ML20207K543

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Determination of Best Estimate Copper Content in Zion 1 & 2 Reactor Vessel Beltline Region Critical Welds
ML20207K543
Person / Time
Site: Zion  File:ZionSolutions icon.png
Issue date: 11/30/1986
From: Balkey K, Kardos Z, John Marshall
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20207K533 List:
References
WCAP-11350, NUDOCS 8701090445
Download: ML20207K543 (34)
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Text

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WESTINGHOUSE CLASS 3 CUSTOMER DEstGMATED DISTRIBUTION WCAP-11350 DETERMINATION OF BEST ESTIMATE COPPER CONTENT IN ZION 1 AND 2 REACTOR VESSEL BELTLINE REGION CRITICAL WELOS K. R. Balkey Z. L. Kardos J. A. Marshall S. E. Yanichko Work performed for Commonwealth Edison Company November 1986

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s H Approved by:

/T.

9. 9M A.~

2 6 4:t 74m Meyer, Manager ,

Structural Materials &

Reliability Technology Approved by: - _ , [

  1. a.F."Enrietto,Ma'nageF~

M terials Technology Although the information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its Licensees without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORATION Power Systems Business Unit P. O. Box 355 Pittsburgh, PA 15230 Q10hb e

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Table of Contents Page Table of Contents i _

List of Tables ii List of Figures ,

iii I. Introduction 1

_=

b II. Sources of Data 3 III. Statistical Evaluation 8 --

1. Evaluation of " Populations" 8 =
2. An Appropriate Statistical Estimate of Copper Content 13 II IV. Technical Evaluation of Chemical Analyses 15 V. Supporting Technical Evidence 18
1. Filler Wire Examination ~

18

2. Impact of Flux Lot 18
3. Evaluation of Surveillance Weld Data 19 VI. Summary and Conclusions 27 VII. References 28 v oc.:io...m i j

List of Tables Paae II-1 Residual Material Chemistry Data for Weld Wire 72105 5 ---

II-2 Sources of Data 6 III.1-1 Statistical Tests for a Given Population 12 III.2-1 Instrument Reproducibility by Sample Groups 14 L

IV-1 Comparison of X-Ray Fluorescence with 17 Other Chemical Analysis Techniques a

V.3-1 Identification of Reactor Vessel Surveillance 21 Program Weld Metal nu. io-m m

List of Figures Page 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 V.3-2 B&W Submerged Arc Welds Made with Linde 80 Flux 23

- RG 1.99 Rev. 2 Mean Curve V.3-3 B&W Submerged Arc Welds Made with Linde 80 Flux 24

- RG 1.99 Rev. 2 Mean with Margin - 0.27 wt %

Copper Content 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

I. INTRODUCTION 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].

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 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 i laboratories conducted the chemical analyses, and that many different samples were used. The second section discusses the statistical treatment 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 m.. +.in n i

p l

for these statements and that from the data and procedures used, no given chemical 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.

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II. SOURCES OF DATA 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 reactor 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 Westinghouse Owners Group (WOG). The scope of work included collecting 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.

In addition to the B&W report BAW-1799, the WOG Reactor Vessel Beltline Region Weld 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 Units 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 for 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 aiu ,io- m m 3 ,

submerged arc weld's 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 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 materials data for Turkey Point Units 3 and 4 (see Reference [4])).

Before proceeding with the discussion of the statistical and technical evaluations, further background on the sources of the chemical measurements for weld wire heat 72105 is presented next.

Chemistry data shown in Table 11-1 was ebtained 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 k 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 conditions, weld metal qualification (WMQ) welds, weld metal qualification retests (WMOR), and reactor vessel nozzle belt dropouts (NBD).

Four laboratories were involved in performing the chemical analyses by the following organizations:

Westinghouse (Spactrochem Laboratory Inc.) (WSCLI)

Westinghouse Advance Reactor Division (WARD)

Southwest Research Inst. (SWRI)

Babcock & Wilcox Co. (B&W) siu. io esini 4

.N Table 11-1 RESIDUAL MATERIAL CHEMISTRY DATA FOR WELD WIRE HEAT 72105 no. _ Coo. No. Flum origin Soure. Tech Ire Maas cu (Wt%) Ni(Wt%) . c(Wt%) S fWts) P (Wtt) S i f W**) _Cr(Wt%) Mo(Wt%)

la wf-2o9-1 8773- s.4 WSCLI XRF ~ N S - 0.350 0.570 - 0.090 0.013 0.020 0.680 0.063 0.390 to WF-209-1 8773 SW WSC1.E XRF M S 0.290 0.550 0.077 0.013 0.017 0.470 0.064 0.390 2A WF-209-1 8773 SW SWRI XRF Y S 0.270 0.570 2A WF-209-1 8773 SW SWRI XRF Y S 0.250 0.490 2A WF-209-1 .8773 SW SWRI XRF Y S O.260 0.560 t 2A WF-209-1 8773 SW SWRI XRF Y S 0.260 0.540 23 WF-OO9-1 8773 SW SWRI XRF Y S 0.240 0.550 23 bF-OO9-1 8773 SW SWRI XRF Y S 0.260 0.530 2A WF-209-1 8773 SW SWRI XRF Y S 0.200 0.560 2A WF-209-1 8773 SW ~ SWRI XRF Y S 0.250 0.540 23 WF-209-1 8773 SW WARD ICP Y V O.220 0.530 0.101 0.019 0.017 0.620 0.071 0.370 23 WF-209-1 8773 SW WARD ICP Y V O.220 0.550 0.104 0.020 0.018 0.660 0.072 0.390 3A WF-209-1 8773 SW SWRI XRF Y S 0.190 0.520 3A WF-209-1 8773 SW SWRI XRF Y S 0.230 0.520 3A WF-209-1 8773 SW SWRI XRF Y S 0.230 0.540 3A WF-209-1. 8773 SW SWRI XRF Y S 0.250 0.530 3A WF-209-1 8773 SW SWRI XRF Y S 0.270 0.530 3A WF-209-1 8773 SW SWRI XRF Y S O.210 0.480 3A bF-209-1 8773 SW SWRI XRF Y S 0.260 0.540 3A WF-209-1 8773 SW SWRI XRF Y S 0.230 0.470 3A WF-209-1 8773 SW SWRI XRF Y S O.220 0.520 3A WF-209-1 8773 SW SWRI XRF Y S 0.200 0.560 3D WF-209-1 8773 SW WARD ATA Y V O.260 0.530 0.024 0.520 3B WF-209-1 8773 SW WARD ATA Y .V O.310 0.520 0.024 0.270 i 33 WF-209-1 8773 SW WARD ATA Y V O.280 0.550 0.026 0.490

.4A WF-209-1 8773 WMQ B&W ATA N V O.300 0.480 0.067 0.005 0.020 0.5o0 0.120 0.330 ,

4B WF-209-1 8773 WMOR B&W ESA N S 0.400 0.590 0.011 0.021 0.570 0.090 0.380 e 4C WF-209-1 8773 SW B&W ESA N S O.350 0.590 0.110 0.010 0.023 0.690 0.090 0.400 ,

l 4C WF-209-1 8773 SW B&W ESA N S 0.360 0.580 0.110 0.010 0.022 0.640 0.092 0.390 1

4C WF-209-1 8773 SW B&W ESA N S 0.350 0.580 0.100 0.010 0.021 0.630 0.088 0.390 4C WF-209-1 8773 SW B&W ESA N S 0.360 0.580 0.110 0.010 0.023 0.690 0.087 0.390 I 4C WF-209-1 8773 SW B&W ESA N S 0.360 0.580 0.110 0.010 0.021 0.600 0.089 0.390 4C WF-209-1 8773 SW B&W ESA N S 0.360 0.570 0.100 0.010 0.022 0.640 0.087 0.380 l 4C WF-209-1 8773 SW B&W ESA N S 0.370 0.590 0.090 0.009 0.018 0.540 0.092 0.410

i. 4C WF-209-1 8773 SW B&W ESA N S 0.350 0.610 0.120 0.009 0.017 0.530 0.095 0.440 4C WF-209-1 8773 SW B&W ESA N S 0.370 0.600 0.100 0.009 0.019 0.560 0.095 0.430 l

l 4C WF-209-1 8773 SW B&W ESA N S O.330 0.620 0.110 0.008 0.019 0.540 0.094 0.430 4C WF-209-1 8773 SW B&W ESA N S 0.;520 0.590 0.100 0.008 0.015 0.590 0.100 0.400 40 WF-209-1 8773 SW B&W ESA N S 0.320 0.590 0.100 0.008 0.014 0.580 0.100 0.400 4C WF-209-1 8773 SW B&W ESA N S 0.310 0.590 0.082 0.007 0.015 0.570 0.110 0.410 4C WF-209-1 8773 SW B&W ESA N S 0.320 0.590 0.073 0.008 0.015 0.560 0.110 0.400 i l CC WF-209-1 8773 SW B&W ESA N S 0.300 0.580 0.090 0.015 0.016 0.560 0.100 0.400 1

- - - - - --- - --. e, m n n.= a nos a en n enn n ann i

_ - -. - . __ - - - - - - .~ ~ - - ._ - - _ - _ _- _ .- - - -

  • C =c-4vv-s o,,, s. - _ . . .

0.016 0.svo u.Avu v. ww G773 SW B&W ESA N S 0.320 0.580 0.000 0.015 4C WF-209-1 0.680 O.100 0.390 i j

4C WF-209-1 8773 SW E&W ESA N S O.310 . O.580 ' O.000 .O.015___O.017 .

0.590' O.OV4 0.390 SW C&W ESA N S O.310 0.580 0.080 0.015 0.017 4C .WF-209-1 0773 0.017 0.590 0.092 0.390 0773 SW B&W ESA N S 0.300 0.580 0.000 0.015 i 4C WF-209-1 0.080 0.015 0.017 0.590 0.095 0.400 j CC WF-209-1 8773 SW B&W ESA N S O.310 0.580 ESA N S 0.300 0.580 0.070 0.015 0.017 0.610 0.094 0.390 (C WF-209-1 8773 SW B&W 0.390 ESA N S O.280 0.570 0.100 0.010 0.016 0.570 0.098 4C WF-209-1 1'73 SW B&W N S 0.280 0.590 0.090 0.010 0.016 0.550 0.100 0.400 4C WF-209-1 6773 SW B&W ESA 4

N S O.300 0.500 0.000 0.010 0.016 0.560 0.100 0.390 4C WF-209-1 8773 SW B&W ESA 0.400 4

ESA N S 0.300 0.590 0.090 0.010 0.016 0.560 0.100 j 4C WF-209-1 8773 SW B&W N S 0.270 0.580 0.070 0.016 0.024 0.770 0.000 0.390 4C WF-209-1 8773 SW B&W ESA 4

N S 0.300 0.580 0.070 0.015 0.023 0.690 0.080 0.390 s 4C WF-209-1 8773 SW B&W ESA N S 0.310 0.580 0.060 0.015 0.022 0.660 0.080 0.390 CC WF-209-1 8773 SW B&W ESA 0.390 l

ESA N S 0.320 0.580 0.070 0.015 0.022 0.660 0.080 4C WF-209-1 8773 SW B&W 0.094 0.380 1

B&W ESA N S O.290 0.570 0.100 0.010 0.018 0.690 g 4C WF-209-1 8773 SW 0.017 0.590 0.098 0.380 '

SW B&W ESA N S 0.000 0.500 0.090 0.010

) 4C WF-209-1 8773 0.010 0.016 0.570 0.096 .O.380  ;

8773 SW B&W ESA N S 0.290 0.570 0.090 4C WF-209-1 0.000 0.010 0.017 0.600 0.100 0.390 l 8773 SW B&W ESA N S 0.290 0.580 1 4C . WF-209-1 0.000 0.010 0.016 0.600 0.097 0.390 L 8773 SW B&W ESA N S 0.300 0.500 4C WF-209-1 0.070 0.010 0.017 0.600 0.100 0.390

  • ! SW B&W ESA N S 0.280 0.580 (C WF-209-1 8773 ESA N S 0.360 0.580 0.110 0.010 0.022 0.650 0.089 0.390 ,

5A WF-209-1 8773 SW B&W '

l B&W ESA Y S 0.340 0.600 0.010 0.013 l SA WF-209-1 8773 SW i N S 0.300 0.580 0.083 0.012 0.017 0.610 0.096 0.390 53 WF-209-1 8773 SW B&W ESA B&W ESA Y S 0.290 0.580 0.010 0.017 5B WF-209-1 8773 SW N S 0.390 0.100 0.080 0.013 0.021 1.000 0.073 0.450  !

I SC WF-209-1 8773 SW B&W ESA l

l N V O.270 0.460 0.070 0.011 0.014 0.400 0.400 1

CD WF-70 8669 WMQ E&W ATA  !

ESA N S 0.340 0.580 0.007 0.019 0.510 0.090 0.380  !

j (E WF-70 8669 WMOR B&W - t ,

N S O.430 0.590 0.070 0.010 0.021 0.650 0.100 0.400 t 4F WF-70 8649 NBD B&W ESA l ESA N S 0.420 0.590 0.070 0.010 0.020 0.600 0.100 0.400 4F WF-70 8669 NBD B&W 0.100 0.400 l

B&W ESA N S 0.400 0.590 0.000 0.010 0.000 0.600 2

4F WF-70 8669 NBD 0.400 B&W ESA N S 0.390 0.590 0.080 0.009 0.019 0.560 0.100 4F WF-70 8669 NBD 0.380 .!'

1 ESA N S 0.350 0.500 0.000 0.009 0.018 0.530 0.090 4F WF-70 8669 NBD B&W 0.100 0.380 NBD B&W ESA N S 0.350 0.580 0.000 0.009 0.018 0.540 ,

4F WF-70 8649 0.018 0.550 0.100 0.390 8669 NBD B&W ESA N S 0.390 0.580 0.090 0.009

! 4F WF-70 0.018 0.540 0.110 0.390 8669 NBD B&W ESA N S 0.370 0.590 0.090 0.009 4F WF-70 0.018 0.530 0.110 0.400 '

8669 NBD B&W ESA N 5 0.360 0.590 0.090 0.009 j (F WF-70 0.018 0.530 0.110 0.390 8649 NSD B&W ESA N 5 0.400 0.590 0.100 0.009 i 4F WF-70 0.100 0.009 0.018 0.520 0.110 0.400 l CF WF-70 8669 NBD B&W ESA N S 0.470 0.600 ESA N S 0.470 0.610 0.100 0.009 0.017 0.490 0.120 0.410 (F WF-70 8669 NBD B&W 0.120 0.410 NBD B&W ESA N S 0.490 0.610 0.100 0.010 0.018 0.490 4F WF-70 8669 0.480 0.120 0.400 4F WF-70 8669 NBD B&W ESA N S 0.470 0.610 0.110 0.009 0.017 0.410 f

ESA N S O.440 0.600 0.120 0.010 0.018 0.490 0.110

4F WF-70 8669 NBD B&W ATA N V O.210 0.590 0.012 0.018 0.590 0.510 4G WF-113 8688 WMQ B&W i 0.017 0.580 0.060 0.370 WMOR B&W' ESA N S 0.300 0.610 0.013  ![

) (H WF-113 8688 1 Average Values 0.316 0.564 1

1 Std Deviations 0.066517 0.059653 , ,_

i i

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TABLE 11-2 SOURCES OF DATA No. Source Time Reference 1 Unirradiated Surveillance Capsules A - Zion 1 WCAP-8064 1973 (5]

B - Zion 2 - WCAP-8123 1973 (6]

2 Irradiated Zion Unit 1 Capsule X - SWRI 06-7484-001 1984 (7]

A - Southwest Research B - Westinghouse ARD 3 Irradiated Zion 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 B - Flux 8773 - Weld Metal Qual. - Retest 1983 C - Flux 8773 - Surv. Weld Archive Mat'l 1983 0 - Flux 8669 - Weld Metal Qual. - Orig. 1969 E - Flux 8669 - Weld Metal Qual. - Retest 1983 F - Flux 8669 - Surv. Weld Archive Mat'l 1983 G - Flux 8688 - Weld Metal Qual. - Orig. 7 H - Flux 8688 - Weld Metal Qual. - Retest 1983 5 B&W Surveillance Reports A - Oconee 2 Sury. Prog. Measurements B - Oconee 3 Surv. Prog. Measurements C - Crystal River Surv. Prog. Measurements vu. so-sen n s

The chemical analysis techniques used in performing the analysis for copper and nickel content were as follows:

1. Inductively Coupled Plasma Emission Spectrometry (ICP)
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.

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III. STATISTICAL EVALUATION III.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 L 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 significantly different from the others.

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 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 one suggests a split -

of the data into the two groups proposed by 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 Q'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 L.. tinghouse", as discussed by the NRC in Reference [2].

r e ,,o- w m g

Figure III.1-1 Histogram: 87 Cu values width of intervot .02 26 Starting Interval = 0.185, 0.205 22 -

20 -

18 -

16 -

14 -

12 -

10 -

7f/ /

r### #

4-2- -

17 9##

/ / / / / / / / 7 7 t - 9 a -

/

O ,

/, /, /, /, /, /, /, /, /, /, /, O, /, , ,

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O, ,

0.17 0.19 0.21 0.23 0.25 0.27 0.29 0.31 0.33 0.35 0.37 0.39 0.41 0.43 0.45 0.47 0.49 0.51 Cu (interval mkJpoints)

Histogram: 87 Cu values width of intervot .02 2, , Starting Interval = 0.175, 0.195 22 -

20 -

18 -

16 - -

14 -

12 -

10 -

e- 7 // 7 -

/f //

6- r- , f f f/

7 f

  • ~

h7 h hh

- . e?f f 44**f f ?sh.

0.160.18 0.2 0.220.240.260.28 0.3 0.320.340.360.38 0.4 0.420.440.460.48 05 Cu (Irderwsl midpoints)

, 9

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Figure III.1-2 CU Content for 3 Groups of Welds plot symbols = no. of welds, O= qualify l 1

l 15 Nozzle belt dro,anut welds 3 2 1 1 2 1 1 1 1 3 1 1 o

.c E

3 l Z -

41 B & W surv. welds E $ 2 1 3 4 9 6 5 1 1 3 5 2 1 2

0 2 -

o 3:

25 Zion 1 & 2 surv. welds 1 1 1 1 3 3 1 3 5 2 3 1 1 O

O O O O O i i 4 4 i i e i i i i i 4 0.18 0.22 0.26 0.3 0.34 0.38 0.42 0.46 0.5 .

CU Content (wt. %)

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

!!I.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 stati!.tical 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 timos, 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 means 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.

m. io-eeo n gg

TABLE III.1-1 STATISTICAL TESTS FOR A SINGLE POPULATION GROUP ID GROUP DESCRIPTION 1 Zion 1 & 2 surveillance welds L 2 B&W surveillance we'd.

3 Nozzle belt dropout welds HYPOTHESIS TESTED RESULT ==

L 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.

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!!!.2 An Appropriate Statistical Estimate of Copper Content The 87 data points can be presented in terms of sample groups as shown in Table 111.2-1. The data fall into 18 groups ranging from a single d3 termination 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, and the present effort is 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 the 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 1argest 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 (theunweightedgroupmean),theresultis0.311wt.%.

(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. ie m m 13

TABLE III.2-1 INSTRUMENT REPRODUCIBILITY BY SAMPLE GROUPS COPPER NICKEL TECH SOURCC OVAN (Wt. %) (Wt. %)

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 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 0.510 Total of 59 analysos, 10 groups by ESA

  • Data point not included in the ostimato e 0 88till

} .2

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 intormediate 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 samplos 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 ano 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, one must recognize the validity of the resultr. A direct comparison of the X-Ray technique with other analytical chemistry measuromonts has boon mado (7,

8) and the results are reproduced in Table IV-1. The various techniques yioided different results when used to analyzo the same sample. However, an evaluation of the data shows no evidence to indicate that the X-Ray technique yiolds results that are biased. Indeed, in one case the X-Ray results were higher and in a second case lower than those obtained by other mothods, and by almost identical amounts. The authors of those reports speculated on the reasons for these diffortnces, but no definitivo tests were run, and therefore, no conclusions can be drawn.

With regard to the early wold qualification results, the only comparison with other techniques that was found is reported in tho D&W Owner's Group Roport (3). Original wold qualification samplos wore romalted, analyzod, and the

m. +m m 15

results compared with those obtained earlier. It is not surprising that the results are different. But ir the absence of any data that indicates a procedure or calibration technique used in the original analysis was in error, the only :enclusion that can be drawn is that the results are different, not that one has more validity than the ether. 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 !!-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, io ee'"'

16

TABLE IV-1

' COMPARISON OF X-RAY FLUORESCENCE WITH OTHER CHEMICAL ANALYSIS TECHNIQUES [7, 8]

XRF Versus ICPS (2 Samples)

Copper Nickel Tech Source II! Wt. % Wt. %

XRF SWRI W 28

  • 0.25 0.49 W 25
  • 0.27 0.57 ICPS WARD W 28
  • 0.218 0.545 l W 25
  • 0.216 0.530 XRF Versus AA (3 Samples)

Copper Nickel Tech Source ID 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.

o The ICPS and AA are volumetric analyses representing the entire specimen elemental content.

aiu.:io- .""

17

~

V. SUPPORTING TECHNICAL EVIDENCE In order to further support the determination of a best estimate copper content value for weld wire heat 72105, three additional 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 copper 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 coating 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 ceppor content present on each wire sample was measured in compliance with ASTM 0-168-77, Method D-Atomic Absorption Spectrophotometry. B&W looked at five different heats of wire, one o'f which was 72105. They used samples that were 8-10 inches long obtained 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 copper 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 presented in Section III of this report.

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 u m. io-s"'

18 l

surveillance weld for several plants, as well as weld WF-113 (Flux Lot No.

8688) , which is a qualification weld that has also been recently retested.

Although copper measurements vary within these weld designations, the weld flux does not affect the copper content, as was described in Reference [1] and 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 ft. Ib. transition temperature versus neutron fluence is presented for the B&W submerged arc welds made with Linde 80 Flux. This figure indicates that radiation damage for the B&W submerged arc welds made with Linde 80 flux tends to saturate at neutron fluences greater than 1 x 10 19 n/cm2 and that the Zion surveillance welds fabricated with weld wire heat 72105 and Linde 80 flux are neither the most sensitive nor the least isnsitive to radiation.

If the data were evaluated as "same wire" surveillance weld, the mean 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

m. ie-un n ig

V.3-3, 4 and 5, respectively. These plots show that use of a 'opper c 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 neutron fluence levels greater than 1 x 10 19 n/cm2 when compared to the surveillance weld data for all welds made with Linde 80 flux.

c 4

r i u.. i o-.." "

20

s 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 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 NA s m.a o-es""

21

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i I

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OO' '-

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l  ;, ,

b to s5 go gg 30 18 2 FLUENCE (10 n/cm )

FIGURE V.3-4 B&W SUBMERGED ARC WELDS MADE WITH LINDE 80 FLUX - RG 1.99' EAN WITH fMRGIN - 0.32 NT% COPPER CONTENT -

1

$00 . - . .

' O Zion 1 and 2 Surveillance Weld _

(Weld Wire 72105 and Linde 80 Flux) h j 400 ii: O Miscellaneous Surveillance Welds _ _ .

.;u .- (Various Weld Wires and Linde 80 Flux) . v i. {.' . !. -l,

.;:1 - .. ', , j .

it. .

r~-

r : y  : -- ---

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(Mean .35 Cu, .59 Ni) r

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o 5 go ,5

\ l i 29 2g 3o 18 FLUENCE (10 i.,cm2 )

FIGURE V.3-5 B&W SUBMERGED ARC WELDS MADE WITH LINDE 80 FLUX - RG TAN WITH MARGIN - 0.35 UT% COPPER CONER h

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.

v oa. s o-w w y

VII. REFERENCES

1. Furchi, E.L. et al., " Zion Units 1 and 2 Reactor Vessel Fluence and RT Evaluations," WCAP-10962, December 1985.

PTS

2. NRC Letter Docket No. 50-295, " Zion Nuclear Power Station, Unit 1 -

Requirements for Protection Against Pressurized Thermal Shock Events,"

from Steven A. Varga to D. L. Farrar of Commonweal'ht Edison Company, August 14, 1986.

3. B&W Owners Group Report, BAW-1799, "B&W 177-FA Reactor Vessel Beltline Weld 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.
5. 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. 06-74S4-001, 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.

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