ML20079M955

Testimony of Mj Hitchler Re Quantitative Assessment of Steam Generator Tube Integrity.Steam Generator Tube Degradation Under Normal Operating & Accident Conditions Is Not Safety Concern

ML20079M955
Person / Time
Site:	Byron
Issue date:	02/28/1983
From:	Hitchler M COMMONWEALTH EDISON CO.
To:
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ML20079M929	List: ML20079M926 ML20079M955
References
ISSUANCES-OLA, NUDOCS 8303030368
Download: ML20079M955 (18)

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DOCKETED U.;W f' UNITED STATES OF AMERICA NUCLEAR REGULATORY) CgiIpSI,9N59 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

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COMMONWEALTH EDISON COMPANY ) Docket No. 50-454-OLA

) 50-455-OLA (Byron Station, Units 1 and 2) )

a TESTIMONY OF MICHAEL J. HITCHLER CONCERNING STEAM GENERATOR TUBE INTEGRITY (QUANTITATIVE ASSESSMENT)

Submitted on behalf of the Applicant, Commonwealth Edison Company in Response to DAARE/ SAFE Contention 9c and League Contention 22 February 28, 1983 n.-

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

COMMONWEALTH EDISON COMPANY ) Docket No. 50-454-OLA

) 50-455-OLA (Byron Station, Units 1 and 2) )

SUMMARY

m The summary of Mr. Michael J. Hitchler sets forth his professional qualifications and provides a quantitative assessment of the frequency of steam generator tube ruptures under normal operating conditions and accident and transient conditions, such as main steam line breaks, loss-of-coolant-accidents'and other events. Mr. Hitchler finds.that the frequency of tube ruptures in such circumstances is sufficiently low that these accidents and other events coincident with tube ruptures need not be considered further in the design of the Byron Station. Mr. Hitchler concludes by confirming Mr. Fletcher's opinion that steam generator tube degradation under normal operating and accident conditions is not a safety concern.

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

)

COMMONWEALTH EDISON COMPANY ) Docket No. 50-454-OLA

) 50-455-OLA (Byron Station, Units 1 and 3) )

TESTIMONY OF MICHAEL J. HITCHLER CONCERNING STEAM GENERATOR TUBE INTEGRITY (QUANTITATIVE ASSESSMENT) 0.1. State your name, address and occupation.

A.l. My name is Michael John Hitchler. I am Manager of Probabilistic Risk Assessment with the Nuclear Safety Department of Westinghouse Electric Corporation, P.O. Box 355, Pittsburgh, Pennsylvania 15230.

Q.2. Please state your educational background.

A.2. I 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.

i I have published over five articles in various technical periodicals and have authored or co-

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authored over eight Westinghouse reports, which g, y

pertained to reactor accident analyses, emergency / - '

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abnormal operating instruction' development and probabilistic risk analyses'. ,

J r, h l Q.3. Please state your professional work experience with si; s \

Westinghouse, , 's , ,

.'i A.3. I joined Westinghouse in June 1975 as dn Engineer.'

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I was promoted to Senior Engineer in' December'1978. .

My responsibilities during that time included #

performing accident analysis for accidents used in licensing dccuments. I have served ~as a Westinghouse interface with the NRC, architect engineers and utilities for . issues concerning ~~

reactor protection system design requirements. .,

Specific areas of specialization include core and 1

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 ,

j purposes, included was the lead responsibility for the transfer of the above technology to various

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utility customers. This included the structuring of - .

i l classroom as well as on-the-job training for a l

number of utility personnel.

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j 'In June 1981, I was assigned responsibilities in the

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, risk assessment area.

These responsibilities tn-

,, involved the development and implementation of strategkeprogramstoenhanceandtoapplyrisk-

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assessmen't technology for use in nuclear power plant design and licensing. .This work included

> ,, develophent;and quantification of event trees for

use in revi' ewing emergency and abnormal operating

' procedures as part of the' Westinghouse Owner's' Group a r e response to post TMI issues. I assisted in the

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development and review of Auxiliary Feedwater System

, . , , Reliability Studies for three nuclear plants.

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'In October 1981, I was promoted to the position of a

Manager,'Probabilistic Risk Assessment Group. I presently have lead responsibility for a I probab listic risk study of the Sizewell B. (British National Nuclear Corporatica) Nuclear Station, which c 4 include development of a risk baseline and an assessment of potential design alternatives. I have

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s also worked on the Zion and Indian Point risk studies, contributing extensively in the following are,as :' plant and containment, event tree co struction, systems success criteria for fault tree ' dev'elopment , external (seismic, wind, fire,

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4-etc.) event analysis and review of the ramification sections.

I am a member of the American Nuclear Society and the American' Society of Mechanical Engineers. I served on two AES Standards committees and contributed to several AIF and IEEE committees on development of Risk. Criteria and Utilization of PRA

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Approach to licensing. I am currently a member of 1

the PRA Fethodology Procedures Handbook Committee.

Q.4.- What is the purpose of your testimony?

A.4. The purposesof my testimony is to address those aspects of thle Rockford League of Women Voters' c, , Contention 22 and DAARE/ SAFE Contention 9c concerning i steam generator tube ruptures both during

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..s' a .main steam line breaks (MSLB) and loss-of-

"l [ coolarit-accidents (LCCA). Specifically, I provide a

/ y v quant'itative assessment of the frequency of these

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l Q.5. Please de, scribe the nature of your assessment.

A.5. I have supervised and participated in the develop-ment of'a model', which is summarized in Attachment

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generator tube rupture frequencies based on the pressure differentials across the steam generator tubes. The model is conservative because it takes no credit for recent modifications in steam generator design or operations. Rather it conserva-tively assumes that steam generator tube degradation rates in the future will be the same as over the previous 2.5 million tube years of operating expe-rience. Also the model assumes that the five major i tube leaks which have occurred were full tube ruptures when in fact a full tube rupture has never occurred. Most of these events had less than 20% of the leak rate of full tube ruptures.

Q.6. What are the results of your assessment?

A.6. The results of this assessment are that single and multiple tube ruptures as initiating events are

-2 predicted to occut at frequencies of 3 x 10 and

-5 The frequency of 3 x 10 per year respectively.

multiple tube ruptures combined or as a consequence

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of (i) large LOCA events is 5 x 10 , and (ii) small LOCA and transient events with normal pressure

-5 differentials is 2 x 10 . The frequency of multiple tube ruptures combined or as a consequence

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. of steam / feed line break events is 3 x 10 ,

Q.7. What conclusions, if any, have you drawn from the results of your assessment?

A.7. The frequency of postulated tube ruptures combined or as a consequence of transient conditions and accident conditions such as MLSB's and LOCA's are extremely low over the 40 year's of plant operation.

Therefore, I would expect the PWR operating experience (presently 350 years) of no steam generator tube ruptures occurring as a result of steam-line breaks, LOCA's or other events, to continue. Finally, I conclude that these events are of sufficiently low frequency that they need not be considered further in the Byron Station design.

Q.8. What is the basis for your opinion that the steam generator tube-rupture frequencies from the accidents and transients discussed above are sufficiently low a exclude further consideration in the Byron Statir design?

A.8. A precise nume al definition for what events are beyond the desi,n basis has not been made. However, numerous events have been defined as being at the design basis limit. From these can be inferred a frequency range below which events need not be considered.

The SSE (Safe Shutdown Earthquakes) is an example.

For the Byron units, the SSE is defined as a 0.2g ground acceleration earthquake. Most nuclear plants are designed near this value. This seismic design 4

basis takes into account those events with recurrence frequencies of one every 1,000 to 10,000 years (10 -3 to 10-4). Probabilistic quantifications utilizing techniques similar to Attachment A (for example, WASII-1400 and the Zion and Indian Point Probabilistic Safety Studies) conservatively show other postulated events, such as, large and medium size LOCA's and steam line breaks to have recurrance intervals of approximately 1000 years per reactor.

Another approach to defining frequency limits for events beyond the design basis is the proposed NRC Safety Goal. The safety goal recently issued by the USNRC for trial use is also consistent with the 10-

-4 to 10 range of recurrence. The severe core damage frequency limit is 10 -4 per year or a recurrence intervpl of once every 10,000 years, and it represents a summatic.t of the frequencies of all scenarios that could lead to severe core damage.

Using the above values as a baseline, the frequency of events described in my answer to question 5. are

beyond the design envelope established by the nuclear industry and the NRC over the years. This does not mean that these events will rever happen, but rather that their frequency of occurrence are sufficiently small relative to the operating life of the units to be cor.sidered incredible.

Moreover, tube rupture events are predicted to result in severe core damage at frequencies of 10-per year for the Byron Station. Stated differently, the likelihood.of a severe core melt occurring at the Byron Station from tube rupture events is one-in-a-billion per year, an extremely low value, well below NRC's proposed safety goal, and a negligible contributor to meeting the goal.

Q.9. Have you read Mr.-Fletcher's testimony in this case?

A.9. Yes.

Q.10. How does your conclusion concerning the consideration of steam generator tube ruptures compare with his assessment?

A.10. My conclusion is consistent with that drawn by Mr.

Fletcher. Mr. Fletcher, using traditional deterministic methodology, concludes that steam generator tubes should not rupture as a result of a

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MSLB or LOCA because the various meastres identified in his testimony assure that steam ge:1erator tube integrity will be maintained. My analysis does not take credit for recent modifications in steam generator design or operation to minimize tube degradation, but it nevertheless shows that such accidents are very low frequency events. Therefore, if the modifications in steam generator design and operation discussed by Mr. Fletcher and the other panel witnesses were taken into account, my frequency values would be even lower. I can only conclude that.my assessment, conservative as it is, confirms Mr. Fletcher's opinion that steam generator tube degradation under normal operating and accident conditions is not a safety concern.

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SGTR PROBABlLXTY MODEL

1. TUBE RUPTURE FREQUENCY The experienced frequency of tube rupture events is about x =

2x10-0 per tube year of operation. This is cased on 5 events in about 2.5 million tube years experience. There are 4578 tubes in each of the four Byron steam generators. The frequency of tube ruptured based on this historical data would then be 0.037 per year of plant operation. This should be a conservative estimate Decause improvements in operation and design as a result of knowledge gained from the experience to date would be expected to reduce the frequency of tube' rupture in the future. In this study, the conservative value of U.037 ruptures per year will be employed.

This approach is taken for two reasons. First of all', this study is attempting to ascertain the effects of tube rupture on risk. The use of a conservative value helps .to set an upper bound on the tube rupture risk impact while avoiding potential concerns about the validity of a more realistic estimate. Secondly, the use of a more realistic value would involve a " fine tuning" of the data and analyses that would be inconsistent with the scope depth of this study.

2. TUBE RUPTURE PROBA8ILITY AS A RESULT OF PRESSURIZATION LUA0

. The x = 2x10-6 per tube year rupture frequency identified above is the frequency of degradation to the extent of rupture under the normal operation tube differential pressure load in the range of 1250 psi . The frequency of degradation to the extent of rupture under faulted condition loads sucn as the approximately 2650 psi secondary pressurization accident differential pressure would.have to be of this magnitude also. It is assumed that for a tube that does degrade to this extent, it may take anywhere from 0 to 40 years of operation with equal probability. The time that a degrading tube spends in the 2650-1250 psi capability range is thus estimated to be (sec Figure 1)

2650-1250 t* *E10,000-1260] t = 0.16t

'n tnis expression,10,000 psi is the minimum virgin tube ourst capability caseo upon laboratory testing, and t is the time for a degrading tube to cegrade to 1250 psi capaoility wnicn is assumed to De a uniform distribution from J to 40 years of service life. Ine average value of t would be 20 years.

uiven an accidental secondary depressurization event, then the probability that a tube exposed to 2650 psi differential pressure would rupture is (1) p = A t* = 0.16xt per tube The steambreak/feedbreak high pressure differentials are applied to only one of the steam generators should uncontrolleo blowdown occur. Based on this and the assumption that each tube's failure probability is random and independent, the probability of various numbers of tubes rupturing can be evaluated from the binomial distribution (2) P(r) =

r'(n r)' Y" * (1-P}" "

where n = no. of tubes in one SG = 4578 r = no. of tubes rupturing, i.e.1 or 2 or J . ..

p = probability of individual tube failure from Eq. (1)

P,( r) = probability of r tubes failing.

This model gives the following results for 1, 2, or 3 tubes rupturing from FL8/SLB loads.

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Probability That steam dreak Event l'ould Causs

-', Tune Rupture YEARS OF PLANT OPERATION No. of Tuoes Ruoturin9 t = 1 yr. t = 20 y r. t = 40 yr. AVG Proc. of one tuce rupture 1.4 x 10-3 2.8 x 10-2 5.6 x 10-2 3 x 10-2 Proo. of two tuce ruptures 1.1 x 10-6 4.2 x 10-4 1.7 x 10-3 6 x 10-4 Proo. of three tuce ruptures b x 10-10 3.9 x 10-0 J.2 x 10-5 g x 10-6 The uncertainty range on 1 is about a factor of 2.5. Allowing an additional factor of 2 to cover uncertainties in the Figure 1 model and tne linear degradation assumption, the 90 percent uncertainty range for p in equation (1) would be a factor of aoout 5. Based on this, the uncertainties on the results for one, two and three tubes rupturing is estimated to be factors of 5, 25, and 125 respectively.

This same approacn can be used to estimate the consequential tube rupture probability for other events in which the differential pressures experienced are greater than for nonnal operation but less than those for severe faulted

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events as above. This includes ATWS where all four steam generators are subjected to lower tube differential pressures. Extensive testing has snown that the strength requirement to guard against tube rupture during a postu-1ated feedwater 1ine or steam 1ine break (FLB/SLB) is always acre-1imiting in tenns of tube wall thickness than for tube collapse for a postulated LOCA.

For this reason the FLB/SLB tube rupture probabilities can be conservatively employed for a LOCA, although it should be realized that the loading in case of a LOCA would be in a direction to seal off leaks or collaspe tubes rather than cause a large leakage path.

3. TUBE RUPTURE LEAR RATE PROBABILITY DISTRiduTION Table 1 summarizes the 5 tube rupture events experienced to date and gives the estimated initial leak rates. Figure 2 provides the probability of

_ exceeding various leak rates based on this experience. It is noted that in only 20 percent of the cases are the leakages as high as the greater

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. than 600 GPM ranga typical of 4 double ended rupture. Tho avsrage has

. btea 273 GPM or less dian half. Furthermore, the experience has b :n on 0.875 incn u0 tubes as opposed to the 0.75 inch UD tubes on the dyron 04 steam geneator. Based on tnis, the leak rates of Figure 2 would be expected to De scaled down by the square of the diameters or a factor of u.735. On this basis, assumption of a 600 GPM leak would already corres-pano to practically 2 to 3 tunes of average expectea leakage. Therefore, it would be appropriate to comoine a distribution of the type shown in Figure 2 with multiple tube rupture frequencies to obtain more realistic frequency vs leak rate relations.

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, FIGURE 1 MODEL FOR PROBABILITY OF TUBE RUPTURE J LLOAD INCREASE 10,000 7------L = Initial tuce pressure capability = 10,000 psi g

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6 8,000 -

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Assumed path of tube that would degrade to failure in

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g t years of operation a

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L slb = Capability of tube failing under FLB/SLB load = 2650 psi E 2,000 -

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Ly--L" = Capability of tube that fails 1,000 -

t  % 'under normal operating load

= 1250 psi 0

10 20 30 40 YEARS OF SERVICE l

A = Frequency of severe degradation to rupture (per tube year) = 2 X 10-6 per year per tube l t = Time to fail under nonnal load (assumed random over period 0 to 40 years) t* = Time vulnerable to feedbrwak/steambreak load (years)

'slb ~ 'no t* = t ( )

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!. Lslb - L no i p = Probability of failure given FLB/SLB loads = At* = At ( )

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. FIGURE 2 PROBABILITY OF EXCEEDING VARIOUS INITIAL LEAK RATES GIVEN A STEAM GENERATOR TUBE RUPTURE

1.0 BASIS

Leak rate estimates from five tube rupture events that have occurred 0.9 -

to December 1982 (see Table 1) 0.8 -

x al 0.7 _ tVENT LEAK RATE o

0.6 1 125 GPM 2! _ _

a 2 80 j 0.5 -

3 135 4 390 3! 0.4 -

5 634 S 0.3 - -

@ AVG 273 a = 235 0.2 -

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, , , , , , i 0 100 200 300 400 500 600 700 LEAK RATE (GPM)

NOTE: The above experience has been with 0.875 inch OD tubes. For 0.75 inch OD Byron tubes the above should be apportioned down by approximately a factor of

= 0.735

( 0.875 )

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TABLE 1 ,

TUBE RUPTURE EXPERIENCES

SUMMARY

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Occurrence NO. Date P iu.. .'startup date) Attributed Cause Estimated Leak Rate 1 Feb. 26,1975 Point Beach 1 (Oct. 70) Phosphate Wastage + SCC 125 gpm III

, 2 Sept. 15, 1976 Surry 2 (Jan. 73) Denting + SCC 80 gpm III 3 June 25, 1979 Doel 2 (June 75) Ovality + SCC 135 apm III l 4 Oct. 2,1979 Prarie Island 1 (Aug. 73) Loose part (spring) 390 gpm III 5 Jan. 25, 1982 Ginna (Sept. 69) Loose part (plate) 634 gpm(2) i W

l Raf.

1. NUREG-0651, Evaluation of Steam Generator Tube Rupture Events, USNRC, Appendices Card H,11 arch 1980

2. Response to Long Term Commitments, Ginna Restart SER, Steam Generator Tube Rupture incident, November 22, 1982 Attachment 8, Analysis of Plant Response During January 25, 1982 Steam Generator Tube Failure at the R.E. Ginna Nuclear Power Plant. ,

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