent base material for a 963(TW) F(Sub Arc) 4 1/2" to 9.9" minimum distance of "T" or 6",
k F,V, H (SMA) whichever is least, where "T"'
is the material thickness.
1261 (SW) F(Sub Arc) 2 3/4" to 8" F,V (SMA)
Maintain preheat temperature until start of post weld heat POST HEAT TREATMENT -
treatment.
I Procedure qualified with 50 hrs. at 1150*F +25'/-50*F.
INTERPASS TEMPERATURE REQUIREMENTE Post weld heat treatment of the weldment shall be in accordance with The interpass temperature shall
.f not exceed 500*F' maximum.
a CB&I approved procedure.
FILLER METAL:
ASME SA-533 Gr B Class 1 or Submerced Arc SA-508 Class 2 Specification - N.A.
Classification - N.A.
ASME Group
- No. P12B Subgroup 1 Analysis - A3 (execpt Ni 0.50 to 1.25)
Usability - F6
~
FILLER METAL - ASME Trade Name - Adcom 1NMM.(1% Nickel or equal See Adjacent Column Shielded Metal Arc Specification - SA-316 Classification - E8018-G
]
. Analysis - A3 (except Ni 0.50 to 1.25) 1t
-F ELECTRICAL CHARACTERISTICS -
See Adjacent Column
^
SHIELDING GAS-None l
SACKUP GAS -
None SMA - DCRP i
1 T.'.':
- y+5Y B
"~
Linde 124 Submerged Are Tandem Wire Lead Wire - DCRP Tg W YF f 28/.2" N "A Trail Wire - AC Single Wire - DCRP Belt /;n s..n.veyL'c 4.g, Af.s.y,y pp,c.A timme.r 1
sy DATE REGd
'~e 1 Os 08 Os Os Os us ou es REG atc MEG 5
E 5,"c5 5."ta !,T " W,
ac 5
,W. "o "a ""5'cT:
m me i l
~
I
l_ _ _ _ _ ____ _ _
g g --
z
=:.,..-
I 4... -
y Asa m
CBI NT, *12AR CDtPAE.
7 g
- stgLD PmOctDURE
. SPf CIF 8 CATION
- .em>
k
.r*
' t CNIO.ACO SA MlE 4 2npu eg.
38w Ailey SMA g gg I
2 g
Crooves & Bu11dge
(}
,,,,e_Ceneral Eleetrie cesseny see4c ag,,,,,
oneoug.,
.m
> a ve,i e iclass 11 e.se - 2 posgsi.evia, J.' a e J ae d ** t a l Are and 34:serted Are 3 3 ene.a. fine of I i:c P133 Suseroun 1 Pat. ora al m.,e 3-17-69
. 5. *
.. a 3 23. N d
==
A&f ensassa spicericatacana
-Ceneral WP5 500 f.atest Revision pg:rgaypyhn;gwy;,;g; Ceneral W:*5 820 !.atest Fevisien MiML*58' Pro.% eat of 200*F sha* 1 he b.-*
- .i.*
eso:f ruet cuauptcati:ss applied uniforesiv to the rull t.51thness of the weld $cint ar.d neo.
- sesmew twiesentis arai cg adjacent base vaaterial for a Par.1-19 9 0 (5M.A)
V 3/1C" to 3*"
'" " *f" "*"*****'I* * "* [. 4*.t.hien-I' '
It31(SMA)
M 3 /16
- ts 8" A8 **
i F-1 i2 (s.A) ca.r an6
- t. 8-
"" " '
- 2 ' h ' ** " " * -
0 1893tSA-1l F
3/16* to ga Maintain 330*T min. preheat te p.
2220 (SA-2l F
3/16* to 8' until start of postweld heat t: e a t-a ment except for acngitudinal a.c i
ecst a. at tatArwavr.
=ircu=*erential shc11 and head j
Proceduto e:ualified with 30 hrs. at does, prehea :.ay k.e disp;ed to 11SD*F
- 5*/ 50*F.
2EG*F min. 8 h=urs a!ss: c: plet; :
Postweld heat t,reatment cf the weld. cf welding. All rur.cf f taas a I w... M at shal! be is acecria.nce with a flux da.-s rus-he :e =ved prier t C3&I approved procedure.
dropping preheat hel=w 2 v'r.
aAstaarTAL-.
IN Ie) As5 I.';I?..?:: n er T :2:v:::. r :
(
ASME SA-533 Cr 3 Class 1 or he interpass te perature s'all
.s-EA.503 Class 2 not amesed 5:0*T :.a.xt=um.
.'t.
Arwr crece> Ns. 7123 Su.bere se T y;=,, 2 3 gg t:,3 FI'
' A MT"4='
- A.nM. -
5 ;::sereet Are Shieldef Meta! lt e Specificataan
< A-5.5 Specificatzen - :../..
Classificati=a - N.A.
./ *,
claesification - 28318-C
'"*?.**
Analysis - A3 (except Ni 0.50 tal.21 ) nal, sis - A3 (em ept Mi 0.5c to A
Usability - F4 1.25) e Usthility - FE Trade 3*t.=a - Alloy Roda ESC 18Mfi e
Trade Name
.CE
!!**.M (11 Nickel) i Suh erved are
?.,
See Adjacent Colu=A..
er equal 2.12c'"m
- CAL C A AAc' " p* S ** c'5 :
i,;,,
ELscTPisaL tenamactamestscs SNA = DCRP
- f'al.
See Adjacent Celse.
Si,2reereed Arcg eseit.nusc sAs - yo.n.
Tandes Wire nacawe sas.
Nene Imed Wire DCRP Trail Wire AC 1.*
PLuz 3.inde 124 Single Wira = DCAP l
f.. e nd > -g o c. :.
i Ta&+
'"c"-"*'"~"
,1*pg
""'gg..... g. J le.u T.rs banT es- ~wn Tse lW ~Ie=== i C.Tw
- I * *** ", ;
i i
-l
" "* e en r..
aa e pt a. t.-t.w,y a
<.an. n. e,as g er..en. " stalmatcaer.*rt=est'
'f.rs 3 /r.er.S -
av r
- ,*, t s
"e -
L El' mo red i
i D
L t* 2' 8 i
. ' * 'l 'I #i "/,', ' -
/_
e n
~me-*-m-o-
-w-weewee--.----etwwg.-,ww.w.
mew.,,a.--v-
e-g w++-e--ww-.
w-.-ew,-wewww
Yo u e O tt
_1MMM E le e 4 rod e_ (%de Ga.m e - %e)
Emde 12 4 Flu x bl,me ed Art
_, Pa s + urcJ d ItSO*F Av 50 %
+v oica I i
_Plam+ C.
( L a c, a ma-Ue.vd e 2 - CST M)
J upr Cha Ene Lakeral
- /o
(*r)
Temp km En-S Espacion She c FI u n. A>o.
(*F)
(s.,T1
( m'ils)
P7397
-50
-70
\\
as,21 i 8, is-c, s-
/ O15 6
~5 42,21,19 33,25,40 to, s 5, so 4-10 G 4, M,55 55,55,54 00, 35,40
\\
t-10 44,70 63, 59 4o 45 t
+AO j
- > I, B 9 103, 9, 85 J S, 6 6, 79 65,9o,97
+ 212
% 94 59,66,59 loo,too,tOO
-80
-80 s I, ai,9 A C, as, I z 5, S, 5 054z
-20 11, s 6, S4 57,57,4s-3o,.25,,20 4-10 65, 8%,1) 6 8, 72, 41 Jo Po, 65
+ lo
% 3, 7(a 42, 64
& b, 3s"
\\
f 4D B 7, 9 I 71, 6 0 2S,80 t 7o loo, /ol, 97 92, 89,7/
9o,9F,90 1 2.12 IOS, lll,168 66,64,86 luo, too,100 f7N
~Go
-80 27, p 4 2 I, I 1.
S-o 0351
~ 70 49, 45,2(,
41,36,22 IS',IS",5-O G3,57 56, 5(66 54, 45, 63 3o, 2 5, 35~
+10 90
& 2, G 2, S6 3 o, R S, 45" tio 67, SS' S5,42-40,30
+ 40 47, 97
~11, 9 o 4 S, SD t 2 f 2.
11 B,102,1I1.
6 6, 71, 11 100,100, loc
)Pb
-20
-80 s, g
(,, 11 5,.5~
ot5G,
~GO 27, IG, I2 2 3, 13, 10 25, 20, 25 i
lo, 10, ID O
r 7, 56, 30 zo, 2.7, 27
+- IO So,56,(T,M,27 25,3B, a2,28,30 15,15,15, Is, ZD t SD 34,46,42.
73, 37,45-2.r,SD, 35
+4D 7%, GO,72.
54,47,49 GD,46 50
)
+ 2.t 7.
4 3, 6t, R3 GI, 66, 61 108, l@/ 'D0 W
l
Lt/e_ e v 1nM M Elec+rodc. (Treulo Ala mu - Ra ch
_Lih d e 12 4 Plu v
% L,n,em ed A ec Pod %)e Iel
!IEb
- P -Co r crh r 4 u pr 'e. I Js u c- (L.- r &z-cozu) v
.a u, upr Chag D'S N
u+,,at
- /,
~
~
M Ted gp Enky g,p..',,
sg
,.c
(*F)y Fluk No.
(s.,T)
(m;13) g/ 093/(>f /-
-60
- e> o 39, 37 a7,37 r, 5-
-Go 19, 2 0, 3 2.
18, 2.2, 27
/o, / o, / o O
SI, 55, 58 SD, 50, 6 3 3 o, 3 0, S[
\\
4-1 0 69,(9,6C G I, 6 s,51 m, s o, 40
+/0 42, S'7 40,63 6 0, 4-0 y90 77,GC 73, 72.
> o, Fo
+ 2 l ?-
6S,9/,85 66, 75, 23
/DC, /00,/M
~20
-Bo S, 8 G, II C, 5~
015 6
- &,0 2.2, I 6, I 2 2 3, l ~6,8 O 10, I o, I O o*
1~7 36, 30 20,2.?j 2P 2F,20,af j
+ 10 30,58, IT 2 S, 3 P, l*L l5',I5,15 t-10 34, a.8 2 8, so 15, 2 o 4 30 34,46,42.
29, 37, 43-2 r, so,35-F40 72, 60, 72.
54, 47, 41 Sb 4 5~ 5D y
t 2 1 2.
9 3, F /, 23 65,66, 67
/00, loo;ICO 1
G
- ~~
~
'~
rat /e ae(%dc klame -
a ERolR oH Elec+rcde' Sh sided h e k t A, c.
_?os.t We Id li f 0*F 4 S O L re h en da f l
I l'
Plan + c (tanum WNe 2 RFV-CSI AO cham-uor se L.+emI
- /.
.[L+Ab.
Temp Yn15 s sp..i.n shw-(v)
)
(ep)
(m;ts) o
-60
-108 10, il y
4 4
4 4216 27Ar
-90 2.5, 3 o,3 2.
5 cg g i o, j o
-3o 19, 2.5, 31 19,23,25 20,25,25 i
-2o 22, 26,5o 2 3, 2./, 27 2 r, 2 s,3 o
-Io 36, 41,43 22,32,30 30,So,30 0
50, TI,57 3 G,3 B,40 30,40,45 1-40 135,137 B4, B 0 90,Bo T-l30 151,16o,16]
60, 6 2,8 j joo,too,IOO 4%
~BO
-S O I A,17 I 5', i G 5, 5
-60 3 3f grff,3 14,16, Z o i s' / G, 2 o i o, i o, i o
~ 40 26,26,40
- 2. G, 2 4, 33 3 o, 3 o,3 c
-20 65, 74,127 44,TP,76 4 o s o, Go 107,108 99, 80 8 d, 70 F25 12 7, g2g j40 24,29,P2.
/ots,/00,9D 1-40 95 lu,PO VO t'50 153,/Y3 /T6 F.f,p/',9 /
s NP
/73, /?],4r fs 7/
/oo,ico,too
-40
-los 19, a 6 3, 3 3,3
~2o 85,20,27 6, 9, I 5 5',1 0,10 g,g 38,42,45 26,31,31 30, S o,3 0
-to 0
SC, s 2,6 2 SS, A 4, 9 s 35, 4o, 40 l
+ 40 5 6, '15 42,SS 5'o 6 0 s
+ \\50 118, 122,I30 8 1, 6 9, 8 2.
100, t oo,lec i
e
-~*
.. *, * - ~.
.-=.:=--~'
strom um F17e E "(TruleAJo-e-Shierded Mokt Are Po -I kidd Itsn*f L so k r.
4 v o,en t_
ia PIant c (I.anun Wedo 2 RPV-CBIN)
J er cg.
e~
t.+.e.i
/.
tn-hy.
ery..l.,
ssw-m Teny 1
1-+ E-(*F)
(psly 9pff0
-60
-Go 42., 45, 23 3r,3G,20 F
Y, r j
-2 eso4827AE G I, F4, 77 4 P, & c. 6 2.
30, 25, 2 f
- 20 6F 67 r/; 5 2.
z r, 25' B o,,89, 8 2.
6 3,6 2, Go 3C,S'o,35 O
t-4 0 9F, 9 7 7t 76 40,75 j
t70
)ll LOT,109 6',92SS, 7 7 60 9o,60 s
s A 212.
122, 114, 13o 92,
, 69 loo,ioc,too 40
~7o 10 II, 7 9, G 5, g 642b 6 27AE
~40 3 3, 5 2,5 2.
2.1,42,22.
10, I S, lo
- 20 6 9, 62.,37 52, 4 P, 3 0 20,so 2b j
- 2.o s2, rr 36,,5g f r, i r
-Io Go, S4,6 8 44,5,,53 40,30,30
~
f40 96,99 s l, 6g 60,60 v 2l2 I19, l2.2, I2+
95,90,G8 100, I 00, e>o slotP W -so
-90 1,so yy 3, 3
%os2ME
- 10 iS,1G,lk I A, IS,16 6,8,IO
- 10 27 39,54 2S 35 35, 55,35 0
27,50,54 25, 4 $,46 j
j 46 A o,4 5, A 5"
+- 10 15,76, 107 40,62,14 4o,50, 60 t-40 9 0,I00 7i,76 10, 60
~
T 130 ISO, I40,142.
91,94,93 100,too, LOO
~70 9,9
'l, 1 S, 5
~b ID E )IE
> S> II E# IO,1O 5403521 %
s 3
O
\\
59, I,70 5' t, 53, S8 SD, SOj GO
+ 40 9 9, IOI 77, 78 90, 75 t 72.
106,110 SS,87 80, BO 1850 I 2 9, 13 I 1 5 2.
St.s 76, SI IOC, loo,ICO L..
L :-
Is h /L 6 t/
EForn um Fle c.tvecic - (T,4 Ua w.
c.
We Ids d Me4d A rc u
Pes + we t &,
Il 50' p L,r s o u,. 4 0 r>; w r i6
._P_I_n a + c. (L a c, o., o tirvd. ? RPv. car;A J
Nor th*Ti E"
l*+ 'r^ l Enk)'
Esp.uion Shur i
(w)
Temp i
aof 2 -
(*F)
Mis) 03
-so
-sos s,9 z,3 3,3 ss2562 747
~ SO lo > lb, I9 3
-20 y I o, I I 80, t03'O 31, F 0, 65 22,31,FD 3 o, 5 o 3 0
~ 10 j
36, 53,58 34,45,W 4o,40,4D O
& I, 25, 7s 44,58,59 SD, 6 0, 6 0 t40
/04, los 75, 77 Po, 60 t-83 0 122, 125,126 69,63,9I 100, loc,ICO OSS-
-Jo
-100 11,I3 3, T 3,5 4
- 90 16,11,19 6, F 7 B, B, /0
~30 j
l'1,50,3) 1S,44, A3 15_ 'o, 20
~ 20 48, 92,44 B3, 34, 35 3o, 30,30
-10 S2, 69, G,6 39,4S,46 40,4-o, 40
+-4 0 84, 87 G 3, G8 60,60 4-13 0 12I, 124, 12.9 91, 9 b, 95~
l00,100, I 00
~80
-loo I o, t 2.
G,7 A, 5
,y ryyp
-So 14, I S, a +
1 o, I ~2 > I &
I *> I O I2 s
- 20 St, 52, 61 31, 40,43 55, 50,40
-20 44,L5,49 51, 4 2, SS~
l5,I5,17
-20 67 SG, 55,45 i S,10 44o 12.0,123 72,73 00)60 t 7 2.
12 P, ifo 7E II 10>
s FIS O 14 7, 154,168
- 90) VI,97 100.,100,900 I
i M
~
n, / /t.
h ~"
//
JuMM Dechode_
( Trod'e t/ame - Techo//w]
L'ha'e /2 4 Flux Su bm e <a ed Arc Jes t We M /t.r6 *f 4 a-
'so s,-r. Aspdo/
/'
Planf A ( 2im er RPV, t*8ZA0 s+ a.
NDr Chag i
le Em La+eral y
'r% Er+ki
(%
Ted ein s h e,,-
En7i,3 en m
c.
KdN
-80
-130 S
7 6
7, 7 f ; S-y 017 I
- 60 34, 18, 22, 32, /6,2 /
4c,55,4c
~ 20 GB,yo sz c s, 51, g6
- n,10,77 i-l O 7S, 72 64, 69-SC 96 j
t40 4 4, B 2.
Bl, 7/
/Do,9r
+ 2 s 2.
94,92,86 74, 80,60 IO C,s o c,s o o
~ 130 T
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W MATERIAL LOCATION SHOREHAM NUCLEAR POWER ST=?'ON-UNIT I FINAL SAFETT ANALYSIS REPORT fle u r't _l - CSM> r&m O
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.SNPS-1 FSAR ATTACHMENT A i
ductility transition temperature (NDTT) is at least 60 F below the lowest servt::e temperature.
When the CVN test was applied, a 30 f t-lb energy level was used in defining the HDTT.
There was no upper shelf CVN energy requirement on the Shoreham beltline material.
The bolting material was qualified to a 30 ft-lb CVN energy requirement at 60 F below.the minimum preload temperature.
Frcan the previous comparison it can be seen that the fracture toughness testing performed on the Shoreham reactor vessel material cannot be shown to comply with 10CFR50 Appendix G.
i Bowever, to determine operating limits in accordance with 10CFR50 l
Appendix G,
estimates of the beltline material RTNDT and the highest RTNDT of all other material were made, as discussed in Section 5.2.4.3.
The method for developing those operating limits is also described therein.
On the basis of the last paragraph on page 19013 of the July 17, 1973 Federal Register, the following is considered an appropriate method of compliance.
t 5.2.4.L2 Intent of Proposed Aporoach The intent of the proposed special method of compliance with Appendix G for this vessel is to provide operating limitations on pressure and temperature based on fracture toughness.
These operating limits assure that a margin of safety against a l
nonductile f ailure of this vessel will be very nearly the same as a vessel built to the Summer 1972 Addenda.
The specific temperature limits for operation when the core is critical are based on a
proposed modification to 10CFR50 Appendix G, Paragraph IV.A.2.C.
The proposed modification and the fastification for it is given in GL Licensing Topical Report NEDO-21778-A.
5.2.4.2 Acceptable Fracture Eneray Levels Not applicable to this backfit approach.
The SNPS-1 approach is l
given in the previous sections (Section 5.2.4.1).
5.2.4.3 Operatino Limits Based on Fracture Touanness Operating limits which define minimum reactor vessel metal temperatures vs reactor pressure during normal heatup,
- cooldown, and inservice hydrostatic
- testing, were established using the
" methods of Appendix G of Section III of the ASME Boller and Pressure Vessel Code, 1971 Edition (Appendix G first appeared in the Summer.197.2 Addenda)...The results are shown, on Fig. 5.2.4-1.
l
+
~ the~ vessel-shell 'and head areas remote trom diiscontinuities 161 1 l
plus the fendwater nozzles were evaluated, and the operating llait curves based on the limiting location. The bolt-up limits a minimum for the flange and adjacent shell region are based on s.
metal temperature. of RT
+
60 F.
The maximum through-wall NDT section for N values and temperature limits are given in this
$9 Estimated RT limiting locations in the reactor vessel.
5.2-25 Revision 18 - June 1980
.v--
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ATTACHMENT A SNPS-1 FSAR temperature gr dient from continmus heating and cooling at 100 F
')
per hour was considered.
The safety factors applied were as
/
specified in ASP 2 Code Appendix G and GE Licensing Topical Report NEco-21778-A.
For the purpose of setting these operating limits the reference l
temperature, RTNDP is determined from the toughness test data taken in accordance with requirements of the code to which this vessel is designed and manufactured. This toughness test
- data, Charpy V-notch (CVN) and/or dropweight nil-ductility transition temperature (NDT) is analyzed to permit compliance with the intent of 10CFR50 Appendix G.
Because all toughness testing needed for strict compliance with Appendix G was not required at the time of vessel procurement, some toughness results are not available.
For
- example, longitudinal CVN's instead of transverse, were tested, sometimes at only a single test temperature of +10 F for absorbed energy.
Also, at the time either CVN or NDT testing was permitted; therefore, in many cases both tests were not performed as is currently required.
To substitute for this absence of certain data, toughness property correlations were derived for the vessel materials in order to operate upon the available data to give a conservative estimate of RT compliant with the ND"'
intent of Appendix G criteria.
l material analyzed, and were derived from the results of NRC
~)
Tnese toughness correlations vary, depending upon the specific Bulletin 217,
" Properties of Heavy Section Nuclear Reactor Steels", and from toughness data from the Shoreham vessel and other. reactors.
In the case of vessel plate material (SA-533 Grade B, Class 1), the predicted limiting toughness property is either NDT or transverse CVN 50 ft-lb temperature minus 60 F.
Longitudinal CVN transition curve results and NDT values are available for all Shoreham vessel plates.
The transverse CVN 50 ft-lb transition temperature is estimated fran longitudinal CVN data in the following manner.
The lowest longitudinal CVN ft-lb value is adjusted to derive a longitudinal CVN 50 ft-lb transition te; rature by adding 2
F per it-lb to the test temperature.? If the actual data equals or exceeds 50 ft-lb, the test temperature is used.
If sufficient data are available as in the case of Shoreham, the 50 ft-lb temperature is derived by interpolation.
Once the longitudinal 50 ft-lb te=perature is derived, an additional 30 F is added to account for orientation affects and to estimate the transverse CVN 50 ft-lb temperature minus 60 F, estimated in the preceding manner.
For forgings (SA-508 Class 2) the predicted limiting property is the same as for the vessel plates.
Both NDT and CVN values are available for the vessel flange, closure head flange, and These forging RT feedwater nozzle materials for Shoreham.in the sanne way as for tN.
}
_ values are derived from the data vessel plates.
5.2-26 Revision 18 - June 1980 w,
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SNPS-1 FSAR ATTACHMENT A For the vessel weld metal the predicted limiting property is the
(
CVN 50 ft-lb transition temperature minus 60 F, as the NDT values are
-50 F or lower for these materials. This temperature is derived in the same way as for the vessel plate material, except the 30 F addition for orientation effects is omitted since there is 'no principal working direction.
When NDT values are available, they are also considered and the RTNDT is taken as the higher of NDT or the 50 ft-lb temperature minus 60 F.
When NDT is not available, the RTNDT shall not be less than -50 F, since lower values are not supported by the correlation data.
For vessel weld heat affected zone (EAZ) material the RT 18 NDT assumed the samr as for the base material as ASME Code weld procedure qualification test requirements and post weld heat treatment indicates this assumption is valid.
l Closure bolting material toughness tert requirements for Shoreham l
were for 30 ft-lb at 60 F below the bolt-up temperature.
Current Code requirements are for 45 tt-lb and 25 mils lateral expansion (MLE) at the preload or lowest service temperature, including bolt-up.
Therefore, since CVN values as low as 40 ft-lb exist for Shoreham closure bolts, 60 F is added to the test temperature in order to derive the bolt-up temperature.
Using the above general approach, an initial RTNDT of 19 F was established for the core beltline ' region for Shoreham Unit 1.
Tne effect of the main closure flange discontinuity was to establish the minimum considered by adding 60 F to the RTNDT temperature for bolt-up and pressurization.
The minimum bolt-up temperature of 70 F for Shoreham Unit 1,
which is shown on Fig. 5.2.4-1, is based on an initial RTNDT of +10 F for the closure flange forgings.
The effect of the feedwater nozzle discontinuities were considered by adjusting the results of a
BWR/6 reactor discontinuity analysis to the Shoreham reactor.
Tne adjustment was made by increasing the minimum temperatures required by the difference between the Shoreham and BWR/6 feedwater nozzle forgang ATNDT's.
The feedwater nozzle adjustment was based on an l
RT of ~20 F.
NDT The reactor vessel closure studs have a minimum Charpy impact energy of 40 ft-lbs and a 23 mil lateral expanslan at 10 F zor snoreham.
The lowest service temperature f or bolt-up is taken to
.be 60 F above the 10 F values or 70 E.
I 5.2.4.3.1 Temperature Limits for Bolt-up A
=4n4=um temperature of 70 F is required f or the closure studs.
A sufficient number of studs may be tensioned at 70 F to seal the closure flange O-rings for the purpose of raising reactor water
(
level above the closure flanges in order to assist in warming l
them.
The flanges and adjacent shell are required to be warmed 5.2-27 Revision 18 - June 1980 l
.c.-
l ATTACHMENT A SNPS-1 FSAR i,
I to minimu:n temperatures of 70 F before they are stressed by the T
full intended bolt preload. The fully preloaded bolt-up limits
)
are shown on Fig. 5.2.k-1.
5.2.4.3.2 Temperature Limits for Preoperational system Evdrostatic Tests and ISI Bydrostetic or Laek Pressure Tests Based on 10CFR50 Appendix G IV.A.2.d, which allows t reduced safety factor for tests prior to fuel loading, the preoperational system hydrostatic test at 1250 paig may be performed at a minimum te=perature of 91 F which is established by the RTNDT Of the botto= head plate plus 60 F.
The fracture tougnass analysis for system pressure tests resulted in the curves labeled A shown
. on Fig. 5.2.4-1.
The curves labeled
' I*
The beltline weld material is expected to be more limiting at end-of-service Iluence levels, and this weld material has an initial RT r;;T of
-50 F.
The predicted shif t in the RTNDT fN Fig. 5.2.4-2 (based on the neutron fluence at 1/4 of the vessel wall thickness) must be added to the beltline curve to account for the effect of fast neutrons.
5.2.4.3.3 operatino Limits Durine Beatup, cooldown, and core Operation The fracture toughness analysis was done for the normal heatup or j
cooldown rate of 100 F/ hour.
The t @ ature gradients and thermal stress eff ecte corresponding to this rate was included.
The results of the analyses are a set of operating limits for nennuclear heatup or cooldown shwn as curves labeled b on Fig. 5.2.4-1.
Curses labeled C co these figures apply whenever the core is critical.
The basis fear Curves C is described in GE BWR Licensing Topical Report NEDO-21778-A.
t 5.2.4.4 Ccrac11ance with
- Reactor vessel Material surveillance Procram Recuiremente=
Charpy in: pact specimens for the reactor vessel surveillance programs are of the longitudinal orientation consistent with the ASME requirernents prior to the issue of the Sunener 1972 Addenda and ASTM E 185-73. Based on GE experience the amount of shift measured by these irradiated longitudinal test specimens is essentially the same as shift in an equivalent transverse specamen.
The program includes three sets of specimens in the reactor. The specimens are manufactured from a plate actually used in the beltline region and a weld typical of those in the beltline region and thun represent base metal, weld metal, and the transition zone between base metal and weld. Sufficient tensile and Charpy-V notch specimens are provided in each of the three in-reacitor sets and in the out-of-reactor set to measure strength, ductility, and toughness of each of the three materials (base,
- weld, EAZ),
both in the unirradiated and irradiated 5.2-28 Revision *18 - June 1980 M --hi*.* "'- ]**
- l 4
SHOREHAM MAIN STEAM PIPING AND FERRITIC VALVES 1
The Shoreham Main Steam Piping was procured to the 1967 ANSI B31.1 Code, which did not require toughness testing.
- However, data are supplied in Tables 12 through 15 to show that the Shoreham NSSS supply steam pipe materials.would possess adequate toughness.
This is concluded from the fact that similar materials for Shoreham Balance-of-Plant, Zimmer Main Steam, and Susquehanna 1 Main Steam have data showing adequate toughness per more current 10CFR50, Appendix G requirements.
These Tables show that the chemical compositions are similar (especially C and Mn:C ratio), tensile properties are similar (indicating similar microstructures), and in some cases the material supplier is the same.
The materials for Zimmer (Table 15) andgusquehanna (Table 14) are generally normalized (about i
1700 F and air cool); whereas, there is no information on the Shoreham material heat treatment.
However, discussions with U.S.
Steel have indicated that the material was hot-finished in a manner which would be similar to a normalized heat treatment.
Also, there is no information on deoxidation practice for any of the materials, although they are all at least Si-killed.
However, as mentioned, it is concluded that their microstructures must be very similar.
Thus, it is concluded that the materials toughness data in these Tables may be considered representative i
of Shoreham.
This is further supported by the fact that the Zimmer and Susquehanna materials were received in 1973, only 3 years later than Shoreham.
Thus, it is likely that these materials, especially from the same supplier (U.S. Steel) are similar, and are adequate.
2.
Shareham Safety Relief Valves (SRV) are in compliance with 10CFR50, Appendix G, since they are exempted by the ASME Code from toughness testing because of their 6-inah size.
3.
Shoreham MSIV (Main Steam Isolation Valves) wert exempt from toughness testing at the time of purchase.
They do not see significant pressures at temperaturas below that of steam.
I Typical information is given in Table 16 for Shoreham MSIV's.
Toughness data on similar materials for MSIV's on other projects, where toughness testing was done, is attached in Tables 17 and 18.
In some cases, the materials and valves vendor are the same as for Shoreham (LILCO).. Especially pertinent are the data in Table 17, which were produced on the same order as Shoreham (LILCO).
These data demonstrate the cap bility of the Shoreham MSIV materials to meet current toughness require-ments.
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Further evidence of toughness for SA-105 forgings (MSIV bonnet or cover material) can be found in the July 1978 issue of Metal Progress, Pages 35-39.
This argicle shows charpy V-notch toughness in excgss of 25 mils at +40 F and NDT values ng greater than -10 F for SA-105 material normalized at 1565 F for 4 hr. and air cooled after forging.
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SHOEEHAM REACTOR VESSEL CLOSURE STUDS Shoreham reactor vessel closure studs were procured to,the specification A540 Grade B24, and were ordered to meet the toughness o
requirement of 30 ft-lb.
Charpy V-notch energy at +10 F.
The cur 5*"t r*Suir***"t i' 45 ft-ld and 25 mils lateral expansion at
+70 F.
Actual charpy tesg values for the Shoreham studs ranged from 40 to 50 ft-lb at 10 F.
Thus, it is predicted'that these studs wculd have get the current requirements at the higher test temperature of 70 F, and bolt-up limits are based on this gigher temperature.
Actual data values for heat no. 66593 at +10 F.
45, 46, 46 44, 44, 46 45, 50, 44 46, 47, 44 46, 49, 49 40, 42, 41 ft-lb i
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18tyMTO: JOHN COPELAND GE kn Jose,
ATTENTION: A.W. ZEUTEN
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gg,/7 May 20. 1981 n.:.
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sumpue. Providias Igeet e-n=+=.. %.
- 1. w a n.gtian ca,any Bill.
Beloware the impact test resnits froa sight (S) valves provided to General Slectric for the Fukushima II Project. On the some order we provided eight (3) each of the mama valves for the Carolina 1. Carolina II. and LILCO. The Pukushima valvas were the only ones that required Charpy-V impact tested material.
Mr. Sanchen of LIIA0 has requested the results of the Fukushima tests for tFPicals of what the LII4D valves would be on the body. bonnet, and disk.
I Flesse review tha ruults and convey to Mr. Zeuthen that these are typical
values. All test results are at +40'F.
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Charpy-V,ft-lb Iateral Expansion. mils 2690942-116 23, 21.5. 21 25, 24, 25 2690919-112
- 39. 42, 45
- 38. 37. 37 2690916-111 26, 27, 31 27, 27. 31 2700070-117 23, 21. 22
- 25. 25, 25 2690913-110
- 26. 28.5. 36 27, 31, 37 2690902-109
- 31. 38, 29 34, 37, 30 1700096-124 41, 38, 36.5 42, 40. 36 2690940-115 36, 38, 29
- 34. 37. 30 Donnet st #
CW-V, ft-Ib tateral tapansion, mils 218006 50.8. 42.1. 40.4
- 46. 41. 37 218006 50.8. 42.1. 40.4
- 46. 41. 37 218006 50.8. 42.1. 40.4
- 46. 41, 37 i
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218006 50.8. 42.1. 40.4
- 46. 41. 37 218006 50.8. 42.1. 40.4 46, 41, 37 21s006 50.8. 42.1. 40.4 46, 41, 37 21s006 50.4. 42.1. 40.4
- 46. 41, 37 M
M h V, h -Ib interal Ixcension, mils 217964 28.8. 41.5. 61.6
- 25. 34, 48 217964 28.8. 41.5. 51.6
- 25. 34. 48 217964 28.8. 41.f. 61.6
- 25. 34, 48 217964 28.8. 41.5. 61.6 25, 34. 48 217964 28.8. 41.5. 61.6
- 25. 34. 48 217964 28.8. 41.5. 61.6
- 25. 34. 48 X.
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[nbis 17]
W. C. Beaks May 30. 1 M1 fL tuos 2 Rockwell Intemational Disk Et #
Cherry-V,ft-lb tatere1 anseesion,mtle 217M4 28.8, 41.5, 61.6 25, 34, 48 I
Flease note that all of the results for the bonnets and disks are identical l
b-=== they are from the same heat. Present Code rules would not, allow one l
test to stand for this many forgings.
If this information meets your satisfaction, plassa trasamit to Mr. Zauthen at the following address:
Mr. A. Zeuthen LILCD 175 East old Country Reed Ricksville, NY 11801 Jh p
B. E. Ca thers, Supervisor ikII r5 cal Process Control 1
Me
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. oo, SHOREEAM RPV SURVEILLANCE SPECIMENS Shoreham Surveillance Specimen Plates are identified in Table 1 and in the response to Q121.23 on SNPS-1.
Mechanical property and compositian data are also included in Tables 1, 2, 3 and 6.
The surveillance weld materials are also identified in the Q121.23 SNPS-1 response.
The Heat'lP3571/ Lot 3958 material is the same as in the LaSalle 1 Surveillance Program.
The weld procedure ic given in Table 7, to represent the Shoreham longitudinal weld seams.
The mechanical properties are given in Table 8, and the composition is 0.37% Cu (unverified).
The charpy v-notch toughness of Type 8018 electrodes EODJ and KACI are 168, 148,153 and 88,76,81 ft.-lb. at +10 F.
These materials are used 0
in the root passes of the weld, and would probably not affect the surveillance program.
Cu contents are 0.02 and 0.04%.
Further information en the surveillance capsules is also given in the SNPS-1 Q121.23 response.
Each capsule will contain 12 specimens (charpy) of each base plate, weld, and weld HAZ.
Also, included are 3 base, 4 weld and 4 HAZ tensiles in capgule 1 (30 ); 3 base, (120 ); and 2 base, 2 weld and 2 HAZ tensiles in capsule 33 weld, and 2 HAZ tensiles in capsulg)2 (300 A response to SNPS-1 Q121.37 is shown as Attachment B, and indicates 1
that plans have been made to add a fourth surveillance capsule.
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e SHOREHAM ATTACHMENT B The Shoreham Surveallance Program was designed prior to the require-ment of ICCFR50, AIpendix H, and three surveillante capsules were provided.
Under 10CFR50, January 1, 1980, Part 50, Appendix H,Section II, Surveillance Program Criteria, Par. C3, Revised With-drawal Schedule, four capsules are required with the fourth capsule indicated as standby.
For Shoreham, three capsule supports are available cn the reactor vessel and it is no longer advisable to perform additional welding on the reactor vessel.
Test coupons are available and it is proposed that a fourth capsule will be installed when the first capsule is removed.
This will serve to continue to provide a standby capsule should one be needed.
The withdrawal schedule will be in compliance with 10CFR50, Appendix H, Section II.
Additional justification for this Surveillance Program in Shoreham-is based on the fact that the weld material, which is limiting, is also used in the LaSalle 1 Surveillance Program.
(See the response to Shoreham Question 121.36).
Thus, between Shoreham and LaSalle 1, a total of seven capsules will be irradiated to study the effects on the properties of the limiting material.
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l SER OPEN ITEM NO, 35 LILCO Response:
1.
Remote Manual Valves Remote manual valves are used in the Shoreham containment isolation design to isolate certain ESF related or potentially beneficial systems.
The system function requires that these isolation valves remain open, or be opened for accident mitigation purposes and/or for safe shutdown.
Subsequent isolation is achieved by operator decision and action, or by process signal.
As discussed in TMI Response III.D.l.1, a preoperational and surveillance leakage reduction program is being implemented at the Shoreham plant.
This program is intended to reduce and maintain leakage to as low as practical for systems outside primary contain-ment that could contain radioactive fluids.
The method available to the operator to close these penetrations is included in the attached Table.
2.
Nitrogen T.I.P. Purge Isolation Provisions As discussed in the TMI Response II.E.4.2 Paragraph 2, a solenoid _ isola-tion valve will be added to penetration I-37A Nitrogen Purge for the T.I.P.
system.
3.
X-44 and I-45 Remote Manual Valves - Locked Closed The remote manual valves for the Drywell Floor Seal Pressurization portion of the X-44 and X-45 penetrations should not be locked closed during operating modes 1, 2, 3,
and 4.
Closure of the 1T23*MOV031A or B isolation valve would isolate the drywell floor seal from its nitrogen accumulator.
Each accumulator has been designed to provide proper seal internal pressure during a-LOCA.
Two seals are provided for redundancy.
Seal accumulator failure is indicated in the control room by a low nitrogen pressure alarm (
55 psig).
The isolation valve would then be closed by the operator to effect containment integrity.
?
The remote manual valves for the Primary Containment Atmospheric Control System (PCACS) Portion of the X-44 and X-45 penetrations are normally closed valves that are only opened for combustible gas control as dictated by emergency operating procedures.
Indi-vidual keylock, momentary contact control switches with CLOSE, OPEN, and CENTER POSITIONS and RED Lopen) GREEN (closedl, and BLUE (out for loss of control power or motor overload) indicating lights are provided in the control room for the X-44, and X-45 PCACS isolation valves.
Each switch spring returns to the center from the CLOSE and OPEN positions.
The key is removable in the center position only.
.c
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. 4.
Instrument Lines Sensing Primary CQntainment Atmosphere Our position regarding these lines is included in FSAR Se'ction 6.2.4.3.5.
All instrument lines will be exposed to the primary containment peak pressure during the Type A test, 5.
Instrument Air to Drywell These penetrations (B-7 and D-5) have been identified as essential as per II.E.4.2. Table 1.
These lines supply nitrogen to air operated isolation and safety relief valves located inside the primary con-tainment.
The system provides an adequate supply of nitrogen from the inerting system, accumulators, and from backup yard connections.
Sufficient nitrogen is available for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following an accident.
Check valves inside containment are provided to prevent back flow out of containment and control room annuciation is provided on low header pressure.
I l
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TABLE 1 CONTAINMENT ISOLATION QUESTIONS 1.
Penetrations X-22A,B RBCLCW to/from Recirc Pump S-23A,B and Motor Coolers These penetrations will be closed by the operator within 30 minutes after automatic containment isolation occurs in accordance with station emergency procedures.
The RBCLCW piping from the Primary Containment penetration to the recirculation pump seal cooler and motor cooler is ASME III Class 3, Seismic Category I.
The pump seal cooler is also Seismic Category I.
The motor cooler, though not designated Seismic Category I, has been analyzed to demon-strate its capability to maintain its integrity under SSE loading.
Therefore, the system forms a closed boundary inside containment satisfying the requirements of GDC 57.
2.
Penetrations X-44, 45, 46, 47 Primary Containment XS-7, 8, 20, 21 Atmospheric Contrc' aystem The isolation valves for the primary containment atmosphere control (PCAC) system are maintained closed during normal operation.
Redundant isolation valves are provided for each penetration.
These valves are Type C tested in accordance with Appendix J to 10CFR50.
Following an accident, at least one train of this system may be in operation to control hydrogen buildup in the primary containment.
In this case, i
the system boundary becomes an extension of the containment.
The PCAC system is a closed system outside containment.
It is designed and constructed to ASME III - Code Class 2 and meets Seismic Category I requirements.
The system is included in the leakage surveillance programs as discussed in TMI Response III.D.l..l.
_.~
1-i Open Item 37 An evaluation of fluid systems was conducted to determine the existence of any potential pathways by which primary coolant or containment atmospheric leakage could bypass the leakage collec-tion and filtration systems of the secondary containment.
An analysis of these pathways determined that their-potential for l
significant leakage to the environment was low.
A.
Feedwater System Primary containment isolation is provided by a check valve inside and a testable check valve outside of primary con-i tainment.
In addition to the containment isolation valves, i
there is a motor operated stop check valve located outside i
the primary containment.
The two check valves located outside of primary containment are equipped with positive closure measures which ensure valve closure and seating l
following an accident.
After isolation of the HPCI and RCIC systems, there are three valves in series between i
either the primary coolant or containment atmosphere and l
the environment.
I' B.
HPCI and RCIC Suctions from the Condensate Storage Tank The condensate storage tank is isolated from suppression pool water by me2ns of a check valve and a motor operated gate valve in the auction lines of the HPCI and RCIC systems.
After system inolation, a second motor operated gate valve provides additional isolation.
Within one day following the accident, the head of liquid from the water in the condensate storage tank would be sufficient to cause condensate to flow into the HPCI and RCIC systems if there is leakage across an isolation valve.
C.
Core Spray Suctions to the Condensate Storage Tank A locked-closed globe valve in each suction line of the core spray system isolates suppression pool water from the i
condensate storage tank.
During initial phases of the t'
accident, containment pressure acts to seat these valves to minimize leakage, while in later phases any leakage is l
into the core spray system from the condensata storage tank.
l
l t
p
' D.
RCIC or HPCI Test Return Lines to the Condensate Storage Tank Two motor operated, normally closed gate valves on the tect lines return isolate the condensate storage tank from either the HPCI or RCIC systems.
Further isolation is provided when HPCI and RCIC are isolated from the RPV.
After one day, elevation differences would normally tend to force liquid into the HPCI or RCIC systems rather than into.the conden-sate systems.
E.
Condensate Fill Connections to the HPCI, RCIC Core Spray and RHR Systems The condensate transfer system is used as the alternate fill source to tho RHR, HPCI, Core Spray and RCIC systems.
The condensate connections are isolated by means of normally closed globe valves as well as check valves in the lines.
Due to the small size of the valves used, bypass leakage is negligible.
F.
RWCU Connection to the Condenser l
During an accident RWCU System is isolate; from the RPV and the RWCU is not expected to contain highly radioactive i
water.
The condenser is further isolated from the RPV by two closed gate valves in the blowdown line.
G.
Suction and Recirculation Lines for the RCIC Loop Level Pump from the Condensate Storage Tank l
Three (3) 1" spring loaded check valves are used to isolata
(
the RCIC loop leve' fill system from the RCIC system.
Leakage past all three vaives would be negligible.
Motor operated valves are not used for the inolation function in order that the reliability of the fill system is not compromised.
H.
CRD System l
The design configuration of this system is discussed in FSAR Section 6.2.4.3.2 (Containment Isolation System).
Bypass leakage concerns are not directly applicar'e to this system due to its unique containrant isolation aspects.
l I
- t
. I.
Service Water System & Ultimate Cooling Connection to the RHR System The ultimate cooling water connection to the RER system is protected against leakage in either direction by dual isola-tion valves with a drain-off connection between the two valves.
J.
Reactor Building Closed Loop Cooling Water System This system including the provisions for leakage detection is duscussed in FSAR Section 9.2.2.
Bypass leakage concerns are negligible due to the easily detected system leakage during normal opcration and the lack of direct leakage paths.
K.
Service Air System This system is discussed in FSAR Section 9.3.1.
Bypass leakage is also negligible for this system due to the minimal containment to environment leakage paths.
L.
Main Steam System Leakage past the Main Steam 13olation Valves is processed via the Main Steam ;eakage Control System as described in FSAR Section 6.5.
l l
M.
Containment Floor and Equipment Drains
(
Two motor operated gate valves provide containment isolation for drywell floor and equipment drains.
Leakage past these valves would be negligible and would be collected in drain tanks vented to secondary containment.
N.
Fuel Pool Cooling and Cleanup System Two normally closed motor operated gate valves provide con-tainment isolation for the suppression pool cleanup line.
Furthermore, a narmally closed diaphragm valve is located further downstream.
Bypass leakage past all three valves would be negligible.
1
SNPS-1 FSAR SER 0.h k Joy Iss*< SZ, M m e><ad & h J A~*,R A.
17.2 QUALITY ASSURANCE DURING THE OPERATIONS PHASE I
17.2.1 Organization The Long Island Lighting Company (LILCO) is responsible for the establishment and execution of the Quality Assurance (QA) Program during the operational phase as required by 10CFRSO, Appendix B.
LILCO will establish the organizational structure shown on l
Fig. 17.2.1-1, to fulfill this respcnsibility.
The LILCO QA Program during design and construction, including the transition from the construction phase to the operational
- phase, is described in.Section 17.1.
.[The overall res m sibility for the
- policies, goals, and objectives of the q tity assurance program during operation has l
been delegated to the Vice President-Cperations.
He is responsible for the safe and reliable operation of all LILCO generating plants including all associated quality, schedule, and l
budget requirements.
The LILCO corporate quality assurance l
policy statement, signed by the two corporate executives b-responsible for all LILCO personnel and organizations performing l
M quality-affecting activities, identifies the Vice President of i
t Operations as the individual responsible for establishment, h
(
implementation, and effectiveness of the QA Program policies, k
goals, and objectives during the operational phase of nuclear power plants.
In addition, this policy statement requires that W
titrict adherence to quality assuring requirements shall prevail j '. (
h over other considerations when dealing with safety-related D
matters.
The Vice President-Operations has delegated the i
l responsibility of the QA Program during operations to the l
This Manager is (organizationally free fron. matters involving schedule and budget./
I
'The responsibilities of the Vice Prcsident-Operations are carried' out through the Director-Production, QA Manarer, Manager-Electric m
Production, Nuclear, and Plant Manager.
l y
The Manager-Electric Production Nuclear, who reports to the Vice y
President - Operations, has responsibility for LILCO Nuclear q
electric generating plants.
He is the immediate supervisor of g
the Plant Manager and has overall responsibility for plant I
g operations including the assurance that the operating organization complies with the QA Program and its implementing l
procedures and instructions.
He is assisted in this effort by i
W the Plant Manager and the
- Manager, Nuclear Operations Support
- Division, whose staffs are comprised of engineers and technical D
and nontechnical personnel.
In addition, supplemental support is available as required from other areas within LILCO such as Engineering, Project Management, and Purchasing organizations, or)
M ernal to LILCO such as consultants and contractors.
f
% V QA Manager is located in headquarters. and he reportsT
{'.*
\\ (The administratively to the Senior Vice President, Engineering and D 5:2 Project Management.
For operating phase activities, he reportsj Ww k
~
17.2-1 Revision 17 - September 1979
9 INSERT A The Corporate Statement of Quality Assurance Policy commits LILCO to the policy of strict adherence to quality requirements in all safety-related matters concerning the Shoreham Nuclear Power Station.
The corporate statement assigns overall respon-sibility for the Quality Assurance Program to the Vice President, Engineering.
The Vice President, Engineering, reports to the Senior Vice President, Engineering and Purchasing, who is directly responsi-ble to the President of the company.
The Vice President, Engineer;rg, is responsible for corporate engineering services such as environmental, power, electrical, design, planning, and systems, and provides engineering support as necessary to the Office of the Vice President, Nuclear.
He is also respon-sible for the establishment of QA policies, goals and objectives during operation of the nuclear power plants, and for assuring the implementation and effectiveness of the QA Program.
He has delegated specific QA Program responsibilities to the Manager, Quality Assurance Department.
i
-._:~._.
i i
INSERT B The Vice President, Nuclear, reports to the Senior Vice President, T&D and Operations, who is directly responsible to the President of the Company.
The Vice President, Nuclear, has overall responsibility for the engineering, construction, licensing, start-up testing, and safe and reliable operation of the nuclear power plant.
He has delegated the responsibilities for the above duties to the Managers of the Shoreham Project, Shoreham Station, Nuclear Engineering, and Nuclear Operations Support.
The Vice President, Nuclear, is also responsible for assuring the implementation of the QA Program requirements by the organizations under his jurisdiction.
He has delegated to the Plant Manager the responsibility to assure implementation of the LILCO QA requirements.
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INSERT C The Manager, Quality Assurance Department (QA Manager), is located offsite and reports directly to the Vice President, Engineering.
3 i
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l l
i
SNPS-1 FSAR Co *U* Zhsa0 0-
{ Operations. functionally to and receives direction from the Vice President-J m,
L f'This reIationship assures that LILCO quality assurance oersonnel who are responsible for GheckimQ auditing, otherwise verliying @ quality related activities gi_ ave oeem or
/Derrormem are independent of personnel directly responsible for performing the activities or any other undue influence associated with schedules or costs.
He is responsible for establishing and assuring implementation of the LILCO QA Program as described in the LILCO OA Manual.
He is responsible for maintaining a working interface and communication within LILCO, regulatory agencies, consultants, contractors, inspection
- firms, and others as l
required to effectively execute the policies stipulated in the QA l
Program.
He is responsible for assuring the establishment and l
continuous implementation of the quality assurance indoctrination and training program for LILCO quality assurance and other l
concerned personnel.
The indoctrination and training will cover the quality related policies, procedures, and requirements applicable to the oersonnel involved.
He is responsible for review and approval of applicable documents to assure the inclusion of appropriate quality requirements as indicated in Section 17.2.6.
He is responsible for the performance of audits as described in Section 17.2.18.
The QA Manager is responsible for defining the content and changes to the LILCO Quality Assurance Manual subject to review and approval as indicated in Table 17.2.6-1.
The QA Manager is authorized to evaluate the manner in which all activities both at the station and offsite are conducted, with respect to
- quality, by means of
- checks, reviews,
- audits, surveillance, and/or inspections.
He will perform this l
evaluation on a planned and periodic basis to verify that the QA Program is being effectively implemented.
He is responsible for periodically evaluating and reporting on the status and adequacy l
of the QA Program to the appropriate LILCO management.
He has the authority and organizational freedczn to identify quality l
problems, to initiate, recommend or provide solutions through designated
- channels, and to verify implementation of solutions.
He has the authority to initiate stop work
- action, or control further
- rocessing,
- delivery, or installation of nonconforming l
material through appropriate channels as described in the l
l applicable QA Procedure.
The minimum qualifications for the position of QA Manager are defined in Section 17.1.1A.
The QA Manager is assisted in carrying out his responsibilities by the QA Department staff consisting of Quality Systems and Field Quality Assurance Divisions.
These Divisions consist of engineers, and technical and nontechnical personnel as required.
In addition, this staff will be supplemented as required from other areas within LILCO, censultants or contractors.
Line l
recpensibility, ccordination, and communication during this time will ne througa the QA Mnager.
17.2-2 Revision 17 - September 1979
I SNPS-1 FSAR TmeA b rThe Plant Manager reports directly to the Manager-Electric
, p Production-Nuclear, and he is directly re.sponsible for the _saf_e l
- j" reliable operation of the plant. f He is responsible for 4
x coordination or activities between regulatory
- agencies, other departments withi, LILCO, consultants, contractors, the public, and others whose activities are applicable to the operational phase.
He is, therefore, responsible for assuring the i
implementation of the QA Program at the station.
He is the authority to approve all station OQA procedures and instructions.
He will continually analyze and evaluate the station QA Program for status and adequacy and report his findings to corporate management..He has the authority to stop work on any activity at the
- station, including removal of the unit from service.
He is l
assisted in carrying out these responsibilities by-the plant operating staff Shich i.%ucescribed in Section 13.1. #
The Operating Quality Assurance Engineer (OQAE) has direct g- [L responsibility for assuring implementation of the LILCO QA y
He *
- Program, and additions and changes thereto at the stati h orking i g
reports directly to the Plant Manager and maintains a
interface and communication with the QA Manager.
He is responsible for establishing and implementing all station QA/QC bnh procedures and instructions.
He is responsible for implementing >S F
(
all LILCO QA procedures and instructions as they apply to the I
station.
He is responsible for review of applicable documents as ) M l
indicated in Section 17.2.6.
He is responsible for the ex. g performance of station audits and inspections as described in )p Sections 17.2.18 and 17.2.10, respectively.
He has the authority
?
and organizational freedom to identify and report quality S-problems;
- initiate, recommend, or provide solutions through b designated channels; and verify implementation of solutions.
He yp 4
has the authority to initiate stop work action through channels h or control further processing,
- delivery, or installation of nonconforming material as described in the applicable station OQA m
procedure.
In the event of a difference of opinion between the e
Plant Manager and the OQAE regarding a
significant quality e
- matter, the OQAE shall refer the problem to the QA Manager for
,k resolution.
d The minimum qualifications for the position of OQAE are as l'
follows:
Bachelor's degree in Engineering or Science plus a
minimum of 4 years experience in a responsible area of operations or quality assurance requiring both technical and administrative ability.
Four years of acceptable erperience will include at least 2 years of quality assurance experience or 2
years of nuclear operations experience under the auspices of an established quality assurance program plus formal quality assurance training.
A high school diploma plus a minimum of 9 l
years experience in a responsible area of operations or quality aesurance activities, at least 5 years of which must be in the area of quality assurance, will be considered equivalent p
qualification tor the position.
These education and experience l (
requirements may be modified by other factors such as previous performance, satisfactory completion of proficiency
- testing,
~
17.2-3 Revision 17 - September 1979
- 9"'
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INSERT D The Manager, Shoreham Nuclear Power Station (Plant Manager),
reports to the Vice President, Nuclear, and has been delegated direct responsibilty for the safe and reliable operation of the station.
I l
- +
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l f
L
SNPS-1 FSAR formal QA education, etc.,
when these other factors provide
- q reasonable assurance that a
person can competently perform required tasks.
i The OQAL is assisted in carrying out his responsibilities by the l
station Operating Quality Assurance (OQA) staff.
This staff consists of engineers and technical or nontechnical personnel as required.
In addition, supplemental support is available, when
- required, from other areas within
- LILCO, consultants, or contractors.
Line responsibility, coordination and communication during this time will be through the OQAE.
The OQAE is responsible for evaluating and reporting the status I
and adequacy of the QA Program at the station to the Plant Manager and the QA Manager on a periodic basis.
l The QA Manager is responsible for the development and implementation of the overall QA Program during
- design, construction, preoperational
- testing, operation, and modifications of nuclear power plants.
A complete description of the program through preoperational
- testing, including organization and responsibility, is presented in Section 17.1.1A.
F Engineering has responsibility fer the l
The Vice President l
design of safety-related modifications for the nuclear power stations in accordance with applicable design bases, regulatory
\\ requirements, codes, and standards.
He shall receive technical id support for these modifications from LILCO Engineering and
's
,% external organizations.
He shall also provide technical support in areas such as licensing, radiation protection, quality of the
- {' environment, fuel management, and nuclear instrumentation.
He is responsible for assuring that their activities are accomplished in accordance with the appropriate operating, engineering, project, and quality assurance policies and procedures.
,/
The Nuclear Review Board and the Review of Operations Committee are responsible for safety
- reviews, approval of
- programs, l
procedures,
- tests, repairs, and modifications.
A complete l
description, including structure and lesponsibility, is presented i
in Section 13.4..
gng;n,,ry o o g pa y go,jn g PresidenthurchasingandStores,hasresp sibility for
- c x l
The Vice the commercial aspects associated with LILCO procurement activities.
He is responsible for assuring that the e activities are accomplished in accordance with the appropriate Purchasing, i
Operating, and QA policies, procedures, and ins ctio'ns.
reports directly to the Senior Vice PresidentpTnn and oppra c i nn_Hef
=T p(
Specia I Suvica.s %parm we, The Manager,e_uas Production and voeratl _
as responsibility for XX 1
providing trained and qualified personnel, as necessary, for the support of nuclear power station maintenance,
- repair, and modification activities.
He is responsible for assuring that these activities are accomplished in accordance with the l
appropriate maintenance services and operating and quality l
17.2-4 Revision 17 - September 1979 l
t
SNPS-1 FSAR
(
assurance policies and procedures.
He reports directly to the Director Production.
4 g"
y f
. n e nort ment The Manaaery G Meter and Test)has responsibility for providing XxX L
carten calibration services to the nuclear power station for calibration L
of measuring and test equipment and shop standards.
He is responsible for assuring that these services are accomplished in e
l accordance with the appropriate meter. test and operating and quality assurance policpas and procedures.
He reports directly to the,Vice President e Operations.
O y
j 17.2.2 Quality Assurance Program l
Responsibility for assuring that SNPS-1 will perform safely and reliably over the life of the station rests with LILCO.
I The LILCO Corporate QA Policy Statementyi endorsed by the two) l g
4 (corporate executives responsible for all personnel and]
X J (organizations performing quality-affecting activities,/ imposes a
% QA Program which is designed to meet the requirements of Title 10 Q of the Code of Federal Regulations, Part 50, Appendix B and t
identifies the QA Manual as the document which establishes the I
t requirements for quality-affecting activities during the operational phase of nuclear power plants.
The QA Manual l
contains this c>rporate policy statement and is distributed on a controlled basis to responsible managers and key supervisory and QA personnel.
The QA Program is designed to assure that activities such as l
operation, maintenance, modification,
- repair, refueling, inspection, and testing, which affect safety-related structures, systems, and components, are accomplished in accordance with the criteria of
The QA Program is applied to l
the safety-related structures, systems and components listed in Table 3.2.1-1.
The QA Program, described in the LILCO QA Manual, is supplemented by QA Procedures and Instructions which provide the detailed instructions and checklists necessary to implement, or verify implementation of QA Program requirements.
These procedures are delineated in Section 17.2.5.
Quality Assurance Procedures and Instructions are issued, reviewed, and approved as shown in Table 17.2.6-1.
The QA Manual, Procedures and Instructions shall be controlled in accordance with the requirements of Section 17.2.6.
The transfer of LILCO QA responsibility from the design and construction phase to the operations phase is described in Section 17.1D.
The QA Program requires that activities affecting quality shall l
be accomplished in accordance with documented
- policies, procedures, and instructions throughout the life of the station.
These activities shall be accomplished under sult;bly controlled conditions.
Controlled conditions
- include, as applicable, l
appropriate equipment, suitable environmental. conditions, and
~
assurance that required prerequisites have been satisfied.
Also l
considered shall be the need for special
- controls, processes, l
17.2-5 Revision 17 - September 1979
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SNPS-1 FSAR test equipment,
- tools, qualification
'of personnel, and e
requirements for verification of quality by inspections,
{
examinations, or tests.
The QA Procedures for operations are derived from the program requirements established in the QA Manual.
Organizations performing quality-affecting activities shall prepare their procedures incorporating requirements of the QA Manual and referenced
- codes, standards, and guides.
These procedures will also receive a quality assurance review to assure that all program requirements have been addressed. I l
.E n o e n aer t w a..
l l
The Cor ate QA Policy Statement contained in the LILCO QA Manual im es the mandatory QA Program requirements on al.1 personnel d
organizations performing activities affeceing the quality of s fety-related structures,
- systems, and components during the operational phase of station life.
The Vice Presidentegg.em;.un$ is responsible for periodically engaging an X
organization, independent of the organization being reviewed, to assess LILCO quality-related activities and evaluate the
- scope, implementation and effectiveness of the QA Program as applied to operations to assure that the program is adequate and complies with corporate QA policies,
- goals, objectives, and 10 CFR Part 50, Appendix B criteria.
The requirement for independent QA Program evaluation is further imposed, as appropriate, on other organizations participating in the LILCO QA Program.
The LILCO l
QA auditing system is described in Section 17.2.18.
E ma i n ee r r'n a. ar& Voce Pasuded >Jxlw laho Direct responslbility
'for establishing and implementing the QA' Program has been deleg ted to the QA Manager and the OQAE.
Provisions have been established for the referral, by these personnel, of quality elated problems to the highest level of l
management necessar) for resolution.
The QA Manager is responsible for regulnrly assessing the status and adequacy of l
the QA
- Program, bo internal and external to LILCO, and for reporting the results of this evaluation to p e Vice Presidentf woerauons.
Tne vice President-coeratuonm will advise Corporate Officers on the status and adequacy of the Program as required.
This regular assessment shall be conducted in accordance with the requirements outlined in Section 17.2.18 and detailed in l
Section 18 of the QA Manual.
The requirement for regular QA Program evaluation shall be extended to other participating organizations for the portions of the program they are executing.
The QA Program requires that procedures be established for the indoctrination and training of-station and offsite personnel performing quality affecting activities.
These procedures will document the scope, objective, and method of implementing the indoctrination and training program and contain provisions for documenting training sessions including
- content, date, attendance, and results.
ene QA indoctrination and training will include instruction as to the purpose, scope, and, implementation m
of quality assurance
- manuals, procedures, and instructions.
Training and qualification in the principles and techniques of 17.2-6 Revision 17 - September 1979
SNPS-1 FSAR.
immediate supervisor, but who may be from the same organization.
Design changes will be subject to design control measures commensurate with those applied to the original design.
Design control measures will provide for the suitable review and selection of standard "off the shelfa commercial or previously approved material,
- parts, equipment and processes that are essential to safety related structures, systems, and components.
Design documents and revisions thereto will be distributed to the responsible individuals in a
timely and controlled manner to prevent inadvertent use of superseded documents.
Control of design documents is further described in Section 17.2.6.
Design documents and reviews, records and changes thereto are collected,
- stored, and maintained in accordance with Section 17.2.17.
Errors or deficiencies which may arise during the design process
~
will be addressed in accordance with Sections 17.2.15 and
'17.2.16.
Organizations supplying equipment and/or services are responsible for imposing the applicable requirements of this section on their internal operations and on those vendors and contractors performing work within the scope of their activity as required by the procurement documents.
They are responsible for assuring by means of audit or surveillance that design control as defined in their respective programs is being effectively implemented.
LILCO is responsible for assuring program adequacy and implementation for external sucoli'ers through planned and C
periodic audits'.
%ce s,w ACt.
t4atlear En4ouersne, ne bluelece Opmts'ons Su pport, In general, internal LILCO organizations will acccmplish design i
activities such as preparat
, review, approval, and issuance of appropriate design docume ts, including changes
- thereto, as described in Table 17.2.6 Design
- changes, including those made by plant operating personnel, shall be governed by design control measures commens ate with those applied to the original design.
Corrective actio, as described in Section 17.2.16, will be applied to design pro ess deficiencies that adversely affect safety-related structur s,
- systems, and components.
The Plant Manager is responsible or determining, initially, wnether or not proposed modification or repairs involve an unreviewed safety l
question or a change technical specifications as described in 10CFR50.59.
Procedur will provide documentation and control of such determinations.
Technical evaluation, including design verification, will e the responsibility off the anoroori gef such as(Electric Production,G,ncineerin l@ W X r a qualified independent organization. g (he internal organizatio N#
T
.g
-p(Pro,ect Managemen_
LILCO Quality Assu ance Department is responsible for assuring l
overall program establishment and implementation through planned and periodic audits.
17.2.4 Procurement Document Control The LILCO QA Program provides for the control of procurement l
l documents for safety-related material, equipment, and services whether purchased by LILCO or their designated agents during the 17.2-9 Revision 17 - September 1979
=
=
= -
SNPS-1 FSAR I
operational phase.
Section 4 of the LILCO QA Manual describes the QA Program requirements established to assure procurement
)
document control.
The program requires that procedures establish measures to assure control of the preparation, review, approval, and concurrence for procurement documents.
Document control as described in Section 17.2.6 and delineated in Table 17.2.6-1 will be applied to procurement documents including changes and revisions thereto.
The procurement documents will be reviewed by qualified personnel, as defined wiS in this section, assuring the adequacy of the quality requirements.
The review will be utilized to assure that the quality requirements, including preparation, review, and
- approval, have been properly
- defined, that the procured items are inspectable and controllable, and that the acceptance criteria are adequately specified.
The program requires that procurement documents such as purchase specifications contain or reference the design bases technical requirements which include
- codes, industry standards, and regulatory requirements; material and component identification requirements; drawings and/or specifications, test and inspection requirements; and special process instructions.
In
- addition, they are required to identify the requirements of 10CFR50 - Appendix B with which the supplier QA Program must comply; the document requirements for drawings, specifications, procedures, personnel and procedure qualifications,
- material, A
- chemical, and physical test results, and inspection and. test (j
records which must be prepared, maintained, submitted, or made available for review and/or approval; the requirements for the retention, control, maintenance, and/or delivery of records; and the procuring agency's right of access to supplier's f acilities and records for source inspection and audits.
Procurement documents for spare or replacement parts will be subject to j
program requirements which are equivalent to those used for the 1
original equipment or those specified by a properly reviewed and approved revision. Nuclear Gna;neermch Nucleew Opore& ions SaDWh D.,cekaa &a44, The LILCO Purchasing organization {is responsible for the commercial aspects associated with proquring items or services l
which includes the processing of purchase orders.
The internalj LILCO organizations such as Glectric Pre r:-1c ; Gngineering),
x j
Skoda Pro 3ect wanagemegn and Quality Assurance are responsible for assuring that the. procurement documents contain technical and*
quality requirements as indicated above.
Authorized release, assuring acceptability of both technical and quality content, is required prior to releasing a
purchase order.
The plant operating staff is responsible for preparing and issuing procurement documents to Purchasing for processing.
The station OQA organization is responsible for reviewing these procurement documents for quality requirements.
The LILCO QA Department is responsible for review of quality requirements in procurement documents prepared by LILCO headquarters organizations.
u 17.2 *0 Revision 17 - September 1979
l h
l-SNPS-1 FSAR, Qualified QA personnel will review and concur with the suppliers' l
[
QA Programs for safety-related items.
Agents such as consultants, architect-engineers, testing companies, etc.,
designated responsibility by LILCO for procurement activities associated with safety-related material, equipment, or services shall impose the control requirements indicated above.
They will establish the requirements in procedures, instructions,
- drawings, etc.
These requirements shall be imposed on their internal operations and on any vendors or contractors performing work within the scope of their activities as required by the procurement documents.
They shall assure the adequacy of program implementation through audit or surveillance.
LILCO will assure program adequacy and implementation for external suppliers through planned and periodic audits consistent with the complexity, importance, and quality of the item or service.
17.2.5 Instructions, Procedures, and Drawings The LILCO QA Program establishes provisions for activities l
affecting the quality of safety-related structures, systems, and i
components during the operational phase to be accomplished and controlled in accordance with instructions, procedures, and drawings.
Section 5 of the LILCO QA Manual describes the QA l
l Program requirements for the control of instructions, procedures, and drawings.
Organizational procedures delineate the sequence of actions to be accomplished in the preparation,
- review, approval, and control of instructions, procedures, and drawings.
Suppliers,
- vendors, and contractors have the responsibility for establishing required ir.structions,
procedures,
- drawings, and other documents to control the quality related activities of their own operations and those of their subsuppliers as required 1
by the procurement documents.
A description of the associated j
procugement document control requirements is in Section 17.2.4.
t e o p(arSmu,
7 LILCO organi_za_tions fsucn as Engineering, Pro 3ect) y Internal (Managementi, Purchasing, and Electric Productionf are responsible for estab ishing (the required) instructions, procedures, and yl
- drawings, or utilizing established procedures instructions, etc.,
and other documents to control t
quality-related gx activities The required station procedures are described in Section 13.5.
All responsible organizations establish provisions such that the development and implementation of instructions, procedures, and drawings, including changes thereto, are clearly identified and controlled.
The LILCO QA Department and Station OQA organization are l
responsible for surveillance and audit to assure that the i
instructions, procedures, drawings, and other documents used for safety-related structures, systems, and components are controlled s
to meet the requirements of 10CFR50 Appendix B.
1 17.2-11 Revision 17 - September 1979 1
\\
.s.A /v-m /* bed SNPS-1 FSAR k GondmStaH. Naclese Gaineerh5Naeleev Oseo+ ions Saport m-stipulated in
'the procurement' documents.
Measures re
,.g established to nrovide for both a. technical and qu'_ity /X
.i i
evaluation of those suppliers providing % iticaB co nents
- The, internal organizations suc13/as (E.ecci
' Preducti=,)
X gnaineerin mr ere'e=
== : = x.; will perform the technical
'g evaluation and QA will perform the quality evaluation.
These X
f functions may also be accomplished through the utilization of qualified independent organizations.
Personnel performing the evaluations, such as
- auditors, will be qualified.
Source evaluation and selection information will be docuse nted and filed.
The program provides for source inspection, surveillance, and audit of suppliers to assure confor: nance to procurement document requirements.
They shall be conducted in accordance with documented procedures.
Source inspection procedures provide for l
instructions to be established for specifying the characterictics to be witnessed, inspected or verified, and accepted; indicating responsibility; and determining documentation requirements.
Source audits or surveillance will be conducted, as necessary, to assure compliance with quality requirements.
Source inspection or audit may not be necessary when the quality of the item can be verified by review of test reports, inspection upon receipt, or other means.
The program requires that receiving inspection be accomplished in accordance with documented procedures and instructions.
The receiving inspection procedures and instructicas establish measures to assure that the item is properly identified and corresponds to the receiving documentation, that the inspection of the item and acceptance records are determined to be acceptable in accordance with the inspection instructions prior to use, that the receiving documentation is available at the plant prior to use, and that the inspection status is identified l
as indicated in Section 17.2.14.
The QA Program specifies that procurement documents require suppliers to furnish documentation identifying any procurement requirements which have not been met together w'.th a
description of these nonconformances dispositioned " accept as is" or " repair" and that responsible QA and technical personnel shall perform a review and approval of the supplier's recommended disposition. ~ Nonconforming items will be identified and controlled as indicated in Section 17.2.15.
Inspections will be conducted based upon the nature of 'the item beug procured.
1,... re required by
- code, regulation, or other contract requirements, documentary evidence that items conform to procurement requirements will be available at the plant and filed.
This documentary evidence will be retrievable and it will specifically identify the item and codes and/or specifications mat by the item.
Where not precluded by other requirements, such documentation may take the form of written certification of conformance identifying the requirements met by the items.
LILCO
%2 QA Procedures require that suppliers' certificates of conformance 17.2-14 Revision 17 - September 1979
SNPS-1 FSAR.
be periodically evaluated by audits or tests to assure that they s
are valid.
u/ h. req u i s et,*o n de m s W/or se rvice s. a d O A o r o o As Suppliers 'of safety-relate material, eg'uipment, and service are responsible for imposing th control requirements indicated above
.on their internal operat ons and on any vendors or contractors performing work within the scope of their acH.vities as required by the procurement documen s.
They shall assure through audit or surveillance the adequacy f program implementation.
The LILCO Purchasing organization is resoonsible for commercial aspects associated with procuring items or services. The internal LILCO [
organizations / such as Electric Production, Engineering, Project) y-CManagement, and QAf are responsinle for assuring tnat the procurement documents contain the information as required above.
Procedures have been established to control the spare and replacement part procurement documents, through technical and QA review, to ensure that the controls for safety-related items are equal to or better than the original equipment.
'Ihe QA Prograa l
requires that a technical evaluation and QA review be performed to determine tne requirements to be applied to the procurement of spare and replacement parts when the original equipment requirements are not known.
Procurement document control is described in Section 17.2.4.
LILCO will assure program adequacy and implementation for external suppliers through planned and periodic audits consistent with the complexity, importance, and quality of the item or service.
The LILCO QA Department will be l
responsible for evaluating suppliers.
This will inelude the l
utilization of qualified independent organization surveys.
j Source inspection, as necessary, will be conducted by LILCO or a
qualified independent organization.
The plant operating staff is responsible for receipt of items at the station.
The station OQA l
organization is responsible for conducting receiving inspection of items, with respect to quality requirements.
The LILCO QA l
l organizations will assure overall program establishment and l
implementation through planned and periodic audit.=
and I
surveillances.
17.2.8 Identification and Control of Materials, Parts, and Components l
The LILCO
-QA Program requires the establishment of an identification and control system to prevent the use of defective, unapproved or incorrect safety-related material, parts and cx>mponents during the operational phase.
Section 8 of the LILCO QA Manual describes the QA Program requirements established l
for this purpose.
The program requires that the identification system, including unique part or mark numbers, developed during the design and construction
- phases, be maintained and expanded as necessary during the operational phase.
A system for identification and control of materials, parts, and components, including partially C*
fabricated subassemblies will be based on documented procedures and/or instre.ations.
Identification is referenced in l
17.2-15 Revision 17 - September 1979
I SNPS-1 FSAR S weLa S L CS Nuc leaf OperatYo ns Su erY CM.+c,,
\\
g/
.,t., m m such as c__ _ __ __________-_. c;nomeerincMXand (Proiect Manacemeng Safety-related nonconformance reports will be analyzed periodically to determine the existence of quality trends.
Trends, if
- any, will be reported to the appropriate LILCO management.
When nonconformances are discovered by LILCO for internally associated activities, it will be the responsibility of the organization issuing the nonconformance to do so in accordance with the requirements indicated above.
The QA Department and the i
station OQA Organization will review their respective I
nonconformr ce reports to determine quality trends and will report sues. trends to appropriace LILCO management.
In general, station nonconformances will be the responsibility of the operating plant staff with a review by the station OQA organization.
If a nonconformance is reported by the station OQA organization, concurrence of the disposition by station OQA is required.
Similarly, if the nonconformance is reported by. a LILCO offsite organization, QA Department concurrence of the disposition is required.
The station OQA organization will review nonconformances associated with safety-related structures, systems, and components.
17.2.16 Corrective Action l
The LILCO QA Program provides measures to assure that ccaditions adverse to qulity are promptly identified, reported and O
l corrected.
Section 16 of the LILCO QA Manual describes the QA V
Program requirements for corrective action and control thereof.
The program provides for a corrective action system implemented through the use of approved written procedures.
The procedures provide for an evaluation of deficiencies including i
c nonconformance
- reports, and determination of the need for corrective action.
They provide for the reporting, to LILCO station and offsite management, the cause of the conditions significant to quality and the corrective action taken.
The program requires that upon determination or significant conditions adverse to quality prompt corrective action be initiated to preclude repetition.
In addition, verification is required to assure that these actions have been implemented.
Follow-up action is conducted to verify that specified corrective action has been properly implemented and corrective action documentation has been closed out.
Supp12ers are re'sponsible for establisaing and implementing a corrective action system.
The supplier systems provide measures t
l which comply with the requirements outlined above and are imposed on internal operations as well as on vendors and contractors performing work within the scope of their activities as required by the procurement documents.
They also assure through audit or surveillance the adequacy of implementation.
LILCO shall assure overall program adequacy and implementation through planned and periodic audits.
v 17.2-24 Revision "J7 - September 1979 l
m e
O SNPS-1 PSAR TABLE 17.2.6-_1 i
OUALITY RELATED DOCUMEttr CONTROL RESPONSIRTI.ITY Doewments a 3
.freoe g (a3 peviewedu3 Aporo edE83 IssuedO3 USD and _ nn.e H a= =}
{ 5I. Vice President QA Manager QA Manual QA Department Appropriate LILCO Managers l
vico Pecelhne -
l 4D Vice President -
Al u c it a +
QA Manager Pngineerin Manager-Nuclear Eroject Management P.Ner Plant 94 (Dept.)
QA Department Responsible QA QA Manager QA Manager Procedures S Division Manager Instructions Manager-Nuclear Power Plar.t j
- l-Proceiurea Staff Engineer Power Plant Engineer Station QA/QC Station OQA Operating QA Manager-Nuclear Operating QA -
l QA Musac-er Station QA/QC Station CQA Operating QA Manager-Nuclear Operating QA i
Dstructions Statf Engineer Powen: Plant Engineer 1
i
,l QA Manager
'I
\\_
Procurement Responsible Marsager-Re-P.esponsible Appropriate j
Documents Offsite sponsible Manager Manager "l
Staf f Offsite 1 l Organization I
QA Department 4
Operating Plant Responsible Plant Manager-Nuclear i
Staff Chief Engineer Power Pluit Station OQA Organization 4
Special Test Operating Plant Review of Oper-Manager-Nuclear Plant Administrative Procedures Staff ations Corznittee Power Plant Coordinatx>r l
Pesponsible Engineer f
1 I
1, P
f i
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