ML19296C935
| ML19296C935 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/19/1980 |
| From: | Boyle M, Campe K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19296C930 | List: |
| References | |
| NUDOCS 8002290087 | |
| Download: ML19296C935 (21) | |
Text
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G ATTACHMENT 1 m
8 00229 0Qg
ATTACHMENT 1 EFFECT OF RECENT WESTINGHOUSE TURBINE CRACKING EXPERIENCE ON ESTIMATED HISTORICAL TURBINE FAILURE RATE By K. Campe and M. Boyle The Staff evaluation of the risks associated with potential turbine missiles for North Anna Units 1 and 2 was based, in part, on the estimated frequency of the expected failure rate of turbine disks by brittle fracture. A failure rate of 6 X 10" per turbine year was used in the Staff's analyses.
This rate is derived from considering the observed number of turbine failures (with missile ejection) and the total number of turbine years of operation.
It was not derived from an application of identifiable failure mechanisms such as metallurgical or mechanical disk properties or crack formation and growth due to environmental conditions involving temperature, stress, and corrosion.
It should be noted, however, that the above environmental failure causes were considered by the Staff when describing the conservatisms asso-ciated with the use of historical turbine failure rate.
The new turbine information stemming from inspection of Westinghouse turbine disks and disc 0very of stress corrosion cracking has the potential for increasing the susceptibility of the disks to brittle fracture failure during operation.
A logical conclusion that can be drawn is that continued operation of a disk with cracks would lead to disk failure within a time
2-period characterized by the initial crack size and the crack growth rate associated with the disk material and operational environment. Consequently, if all Westinghouse units suspected of cracked disks were pemitted to
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operate indefinitely without any corrective measures, the failures occurring presumably within the next two or three years would affect the current estimate of the average turbine failure rate by brittle fracture.
- However, this is not the case, as evidenced by the Staff activities currently being pursued with respect to all operating reactors, including North Anna Unit 1.
The inservice inspection, the crack growth rate estimation, and the correc-tive measures being taken with respect to defective disks, give assurance that the conservatism associated with the use of the previously estimated turbine failure rate is not invalidated. The following considerations support this view:
a.
Current inservice inspe: tion intervals are based on crack growth rates which incorporate the recent stress corrosion cracking experience. These intervals are much shorter than those that were recommended at the time when the Staff did the turbine missile risk analysis for North Anna Units 1 and 2.
b.
The ability to detect key way cracks in the field, as recently developed by Westinghouse, permits a more thorough inservice inspection than previously possible.
8 c.
Corrective measures taken by the utilities with respect to defec-tive disks returns the rotor integrity to an acceptable level until the subsequent inservice inspection.
In view of the above, the Staff believes that the continued use of the historical turbine failure rate in nuclear plant risk assessments, including North Anna Units 1 and 2, is justified.
II.
EFFECT OF RECENT WESTINGHOUSE TURBINE CRACKING EXPERIENCE ON FACTORS OF IMPROVEMENT IN P y In "NRC Staff Testimony Regarding Turbine Missiles" (dated April 27,1979)a factor of improvement in the probability of disk failure and missile ejection, P, at North Anna Units 1 and 2 was presented.M The P failure probability 3
y is made up of two distinct modes (pp. 4-5), the design overspeed failure and the destructive overspeed failure. The recent Westinghouse cracking experience does not have an effect on the destructive overspeed failure probability since this is a failure mechanism that is not material dependent, rather it occurs because of a component or system failure. Therefore, the factors of improvement in the value of the destructive overspeed failure probability presented in our earlier testimony (pp.19-20) are still valid. However, the contribution of the design overspeed failure to the overall value of P 3
may be affected by the recent Westinghouse operating experience.
If All further page references are to this April 27, 1979 testimony.
, The factor of improvement in the design overspeed failure probability con-sists of two parts, which are:
(1) improvement due to an increase in material fracture toughness (pp. 28-33) and (2) improvement due to preservice inspection (pp.34-35).
In the discussion of the factor of improvement due to materials toughness, it was assumed that crack growth was due to stress corrosion assisted fatigue.
However, it appears that the crack growth mechanism that has actually been found in Westinghouse turbine disks is primarily stress corrosion cracking.
The factor of improvement in the design overspeed failure probability was determ'ned by taking the ratio of the density function of cracks that could grow tc critical size in two materials with differing materials toughness (p.31). The initial crack size was calculated from stress corrosion assisted fatigue loading over the expected life of the turbine. This same calculation of the initial flaw size necessary for a failure to occur may be made for any assumed crack growth mechanism and for any period of time.
Or, stated in another way, the factor of improvement that was calculated in the design overspeed failure probability does not rely on the method of crack growth.
Rather, over a given period (load cycles or time), it will depend upon the disk material fracture toughness level.
The factor of improvement due to preservice inspection (pp. 34-35) will not be affected by the recent cracking experience, since the cracks appear to be service induced flaws.
B Other factors of improvement in the design overspeed failure probability were presented in our previous testimony in a nonquantitative manner (pp. 35-40).
Of these additional factors, only inservice inspection of the turbine disk fill change.
Westinghouse has instituted an inservice inspection technique, which is based on ultrasonic non-destructive examination of the disk bore and keyway, that will affect our previous testimony (pp. 38-40). After a given period of turbine service, this technique is used to detect flaws in the disk bore and keyway design. The technique is capable of detecting flaws much smaller than critical size with good reliability, thus allowing the affected material to be repaired, replaced, or returned to service with a known flaw for a given period of time, as determined from a failure evaluation.
When this technique is implemented in a permanent periodic manner on all nuclear turbines, it will have the effect of reducing the value of P, since 1
the turbine material degradation, if any, will be monitored and corrected throughout the proposed forty year life of the plant.
In our previous testimony, we assumed that if a crack was present in the turbine disk, it will never be detected and repaired.
On pages 20 and 40 of our previous testimony, we made conclusions about the improvements in the value of P for North Anna Units 1 and 2.
In our judg-3 ment, based upon what is presently known about the Westinghouse turbine disk cracking experience, the conclusions reached in our previous testimony are
. still vaiid. The absolute value of the calculated factor of improvement due to increased material fracture toughness may have changed due to recent operating experience, but the magnitude of the value should be at least as great as previously calculated.
In view of the above, we believe that the recent operating experience of Westinghouse turbines will not appreciably change our estimate of turbine missile risks for North Anna Units 1 and 2.
WESTINGHOUSE REVISION OF TURBINE MISSILE EXIT ENERGY ESTIMATES Westinghouse has indicated that the results of recent scale model tests with turbine disk ruptures lead to some revisions of the estimated disk fragment exit energies for their turbines.
The impact of this on the Staff's evalua-tion of turbine missile risks is believed to be negligible for the following reasons:
a.
Westinghouse scale model tests were conducted to determine the effects of non-symmetric impacts of disk fragments on turbine blade stator rings. This effect was anticipated by the Staff during its evaluation of North Anna Units 1 and 2, and prior to the Westinghouse tests. The Staff missile energy estimates were based on the assumption of non-symmetric impacts.
, b. ' Staff evaluation of North Anna Units 1 and 2 was based on neglect-ing the presence of plant structures and barriers.
Hence if there were some changes in the estimated missile energies, this would not affect significantly the risk evaluation.
In view of the above, we do not believe that the revised missile energies will change our estimate of the turbine missile risks for North Anna Units 1 and 2.
CONCLUSIONS For the reasons indicated above, we do not believe that the recent informa-tion regarding Westinghouse turbines has any substantial effect on our previous testimony discussing turbine missile risks for North Anna Units 1 and 2.
Nevertheless, we are in the process of reviewing the Westinghouse information as it becomes available.
If additional findings relevant to North Anna Units 1 and 2 are made, we will notify and inform the Board of their nature.
4 ATTACHMENT 2 M
STAFF'S BASES FOR THE CONTINUED OPERATION 0F NORTH ANNA UNIT 1 During November,1979, the NRC became aware of a problem of stress corrosion cracking in Westinghouse turbines. Meetings were held with Westinghouse to ascertain the probable extent and severity of the problem.
Westinghouse was recommending early inspection of turbines that had long times, and particularly those machines with discs of marginal material properties or history of sceondary water of steam chemistry problems. Since then, inspec-tions have been performed on about eight more Westinghouse turbines, with indications of cracking, some severe, found in most of them.
Investigations are continuing.
Westinghouse has developed a procedure to predict what the maximum crack size in a turbine would be expected to be. This procedure is based on evaluating all of the cracks found to date in Westinghouse turbines, past nistory of similar turbine disc cracking, and results of laboratory tests. This prediction method takes into account two main parameters; the yield strength (and stress) of the disc, and the temnperature of the disc at the bore area where the cracks of concern are occurring.
The higher the yield strength of the material and the higher the temperature, the faster the crack growth rate will be.
Westinghouse has also developed a calculational method to determine what the critical crack size of any specific disc will be. This is the size - primarily depth - of crack that could cause disc f ailure at design overspeed of the turbine.
Their calculations for the North Anna No.1 plant have been submitted to the staff.
The NRC staff also developed crack size prediction methods and critical crack size calculational procedures. These differ slightly from those used by Westinghouse, but provide similar evaluations in most cases. The chief differences are that the NRC crack growth rate curves (See Fig 1.) although based on the same data as the Westinghouse curves, take more account of disc material yield strength and related stress level, thereby predicting larger postulated cracks in some cases. Another difference is that the NRC procedure for calculating critical flaw sizes uses assumed ' law shapes that are more realistic, that is, are more similar to the shape of cracks actually found.
The resulting calculated critical crack depths are slightly larger than those calculated by Westinghouse. We feel that the NRC staff procedure is still highly conservative, but presents a more realistic situation when evaluating the integrity of any given turbine disc.
The results of the staff calculations are given in Tables 1, 2 and 3.
Results can be summarized as follows:
The two most critical stages in the North Anna turbines are the first and second stages. The later stages run at such low temperatures that stress corrosien will proceed at a very slow rate, if at all.
e We have calculated that the most critical first stage disc could now have cracks approximately 19% of the critical crack depth at design overspeed and this will only increase to 31% of critical size by December 1,1980, the time of the next refueling outage. The most critical second stage disc could now have a crack 27% of the critical depth at design overspeed, which could grow to 45% of the critical crack depth by December 1,1980. The current staff position is that when the postulated crack is less than 50% of the critical crack size, there is sufficient margin to account for any un-certainties in the calculations and ther6 fore no necessity to become concerned about turbine operation.
We therefore conclude that barring additional and new information that changes our calculational methods, there is reasonable assurance that the North Anna Unit 1 plant can continue operation until December 1,1980 with a very low probability of a turbine disc rupture.
Sigr.ificance of Yankee Rowe Turbine Failure.
The turbine of Yankee Rowe failed on February 13, 198L. Although only very prelirinary information is available at this time, it appears that both first stacs cisc in the. low pressure rotor failed.
There are indications of bore crac<.s, probably stress corrosion, on the fractured disc segments.
There are :ertain differences between North Anna and Yankee Rowe:
A.
Tire of Operation - f orth Anna has operated approximately 14 months.
The Yankee Rowe turoine operated for 187 months, and has never been inspected,
- rimarily because the rotor design did not provide access for inspection of ciscs.
3.
- ist's ':aterial - The material used for the Yankee Rowe discs had f airly low yield strength, but also probably had poor fracture toughness.
In terms of our cor.siderations cf the North Anna No. I turbine, we would expect slower crack growth, bec.'use of the lower disc strength and stress, but the critical crack size for failure would be signifi antly smaller.
Although we do not ye have as detailed information on the Yankee Rowe materials as we have for
' orth Anna.
ap;roximate calculations for Yankee Rowe using the conservative
- recedure ' described earlier, and used for our evaluation of North Anna, would
- r. ave predicted that the critical crack size would have been reached several years ago.
This te".ds to confirm our belief that our evaluation of the integrity of the f; ort". Anna turbine discs is conservative.
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TABLE 1 f40RTH AfifiA YS MG TO MAX M0. T0.
MAX.
DISC NO. TEMP.
2/15/80 RATE RATE REFUELING CRACK 1
329 Med 14.2
.028
.40 23.7
.66 2
267 Hi 14.2
.024
.34 23.7
.57 3.
213 Med 14.2
.004
.06 23.7
.09 4.
182 Med 14.2
.002
.03 3.7
.05 TABLE 2 WORST DISCS, CRITICAL CRACK SIZES
.1 Shape (Westinghouse).25 Shape @taff)
Kic Stress a Crit Crit Key a Crit Cr. Keyway Crack Crack DISC NO. LOCATI0f:
1 LP-1 Gov 200 76 1.8 1.43 2.46 2.1 2
LP-1 Gov 187 87 1.21 0.84 1.63 1.26 3
LP-2 Gov 228 72.3 2.6 2.2 3.5 3.1 4
LP-2 Gov 178 70 1.7 1.33 2.3 1.9 4
Kc Acrit =
Q i
1.21n a
for.1 Shape Flaw Q = 1 for.25 Shape Flaw Q = 1.35
NORTH ANNA TAE_E 3
% OF CRITICAL CRACK SIZES hestinghouse NRC
.1 Shape
.25 Shape LI"ITI!iG 2/15/80 CRITICAL CRITICAL DISC MAX CRACK KEYWAY of KEYWAY of CRACK CRITICAL CRACK CRITICAL 1
LP-1 GOV
.40 1.43 28%
2.1 19%
2 LP-1 GOV
.34
.84 40%
1.26 27%
12/1/80 MAX CRACK 1
LP-1 GOV
.66 1.43 46%
2.1 31%
2 LP-1 GOV
.57
.84 67%
1.26 45%
O
PROFESSIONAL QUALIFICATIONS VINCENT S. NOONAN U. S. NUCLEAR REGULATORY COMMISSION ENGINEERING BRANCH DIVISION OF OPERATING REACTORS I am currently the Chief of the Engineer.43 Branch, Division of Operating Reactors, responsible t )r directing and sus rvising the engineering safety review, analyses and evaluations of structural and mechanical components for reactor facilities licensed for operation, for the evaluation of applications and issuance of construction permits and operating licenses for non-power reactors and for the evaluation of operational and design modifications of ERDA and D0D-owned operating facilities exempt from licensing as requested.
,I was a member of the Mechanical Engineering Branch, responsible for the review and evaluation of design criteria for mechanical components, the
- dynamic analyses and testing of safety related systems and components and the criteria for protection against the dynamic effects associated with postulated failurcs of fluid systems foe nuclear facilities.
I was also one of three members of the Mechanical Engineering Seismic Qualification Test Review Team responsible for the review of. test methods for the seismic qualification of electrical and mechanical equipment. I also had the responsibility as the Mechan-
-ical Engineering Branch's principle reviewer on the issue of asymmetric leading on reactor vessel supports resulting from transient differential pressures around the exterior of the vessel and from forces induced in the internal structures within the reactor vessel.
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d 2-From March 1951' to March 1956 I served as a commissioned of ficer, Air Force Cadet and Airman witti the U. S. Air Force.
I worked full time During the period that I attended college (1956-1959) at night as a navigational analyst niid computer programmer at the
'U. S. Air Force Aeronautical' Chart and Information Center.
From August 1959 to August 1974 I was employed by McDonnell Douglas Aircraft i-
, and Astronautics Corporation as a group engineer in structural dynamics i
l responsible for short time dynamic transient (Shock), vibration and acoustics I
structural analyses and dynamic tests of both aircraft and spacecraft.
I was responsible for fin'ite element math modeling of complex spacecraf t i
structures and for analyzing random vibration and sonic fatigue type problems.
I have personally witnessed and was responsible for over 300 vibration and shock tests on both electronic and mechanical equipment and
' full scale spacecraft and aircraft.
In August of 1974 I joined the U. S. Atomic Energy Commission and have remained
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with this organization through the transition.to the U. S. Nuclear Regulatory Commission performing the type of wrk as previously described.
I serve as a member of the ASME Committee on Design Technology Transfer and I
while at St. Louis University I was elected'to the Pi Alpha Delta La0
, Fraternity.
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1 1 have written and published papers on such subje cts as structural respotst. to ic:pulsive loads, missile flight environmental measurement program, full scale spacecraft vibration testing and e.ner dynamic response related subjects.
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WARREN S. H'42ELTON
' Professional Qualifications My name is Warren'S. Hazelton.
In my capacity as Section Leader, Engineering Branch, Division of Operating Reactors, I am responsible for reviewing materials related aspects of operating nuclear power plants.
In conjunction with this work, I am also responsible for aiding in the preparation of Federal Regulations and Regulatory Guides relating to materials, inservice inspection, and opera-tional licitations important to the safety of nuclear power plants. Another primary responsibility is reviewing research, programs on reactor safety, evaluating results of these programs, naking recommendations for new programs, and factoring the results of these programs into our other review activities.
I was born in Cutler, Minnesota on October 20, 1916, and attended public schools in Duluth and Uchkon, Minnc;ota.
After attending the University of Minnesota intermittently, I joined the Armed Services in 1941.
I was cischarged in 1945 after serving as an Army Air Force pilot.
I then resumed my education at the University of Minnesota, was honored by beir.; selected for "Plur.b Bob," and graduated in 1949 with a Bachelor ofMetallurgica5Engineeringdegree,withdistinctior..
~
From 1949 to 1960 I was employed in the Westinghouse Aviation Gas Turbine Division, at South Philadelphia and at Kansas City, Missouri.
From 195' to 1960 I was manager of the materials application and O
development activity, responsible for the materials aspect of design, materials properties, failure analysis, and the development of new materials.
From 1960 to 1963 I was Supervising Engineer of the Materials Development Section at the Westinghouse Bettis Atomic Power Laboratory.
In this capacity I was responsible for development programs in the fields of stress corrosion, brittle fracture prevention, and radiation damage.
From 1963 until 1972, when I assumed my present position, I held various management positions in the Westinghouse PRR Systems Division. My responsibilities included the development and application of improved fracture prevention technology, evaluation of radiation damage, stress corrosion prevention, and involved close interface with design groups.
I was responsible for the detailed failure analysis performed on the internals at the Yankee Rowe, Connecticut Yan::ee, Trino (Italy), and SEfM (Franco-Belg.) plants.
I also participated actively in the redesign and repair work performed for these plants.
I have been active in the preparation of Codes and Standards relating to reactor safety.. Specifically, I am a member of several ASME Boiler and Pressure Vessel Code committees, the Pressure Vessel Research Committee Task Group on Fracture Toughness Requirements, and several ASTM committees developing standards for evaluating radiation damage of metals.
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ATTACHMENT 3
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