ML20072T831
| ML20072T831 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 07/31/1994 |
| From: | Devan M, Gross L, Moore K BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20072T824 | List: |
| References | |
| BAW-2224, GL-92-01, GL-92-1, NUDOCS 9409150395 | |
| Download: ML20072T831 (32) | |
Text
.
BAW-2224 JULY 1994 l
l NORTH ANNA UNITS 1 AND 2 RESPONSE TO CLOSURE LETTER FOR NRC GENERIC LETTER 92-01, REVISION 1 l
NED@5098&WNUCLEARETEC U3
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l BAW-2224 l
July 1994 l
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NORTH ANNA UNITS 1 AND 2 RESPONSE TO CLOSURE LETTER FOR NRC GENERIC LETTER 92-01, REVISION 1 l
by l
M. J. DeVan l
Prepared for Virginia Power BWNT Document No. 77-2224-00 (See Section 7 for document signatures) l l
t Prepared by B&W NUCLEAR TECHNOLOGIES, INC.
Engineering and Project Services Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 BWit&Malag
l CONTENTS Page 1.
INTRODUCTION 1-1 2.
ORGANIZATION OF RESPONSE 2-1 3.
UPPER-SHELF ENERGY FOR FORGING 3-1 4.
NICKEL CONTENT FOR ROTTERDAM WELD.................
4-1 5.
RESPONSE TO CLOSURE LETTERS....................
5-1 6.
REFERENCES 6-1 7.
CERTIFICATION......
7-1 List of Tables Table 4-1 Nickel Content Data for Rotterdam Weld Metal Used to Determine Conservative Estimate Nickel Content 4-2 5-1.
North Anna Unit 1 -- Data Summary for Pressurized Thermal Shock C al c ul a t i o n............................
5-2 5-2.
North Anna Unit 1 -- Data Summary for Upper-Shelf Energy Calculation 5-4 5-3.
North Anna Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation............................
5-6 5-4.
North Anna Unit 2 -- Data Summary for Upper-Shelf Energy Calculation 5-8 ii BWTECHNOLOGIES BG W NUCLEAR
- 1. INTRODUCTION e
This report provides a response to the Generic Letter 92-01, Revision 1, closure letter recently issued by the U. S. Nuclear Regulatory Commission for Virginia Power's North Anna Unit 1 (NA1) and North Anna Unit 2 (NA2) Nuclear Station.
1-1 B W iennaitle:
b 5
4 i
I 3
f 1
2.
ORGANIZATION OF RESPONSE l
1 i
The Generic letter closure letter requested information which included the l
following:
1.
For, aclosure 1,
" Data Summary - fm Pressurized Thermal Shock,"
l verify that the information therein is accurate.
2.
For Enclosure 2, " Data Summary for Upper-Shelf Energy," verify that -
i li the information therein is accurate.
3.
Additional information.
i
]
(a)
Upper-shelf energy for forgings (b)
Nickel content for Rotterdam weld.
I I
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Expanded explanations for the additional information are presented La following-Sections 3 and 4.
l In Section 5 the full Data Summary Tables for Pressurized Thermal Shock and l
Upper-Shelf Energy are presented. Those' values that are unchanced are shown in t aded boxes. Where applicable, the table presentation was reordered from the h
Generic Letter 92-01 closure letter enclosures supplied by the Nuclear Regulatory 1
l Commission (NRC).
The order of presentation followed in this report is at
]
}-
follows:
I 1.
Wrought materials (forgings and plate) arranged from top to bottom as located in the reactor vessel 2.
Circular welds arranged from top to bottom as located in the reactor vessel
)
The NRC's closure letter for NA1 and NA2 is shown following the above information.
2-1 1
3 i
BWfisafEi!
3.
UPPER-SHELF ENERGY FOR N0ZZLE BELT FORGING The North Anna (NA) Updated Final Safety Analysis Report' (UFSAR) Table 5.2-26 indicates a Charpy absorbed energy value of 60 ft-lbs (tangential, unirradiated) for the reactor vessel nozzle belt Forging 05 at a test temperature of :s58*F.
(This upper-shelf data is repeated in WCAP-ll791,2 Table III.) If this test data were assumed to characterize the upper-shelf of this material, the initial Charpy V-notch upper-shelf energy (CvUSE) of this material in the axial orientation might be conservatively estimated by application of the guidelines in the Standard Review Plan, Branch Technical Position MTEB 5-2,
" Fracture Toughness Requirements," for estimating the Cv0SE which requires the tangential CvuSE values to be multiplied by 0.65.
(See also Memo from C. Z. Serpan, Jr. to C. Y.
Cheng, " Ratio of Transverse to Longitudinal Orientation Charpy Upper-Shelf Energy," dated June 25, 1990.) Applying this method would result in an estimated initial CvUSE of 39 ft-lbs in the axial orientation.
However, the 60 ft-lbs unirradiated Charpy absarbed energy value for Forging 05 in the tangential orientation is not representative of the material strength at upper-shel f conditions.
Further, there is evidence which supports the conclusion that the initial CvUSE in the axial orientation for the NAl Forging 05 is 74 ft-lbs or more. The basis for this conclusion is presented in the following paragraphs.
An unirradiated CvuSE estimate of 74 ft-lbs for Forging 05 is supported by consideration of the data in Table 5.2-26 of NA UFSAR (or Table III of WCAP-11791) and Appendix C of BAW-1638,4 " Analysis of Capsule V Virginia Electric and Power Company North Anna Unit No. 1 Reactor Vessel Materials Surveillance Program."
Figure C-2 of BAW-1638 presents Charpy curves which demonstrate an average CvuSE of 135 ft-lbs and a minimum CvuSE value of 120 ft-lbs in the tangential orientation for the Forging 03 surveillance material. The NA UFSAR Table 5.2-26 (or Table III of WCAP-ll791) presents an absorbed energy value of 74 ft-lbs in the tangential orientation for Forging 03 and a comparable absorbed 3-1 B W ?! n M e sta
energy value of 60 ft-lbs in the tangential orientation for Forging 05.
The values presented in Table 5.2-26 of the NA UFSAR are the minimum absorbed energy values at the highest test temperature (s68'F).
It is evident by inspection of the surveillance data for Forging 03 that the maximum test temperature (68'F) is considerably below the upper-shelf, since the tangentially oriented specimen data at 68'F (i.e., 74 ft-lbs) is less than the axially oriented specimen data at the upper-shel f (i.e., 85 f t-lbs). Similarly, it is evident that a 60 ft-lb absorbed i
energy value for a tangentially oriented specimen tested at 68'F is not consistent with the value which would be obtained for an axially oriented specimen tested at the upper-shelf, and certainly not for a tangentially oriented specimen tested at the upper-shelf.
Given th mon vendor and nearly identical chemical compositions of Forgings 03, 04.
no 25, the axially criented CvUSE for either Forging 03 (85 ft-lbs) or Forging 04 (92 ft-lbs) would serve as an appropriate surrogate for the unirradiated CvuSE for Forging 05 in the axial orientation.
As described in a Virginia Electric and Power Company letter to the NRC dated December 29, 1992,5 the CvUSE for NA2 Forging 05 in the axial orientation was equated to that of the NA2 surveillance Forging 04 (i.e., 74 ft-lbs) in the axial orientation on the basis of a similar, quantitative comparison of tangential Charpy test data taken below 68'F. The NA2 surveillance Forging 04 CvUSE value in the axial orientation (i.e., 74 ft-lbs) is coincidentally equal to the NAl surveillance Forging 03 measured Charpy data taken at a test temperature below 68'F in the tangential orientation. Based on this evidence, it can be concluded that the unirradiated i
CvUSE for the NAl Forging 05 in the axial orientation is greater than the unirradiated CvuSE for the NAl surveillance Forging 03 material in the tangential orientation at test temperature below 68'F.
Therefore, a conservative estimate for the unirradiated CvuSE for Forging 05 in the axial orientation is 74 ft-lbs.
Because no known Rheinstahl Huttenwerke reactor vessel forgings are considered
" low upper-shelf energy materials" (i.e., initial upper-shelf energy less than 8
the minimum 75 ft-lbs required by 10CFR50 Appendix G at the time of fabrication), it is concluded that an initial CvuSE of 74 ft-lbs for Forging 05 is a conservative estimate.
3-2 GB WTECHNOLOGIES IS B&W NUCLEAR
i l
u i
i 4.
NICKEL CONTENT i
The Pressurized Thermal Shock (PTS) Rule,10CFR50.61,7 requires that the nickel
{
content be the best-estimate value. The nickel content for Weld 058 in the NA2 reactor vessel is not available, therefore, it is necessary to determine a best-estimate nickel content based on data from similar material.
j According to the PTS Rule, the nickel content shall be obtained as follows:
i i
1.
Determine the mean value of measured. values for welds fabricated i
using the same heat number as that of the critical reactor vessel weld.
i 2.
If the above values are not available, the upper limiting values given in the material specifications to which the reactor vessel was j
built may be used.
l 3.
If the above are not available, conservative estimates (mean plus
{
one standard deviation) based on generic data (data from reactor vessels fabricated to the same material specification in the same shop and in the same time period) may be used if justification is l
provided.
d i
4.
If none of the above alternatives are available,1.0 percent nickel j
must be assumed.
i 1
For Weld 058 in the NA2 reactor vessel, no data is available to satisfy the first two alternatives described above, however, nickel content data are available for l
Rotterdam welds that satisfy the requirements of Alternative 3.
The nickel j
content for all welds fabricated in the same material specification in the same j
shop and in the same time period are listed in Table 4-1.
The mean was found to l
be 0.10 weight percent with a standard deviation of 0.01.
By the definition i
described in Alternative 3, the conservative estimat<; (mean plus one standard deviation) for the nickel content of NA2 Weld 058 is 0 11 percent.
}
3 i
i 4-1 1
1 4
B W 'ls M itlei
Table 4-1.
Nickel Content Data for Rotterdam Weld Metals Used to Determine Conservative Estimate Nickel Content Weld Wire Flux Ni, Pl ant Tvoe Heat No.
TYDe lot No, wt%
Reference 8
Ringhals Unit 2 S4 Mo 1725 SMIT 89 2275 0.084 WCAP-8216 8
Seo.uoyah Unit 2 S4 Mo 4278 SMIT 89 1211 0.11 WCAP-8513 Mean - 0.10 a = 0.01 Hean + a = 0.11 l
4-2 BWits&"e:a
4 5.
REVISED TABLES 1 AND 2 This section contains the revised Tables for Pressurized Thermal Shock and Upper Shelf Energy in the closure letter to Generic Letter 92-01, Revision 1, for NA1 and NA2.
5-1 B W "la naia n
Table 5-1.
North Anna Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation IS Neutron Method of Method of Beltline Fluence at IRTm Determin.
Chemistry Determin.
Material Heat No.
30.7 EFPY F
IRTy Factor CF
%Cu
%Ni Nozzle Belt 990286/
2. 7 7E + 18
+6 MTEB 5-2*'
121.5 RG1.99 0.16'S8 0.74'S' Shell 295213 Table 2
Forging 05 Interm. Shell 990311/
3. 9 5E+ 19
+17 Plant 86 RG1.99 0.12'S' O. 8 2 'S' Forging 04 298244 Specific
Table 2
Lower Shell 990400/
3. 9 5E +19
+38 Plant 88.9 Calculated'"
0.15 S' O. 8 0 'S' 8
Forging 03 292332 Specific
t Weld 05A 25295 2. 7 7E+ 18
O Generic'd) 143 8 RG1.99 0.30'S' O.17*)
NB to IS Table l
m 4
Weld 05B 4278 2. 7 7 E + 18
O Genericid '
59.5 RG1.99 0.11'S' O.11*)
NB to IS Table l
Cire. Weld (ID 6%)
Weld 04 25531 3. 9 5E + 19
+19 Plant 93.1 Calculated'"
0.09'S) 0.11'S' IS to LS Specific
Cire. Held i
i I
1
.m
-. _ -. ~.
Table 5-1.
(cont.)
North Anna Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation NOTES:
a.
Values obtained from WCAP-11777.20 (Nozzle belt shell forging and nozzle belt shell-to-intermediate shell circumferential weld fluences are 7% of maximum vessel inner surface fluence.)
b.
Initial reference temperature was determined in accordance with MTEB 5-2 guidelines for cases where the reference temperature was not determined using ASME Boiler and Pressure Vessel Code,Section III, NB-2331,2 methodology.
Initial reference temperature was determined from tests on material fabricated from the same heat of the c.
beltline material.
d.
Initial reference temperature was determined from the mean value of tests on material of similar types.
Chemistry factor was determined from the chemistry factor tables in Regulatory Guide 1.99, Revision 2.22 e.
f.
Chemistry factor was determined from surveillance data (WCAP-11777) via procedures described in Regulatory Guide 1.99, Revision 2.
I g.
Chemical content obtained from BAW-1911, Revision 1.22 w
h.
Chemical content obtained from Sequoyah Units 1 and 2 data.' 2*
(Same weld wire heat numbers. )
l l
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+
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Table 5-2.
North Anna Unit 1 -- Data Summary for Upper-Shelf Enerov Calculation 1/4T Method of 1/4T USE Neutron Determin.
Beltline Material at 30.7 Fluence at Unirrad.
Unirrad.
Material Heat No.
Type EFPY 30.7 EFPY USE USE Nozzle Belt 990286/
A 508-2 62
- 1. 7 5 E + 18
74 See Section 3*
Shell 295213 Forging 05 Interm. Shell 990311/
A 508-2 68 2. 4 9E+19
92 Direct *8 i
Forging 04 298244 Lower Shell 990400/
A 508-2 60 2. 4 9E+19
85 Direct
i Forging 03 292332 Weld OSA 25295 SMIT 89, 78~
- 1. 7 5E+18
111 Sister NB to IS SAW Plant'd' Cire. Weld
<n (OD 94%)
Weld 05B 4278 SMIT 89,
--i
1.75E+18
105 Sister NB to IS SAW Plant'd' Circ. Weld (ID 6%)
Weld 04 25531 SMIT 89, 73 2. 4 9E +19 '*'
102 Direct"'
__....m
.___.__.___.__.__..-._.--m._-.
Table 5-2.
(cont.)
North Anna Unit 1 -- Data Summary for Upper-Shelf Enerciv Calculation NOTES:
a.
End-of-life neutron fluence at T/4 from inner wall calculated using Regulatory Guide 1.99, Revision 2, neutron fluence attenuation methodology from ID value. (Vessel thickness = 7.667 in.)
b.
The unirradiated USE was determined on the basis of a comparison with similar materials to the beltline material. (See Section 3.)
c.
The unirradiated USE for the forgings was determined from weak oriented specimens. The unirradiated USE for the weld was determined from test data.
d.
The unirradiated USE was determined using reported data from other plants with the same weld wire heat number (Sequoyah Units 1 and 2' ").
e.
Weld 05A is 94% of the thickness of the nozzle belt shell-to-intermediate shell circumfere..tial weld and Weld 05B is the remainder. Therefore, it is not necessary to evaluate the end-of-life USE for Weld 05B because it is not at the T/4 location.
us e
ul i
i I
i
.m.
m m.
s Table 5-3.
North Anna Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation IS Neutron Method of Method of Beltline Fluence at IRTm Determin.
Chemistry Determin.
Material Heat No.
32 EFPY F
IRTyy Factor CF
%Cu
%Ni Nozzle Belt 990598/
3.13 E + 18
+9 MTEB 5-2'"'
51 RG1.99 0.08'S) 0.77'S' Shell 291396 Table 2 *'
4 Forging 05 Interm. Shell 990496/
4. 4 7 E + 19
+75 Plant 47.9 Calculated'"
0.09(S' O. 8 3 'S' Forging 04 292424 Specific'*'
Lower Shell 990533/
4. 4 7E+19
+56 Plant 96 RG1.99 0.13(8) 0.83'S' Forging 03 297355 Specific'*'
Table 2'*'
Weld 05A 4278 3.13 E + 18
O Generic'd' 59.5 RG1.99 0.11'S' O.11'h' i
NB to IS Table l*'
m Weld 05B 801 3.13 E +18 '*'
O Generictd' 87.8 RG1.99 0.18'S' O.11'"
i*
NB to IS Table l
Cire. Weld (ID 6%)
Weld 04 716126 4. 4 7E+19
-48 Plant 10.4 Calcula ted'"
0.09 S' O. 08 'S' t
tei IS to LS Specific Circ. Weld l
l
Table 5-3.
(cont.)
North Anna Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation NOTES:
a.
Values obtained from WCAP-12497."
(Nozzle belt shell forging and nozzle belt shell-to-intermediate shell circumferential weld fluences are 7% of maximum vessel inner surface fluence.)
b.
Initial reference temperature was determined in accordance with MTEB 5-2 guidelines for cases where the reference temperature was not determined using ASME Boiler and Pressure Vessel Code,Section III, NB-2331, methodology.
l c.
Initial reference temperature was determined from tests on material fabricated from the same heat of the beltline material.
d.
Initial reference temperature was determined from the mean value of tests on material of similar types.
e.
Chemistry factor was determined from the chemistry factor tables in Regulatory Guide 1.99, Revision 2.
f.
Chemistry factor was determined from surveillance data (WCAP-12497) via procedures described in Regulatory Guide 1.99, Revision 2.
g.
Chemical content obtained from BAW-1911, Revision 1.
h.
Chemical content obtained from Sequoyah Unit 2 data.'
(Same weld wire heat number.)
i.
Conservative estimate (mean plus one standard deviation) determined using data from other plants (Ringhals Unit 2 and Sequoyah Unit 2) with similar materials to the beltline material, i.e.,
data from reactor vessels fabricated to the same material specification in the same shop and in the same time period).
(See Section 4.)
l
' ~ - - - - - - - -
m Table 5-4.
North Anna Unit 2 -- Data Summary for Upper-Shelf Energy Calculation 1/4T Method of 1/4T Neutron Determin.
Beltline Material USE at Fluence at Unirrad.
Unirrad.
Material Heat No.
Type 32 EFPY 32 EFPY USE USE Nozzle Belt 990598/
A 508-2 64
- 1. 9 8 E+ 18
74 Equiv. to Shell 291396 Forging 04'b' Forging 05 Interm. Shell 990496/
A 508-2 56 2.82E+19*
74 Direct
- Forging 04 292424 Lower Shell 990533/
A 508-2 58 2. 8 2 E +19 '*2 80 Direct
Forging 03 297355 Weld 05A 4278.
SMIT 89, 87 1.98E+18
105 Sister NB to IS SAW Plant
- Circ. Weld f
(OD 94%)
co Weld 05B 801 SMIT 89, 1.98E+18'**
---(*)
-- N NB to IS SAW Circ. Weld (ID 6%)
Weld 04 716126 LW 320, 76 2.82E+19*
107 Direct
~.m
..m__._
_.m._. _.......
Table 5-4.
(cont.)
North Anna Unit 2 -- Data Summary for Upper-Shelf Enerov Calculation NOTES:
a.
End-of-life neutron fluence at T/4 from inner wall calculated using Regulatory Guide 1.99, Revision 2, neutron fluence attenuation methodology from ID value. (Vessel thickness = 7.667 in.)
b.
Letter from W.
L.
Stewart, Virginia Electric and Power Company,-to U.
S. Nuclear Regulatory Commission,
Subject:
Virginia Electric and Power Company North Anna Power Station Unit 2 Selection of Limiting Forged Material for Low Upper-Shelf Energy Considerations, dated December 29, 1992.5 c.
The unirradiated USE for the forgings was determined from weak oriented specimens. The unirradiated USE for the weld was determined from test data.
d.
The unirradiated USE was determined using reported data from other plants with the same weld wire heat number (Sequoyah Unit 2').
e.
Weld 05A is 94% of the thickness of the nozzle belt shell-to-intermediate shell circumferential weld and Weld 05B is the remainder. Therefore, it is not necessary to evaluate the end-of-life USE for Weld 05B because it is not at the T/4 location.
tn e
10
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cwey 31 e-7 E
UNITED STATES i
f NUCLEAR REGULATORY COMMISSION kD E
WASHINGTON, D.C. 20555-0001 May 31, 1994 Docket Nos. 50-338 and 50-339 Mr. J.P. O'Hanlon Senior Vice President - Nuclear 5000 Dominion Blvd.
Glen Allen, Virginia 23060
Dear Mr. O'Hanlon:
SUBJECT:
GENERIC LETTER (GL) 92-01, REVISION 1, " REACTOR VESSEL STRUCTURAL INTEGRITY," VIRGINIA ELECTRIC AND POWER COMPANY (VEPCO) NORTH ANNA POWER STATION, UNITS NO. 1 AND N0. 2 (NA-l&2) (TAC N05. M83488 AND M83489)
By letters dated May 29, 1992, September 28, 1992, October 22, 1992, December 29, 1992, September 23, 1993, and February 9, 1994, you provided a response to GL 92-01, Revision 1.
The NRC staff has completed its review of your responses.
The GL is part of the staff's program to evaluate reactor vessel integrity for Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs).
The information provided in response to GL 92-01, including previously docketed information, is being used to confirm that licensees satisfy the requirements and commitments necessary to ensure reactor vessel integrity for their facilities.
A substantial amount of information was provided in response to GL 92-01, Revision 1.
These data have been entered into a computerized data base designated the Reactor Vessel Integrity Database (RVID).
The RVID contains the following tables:
a pressurized thermal shock (PTS) table for PWRs, a pressure-temperature limits table for BWRs and an upper-shelf energy (USE) table for PWRs and BWRs. provides the PTS tables.
provides the USE tables for NA-1&2, and Enclosure 3 provides a key for the nomenclature used in the tables.
The tables include the data necessary to perform USE and RT,iously docketed information.These data were taken from your respo evaluations.
g to GL 92-01 and prev References to the specific source of the data are provided in the tables.
We have determined that additional data is required to confirm that the USE at end-of-life (E0L) for one of your beltline materials, forging 05, for NA-1, is greater than 50 ft-lb because you have provided a generic mean value for the unirradiated USE. These types of values are unacceptable because they do not consider material variability.
When the unirradiated USE for a particular material has not been determined, you can set the' USE equal to the lower tolerance limit calculated for the group of similar materials.
The unirradiated USE should be determined such that there exists 95% confidence 1
5-10 1
Mr. J. P. O'Hanlon that at least 95% of the population is greater than the lower tolerance limit.
If the lower tolerance limit results in a projected USE at E0L of less than 50 ft-lb, then you must demonstrate, in accordance with Appendix G, 10 CFR Part 50, that lower values of USE will provide margins of safety against fractura
)
equivalent to those required by Appendix G of Section III of the American i
Society of Mechanical Engineers Boiler and Pressure Vessel Code. We request that, within 30 days receipt of this letter, you submit a schedule for providing this required data.
Additionally, we have determined that additional data is required to confirm the value provided for the nickel content of weld 058 of the NA-2 reactor vessel.
The value of 0.10 provided in the GL 92-01 submittal was cited as an
" estimated" value.
However, the supporting data and methodology for determining the estimated value were not provided.
The Pressurized Thermal Shock (PTS) Rule,10 CFR 50.61, requires that the amounts of copper and nickel be best-estimate values.
According to the PTS Rule, a mean value is acceptable for welds fabricated using the same heat number as that which matches the critical reactor vessel weld.
If these values are unavailable, upper limiting values given in the material specifications to which the reactor vessel was built may be used.
If not available, conservative estimates (mean plus one standard deviation) based on generic data (data from
(
reactor vessels fabricated to the same material specification in the same shop as your vessel and in the same time period) may be used if justification is provided.
If none of these alternatives are available,1.0 percent nickel must be assumed.
We request that you provide the Westinghouse Owners Group (WOG) data that was used to determine the amount of nickel and that you determine the best-estimate amount of nickel in accordance with the PTS Rule, 10 CFR 50.61, within 30 days of receipt of this letter.
Further, we request that you verify that the information you have provided for i
NA-l&2 has been accurately entered in the summary file.
If no comments are made in your response to this request, the staff will use the information in the tables for future NRC assessments of your reactor pressure vessels.
Once your response is received and your schedule is determined to be satisfactory, the staff will consider your actions related to GL 92-01, Revision 1, to be complete.
When your analyses are submitted, they will be reviewed as plant-specific licensing actions.
The information requested by this letter is within the scope of the overall burden estimated in G!. 92-01, Revision 1, " Reactor Vessel Structural Integrity,10 CFR 50.54(f)." The estimated average number of burden hours is 200 person hours for each addressee's response.
This estimate pertains only to the identified response-related matters and does not include the time i
5-11
i Mr. J. P. O'Hanlon required to implement actions required by the regulations.
This action is covered by the Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994.
Sincerely, s-90sdfbi%
\\~
Lenn B. Engle, aject Manager Project Direct te II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Pressurized Thermal Shock Tables 2.
Upper-Shelf Energy Tables 3.
Nomenclature Key cc w/ enclosures:
See next page i
5-12
Mr. J. P. O'Hanlon North Anna Power Station Virginia Electric & Power Company Units 1 and 2 CC*
Mr. William C. Porter, Jr.
Robert B. Strobe, M.D., M.P.H.
County-Administrator State Health Commissioner Louisa County-Office of the Commissioner P.O. Box 160 Virginia Department of Health i
Louisa, Virginia 23093 P.O. Box 2448 i
Richmond, Virginia 23218 J
. Michael W. Maupin, Esq.
~
Hunton and Williams Regional Administrator, RII j
Riverfront Plaza, East Tower U.S. Nuclear Regulatory Commission 951 E. Byrd Street 101 Marietta Street, N.W., Suite 2900 Richmond, Virginia 23219 Atlanta, Georgia 30323 1
i Dr. W. T. Lough j
Virginia State Corporation Commission Mr. J. A. Stall, Manager l
Division of Energy Regulation North Anna Power Station i
P.O. Box 1197 P.O. Box 402 4
Richmond, Virginia 23209 Mineral, Virginia 23117 Old Dominion Electric Cooperative 4201 Dominion Blvd.
Glen Allen, Virginia 23060 4
I j
Mr. M. L. Bowling, Manager 1
Nuclear Licensing & Programs j
Virginia Electric and Power Company.
Innsbrook Technical Center 4
5000 Dominion Blvd.
j Glen Allen, Virginia 23060 i
Office of the Attorney General Supreme Court Building 4
101 North 8th Street Richmond, Virginia 23219 i
+
i Senior Resident Inspector North Anna Power Station U.S. Nuclear Regulatory Commission i
Route 2, Box 78 i
Mineral, Virginia 231172 4
i i
5-13
.- ~.
Sumary File for Pressurized Thermal Shock Ptsnt Settline Heat No.
10 Neut.
I R T.
Method of Chemistry Method of
%Cu 2Ni Name Ident.
Ident.
Fluence at Determin.
Factor Determin.
EOL 127 CF North Nozzle 990286/
2.51E18 6*F MTEs 5 2 121.5 Table 0.16 0.74 1
Anna 1 shell 295213 forging 05 EOL:
Int, shell 990311/
3.95E19 17'F Plant 86 Table 0.12 0.82 4/1/2018 forging 04 298244 specific Lowr 990400/
3.95E19 38'F Plant 73.503 Calculated 0.16 0.80 shelt 292332 specific forging 03 Weld 04 25531 3.95E19 19'T Plant 93.089 calculated 0.09 0.11 specifle j
Weld 05A 25295 2.7BE18 0'F Generic 138.5 Table 0.30 0.17 l
Weld 058 4278 2.78E18 0*F Ceneric 58.5 Table 0.11 0.11 i
References a
j The nickel contents for weld 05A and 058 are values from sequoyah 1&2. (same weld wire heat nu@ers).
1 Chemical concosition arc IRT. data are f rom SAW 2168, iAlch is attached to the CL 92-01 response.
I j
Fluence cates WCAP 11777: 10 EOL fluence is 3.95E19 n/ca' j
Tabte 2 1 of SAW 2144, d ich is attached to Decameer 27, 1991, letter from W. L. stewart (VPCo) to UsNRC a
Occument Controt Desk, adaject:
Re< pest to Change Technical specifications: Pressure / Temperature Limitations, Low Topreture/ Overpressure Protection System Setpoints, states that forging 05, and welds 05A ard 058 have fluences that dif fer f rom forgings 03 and 04, and weld 04 i
Note:
Chemical ccuposition values for forging as are averages from the beltline material data and the surveillance
- cata, f
1 A margin of 69'F (a, = 20*F aA = 28'F) has to be used f or wld 05A and weld 05B for which a generic 127, of O'F has been oerived.
I 3
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l 4
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i 2
5-14 s
5'
l Sunenary File for Prassurized Therinal Shock l
Plent Beltline Heat No.
10 Neut.
I R T.
Method of Chemistry Method of
%Cu 1Ni Nane Ident.
Ident.
Fluence at Determin.
Factor Deternin.
ECL I R T.
CF xorth upper 990598/
4.47E19 9'F MTE8 5 2 51 Table 0.08 0.77 Anna 2 shell 291396 forging 05 EOL:
Int. shell 990496/
4.47E19 75'F Plant 35.112 Calculated 0.10 0.85 8/21/2020 forging 04 292424 spectfIe Lower 990533/
4.47E19 56'F Plant 96 Table 0.13 0.83 shell 207355 specific forging 03 Weld 04 716126 4.47E19 48'F Plant 10.398 Calculated 0.09 0.08 soecific Weld 05A 4278 4.47E19 0*F ceneric
$8.5 Table 0.11 0.11 Weld 058 801 4.47E19 0*F Generic 97.0 Table 0.18 0.10 n,fuences The nickel content for weld 05A is from segaoyah 2 (the same weld wire heat romber and the same flux).
8AW-2168, which is attached to Jww 29, 1992, letter from W. L. Stewart (VPCo) to USNRC Document Control Desk, s@ ject:
Response to Generic Letter 92 01, Reactor Vessel structural Integrity, contains chemical careposition and the initial RT,
(IRT.) data for all the beltline materials Fluence is fecue Table 6-13 of WCAP 12497 Note:
Chemical coeposition values for forging 04 are averages from the beltline material data and the surveillance data.
A margin of 69'F (a, = 20*F aa = 28'F) has to be used for weld 05A and weld 058 for which a generic IRT., of O'F has been derived.
7Additional information required to confirm value 5-15
Sumary File for Upper Shelf Energy Plant Name Beltline Meat No.
Material 1/4T USE 1/4T unirrad.
Method of Ident.
Type et EOL Woutron USE Determin.
Fluence at Unirrad.
E0L USE North Arina Nozzle 990286/
A 508 2 62 1.74E18 75' Generic 1
shell 295213 j
forging 05 ECL Int. shell 990311/
A 508-2 68 2.49E19 92 Direct 4/1/2018 forging 04 298244 1
Lower 990400/
A 508 2 58 2.49C19 85 Direct shelL 292332 forging 03 i
cire. Wold 25531 SMIT 89, 71 2.49E18 102 Direct i
Weld 04 sAW 1
Nozzle to 25295 SMIT 89, 78 1.74E18 111 sister Int. Shell sAW Plant
~
Weld 05A Wozzle to 4278 SMIT 89, 1.74E18 105 Sister Int. SheLL SAW Plant Weld 058 e
References The fluence data for weld 04 is from June 29, 1992 letter to WRC (tesponse to GL 92 01);
fluence data for other materials are from september 23, 1993 letter to unc (tesponse to CL 92 01 ItA!).
Chemical composition and UUSE data for forging 03 are f rcan BAW-2168, sich is attached to the CL 92-01 response.
WSE data f or forging 04 and weld 04 are f rom BAW 1911, Rev.1.
Note: Weld 05A is 94% of thickness of the Nozzle to intermediate shell weld and 058 is the remainder. Therefore, it is not necessary to evaluate the EOL USE for weld 058 because it is not at the 1/4T location.
7Additional information required to confirm value 5-16
Sumary File for Upper Shelf Energy Plant dame Beltline Heat No.
Material 1/4T USE 1/47 Unirred.
Method of Ident.
Type at EOL Neutron USE Determin.
Fluence at Unfered.
EOL USE North Ame Upper 990598/
A 508 2 64 2.0E18 74 Egulv. to 2
shell 291396 forsing 04 forging 05 E0L:
Int sheLL 990496/
A 508-2 51 2.82E19 74 Direct 8/21/2020 forging 04 292424 Lower 990533/
A 508 2 58 2.82E19 80 Direct shell 207355 forging 03 Weld 04 716126 LW 320, 69 2.82E19 107 Direct SAW Weld 05A 4278 SMIT 89, 86 2.82E19 105 Siser SAW Plant Weld 05B 801 SMIT 89, 2.82E19 SAW Refererwes BAW 2168, which is attached to Jme 29, 1992, letter from W. L. Stewart (VPCo) to USNRC Document Control Desk, subject: Response to Generic Letter 92-01, Reactor vessel structural Integrity, contains chemical composition data for all the beltline materials. However, it contains WSEs for forging 04 and weld 04 only.
Fluence and WSEs for forgings 05 and 03 are from Deceeer 29, 1992 letter to NRC.
Note: Weld 05A is 94% of thicknees of the Wozzle to intermediate shell weld and 058 la the remainder. Therefore, it is not necessary to evoluete the EOL USE for weld 055 because it is not at the 1/4T location.
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5-17 i
i PRESSURIZED THERMAL SHOCK AND USE TABLES FOR ALL PWR PLANTS NOMENCLATURE
{
Pressurized Thermal Shock Table i
Column 1: Plant name and date of expiration of license.
2 Column 2: Beltline material location identification.
i l
Column 3: Beltline material heat number; for some welds that a single-i i
wire or tandem-wire process has been reported, (S) indicates j
single wire was used in the SAW process, (T) indicates tandem wire was used in the SAW process.
Column 4:
End-of-life (E0L) neutron fluence at vessel inner wall; cited directly from inner diameter (ID) value or calculated by using i
Regulatory Guide (RG) 1.99, Revision 2 neutron fluence attenuation methodology from the quarter thickness (T/4) value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals) j Column 5: Unirradiated reference temperature, j
Column 6: Method of determining unirradiated reference temperature 4
(IRT).
Plant-Soecific This indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.
MTEB 5-2 This indicates that the unirradiated reference temperature was determined from following MTEB 5-2 guidelines for cases where the IRT was not determined using American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, NB-2331, methodology.
Generic This indicates that the unirradiated reference temperature was determined from the mean value of tests on material of similar types.
Column 7: Chemistry factor for irradiated reference temperature evaluation.
Column 8: Method of determining chemistry factor Table This indicates that the chemistry factor was determined from the chemistry factor tables in RG 1.99, Revision 2.
Calculated This indicates that the chemistry factor was determined from surveillance data via procedures described in RG 1.99, Revision 2.
5-18
Column 9:
Copper content; cited directly from licensee value except when more than one value was reported.
(Staff used the average value in the latter case.)
No Data This indicates that no copper data has been reported and the default value in RG 1.99, Revision 2, will be used by the staff.
Column 10: Nickel content; cited directly from licensee value except when more than one value was reported.
(Staff used the average value in the latter case.)
No Data This indicates that no nickel data has been reported and the default value in RG 1.99, Revision 2, will be used by the staff.
Upper Shelf Energy Table Column 1:
Plant name and date of expiration of license.
Column 2:
Beltline material location identification.
Column 3: Beltline material heat number; for some welds that a single-wire.or tanden-wire process has been reported, (S) indicates single wire was used in the SAW process.
(T) indicates tanden wire was used in the SAW process.
Column 4:
Material type; plate types include A 5338-1, A 3028, A 302B Mod., and forging A 508-2; weld types include SAW welds using Linde 80, 0091, 124, 1092, ARCOS-B5 flux, Rotterdam welds using Graw Lo, SMIT 89, LW 320, and SAF 89 flux,' and SMAW welds using no flux.
Column 5:
EOL upper-shelf energy (USE) at T/4; calculated by using the EOL fluence and either the cooper value or the surveillance data.
(Both methods are described in RG 1.99, Revision 2.)
f2B This indicates that the USE issue may be covered by either owners group or plant-specific equivalent margins analyses.
Column 6:
EOL neutron fluence at T/4 from vessel inner wall; cited directly from T/4 value or calculated by using RG 1.99, Revision 2 neutron fluence attenuation methodology from the ID value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
5-19 h
L-_.
Column 7: Unirradiated USE.
Ebb This indicates that the USE issue may be covered by either owners group or plant-specific equivalent margins analyses.
Column 8:
Method of determining unirradiated USE Direct For plates, this indicates that the unirradiated USE was from a transverse specimen.
For welds, this indicates that the unirradiated USE was from test date.
.ftil This indicates that the unirradiated USE was 65% of the USE from a longitudinal specimen.
~
Generic This indicates that the unirradiated USE was reported by the licensee from other plants with similar materials to the beltline material.
NRC aeneric This indicates that the unirradiated USE was derived by the staff from other plants with similar materials to the beltline material.
- 10. 30. 40. or 50 *F This indicates that the unirradiated USE was derived from Charpy test conducted at 10, 30, 40, or 50 'F.
1 i
Sury. Weld i
This indicates that the unirradiated USE was from the surveillance weld having the same weld wire heat number.
Eauiv. to Sury. Weld This indicates that the unirradiated USE was from the surveillance weld having different weld wire heat number.
Sister Plant This indicates that the unirradiated USE was derived by using the reported value from other plants with the same weld wire heat number.
Bl ank indicates that there is insufficient data to determine the unirradiated USE.
5-20
.-m y-,.
, -, -. y r
w-g y
- l 6.
REFERENCES 1.
VEPC0 North Anna Power Station Units 1 and 2, Final Safety Analysis Report, USNRC Docket Nos. 50-338 and 50-339.
2.
J. C. Schmertz, " Analysis of Capsule U from the Virginia Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program, North Anna Unit 1 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation," WCAP-ll791, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, May 1988.
3.
U.S.
Regulatory Commission, Standard Review Pl an, Branch Technical Position MTEB 5-2, Revision 1, " Fracture Toughness Requirements," NUREG 0800, July 1981.
4.
A. L. Lowe, Jr., et al., " Analysis of Capsule V Virginia Electric and Power Company North Anna Unit No.1 Reactor Vessel Materials Surveillance Program," BAW-1638, Babcock & Wilcox Nuclear Power Generation Division, Lynchburg, Virginia, May 1981.
5.
Letter from W. L. Stewart, Virginia-Electric and Power Company, to V. S.
Nuclear Regulatory Commission,
Subject:
" Virginia Electric and Power Company North Anna Power Station Unit 2 Selection of Limiting Forged Material for Low Upper-Shelf Energy Considerations," December 28, 1992.
6.
Code of Federal Regulations, Title 10, Part 50, Domestic. Licensing of Production and Utilization Facilities, Appendix G, " Fracture Toughness Requirements."
7.
Code of Federal Regulations, Title 10, Part 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Event.s."
8.
S. E. Yanichko and D. J. Lege, "Swedish State Power Board Ringhals Unit No.
2 Reactor Vessel Radiation Surveillance Program,"
WCAP-8216, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1973.
i 9.
J. A. Davidson, J. H. Phillips, and S. E. Yanichko, " Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-8513, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1975.
6-1 B W its :salet
l 10.
S. E. Yanichko, et al., " Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-ll777, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, February 1988.
11.
American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, " Nuclear Power Plant Components," NB-2331,1989 Edition.
12.
U. S. Nuclear Regulatory Commission, " Radiation Embrittlement of Reactor Vessel Materials," Reaulatory Guide 1.99. Revision 2, May 1988.
13.
A.
L.
- Lowe, Jr.,
" Reactor Pressure Vessel and Surveillance Program Materials Licensing Information for North Anna Units 1 and 2," BAW-1991.
Revision 1, Babcock & Wilcox Nuclear Power Division, Lynchburg, Virginia, August 1986.
14.
S.
E.
- Yanichko, D.
J.
Lege, and J.
H.
Phillips, " Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel Radiation Surveillance Program," WCAP-8233, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1973.
15.
E. Terek, S. L. Anderson, and L. Albertin, " Analysis of Capsule U from the Virginia Electric and. Power Company North Anna-Unit 2 Reactor Vessel Radiation Surveillance Program,"
WCAP-12497, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, January 1990.
6-2 B W ilc %"H e lai
[
7.
CERTIFICATION This report accurately responds to the request for information stated in the closure letter to Generic Letter 92-01, Revision 1, for North Anna Unit I and Unit 2.
3b
/
7/27/W M. J. IreVan, Engineer III Date Materials and Structural Analysis Unit This report was reviewed and found to be accurate.
Y A? ?f L. 2. Gross, Advisory Engineer'
/ Date Materials and Structural Analysis Unit Verification of independent review.
78Y
'K. E. Moore, Manager Dhte Materials and Structural Analysis Unit This report is approved for release.
f
})ill. Howe l ML
]
7rogram Manager 6
'Date wners Gr up Projects x
P-1
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