ML20063C657

Testimony of Wd Fletcher Re Issues Raised in Contentions 1,2,3(a)-(e),4 & 5 in Wi Environ Decade 820721 Motion Concerning Litigable Issues.Prof Qualifications & Certificate of Svc Encl

ML20063C657
Person / Time
Site:	Point Beach
Issue date:	08/04/1982
From:	Fletcher W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP., WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20063C632	List: ML20063C629 ML20063C641 ML20063C657
References
NUDOCS 8208270323
Download: ML20063C657 (30)

Text

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August 4, 1982 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

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WISCONSIN ELECTRIC POWER COMPANY ) Docket Nos. 50-266

) 50-301 (Point Beach Nuclear Plant, ) (OL Amendment)

Units 1 and 2) )

STATEMENT OF W. D. FLETCHER

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I am presently Manager, Steam Generator Development and Performance Engineering in the Nuclear Technology Division of the Westinghouse Electric Corporation. A statement of my qualifications and experience is attached.

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2. The purpose of this Statement is to address the l

issues raised in Contentions 1, 2, 3(a)-(e), 4 and 5 by Wisconsin's Environmental Decade in its July 21, 1982 Motion Concerning Litigable Issues. .

Contention 1

. Degradation of as few as one to ten steam l generator tubes in a pressurized water reactor l such as at Point Beach Nuclear plant (" Point Beach") could induce essentially uncoolable conditions in the course of a loss-of-coolant-accident ("LOCA"). .,

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3. Sleeving a steam generator tube does not increase the probability that the tube will rupture during a loss-of-coolant accident (LOCA).

4. Contention 1 is based on a stateme,nt in " Report to the American Physical Society by the Study Group on Light Water Reactor Safety," pablished in " Reviews of Modern Physics, Vol.

47, Supp. 1, Summer 1975, p. S85: ,

1. .'III]t was the consensus of the group that' steam generator tube failure during a severe LOCA could occur fre-quently. Moreover, it appears that rupture of a few tubes (on the order of one to ten) dumping secondary steam into the depres-surized primary side of the reactor system could exacerbate steam binding problems and induce essentially uncoolable conditions in the course of a LOCA. . . (emphasis supplied).

5. The cited statement is not relevant to sleeving of steam generator tubes at Point Beach or to the potential for rupture of the steam generator tubes at or in the vicinity of .

the tube where sleeving will occur. The statement was not based on any cited analytical or empirical studies, and was not related to the sleeving of steam generator tubes. No mechanism i

was postulated to explain how the rupture would occur as a result of a LOCA.

6. Increased loads on steam generator tubes during postulated design basis accident sequences have been evaluated I

by Westinghouse in an analytical and mechanical test veri-fication program. The results of these analyses and tests shes

that unsleeved steam generator tubes will maintain their integrity for all postulated design basis accident sequences, including the LOCA.

7. For a LOCA, the maximum steam generator tube stresses caused by rarefaction waves, blowdown and vibration forces occur in the U-bend (upper) region of the steam generator.

This region is not affected by the sleeving process. The stresses near the tubesheet, where the sleeving takes place, are lower.

8. A steam generator tube rupture (as postulated by the group reporting to the American Physical Society) is generally .

considered to be an open-ended guillotine break or its equivalent. However, the pressure forces during a LOCA are not a

s consistent with producing a break of that type or magnitude.

During a LOCA, the pr,imary side of the primary-secondary pressure boundary, i.e., the inside of the steam generator l

! tube, is depressurized. This means that the pressure force would press inward, with a tendency toward tube collapse, rather than rupture. Moreover, most of the sleeved length of a,c a e the tube [ J is within Lne 22"-thick tubesheet. The tubesheet provides additional resistance to the potential for rupture. Thus, rupture of a steam generator tube in the vicinity of the tubesheet, where the sleeve would be installed, i

l would not occur as a result of a LOCA, irrespective of whether or not the tube is sleeved.

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9. A postulated tube collapse has only a small potential for creating a leak path between the secondary and primary sides of the steam generator. Any leak resulting from a tube collapse would be significantly lower than a double-ended break or rupture due to the reduction in steam generator tube flow area. Such a leak would not be enough to affect ECCS perfor-mance. ,

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10. Sleeving a steam generator tube provides even greater margin against tube rupture along the sleeved length of the tube during a LOCA. The presence of the sleeve inside the tube strengthens the tube-sleeve assembly along the sleeved length of the tube, and provides additional support in that region, including added resistance to tube collapse.

11. Even if, for the sake of argument, one were to assume a tube rupture in the sleeved region during a LOCA, leakage would be prevented or limited.by the presence of the sleeve and i

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the sleeve-to-tube joints. Moreover, even if the tube were to sever completely around its circumference above the upper j oint (a region where corrosion has not been significant at Point Beach), the sleeve would act as a restraint, and thus would not allow a double-ended, unobstructed leak path, by: holding the end of the tube in place. Leakage would be limited by the small annulus between the sleeve and tube to approximately 5%

(approximately 12.5 gpm) of the rate which would be expected from the unobstructed leak path of a double-ended break. The l

NRC Staff concluded in the " Safety Evaluation Report Related to i

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Point Beach Unit 1 Steam Generator Tube Degradation Due to Deep Crevice Corrosion," November 30, 1979, at page 21, that j

" critical overheating of the fuel during a LOCA could only occur for leakage rates in excess of 1300 gpm." Therefore, over 100 sleeved tubes would be required to fail simultaneously and circumferentially, in a region where corrosion has not been significant, to even begin to affect the abi'lity of the ECCS to cool the core.

12. Thus, it is not considered credible to postulate

" essentially uncoolable core conditions in the course of a. . .

[LQCA]" as a result of sleeving steam generator tubes at Point Beach. In fact, because of the added structural integrity of the tube-sleeve assembly, and the more restricted secondary-to-primary pathway due to the presence of the sleeve in the hot leg end of the tube, sleeving would actually mitigate the consequences alleged in Contention 1.

Contention 3(a)

The process of sleeving steam generator tubes increases the probability of tube failures generally, and, of even greater significance, it substantially increases the risk of failures in the unconstrained free standing region of the steam generator specifically in, among other things, the following manner:

I (a) Inspectability. Present inspection

- methods in unsleeved tubes have been shown to be inadequate to dete.ct defects, and the complicat-ing presence of the sleeve inside the tube will make the detection of degradation, especially at the joints, even more difficult. Over time, the detection capability will continue to degrade.

Scaling will occur on the outer surface of the sleeve inside those tubes with through-wall defects because the all-volatile water chemistry

treatment used in lieu of phosphate chemistry can no longer maintain the secondary water completely free of solids. In the narrow confines of the crevice-like annulus, the rate of scaling will be accelerated by concentration effects beyond any scaling on the outside of the tubes in the free standing region where there is no crevice. Combined with the scaling will be other conductive impurities from the feedwater train and elsewhere that are also an unintended byproduct of all-volatile treatment and that will further degrade and confuse the eddy current signal. The inability to adequstely detect defects that can lead to primary-to-secondary or secondary-to-primary pathways for leakage will exacerbate the problems indicated in [ Contentions 3(b), (c), (d) and (e)].

13. The Point Beach steam generator tubes are vertically configured in an inverted U-shape, with each end placed into holes bored in a 22"-thick horizontal metal plate, ' called the tubesheet, at the bottom of the steam generator. Each tube end is secured by a hard roll and a weld at the bottom of the a,c,e tubesheet. The sleeves, which are either[ } in length, are installed in the ends of the tubes which are anchored in -

the tubesheet, and extend upward into the tube beyond the top a r c,e of the tubesheet no more thanl ,} respectively. The

! entire length of the sleeve constrains the tube, and sleeving in no way affects inspectability, corrosion, leakage, structural integrity, or ability to withstand accidents in the

" unconstrained free standing region" of the steam generators, i.e., the portion of the tubes above the sleeved region.

14. Sleeving is performed to repair steam-generator tubes and to provide resistance to the future potential occurrence of

primary pressure boundary degradation. There is no known mechanism in the sleeving of tubes which would increase the probability of tube rupture generally, or increase this possibility in the unconstrained free standing region of the steam generator tubes.

15. The structural integrity of the sleeved tube (i.e.,

the sleeve-tube assembly) for maintaining the primary-to-secondary pressure boundary under no.rmal and accident condi-tions has been extensively evaluated, by both testing and analytical verifications. The testing, described in Section 6.1 of the Point Beach Steam Generator Sleeving Report, WCAP-9960, Revised February 1982, involved the application of a variety of loading conditions to the assembly, including internal proof pressure tests, external pressure tests, axial and pressure loading cyclic fatigue tests, and tensile tests at operating temperatures. These- tests have demonstrated that the sleeve-tube assembly is well within the design limits for both normal operating conditions and postulated accident conditions, including the LOCA.

16. The testing, and the analytical verifications described in Section 6.2 of the Sleeving Report,' clearly show that the sleeve-tube assembly meets the requirements of the ASME Boiler and Pressure Vessel Code. The Code specifies the allowable limits of material stresses. The sleeve material is of higher strength than the original tube material, and the sleeve and tube assembly has a higher structural capability than the original unsleeved tube.

17. Because tube rupture along the sleeved length of a steam generator tube is not expected to occur as a result of LOCA forces, eddy current inspections of the sleeved length of the tube are not necessary for the purpose of avoiding this event. Nevertheless, effective eddy current inspections can be

-- and will be -- performed on the sleeved tubes, including the I

sleeved portions of the tubes.

18. The potential concentration of impurities, including scaling, is not expected to be increased by the presence of the t

sleeve. Since the temperature in the tube-sleeve annulus is not significantly.different from that in a crevice of an unsleeved tube, the upper limit of concentration of impurities remains essentially the same.

19. Dissolved solids in'the steam generator are not more prevalent because of the use of all-volatile water chemistry treatment. To the contrary, the objective of all-volatile chemistry is to minimize the concentration of solids in the steam generator, since solids are not added as in phosphate chemistry control. In any event, the feedwater portions of the secondary systems at Point Beach have utilized all-volatile control since initial start up, regardless of the steam

' generator chemistry control method.

20. Deposition of impurities has not, and is not expected i to, interfere significantly with eddy current inspectability.

Signals from conductive deposits such as copper or magnetite which might enter the annulus or be present on the tube are

l small, and can be eliminated by use of the same multi-frequency data processing techniques used for unsleeved steam generators.

Thus, the presence of impurities will not significantly affect eddy current inspectability of the sleeve and tube and, in any event, any effect will be no different than the effect on inspectabili.ty of unsleeved tubes.

21. Thus, for all the foregoing reasons, detection capability is not expected to be degraded over time as a consequence of sleeving, and therefore it is not believed that ,

sleeving will lead to primary-to-secondary or secondary-to-primary pathways which will exacerbate the problems postulated in Contentions 3(b), (c),. (d) and (e).

22. Steam generator tubes at Point Beach are inspected routinely, on a periodic basis, for tube degradation by standard eddy current inspection techniques. These inspection techniques enable the location of tube degradation with sufficient sensitivity to comply with the current NRC licensing requirements to identify and plug tubes exhibiting 40% or more wall penetration. Sleeving does not prevent inspection of the unsleeved portions of the tube above the sleeve in satisfaction of the NRC's 40% degradation limit.

23. After sleeve installation, each tube will be inspec-ted by eddy current techniques. This will serve two purposes.

The first is to verify the sleeve expansion and hard rolling of the upper and lower joints. The second purpose is to obtain a baseline inspection " signature" for each sleeved tube against

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which to evaluate subsequent eddy current inspections of the tubes.

24. The sleeves have been designed to span the lengths of the tube within and in the vicinity above the tube sheet.

These locations have been subject to corrosion from a caustic environment within the crevice between the tube and the tubesheet. After sleeving, the primary-secondary pressure boundary in this region is the portion of the sleeve below the upper tube-to-sleeve joint. In the tubesheet region, the eddy current inspectability of the sleeve is actually enhanced relative to the unsleeved tube; the unsleeved tube inspection produces a lower " signal-to-noise" ratio because the tubesheet is closer to the tube than it is to the sleeve.

25. At the upper joint, the sleeve wall has geometric transitions at the top and the bottom of the hydraulic expan-sion, and at the hard roll within the hydraulic expansion zone.

At these points, and at the upper tip of the sleeve, eddy current sensitivity is lessened. Degradation can be detected at these points using multi-frequency inspection with the

! standard bobbin-type eddy current probe, although the transi-tions make quantitative evaluation more complex. As stated in

! the Staff's SER, the amplitude of eddy current signals at these points was approximately 50% of the amplitude for non-sleeved tubes. However, the Westinghouse testing program has demon-l strated, by use of standird eddy current techniques, the l

-detectability of degradation which is smaller than that which

could cause a tube rupture during normal operation or postulated accidents.

26. In any event, Westinghouse has developed eddy current techniques which can be used, if necessary, to, gain additional information if degradation occurs in the transition areas. The use of eddy current probes consisting of " cross-wound" coils will provide inspectability of the transition regions compara-ble to that of the non-transition regions _using the standard bobbin-type probe. Sensitivity can be further enhanced by the use of multi-frequency data processing techniques.

27. Thus, eddy current inspection of sleeved tubes is

expected to provide adequate sensitivity for the detection of tube degradation before such degradation becomes a safety concern. Eddy current inspectability of sleeved tubes is sufficient to locate degradation with the potential for tube rupture during normal or postulated accident conditions.

28. The region of the tube where the upper joint is located has virtually been free of corrosion degradation in the past at Point Beach. In addition, the pre-sleeving eddy current inspection will be employed so that a joint will not be placed where degradation is indicated to exist. .Thu s , degrada-tion is not expected to occur at or in the immediate vicinity of the upper joint.

I 29. Even assuming, for the sake of argument, that degradation in the joint region would occur undet'ected, it would not be a significant safety concern. Assuming 100%

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through-wall degradation around the entire circumference at the joint, the presence of the sleeve would hold the tube in place.

Primary-to-secondary leakage would be linited by the small annulus between the sleeve and the tube to approximately 5% of the rate which would be expected from the unobstructed leak path of a double-ended break. This leakage could be detected by normal radiation monitoring systems and wo.uld allow for an orderly planned shutdown if technical specification limits were exceeded.

30. Taking all of the foregoing into consideration, sleeved tubes can be inspected, and it is concluded that sleeving does not increase the probability of tube rupture generally and does not increase this risk in the unconstrained free standing region of the steam generator tubes.

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- Contention 3(b)

The process of sleeving steam generator tubes increases the probability of tube failures generally, and, of even greater significance, it substantially increases the risk of failures in the unconstrained free standing region of the steam generator specifically in, among other things, the following manner:

(b) Annulus. The annulus between the original tube and the sleeve may give rise to a corrosive environment in the unconstrained free standing region of the steam generator in cases where the original tube is or may be suffering in the future from a through-wall crack permit-ting secondary water impurities (including copper and iron oxides from the feedwater heaters that are an unintended byproduce of the conversion to all volatile treatment) to seep into the narrow space and concentrate to eventually corrode the sleeve as well.

31. There is no mechanism by which secondary water impurities entering the annulus between the tube and the sleeve through a degraded tube could conceivably give rise to a corrosive environment in the unconstrained free standing region of the steam generator, i.e., the portion of the tube above the sleeve.

32. There is no known mechanism whereby such secondary water impurities entering the annulus would be concentrated to a greater extent, or produce a more corrosive environment, because the temperature in the tube-sleeve annulus is not significantly different from that in a crevice of an unsleeved tube.

33. In any event, the sleeving material to be utilized provides greater protection of the primary-secondary pressure boundary from a corrosive environment than the original tube material. Results of the Westinghouse corrosion testing

' program have demonstrated that, compared to the mill annealed Inconel 600 tube, the thermally treated Inconel 600 sleeve provides additional resistance to stress corrosion cracking by l

caustic impurities.

l l 34. Thus, the presence of an annulus in the tube-sleeve region is not expected to increase the probability of tube rupture generally and is not expected to increase the risk of l

rupture in the unconstrained free standing region of the steam generator tubes.

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Contention 3(c)

The process of sleeving steam generator tubes increases the probability of tube failures generally, and, of even greater significance, it substantially increases the risk of failures in the unconstrained free standing region of the steam generator specifically in, among other things, the following manner:

(c) Quality Assurance. The dependence on a large number of transient workers to install the sleeves will make it impossible to assume that the installation in the field matches the performance of test installations in the laboratory and will increase the probability of the kinds of problems indicated in [ Contentions 3(d) and 3(e)). .

35. Sleeving of the Point Beach steam generator tubes will be performed by the trained technicians and engineers who are employees of Westinghouse Electric Corporation, the contractor for the Point Beach sleeving program. They are assisted by trained temporary employees known as " channel head l workers."

36. Installation of nearly all of the sleeves, which includes decontamination, plug removal, inspections, honing of tubes to be sleeved, cleaning of tubes, insertion of sleeves, sleeve expansion, and hard rolling the joints is not done by channel head workers; these activities are performed remotely, from outside of the channel head, by the use of computerized automated equipment operated by the trained technicians and engineers who are employees of Westinghouse. A description of the procedures and equipment for the decontamination, sleeving, and inspection activities is contained in Sections 4, 5 and 7 of the Point Beach Steam Generator Sleeving Report.

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37. The duties of the channel head workers are primarily to install and remove the automated equipment within the channel head, perform manual sleeving operations for those tubes, if any, which are not accessible by the remotely operated equipment, and to perform the unskilled, non-sleeving activities which are required within the channel head. In no cases do the channel head workers perform any activities which require skill or discretion to effect proper installation of a sleeve, nor do they conduct inspections or make decisions or exercise judgment on the adequacy of the sleeving operations performed.

38. No work performed by channel head workers has any effect on the unconstrained free standing region of the steam generator.

39. To the extent that sleeving operations which cannot be performed remotely by Westinghouse technicians and engineers are done by channel head workers, all operations are performed with tools which are precalibrated. Thus, for example, the proper placement and applied forces for tube honing, sleeve insertion, sleeve expansion, and hard rolling of joints are

. automatically accomplished by the tools used, and do not require any judgment or discretion from the worker using the tools.

40. Examples of automatically controlled processes are the insertion and hydraulic expansion of the tubes. For most, if not all, of the tubes to be sleeved, the automatic, l .

computerized coordinator transport machine ("CTM") is used for sleeve insertion and expansion. The CTM is described in Section 5.0 of the Point Beach Sleeving Report. Sleeve ,

insertion and expansion with the CTM is described in Section 4.2.1 of the Sleeving Report. With the CTM, the sleeve must be fully inserted before the expansion step can take place. A limit switch must be engaged, and a pressure connection must be made, before the expansion process can begin; neither can occur unless the tube is fully inserted. For manual sleeve inser-tion, described in Section 4.2.2 of the Point Beach Sleeving Report, the sleeve must be fully inserted before a latch mechanism on the tubesheet can be engaged; the expansion tool cannot operate without engaging the latch mechanism.

41. In addition, each sieeve is visually inspected by qualified Westinghouse technicians or engineers by remote

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closed-circuit television.

42. Thus, the procedures, equipment used, and inspections provide that all sleeves are fully inserted within the tubes before they are held in place by sleeve expansion.

43. The expansion equipment, for both automatic and manual operations, is sized to limit the maximum: inner diameter

'of the sleeve. The expansion is determined by the dimensional limits of the precalibrated tooling. The dimensional limits and calibration of the tools, as defined by the installation proced~res u and process qualifications, are verified by Westinghouse Quality Assurance personnel. In addition,

.xpansion will also be verified by a sample inspection program, and by 100% eddy current inspection which is effective in determining the sleeve expansion.

44. Thus, the sleeving procedures, equipment limit settings, and inspections will provide that sleeves are neither over-expanded nor under-expanded.

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45. All work done by the channel head workers is observed by a trained Westinghouse technician (platform supervisor) who stands on the work platform at the entrance to the channel head. In addition, Westinghouse engineers in the on-site control trailer are in voice communication with both the platform supervisor and the channel head workers inside the channel head, and closely monitor the work through closed-circuit television.

46. All sleeving operations are performed in strict compliance with the Wisconsin Electric and Westinghouse quality assurance programs. In addition to the constant monitoring and control of all operations within the channel head during performance, each sleeved tube is eddy current inspected for proper sleeve expansion and hard rolling of the joints, a sleeve diameter inspection program is utilized, and all of the sleeved tubes are subjected to different'ial pressure tests.

l This total monitoring, control, inspection, and testing provides that all quality standards are met for the work performed by the channel head workers.

47. Channel head workers generally are not permanent employees of Westinghouse, the contractor which is performing the sleeving at Point Beach. The channel head workers are recruited either by Atlantic Nuclear Services (ANS), who acts as a recruiting contractor for Westinghouse, or by Westinghouse itself. Channel head workers recruited by ANS are not hired unless approved by Westinghouse. Applicants must pass both a mechanical and psychological aptitude test, and are screened for character, stability, and aptitude. Whether initially hired by ANS or Westinghouse, all channel head workers are trained and supervised solely by Westinghouse.

48. All channel head workars receive extensive training by Westinghouse prior to performing work on the job. This includes between 50 and 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> of classroom instruction, testing, and hands-on training in a full-scale mockup of the channel head. The channel head workers are trained and examined in all'of the sleeving activities, followed by a complete dress rehearsal in the channel head mockup. Channel head workers who do not perform ac quately during training, are not attentive, or exhibit a poor attitude are dismissed. Only those channel head workers who satisfactorily complete the training are permitted to work in the sleeving program.

49. Channel head workers will be under close supervision and observation by qualified Westinghouse training instructors or supervisors. The use of alcohol er drugs is strictly prohibited; channel head workers who violate this restriction are immediately dismissed. ,

50. Thus, considering the tasks performed by the channel head workers, the recruitment and training standards to be employed, the rigid on-the-job restrictions and close super-vision, the precalibrated tools and equipment used for the sleeving process, the close and continuous observation of the work performed by the channel head workers, and the 100%

inspection and testing of the sleeved tubes, all reasonable measures will have been taken to provide for the quality of the sleeving work to be performed by the channel head workers. The use of channel head workers in the sleeving program does not increase the probability of tube ruptures generally and does not increase the risk of ruptures in the unconstrained free standing region of the steam generator.

Contentions 3(d) and (e)

The process of sleeving steam generator tubes increases -

the probability of tube failures generally, and, of even greater significance, it substantially increases the risk of failures in the unconstrained free standing region of the steam generator specifically in, among other things, the following manner:

(d) Under Expanded Sleeve. In a LOCA accident condition which stresses the system and in which the suddenly depressurized primary system no longer acts'to compress together the aleeve and tube at the upper joint, an undetected insufficiently expanded sleeve, that may have been functional in normal operation, may leak and, if the original tube is also defected through-wall, form a secondary-to-primary pathway for in-leakage in excess of the allowable leak rates for model 44 steam genera-tors or otherwise sufficient to retard reflood

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of the core.

(e) Over Expanded Sleeve. If the reference upper joint is excessively expanded

and not detected in the sample verification process, the residual stresses in the transition zones will become more prone to degradation that can yield under the stress of a tube rupture event or accident conditions.

51. The sleeving procedures, the automatically calibrated equipment used in expanding the sleeve in the tube, and the observations and inspections conducted will provide that the sleeves will be neither under-expanded nor over-expanded.

52. Testing and analysis of the sleeve-tube assembly has dem6nstrated that it will withstand the stresses of normal operation and accident conditions, including the LOCA, and that the assembly meets the material stress limits of the ASME Boiler and Pressure Vessel Code. Sealing of the sleeve-tube assembly does not depend upon the primary side pressure within the tube, either for structural integrity or leak limiting capability at the joints. Primary side pressure within the sleeved tubes during normal operation has no effect on the physical leak limiting characteristics of the sleeve-tube i assembly. Thus, depressurization of the primary side during a l

l LOCA would have no significant effect on the leak tightness of l

the sleeve-tube assembly.

53. Hydrostatic leak tests conducted at the completion of the sleeving program will detect any leakage prior to resump-tion of operation. If, nevertheless, one were to postulate

( leakage in excess of allowable leak rates during subsequent operation due to an under-expanded sleeve installed in a .

l leaking tube, the leakage would be detected, and the tube would be repaired or plugged in accordance with the Technical Specifications of the Point Beach licenses. Leakage during a LOCA (even assuming both breach of the original tube and no expansion of the sleeve in the upper joint) would be small, in that it would be limited by the presence of the sleeve (i.e.,

the small tube-sleeve annulus) to no more than 5% of the flow which would occur through the maximum double-ended break, or about 12.5 gpm. As stated in paragraph 11 above, over 100 tubes would have to leak simultaneously in this manner to potentially affect the capability of the ECCS to cool the core.

It is not credible to postulate that post-sleeving hydrostatic tests, process control, sleeve joint diameter inspections, and eddy current inspections would not detect this number of tubes with under-expanded sleeve joints. Thus, taking all of the above into consideration, under-expansion of the sleeve would not create a credible threat to the ECCS performance.

I 54. Over-expansion of the sleeve would be unlikely to increase the likelihood of degradation. Reference to the Westinghouse test results presented in Figures 6.1-1 and 6.1-3 show that resistance of the thermally treated Inconel 600 .

sleeves to stress corrosion is not significantly affected by residual stresses due to sleeve expansion. In any event, the consequences of tube degradation at an over-expanded joint would be as discussed above in paragraph 53. There is no known mechanism by which under-expansion or over-expansion of the

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sleeve at the upper joint could significantly affect tube integrity in the unconstrained free standing region of the steam generator tubes.

55. Thus, under-expansion or over-expansion of the sleeve at the upper joint would not increase the probability of tube ruptures generally, or increase the risk of ruptures in the unconstrained free standing region of the steam generator tubes.

Contention 4 Pre-existing explosive plugs in tubes with through-wall defects, or which are incipient failures, may rock loose in the course of a LOCA accident condition providing a pathway for secondary-to-primary in-leakage, by itself or in combination

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with tube failure pathways, in excess of allowable leak rates for model 44 steam generators or otherwise sufficient to retard reflood of the core. ,

56. Pre-existing explosive plugs will not be removed, or in any way affected, during the sleeving program. The sleeving of steam generator tubes will in no way affect the leak tightness of pre-existing explosive plugs.

Contention 5 Loose parts left behind from steam generator repair work may impact upon and rupture tubes in the unconstrained free standing region, including the region where there is no double

' primary-to-secondary boundary of sleeve and tube during normal or accident conditions. This will increase the leakage rates which worsen the problem identified in [ Contentions 1 and 2].

57. Preparation of the steam generators for sleeving, installation of sleeves in the steam generator tubes and

post-sleeving inspections and tests are performed entirely from the channel head (primary side) of the steam generator. There are no sleeving operations which are performed in the secondary side of the steam generator. Thus, repair of the steam generators by sleeving will not result in loose parts or other debris in the tube bundle which could impact upon or rupture the tubes during subsequent operation.

58. In early 1982, a piece of wire was found inside one of the Unit 2 steam generators (unrelated to sleeving). That piece was removed, and there are no indications of loose or foreign objects within the Point Beach steam generators. In any event, even assuming the presence of a foreign object in the secondary side of the steam generator, the likelihood or degree of steam generator tube degradation by the foreign object would in no way be affected by the presence of the sleeve, which is inside the tube on the primary side.

Attachment to Affidavit of W. D. Fletcher STATEMENT OF QUALIFICATIONS AND EXPERIENCE W. D. Fletcher EXPERIENCE My name is W. D. Fletcher; I am presently Manager, Steam Generator Development and Performance Engineering in the Nuclear Technology Division of the Westinghouse Electric Corporation.

I graduated from Hardin-Simmons University in 1950 with a Bachelor degree in Chemistry a'nd from Fordham University in 1960 with a Masters degree in Chemistry.

I was employed with the Vitro Laboratories from 1951 to 1955, where I performed research on organo-phosphorus compound synthesis, reaction kinetics and mechanisms of organo-phosphorus compounds, phase studies, bench scale and pilot i

i plant production of organo-phosphites, high and low temperature kinetic studies of boron hydride synthesis, and electro-kinetic i studies of electrophoretic deposition of inorganic oxides in the manufacture of reactor fuel elements.

In 1957 I began my employment with Westinghouse and have been engaged in development work on the heterogeneous catalysis of reactions between hydrogen and oxygen produced through -

radiolysis of reactor coolants, reaction kinetics and I

mechanisms, catalyst development and evaluation in-high temperature and pressure aqueous solutions; evaluation and study of reactor coolant contaminants and means of coolant purification; study of behavior of fission and corrosion products in reactor coolants; in-pile studies of reactor coolants as pertains to chemical shim technology; reactor plant chemistry control, analyses, and data collection and inter-.

pretation of all operating reactor systems designed by Westinghouse.

Since 1970, I have been directly involved in development and design activities related to Westinghouse steam generators.

Under my direction, steam generator programs related to operations have been executed involving chemistry and materials as well as specific design configurations.

As Manager, Steam Generator Development and Performance Engineering, I am responsible for three design-development groups that involve steam generator thermal / hydraulics, advanced concepts design and analysis and design of field modification to steam generators.

I am a member of the American Chemical Society, the l National Association of Corrosion Engineers, the American

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l Nuclear Society, and the American Society of Mechanical Engineers.

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PUBLICATIONS

" Update of Operations with Westinghouse Steam Generators,"

American Nuclear Society, 1977, D.C. Malinowski and W.D. Fletcher.

" Operating Experience with Westinghouse Steam Generators,"

Nuclear Technology, 1975, W.D. Fletcher and D.C. Malinowski.

" Water Technology for Nuclear Power /PWR's," Industrial Water Engineering, 1971, W.D. Fletcher.

" Primary Coolant Chemistry of PWR's," W.D. Fletcher, the International Water Conference of the Engineers Society of Western Pennsylvania, Pittsburgh, October 1970.

" Post Accident Iodine Cleanup by Containment Filters and Sprays." Presentation at Tampa, Florida, May 21, 1968, J.D. McAdoo and W.D. Fletcher.

" Effects of Coolant Chemistry on Corrosion and Corrosion Products," W.D. Fletcher, Am. Nuc. Soc., Seattle, June 1969.

EURAEC-1972 (WCAP-3690-4) - " Description and Evaluation of the Boron Concentration Meter Utilized at the SENA (Franco-Belge)

Reactor Plant," January 1968, W.D. Fletcher.

WCAP-3269 "The Post-Irradiation Examination of Saxton Fuel Cladding Corrosion Products," March 1966, L.F. Picone and -

W.D. Fletcher.

WCAP-3269 " Fission Products fr'mo Fuel Defect Test at Saxton," April 1966, W.D. Fletcher and L.F. Picone.

WCAP-2964 - " Stability of Alkali in Reactor Coolant," 1964, W.D. Fletcher.

WCAP-2656 " Analysis of Fission Products in Saxton Primary Coolant," August 1964, W.D. Fletcher. -

" Water Technology of the Saxton Nuclear Experiment," Division of Water and Waste Chemistry, 4, 46 (1964), W.D. Fletcher and R.F. Swift.

" Flame Photometric Determination of Lithium Produced by B-10 (n,a) Li-7 to Mea'sure Boron-10 Burnup in Reactors Utilizing Chemical Shim Control." Presentation at Gatlinburg, Tenn., Oct.

6-8, 1964, B.D. LaMont and W.D. Fletcher.

- WCAP-3716 " Ion Exchange in Boric Acid Solutions with Radioactive Decay," November 1962, W.D. Fletcher.

WCAP-1689 Rev. -

"The Behavior of Stainless Steel Corrosion Products in High Temperature Boric Acid Solut' ions," May 1961, W.D. Fletcher, A. Krieg and P. Cohen.

WCAP-4097 " Inorganic Ion-Exchanger Materials for Water Purification in CVTR," August 1961 (CVNA-135), N. Michael, W.D. Fletcher, et al..

WCAP-3730 - " Interactions Between Stainless Steel Corrosion Products and Boric Acid Solutions," March 1960, W.D. Fletcher.

"Some Performance Characteristics of Zirconium Phosphate and Zirconium Oxide Ion Exchange Materials," Trans. Am. Nuc. Soc.,

3, 46 (1960), N. Michael and W.D. Fletcher.

WCAP-1206 - " Internal Recombination Catalyst Studies," May 4, 1959, W.D. Fletcher and D.E. Byrnes.

WCAP-1110 "A Semi-Flow System for the Study of Catalytic Combination of Hydrogen and Oxygen in Aqueous or Slurry System," February 1959, W.D. Fletcher and W.E. Foster.

" Electrophoretic Deposition of Metallic and Composite Coatings," Plating 42, 1255 (1955).

" Post LOCA Hydrogen Generation in PWR Containments," American Nuclear Society, W.D. Fletcher, M.J. Bell, R.T. Marchese, and J.L. Gallagher.

PATENTS U.S. Patent, "Information Storage Systems and Methods for Producing Same."

U.S. Patent, " Boron Concentration Meter."

U.S. Patent, " Electrophoretic Coating Dispersion Formulations."

l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

WISCONSIN ELECTRIC POWER COMPANY ) Docket Nos. 50-266

) 50-301 (PointBeachNuclearPlant, ) (0LAmendment)

Units 1 and 2) )

AFFIDAVIT OF W. D. FLETCHER County of Allegheny )

ss Commonwealth of Pennsylvania )

W. D. FLETCHER, being duly sworn according to law, deposes and states:

This information contained in the foregoing " Statement of W. D. Fletcher" is true and correct to the best of my knowledge and belief.

I f"f W. D. Fletcher Suliscribed and sworn to before me this # 4 day of August, 1982.

?7,e cenrLC '

Notar Public l ~ Nm " a. dnIAny Pusuc My Commission Expires: causcant ww. Attrenny can.

"'1 NNa UPl#f5 MAy 4. IMS U M*vivenia Associeth W hers .

(-

4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION t

Before the Atomic Safety and Licensing Board

(

In the Matter of )

)

WISCONSIN ELECTRIC POWER COMPANY ) Docket Nos. 50-266

) 50-301 (Point Beach Nuclear Plant, ) (OL Amendment)

Units 1 and 2) )

CERTIFICATE OF SERVICE This is to certify that copies of the foregoing

" Licensee's Response to Decade's Motion Concerning Litigable Issues," " Statement of W. D. Fletch - and " Affidavit of W. D. Fletcher" were served, by depvait in the U.S. Mail, first class, postage prepaid, to all those on the attached Service List, except those marked with an asterisk were served by hand delivery or by deposit with Federal Express, this 9th day of August, 1982.

~

{

, - o

.hceW Churchill Dated: August 9, 1982 P

e 2

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

WISCONSIN ELECTRIC POWER COMPANY ) Docket Nos. 50-266

) 50-301 (Point Beach Nuclear Plant, ) (OL Amendment)

Units 1 and 2) )

SERVICE LIST

• Peter B. Bloch, Chairman Stuart A. Treby, Esq.

Atomic Safety and Licensing Office of the Executive Board Panel Legal Director U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Wasington, D.C. 20555 Washington, D.C. 20555 CDr. Hugh C. Paxton

• Richard G. Bachmann, Esq.

1229 - 41st Street Office of the Executive Los Alamos, New Mexico 87544 Legal Director U.S. Nuclear Regulatory Commission

• Dr. Jerry R. Kline Wasington, D.C. 20555 Atomic Safety and Licensing Board Panel *Kathleen M. Falk, Esq.

U.S. duclear Regulatory Commission Wisconsin's Environmental Decade Washington, D.C. 20555 114 North Carroll Street Suite 208 Atomic Safety and Licensing Madison, Wisconsin 53703 Board Panel U.S. Nuclear Regulatory Commission Francis X. Davis, Esq.

Washington, D.C. 20555 Monroeville Nuclear Center Westinghouse Electric Corporation Atomic Safety and Licensing P. O. Box 355 Appeal Board Panel Pittsburgh, PA 15230 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Barton Z. Cowan, Esq.

John R. Kenrick, Esq.

Docketing and Service Section Eckert, Seamans, Cherin & Mellott Office of the Secretary Forty-Second Floor U.S. Nuclear Regulatory Commission 600 Grant Strcet Washington, D.C. 20555 - Pittsburgh, PA 15219 1