ML20235J975
| ML20235J975 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 09/18/1987 |
| From: | Delgeorge L COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8710020185 | |
| Download: ML20235J975 (92) | |
Text
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f Commonwealth Edison 3
O k'v} One First Chicago, Nationit Illinois 60690 0767 Plazm, Chicago, lilanois Address R: ply to: Post Office Box 767 i September 18, 1987 U.S. Nuclear Regulatory Commission i
Attn: Document Control Desk l Washington, DC 20555
Subject:
Zion Nuclear Power Station Units 1 and 2 Pressurized Thermal Shock NRC Docket Nos. 50-295 and 50-304 References (a): January 17, 1986 letter from G.L. i Alexander to H.R. Denton (b): August 14, 1986 letter from S.A. Varga to D.L. Farrar (c): December 29, 1986 letter from P.C.
LeBlond to H.R. Denton (d): May 7, 1987 letter from D.R. Muller to D.L. Farrar i
Gentlemen:
The purpose of this letter is twofold. First, Commonwealth Edison company is expressing concern regarding the NRC Staff's evaluation of the submittals made for Zion Station in support of 10 CFR 50.61. Secondly, Conunonwealth Edison company is requesting that an independent technical review be performed of the information contained in the referenced letters and in tH s submittal.
Commonwealth Edison Company and the }/RC Staff have been involved in a lengthy exchange regarding Zion Unit l's compliance with the Pressurized Thermal Shock (PTS) rule, 10 CFR 50.61. A short summary of this discussion is included as Attachment A. j i
The value of the copper content for the limiting Unit I reactor )
vessel weld is the matter in controversy. Commonwealth Edison believes that )
l it has accurately determined the copper content of the limiting vessel weld to j be 0.32 weight percent. The NRC Staff disagrees, believing that the correct value is 0.35% and thus has requested in references (b) and (d) that a flux reduction program be prepared in accordance with 10 CFR 50.61.b.3.
I 0Y 8710020185 870918 no !
t US NRC September 18, 1987 The implementation of a flux reduction program, while costly, may eventually be consistent with Commonwealth Edison Company's overall strategy for the maintenance of our generating capability. However, as discussed in reference (c), Commonwealth Edison's final decision regarding the life extension of Zion Station is still pending. Thus, any steps, such as flux reduction programs, taken in support of life extension are premature.
Should the decision be made to extend the life of Zion Station, then the NRC Staff's proposed value for copper content of 0.35% would represent an unnecessary reduction in the calculated capability of the Zion Unit I reactor vessel to withstand neutron radiation damage. This could eventually result in the premature retirement of Zion Unit 1.
Commonwealth Edison Company has considerable concern regarding the NRC Staff's evaluation contained in reference (d) which forms the basis for their rejection of our proposed value of 0.32 weight percent copper content.
These concerns focus on two issues. These issues are:
- 1. The statistical techniques to be used in reducing the total chemistry data set of 99 measurements down to a single value for copper content.
- 2. The validity of 31 chemistry measurements obtained through a variety of techniques and laboratories.
Attachment B demonstrates that the statistical technique utilized by the NRC Staff in reference (d) is not an appropriate treatment of the data. 7.
rigorous treatment of the available data was presented in Section I11.2 of WCAP 11350. Commonwealth Edison Company contends that this methodology is the proper technique to be utilized.
Attachment C refutes reference (d)'s arbitrary exclusion of 31 valid measurements. Although these 31 measurements are located in the lower half of the data's distribution, the analysis shows that there is no valid justification for excluding these measurements.
Attachment D provides a summary of the values for copper content that would result for nine possible combinations of statistical technique and data inclusion / exclusion. It demonstrates that the use of both an improper statis-tical technique and the exclusion of valid data is necessary to support a value different from 0.32 weight percent copper content.
In addition, major segments of Commonwealth Edison's technical justi-fication have not been addressed in reference (d). These segments include evaluations presented at the October 3, 1986 meeting and in Section III.1 of WCAP 11350 (Attachment G) which demonstrate that the NRC Staff's exclus. ion of valid data while continuing to endorse the highest measurements is not we?.1 founded in basic stat'Istics. For example, given the NRC Staff's assertion regarding the existence of distinct populations within the data, the highest data endorsed by the NRC Staff has less than a 1 in 100,000 chance of belonging to the other NRC postulated populations.
US NRC September 18, 1987 Finally, reference (d) includes a number of statements containing factual errors that have produced additional concern. Attachment E provides a point-by-point review of reference (d). While some of these instances do not directly relate to the central issues discussed in Attachments B, C, and D, collectively they have produced an impression of incompleteness in the review of this very important issue.
As a result, Commonwealth Edison Company is requesting that an independent review of the information contained in the references and this submittal be performed by personnel not associated with the earlier reviews.
This review should include consideration of the statistical treatment of the I data. Attachments A through J are provided to facilitate this review. j specifically, Attachment J contains a suggested review outline to integrate the conclusions of the individual Attachments.
As discussed above, commonwealth Edison Company considers this to be a serious issue. The broader issue of plant life extension is being properly addressed as described in reference (c). 10 CFR 50.61 is not a management tim 1 to be utilized to control the life extension process. Rather, it is intended to provide assurance that proper protection exists for postulated Pressurized Thermal Shock events. Commonwealth Edison has provided that assurance in references (a) and (c), at the October 3, 1986, meeting, and in this submittal.
Commonwealth Edison believes that a meeting with the NRC Staff to discuss this issue would be very beneficial. Thus, Commonwealth Edison Company requests that such a meeting be held with the review participants at your convenience.
One signed original and fifty copies are provided for your review.
Please contact Commonwealth Edison Company's Nuclear Licensing Department with any further questions regarding this matter or to arrange the requested meeting.
Very truly yours, o
L. O. DelGeorge Assistant Vice President Attachments cc: Resident inspector - Zion j J. A. Norris - NRR 3419K l
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[ Commonwealth' Edison q ' One Fir:t Nabonal Plaza. Chicago, Illinois
('
\ is n /'f ' Address Repty to: Poet Omco Box 767
\j - Chicago, Illinois 60690 0767 :
September. 18, 1987 U.S. Nuclear Regulatory Commission .
Attn: Document Control Desk l .Wa2hington, DC 20555
Subject:
Zion Nuclear Power Station Units 1 and.2 pressurized Thermal Shock NRC Docket Nos.'50-295 and 50-304 References (a): January, 17, 1986 letter from G.L.
Alexander to H.R. Denton q l (b): August 14, 1986 letter.from S.A. Varga l- to D.L. Farrar (c): December 29, 1986 letter from P.C.
LeBlond to H.R. Denton (d): May 7, 1987 letter from D.R. Muller to D.L. Farrar Grntlemen:
The purpose of this letter is twofold. First, Commonwealth Edison-Company is expressing concern regarding the NRC Staff's evaluation of the submittals made for Zion Station in support of 10 CFR 50.61. Secondly, commonwealth Edison Company is requesting that an independent technical review bn performed of the information contained in the referenc,ed letters and in this submittal.
Consnonwealth Edison Company and the NRC Staff have been it.volved in a lengthy exchange regarding Zion Unit l's compliance with the Pressurized Tharmal Shock (PTS) rule, 10 CFR 50.61. A short summary of this discussion is included as Attachment A.
The value of the copper content for the limiting Unit I reactor vaarel weld is the matter in controversy. Commonwealth Edison believes that it has accurately determined the copper content of the limiting vessel weld to be 0.32 weight percent. The NRC Staff disagrees, believing that the correct value is 0.35% and thus has requested in references (b) and (d) that a flux rsduction program be prepared in accordance with 10 CFR 50.61.b.3.
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US NRC- September 18, 1987 The implementation of a flux reduction program, while costly, may svantually be consistent with Commonwealth Edison Company's overall strategy for the maintenance of our generating capability. 1However, as discussed in.
reference (c), commonwealth Edison's final decision regarding the. life
. extension of Zion Station is still'pending.' Thus, any steps, such,as flux reduction programs, taken in support'of life extension are premature.
Should the decision be made to extend the life of Zion Station,.then' tha NRC Staff's proposed value for copper content of 0.35% would represent an unnecessary reduction in the calculated capability of the Zion Unit-1 reactor vasesi to withstand neutron radiation damage. This could' eventually result in-tha premature retirement of Zion Unit 1.
Commonwealth Edison Company has considerable concern regarding the NRC Staff's evaluation contained in reference (d).which forms the basis for th3ir rejection of our proposed value of 0.32 weight percent copper content.
Thssa concerns focus on two issues. These issues are:
- 1. The statistical techniques to be used in reducing the total chemistry
- data set of 99 measurements down to a single value for copper content.
- 2. The validity of 31 chemistry measurements obtained through a variety-of techniques and laboratories.
Attachment B demonstrates that the statistical technique utilized by tha WRC Staff in reference (d) is not an appropriate treatment of the data. A rigorous treatment of the available data was presented in Section III.2 of WCAP 11350. Commonwealth Edison Company contends that this methodology is the propar technique to be utilized.
l Attachment C refutes reference (d)'s arbitrary exclusion of 31 valid I measurements. Although these 31 measurements are located in the lower half of I tha dsta's distribution, the analysis shows that there is no valid justification for excluding these measurements.
l Attachment D provides a summary of the values for. copper content that would result for nine possible combinations of statistical technique and data-inclusion / exclusion. It demonstrates that the use of both an improper statis-tical technique and the exclusion of valid data is necessary to support a value different from 0.32 weight percent copper content.
In additiori, major segments of Commonwealth Edison's technical justi-fieraion have not been addressed in reference (d). These 'egments include 4 valuations presented at the October 3, 1986 meeting and Section III.1 of WCAP 11350 (Attachment G) which demonstrate that the NRC bcaff's exclusion of valid data while continuing to endorse the highest measurements is not well i foundid in basic statistics. For example, given the NRC Staff's assertfon-regarding the existence of distinct populations within the data, the highest L dato andorsed by the NRC Staff has less.than a 1 in 100,000 chance of belonging to th3 other NRC-postulated populations. ,
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US NRC
- September 18, 1987 Finally, reference (d) includes a number of statements containing factual errors that have produced additional concern. Attachment E provides a point-by-point review of reference (d). While some of these instances do not directly relate to the central issues discussed in Attachments B, C, and D, collectively they have produced an impression of incompleteness in the review of this very important issue.
As a result, Commonwealth Edison Company is requesting that an independent review of the information contained in the references and this submittal be performed by personnel not associated with the earlier reviews.
This review should include consideration of the statistical treatment of the data. Attachments A through J are provided to facilitate this review.
Specifically, Attachment J contains a suggested review outline to integrate the conclusions of the individual Attachments.
As discussed above, Commonwealth Edison Company considers this to be a serious issue. The broader issue of plant life extension is being properly addressed as described in reference (c). 10 CPR 50.61 is not a management tool to be utilized to control the life extension process. Rather, it is intended to provide assurance that proper protection exists for postulated Pressurized Thermal Shock events. Commonwealth Edison has provided that assurance in references (a) and (c), at the October 3, 1986, meeting, and in this submittal.
Commonwealth Edison believes that a meeting with the NRC Staff to discuss this issue would be very beneficial. Thus, Commonwealth Edison Company requests that such a meeting be held with the review participants at your convenience.
One signed original and fifty copies are provided for your review.
Please contact Commonwealth Edison Company's Nuclear Licensing Department with any further questions regarding this matter or to arrange the requested meeting.
Very truly yours,
. c(?
L. O. De1 George Assistant Vice President Attachments cc: Resident Inspector - Zion J. A. Norris - NRR 3419K
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ATTACHMENTS DESCRIPTION A Summary of previous correspondence )
i B Weld chemistry Data reduction techniques- I C Validity of data.
D Summary of' potential copper content values.
E Detailed ~ review of the NRC Staff's SER (Reference (d))
F Reference (d)
G Reference (c) (Including WCAP 11350) ,
H Pages 6-5 and 6-6 of BAW - 1799,-
July, 1983. ,
I Pages 33, 34 and 35 from SWRI Report of Unit 1, Capsule X.(SWRI-7484-001) ,
i J Review outline and summary ]
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i ATTACHMENT A l
Reference (a) provided Commonwealth Edison Company's response to 10 CFR 50.61.b.1 for the Byron, Braidwood, and Zion Stations. Included with this submittal was WCAP 10962, " Zion Units 1 and 2 Reactor Vessel Fluence and RT pts Evaluations".
Section III.3 discussed the calculation of the values for copper and nickel contents that were utilized.Section III.3.1.1 specifically identified thirty additional data points that were obtained via the i Westinghouse Owner's Group (WOG) data base. These 30 measurements were in addition to the 57 measurements considered in BAW - 1799.
WCAP 10962 utilized a copper content of 0.32 weight percent and a nickel content of 0.56 weight percent for welds made with heat 72105 wire.
These values came from a simple average of all 87 values. This work resulted in a limiting value for the Unit 1 RTpts of 297 F after 32 EFPY, which is below the screening criteria of 300'F.
Reference (b) transmitted the NRC Staff's review of reference (a).
It stated that the controlling material had been correctly identified, but disagreed with the copper and nickel contents utilized in WCAP 10962.
A meeting was held on October 3, 1986, in Bethesda, Maryland, to discuss Commonwealth Edison's treatment of the limiting Unit I reactor vessel weld. This di;;"ssion centered on the value of copper content that l was utilized in WCAP 10962, since RT pts is more sensitive to copper I l content variations than to variations in nickel content.
The material provided at that meeting provided a complete technical justification for the material properties utilized. Nevertheless, at the conclusion of that meeting, NRC staff members indicated their conti1uing disagreement with Commonwealth Edison Company's technical position.
However, there was insufficient information provided by the NRC Staff personnel for Commonwealth Edison to adequately identify the NRC Staff's j specific areas of concern. l In an attempt to clarify the NRC Staff's views, WCAP 11350 was ,
prepared to document the information presented at the october 3, 1986, '
meeting. Reference (c) transmitted this information. WCAP 11350 also provided an analysis of the appropriate statistical treatment for the chemistry data base in use. That analysis, presented in Section III.2, resulted in an estimate of 0.311 weight percent copper content.
Commonwealth Edison Company did not rely on the lower 0.311% value to reevaluate the RTpts, but rather utilized it to provide additional assurance that Commonwealth Edison's stated value of 0.32% was conservative.
Finally, reference (d) provided Commonwealth Edison with the NRC Staff's evaluation of reference (c). The NRC Staff continued to find the material properties unacceptable and proposed the use of higher values for both copper and nickel.
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RTTACHMENT B There are currently 99 measurements of the copper content of weldments made with wire from heat no. 72105. This includes the 12 new measurement of Midland test blocks discussed in reference (d). These 99 measurements were obtained with differing techniques at different laboratories, and at different times. The only commonality is that all of the weldments were made with wire from heat No. 72105. The problem of reducing these 99 separate measurements into a single value that is representative of 72105 welds can be solved by a straight forward application of classical statistics.
l Three different reduction techniques have been utilized in references (a) through (d). These three techniques are as follows:
- 1. All of the data considered to be valid is simply averaged. Each measurement is equally weighted. This technique was utilized in BAW-1799 and reference (a).
- 2. The data is grouped by weldment. That is, all measurements taken from a particular weldment are averaged together. Then, the results for each of the individual weldments are averaged together. This effectively determines a copper value for each weldment, even though all weldments were made with the same wire.
This technique was utilized in reference (d).
- 3. The data is grouped according to the variables that could affect the measurement results. These variables are measurement technique, laboratory, and time of measurement. The data points in each group are averaged to determine a value for that set of i
variables. The individual values are then averaged together. This technique was utilized in reference (c).
As stated above, a single value for the copper content of the limiting Zion Unit I reactor vessel weld is needed. This value is obtained I by examining measurements made on any weldment that was produced with wire !
72105. The entire analytical approach of making copper measurements on independent weldments to assess the copper content of a reactor vessel weldment made with the same wire is based on the tacit assumption that all weldments made with the same wire will have essentially the same copper i content. The average copper value obtained should be independent of the particular weldments on which the analyses were performed.
I However, every measurement result has a bias associated with the method of measurement. Thus, the statistical data reduction is intended to average these biases, thereby obtaining a more accurate (. determination of the "true" value. These conventional statistical concepts were outlined in Section III.2 of WCAP 11350.
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1 The variables to be reconciled are analytical techniques, laboratories, and time of measurement since these variables can affect the '!
I measurement result. Given the basic assumption that all weldments have the same copper content, it is irrelevant whether ten measurements were made on one weldment or one measurement was made on each of ten weldments. An f attempt to sort by weldments when all weldments were made with the same weld wire is inappropriate and will yield erroneous results. Sorting by weldments would allow the biases of a single technique to unduely affect the I results. Sorting by techniques, laboratories and time of measurement is the ,
appropriate methodology to minimize the error of the final result. )
i For example, technique #2 was utilized in reference (d). The ESA technique was incorporated in nine of the eleven groups. This effectively assigns a weighting factor of 82% to the biases associated with the ESA technique. Technique #1 suffers from the same shortcoming. The fact that there were more ESA measurements performed is not justification for more heavily weighting those results.
Thus, techniques #1 and #2 above do not properly account for the I measurement biases. Technique #3 is the correct methodology and is shown in Attachment D as Table 2.
In summary, the proper statistical treatment for these data sets involves assuming that the copper content of every weldment made with the same filler wire has essentially the same overall copper content. Thus, the stlstistical grouping of the data should be based upon the analytical techniques, laboratories, and time of measurement. If the sorting is pr.rformed by weldment or if a simple average is performed, then the biases I of: the most frequently utilized laboratory or technique would unduely influence the results. Thus, technique #3 is the only appropriate methodology.
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ATTACHMENT C !
I VALIDITY OF DATA I
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The analysis contained in reference (d) eliminated 31 of the available 99 chemistry measurements. These 31 points were eliminated for one of three reasons outlined in reference (d).
- 1. The measurement utilized the XRF technique on irradiated material.
(Groups 2A and 3A on attached Table 1)
- 2. The measurement was an early weld metal qualification by B+W.
(Groups 4A, 4D, 4G)
- 3. The measurements did not consist of " credible multiple (four or more) measurements". (Groups 2B, 3B, 5A, SB, SC) t Each of these justifications will be discussed individually below.
l l Reason #1 Reference (d) discusses the validity of the 18 measurements of broken charpy samples utilizing the XRF technique (Grcups 2A and 2B). These discussions are found in reference (d) on Page 5-second paragraph, Page 5-third paragraph, and page 5-fourth paragraph. Commonwealth Ediron Company's comments regarding these paragraphs are as follows:
Page 5 - Second Paragraph. I The statement in reference (d) is made that "The author of the SWRI reports ... questioned their validity ..." l The author of the Zion I, Capsule X surveillance report by SWRI did not question the validity of the X-ray fluorescence (XRF) data. He merely j speculated on the reasons why the XRF data were different from those l obtained on the same samples by Inductively coupled Plasma (ICP) techniques.
Attachment I contains the pertinent pages from SWRI-7484-001/1. On page 35 of this report, Table XIII lists the copper values obtained by the two techniques on the weld metal in question.
Specimen Technique % Copper W-25 ICP .216 W-25 XRF .27 W-28 ICp .218 W-28 XRF .25 Note that there were two direct comparisons between the XRF and ICP techniques on specimens W-25 and W-23. In both comparisons, the XRF technique yielded a higher result.
It is not logical to reject the XRF data as biased low when a direct comparison with a different technique on the same specimen shows that XRF yields higher results. The subject of direct comparisons with other techniques was extensively discursed in Section IV of WCAP 11350. The NRC's Staff's review has not addressed that discussion.
Commonwealth Edison Company is not suggesting that it is improper to compare results obtained from different specimens. On the contrary, since a basic assumption is that all weldments made with the same wire have essentially the same copper content, it is logical to compare different samples. However, Commonwealth Edison does intend to highlight the incomplete comparison of results contained in reference (d). This facet of reference (d) is also evident in the second and third paragraphs of Page 5.
Page 5 - Third Paragraph.
Reference (d) further questions the accuracy of the XRF technique by observing that the nickel contents obtained by the XRF technique in the SWRI report referenced above are different from results obtained by Emission Spectrographic Analysis (ESA) on different specimens.
f However, if one exacines the data given in Table XIII on page 35 of the SWRI report, one finds the following nickel contents for two techniques used on the same specimen.
Specimen Technique % Nickel W-25 ICP .53 W-25 XRF .57 W-28 ICP .545 W-28 XFR .49 Note that the average nickel content obtained by the XRF method was
.53% while that obtained on the same specimen by the ICP method was .537%.
Moreover, the ICP technique measured nickel contents of 0.53 to 0.545%, which is also outside the range of 0.57 to 0.62% measured by the ESA technique as cited in the Staff's review. Merely because one set of measurements fall outside the range of another is insufficient reason to cast doubt on either.
As further confirmation that the XRF technique yields accurate results on irradiated specimens, the reader is referred to Table XIII of the Zion 2 Surveillance capsule T report by SWRI cited in the subject letter and reproduced as Table IV-1 in WCAP 11350. This table compares XRF nickel determinations with those done by ATA on the same specimens. The results are in excellent agreement.
1
Thus, contrary to what is implied in reference (d), direct.
comparisons of the XRF technique with two independent measurements of nickel show that the XRF technique gives accurate results regardless of the state of irradiation.
Page 5 - Fourth Paragraph.
The statement is made that "For the Zion 1, Capsule.X material, the copper measurement made by Westinghouse using XRF on unitradiated material was 0.35% compared to an average of 0.26% by SWRI and 0.22% by Westinghouse, using the XRF method on irradiated Charpy bars. The SVRI measurements on the Zion 2 Capsule T material also seemed low ..."
There is a factual error in these statements. .The 0.22%
measurement attributed to the XRF method was in fact obtained by the ICP technique as documented in the SWRI Capsule X report on page 33 (Attachment I). There is no technical reason why irradiation would bias ICP results. Further, if one examines all the data in the two SWRI reports cited for Zion Units 1 and 2, the following is found:
Zion Independent Technique-AVG XRF-AVG XRF-single meas.
Unit (ICP or ATA) (Irrad) (Unirrad)
No.1 .22%Cu .26%Cu .35%Cu No.2 .28%Cu .234Cu .28%Cu Avg. .26%Cu .2454Cu .315%Cu Again, one should note that the XRF (Irrad) and the independent ;
techniques were both used on the same specimens and the average results are in excellent agreement. There is no technical reason why either the ICP or the ATA methods should be biased because of irradiation effects.
The XRF (Unitrad) analyses were 9? tformed on totally different specimens at different times and in differ-ent laboratories. Thus, it is not surprising that the agreement with the other techniques is not as good. This should not imply that one result is any more credible than the other.
These measurements are largely located in the lower half of the data's overall distribution. However, that is r.ot sufficient reason to question the data's credibility. Likewise, it would not be appropriate to exclude Group 4F simply because it appears to be too high.
In summary, the XRF data represents valid measurements of welds made with filler wire 72105.
Reason #2 These three data points have been excluded because a subsequent test of the same samples yielded higher results. Reference (d) discusses this issue on page 4 -First paragraph.
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1 Page 4 - First Paragraph.
The statement is made in reference to weld qualification welds that
" Singly measurements made on them in 1969 are not shown in Table I ...
because retests made by B&W in 1983 of these and other wire / flux combinations ;
... and in their judgement the modern measurements were more credible (see B&W 1799, p.6-5)."
As previously discussed in Section IV of WCAP 11350, the actual words on p.6-5 of the B&W report are reproduced below and in Attachment H.
"Without attempting to pass judgement on the original methods, it can be stated that the current methods provide more conservative values .."
Thus the report explicitly states that no attempts were made to pass judgement much less to assign credibility to one set of measurements over another. More conservative does not mean more credible.
BAW 1799 did not render a judgement regarding the credibility of these B&W ATA weld qualifications measurements on pages 6-5 and 6-6. As a result, they have been included in the groupings in Table 3 of Attachment D.
This justification carries with it the presumption that the subsequent measurements (4B, 4E, 4H) are more accurate than the earlier results (4A, 4D, 4G). A close examination of these six data points yields logical inconsistencies.
Note that the 4G value of 0.21 weight percent copper was discarded j
in favor of the 0.30% value of 4H. However, the 4A value of 0.30% has been discarded also in favor of the 0.40% 4B value. Thus, 0.30% is an acceptable result if it was generated from the B & W retests, but it is not if the value came from the earlier tests.
This example demonstrates that Reason #2 is an illogical basis for excluding data. Merely because one measurement results in a higher value than the other is insufficient reason to exclude either.
Reason #3 This justification was utilized to delete 10 data points. The logic behind these deletions was incompletely explained. Reference (d) provides a very short discussion on this issue on Page 7. There is no other justification provided for the deletion of these 10 measurements. Page 7 stated;
" Alternatively if only the six weldments for which credible multiple (four or more) measurements are considered,..."
i l
Evidently, the 10 measurements contained in Groups 2E, 3B, 5A, 5B, l end SC were deleted because they individually did not consist of four or '
more measurements, or because they were not considered to be credible.
These two possibilities are discussed below.
4.4 Measurements If this was the criteria utilized, then reference (d) is internally inconsistent. Table 1 of reference (d) identifies 5 groups consisting of a sole measurement. Thus, it would be inconsistent to delete 2B, 3B, 5A, and 58 while endorsing 1A, 1B, 4B, 4E, and 4H.
l In addition, reference (d) sorts the data according to weldment.
Thus, a single group would be allowed to contain measurements produced by different techniques at different times. (The impropriety of this method is discussed in Attachment B.) Note that Groups 5A, 5B, and 5C are measurements of the same weldments that produced the three groups that l comprise the 4C values. (See Table 2 in this attachment). Thus, to be l
consistent these six groups should have been combined into three larger j groups. Table 1 of reference (d) clearly shows that this was not performed, l l but rather these measurements were discarded. I Thus, the " 4'4 measurement" test is not a logical basis for excluding these 10 data points.
l l
Credible Data The 10 measurements contained in Groups 2B, 3B, 5A, 5B, and 5C were produced through the use of the ICP, ATA, and ESA techniques. There has been no discussion in any of the references regarding the validity of these three techniques. The NRC Staff has expressed concern regarding the use of I the XRF technique. However, the XRF technique was not utilized on these 10 l measurements.
In summary, no sound justification for the exclusion of thene 10 i measurements has been provided. Their exclusion appears to have been crbitrary.
Summary of Attachment C None of the three justifications provided in reference (d), and discussed above, contain a reasonable basis for excluding data from consideration. In addition, a close examination of how these justifications miy have been applied to the data base result in the identification of logical inconsistencies. Thus, the exclusion of these 31 data points, c1most one-third of the available data, is not justifiable for the reasons cited by the NRC Staff in reference (d).
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. l ATTACHMENT D Summary of potential Copper Content Values- .]
.i This attachment examines the sensitivity of the calculated value for copper content to the issues previously discussed in Attachments B and C. It demonstrates that the.NRC Staff is endorsing both an improper statistical technique and the exclusion of valid data to support the Staff's proposed copper content of 0.35 weight percent, j Attachment B discussed the three statistical techniques that have been utilized in the four references. It identified two of those' techniques 1 as being statistically inappropriate for use in this application. Technique l
- 3, sorting by measurement variables, is the mostivalid methodology.
l l Attachment C discussed the three justifications provided by the NRC Staff for the exclusion of 31 out of 99 available measurements. None of those three justifications have adequate technical. basis.
Tables I and 2 of this Attachment illustrate the difference in the grouping that would result from sorting by weldment or sorting by technique, laboratory, and time. Each group's average material properties is included in Tablec 1 and 2. The third technique, simple averaging of- all data, is obvious j and is not illustrated. 1 Attachment C discussed the three justifications provided in reference
]
(d) for th's exclusion of 31 sets of chemistry measurements.
These three reasons were;
- 1. Measurement was XRF of irradiated Charpy Specimens.
- 3. The measurements were not " credible multiple".
Reason #2 was shown to be based upon a misreading of BAW-1799 (Attachment H). Reason #3 was shown to be either not technically justifiable or inconsistently applied. Therefore, the exclusions based on reasons #2 and
- 3 are not justifiable. This renders the statistical analysis contained in reference (d) invalid. Table 3 illustrates the data excluded by-the NRC Staff ,
in reference (d). i Reason #1 was shown to be based upon faulty comparisons and a ,
misreading of a SWRI report, (Attachment.I). Reason #1 was evidently {
responsible for the exclusion of 18 data points. (groups 2A and 3A). 1 i
Table 4 contains the calculated copper content that would result from )
the application of each of three statistical techniques discussed above in-combination with the inclusion or exclusion of the disputed data points.
l l
l
Note that 10 of the 12 values are .32 weight percent copper. The-two values that differ from 0.32 weight percent both result from the exclusion of data and the use of an improper statistical technique. )
The NRC staff's value.of 0.35 weight percent discussed in reference (d) was obtained.by excluding all of the data' discussed in Attachment C and sorting the remaining data by.weldment as discussed in Attachmeret B.y As I discussed above and in Attachments B and C, the. technical logic behind this calculation is not evident.
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- 00000000000000000 00 00 0 0 0 000000000000000 0 0 2 01 08800701 2909908 64 09 9 7 4 32095597 687797 4 0 33322332333232232 33 32 3 2 3 444333333444444 12 3 6" 00000000000000000 00 00 0 0 0 000000000000000 0 0 3.
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AAAAAAAAAAAAAAAAA SSSSSSSSSSSSSSSSS AA SS AA SS A
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- SSSSSSSSSSSSSSSSS WWWWWWWWWWWWWWWWW DDDDDDDDDDDDDDD O OA WW SS SS WW W S MM BBBBBBBBBBBBBBB M -
W W NNNNNNNNNNNNNNN W W" 33333333333333333 33 33 3 9 9 999999999999999 8 8 3 777 777777 7 7777777 77 7 7 7 6 6 666666666666666 8 8 7 7 7 7 77 7 777 7 7777777 77 7 7 7 6 6 88888888888888888 38 88 8 8 8 666666666666666 6 6 7 888888888888888 8 8 8 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 93999999999999999 99 99 9 3 3 0G000000000000000 00 00 0 0 0 000000000000000 1 1 22222222222222222 22 22- -
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M B (6 CCCCCCCCCCCCCCCCC AA BB C D E F iFeFFFFFFFFFFF G H 4444 44444444 44444 SS S5 S 4 4 4d 4444444444A44 4 4
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TABLE 4 l
Copper values Resulting from Possible Combinations of Data Exclusion and Statistical Technique.
1 1
'l Data Sort Data Sort Data Simple Reduction By By Average Method Weldment Technique, (Ref (a)
(Ref (d)). Lab, and Time and BAW Data (Ref. (c)) 1799)
Excluded i
NONE 0.32% 0.32% 0.32%
XRF 0.32% 0.32% 0.34%
(Groups) 2A, 3A) ;
l l
Initial 0.33% 0.32% 0.32%
B+W Weld i Samples i (Groups 4A,4D, 40)
" Credible 0.32% 0.32% 0.32%
Multiple" (Groups 2B,3B, SA,5B,5C)
Notes:
- 1) Reason #1 was responsible for the NRC's exclusion of the 18 measurements utilizing the XRF technique on broken charpy samples.
(2A, 3A)
- 2) Reason #2 was responsible for the NRC's exclusion of the threr initial B+W weld qualification measurements. (4A, 4D, 40)
- 3) Reason #3 was responsible for the NRC's exclusion of ten measurements th,at were not " credible multiple". (2B,3B,5A,5B,5C)
- 4) The NRC's value of 0.35% is obtained by excluding all of the above data (Reasons #1,2,3) and sorting the remaining data by weldment.
- 5) Nickel contents of 0.57 wt.% result from treatment similar to that described in reference (c). This would correspond to less than a l'F rise in RTpts
ATTACHMENT E -
I Reference (d) provided Commonwealth Edison Company with the NRC staff's evaluation regarding the acceptability of the reported values of the Copper and Nickel contents of the Zion Unit I reactor vessel. This evaluation concludes that these reported values are not acceptable.
However, this evaluation also contains numerous misquotes and factual errors. These instances are individually summarized and discussed below.
The page and paragraph numbers will refer to the SER contained in reference (d). Reference (d) is reproduced here as Attachment F to aid in the review.
Page 3 - second paragraph.
THE NRC STAFF'S CALCULATIONAL METHODOLOGY IS STATISTICALLY ERRONEOUS AND IS l NOT CONSISTENT WITH CECO'S TREATMENT.
The statement is made that "The approach used . . is to first i cverage the measurements made on each individual weldment as indicated in l Table I, keeping separate the measurements made by different analytical I techniques or at different t imes . This approach is consistent with that used in Section III.2 in WCAP-11350."
This approach is not consistent with the method used in the WCAP, I which is reproduced in Attachment G. Nowhere in the WCAP, and certainly not in Table III.2, were the measurements on individual weldments averaged.
l Purther, it is statistically erroneous to do so. The statistically correct way to average measurements made by different techniques and at different l times is not to average measurements on individual weldments, but rather to cort the data by laboratory, technique, and time of measurement. !
l 1 l l This issue is discussed in more detail in Attachment B to this )
I letter. That discussion will not be repeated h,ere. 1 Pagt 4 - First paragraph.
) BABCOCK AND WILCOX DID NOT STATE IN B&W 1799, P.6-5 THAT "...IN THEIR JUDGMENT THE MODERN MEASUREMENTS WERE MORE CREDIBLE".
The statement is made in reference to weld qualification welds that
" Single measurements made on them in 1969 are not shown in Table I ...
because retests made by B&W in 1983 of these and other wire / flux combinations
... and in their judgement the modern measurements were more credible (see
> B&W 1799, p.6-5)."
This issue is also discussed in Attachment C to this letter. As previously discussed in Section IV of WCAP 11350, the actual words on p.6-5 of the B&W report are reproduced belos and in Attachment H.
"Without attempting to pass judgement on the original methods, it l can be stated that the current methods provide more conservative values .."
Thus the report explicitly states that no attempts were made to pass judgement much less to assign credibility to one set of measurements over another. More conservative does not mean more credible.
t l
/
L__________
i BAW 1799 did not render a judgement regarding the credibility of these B&W ATA weld qualification measurements on pages 6-5 and 6-6. As a result, they have been included in the groupings in Attachment D.
I Page 4 - First Paragraph. /
THE ORIGINAL ATA SPECIMENS WERE NEVER MELTED The statement is made that "The retests were done on archive material in the form of weld chips, which were melted to form a button for analysis purpoJes, following the method used for the original measurements."
It is an incorrect and potentially misleading statement to assert that the procedure followed the original method. The original method was the Atomic Absorbtion Technique (ATA) and this technique does not use tcmelted buttons. In fact, the technique depends on dissolving the sample into a liquid solution. It would be a poor practice to remelt chips into a button in order to subsequently dissolve them.
Pege 4 - Second Paragraph.
THE MIDLAND SAMPLES HAVE BEEN CONSIDERED.
Reference is made in the subject letter to B&W copper measurements of 0.35 to 0.49% on the Midland Vessel Nozzle Drop-out.
Commonwealth Edison acknowledges that these values of copper content are indeed high. In fact they collectively constitute the upper regimes of the data's overall distribution. Nevertheless, since there is no technical reason for excluding these data, we have consistently included them in our statistical analyses provided in references (a), (c) and ;
l Attachment D. j In addition, reference (d) discusses the acquisition of twelve new measurements of Midland test blocks. These twelve new measurements have been placed in two groups based upon their separation in time.
Psge 5 - Second Paragraph.
l THE SWRI'S AUTHOR DID NOT QUESTION THE XRF MEASUREMENTS' VALIDITY XRF MEASUREMENTS ACTUALLY PRODUCED HIGHER RESULTS WHEN THE SAME SAMPLE IS d ANALYZED WITH DIFFERENT METHODS.
The statement is made that "The author of the SWRI reports ... i questioned their validity ...". This issue is also discussed in Attachment C to this letter. l The author of the Zion I, Capsule X surveillance report by SWRI did not question the validity of the X-ray fluorescence (XRF) data. He merely speculated on the reasons why the XRF data were different from those obtained on the same samples by Inductively coupled Plasma (ICP) techniques.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ - _ _ __ _ a
Attachment I contains the pertinent pages from SVRI-7484-001/1. On page 35 of this report, Table XIII lists the copper values obtained by the two techniques on the weld metal in question.
Specimen Technique % Copper W-25 ICP .216 W-25 XRF .27 W-28 ICP .218 W-28 XRF .25 Note that there were two direct comparisons between the XRF and ICP techniques on specimens W-25 and W-28. In both comparisons, the XRF technique yielded a higher result.
It is not logical to reject the XRF data as biased low when a direct comparison with a different technique on the same specimen shows that XRF yields higher results. The subject of direct comparisons with other tschniques was extensively discussed in Section IV of WCAP 11350. The NRC's Staff's review has not addressed that discussion.
P'ge 5 - Third Paragraph.
l THE ASSERTION OF " DIFFICULTIES WITH XRF MEASUREMENTS" IS BASED UPON FAULTY COMPARISONS.
Attachment C to this letter also contains a discussion on this 12 sue. Reference (d) cites the nickel contents obtained by the XRF technique in the SVRI report referenced above as being different from results obtained by Emission spectrographic Analysis (ESA) on different specimens as a further indictment of the XRF technique.
However, if one examines the data given in Table XIII on page 35 of the SVRI report, one finds the following nickel contents for two techniques u:ed on the same specimen.
Specimen Technique % Nickel W-25 ICP .53 W-25 XRF .57 W-28 ICP .545 W-28 XFR .49 Note that the average nickel content obtained by the XRF method was
.33% while that obtained on the same specimen by the ICP method was .537%.
Moreover, the ICP technique measured nickel contents of 0.53 to 0.545%, which is also outside the range of 0.57 to 0.62% measured by the ESA technique as cited in the Staff's review. Merely because one set of measurements fall outside the range of another is insufficient reason to cast doubt on either.
As further confirmation that the XRF technique yields accurate results on irradiated specimens, the reader is referred to Table XIII of the Zion 2 Surveillance Capsule T report by SWRI cited in the subject letter and reproduced as Table TV-1 in WCAP 11350. This table compares XRF nickel determinations with those done by ATA on the same specimens. The results are in excellent agreement.
Thus, contrary to what is implied in reference (d), direct comparisons of the XRF technique with two independent measurements of nickel show that the XRF technique gives accurate res'alts regardless of the state of irradiation.
Page 5 - Fourth Paragraph.
THE .22% VALUE WAS NOT OBTAINED FROM XRP.
COMPARISON OF XRF WITH OTHER TECHNIQUES SHOWS EXCELLENT AGREEMENT.
The statement is made that "For the Zion 1, Capsule X material, the copper measurement made by Westinghouse using XRF on unirradiated material was 0.35% comparea to an average of 0.26% by SWRI and 0.22% by Westinghouse,' using the XRF method on irradiated Charpy bars. The SWRI measurements on the Zion 2 Capsule T material also seemed low ..."
There is a factual error in these statements. The 0.22% measurement attributed to the XRF method was in fact obtained by the ICP technique as documented in the SWRI Capsule X report on page 33 (Attachment I). There is no technical reason why irradiation would bias ICP results. Further, if one examines all the data in the two SWRI reports cited for Zion 1 and Zica 2, the following is found:
Zion Independent Technique-AVG XRF-AVG XRF-Single Meas Unit (ICP or ATA) (Irrad) (Unirrad)
No.1 .22%Cu .264Cu .35%Cu No.2 .28%Cu .234Cu .28%Cu Avg. .26%Cu .245%Cu .315%Cu Again, one should note that the XRF (Irrad) and the independent techniques were both used on the same specimens and the average results are in excellent agreement. There is no technical reason why either the ICP or the ATA methods should be biased because of irradiation effects.
The XRF (Unitrad) analyses were performed on totally different cpecimens at different times and in different laboratories. Thus, it is not surprising that the agreement with the other techniques is not as good. This should not imply that one result is any more credible than the other.
As further confirmation that the XRF technique yields accurate results on irradiated specimenc, the reader'is referred to Table XIII of the .i Zion 2 Surveillance capsule T report by SWRI cited in the subject letter and reproduced as Table IV-1 in WCAP 11350. This table compares XRF nickel determinations with those done by ATA on the same specimens. The results are in excellent agreement, j Thus, contrary to what is implied in reference (d), direct comparisons of the XRF technique with two independent measurements of nickel i show that the XHF technique gives accurate results regardless of the state of J irradiation.
Page 5 - Fourth Paragraph. i THE .22% VALUE WAS NOT OBTAINED FROM XRF.
1 COMPARISON OF XRF WITH OTHER TECHNIQUES SHOWS EXCELLENT AGREEMENT.
The statement is made that "For the Zion 1, Capsule X material, the copper measurement made by Westinghouse using XRF on unirradiated material was 0.35% compared to an average of 0.26% by SWRI and 0.22% by Westinghouse,' using the XRF method on irradiated Charpy bars. The SWRI measurements on the Zion 2 Capsule T material also seemed low ..."
There is a factual error in these statements. The 0.22% measurement attributed to the XRF method was in fact obtained by the ICP technique as documented in the SWRI Capsule X report on page 33 (Attachment I). There is no technical reason why irradiation would bias ICP results. Further, if one examines all the data in the two SWRI reports cited for Zion 1 and Zion 2, the following is found:
Zion Independent Technique-AVG XRF-AVG- XRF-Single Meas Unit 1ICP or ATA) (Irrad) (Unirrad)
No.1 .22%Cu .264Cu .35%Cu i No.2 .28%Cu .234Cu .28%Cu i Avg. .26%Cu .2454Cu .315%Cu !
Again, one should note that the XRF (Irrad) and the independent techniques were both used on the same specimens and the average results are in excellent agreement. There is no technical reason why either the ICP or the ATA methods should be biased because of irradiation effects.
The XRF (Unirrad) analyses were performed on totally different specimens at different times and in different laboratories. Thus, it is not surprising that the agreement with the other techniques is not as good. This should not imply that one result is any more credible than the other.
page 6 - First Paragraph.
SECTION V.3 OF WCAP 11350 HAS BEEN MISINTERPRETED The statement la made that "In en attempt to make an estimate of the copper and nickel content of the Zion 1 and 2 surveillance weldments, WCAP-ll350 contains a comparison of the Charpy 30 ft. Ib shift data from the surveillance reports with predictions based on Revision 2 to Regulatory Guide 1.99. This analysis indicates a copper content of about 0.27% if the Zion data are assumed to fall on the mean curve. This is an estimate with considerable uncertainty because for the correlation used in developing the basis for Revision 2, one standard deviation is 28F and this corresponds to about 0.08% copper. Yet, it is an indication that the measurements of copper content made by SVR1 may indeed be low."
This interpretation of the data presented in Section V.3 of WCAP-11350 is evidently a misunderstanding of the intention of this information. This was not an attempt to make an estimate of the copper and nickel content of the Zion 1 and 2 surveillance weldments, but rather to show the over-conservatism associated with assigning a copper content of 0.35 wt.%
to the Zion 1 and 2 critical reactor vessel welds.
Surveillance measurements were presented for all reactor vessel submerged arc welds fabricated by Babcock & Wilcox with Linde 80 flux. That is, data were presented from seventeen reactor vessels containing Linde 80 welds in addition to Zion 1 and 2. The mean copper content of 0.27 wt % was obtained by fitting the surveillance measurements from all nineteen reactor vessels using the proposed RG 1.99-Revision 2 methodology. It was never intended that this value be defined as the copper content for Zion 1 and 2, although the actual Zion surveillance measurements do fall close to a mean irradiation prediction curve with 0.27 wt.% copper (0.60 wt.% nickel).
We do agree with the NRC that considerable uncertainty exists in estimating chemical content from surveillance measurements and that this was considered in the development of proposed RG 1.99-Revision 2. If the margins per the proposed regulatory guide are added to the mean curve (with 0.27 wt.%
copper) to account for this uncertainty, the "mean + margin" curve just about envelopes the surveillance measurements from the nineteen reactor vessels.
Moreover, a "mean + margin" curve with 0.32 wt.% copper conservatively envelopes all of the measurements, including Zion 1 and 2, and a "mean +
margin" curve with 0.35 wt.% copper is a gross over-prediction of not only Zion 1 and 2, but all measurements from Linde 80 welds.
The arguement that the Charpy surveillance data on B&W Linde 80 welds shows that the SWRI copper measurements are too low can equally well be used to show that the B&W ESA measurements are too high. Commonwealth Edison maintains that it is technically and logically incorrect to use this arguement to justify the rejection of any analytical chemistry results. For a complete discussion and interpretation of the Charpy surveillance data, the reader is referred to Section V.3 of WCAP 11350.
However, we reemphasize that the above discussion was never intended to be used as the basis for assigning the value of the copper content. It is merely an attempt to demonstrate the excessive amount of conservatism associated with the proposed copper content of 0.35 weight percent.
page 7 - Last Two Paragraphs THE NRC STAFF'S STATISTICAL METHODOLOGY IS NOT APPROPRIATE These two paragraphs contain a discussion of the results obtained by averaging the copper contents in individual weldments.
l This issue was extensively discussed in Attachment B to this letter. In summary, the proper statistical treatment for this data sets involves assuming that the copper content of every weldment made with the l same filler wire has essentially the same overall copper content. Thus, the statistical grouping of the data should be based upon the analytical techniques, laboratories, and time of measurement. If the sorting is performed by weldment, then the biases of the most frequently utilized laboratory or technique would unduely influence the results.
In addition, even if most of the data in question were to be excluded, and the appropriate statistical grouping utilized, a value of approximately 0.32 copper weight percent is obtained. This is thoroughly explored in Attachment D.
PAGE 9 AND TABLE I THERE ARE FOUR ERRORS IN THESE TABLES.
TABLE I There are two factual errors in this table. Under Zion 1 .
Surveillance Veldments, the third line has the wrong method of chemical analysis listed. Instead of XRP, it should be ICP, and we don't understand why these data are rejected. Similarly, under the Zion 2 Surveillance Veldments, the third line also has the wrong method of chemical analysis listed. Instead of XRP, it should be ATA, and it is unclear why these data are rejected either.
Page 9 The "M" and "CF" terms under the Utility Submittal data were omitted. The M term is 59F, which is the same as tha NRC margin term. The CF term is 203.1F, which is lower than the NRC CF term.
In addition to the above comments, reference (d) has not addressed a number of technical issues raised by Commonwealth Edison Company. These issued are briefly summarized below.
- 1. The statistical discussion presented in Section III.1 of WCAP 11350 demonstrated that the NRC's Staff's exclusion of valid data while continuir.g to endorse the highest measurements is not well founded in basic statistics. For example, given the NRC Staff's assertion regarding the existence of distinct populations within the data, the highest data endorsed by the NRC Staff has less than a 1 in 100,000 chance of belonging to the other NRC-Postulated populations. .
0% =
+
)
-l l
<l
- 2.Section III.2 of WCAP 11350 discussed the appropriate statistical .
grouping for data sets such as.the set of copper measurements. No !
mention of this discussion, other than the erroneous reference discussed in the "Page 3 - Second Paragraph" section.
- 3.Section IV of WCAP 11350 raised the general issue of the technical adequacy of the X-ray fluorescence technique. Nowhere in reference' (d) is the question of a definitive study of the accuracy of.the l XRF method addressed. In fact, .the only comparative measurements I made on the same specimens are those cited in these comments. As. l discussed previously, the XRF method yields results in excellent ;
agreement with those obtained by other techniques. . Thus there is i no more reason to reject XRF data because it is "too low"~than there is to reject the ESA data because it is "too high". Note )
that Attachment D includes all the data available including, "high" !
values as well as " low". j a
- 4. Reference (d) does not address the issue of the Filler Wire )
Examination that is discussed in section V.1 of WCAP'11350. That evaluation was based on a chemical analysis of bare filler wire as i reported in the B&W Owners Group Report. That analysis showed a j copper content of 0.30% for the weld wire used to produce the Zion ;
welds. H Commonwealth Edison Company acknowledges that not all of the.
preceding seven pages of comments directly relate to the central arguements contained in Attachments B, C, and D. However, the scope of the 1 inaccuracies contained in reference (d) reasonably calls into question all j aspects of reference (d). Under these circumstances, an independent review ]
of the technical issues should be performed to provide a technically sound ,
l bcsis for the resolution of the PTS issue for Zion Unit 1. '
l i
j 4
t I
b l
1 3419K l
l l
- [ $ 1E:o o,, UNITED STATES y g NUCLEAR REGULATORY COMMISSION g a WASHINGTON, D. C. 20886 9 gg g A.,,../ May 7, 1987 Dockets Nos. 50-295/304 Mr. D. L. Fa rra r Director of Nuclear Licensing Commonwealth Edison Company 1 Post Office Box 767 Chicago, Illinois 60690.
Daar Mr. Farrar: ,
SUBJECT:
ZION NUCLEAR POWER STATION, UNITS 1 AND 2 - MATERIAL PROPERTIES FOR FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST. PRESSURIZED THERMAL SH0CK EVENTS, 10 CFR 50.61.(TAC NO. 60777/8)
By letter dated January 17, 1986, the Commonwealth Edison Company (CECO) '
submitted infonnation required by 10 CFR 50.61. The staff reviewed the j submittal and found the copper and nickel contents unacceptable. A follow-up '
meating on October 3,1986, did not resolve the issue and another submittal, including WCAP-11350, was received by letter dated December 29,-1986. The. i 1
staff reviewed your December 29 submittal and again concludes that the reported values of copper and nickel content are not acceptable and that when RT is evaluated using proper values, the Zion Unit I vessel will reach the schning criterion before the end of licensed life. Our detailed evaluation is contained in the enclosed Safety Evaluation.
CECO, therefore, should submit plans for further. flux reduction'for Zion Unit 1.
The Zion Unit 2 vessel has the same controlling material,- but the reported fluence is low enough to prevent reaching the screening criterion at end of licensed life, i Sincerely, Daniel R. Muller, Director Project Directorate III-2 Division of Reactor Projects - III, IV, '
V and Special Projects
Enclosure:
As stated cc: See next page I
i
_ -_m
f i
Mr. D. L. Farrar Commonwealth Edison Company Zion Station .j cc:
- Robert J. Vollen, Esquire. Mr. Michael C. Parker, . Chief-109 North Dearborn Street Division of Engineering Chicago, Illinois 60602 Illinois Department of Nuclear 3
(
Safety Dr.. Cecil Lue-Hing J 1035 Outer Park Drive, 5th Floor i Director of Research and Development Springfield, Illinois 62704 '
Metropolitan Sanitary District 1 i of Greater Chicago l
100 East Erie Street Chicago, Illinois 60611 Mr. P. Steptoe' 1 Isham, Lincoln and Beale Counselors at Law ,
Three First National Plaza .l Sist Floor Chicago, Illinois 60602 .)<
Mayor of Zion .I i
Zion, Illinois 60099 Illinois Department of Nuclear Safety ATTN: Manager, Nuclear Facility Safety 1035 Outer Park Drive. 5th Floor i Springfield, Illinois 62704 U.S. Nuclear Regulatory Comission l R:sident Inspectors Office 105 Shiloh Blvd.
Zion, Illinois 60099 4
Rsgional Administrator, Region III U.S. Nuclear Regulatory Connission 799 Roosevelt Road Glen Ellyn, Illinois 60137
I 9,, UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION 3 waawamorow. o.c.aossa 4
oooe*
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO HATERIAL PROPERTIES FOR FRACTURE j TOUGHNESS REQUIREMENT 5 FOR PROTECTION AGAINST I PRES 5URIZED THERMAL 5 HOCK EVENTS - 10 CFR 50.61 COMMONWEALTH EDISON COMPANY ZION NUT. TEAR POWER STATION, UNITS 1 AND 2 00CKETS N0s. 50-295/304 J
l
1.0 BACKGROUND
l i
The PTS submittal for Zion 1 and 2 submitted by Commonwealth Edison on January 17, 1986 was based on a report by Westinghouse Electric Corporation WCAP-10962. The controlling reactor vessel material'from the standpoint of pressurized thermal shock evaluations was identified as the circumferential beltline weld in the Zion 1 reactor vessel and j the lower shell longitudinal welds in Zion 2. Both critical welds were-made with weld wire heat number 72105 and weld flux 8669 and are desig-nated WF-70 by Babcock and Wilcox (B&W), the vessel manufacturer.
To arrive at the best estimate copper and nickel contents for these vessel welds, Westinghouse averaged 87 measurements that had been reported for several weldments made with wire 72105. The majority of these came from studies made by B&W and reported in BAW 1799. Weld i
wire 72105 also had been used with weld flux 8773 to make a number of surveillance welds for B&W vessels including Zion 1 and 2. All were designated WF-209-1. The unusually large number of measurements of copper and nickel was also the result of some special studies conducted by B&W to determine throughwall variability in a nozzle dropout that contained WF-70 weld material and other studies on pieces of WF-209-1 material from the archives. '
l .
0 1
The chemical composition given in BAW 1799 was 0.35% Cu and 0.59% Ni.
However, in WCAP-10962, Westinghouse added about 30 data taken mainly .
l from two surveillance reports by Southwest Research Institute'(SWRI) j t
which gave the reselts of X-ray fluorescence measurements on irradiated broken Charpy bars. The measurements of copper content were signifi-cantly lower than the values given in BAW 1799. The average of 87 values was 0.32% Cu and 0.56% N1. The differences in the averages seem smell when one considers that the standard deviations reported in BAW 1799 were- l about 0.05% Cu and about 0.01% N1. However, the difference is crucial -
with regard to meeting the screening criteria at the end of licensed life for Zion 1.
In its review of the submittal for Zion 1, dated June 24, 1986 .the PWR-B Engineering Branch concluded that many of the 30 measurements added: to the list in WCAP-10962 were not credible, and the copper and nickel contents should be 0.35% and 0.59% as reported in BAW 1799. This finding.was trans-mitted to Commonwealth Edison by letter of August 14, 1986. At a meeting on October 3, 1986, representatives of Commonwealth Edison and Westinghouse presented additional data and arguments in support of their position. These arguments were rejected by the staff for reasons given below. The Connonwealth Edison position was put in writing in another submittal dated December 29. 1986 which referenced WCAP-11350, a compila-tion'of the technical information presented at the meeting.' This SER_is a review of the December 29, 1986 submittal.
1.1 Review of the December 29, 1986 Submittal and WCAP-11350 It is important to begin this review by restating the issue: to determine best estimate values for the copper and nickel contents of weld WF-70 in the Zion 1 circumferential beltline weld and the Zion 2 lower shell longitudinal welds. Having no measurements on those weldments per se, it is our practice to use measurements on other weldments made with weld wire of the same heat number 72105.
I Having no evidence that the weld flux lot affects either copper or nickel content, differences in flux have been ignored. Bec'ause most of the copper in the weld comes from the plating on the welci rod i from which the wire is drawn, plating thickness variability con- l tributes to the differences in copper content between weldments and within weldments as well. Nickel, on the other hand, is present' as an alloying element, and in heat 72105 the variability of nickel content is very small when measured by the emission spectrographic analytical method.
The approach used in this SER for arriving at a best estimate of the chemistry is to first average the measurements made on each indi-vidual weldment as indicated in Table I, keeping separate the measure-ments made by different analytical techniques or at different times.
This approach is consistent with that used in Section III.2 in WCAP-11350. From the discussion and data tabulated in WCAP-11350
l
]
. . ;,j l
]
there appear to have been measurements made on nine weldments.
1 Three of the nine weldments were weld metal qualification welds WF-70, WF-209-1, and WF-113. Single measurements made on them in 1969 are not shown in Table I (although they were included in the. '
listing in WCAP-11350) because retests made by B&W in 1983 of these and other wire / flux combinations showed the early measurements to be.
consistently lower than the retests, and in their judgment the modern
measurements were more credible (see BAW 1799, p. 6-5). The retests were done on archive material in the form of weld chips, which were melted to form a " button" for analysis purposes, following'the method used for the original measurements.
1 One of the nine weldments was from a nozzle dropout from the fabri-cation of a Midland vessel nozzle shell course. The average of 15 measurements made by B&W to study through thickness variability is shown in Table I. Copper content ranged from 0.35% to 0.49%. Nickel content ranged from 0.58% to 0.61%
4 i
The remaining five of the nine ,weldments were made to provide samples for surveillance tests of the effect of radiation on the material.
The averages of the copper and nickel measurements made by different techniques, different laboratories, or at different times are given in Table I.
L
i The measurements made on irradiated Charpy bars from the Zion I sur-veillance Capsule X and Zion 2 surveillance Capsule T are shown in' i
Table I, but they are not sufficiently credible to be included in this analysis for the following reasons: l l
- 1. The author of the SWRI reports from which these values come questioned their validity for reasons given in the Zion 1
, Capsule X surveillance report by Southwest Research Institute, i
1 In brief, the small size and the gama activity of the sample I complicate the analytical procedures.
1
- 2. Measurements of nickel content by X-ray fluorescence on J irradiated broken Charpy bars fall in the range from 0.47% to 1
1 0.57% Ni, whereas measurements by emission spectrographic analysis ranged from 0.57% to 0.62%, with 44 of 57 values being either 0.58% or 0.59% Ni. Th5s is an indication that there were difficulties with the X-ray fluorescence technique when applied to irradiated 'oroken Charpy bars.
- 3. For the Zion 1, Capsule X material, the copper measurement made
- by Westinghouse using X-ray fluorescence on unirradiated material i was 0.35% compared to an average of 0.26% by SWRI and 0.22% by l Westinghouse, using the X-ray fluorescence method on irradiated Charpy bars. The SWRI measurements on the Zion 2 Capsule T l
material also seemed low in comparison to the measurements made l by Westinghouse (see Table I).
___ _ _ - - _ L
I
)
4., In an attempt to make an estimate' of the copper and nickel con-tent of tha Zion 1 and 2 surveillance weldments, WCAP-11350 con- -
.1 tains a comparison of the.Charpy 30 ft. Ib. shift data from the
- d 1
surveillance. reports with predictions based on Revision 2 to "
Regulatory Guide 1.99.- This analysis indicates a copper content ,
of about 0.27% if the Zion data are assumed to fall on the mean -d curve. This is an estimate with considerable uncertainty because, for the correlation used in developing the basis for Revision 2, one standard deviation is 28 F and this corresponds to about 1
( 0.08% copper. Yet, it is an indication that the measu'rements of copper content made by SWRI may indeed be' low.
i The letter of December 29, 1986 from Commonwealth Edison which sub-mitted WCAP-11350 said that the staff's reservation about the credi-bility of X-ray fluorescence (XRF) measurements-was inconsistent with staff's position in its review of Amendment 100 from Fort Calhoun.
The latter cited studies of the chemistry of welds in the vessel head that were made with weld wire from the same heat number as certain girth welds. In response, the staff notes that the Fort Calhoun samples tested by XRF were also tested by emission spectrographic means and were not pieces of irradiated Charpy bars, which suffer from their small size and the effort of gamma activity of the sample on the measurement of the fluorescent peak.
_c_n-.-
Finally, there was some new information provided by B&W in the form of f emission spectrographic analysis results obtained from two test blocks taken from the Midland 1 Reactor Vessel Surveillance Weld. For Test Block BN 10 there were six measurements ranging from 0,35% to 0.38% Cu (average of 0.36% Cu) and 0.59% Ni, the same for all six measurements. '
For Test Block BN 6 there were six measurements. ranging from 0.32% to 0.41% Cu (average of 0.36% Cu). Nickel content ranged from 0.58% to 0.59%. These averages are given at the bottom of Table I.
The sumary of data in Table I plus the new information given above.provides data on three weld metal qualification weldnents, one nozzle belt' dropout (a full-thickness weld), and five surveillance weldments reported in WCAP-11350, plus two test blocks from another surveillance weldment. For the three weld metal qualification weldments there is only one measurement i for each. For the Zion 1 and Zion 2 surveillance weldments there is only one measurement each on unirradiated material, the others being deemed not credible as described above. For the remaining six weldments the average ,
copper and nickel values used in this analysis are the average of four or more measurements. The grand average of the 11 copper values is 0.348% Cu l
and for the nickel it is 0.585% N1.
Alternatively, if only the six weldments for which credible multiple (four or more) measurements are considered, the average copper contents are:
u l
l
8 0.419, 0.355, 0.357, 0.302, 0.362, and 0.360. The grand average of these six values is 0.359% Cu. For the nickel it is 0.589% Ni. Viewed another
- way, of these six values five are above 0.35% Cu and only one is below.
l Even if it is conceded that the two Zion surveillance welds have copper contents below 0.35%, and there is some evidence that they do, there are only three weldments below 0.35% Cu and five above.
2.0 CONCLUSION
The staff concludes from this analysis that its evaluation given in the earlier SER and repeated below is supported by this evaluation. The licensee's estimated copper and nickel values are not acceptable because they averaged suspect data with the credible data. We believe the more credible data should be given greater weight.
For Zion 1, the controlling beltline material from the standpoint of PTS susceptibility was identified to be the intermediate shell circumferential weld WF-70 (weld wire heat number 72105). For Zion 2, the controlling material was identified to be the lower shell longitudinal welds which were also WF-70.
The material properties of the controlling material and the associated margin and chemistry factor were reported to be:
.g.
Utility Submittal Staff Evaluation Cu (copper content, %) 0.32 0.35 Ni (nickel content, %) 0.56 0.59 I (Initial RTtlDT , F) 0 0 M (Margin,'F) --
59 CF (Chemistry Factor, 'F) --
226.8 Principal Contribution: P. N. Randall Dated: MAY 7 1987 O
_ __m. _ _ _ - . - . - - - --
TABLE I
SUMMARY
OF COPPER AND NICKEL CONTENTS OF WELDMENTS MADE WITH WIRE HEAT 72105 Code Weld No. of No. Flux Source Technique % Cu % Ni Measurements 3 Weld Metal Qualification Weldments WF-70 8669 B&W ESA 0.340* 0.580* 1 WF-209-1 8773 B&W ESA 0.400* 0.590* 1 WF-113 8688 B&W ESA 0.300* 0.610* 1 !
h 1 Nozzle Belt Dropout (Midland Vessel) !
i WF-70 8669 B&W ESA 0.419* 0.593* 15 5 Surveillance Weldments, WF-209-1 Crystal 8773 B&W Sury. ESA 0.390 0.100** 1 River 3 8773 B&W Archive ESA 0.355* 0.605* 4 3 B&W Sury.
Oconee 2 8773 ESA 0.350 0.590 2 l 8773 B&W Archive ESA 0.357* 0.580* 6 Oconee 3 8773 B&W Sury. ESA 0.295 0.580 2 !
8773 B&W Archive ESA 0.302* 0.582* 26 )
1 Zion 1 8773 WSCLI XRF 0.350* 0.570* 1 l 8773 SWRI XRF Irr. *** 0.259 0.543 8 I 8773 WARD XRF Irr. *** 0.220 0.540 2 )
Zicn 2 8773 WSCLI XRF 0.280* 0.550* I 8773 SWRI XRF(Irr.)*** 0.229 0.521 10 8773 WARD XRF(Irr.)*** 0.283 0.533 3 !
Two Test Blocks From Midland I Surveillance Weldment Test Block 8773 B&W ESA 0.362* 0.590* 6 BN 10 '
Test Block 8773 B&W ESA 0.360* 0.583* 6 BH 6 ESA = Emission Spectrometry Analysis XRF = X-ray Fluorescence Spectrometry l IRR = Irradiated Charpy Specimen B&W =
Babcock and Wilcox WSCL1 = Westinghouse (Spectrochem Laboratory, Inc.)
SWR 1 =
Southwest Research Institute WARD = Westinghouse Advance Reactor Divison
- The 11 weldments for which measurements were averaged.
- A-typical weld (See BAW 10144A), not included in NRC averages.
l' ***Not included in NRC averages.
_________-_-___a
Commenw 2li.h Edison One First National P!aza. Chicago. Illinois Adores > Reply to: Post Office Box 767 Chicago, Ininois 60690 0767 ,
December 29, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 ;
Subject:
Zion Nuclear Power Station Units 1 and 2 Pressurized Thermal Shock NRC Docket Nos. 50-295'and 50-304 References (a): January 17, 1986 letter from G. L. Alexander to H. R. Denton.
(b): August 14, 1986 letter from S. A. Varga to D. L. Farrar.
(c): July 10, 1986 letter from R. L. Andrews to A. C. Thadani.
(d): September 8, 1986 letter from D. 3. Sells to R. L. Andrews.
l
Dear Mr. Denton:
. Reference (a) contained Coramonwealth Edison Company's submittal for Zion Station in compliance with the requirements of 10 CFR 50.61. This document transmitted a copy of WCAP-10962, " Zion Units 1 and 2 Reactor Vessel Fluence and RTpts Evaluations". WCAP-10962 demonstrated that the Zion Units 1 and 2 Reactor Vessels remained below the applicable screening criteria. Reference (b) provided the NRC's review of reference (a) and concluded that Zion Unit 1 did not meet the fracture toughness requirements j of 10 CFR 50.61.
The NRC Staff stated in reference (b) that the limiting weld material properties utilized in the analysis contained in WCAP-10962 were l not acceptable. In their place, the NRC Staff has suggested the use of significantly higher values.
I l
{
Mr. H. R. Denton December 29, 1986 I
commonwealth Edison Company has thoroughly reviewed this issue and 4 has concluded that the copper and nickel contents of .32 and .56 percent, j respectively, that were utilized in reference (a), are the appropriate material properties of the limiting weld material for Zion Unit 1.
Commonwealth Edison personnel met with Mr. J. A. Norris and other members of ;
your staff in Bethesda, MD on October 3, 1986. Commonwealth Edison Company j presented a detailed and complete justification of those material properties j at that meeting.
1 At the conclusion of that meeting, NRC Staff members indicated their l continuing disagreement with Commonwealth Edison's technical position.
However, there was insufficient information provided by the NRC Staff )~
personnel for Commonwealth Edison to adequatell' identify the NRC Staff's specific areas of concern. Therefore, the technical material presented at the october 3 meeting has been compiled into WCAp-ll350, " Determination of Best Estimate Copper Content in Zion 1 and 2 Reactor Vessel Beltline Region l Critical Welds". The NRC Staff is requested to review this material and to document those areas which the NRC has found to be unacceptable.
l l Attachment I demonstrates that there is no statistical basis to l support the exclusion of a segment of the chemical analysis data. In addition, a detailed review of the various analytical chemical methods is performed. No reason is found to justify the preference or rejection of a particular method. Thus, on a technical basis, one must consider all the data to arrive at the best estimate of the copper content.
Appropriate statistical treatments, trend curve analysis of surveillance welds, and chemical analysis of filler wire all provide confidence that Commonwealth Edison Company's stated value of .32 weight percent copper content is conservative. All of the above evaluations support copper concentrations belcw Commonwealth Edison Company's stated value of 0.32 weight percent. In addition, if one were to exclude the values contested by the NRC Staff and apply the appropriate statistical treatments discussed above, the resulting copper content remains consistent with Commonwealth Edison's stated value. Thus, the NRC Staff's suggested copper content of 0.35 weight percent is unwarranted.
The NRC Staff also expressed reservation regarding the use of X-Ray Florescence (XRF) technique as a valid means of analyzing weld material for reactor vessels. This position appears inconsistent with the NRC Gtaff's previous acceptance of the XRF technique in support of Amendment #.100' for the Fort Calhoun Station Unit 1 (references (c) and (d)).
j Mr. H. R. Denton ' December 29, 1966 :
In addition, the NRC Staff. expressed concern at the october 3, 1986 1
meeting that Commonwealth Edison Company was not treating the issue of Zion life extension in an aggressive fashion. On the contrary, Commonwealth Edison Company is addressing the issue of plant life extension in a comprehensive-and complete fashion. Attachment 2 provides a summary statement of <
Commonwealth Edison's efforts in this complex area.
Five (5) copies of this submittal and the attachments are provided .
l for your review.
Please direct your response to this document and any questions that.
may arise to this office.
Very truly yours, PA dW P. C. LeBlond Nuclear Licensing Administrator 1m I
Attachments ec: J. A. Norris - NRR i
Resident Inspector - Zion 2535K er!
}
i ATTACHMENT 1 VCAP-11350 DETERMINATION OF BEST 8
ESTIMATE COPPER CONTEb'T IN ZION 1 AND 2 REACTOR VESSEL BELTLINE LEGION CRITICAL VELDS 1
1 i
)
[
WESTINGHOUSE class 3 CusTOfAER DESIGNATED DISTRIBUTION WCAP-11350
{
DETERMINATION OF BEST ESTIMATE COPPER CONTENT IN ZION 1 AND 2 REACTOR VESSEL BELTLINE REGION CRITICAL WELDS
(
K. R. Balkey Z. L. Kardos i J. A. Marshall S. E. Yanichko
{
Work performed for Commonwealth Edison Company November 1986 Approved by: Is ji. A. Meyer, Manager
-Cb-he c._
Structural Materials &
Reliability Technology Approved by: M_ _
f F. Inrietto, Ma'nageF
[
aterials Technology Although the information contained in this report is nonproprietary, no k distribution shall be made outside Westinghouse or its Licensees without the cus2omer's approval. i f
WESTINGHOUSE ELECTRIC CORPORATION !
Power Systems Business Unit P. O. Box 355 Pittsburgh, PA 15230 a i o... i o-..n ii
- _ _ _ _ _ - =
Table of Contents i
Pace 3 Table of Contents i l
List of Tables ii List of Figures iii
- 1. Introduction 1 If. Sources of Data 3 TIZa Statistical Evaluation 8
- 1. Evaluation of " Populations" 8
- 2. AnfAppropriate Statistical Estimate of Copper Content 13 i
8V. Technical Evaluation of Chemical Analyses 15 1, V. Supporting Technical Evidence 18
- 1. Filler. Wire Examination 18
- 2. Impact of Flux Lot 18
- 3. Evaluation of Surveillance Weld Data 19 VZ. Summary and Conclusions 27 1
V!2. References 28
{
i e
siw. i nn 3
I List of Tables
?
f Pace
]
II-1 Residual Material Cheinistry Data for Weld Wire 72105 5 II-2 Sources of Data 6 l III.1-1 Statistical Tests for a Given Population 12 Ill.2-1 Instrument Reproducibility by Sample Groups 14 i
XV-1 Comparison of X-Ray Fluorescence with 17 l l Other Chemical Analysis Techniques V.3-1 Identification of Reactor Vessel Surveillance 21 Program Weld Metal 1
i i
i
- (
(
p .io...i m
List of Figures Lake
[ III.1-1 Histograms: 87 Cu Values 9 111.1-2 Cu Content for Three Groups of Welds 10 V.3-1 B&W Submerged Arc Welds Made with Linde 80 Flux 22 i
V.3-2 B&W Submerged Arc Welds Made with Linde 80 Flux 23 ,
- RG 1.99 Rev. 2 Mean Curve l 1
V.3-3 E&W Submerged Arc Welds Made with Linde 80 Flux 24 i
- RG 1.99 Rev. 2 Me' a n with Margin - 0.27 wt %
Copper Content 4 f
l V.3-4 B&W Submerged Arc Welds Made with Linde 80 Flux 25
- RG 1.99 Rev. 2 Mean'with Margin - 0.32 wt %
Copper Content V.3-5 B&W Submerged Arc Welds Made with Linde 80 Flux 26
- RG 1.99 Rev. 2 Mean with Margin - 0.35 wt %
Copper Content e
+
v u .so-oes m 2
I 1
i l
l I. INTRODUCTION i
The purpose of this report is to document material presented at the October 3, 1986 meeting between Commonwealth Edison Company and the U. S. Nuclear Regulatory Commission concerning the best estimate of the copper content in the limiting welds of the Zion Unit 1 and 2 reactor vessels. The background material relative to the identification of the critical beltline region welds relative to pressurized thermal shock was previously documented (1].
A number of chemical analyses have been conducted on weldments made with the same weld wire used in these Zion welds. The NRC has suggested that some of the data be rejected on the bases of statistical and technical arguments (2].
1 This report treats the copper content data rigorously from both a statistical and technical point of view in an effort to determine whether there is in fact l any basis for rejecting or favoring one or more particular data sets. From this evaluation, it is concluded that there is no valid reason of any kind to support the rejection or favoring of any given data set or sets, It follows that all of the data must be considered in arriving at the best estimate of the copper content of the welds in question, as required by 10CFR50.61.b.2.iii.
For convenience, the report is divided into four sections. The first section describes the various sources of the data. It will be seen that several different analytical methods were employed, that several different laboratories conducted the chemical analyses, and that many different samples were used. The second section discusses the statistical treatmer.t of the data. It is shown that the data cannot be divided into discrete populations by a statistical analysis alone. Further, even if the data are grouped by non-statistical means, there is no statistical method by~which any particular group or combination of groups can be favored over the others. In the third section, the various methods of chemical analyses are discussed. The NRC has stated that some of the analytical methods are not credible and that some of the analyses are in error (2). It is shown that there is no technical reason vu, w-w m 1
aw
_ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ . . _ _ . . _ _ _ _ ____ ._ .____J
for these statements and that from the data and procedures used, no given chomical analytical method or result can be judged inferior to any other.
Finally, in the fourth section, other technical evidence will be presented to support the contention that all the data must be used in assessing the copper concentration in the critical welds of the Zion reactor vessels.
l t
l 9
sw. ws m, y
II. SOURCES OF DATA l
Fast neutron irradiation-induced changes in the tension, fracture, and impact properties of reactor vessel materials are largely dependent on chemical composition, particularly in the copper concentration. The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper concentration within the reactor vessel weldments.
To address the variation in chemistry, Babcock & Wilcox (B&W) performed a reacter vessel beltline weld chemistry study of eight B&W vessels, including Zion Units 1 and 2, and reported the results in BAW-1799 (3) for the i Westinghouse Owners Group (WOG). The scope of work included collecting I existing sources of chemistry data, performing extensive chemical analysis on the available archive reactor vessel weldments, and developing predictive methods with the aid of statistical analyses to establish the chemistry of the reactor vessel beltline weldments in question.
2n addition to the B&W report BAW-1799, the WOG Reactor Vessel Beltline Region !
Eeld Metal Data Base was used by Commonwealth Edison in the preparation of their pressurized thermal shock submittal. The WOG data base, which was developed in 1984 and is continually being updated, contains information from weld qualification records, surveillance capsule reports, B&W report BAW-1799, and Materials Properties Council (MPC) and NRC Member MATSURV. data bases.
For each of the welds in the' Zion IJnits 1 and 2 beltline region, a material data search was performed using the WOG data base. Searches were performed for materials having the identical weld wire heat number as those in the Zion vessels, but any combination of wire and flux was allowed. For all of the data found for a particular wire, the values of copper and nickel content were averaged and the standard deviations were calculated. Although several other additional elements were not needed for PTS considerations, they were tabulated fer the sake of completeness.
Data was obtained for weld wire heat 72105, which is associated with the critical vessel weld seams in both units, as well as weld wire 71249, which is associated with the Unit 2 girth weld. RTPTS calculations (1) show that the m .0ae-..i m 3
~
l
- f 1
i submerged arc welds fabricated with filler wire heat number 72105 are limiting in regard to PTS for both Zion Units 1 and 2 reactor vessels. Since the copper content of these welds significantly impacts the PTS concern for the Zion vessels, statistical and technical evaluations have been performed to I define the best estimate copper content for these critical welds, considering the large scatter in data. (The chemical composition values for heat 71249 have already been addressed by the NRC staff via evaluations of reactor vessel materiais data for Turkey Point Units 3 and 4 (see Reference (4])). i j Before proceeding with the discussion of the statistical and technical l l evaluations, further background on the sources of the chemical measurements l l for we!d wire heat 72105 is presented nort.
Chemistry data shown in Table II-1 nss obtained from nineteen different '
sources of measurements on submerged are welds fabricated by American Chain and Cable with Mn-Mo-Ni weld wire heat 72105 and Linde 80 flux. The sources ;
of data were generated in a time period from 1969 to 1984. Table II-2 identifies the various sources of data, which include B&W Owners group data on submerged are welds made with three different lots of Linde 80 flux, as well as Zion 1 and 2, Oconee 2 and 3, and Crystal River surveillance program welds.
The chemical analyses were performed on surveillance welds (SW) in both unirradiated and irradiated cor.ditions, weld metal qualification (WMO) welds, weld matal qualification rotests (WMOR), and reactor vessel nozzle belt dropouts (NBD).
Four laboratories were involved in performing the chemical analyses by the following organizations:
Westinghouse (Spectrochem Laboratory Inc.) (WSCLI)
Westinghouse Advance Reactor Division (WARD)
Southwest Research Inst. (SWRI)
Babcock & Wilcox Co. (B&W) w.ao-san n 4
x
D d
s
_ . . .. . _ .., ... . . . . = . _ = . . _ .
_. 7 Table !!=I 9F!fSUAL MATERfAL CututsTPV DATA POR WELS Vf#t ifEAT '71'5 xu wr.250=i d,~ ,i = % _S-*
- Cm e 77U 7w =6C3. { qEe ARP e m w . e,,,p
$ 0.3V4ot;s . s0.090
- 0. m_m _ m 0.020 0.043 m . m.400 0 0.063 0.390 .j 18 esF= 209* l 8773 SW waCI.! IMF N 8 0.200 f 530 0.077 0.047 0.013 0.470 0.064 0.39e 22 WFe20*=a SM3 Su Sung saut v 8 0.270 6. ff70 22 WF*20*=4 SM3 Su SwMt snr v 8 0.250 0.490 2A WF=20**l 8773 SW Swat user V 0 0.240 0.See 22 6sF=20* = t 9773 Sw Swat ser v 9 0.260 0.940 <
2A esF = 209- 8 8773 Sw Sem t smF V e 0.240 0.s50 22 WF.20**l 5773 su swat smP V E 0.240 0.530 22 esF
- 2 09= 1 SM3 Sw Smut ser Y 8 0.200 0.S40 22 wF=209 8 SM3 SW Sosnt asur v 8 0.230 0.940 28 wr=20ht 8773 Sw teARO ICP V V 0.220 0,330 0.308 0.01* O.017 23 WF=209*n BM3 Su eseno 3CF v V O.220 0.420 0.07t 0.370 0.590 0.t04 0.020 0.048 0.ee0 0.072 0.390 32 esF. 20*+ 4 8773 gw twat ser y B 0.100 0.520 Sa wr=209*1 3773 su sesFt ser v 8 0.230 0.320 I 32 wr=20**n a773 sw Swnt asur Y S 0.230 0.340
- 3a hsF=209* 4 , S773 $w Sum! smP Y S 0.230 0.530 3a wr=209-8 8773 SW tw#1 3 F8' y e 0.270 0.330 3a wr=20*=3 3773 Ses Seat t anF v 8 0.210 0.4e0 3a wr*20*=8 8773 Sw Se8s t sRF V 3 0.240 0.240 3a wF.20*.3 8773 su gung smP V 5 0.230 0.470 l 3a wP=209-8 8773 su test s snF V e 0.220 0.520 3a wr.209 1 8773 Eu Suna asWP V 8 0.200 0.360 70 wr*200 5 9773 su meme a7a v V 0.240 0.530 0.024 0.820
- 38 esF=20h 8773 Su es4RO a7a v V 0.340 0.E20 30 wr=209.4 8773 Su 0.034 0.370 8 esaRO ATA Y V 0.300 0.500 0.026 0.490 am esF.209-t e773 este Dam ATA m V 0.300 0.4e0 0.067 0.005 0.020 0.360 0.120 0.330 49 WF = 20h l 3773 wMen abw Esa es S 0.400 0 990 0.081 0.024 0.370 0.0*O 0.380 .
AC WF=20**4 SM3 su thes Esa es e 0.330 0.900 0.410 0.010 0.023 0.690 AC ear.20*=t 8773 0.090 0.400 See thes Ese a S 0.340 0.S00 0.480 0.080 0.022 0.e40 0.092 0.390 4C esr= 20*= 8 8773 tu abas Esa es e 0.330 0.3e6 0.300 0.080 )
oc wF=20*=8 SM3 Sw 0.02: 0.630 0.000 0.390 ,a 86es Esa u 8 0.340 0.300 0.110 0.030 0.023 0.490 0.087 4C edF=209 8 8773 Sw thes tu a 0.390 3 0.360 0.0e0 0 430 0.010 0.021 0.600 0.089 0.390 4C esF= 209.s 3773 Su esas Sea a S 0.340 0.570 4C esF = 209* l 9773 Sw Deaf ESA N 0.800 0.0t0 0.022 0.440 0.087 0.380 S 0.370 0.390 0.090 0.000 0.015 0.S40 0.092 0.450 4C WW 20**t SM3 see has FSA m 5 0.330 0.410 4C est 209*l 8773 Su Shas See m 0.820 0.009 0.087 0.530 0.099 0.440 i
0 0.370 0.400 !
AC ear. 20*= 5 0773 Su thes Eta es S 0.330 0 400 0.009 0.089 0.See 0.099 0.430 i 4C esp.209 8 e773 ees a6es Esa a 0.620 0.110 0.00e 0.019 0.540 0.0*4 0.430 I 3 0.320 0.990 e.800 0.000 0.0&S 0.S*0 0.100 0.400 4C bdP* 20** l SM3 tw thes Esa m 3 0.320 0.890 0.s00 0.000 4C esF= 20*= t 8773 Sw 96ml 0.054 0.S80 0.800 0.400 ESA m 9 0.330 4C WF*200 8 S??3 Su has 0.900 0.032 0.007 0.013 0.570 0.830 0.450 EBA N 9 0.320 9.S*0 9.473 9.00e 9.088 9.See O.110 0.400 4C W*20*=4 8773 Eu Dhes Ese a S 0.300 0.330 0.080 q 4C wr=20h t an3 ou hw Esa a 0.013 0.036 0.840 0.100 0.400 e 0.3:0 0.900 0.090 0.0:3 0.084 0.c70 0.s00 1
4C WF=20**t 8773 SW thes ESa 0.400 M S 0.320 0.330 0.0e0 0.083 0.086 0.990 0.100 0.400 1
4C heFe200 8 8773 Su thw Esa a S 0.350 0.380 AC WF= 20h 8 SM3 su hw Esa 0.0e0 0.01S 0.,087 0.400 0.800 0.390 4C W =20**8 5773 es 5 0.350 0.See 0.Geo 0.088 _ 0.087 0.390 0.044 0.390 has Dhes Esa a S 0.300 0.See 0.0e0 0.013 0.087 0.390 AC ed'=20* 1 8773 tu thes ESA m c.091 9.390 S 0.340 e.neo 0.0e0 0.015 0.087 0.990 0.0*5 4C esF.209 8 8773 SW Been Esa m 0 0.300 0.3a0 0.400 AC esF = 20h l s773 . su a6m Esa a 0.070 0.038 0.087 0.eno 0.004 0.390 8773 e 0.2o0 0.890 0.800 0.0 0 0.016 0.570 0.090 0.390 4C hsF = 20*= t Ses Bhw Enh a 3 0.200 0.390 AC ter= 209- 8 8773 gas D6m 0.080 0.080 0.016 0.550 0.100 0.400 ESA m 0 0.300 0.930 0.000 0.010 0.016 0.S*O 0.800 0.390 4C edF=209 4 8773 Su Shw Eam u 8 0.300 0.890 4C esF = 20*= 4 5773 u 0.000 0.0t0 0.036 0.560 0.100 0.400 She Dbas ESA 8 0.270 0.Smo 0.070 0.016 0.034 0.770 0.000 0.390 AC 88'= 20** l 8773 Sas phes ESA u 8 0.300 0.Ree AC WF=20*=4 0773 Sw ptes sea a 0.070 0.083 0.023 0.490 0.080 0.390 5 0.350 0.Sec 0.040 0.013 0.022 0. MO 0.000 AC wr=20*=t SM3 Ses Seas ESA u E 0.330 0.300 0.390 AC edr=209-8 8773 See Ese m 3 0.070 0.013 0.022 0.e40 0.000 0.390 4C t6es
- 0. 290 0.870 0.t00 0.010 0.018 0.600 0.004 0.3e0 ofr= 20*= 3 4773 Ses thes Eta m 0 0.300 0.S00 0.010 i
4C WP-209 1 5773 Su Bhas ESA 0.090 0.017 0.390 0.0*s 0.380 d AC asW-20h l 8773 at 0 0.290 0.570 0.090 0.080 0.046 0.370 0.094 0.390 su thes sea a S 0.290 0.500 AC edr=209 1 e773 tu thes 0.0e0 0.080 0.017 0.600 0.400 0.290 tan es S 0.300 0.300 0.0e0 0.040 0.016 0.e00 0.097 0.390 4C edr-20hl 8773 Su Dbas Een m 8 0.330 0.380 0.070 0.080 0.017 0.400 0.800 0.390 SA wr=20h t 8773 tu this Een m 8 0.360 SA esF= 20*= 1 8773 SM ESA 0.300 0.t&O 0.010 0.022 0.680 0.00* 0.390 D6as 7 0 0 340 0.600 0.010 0.013 98 esF=209-1 0773 Gu Dbas ESA es 8 0.300 0.900 0.003 0.012 S8 esF.20*= t 5773 su Dbas San 0.087 0.610 0.096 0.390 V E 0.290 0 300 0.040 0.017 SC esF 20hg S??3 su had ESA N 5 0.390 0.800 0.000 0.013 0.621 6.000 0.073 0.430 43 wF=70 8660 eens has a7A a V 0.270 0.ee0 0.070 0.011 0.084 0.400 0.400 4E ut=70 See9 eFet t6es Esa a S 0.340 0.380 0.007 0.0t9 0.880 0.090 0.300 4F esF.70 M9 esSD thes a 40 Eta S 0.430 0.990 0.070 0.050 0.028 0.630 0.100 0.400 asF= 70 8649 N00 has Eta m S t.420 0.390 0.070 0.010 or edr=70 See9 N90 thes San m 8 0.020 0.e00 0 400 0.400 est 0.400 0.990 0.&30 0.080 0.020 0.e00 0.100 0.400 edF=70 See9 NGO Bbes Een as 0 0.390 0.390 0.0e9 0.00*
aJP wp.70 Seet NB0 0.019 0.See 0.100 0.400 D6m Bea m e 0.350 0.300 0.0e0 0.00* 0.018 0.330 4F W= 70 W9 M99 hw 0.000 0.380 (F
tes M t 0.330 0.380 0.0e0 0.009 0.018 0.340 0.800 0.380 ed'* 70 See9 N80 96es Eta m 3 AF ap.70 6649 NSO 0.390 0.800 0.090 0.000 0.010 0.390 0.100 0.390 4F D6es ESA N 8 0.370 0.390 0.090 0.009 0.080 0.S40 0 840 0.390 edP*70 h9 ssee MIB ERA N 8 0.340 0.990
- F uP=To h9 shes 0.090 0.00* 0.010 0.330 0 380 0.400 or sett Eta m 3 0.400 0.990 0.800 0.009 0.010 0.330 0.180 0.390 esF = 70 M9 NGO Dbas Esa m 8 0.470 0.600 AF edF= 70 See9 8e00 0.100 0.009 0.018 0.220 0.180 0.400 er us' 70 W9 Sbas tem m 8 0.470 0.650 0.600 0.009 0.0l? 0.4*C 0.820 0.480 hee hw ESA N S 0.490 0.610 0.100 0.080 0.018 er wr*TO See9 NGO thes 0.400 0.820 0.4e0 4F Eta m 0 0.470 0.630 0.110 0.00* 0.017 0.4e0 0.820 0.420 esFa?O Seet mee has Eta N O 0.440 0.600 0.120 0.010 0.010 0.400 0,180 0.ene 43 esF* l l 3 Sees are0 66es a7a m V 0.230 0.390 0.042 0.010 0.590 0.380 em ed'= l l 3 9600 hw eF40m Eam es 5 0.300 0.610 0.013 0.017 0.880 0.0e0 0.370 everees weawee 0.316 0 3e4 ses Dewsettene 0. h St? 0.0T*653 =
yt TABLE II-2 SOURCES OF DATA I No. Source Time Reference 1 Unirradiated Surveillance Capsules I A - Zion 1 - WCAP-8064 1973 [5]
B - Zion 2 - WCAP-8123 1973 (6) 1 l
l 2 Irrdiated Zion Unit 1 Capsule X - SWRI 06-7484-001 1984 (7)
A - Southwest Research B - Westinghouse ARD 3 Irradiated Zier Unit 2 Capsule T - SWRI 06-6901-001 1983 (7)
A - Southwest Research 8 - Westinghouse ARD 4 B&W Owners Group Program - BAW-1799 1983 (3)
A - Flux 8773 - Weld Metal Qual. - Orig. 1969 l
B - Flux 8773 - Weld Metal Qual. - Retest 1983 C - Flux 8773 - Sury. Weld Archive Mat'l 1983 D - Flux 8669 - Weld Metal Qual. - Orig. 1969 E - Flux 8669 - Weld Metal Qual. - Retest 1983 F - Flux 8669 - Surv. Weld Archive Mat'.1 1983 G - Flux 8688 - Weld Motal Qual. - Orig. 7 H - Flux 8688 - Weld Metal Qual. - Retest 1983 5 B&W Surveillance Reports l A - Oconee 2 Surv. Prog. Measurements B - Oconee 3 Sury. Prog. Measurements C - Crystal River Sury., Prog. Measurements
- u. ie..' " '
6
- e
I 1
The chemical analysis techniques used in performing the analysis for copper and nickel content were as follows:
- 1. Inductively Coupled Plasma Emission Spectrometry (ICP) i
- 2. X-ray Fluorescence Spectrometry (XRF) .
. 3. Atomic Absorption Spectrometry (ATA)
- 4. Emission Spectrometry Analysis (ESA)
The chemical analysis techniques represent either a surface (S) or volumetric (V) type of measurement. The ICP and ATA methods measure chemical content of the full volume of the sample material, whereas XRF and ESA methods measure chemical content only at the surface of the sample.
The next three sections discuss the statistical and technical evaluation of the chemistry data.
1 m .i3-..i m 7
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1!-
III. STATISTICAL EVALUATION 111.1 Evaluation of " Populations" in the " Discussion" section of Reference [2], the NRC staff states that of the 87 welds submitted to the NRC, ". . . Clearly, the values added by Westinghouse constitute a different population. . . ". It is not clear whether the NRC staff considers this statement to be deriveable on purely statistical grounds, that is without reference to prior knowledge of which welds belong to the " values ~ added by Westinghouse", or whether it claims that given the prior grouping, the values in question can be shown to be I f significantly different from the others.
I I
With respect to the former, it must be stated that there is no statistical test that will take a set of data with no prior groupings and decide whether l l all the values came from a single population or not. Thus, the existence of prior groups is a prerequisite to any meaningful statistical test.
Reference [2] discusses a histogram that leads the NRC to postulate the existence of two distinct groupings of the weld values. Figure III.1-1 shows two possible histograms of the 87 weld values. They differ in the choice of class, intervals; the top one begins with the interval (.185, .205) and the bottom one begins with the interval (.175, .195). The choice of interval influences the appearance.of the histogram, but neither on'e suggests a split of the data into the two groups proposed b'y the NRC.
Of course there are prior groupings in these data. Figure III.1-2 is a representation of the data showing meaningful groupings. At the bottom of the figure are 6 0's which represent the values of qualification and requalifi-cation welds. Next are the 25 welds denoted as Zion 1 and 2 surveillance welds. Notice that the plotting points are integers representing the number of welds at a given value. Group 2 consists of 41 welds marked B&W surveillance welds. Finally, there are 15 welds denoted as nozzle belt dropout welds. The 25 welds called Group 1 in the figure are the ones "added by Westinghouse", as discussed by the NRC in Reference (2].
w.. ....n n g
Figure III.1-1 Histogram: 87 Cu values s wdh of hiervol: 02 f, 26 '
Start.2.ng Interval = 0.185, 0.205 t 22 -
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0.170.190.210.230.250.270.290310.330.350.370.390.410.430.450.470.490.51 Cu (itdard tWeds)
Histogram: 87 Cu values
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, , , Starting Interval = 0.175, 0.195 22 -
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, ? YY? Y, 0.160.18 0.2 0.220.240.260.28 0.3 0.320.340.360.38 0.4 0.420.440.460.48 0.5 Cu (lederval trudpoirds) 9 1 -
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I With these three groups in view, we can now turn to the interpretation in the
" Discussion" section of Reference (2) that was mentioned above. Namely, there are statistically significant differences between the 25 Zion surveillance weld measurements and the remaining weld measurements. It is of course quite true that if the data are split into Group 1 versus Groups 2 and 3 combined then there are statistically significant differences. However, the same
{'
conclusion is also reached if the data are split into Groups 1 and 2 combined versus Group 3. In other words, the group of low values is significantly low and the group of high values is significantly high. A sensible test for these hypotheses would seem to be the well known Mann-Whitney non parametric test.
It has the advantage of not requiring the assumption that the data are (
normally distributed. The results of these tests are summarized in Table 111.1-1. The tests clearly indicate that if Group 1 is rejected, then Group 3 must also be rejected on the same basis.
From a statistical point of view, the rejection of the Group 1 data is arbitrary. There are at least three groups present and probably more. In fact, it is not unexpected that a series of chemical analyses done at different times, at different labs,'and by different analysts would exhibit group differences. A well known feature of round robin tests, as sponsored for example by ASTM committees, is the existence of a strong inter-laboratory component in the results. Specifically, when a collection of standard specimens (knowns) are sent to a number of laboratories for analysis, the errors contain a component representing the variability of the individual analyst about his mean and a component representing the variability of the m3ans of the analysts (that is, the laboratories) about (presumably) the true mean. In other words, if a set of specimens were sent to each of a series of laboratories for analysis, the results would be expected to fall into groups corresponding approximately to the laboratories.
no.. io-..n n 11 4
TABLE 111.1-1 STATISTICAL TESTS FOR A SINGLE POPULATION GROUP ID GROUP DESCRIPTION 1 Zion 1 & 2 surveillance welds 2 B&W surveillance welds 3 Nozzle belt dropout welds HYPOTHESIS TESTED RESULT Ho: 1=2&3 reject strongly*
Ho: 3=1&2 reject strongly*
- Note: Pr(discrepancy > observedlHo) :
100 000 Therefore, if Group 1 is rejected, then Group 3 must be rejected on the same basis. ,
l l
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12 l .-
III.2 An Acoropriate Statistical Estimate of Cocoer Content The 87 data points can be presented in terms of sample groups as shown in Table III.2-1. The data fall into 18 groups ranging from a single determination to 26 determinations. Each group of results provides a single estimate of copper content with a precision that varies not only because some groups have more determinations than others but also possibly because the precision per determination may vary across groups. The problem of making a combined estimate in such a situation is a classical one in statistics.
Cochran (9) is an excellent reference on this topic,:nnd the present effort is l
based on it.
Briefly, an examination of these data shows that the per observations variability differs across groups, and that group means differ by more than can be accounted for by tho variability evident within groups. In the language of Cochran's paper, this means that the overall estimate should be the semi-weighted mean of the group means. However, due to the large variation of the group means compared to the variation within groups, the group weights for the combined estimate turned out to be virtually constant.
In other words, the unweighted mean of the group means is equivalent. In carrying out the analysis on which these statements are based only the 5 largest groups were used. The.. reason is that many groups are not large enough to give useful information on the structure of the variation that effects the data. Specifically, some approximations on which the analysis rests require that groups not be too small.
The combined estimate using only the means of the 5 largest groups is 0.313 wt. %. These groups contained 26, 15, 10, 8, and 6 observations and had individual means of 0.302, 0.419, 0.229, 0.259, and 0.357, respectively. This accounts for 65 of the 87 observations overall. If all groups are averaged using the method of combination determined as appropriate for the largest 5 (the unweighted group mean), the result is 0.311 wt. %.
(Using the same approach for the nickel content, the result is 0.564 wt. % for the five largest groups and 0.561 wt. % if all groups are averaged.)
m.. m-m m 33
1 TABLE III.2-1 5
INSTRUMENT REPRODUCIBILITY BY SAMPLE GROUPS COPPER NICKEL .,-
TECH SOURCE OUAN (Wt. %) (Wt. %) g XRF WSCLI 2 0.315 0.560 SWRI 8 0.259 0.543 _
SWRI 10 0.229 0.521 ~_
Total 20 analyses, 3 groups by XRF i i
l ICPS WARD 2 0.217 0.538 Total of 2 analyses,1 group by ICPS ATA WARD 3 0.283 0.533 B&W 1 0.300 0.480 ;.
B&W 1 0.270 0.460 B&W 1 0.210 0.590 Total 6 analyses, 4 groups by ATA ESA B&W 1 0.400 0.590 6 0.357 0.580 4 0.355 0.605 26 0.302 0.582 2 0.350 0.590 2 0.295 0.580 1 0.390 0.100*
1 0.340 0.580 15 0.419 0.593 1 0.300 0.610 Total of 59 analyses, 10 groups by ESA
- Data point not included in the estimate nu.w.ini 34
1 IV. TECHNICAL EVALUATION OF CHEMICAL ANALYSES The NRC has questioned the technical adequacy of some of the analytical chemistry methods and results used to estimate the copper content of the intermediate to lower shell girth weld (2). Specifically, the X-Ray fluorescence technique used by Southwest Research Institute has been questioned on the grounds that calibration is difficult when using this technique. In addition, results reported in surveillance capsule reports (7, 8), where X-Ray fluorescence data and data obtained from the same specimens by other techniques were compared, were cited to support the argument that the X-Ray technique produced biased results. Reference (2) also stated that results obtained from early measurements on weld qualification samples by atomic absorption techniques "... are now known to be in error on the low side." The NRC staff did not cite a reference to support that statement.
Finally, the results obtained by Westinghouse using atomic absorption techniques and inductively coupled plasma methods (see Table IV-1) were not included in the NRC evaluation (2) for unknown reasons.
The procedures used by Southwest Research Institute were reviewed and no evidence was found to support the contention that their methods would yield biased results. Corrections for small specimen and radiation effects were made in the prescribed manner. In the absence of any data to indicate the contrary, j one must recognize the validity of the results. A direct comparison of the X-Ray technique with other analytical chemistry measurements has been made (7,
- 8) and the results are reproduced in Table IV-1. The various techniques l
yielded different results when used to analyze the same sample. However, an evaluation of the data shows no evidence to indicate that the X-Ray technique l yields results that are biased. Indeed, in one case the X-Ray results were l higher and in a second case lower than those obtained by other methods, and by almost identical amounts. The authors of these reports speculated on the reasons for these differences, but no definitive tests were run, and therefore, no conclusions can be drawn.
With regard to the early weld qualification results, the only comparison with other techniques that was found is reported in the B&W Owner's Group Report (3). Original weld qualification samples were remelted, analyzed, and the p.+"""
15
1 results compared with those obtained earlier. It is not surprising that the results are different. But in the absence of any data that indicates a procedure or calibration technique used in the original analysis was in error, the only conclusion that can be drawn is that the results are .
different, not that one has more validity than the other. In fact, the referenced report does not attempt to judge the original methods, but points out that the latter technique yields more conservative results (see pages 5-6 in Reference (3]). That is certainly true, but 10CFR50.61.b.2.iii requires that one obtain the best estimate, not the most conservative estimate of the copper content.
In summary., no rigorous technical studies or data,have been presented as evidence to show that any of the chemistry data cited in Table 11-1 should be rejected. In the absence of such evidence, all data from all sources must be considered to arrive at the best estimate of the copper content in the weld in question.
m.. i o-.. " "
16
W2 TABLE IV-1 . u COMPARISON OF X-RAY FLUORESCENCE WITH OTHER CHEMICAL ANALYSIS TECHNIQUES [7, 8)
Cepper Nickel p Tech Source 0 I_0 Wt. % Wt. %
XRF SWRI W 28
- 0.25 0.49 .
I W 25
- 0 27 0.57 I ICPS WARD W 28
- 0.218 0.545 -
W 25
- 0.216 0.530 ,
l 4
Copper Nickel Tech Source 10 Wt. % Wt. %
XRF SWRI W 37A ** 0.19 0.52 SWRI W 38A ** 0.23 0.54 SWRI W 39A ** 0.27 0.53 AA WARD W 37A ** 0.257 0.526 W 38A ** 0.309 0.518 W 39A ** 0.281 0.545 Notes:
o The sample specimens were first analyzed at SWRI by XRF, then dissolved and analyzed by ICPS (*) and AA ("*) at WARD.
o The XRF analysis represents a surface measurement. f o The ICPS and AA are volumetric analyses representing the entire i specimen elemental content, j i
a i o..: i o-..i n i 17 k .
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i.
.a ti V. SUPPORTING TECHNICAL EVIDENCE In order to further support -the determination of a best estimate copper content value for weld wire heat 72105, three additionci arguments are presented in this section. This evidence includes an examination of the ,
filler wire, the impact of weld flux lot, and the evaluation of irradiated surveillance weld data for all weld wires made with Linde 80 flux.
V.1 Filler Wire Examination ,
To examine the potential for a biased set of data, the cepper content in the filler wire, which is the principal source of copper in the weldment, was evaluated. In Appendix A of'B&W report BAW-1799, an examination of the filler wire heat number 72105 was performed. In this analysis, the filler wire was stripped of its copper coating. .The copper quantity in both the coatin.g and the bare metal was measured to determine the principle source of copper in the as-deposited weld metal, which is usually the surface coating. The quantity of copper content present on each wire sample was measured in compliance with ASTM D-168-77, Method D-Atomic Absorption Spectrophotometry. B&W looked at five different heats of wire, one of which was 72105. They used samples that were 8-10 inches long ebtained from two or more spools of filler wire. As shown in Tables A-3, A-4, and A-5 in BAW-1799 (3), B&W obtained a copper concentration of 0.230 wt% for the coating and a coppor concentration of 0.075 wt% for the bare filler wire for heat 72105. Thus, the total copper concentration found for wire heat number 72105 in this analysis was 0.30 wt%.
This value is in agreement with the best estimate copper values determined from the evaluation of all available measurements for weld wire heat 72105, as l l presented in Section III of this report.
l 1
V.2 Impact of Flux Lot As can be seen from Table 11-1, heat 72105 was used with one weld flux (Lot No. 8669) to make welds designated WF70, which was used in the Zion Unit 1 and 2 reactor vessel welds, as well as in several other vessel welds. It was used with another flux (Lot No. 8773) to make weld WF-209-1, which was used as the ,
\
i m e...'"'
18
surveillance weld for several plants, as well as weld WF-113 (Flux Lot %. :e 8688) , which is a qualification weld that has also been recently retested.
Althcugh copper measurements vary within these weld designations, the weld flux does not affect the copper content, as was described in Reference [1] and i also stated by the U.S. Nuclear Regulatory Commission in Reference (2].
V.3 Evaluation of Surveillance Weld Data A review of irradiation data from reactor vessel material surveillance programs was conducted to evaluate the effects of radiation on reactor vessel submerged arc welds fabricated by Babcock & Wilcox with Linde 80 flux.
Surveillance program weld data from Zion Units 1 and 2 and seventeen other plants shown in Table V.3-1 were evaluated. This assessment included data from PWRs as well as BWRs and also included data from one foreign plant (KORI Unit 1). !
In Figure V.3-1 a plot of the increase in 30 f t. Ib. transition temperature versus neutron fluence is presented for the B&W submerged are welds made with Linde 80 Flux. This figure indicates that radiation damage for the B&W submerged are welds made with Linde 80 flux'tends to saturate at neutron fluences greater than 1 x 10 19 n/cm and that the Zion surveillance welds 2
fabricated with weld wire heat 72105 and Linde 80 flux are neither the most sensitive nor the least sensitive to radiation.
If the data were evaluated as "same wire" surveillance weld, the mean s copper content for the weld would be 0.27 wt. % (with a nickel content of 0.60 wt. %), as shown in Figure V.3-2, using the methods.of proposed Reg. ,
Guide 1.99 Rev. 2 (10).
A comparison of the mean plus margin for a copper content of 0.27 wt. % (as determined using Reg. Guide 1.99 Rev. 2), 0.32 wt. % copper (as determined from the average of all the copper analysis performed on the welds made with wire heat 72105 (see Section II)), and 0.35 wt. % copper (as recommended by the U.S. Nuclear Regulatory Commission in Reference (2]) are shown in Figures
- w. io-m m gg
i V.3-3, 4 and 5, respectively. These plots show that use of a ecpper content of 0.35 wt. % for the Zion Unit 1 and 2 critical reactor vessel welds, as recommended by the U.S. NRC in Reference (2), is overly conservative in predicting radiation damage at neutren fluence. levels greater than 1 x 1019 i n/cm2 when compared to the surveillance weld data for all welds made with Linde 80 flux.
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20
i TABLE V.3-1 IDENTIFICATION OF REACTOR VESSEL SURVEILLANCE PROGRAM WELD METAL Weld Wire Linde 80 Flux Weld Plant Heat No. Lot No. Code No.
Zion 1 72105 8773 WF 209-1 Zion 2 72105 8773 WF 209-1 l R. E. Ginna 61782 8436 SA-1036 Point Beach 1 72445 8504 SA-1263 Point Beach 2 406L44 8773 WF-193 Turkey Point 3 71249 8445 SA-1101 Turkey Point 4 71249 8457 SA-1094 Surry 1 299L44 8596 SA-1526 Kori 1 T29744 8790 WF-233 Oconee 1 406L44 8688 WF-112 Oconee 2 72105 8773 WF 209-1 Oconee 3 72105 8773 WF 209-1 Three Mile Island 1 299L44 8650 WF 25 Arkansas 1 406L44 8773 WF 193 David Bessie 821T44- 8754 WF-182-1 Rancho Seco 406L44 8773 WF 193 Dresden 3 NA NA NA Quad Cities 1 NA NA NA ,
Quad Cities 2 NA NA s NA i
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7 VI.
SUMMARY
AND CONCLUSIONS From the results and evaluations discussed in Section III of this report, it-is concluded that there is no statistical basis for rejecting any of the data or chemical analyses described in Section II. Although one may make arbitrary combinations of groups of analyses and portray that they do not come from the same population as another arbitrary group, this effort results in a predetermined conclusion. For it has been demonstrated that other combinations, also arbitrary, can lead to a totally opposite conclusion; for example, that the 15 nozzle belt dropout weld measurements should be excluded. It is therefore submitted that on a statistical basis, the entire set of data must be considered in arriving at a best estimate of copper content.
In Section IV, the technical aspects of the various analytical methods are discussed. No reason could be found as to why any given method should be favored or rejected. Further, no evidence in the form of studies, tests, published papers, etc. to substantiate that a bias exists has been presented.
It is true that when different analytical techniques were used on the same or similar samples, different results were obtained, and the chemists have speculated on the reasons for those differences. But at this point, it can only be concluded that the results were different, i.e., there is no reason to say that one method was right and the other was wrong, much less to judge which was which. Thus, on a technical basis one must consider all the data to arrive at the best estimate of the copper content.
Appropriate statistical treatments, trend curve analysis of surveillance welds, and chemical analysis of filler wire all provide confidence that 0.31 wt % is an appropriate representation of the copper content. Therefore, the value of 0.32 wt % for the copper content cited in Reference [1] is considered to be conservative. '
27 m . . .
7 VII. REFERENCES
- 1. Furchi, E.L. et al., " Zion Units 1 and 2 Reactor Vessel Fluence and RT PTS Evaluations," WCAP-10962, December 1985.
i 2. NRC Letter Docket No. 50-295, " Zion Nuclear Power Station, Unit 1 -
l Requirements for Protection Against Pressurized Thermal Shock Events,"
i frcm Steven A. Varga to D. L. Farrar of Commonwealth Edison Company, August 14, 1986.
- 3. B&W Owners Group Report, BAW-1799, "B&W 177-FA Reactor Vessel Beltline Keld Chemistry Study," July 1983.
- 4. NRC Letter Docket Nos. 50-250 and 50-251, " Evaluation of Reactor Vessel Materials Data for Turkey Point Plant Units 3 and 4 Reactor Vessels", from S. A. Varga to J. W. Williams, Jr., of Florida Power and Light Company, April 26, 1984
- 5. Yanichko, S.E. and Lege, D. J., " Commonwealth Edison Co. Zion Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-8064, March,1973.
- 6. Yanichko, S.E. and Lege, D. J., " Commonwealth Edison Co. Zion Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-8132, May,1973,
- 7. " Reactor Vessel Material Surveillance Program for Zion Unit No.1 Analysis of Capsule X," Final Report SWRI Project No. 06-7484-001, March,1984.
- 8. " Reactor Vessel Material Surveillance Program for Zion Unit No. 2 Analysis of Capsule T," Final Report SWRI Project No. 06-6901-001, July 6,1983.
- 9. Cochran, W.G., "The Combination of Estimates from Different Experiments",
Biometrics, Vol. 10, 1954, pp. 101-129.
- 10. " Proposed Revision 2 to Regulatory Guide 1.99 Radiation Damage to Reactor Vessel Materials," U.S. NRC, February,1986.
m.u-m m 28
ATTACHMENT 2 DESCRIPTION OF COMMONWEALTH EDISON COMPANY'S l
[ PLANT LIFE EXTENSION STRATEGY 1
l l
power plant life extension is being addressed on a Corporate level for all of Commonwealth Edison's existing generating units. An integrated working group has been formed consisting of both upper corporate management and technical expertise from a variety of disciplines within the Company.
This group has embarked upon a five step process to formulate the appropriate strategy for plant-life extensions, both fossil and nuclear.
This five step process consists of:
(1) Assess the various issues involved (2) develop the appropriate alternatives (3) evaluate the level of risks and potential return for each of the postulated alternatives (4) select one of the alternatives (5) implement the specific plan.
Steps 1 and 2 are anticipated to be complete by April of 1987. The implementation of a specific plan for Commonwealth Edison Company is anticipated to occur in early 1990.
The task of developing a corporate strategy regarding plant life extension has resulted in the identification of several broad areas which contain a variety of individual issues. Among these three areac are:
l (a) Technical (b) Regulatory (c) Economic The technical issues surround the consideration of key system failures, repair or replacement of major components and the operating /
maintenance performance of the various generating facilities. The Regulatory issues include consideration of our relationship with our governing commerce commissions, consideration of pending environmental regulations, and Commonwealth Edison's relationship with the Nuclear Regulatory Commission.
The Economic issues involved include consideration of Northern Illinois load growth, investment costs, and the construction time required for the implementation of new capacity.
Each of the issues discussed above requires careful and c'omplete consideration to insure that the appropriate corporate decision is arrived upon. For example, to address only the first issue discussed above would require the analysis of any additional investment required to maintain major plant systems.
2-Key fossil plant systems include the steam generators, turbine, boiler control, feedwater heaters, percipitators, air heaters, condenser, water induction prevention, cables, transformers, and plant switchgear. Key nuclear plant systems include the ree.ctor vessels, steam generators, containment, cables inside of containment, reactor coolant system main Piping, and the steam turbines.
Each of the components discussed above requires a detailed technical and financial assessment of the various options available. Thus, it may not be prudent to invest prematurely in any given component when future work may result in an identification of a more limiting component.
The plant life extension working group is progressing towards its first milestones of completing the development of a range of possible alternatives by April of 1987.
L 2535K
6.3. Grouping by Filler Wire Heat Table 6 presents the elemental concentrations for the data set grouped by heat of filler wire. Each value is expressed as sample mean and standard deviations.
i Comparisons of the values for chromium, nickel, and molybdenum in Tables 5 and 6 illustrate the point that the concentrations of these elements are vir-tually independent of flux lot. This observation is suppo rt ed by previous B&W studies that showed that the concentrations of these element s in the as-depos it ed weld are the same as their concentrations in the filler wire, i.e., independent of flux lot and differences in welding parameters (heat input, travel speed, etc.) within the allowed ranges 10, As will be noted in section 7 this observation was used as the basis for pre-dicting the concentration of Cr, Ni and Mo in the beltline welds distributed among the three categories.
6.4 Weld-Metal Qualification Samples The original chemistry sample obtained during the weld-metal (w/m) qualifica-tion of 14 of the WF series weld / wire flux combinations (refer to Table 2).
These samples were re-analyzed by emission spectrographic analysis. The re-
/ suits of these analyses and the original e,hemistry obtained during w/m quali-fication were tested for equivalence. The significance levels were quanti-fied with the paired "t" test as explained in Appendix C. The results of this comparison are presented in Table 7. As can be obs e rved , significant deviations exist between the original and re-test analyses for c op pe r ,
The chemistries obtained in re-test on the w/m qualification test and the ac-tual reactor vessel archive nr.t eri als were also compared. This comparison was limited to seven WF wire / flux combinations and involved using the estimat-ed mean values from Table 5 for the chemistry of the actual reactor vessel ma-terials.
These comparisons indicate that original test results on WF series welds yielded Cu and Ni concentrations of 0.07 and 0.06 wt I, res pectively, lower than the actual value. Without attempting to pass judgement on the original 6-5 Babcock s,Wilcox
.me a .,
..)
j l
. )
')
cethods, it can be stated that the current methods ' provide morel conserva tive .
values ' in t ems . of ' irradiation damage. I ' 5'oth comparisons - show equivalence of .
methods and material sources for phosphorous, sulfur, s ilicon, and molybde-num. No consistent trend can be inferred from the test results on manganese.?
.c.
,3 1
1
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- i
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i o
.g sl 1
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i l
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1 6-6 Babcock & Wilcox
__.____1_.______ _ _ _ _ _ _ _ _ __ _!____.____________. .
cith an Instron Catalogue No. G-51-13A 2-in. strain gage 'extensometer and Hewlett Packard Model 70048 X-Y autographic recording equipment. Tensile tests on the plate material and the weld metal were run at 250*F and 550'F ;
l at a strain rate of 0.005 in/in/ min. through the 0.2% offset yield strength using servocontrol and ramp generator. The results, along with tensile data reported by Westinghouse on the unieradiated' materials (123, are presented in Table XII. The load-strain records are included in Appendix A.
Testing of the WOL specimens was deferred at the request of Common-eealth' Edison Company. The specimens are in storage at the SwRI radiation laboratory.
l D. Chemical Analvses Check chemical analyses were run en samples cut from the fracture end of selected tested Charpy specime'ns. All of the weld specimens were tested for copper and nickel content at SwRI using an X-ray flourescent technique. ,
Two each base plate and weld specimens were then sent to Westinghouse' Advanced Reactors Division Analytical Laboratory for complete analyses using gravi-metric (Si), combustion (C and S) and ICP Plasma (remainder of elements) .I 1
methods of analyses. The results are summarized in Table XIII. The differ-1 ences in the copper and nickel results can be attributed to at least three j i
l factors: !
1 (1) The ICP Plasma method measures the chemical content of the full volume of sample material while the X-ray flourescent methoo locks only at the surface of the sample; 1
l (2) The 1 square centimeter cross sectional area available .
from the Charpy samples is smaller than desired' for.da- l tecting small amounts of residuals with an bray flour-escent technique.
(3) The gamma activity of the sample increases the difficulty in accurately measuring the flourescent pon.
33
t a) o g% 3 9 06 773343 643311 Tn( 21 21 222222 222222 o
l E
n o
mi 93 548280 544632 rt 75 oa) .
543333 f g% 74 54 553344 i n( 11 1 1 111111 111111 no Ul E
e _
rs) 42 57
)
ust p S t es 50 46
(
c L crk 55 66 A at( 11 1 1 .
I rS R F
_ E T
A M e
_ r '
_ E ud) 75 55
_ C t ab 49 38 )
l l co1 20 77
(
c n aL( 33 33 A. 1 6
L F r
_ I I E0
_ I V 1 1
X R ) 85 36 057244 260 840 U T St 712 378 E S I T s 58 21 325631 L l t Uk 98 00 887778 898 888 B F U ( 11
_ A O
_ T t f
S0 EI S
_ I Z Y)
T t 27 84 618764 865 638 R % s E 34 43 115523 955 866 P 2. (k 76 88 665555 676 666 O 0 R
P E .
L p) m mm mm oo0000 I mF o0 00 oo0000 S e* o5 05 oo0000 oo0000 l
l T( R5 35 RR3366 RR3366 E
T
. 78 c . - 78 eo T T- - - - - - - - - - - -
pN EE WW S .
3 2
1 1 1
-) -) .
[
5 e l 5 e l l 3 s a 3 s a 3d a 8 r t 8 r t 2e ti 7 e e 7 e e
- 1t sr 8 v MI Bv " M "
[ r ee s s o Tt en d en d d p ee a t a l t a l M ar e ar e t r l T W l T W a P( P( it d o aH r
n X ra o t t na i e t l ) UD i
u b d s (
n p ) )
o a b c
( (
C C
e ,
, 1 l
I TABLE XIII I
CHECK CHEMICAL ANALYSIS RESULTS 1
Specimen Weicht Percent of Element' o 7 identification (a) Source (b) C S P Si Cu Ni Cr Mo ET-32 W .270 .019 .007 .176 .117 .476 .075 .460 S .16 .43 ET-37 W .247 .020 .007 .213 .113 .. 477 .073 .469.
S .13 .42 W-25 W .101 .019 .017 .619 .216 .530' .071 .374 S .27 .57 W-28 W .104 .020 .018 .661 .218 .545 .072 .390 S .25 .49 W-26 S - - .- - .26 .56 - -
W-27 S .- - - - .26 .54 - -
t W-29 S - - - - .24 .55 - -
W-3 0 S - - - - .26 .53 - -
W-31 S - - - - .28 .56 - -
W-32 S - - - - .25 .54 - -
(a) Cut from tested Cy specimen *
(b) W = Westinghouse Atomic Power Division S = Southwest Research i
35
q
\
.1 l
4 ATTACHMENT J REVIEW OUTLINE AND SU W RY i
Initial Ceco Submittal j September 7, 1977 - O'Brien to Schwencer Cu = 0.27 wt.%
v l'
Conynonwealth Edison PTS Submittal January 17, 1986 (Reference (a))
Cu = 0.32 wt.5 V
First NRC SER/ Rejection August 14, 1966 (Reference (b))
Cu = 0.35 wt.5 l
, Meetir.g in Bethesda October 3, 1996 V
Commonwealth Edison Rebuttal (included WCAP l1350)
December 29, 1986 (Reference (c) and Attachment G)
-] '
l v Second NRC SER/ Rejection May 7, 1987 (Reference (d) and Attachment F) )
Two Ceco Concerns
- 1) Statistical treatment
- 2) Exclusion of data e _
.M M _
Statistical Treatment Validity of Data Dstalled Review of (Attachment B) (Attachment C) Reference (d)
Improper Technique was used Valid Data was excluded (Attachment E)
In Reference (d) in reference (d) l} Ref. (d) contains errors
- 2) Some issues are not addressed Summary of Potential Values (Attachment 0)
Cu = 0.32 wt.5 3419K
. _ _ _ _ . _ _ - -