ML20248E709
| ML20248E709 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/01/1998 |
| From: | John Hickman NRC (Affiliation Not Assigned) |
| To: | Kingsley O COMMONWEALTH EDISON CO. |
| References | |
| GL-92-01, GL-92-1, TAC-MA0527, TAC-MA0528, TAC-MA0529, TAC-MA0530, TAC-MA527, TAC-MA528, TAC-MA529, TAC-MA530, NUDOCS 9806030421 | |
| Download: ML20248E709 (11) | |
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UNITED STATES j,
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NUCLEAR RE(3ULATORY COMMISSION L'
WASHINGTON, D.C. 20666-0001 June 1,1998 I
Mr. Oliver D. Kingsley, President Nuclear Generation Group Commonwealth Edison Company Executive Towers West til 1400 Opus Place, Suite 500 Downers Grove, IL 60515
SUBJECT:
REQUEST FOR ADDITIONAL INFORMAT!CN REGARDING REACTOR PRESSURE VESSEL INTEGRlTY AT BRAIDWOOD, UNIT NOS.1 AND 2 AND BYRON, UNIT NOS.1 AND 2 (TAC NOS. MA0527, MA0528, MA0529 AND MA0530)
Dear Mr. Kingsley; Generic Letter (GL) 92-01, Revition 1, Supplement 1 (GL), " Reactor Vessel Structural Integrity,"
was issued in May 1995. This GL requested licensees to perform a review of their reactor pressura vessel (RPV) structural integrity assessments in order to identify, collect, and report any new data pertinent to the analysis of the structuralintegrity of their RPVs and to assess the impact of those data on their RPV integrity ana;yses relative to the requirements of Section 50.60 of Title 10 of the C_gde of Federal Rec 2ations (10 CFR 50.60).10 CFR 50.61, and Appendices G and H to 10 CFR Part 50 (which encompass pressurized thermal shock and upper shelf energy (USE) evaluations), and any potential impact on low temperature overpressure limits or pressure-temperature limits.
After reviewing your response, the NRC issued you a letter dated August 29,1996. In this letter we noted that additional RPV information may become available as a result of Owners Group efforts and requested that you provide us with the results of the Owners Group's programs relative to your plant. We further indicated that a p' ant-specific TAC number may be opened to review this material _ Subsequent to issuing this letter, the NRC conducted an inspection of Framatome Technologies, Inc. (FTI) in May 1997. This inspection focused on obtaining all available RPV weld chemistry data for RPVs fabricated by Babcock & Wilcox (B&W). As a result of this inspection, additional dats were identified that may affect previous RPV integrily analyses supplied by licensees with B&W-fabricated RPVs. As a follow-up to the letter and the FTl inspection, and :n order to provide a complete response to items 2. 3, and 4 of the GL, the NRC requests that you prnvide a response to the enclosed request for additional information (RAI) within 90 days of receipt of this letter, if a question does not apply to your situation, please indicate this in your RAl response along with your technical basis and, per the GL, provide a certification that previously submitted evaluations remain valid.
The informa' tion provided will be used in updating the Reactor Vessel Integrity Database (RVID).
Also, please note that RPV integrity analyses utilizing newly identified data could result in the need for license amendments in order to maintain comp;iance with 10 CFR 50.60,10 CFR 50.61 (pressurized thermal shock), and Appendices G and H to 10 CFR Par 150, and to address any potentialimpact on low temperature overpressure limits or pressure-temperature limits. If additionallicense amendments or assessrnents are necessary, we request that you provide a schedule for such submittals as stated in the enclosure.
9006030421 980601 bb
@W PDR ADOCK 05000454 p
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O. Kingslay Should you have any questions on this issue, please contact me at (301) 415-3017.
Sincerely, John B. Hickman, Project Manager Project Directorate 111-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation Docket Nos. STN 50454, STN 50-455, STN 50-456, STN 50-457
Enclosure:
At Stated cc w/ encl: See next page
O. Kingsley June 1, 1998 Should you have any questions on this issue, please contact me at (301) 415-3017.
Sincerely, ORIG. SIGNED BY John B. Hickman, Project Manager Project Directorate lll-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation Docket Nos. STN 50-454, STN 50-455, STN 50-456, STN 50-457
Enclosure:
As Stated cc w/ encl: See next page I
DISTRIBUTION:
Docket File PUBLIC PDill-2 R/F '
EAdensam (EGA1)
SRichards JHickman CMoore OGC ACRS,T2E26 AHiser (ALH1)
ESullivan, EMCB (EJS)
MMitchell(MAM4)
GVissing (GSV)
DOCUMENT NAME: G:\\CMNTSPBYRONEYA0527.RAI T2 RECEfVE A COPY OF THis ooCUMENT INDICATE IN THE BOX: "C" = COPY WITHoUT ENCLOSURES "E" = COPY NTH ENCLOSURES "N" OFFICE PM:PD3-2 6 QPp3-2 L PM:PD3-2 6
BC:EMCB D:PD3-2 Ir-NAME JHickmanggp/ Oooh SBailey f4 ESullivan M SRichards S&
U DATE 5/ 2 6 /98 5/ M/98 5/ 7 J /98 5/ lf /98 hI
/98 OFFICIAL RECORD COPY
l l
O. Kingsley Byron /Braidwood Stations I
Commonwealth Edison Company cc:
Ms. C. Sue Hauser, Project Manager George L Edgar Westinghouse Electric Corporation Morgan, Lewis and Bochius Energy Systems Business UnM 1800 M Street, N.W.
Post Office Box 355 Washington, DC 20036 Pitts' urgh, Pennsylvania 15230 o
Attomey General Joseph Gallo 500 S. Second Street Gallo & Ross Springfield,lilinois 62701 1250 Eye St., N.W., Suite 302 Washington, DC 20005 lilinois Department of Nuclear Safety Office of Nuclear Faci:ity Safety l
Michael 1. Millsr Esquire 1035 Outer Park Drive Sidley and Austin Springfield, Illinois 62704 i
One First National Plaza l
Chicago, Illinois 60603 Commonwealth Edison Company l
Byron Staticn Manager Howard A. Leamer 4450 N. German Church Road l
Environmentallaw and Policy Byron, Illinois 61010-9794 l
Center of the Midwost 35 East Wacker Dr., Suite 1300 Commonwealth Edison Company i
Chicago, Illinois 60601 Site Vice President-Byron 4450 N. German Church Road U.S. Nuclear Regulatory Commission Byron, Illinois 61010-9794 l
Byron Resident inspectors Office 4448 N. German Church Road U.S. Nuclear Regulatory Commission Byron, Illinois 61010 9750 Braidwood Resident inspectors Office RR 1, Box 79 Regional Administrator, Region lli Braceville, Illinois 60407 U.S. Nuclear Regulatory Commission l
801 Warrenville Road Mr. Ron Stephens Lisle, Illinois 60532-4351 lilinois Emergency Services and Disastar Agency Ms. Lorraine Creek 110 E. Adams Street RR 1, Box 182 Springfield, Illinois 62706 Manteno, Illinois 60950 Chairman Chairman, Ogle County Board Will County Board of Supervisors Post Office Box 357 Will County Board Courthouse Oregon, Illinois 61061 Joliet, Illinois 60434 l
Mrs. Phillip B. Johnson Commonwealth Edison Company 1907 Stratford Lane Braidwood Station Manager Rockford, Illinois 61107 RR 1, Box 84 Braceville, Illinois 60407
i O. Kingsley
{
Commonwealth Edison Company Byron /Braidwood Stations l
Ms. Bridget Little Rorem Ms. Irene Johnson, Licensing Director Appleseed Coordinator Nuclear Regulatory Services 117 N. Linden Street Commonwealth Edison Company Essex,Illiaois 60935 Executive Towers West lli 1400 Opus Place, Suite 500 Document Control Desk-Licensing Downers Grove, IL 60515 Commonwealth Edison Company 1400 Opus Place, Suite 400 Commonwealth Edison Company I
Downers Grove, Illinois 60515 Reg. Assurance Supervisor-Braidwood 1
RR 1, Box 79 Commonwesith Edison Company Braceville, Illinois 60407 Site Vice President - Braidwood RR 1, Box 84 Commonwealth Edison Company Braceville,IL 60407 Reg. Assurance Supervisor Byron 4450 N. German Church Road Mr. David Helwig Byron, lilinois 61010-9794 Senior Vice President Commonwealth Edison Company Mr. Michaal J. Wallace Executive Towers West lll Senior Vice President 1400 Opus Place, Suite 900 Commonwealth Edison Company Downers Grove,IL 60515 Executive Towers West til 1400 Opus Place, Suite 900 Mr. Gene H. Stanley Downers Grove,IL 60515 PWR's Vice President Commonwealth Edison Company j
Execu'ive Towers West lli 1400 Opus Place, Suite 900 Downers Grove,IL 60515 N r. Steve Perry BNR's Vice President Commonwealth Edison Company Executive Towers West lll 1400 Opus Place, Suite 900 Downers Grove, IL 60515 Mr. Dennis Farrar Regulatory Services Manager i
Commonwealth Edison Company Executive Towers West lli
)
1400 Opus Place, Suite 500 l
Downers Grove,IL 60515 j
l i
REQUEST FOR ADDITIONAL INFORMATION (RAIJ REACTOR PRESSURE VESSEL INTEGRITY Section 1.0, Assessment of Best-Estimate Chemistry The staff recently received additionalinformation that may a3ect the determination of the bast-
)
estimate chemistry composition for your reactor pressure vessel (RPV) welds or your i
surveillance weld material. This data was provided by Framatome Technologies, Inc. (FTI) in J
letters from Mr. Matthew J. DeVan (FTI) to Mr. Barry J. Elliot (NRC) dated June G,1997 (INS 97-
)
2262), June 19,1997 (INS-97-2450), and July 10,1997 (INS-97-2741). In addiuon, it is the NRC staffs understanding that an evaluation of this data was provided to members of the B&W Owner's Group, Mr. R. E. Jaquin (Rochester Gas and Electric), and Mr. P. S. Askins (Tennessee Valley Authority) via letter dated June 30,1997 (INS-97-2526).
I Based on this information, in accordance with the provisions of Generic Letter 92-01, Revision 1, l
Supplement 1, the NRC requests the following l
1.
An evaluation of the information in the referer ce above ar,d an assessment of its 1i, applicability to the determination of the best-estimate chemistry for all of your RPV bonline welds. Based upon this reevaluation, supply the information necessary to completely fill out the data requested in Tab;e 1 for each RPV beltline weld material. Also provide a discussion for the copper and nickel valuos chosen for each we:d wire heat noting what heat-specific data were included and excluded from the analysis and the analysis method l
l chosen for determining the best estimate. If the limiting material for your vessel's l
pressurized thermal shock / preen-temperature (PTS /PT) limits evaluation is not a weld,
)
include the information requested in Table 1 for the limiting material also. Furthermore, you should consider the information provided in Section 2.0 of this RAI on the use of survelHance data when responding.
)
With respect to your response to this question, the staff no'.es that some issue.s regarding the evaluation of the data were discussed in a public meeting between the staff, NEl, [ Nuclear Energy institute) and industry representatives on November 12,1997. A summary of this j
meeting is documented in a meeting summary dated November 19,1997, " Meeting Summary for 1
l November 12,1997, Meeting with Owners Group Representatives and NEl ReGarding Review of j
Responses to Generic Letter 92-01, Revision 1, Supplement 1 Responses" (Reference 1). The j
'information in Reference 1 may be useful in helping you to prepare your response.
In addition to the issues discussed in the referenced meeting, you should also consider what l3-method should be used for grouping sets of chemistry data (in particular, those from weld j
qualification tests) as being from "one weld" or from multiple welds. This is an important consideration when a mean-of-the-means or coil-weighted average approach is determined to be i
the appropriate method for determining the best estimate chemistry. If a weld (or welds) were fabricated as weld qualification specimens by the same manufacturer, within a short time span, using similar welding input parameters, and using the same coil (or coils in the case of tandem are welds) of weld consumables, it may be appropriate to consider all chemistry camples from that weld (or welds) as samples from "one weld" for the purposes of best-estimate chemistry determination, if information is not available to confirm the aforementioned details, but sufficient evidence exists to reasonably assume the details are the same, the best-estimate chemistry ENCLOSURE
i should be evaluated both by assuming the data came from "one weld" and by assuming that the i
data came from an appropriate number of " multiple welds." A justification should then be provided for which assumption tvas chosen when tin test-erjimate chemistry was determined.
Section 2.0: Evaluation and Use of Surveillance Data The chemical composition report referenced in Section 1.0 includes updated chemistry estimatss for heats of weld metal. These reports provide infonnation regarding a best-estimate value and i
c the source of the data used in estimating the chemical composition of the heat of material. This permits the determination of the best-estimate chernical composition for the various sources of c#a including surveillance welds. Sinc 3 the evaluation of surveillance data rely on both the best-estimate chemical composition of the RPV weld and the surveillance weld, the information in these reports may msuit in the need to revise previous evelustions of RPV integrity (including low temperature overprenure (LTOP) setpoints and PT limits) per the requirements of 10 C"R
)
50.60,10 CFR 5041, and Appendices G and H to 10 CFR Part 50.
Based on this information and consistent with the provisicas of Generic Letter 92-01, Revision 1, Supplement 1, the NRC requests the following:
i 2.
That (1) the infonnation listed in Table 2, Table 3, and the chemistry factor from the surveillance data be provided for each heat of ma4erial for which surveillance weld data are available and a revision in the RPV integrity analyses (i.e., current licensing basis) is needed or (2) a certification that previously submitted eva!uations remain valid. Seperate tables should be used for each heat of material addressed. If the limiting material for your vessel's PTS /PT limits evaluation is not a weld, include the information requested in j
the tab'es for the imiting materist (if surveillance dafe are available for this material).
The information discussed in Section 1.0 of this RAI rogarding the chemistry repor's should be considered in this response a!ong with the following questions and comments.
All surveillance program results for the heats of materia! in an RPV should be considered in evaluating its integrity regardiess of source per 10 CFR 50.61 (" Surveillance program results means any data that demonstrates the embrittlement uends for the limiting beltline material, including but not limited to data from test reactors or from surveillance programs at other plants with or without surveillance program integrated per 10 CFR Part 50, Appendix it"). If any of the data provided in Table 2 are rot used in the calculation of the embrittlement trend for a particular RPV weld, the technical basis for not including /using the data snould be provided.
When assessing credibility of surveillance data that come from more than one source, adjustments to the surveillance data may be needed to account for differences in the chemical composition and irradiation environment of the different sources consistent with the requirements in 10 CFR 50.61. A method for accounting for these differences !s discussed in Reference 1.
Based on the information p ovided in Table 2, the credibility of the surveillance data can be evaluated. The results of these analyses including the slope of the best-fit line through the surveillance data can be provided in a format similar to that of Table 3. If the method for j
adjusting and/or normalizing the surveillance data when assessing credibility differs from the j
methods documanted in Reference 1, providu the technicsi basis for the adjustment and/or the 1
4 1
l 3
I normalization procedure. If the chemical composition of the surveillance weld is not detoimined 1
in accordance with Reference 1 (i.e., the mean of all chemistry analyses performed on the surveillance weld), provide the technical basis for the estimate.
l When determining the chemistry factor for an RPV weld from surveillance data, adjustments to i
the surveillance data may be needed to account for differences in the chemical composition and i
irradiation environment between the surveillance specimens and the vessel being assessed i
consistent with the requirements in 10 CFR 50.61. A method to account for these differences is provided in Reference 1.
1 in addition,10 CFR 50.61(c)(2) specifies that licensees shall consider plant-specific information (e.g., operating temperature and surveillance data) to verify that the RTer for each vessel beltline material is a bounding value. Regulatory Guide (RG) 1.99, Revision 2, describes two methods for determining the amount of margin and the chemistry facter used in determining i
RTmy. Position 1.1 describes the use of the Generic Tables in the regulatory guide. Position 2.1 I
describes the use of credible surveillance data. If the surveillance data are credible, the o, may be reduced in half to calculate the margin term and the chemistry factor is to be determined from the best-fit line of the surveillance data, if the evaluation of the surveillance data indicate that the surveillance data set is not credible and the measured values of ARTm7 are less than the l
projected mean from the tables plus the generic 20, the chemistry factor may be calculated 3
using either Fosition 1.1 or Position 2.1; however, the full margin term must be applied. The method chosen must bound all the surveillance data to be in compliance with 10 CFR I
50.61(c)(2).
Based on the information provided in Table 2 along with the best-estimate chemical composition
)
of the heat of material and the irradiation temperature of the plant whose vesselis being assessed, the chemistry factor of the RPV weld can be determined. Note that the adjusted ARTm7 or a particular surveillance data point may be one value when determining credibility and l
f another value when determining the chemistry factor as a result of the different normalization I
procedures. If the method fcr adjusting and/or normalizing the surveillance data when determining the chemistry factor differs from the methods documented in Reference 1, provide the technical basis for the adjustment and/or the normalization procedes.
in a meeting between the staff and industry representatives at the NRC on February 12,1998, an industry representative requested a clarification as to when the ratio procedure should be used to j
evaluate surveillance data. The ratio procedure is described in the PTS rule and RG 1.99, Revision 2. The ratio procedure is used to adjust the measured value of ARTer to account.for differences in the chemical composition between the surveillance weld and the vessel beltline weld. The PTS rule and RG 1.99, Revision 2, indicate that when there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, i.e., differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the ratio procedure must be used.
Section 3 0, PTS /PT Limit Evaluation 3.
If the limiting material for your plant changes or if the adjusted reference temperature for the limiting materialincreases as a result of the above evaluations, provide the revised RTers value for the limiting materialin accordance with 10 CFR 50.61. In addition, if ths adjusted RTer value increased, provide a schedule for revising the PT and LTOP limits.
4 The schedule should ensure that compliance with 10 CFR Part 50, Appendix G is maintained.
Reference 1.
Memorandum from Keith R. Wichman to Edmund J. Sullivan, " Meeting Summary for November 12,1997, Meeting with Owners Group Representatives and NEl Regarding Review of Responses to Generic Letter 92-01, Revision 1, Supplement 1 Responses,"
dated November 19,1997.
Attachments: 1. Table 1
- 2. Tables 2,3 l
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Table 2: Heat xxxx Capsule ID Cu Ni irredebon Fluence Measured Date Usedin (including Temperature (x10n/ctn')
ART.
^--
'-g Vessel enurce)
(*F)
(*F)
(YcrN)
Table 3: Heat xxxx Capsule ID Cu Ni irredehon Fluence Measured A4usted Prodseted (Aquoted -
(including Temperefure Factor ART.
ART.
ART.
Predicted) ART.
source)
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _