ML20073S775
| ML20073S775 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/30/1994 |
| From: | Shelton D CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2233, GL-92-01, GL-92-1, NUDOCS 9407060180 | |
| Download: ML20073S775 (6) | |
Text
_m.__.______
!(*
e' CENTE g i
f200 Och he Bouievo'o V34 800'est Donau C,Shelion l
indetenceveOH PO 8cs 94001 semo, y,ce President
{
716 447 3'0 0 CweW N 44'01400 Num m i
i t
Docket Number 50-346 l
License Number NPF-3 L
Serial Number 2233 l
June 30, 1994 i
i United States Nuclear Regulatory Commission Document Control Desk.
Vashington, D. C.
20555 i
Subject:
Response to NRC Generic Letter 92-01, Revision 1,. Reactor l
Vessel Structural Integrity, for the Davis-Besse Nut Icar j
Power Station (TAC No. MH3732)
Gentlement i
i This letter provides Toledo Edison's response for the Davis-Besse
% clear Power Sta'. ion, Unit 1 (DBNPS), to the Nuclear Regulatory-l Commission (NRC) Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity, dated May 13, 1994 (Toledo Edison Log Number 4216). This letter requested confirmation on the applicability of the Topical Reports BAV-2178P, Lov Upper-Shelf Toughness Fracturt Mechanics Analysis of Reactor Vessels of B&'J Ovners Reactor Vessel Vorking Group for Level C & D Service Loads, and BAV-2192P, Lov Upper-Shelf Toughness Fracture Analysis _of Reactor Vessels of B&V Ovners Gtoup Reactor Vessel Vorking Group for Level A & B Conditions, to the DBNPS and verification of the information entered in the summary data files for the DBNPS.
Topical Reports BAV-2178PA and'BAV-2192PA vere prepared by the B&V Nuclear Technologies Company for the B&W Owners Group Reactor Vassel l
Vorking Group.
The B&V Ovners Group submitted BAV-2178P and BAV-2192P to the NRC by letters dated March 9, 1993 (OG-1158) and August 26, 1993 i
(0G-1272), respectively.
By letters dated March-29, 1994 to the B&V-Ovners Group Reactor Vessel Vorking Group, the NRC provided the acceptance of the Topical Reports.
The detailed informathn contained in the B&V Ovners Group Reports BAV-2178PA and BAV-2192PA is applicable i'
to the DBNPS.
?
j 9407060180 940630
)
,PDR ADOCK 05000346 t
P PDR
)
Operotmg compomes Cleveiond Dectoc Nummotmg loiedofd' Son f
Doth t Number 50-346 v
License Number NPF '
serial Number 2233 rage 2 In addition, Toledo Edison has accepted the B&V Owners Group Topical Repott BAV-2222, Reactor Vessel Votking Group xesponse to Closurt Letters to NRC Generic 92-01, Revision 1, which was submitted to the MRC by letter dated June 23, 1994 (CG-94-1393). This report was also prepared by the B&V Nuclear Technologies Company for the B&V Nuclear Owneis Group Reactor Vessel Vorking Group in response to Generic Letter 92-01, Revision 1 and summarizes data from other Topical Reports.
The above topical reports provide the basis for demonstrating compliance with 10 CFR Part 50, Appendix G, Paragraph IV.A.1. provides the data summary for the Pressurized Thermal Shock (PTS) calculation and Enclosute 2 provides the data summary for the tipper-Shelf Energy (USE) calculation for the DBNPS.
The tables include the data necessary to perform JSE and RT evaluations.
Please note that certaindatainbothtablesarediff0!!ntfromToledoEdison's previous letter datt'd July 1, 1992 (Serial Number 2060). The enclosed tables are consistent with the tables contained in the Topical Report BAV-2222. Accordingly, Toledo Edison requests NRC approval of the above as the basis for demonstrating compliance with 10 CFR Part 50, Appendix G, Paragraph IV.A.1 for the DBNPS.
On May 19, 1994, Toledo Edison informed Mr. G. Vest, Jr., NRC Acting Project Manager, that based on the supporting work being performed by the B&V Nuclear Technologies Company for the B&V Owners Group, the response to the Generic Letter 92-01, Revision 1 vould be submitted by June 30, 1994, rather than 30 days after receipt of the NRC's request.
Should you have any questions or require additional information, please contact Mr. Villiam T. O'Connor, Manager - Regulatory Affairs, at (419) 249-2366.
Very truly yours, b( MC
/ +-
A
- dl j Enclesures cc:
J. B. Martin, Regional Admidistrator, NRC Region III S. Stasek, DB-1 NRC Senior Residen; Inspector R. J. Stiansky Jr., DB-1 NRC/NRR Projec. Manager Utility Radiological Safety Board i
1 i
l 1
m tn v1 &-tt S$$$0 y7@$
T@le 1, Davis-Besse -- Data Summary for Pressurited Themal Shock Calculation C
~ g +-. g n
,z x
I
$kb IS Neut.
Method of Lethod of
- 7 E o' Beltline Fluence at IRTm Detemin.
Chemistr/
Detemin.
O$$
Material Heat No.
EOL/EFPY F
IRTmr Factor CF
%Cu
%Ni w"w w2o Nozzle Belt ADB 203 1.50E+18'
+50 Plant 26 RG1.99 0.04' 0.68 Ukb 2
7 Forgin_g (123Y317)
( a,= 0 )
Specific-Table 2 2
7 7
Upper Shell AKJ 233 1.07E+19'
+20 Plant 26 RG1.99 0.04 0.77 Forging (123I244)
(a,=0 )
Specific Table 2 2
7' Lower Shell BCC 241
- 1. 07E + 19'
+50 Plant 15.5*
Calculated 0.02 O.81' Forging
. (SP4 086) --
(a =0)
St'acific r
8 8
Generic' 160.6-RG1.99 0.18' O.64' 2
WF-232 8T3914 1.50E+18
-5 NB to US (vi=19. 7) -
Table.1 Circ. Weld.
(ID 9%)
WF-233 T29744 N/A'
-5 Generic.
207.44s Calculated 0.23' O.68' 8
NB to UC' (ag=19. 7) 1 Ci c. Weld iOD'91%)
2 WF-182-1 821T44 1.07E+13'
+2 Plant.
'162.09' Calculated 0.24' O.63' US to LS (vi=0 )
Specific.
Cire. Weld f.
u____...
b
.m _ _..
.s._ _. _ __ _
i k
i e tn vs
- ts l m :s c - o h
Table L.-(cont.)
Davis-Besse -- Data Summary for Pressurized Thermal Shock Calcu1Fli.23 te on r
o e r tu pr i
o o :s c wmem" l
c c
t 8
i NOTES:
n Le
- s '.
e e c :z c t
9C9
- 1.
Values obtained from BAW-2108, Revision 1.**
~{${ j non i
2.
Values deter:Mncd from data in Materiel Test Reoort/ Weld Tesst Data.
u ' v, j
N: o }
3.
Mean value frcsu data in BAW-1803, Ravi61on 1.*2 U$0 i t
8 + 1 we :
4.
Chemistry Factor for. forging BCC 241 was determined using DB1 surveillance data as reported in BAW-2125."
t 5.
Chemistry Factor for weld metal WF-233 w?h determined using Kori Unit 'l surveillance data as reported in f
f BAW-1803, Revision 1.
4 6.
Chemistry Factor for weld metal WF-182-1 was determined using DB1 surveillance data as reported in BAW-2125.
[
t
~1 Values obtained from BAW-1820."
f I
8.
Values obtained from BAW-2121P."
l 4
1 I
r i
I t
4 t
i
?
I i
i h
i f
i -
l 1
-i
(
r 1
g-1 i
]
l t
t I
i i 1
nW w;a /w a YE'$CE Iable 2. Davis-Besse -- Data Stan ':ry for Urner-Shelf Enerov Calculatiqn nanoo ". o E $ [ .- u> w w r, c o ~ 2 [z[ 1/4T Method of Neutron Determin. HC 9 "[@[ Beltline Material 1/4T USE Fluence at Unirrad. Unirrari. Material Heat N). Type at EOL EOL USE l UL.:. Nozzle Belt ADB 2';3 A 508-2 122.6 8.40E+17 132' Direct' u 2 (Matl. Cert.) "7$ Forging (123'.317) wt-Upper Shell ATJ 233 A 508-2 125.0 5.99E+18 140 Direct *~ 2 (Sury. Matl.) Forging {123X244) Lower Shell BCC 241 A 508-2 110.4 5.99E+18 122 Direct
- 2 (Surv. Matl.)
_ Forging (5P4086) WF-232 BT3914 Linde 80 EMA' N/A EMA Generic' 18 to US SAW Cire. Weld (ID 9%) 8 2 EMA' Generic WF-233 T29744 Linde 80 . EMA' 8.40E+17 1m to US SAW Cire.' Weld-l (OD 91%) 2 EMA' Direct' WF-162-1 821T44 Linde 80 EMA' 5.99E+18 US to LS SAW (Suppl. Tect - Report) l Cire. Weld
~ - -. -. -.... -.. - -.. - -... -... ~. ...... -.... _.... ~.. _.. _... _ - - -.... -.. -... - -.. =.. -....... ,- l i m m m t" b ' m :s m
- o i
Table 2. (con'.) Davis-Besse -- Data Sumtarv for Uooer-Shelf Eneroy Calculation va anan m**mF o o :s a wm~mn e c NOTES: wz z 1 ecZc scp USE issue covered by t.he approved equivalent margins analysis in the Topical Reports BAW-2192P, $[@ uVUT 1. Re* 'on 1** and BAW-2178P" y n u, i wZo i 2. Values obtained from BAW-2192P, Revision 1. w" " u' ' j ml & 3. Unirradiated USE is determined from transverse oriented specimens. r Unirradiated USE is determined using data from other plants with similar materials to the beltline 4. I material. i L h l l b ( ) 4 + + 'i t i i i t + l i r .}}