ML20045F036
| ML20045F036 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/02/1993 |
| From: | Horn G NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-92-01, GL-92-1, NSD930782, NUDOCS 9307060341 | |
| Download: ML20045F036 (10) | |
Text
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GENERAL OFFICE
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ipg' P O. BOX 499. COLUMBUS. NEDRASKA 686024499 M.g g Nebraska Public Power District TELEPHONE (402) s64-Bs61
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NSD930782 July 2, 1993 U.S. Nuclear Regulatory Conunission Attention:
Document Control Desk k'a sh i ng t on, DC 20555 Gentlemen:
Subject:
Response to Request For Additional Information Generic Letter 92-01, Revision 1 Cooper Nuclear Station, NRC Docket 50-298, DPR-46
Reference:
Letter from Harry Rood (NRC) to G.
R. Horn, NPPD, dated April 27, 1993, " Request For Aduitional Information Regarding Response to Generic Letter 92-01, Revision 1" The Nebraska Public Power District (D1 strict) hereby provides its response to the NRC's request for additional information concerning the District's initial response to Generic Letter 92-01, Revision 1.
Generic Letter 92-01, Revision 1 requested licensees to provide information to demonstrate compliance with NRC regulations concerning reactor vessel f racture toughness requirements and vessel material surveillance programs.
In the District's initial response to GL 92-01, Revision 1 the District provided a detailed discussion of the bases for its compliance with these NRC regulations.
The attached discussion supplements the initial submittal and responds to the questions raised in the referenced request.
Please contact me if you have any questions.
Sinc # rely, c.,
G tt. Ilorn h% lear Power Group Manager Attachment cc:
NRC Regional Administrator Region IV Arlington, TX NRC Resident Inspector Cooper Nuclear Station
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Attachment to NSD930782 Page 1 of 6 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GENERIC LETTER 92-01, REVISION 1 REACTOR VESSEL STRUCTURAL INTEGRITY I.
INTRODUCTION The information in this attachment provides the Nebraska Public Power District's (District's) response to the NRC's Request for Additional.
Information (RAI)U concerning the District's July 1, 1992 responseU to Generic Letter 92-01, Revision 1.U Generic Letter 92-01, Revision 1 requested licensees to respond to a number of questions intended to verify that licensees have adequately. addressed the. requirements of the following:
.t 10 CFR 50.60, " Acceptance Criteria for Fracture Prevention Measures
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for Lightwater Nuclear Power Reactors for Normal Operation" 10 CFR 50 Appendix G, " Fracture Toughness Requirements"
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10 CFR 50 Appendix H, " Reactor Vessel Material Surveillance Program
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Requirements" i
And for Pressurized Water Reactors, 10 CFR 50.61,
" Fracture
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Toughness Requirements for Protection Against Pressurized Ihcrmal-Shock Events" The District's initial response to the generic letter discussed the-District's compliance with these requirements with respect to Cooper Nuclear Station.
While it is the District's position that these requirements are met for CNS, and this compliance has been adequately demonstrated to the NRC, the District provides the following response to the NRC's RAI, which discusses in greater detail,' the District's bases for demonstration of compliance with the above requirements.
II.
DISCUSSION The following discussion restates the questions raised'in the NRC's RAI, and provides the corresponding District response ~.
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'h 1.
Letter from H. Rood (NRC) to C. R. Horn (NPPD) dated April 27, 1993,.
" Request For Additional Information Regarding Response to Generic Letter 92-01,' Revision 1 "
2.
Letter.from G. R, Horn (NPPD) to NRC dated July 1,1992, " Response to Generic Letter 92-01, Revision 1."
3.
NRC Generic Letter 92-01, Revision 1, dated March 6,1992, " Reactor -
Vessel Structural Integrity, 10 CFR 50.54(f)."
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Attachment to NSD930782 i
Page 2 of 6 i
OUEST10N RECARDING ITEM 2.a j
"Your response to GL 92-01 indicates that the initial upper-shelf energy (USE) for all beltline welds, except for the surveillance weld, is not known.
Either provide the Charpy USE for each beltline weld or provide same
.i the Charpy USE and analysis from welds that were fabricated by the vendor, in the same time frame, using the same fabrication process and material specification to demonstrate that the surveillance weld is representative of beltline welds and that all beltline welds will meet the USE requirements of Appendix G to 10CFR Part 50. The analysis should take into account that the measured percent drop in USE exceeded the predicted percent drop obtained by using Regulatory Guide 1.99, Revision 2."
DISTRICT RESPONSE l
While it is the District's position that CNS currently is. in compliance with the requirements of 10 CFR 50, Appendix G,
the District has undertaken or been involved with several efforts to further demonstrate to.
i the NRC that these requirements are met, and ultimately that adequate j
margin exists to ensure that the CNS reactor vessel will not be subject to brittle fracture at any time during its design lifetime As discussed in the District's initial response to Generic Letter 92-01 l
Revision 1, the District does not have initial USE data for the CNS vessel l
beltline weld materials.
Development of a full Charpy impact testing curve to obtain USE was not a requirement of the ASME code at the time of CNS vessel fabrication.
Rather, the weld materials were tested at 10*F with a 30 ft-lb minimum requirement. Likewise, other vessels constructed same time frame had similar testing requirements; therefore, l
during the initial USE data was typically not generated during fabrication for other t
vessels constructed with similar materials during the same time frame.
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I However, as a result of subsequent research efforts, CE has acquired some initial USE data for Linde 1092 welds similar to those used in the CNS l
vessel beltline. Therefore, the District's best current prediction of CNS beltline weld USE at 32 EFPY is conservatively based on the lower bound initial USE value in that data base of 98 ft-lbs.
Assuming that the CNS l
vessel beltline welds initial USE equals this lower bound, the District currently predicts the USE for the limiting CNS beltline weld to be 68.1 ft-lbs at 32 EFPY, well above the 10 CFR 50 Appendix G 50 ft-lb requirement.
This figure is based on an estimated decrease in USE calculated in accordance with the guidance of Regulatory Guide 1.99, r
Revision 2, and adjusted conservatively for the CNS surveillance testing a
results.
A detailed discussion of the development of this estimate is t
provided in the District's February, 1993 vessel materials surveillance report,i' submitted to the NRC by-letter dated February 25, 1993.2' l
4 GE Nuclear Energy Report No. GE-NE-523-159-1292, dated February, 1993, " Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis."
5.
Letter from G. R Horn to NRC dated February 25,1993, " Submittal of Reactor Vessel Surveillance Test Results."
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c-4, Attachment-to NSD930782 1'
-Page 3 of 6 In addition, the District does have initial USE data for the CNS vessel surveillance weld, which was fabricated by the same vendor..in the same.
time frame, using the same fabrication process and material specification as the CNS beltline welds.
Using~ the initial USE energy of the surveillance weld,110 ft-lbs, would yield even greater margin over the 50 ft-lb requirement than the lower bound estimate.
While the District believes that its current prediction of USE at 32 EFPY I
is justifiable, the District has participated in a recent' effort sponsored by the BWROG to demonstrate equivalent margin using alternate analytical 5
methods. This effort resulted in the preparation of NEDO-32205, "10CFR50 '
Appendix G Equivalent Margin Analysis For Low Upper Shelf Energy In BWR/2
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Through BWR/6 Vessels,"U which was submitted to the NRC by letter dated April 30, 1993.F l
NED0-32205 provides an equivalent margin analysis showing that, for 32 Effective Full Power Years (EFPY) of operation, the USE levels for CNS's beltline welds are higher than that required to demonstrate equivalent i
margin in accordance with 10 CFR 50 Appendix G, This Topical Report contains the required analysis and comparison of BWR/2 i
through BWR/6 welds, which includes those that were fabricated by the same 1
vendor, in the same time frame, and using the same fabrication process and material specification as the CNS beltline welds. The CNS vessel (as with many BWR/4 vessels) was fabricated by CE in the 1960s, using a submerged j
are welding process.
j While this BWROG analysis was performed using assumptions which bound all BWR/2 through BWR/6 vessels, Appendix B of NEDO-32205 contains applicability verification forms for each licensee to complete to confirm t
that the plant-specific beltline materials at their plant are enveloped by the equivalent margin analysis contained in the Topical Report. Attached l
to this letter (Appendix A) are completed copies of these forms demonstrating that the beltline plates and welds at CNS are enveloped by the BWROG analysis and therefore meet the requirements for demonstrating equivalent margin as allowed by 10 CFR 50 Appendix G.
j As shown on the attached applicability verification form'(Appendix A to f
this letter), the District's conservatively estimated end-of-life decrease in USE for weld material exceeds that estimated using Regulatory Guide l
1.99, Revision 2 based on the surveillance adjustment. However, as also shown on the verification form, the material is still determined to be' acceptable per the Equivalent Margin Analysis.
i The District would like to note that while it maintains that justifiable basis exists to show that the CNS beltline weld materials will-remain i
above the 50 ft-lb limit and that with the District's participation in the l
6.
NEDO-32205, "10CFR50 Appendix G Equivalent Margin Analysis For Low Upper Shelf Energy In BWR/2 Through BWR/6 Vessels," dated April 1993.
l 7.
Letter from C. L. Tully (BWROG) to NRC dated April 30, 1993 "BWR Owners' Group Topical Report on Upper Shelf Energy Equivalent Margin i
Analysis."
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i Attachment to I
NSD930782 Page 4 of 6' recent equivalent margin effort additional basis exists to demonstrate compliance with 10 CFR 50 Appendix G,
the District continues to
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participate in other industry-sponsored -efforts to obtain further 4
information on this important issue.
The District is currently participating in an effort sponsored by the ABB-CE Reactor Vessel Group to catalog i
matrix available CE vessel material information,. as well as existing archived vessel material fabricated during the time of. CNS vessel i
construction. This effort is expected to provide additional information to further support the District's conclusion that the CNS vessel meets 10 r
CFR 50 Appendix G requirements.
In addition, the District is also participating as a host reactor in the BWROG Supplemental Surveillance Program.
This effort will provide additional surveillance data for industry use in' characterizing vessel embrittlement.
Therefore, while the District maintains that all CNS vessel beltline materials will meet the 10 CFR 50 Appendix G 50 ft-lb requirement at 32 EFPY, the District has also determined through its participation in the recent BWROG equivalent margin effort that the 10 CFR Appendix G requirements are met for CNS even assuming a 32 EFPY USE as low as 35 ft-
- lbs, lioweve r, the District will continue to maintain a proactive stance on this issue, and will continue to participate in related industry programs..
i DRST OUESTION RELATED TO ITEM 2.B "Your response indicates that data from the drop-weight test and Charpy i
test for beltline materials is either absent or incomplete for initial RT,,
i determination.
An alternative method developed by the General Electric Company (GE) was usec in deriving the initial RT,,, for these materials.
In the GE method, the establishment of the slope for the transition zone of the Charpy curve is crucial in deriving the initial RT,, from incomplete test data.
Provide all plate and weld Charpy test curves compiled by GE for establishing the 2"F per ft-lb slope for the transition zone of the 1
Charpy curve. All test data must be from materials-equivalent to ( i '. e.,
having the same vendor, fabrication time frame, fabrication process, material specification, etc., as) the beltline materials of this reactor vessel."
pl.S_IBICT RESPONSE The 2"F per f t-lb adjustment was used to develop conservative initial RT,,,
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values for the CNS beltline plate material. This adjustment was not used
.l in developing initial RT., for CNS vessel weld material because all weld.
Charpy data taken at fabrication was above '50 ft-lbs.
A detailed i
discussion of the development of these values is provided in the first CNS
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vessel material surveillance testing reportF which was submitted to the 8.
CE Report MDE-103-0986 dated May 1987, " Cooper Nuclear Station Reactor Pressure Vessel Surveillance Materials Testing and Fracture Toughness Analysis."
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Attachment to NSD930782 Page 5 of 6 l
NRC by letter dated July 6, 1987.F This methodology was developed to I
analytically convert fracture toughness data obtained under ASME code editions prior to the Summer 1972 Addendum to current ASME code requirements. These methods have been presented to the NRC in a number of.
Final Safety Analysis Reports for other licensees, reviewed on a case by case basis, and in each case approved. This methodology was also reviewed by the NRC during their evaluation of Amendment No.
120 to the CNS Operating license, and is discussed in the NRC Safety evaluation which l
accompanied that amendment.B
'l However, in recent discussions with GE, the District has learned that due to a number of plants receiving the above question, the BWROG is sponsoring an effort to have GE document the basis for its development of the 2*F per ft-lb correlation. This effort should be complete by the end of August, 1993, and is expected to resolve NRC concerns in this area.
f SECOND OUESTION RE1ATED TO ITEM 2.B "The response also indicates that the chemistry data for axial welds 2-233 l
A, B and C are not available. What values of copper, nickel and neutron fluence were used to determine the increase in transition temperature and the drop in USE for these welds? What are the justifications for using these values? What methods were used to determine the copper and nickel values?"
DISTRICT RESPONSE Since submittal of the District's initial response to Generic Letter j
92-01, Revision 1, the District has obtained chemistry data' for these j
welds of the same type, heat, and flux lot number from the Salem 1 PTS submittal, dated January 10, 1986. This information is documented in the District's February, 1993 vessel materials surveillance test report,F which reports a copper content of 0.22% and a nickel content of 1.02%.-
Prior to locating this information, the District assumed an " upper limit" copper content of 0.35% in accordance with.the guidance of-Regulatory i
Guide 1.99 Revision 1.
The. methodology used in Revision 1.of Regulatory Guide 1.99 did not require determination of nickel content.
Based on the results of flux wire analyses and fluence evaluation conducted following the removal of the initial CNS surveillance-capsule
.i and reported in the District's May 1987 surveillance report,F the peak end-of-life fluence at 1/4 T was estimated to be 1.5 X 10a.n/cm.
The 1
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District's current estimate, based on the flux wire analyses and fluence evaluation conducted following the removal of the second surveillance '
capsule and reported in the District's February 1993 surveillance report,
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is 1.1 X 10 n/cm at 1/4 T.
9.
Letter from G. A. Trevors to NRC dated July 6,1987, " Reactor. Vessel Material Surveillance Program, NRC Docket No. 50-298/DPR-46."
10.
Letter from W. O. Long (NRC) to G. A. Trevors (NPPD) dated April 26, Amendment-No.
120 to Facility.
i 1988, " Cooper Nuclear Station Operating License No. DPR-46 (TAC NO. 65793),"
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Attachment to NSD930782 Page 6 of 6 With respect to the USE, the equivalent margin analysis sponsored by the BWROC and discussed above demonstrates CNS's compliance with the requirements of 10CFR50 Appendix G, and as discussed further above, the 32' i
EFPY USE for these welds is expected to be above 50 ft-lbs.
III.
CONCLUSION The District's original response to GL 92-01 Revision 1 used data from our Surveillance Test Report issued in 1987.
Since the submittal of the District's initial Generic Letter 92-01, Revision 1 response, the District has submitted its second CNS vessel materials surveillance test report, for specimens withdrawn from the CNS reactor in 1991.
As such, some of' the additional information provided in this response has been obtained from this more recent test report.
Since the issuance of.the NRC's request for additional information, the BWR Owner's Group has issued Topical Report NEDO-32205.
This report, "10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels," was completed in anticipation' of additional questions'on this topic for utilities which do not have the necessary initial USE data to demonstrate USE requirements in accordance i
with current NRC methods. As allowed by 10 CFR 50 Appendix G, the topical report documents an alternate means of demonstrating acceptable margin -
exists to preclude brittle vessel fracture.
Accordingly, based on the District's previous submittals, and as supplemented by this discussion, it t
is the District's position that CNS is in full compliance with l
10 CFR 50.60 and 10 CFR 50 Appendices G and H.
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i APPENDIX A EQUIVALENT MARGIN ANALYSIS I
PLANT APPLICABILITY VERIFICATION FORMS FROM GE REPORT NEDO-32205, APRIL 1993 l
COMPLETED FOR COOPER NUCLEAR STATION A
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bHU)O-32205' EQUIVALENT MARGIN ANALYSIS PLANT APPLICABILITY VERIFICATION FORM BWR/3-6 PLATE i
Surveillance Plate VSE:
%Cu -
0.21%
(5-5) 17 Capsule Fluence -
2.4 x 10 (5-5) e Measured % Decrease -
6%
'(Charpy Curves)(5-9)
I R.G. 1.99 Predicted % Decrease -
12.5%
(R.G. 1.99, Figure 2)
(5-9) 1 Limitino Beltline Plate USE:
%Cu -
0.21 (7-15) 18 32 EFPY Fluence -
1.1 x 10 (7-15) t R.G. 1.99 Predicted % Decrease -
18 (R.G.1.99, Figure 2)(7-15)
Adjusted % Decrease =
NA (R.G, 1.99, Position.2.2) 18 % s 21%, so vessel plates are bounded by equivalent margin analysis Note: The reference numbers in parentheses are the page numbers from GE Report No. GE-NE-523-159-1292 dated February 1993, " Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis," transmitted to NRC by letter dated February' 25, 1993.
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NEDO-32205 EQUIVALENT MARGIN ANALYSIS PLANT APPLICABILITY VERIFICATION FORM BWR/2-6 WELD t
i Surveillance Weld VSE:
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%Cu =
0.237 (5-5) 17 Capsule Fluence -
2.4 x 10 (5-5)
Measured % Decrease -
22:
(Charpy Curves)
(5-9) j R.G. 1.99 Predicted % Decrease -
15.5%
(R.G. 1.99, figure 2)
(5-9)
Limitina Beltline Weld USE:
%Cu =
0.19%
(7-15)
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18 32 EFPY Fluence =
1.1 x 10 (7-15)
R.G. 1.99 Predicted % Decrease =
20%
(R.G. 1.99, Figure 2)
(7-20)
Adjusted % Decrease -
30.5%
(R.G. 1.99, Position 2.2)-
(7-15) i 39,3% 133%, so vessel welds are bounded-by equivalent margin analysis i
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