ML20058G839

Response to Applicant 820719 Answer Supporting NRC Motion for Summary Disposition of Dekalb Area Alliance for Responsible Energy/Sinnissippi Alliance for Environ Contention 9c.Only Items 1,2,7,11,12,15 & 17 Admitted True

ML20058G839
Person / Time
Site:	Byron
Issue date:	07/30/1982
From:	Chavez D SINNISSIPPI ALLIANCE FOR THE ENVIRONMENT (SAFE)
To:	Atomic Safety and Licensing Board Panel
Shared Package
ML20058G812	List: ML20058G802 ML20058G839 ML20058G860
References
ISSUANCES-OL, NUDOCS 8208030430
Download: ML20058G839 (11)

Text

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COM!'ISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

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COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454 OL l 50-455 OL (Byron Nuclear Power Station, )

Units 1 & 2) __

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INTERVENOR'S RESPONSE TO. COMMONWEALTH EDISON

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COMPANY'S " ANSWER TO THE NRC STAFF'S MOTION FOR

SUMMARY

DISPOSITION OF DAARE/ SAFE CONTENTION 9e" Referring to Applicant's " Statement of Material Facts as to which there is no Genuine Issue to be Heard," Exhibit I B. Intervenor admits as true only items numbered 1., 2., 7., 11., 12., 15., and 17 Intervenor's'pecifically denies the remaining items, and posits in their place short and concise statements of material facts as which there exists a genuine issue 8208030430 820730 .

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to be heard. Each such statement is supported by affidavit or other supporting document. Intervenor's statements of material facts are organized in sets by Applicant's numbered statements, e.g. Applicant's Fact 3 is controverted by items _below numbered 3A through 3E.

Statement of Material Facts in Issue 3A. The mechanism responsible for general inter-granular attack and its relationship to previously .

experienced caustic-induced, stress-assisted cracking is not yet clear. (" Materials Performance in Nuclear ..

Pressurized Water Reactor Steam Generators," hereinafter

" Mat. Perf. in PWR SG's, by Stanley J. Gr'en o and Peter N.

Paine, Nuclear Technology, vol.55, Oct. 1981, pgg. 10-29,

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19).

3B. The presence of an oxidizing agent can produce acid solutions in crevices by anodic dissolution even when the bulk water is neutral or slightly alkaline.

(" Corrosion of PWR Steam Generators," hereinafter " Corr. of PWR 's ," by R . Garnsey , Nucl . Energy , 1979, vol.18, April, pgg. 117-132.)

3C. AVT water chemistry treatment has been shown to not be capable of providing buffering capacity to prevent corrosion at locations having high concentrations of impuritie s . (" Mat. Perf. in PWR SG's," pg.21).

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i 3D. There is no reliable quantitative relationship between the compositions of solutions generated locally, the bulk water chemistry, and the thermal-hydraulic par-ameters. ( "C o rr . of PWR ' s , " pg . 121.)

3E. See Intervenor's Answer to NRC's Motion for Summary Disposition, at 105a, lo5b, lo5d,1071, and iloa.

4A. The occurrence of SCC on the inside surface of the SG tubes may be slightly reducad by limiting cold work techniques during'~ tube fabrication or by reducing re- ,

sidual stresses by thermal treatment. ("Effect of Some environmental Conditions on Stress Corrosion Behavior of '-

r Ni-Cr-Fe Alloys in Pressurized Water," H.R. Copson and G.

Economy , Corrosion, 24,55 (1968))

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4B. The improvement in SCC resistance to high tem-Derature water has been minimal, even after the use of a low temperature heat treatment during tube fabrication.(Id.)

40 . Extensive denting may produce plastic deformation l or ovalization of the tubing. (Malinowski Affidavit, og.8, A.9).

4D. Plastic deformation, i.e. denting, can lead to stress corrosion cracking of the tubing from the inside of the tub'e ' (" Corr of Pwr's," pg. 118, and Malinowski Affidavit, pg.10 A.12).

4E. See Intervenor's Answe to NRC 's Motion for Summary Disposition, at lo5c and 105d.

4F. See Facts 3A and 3D above.

4G. Byron Unit 1 is a D4 model which apparently has not had its Inconel 600 SG tubes thermally treated.

(Malinowski Affidavit. pg., A.7).

5B. Thinning has . occurred in the following plants using AVT water chemistry: Beznau 1. Ginna, Point Beach 1, Surry 1, Turkey Point 3 and 4, Indian Point 2, Prairie Island 2 and Tihange 1. (Correlation of following two tables: Table 2 in " Operating Experiences: Steam Gen- .

erator Tube Performance During 1979," hereinafter "Op.

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Exp. 1979," by 0.S. Tatone and R.S. Pathania, Nuclear ,,

Safety, vol. 22. no. 5, pg.636-655, and Table III in

" Mat . Perf. in PWR SG 's . ")

5B. Through 1979, the plants noted in Fact 5A had a tube failure rate range of .4 X 10- to 441 3 X 10-failures per tube year. (Id.)

SC . In 1979, the following plants using AVT water chemistry had tubes plugged due to wastage (thinning): .

Plant Tubes Plugged Cause of Failure Bezhau 1 15 6 Soc 8 wastage 1 fretting Ginna 19 13 SCC 4 wastage 2 UD

Plant Tubes PluRRed Cause of Failure Turkey Point 3 740 Denting 40 wastage Turkey Point 4 185 Denting 6 wastage (Table 3, "Op. Exp. 1979")

5D . The continued occurence of failures attributed -to SCC at Bezhau 1,and Point Beach requires explanation because although residual alkalinity may not have been .

completely eliminated, the caustic concentrations must have been reduced by prolonged operation with AVT chemistry. ,,

(Corr. of PWR's," pg. 124).

SE. See Intervenor's Answer to NRC's Motion for Summary l Disposition, at 105a,105b, and 105d.

6A. See 3A, 3D, SC , and SD . -

! 6B. The reasons for wall thinning, similar to phosphate wastage, occurring at the support plate intersections in plants using AVT, are not understood. (Malinowski Affidavit,

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pg.8-9, A.10).

8A. The features of denting are heavy magnetite l

buildup in support plate tube holes, collapsed tubes.

1 plastic st'ra'in of support plates, and cracking of support plate hole-to-hole ligaments, and SCC on the inside of the tubes. (" Effects of Copper and Nickel Compounds on

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Corrosion of PWR Steam Generator Materials," Earl L. White and Warren E. Berry, hereinafter " Effects of Cu. and Ni.,"

Nuclear Technolory, vol.55, Oct. 1981, pgg.135-150, 136).

8B. The roles of copper, nickel, and oxygen in the denting process are not fully understood. In laboratory testing, copper, nickel, and iron ions have acted as accelerators of the denting corrosion process. (Nat.

Perf. in PWR SG's," pg.25).

80. Concentration of chl'oride and sulfate salts -

alone is not sufficient toexplain denting on riverwater cooled stations with a history of alkaline boiler water. ,,

(" Corr. of PWR's," pg. 131).  :

8D. Denting has been known to have occurred when only one ppm of chloride was in the bulk boiler water. ( System Chemistry Considerations for Nuclear Steam Generators,"

hereinafter "Sys. Chem. for SG's," Frederick J. Pocock, Nuclear Technology, vol. 55, Oct. 1981, pgg. 117-123,119).

8E. One part per billion of iron in the feedwater can result in 40 to 50 lbs. per year of accumulated corrosion products in a 1000-MW plants SG. ("Sys. Chem, for SG's,"

pg. 122).

8F. In'1979, Indian Point 3, a PWR with no prior record of poor water chemistry control, which had been operating only since 8/76, and which had no condensor leaks that year (1979), had 437 tubes plugged due to denting.

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(Tables 1,2,and3, "Op. Exp. 1979")

8G. See Intervenor's Answer to NRC's Motion for Summary Disposition, at 105b and 105c.

9A. Potential sources of foreign objects include devices used to aid in SG fabric ~ation, objects carried in and lost by workers during fabrication or maintenance, and objects that became dislodged along the secondary system. (" Mat.

Perf. in PWR SG's," pg. 26).

9B. Sources of internally entrained particles which .

can cause mechanical tube damage are previously deposited sludge, corrosion products such as magnetite, and materials ,

from manufacturing. (" Mat. Perf. in PWR SG's," pg. 27).

90. In 1979, at Salem 1, 40 tubes suffered wear damage when devices used in the SG for tube- lane blocking came loose. ("Op. Exp. 1979," pg. 653).

9D. Following the Ginna accident on January 25, 1982, a TV optics inspection of the B steam generator revealed the following " foreign objects": a piece of wire near R43C34, a piece of tube wedged between R32C15 and stay bar, and two 30 inch tube segments, among other items. (NUREG-0909, "NRC Report on the 1/25/82 SG Tube Rupture at R.E. Ginna Nuclear Power Plant," pg7-18).

l 9E. Some of the items revealed by the TV optics inspec-tion of the Ginna plant were not foreign objects (Id, at 7-20)

10. Steam generator tube wear at antivibration bar

locations has been reduced, bet not fully solved..

(Malinowski Affidavit p.10, A.5) 13A. Unit 1 Byron is a D4 Model employing Inconel 600 SG tubes and carbon st. eel support plates (Malinowski affidavit, pg.5). -

13B. Use of Alloy Incoloy 690 instead of Inconel 600 would add corrosion resistance of chromium (30%) to acids in flowing high temperature water, and to resistance of nickel (60%) to chloride SCC. (" Material Options for ,

Steam-Cycle Heat Exchangers," Robert G..Schwleger,-Nuclear Power, June 1979, pgg. S5-S11, 59). ,,

13C. Through 1979, the following plants used Incoloy 690 or 800 tube material and stainless st' eel' support plates: Atucha 1, Biblis A and B, Borssele, GKN Neckar, Goesger, KKS Stade, and KKU Unferne.sser. Of these plants, all were on PO4 water chemistry yet only two had any cummulated tube defects. They were Borssele with 12 3 X Stade with .2 X 10-

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10 and KKS failures per tube year.

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(Tables 1 and 2, "Op. Exp. 1979.")

13D. Alloy 690 has the following benefits versus Alloy 600 heat treated at 700 C.:

l 1. B'etter resistance to SCC in pure water or low l

concentration caustic solutions.

2. Much lower corrosion product release in primary coolant.

3 Shorter 700 C stress relieving treatment possible 4

without sencitizing the alloy.

4. Good resistance to chemical cleaning agents.

(" Materials Requirements for Pressurized Water Reactor Steam Generator Tubing," hereinafter " Mat. Req. for PWR's,"

Nuclear Technology, by P.L. Berge and J.R. Donati, vol.55, Oct.1981, pgg.88-104, 103.)

13E. Alloy 690 does not represent any major difference with respect to Alloy 600 for industrial tube fabrication.

(" Mat. Req. for PWR's," pg.89).

13F. SCC has never been observed on Alloys 80d or 690 nor on stainless steels. Coriou has indicated that it only affects austentitic alloys with high nickel content.

(" Mat. Req. for PWR's," pg. 93).

14A. The D5 model incorporates design features which are improvements in reducing the potential for tube corrosion.

(Malinowski Affidavit pg.5. A7.)

14B. The D5 is not a design evolution from the D4.

The improved design features in the D5 mentioned in the Malinowski affidavit are employed at other steam generators.

, (See 130 above See also Tables 1 and 2. "Op. Exp. 1979")

, 16A. Tube wear has been less at the D4 KRSK0 plant 1

then in D3 operating plants for equivalent operating periods and power levels because the onset of unacceptable tube vibration in the D4 model occurs at a higher power level and power flow. The equivalent power level and l

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operating period refers to a power level and power flow which is above the threshold for fluid elastic instability in the D3, but below the threshold--for a D4 model.

(Burns Affidavit, pg.4-5, A.5) 16B. See Intervenor's Answer to NRC's Motion for Summary Disposition , at 105e , 105f , 1071, 108a , 108b , 108k , 110h.

18A. Commonwealth Edison has not identified the cause of preheater tube vibration. (Burns Affidavit, p.6 A.7.)

18B. The fretting in the Ringhals 3 steam generators .

(Model D3) is a new phenomenon, and is of a generic nature. Computer models have' failed to predict this ,,

phenomena. (Memo form Frigyes Reisch, SKI, to Joseph D.

Lafleur, NRC, Janary 20, 1982.)

180. Full scale model tests are very important in addition to the reduced scale model tests. (Id.)

18D. See Intervenor's Answer to NRC's Motion for Summary Disposition, at 104d ,108d ,1081, and 108m.

19/. Expansion of tubes at the support plates and sleeving tubes will make the tubes stiffen. It will not decrease the water flow velocity across the tubes.

(Murphy /Rajan Depo, pg.52) 19B. Modifications which divert a portion of the feedwater to the bypass line decrease the flow and increase the power level at which unacceptable vibration occurs. They do not decrease the water velocity. (Burns Affidavit, pg. 4-5, A.5) ,

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19C. See Intervenor's Answer to NRC 's Motion. for Summary Disposition, at 1071, 10Sc, 108d, 108f, 108g, 108h, 1081, 108j, 108k.

20A. Westinghouse has not released its work program or schedule for testing on the D4/D5 scale model.(Burns Affidavit, pg. 6-7, A.7,8. as indicated by the models'incompletion.)

20B. Westinghouse's evaluation can not be conducted until the test results are in. The scale model tests will not start uniil August or September, 1982, (Burns .

Affidavit, pg. 7, A. 8 & 9).

20C. Without the test results, it is impossible to judge ,,

what the evaluation will show or how long it will take.

(Id.p. 9, A. 13.)

21. In summary, it is impossible to judge the length of time it will take to implement an appropriate design modification at Byro.n because none of the steps mentioned in Facts 20A-200 above have been completed.

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