ML20084P172

Affidavit of Mj Hitchler Addressing Aspects of Joint Intervenors Contention 7 Re Safe Operation of Steam Generators.Strong Disagreement Expressed W/Basic Allegation of Contention.Related Correspondence

ML20084P172
Person / Time
Site:	Harris
Issue date:	05/05/1984
From:	Hitchler M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20084P102	List: ML20084P094 ML20084P120 ML20084P133 ML20084P149 ML20084P163 ML20084P172 ML20084P183 ML20084P188
References
OL, NUDOCS 8405170437
Download: ML20084P172 (17)

Text

,

e RELATED CORRESPONDENCE

' UNITED STATES OF AfER8CA NUCLEAR REGULATORY COMMISS10N BEFORE THE ATOMIC SAFETY AND LICENSING BOARD [hD In the Matter of 17 g,,

CAROLINA POWER & LIGHT COMPANY ) Docket Nos. 50 400f0L v and NORTH CAROLINA EASTERN ) 50-401 OLi J -

MUNICIPAL POWER AGENCY )

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(Shearon Harris Nuclear Power )

Plant, Units 1 and 2)

AFFIDAVIT OF MICHAEL J. HITCHLER County of Allegheny )

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Commonwealth of Pennsylvania )

MICHAEL J. HITCHLER, being duly sworn according to law, deposes and says as follows:

1. My name is MICHAEL JOHN HITCHLER. I am Manager of Plant Risk Analysis with the Nuclear Safety Department of Westinghouse Electric Corporation, P.O. Box 355, Pittsburgh, Pennsyfvania 15230.

2. I was graduated from Lowell Technological Institute in 1974 with _a Bachelor of Science Degree in Nuclear and Mechanical Engineering and from Carnegie-Mellon University in 1978 with a Master of Science Degree in Mechanical Engineering.

3. I have published five articles in various technical periodicals and have authored or coauthored eight Westinghouse reports which pertained to reactor accident analyses, emergency /abnonnal operating instruction development and probabilistic risk analyses.

4. I joined Westinghouse in June 1975 as an Engineer. I was promoted to Senior Engineer in December 1978. My responsibilities during that time included perfonning accident analyses for use in licensing kkD 0

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i documents. I have scrved as a Westinghousa liaison with the NRC, architect enginters and utilities for issuns concerning reactor protection system design requirements. My specific areas of specialization included core and systems response to transients initiated in the primary system, development of methodology for safety analysis of reload cores, and simulation of actual plant transients for computer verification purposes. I also had the lead responsibility for the transfer of the above technology to various utility customers. This responsibility included the structuring of classroom as well as on-the-job trair.ing for a number of utility personnel.

5. In June 1981, I was assigned responsibilities in the risk assessment area. These responsibilities involved the development and implementation of strategic programs to enhance and to apply risk assessment technology for use in nuclear power plant design and licensing. This work included development and quantification of event trees for use in reviewing emergency and abnormal operating procedures as part of the Westinghouse Owner's Group response to post TMI issues. I assisted in the development and review of Auxiliary Feedwater System Reliability Studies for three nuclear plants.

6. In October 1981, I was promoted to the position of Manager, Probabilistic Risk Assessment (PRA) Group. I presently have lead responsibility for a probabilistic risk study of two non-domestic, pre-construction nuclear stations,.which includes development of a risk baseline and an assessment of potential design alternatives. I have also worked on three domestic station risk studies, contributing extensively in the following areas: plant and containment event tree construction, systems success criteria.for fault tree development, external (seismic, wind, fire, etc.) event analysis and review of the results sections.

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7. I am a memb:r of the American Nuclear Society (ANS) and th2 American I

• Society of Mechanical Engineers. I served on two ANS Standards committees and contributed to several Atomic Industrial Forum (AIF) and Institute of Electrical and Electronics Engineers (IEEE) committees on development of risk criteria and utilization of PRA approach to licensing.

8. The purpose of my affidavit is to address those aspects of the Joint Intervenors' Contention VII concerning safe operation of steam generators. I strongly disagree with the basic allegation in Contention VII, based on the analysis which follows and my experience in other assessments.

9. Steam generator tube ruptures are defined as Condition IV events in the Harris FSAR (Section 15.0). The consequences of these events have been demonstrated to be acceptable in the FSAR and Environmental Impact Statement (EIS). Condition IV events are defined to have frequencies of less than once per plant lifetime (i.e.,

2.5x10-2/ year) . PWR performance to date with -233 years of experience is comparable (i.e., 2x10-2 er year). In the following paragraphs I have reviewed in detail this experience base which -

focuses on the various causes of steam generator tube degradation and the applicability of these causes to the Harris Plant design.

10. The total number of tube years of experience in Westinghouse-design plants with Inconel steam generator tubes was determined, as shown in Tables 1 through 6. .

11. The tables cover different categories of plants and set forth plant designation, number of tubes, date of connercial operation, total calendar years between beginning of commercial operation and July 1983, and tube years to July 1983. This data shows a total of over four million tube years of experience since beginning of connercial operation.

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12. Table 7 presents a list of tuba rupture events that have occurred in i Westinghouse steam generators. All five of' these events had flow rates large enough to cause plant trip and initiate safety inj ection. Only one event, however, had a flow rate equivalent to a full double-ended tube rupture as described in the FSAR; the other four events were much smaller in magnitude.

13. With five tube ruptures in an experience base of 3.6 x 105 tube years, the experienced tube rupture failure rate would be A = 5 + (3.6 x 106) = 1.4 x 10 6/ tube-year or, using Chi-square tables, the 50 percent confidence value would be 1

50 percent " c = 1.6 x 10-6/ tube-year 2 x 3.6 x 10 with upper and lower 95 percent confidence limits of 21.03 , ,

5.23 6 -

2 x 3.6 x 10 6 2 x 3.6 x 10 2.9 x 10-0 111 0.73 x 10-6 per tube-year

14. Based on this calculation, the tube failure rate derived from experience is 1.6 x 10'0/ tube-year and has about a factor of two uncertainty.

15. Because of advances in the state of the art in the design, operation, and inspection of steam generators, it is believed that experience in the Model O plants will provide additional margin to that observed to date. Cogent reasons can be given as to why certain of the five tube ruptures experienced should not occur in the Model O's since the operating conditions are not applicable, or why the occurrence rate should be substantially less because of such design and inspection advancements. These are described below.

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16. At Plant E in February 1975, phosphate tastage had thinned tubas in a zone just above the tubesheet where sludge had collected. In addition to thinning, some stress corrosion cracking was also present. The events at Plant I in September 1976, and Plant bb in June 1979, show some similarities. In both cases, the tubes had suffered stress corrosion cracking starting from the primary side. At Plant I, this was due to denting accompanied by " hour glassing" of the flow slots. At Plant bb, the affected tube had excessive ovality which led to high stresses at the U-bend. The two remaining events, at Plant N in October 1979, and Plant C in January 1982, were bocn due to foreign objects fretting and wearing the tube along one side.

17. Due to advances in the design of Model D steam generators and in maintenance procedures at Harris, these are judged to be reduced in frequency. The phosphate wastage, for example, has been l

eliminated since phosphates will not be used at Harris, thus the tube rupture frequency attributeo to wastage is judged to be lowered by at least a factor of 100. Reduction factors are utilized, since although phosphate wastage is impossible at Harris other types of chemical wastage (currently unobserved) may still be possible.

18. Denting of tubes, if it occurs at all, will develop much more slowly and with more limited extent than at previous stations due to:

- operation with only AVT chemistry control;

- reduction of copper in the secondary side systems, compared to other plants; d

- fresh water condenser cooling with resultant decrease in chloride concentrations as compared to plants operating on sea or brackish water.

19. Stress corrosion cracking (SCC) at Harris is judged very unlikely due to the following:

- minimization of copper which reduces the rate of SCC by minimizing the corrosive environment, i.e., reducing the concentration of alkaline salts; and

- design advances which (a) minimize crevices between the tube and tubesheet through full depth expansion of tubes and (b) provide features to reduce the accumulation of overlying sludge.

20. In addition, tube degradation will most likely be identified before rupture could occur due to extensive In-Service Inspection (ISI) which includes: eddy current testing, ultrasonic techniques, profilometry probes, full inspection of all tubes before the plant is put into operation, and continuous monitoring

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of water quality, radioactivity, leakage rates, etc. For these reasons, tube rupture due to denting and SCC is judged to be reduced by a factor of two.

21. One type of tube leakage event which is not affected by design advances is wear due to foreign objects, which was responsible for the two largest tube rupture events which have occurred. However, due to rigorous quality assurance procedures as well as monitoring for loose parts at Harris, this type of tube leakage event is judged to be much less,likely than historical frequency indicates, and a lowering by a factor of five is assumed in this study.

22. Implementation of the modifications to minimize tube vibration in the Model D steam generators as discussed in Mr. Timmon's affidavit should reduce tube vibration levels such that they will be at or below the levels contained in the experience base used in this analysis.

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23. Given the design and maintenance advances described above, the number of historical tube rupture incidents which are applicable to Harris for this analysis can be decreased from five to about l 1.5 (virtually none due to phosphate wastage, one due to denting and SCC, and 0.4 due to loose parts).

24. Table 8 shows how the 50 and 95 percent confidence level failure rate decreases as the number of tube ruptures in the experience base to the present decreases.

25. To account for tne above-mentioned abvances,1.5 rather than 5 ruptures in the experience to date will be assumed in this analysis.

26. On this basis, the median (50 percent confidence level) failure rate would be A 50 percent = 0.6 x 10-6/ tube-year with an uncertainty of less tnan a factor of 3. Although the above approach utilizes some engineering judgment in conjunction with the experience base, the data available and identified advances provide reasonable support for this. In fact, engineering i

judgment would suggest that the advances in the state of the art should yield an even lower f ailure rate.

27. This failure rate of 0.6 x 10'0/ tube-year corresponds to an annual frequency of 8.2 x 10~3 per year g0 .6 x 10-6 4578 tubes x 3 SG = 8.2year x 10 )

tube-year SG at Harris. This predicted value is significantly below the historical base. Thus the operation of Model 04 steam generators at Harris as compared with previous experience should result in an even higher degree of public safety with respect to these issues.

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23. Based on the above analysis and ray experience in other assessments, I am confident that the Model D-4 steara generators used at the Harris Plant are adequately designed and are capable of being operated in a safe canner.

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M6t/ b O idichaelJ.'itchler H

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Sworn to and subscribed before me this [ day of ////f' , 1984.

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tiotary Public fly Coirc.ission Expires:

LORRAINE W. PirLICA h0fARY PUBilC MONR0EVILLE 80R0. ALLEGHEhY COUNTY MY CON #15$10N EXPtRES OfC 14.1981 Neaster Pennsylvania Asso:sstion of Notasesen s

TABLE 1 STEAM GENERATOR TUSE EXPERIENCE U.S. WESTINGHOUSE INCONEL PLANTS No. of Comercial Tubes Operatjon Years Tube-Year Plant A 11,382 1/68 15.4 17.5 x 10 '

15,176 1/68 15.4 23.4 x 10 B

• 4 6,520 3/70 13.2 -

8.6 x 10 C *4 9,780 3/71 12.2 11.9 x 10 0 *4 6,520 12/70 12.5 8.2 x 10 E *#

6,520 10/72 10.7 7.0 x 10 F- *4 10,164 12/72 10.5 10.7 x 10 G *4 9.780 12/73 9.5 9.3 x 10 H *4 10,164 5/73 10.1 10.3 x 10 I *4 13,040 -

7/74 8.9 11.6 x 10 J

• K 13,552 10/73 9.7 13.1 x 10 *

• 4 L 9,780 ' 9/73 9.7 9.5 x 10

• 4 i

M 13.552 9/74 8.7 11.8 x 10

• 4 N 6,776 12/73 9.5 6.4 x 10

• 4 0 6,776 6/74 9.1 6.2 x 10

• 4 P 6,776 12/74 8.5 5.8 x 10

• 4 13,552 8/75 7.8 10.6 x 10 0 *4 R 13,552 5/76 7.1 9.6 x 10

• 4 5 13,040 8/76 6.8 8.9 x 10

• 4 7 10,164 4/77 6.2 6.3 x 10

• 4 U 13,552 6/77 6.0 8.1 x 10

• 4 V 10,164 12/77 5.5 5.6 x 10 W 13,552 7/78 4.9 6.6 x 10 X 10,164 6/78 5.0 5.1 x 10

• 4 Y

10,164 12/80 2.5 2.5 x 10 '

Z 13,552 7/81 1.9 2.6 x 10 l A1 13,552 10/81 1.7 2.3 x 10

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TABLE 1 (Continued)

STEAM GENERATOR TU8E EXPERIENCE U.S. WESTINGHOUSE INCONEL PLANTS No. of Connemial Plant Tubes Operation Years Tube-Year A2 10,164 7/81 1.9 1.9 x 10 '

A3 18,696 12/81 1.5 2.8 x 10 A4 13,552 6/82 1.0 1.4 x 10 '

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Total 233.4 245.6 x 10 Tube Years 5

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TA8LE 2 STEAM GENERATOR TUBE EXPERIENCE j

WESTINGHOUSE FOREIGN PLANTS (INCONEL) l No. of Commercial Plant Tubes Ooeration Years Tube-Year AA 2,604 8/69 13.8 3.6 x 10*4

• 4 88 5,208 12/69 13.5 7.0 x 10 CC 5,208 3/72 11.2 5.8 x 10*'

00 10,164 11/74 8.6 8.7 x 10 l

EE 10,164 5/75 8.1 8.2 x 10*4

• 4 FF G,776 4/78 5.2 3.5 x 10 GG 13,552 3/79 4.2 5.7 x 10*4

• 4 i HH 14,022 4/81 2.2 3.1 x 10 II 14.022 12/81 1.5 2.1 x 10*4

*4 JJ 9,156 12/81 1.5 1.4 x 10 l -

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. Total' 69.8 49.1 x 10' i

Tube Years i

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. TABLE 3 STEAM GENERATOR TUBE EXPERIENCE MHI PLANTS No. of Comen:ial Operation Years Tu be-Ye a r Plant Tubes

• 4 6,520 7/72 10 .9 7.1 x 10 l ZZ *#

10,164 11/75 7.6 7.7 x 10 YY

• 4 6,776 10 /7 5 7.7 5.2 x 10 XX *4 10,164 12/76 6.5 6.6 x 10 WW *4

! 6,776 9/77 5.7 3.9 x 10 VV 6,776 3/81 2.2 1.5 x 10*4 UU *4 6,776 3/82 1.2 0.8 x 10

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4 41.8 32.8 x 10

! Total Tube Years I

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4 TABLE 4 STEAM GENERATOR TUBE EXPERIENCE FRAMATO M PLANTS No. of Consen:ial Goeration Years Tube-Year Pl ant Tubes

• 4 10,164 12/77 5.5 5.6 x 10 a

10,164 3/78 5.2 5.3 x 10

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• 10,164 2/79 4.3 4.4 x 10 '

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10,164 2/79 4.3 4.4 x 10 d

10,164 7/79 3.9 4.0 x 10'4 e *4 10,164 12/79 3.5 3.6 x 10 f

10,164 11/80 2.6 2.6 x 10*4 g *4 2.5 2.5 x 10

[ h 10,164 10,164 12/80 9/80 2.7 2.7 x 10

• 4 i *4 12/80 2.5 2.5 x 10 j 10,164 -

• 4 10,164 12/80 2.5 2.5 x 10 k

• 10,164 6/81' 2.0 2.0 x 10

• 1 *4 10,164 5/81 2.1 2.1 x 10 m *4 10,164 2/81 2.3 2. 3 ), 10 n *4 10,164 5/81 2.1 2.1 x 10 o

' 10,164 12/82 1.5 1.5 x 10

p 10,164 10/81 1.7 1.7 x 10

q *4 10,164 11/81 1.6 1.6 x 10 r

10,164 12/81 1.5 1.5 x 10*'

s 10,164 11/82 0.6 0.6 x 10'4 t

4 54.9 55.5 x 10 Total Tube Years

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1 TABLE 5 STEAM GENERATOR TUBE EXPERIENCE MISCELLANEOUS WESTINGHOUSE LICENSEE PLANTS ACEC0 KEN No. of Cosmercial Operation Years Tube-Year Plant Tubes i i

• 4 aa 6,520 2/75 8.2 5.3 x 10 bb 6,520 11/75 7.6 5.0 x 10 ACLF

• 4 cc 10,164 9/75 7.7 7.8 x 10 4

Total _

23.5 18.1 x 10 Tube Years

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SUMMARY

OF STEAM GENERATOR TUBE EXPERIENCE

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No. of Plants Plant-Years Tu be-Ye a rs t 7 ,

Westingnouse (Inconel Tube) ,<. .

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US p1 ants 31 233.4 2,456,000 \'"

69.9 491,000 Foreign plants 10 l  !

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Subtotal 41 30's. i' , 2,947,000 4

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Westinghouse Licensee plants  !, ,

7 41.8 328,000 MHI 54.9 Sh5,000 FRA 20 t23.5 181,00J ACECOWEN, ACLF J +

[, ,, 1 Subtotal 30 $ 120.2 .- 1,064,000 ,

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.i TOTAL

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. 71 42!.4 M 4,011,000 1.

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TUBE RUPTURE EXPERIENCES

SUMMARY

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s 4 jt s g Occurrence Estituated Date plant Attributed Cause Leak Rate

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' g %\ l, g Feb. 26, 1975 E' [ Phosphate Wastage + SCC 125 gpu (1 N2 i Sept. 15, 1976 I Denting + SCC 80 gpm (1 June 25, 1979 135 gpm (1

, ( 3 ' bb Ovality + SCC , ,,

i' 4 Oct. 2,1979 N Loose part (spring) 390 gpm (1 5 Jan. 25, 1982 C Loose part (plate) 634 gpm (!

(k.- Ref.

1. NUREG-0651, Evaluation of Steam Generator Tube Rupture Events, USNRC, Appendices C and H, March 1980.  ?,

2. Response to Long Term Comitmerts, Ginna Restart SER, Steam Generator Tube Rupture Incident,' November 22, 1982 Attachment B, Analysis of Plant Response Ouring Januar.

25, 1982, Steam Generator Tube Failure at the R. E. Ginna Nuclear Power Plant'.

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? TABLE 8 SENSITIVITY OF TUBE FAILURE RATE,TO NUMBER OF FAILURES EXPERIENCED Assumed No. of Failures Corresponding Failure Rate 6

Experienced in 3.6 x 10 at Indicated Conficence Level 95 percent Tube Years of Operation 50 percent 5 1.6 x 10'0/ Tube Year 2. 9 x 10 -6/ Tube Year 4 1.2 x 10-6 2.5 x 10-6 3 1.0 x 10-6 2.2 x'10-6 2 0.74 x 10-6 1.8 x 10-6 1.5 0.60 x 10-6 1.5 x 10-6 1 0.47 x 10-6 1.3 x 10-6 0 0.19 x 10-6 0.83 x 10-6 e

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