ML20136B857

Requests Permission to Perform Insp Program Per IE Bulletin 79-13 During Jan 1980 Refueling Outage.Cites Financial Pressure.Submits Documentation to Support Contention That Facility Will Not Suffer Weld & Pipe Failures

ML20136B857
Person / Time
Site:	Fort Calhoun
Issue date:	08/08/1979
From:	Short T OMAHA PUBLIC POWER DISTRICT
To:	Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
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ML20136B855	List: ML20136B851 ML20136B857
References
NUDOCS 7909130010
Download: ML20136B857 (70)

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... . n . m s Omaha Public Power District .

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I August 8,1979 Mr. K. V. Seyfrit. Director U. S. Nuclear Regulatory Comission Offica cf Inspection and Enforcement Region IV 611 Ryan Plaza Drive '

Suite 1000 Aritngt:m, Texas 76011

Reference:

Docket No. 50-285 1

Gentlemen:

The Omaha Public Power District hereby requests pemission to perform the inspection program specified in IE Bulletin 79-13 during the next scheduled l refueling outage at the Fort Calhoun Station in January 1980. The District recalved IE Bulletin 79-13, dated June 25, 1979, requesting that an inspec-tion program be conducted within 90 days of the date of the Bulletin, to evaluate indications of all feedwater nozzle-to-piping welds and of adfa-cent pipe and nozzle areas.

Since the receipt of the Comission's letter, the District has evaluated the design and operational characteristics of the Fort Calhoun Station with respect to the potential for weld and pipe failures as described in the Bulletin. In particular, the District examined: steam generator nozzle design, secondary water chemistry, piping and nozzle stress levels, thermal transients, and crack propogation parameters. The results of this examina-tion demonstrate that the Fort Calhoun Station is not susceptable to the kinds of failures described in the Bulletin. A detailed technical discus-sion is attached in support of this position.

Although it is felt that the Fort Calhoun Station'has not experienced any l cracking in the vicinity of the feedwater piping-to,-nozzle welds, it is ~

recognized that these types of failures have occurred at other facilities, and there is, therefore, sufficient cause for concern. However, as further .

stated in the attached information, the potential for cracking is sufficiently small at our station so as to warrant an extension in the inspection schedule from September 23, 1979 (90 days after the date of Bulletin 79-13) to January, 1980. Since a reactor shutdown for a minimum of eight days will be required to perfona these inspections, extending the schedule to January,1980, will permit the District to take advantage of a scheduled refueling outage to fulfill the bulletin requirements and, at the same time, assure contfaued station availability during the sumer months without jeopardizing the health and safety of the public. In addition, extending the schedule would result in a savings to the District of approximately $722,000 in net costs for re-placement power.

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Mr. K. V. Seyfrit August 8,1979 Page Two Considering the foregoing discussion, the Comission is respectfully re-quested to grant a schedule extension to January,1980, for the perfonnance of the subject inspection (designated in items la, b, c, of the bulletin) and provide such determination ca a timely basis. Should a cold shut down of sufficient duration occur prior to the next scheduled refueling outage, then the exams would be performed at that time. The District's staff is available to discuss this matter, should further information be desired.

Sincerely.

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, [2,T.E./;Short Assistant General Manager TES/KJM/BJH/rh Attachment xc: U. S. Nucle'ar' Regulatory Comission Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D. C. 20555 Director of Nuclear Reactor Regulation Attn: Mr. Robert W. Reid. Chief Operating Reactors 3 ranch No. 4 U. S. Nuclear Regulatory Comission Washington, D. C. 20555 LeBoeuf. Larb, Leiby & MacRae 1333 New Harcpshire Avenue, N. W.

Washington, D. C. 20036

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REQUEST TO EXTEND THE DEADLINE FOR THE FEEDWATER N0ZZLE-TO-PIPE AREA RADIOGRAPHIC EXAMINATION OF IE BULLETIN 79-13 1.0 FOTENTIAL FOR CRACK OCCURRENCE The Fort Calhoun Station steam generators have design features and an coerating history that differs from those units in which cracks ,

have been observed. These design features and the plant operating history,, which are discussed below should significantly reduce the likelihood of cracks that have been observed in other units.

I.I Nozzle Design The stema generator nozzle cracks observed at other units may be largely due to desi Fort,Calhoun Station. gn features that do not exist at the

• The Fort Calhoun Station nozzle design provides an excellent match to the feedwater piping and is dissimilar from the designs shown in References 1 and 2. Table 1 lists the .

pertinent Fort Calhoun piping and nozzie parameters. .

, TABLE 1

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Instde Outside Ma11 Diameter N ameter Thickness l Feedwater Nozzle 14.312 in. 16.562 in.

i 1.125 in. l (safe end) .

Feedwater Piping 14.312 in. 16.000 in. 0.844 is.

(16in.Sched.80)

This connection has the following features:

a) An even alignment of inside diamet'ers; .

b) Slight mismatch of outside diameters and wall thickness (standard design practice), but never --

reduced to less than design wall thickness for .

schedule 80 pipe; and c) No backing rings were used to fabricate this -

connnection. -

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'It can 'eadily r be seen that no significant discontfnuity stresses on the .inside surface should exist at the connection of the piping to the nozzle. Detafis of the nozzle design may be seen on drawing CE-E-232-485-5, which was submitted to the Consission on June 18, 1979, in response to the Cosmission's letter' of May 25, 1979, on this subject.

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.e 1.0 POTENTIAL FOR CRACK OCCURRENCE (Continued) 1.1 Nozzle Design (Continued)

This design must be compared to other designs in which cracking has been observed. In a design prevalent at plants in which cracking has been observed, the riozzle wall thick-ness is about .656 inches, approximately 22% less than minimum pipe wall thickness for 16-inch schedule 80 pipe.

In such a design, not only is there a stress disconti-auity where the piping,was machined down to match the l

nozzle inside diameter, but that portion of the pipe is considerably weaker than the remainder of the piping. '

This situation does not exist at the Fort Calhoun .

Station. .

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1.2 Water Chemistry Stress assisted corrosion cracking can be caused by improper '.

water chemistry control. The Fort Calhoun Station. Lar. -

has had excellent chemistry control. Data taken from January,1974, through July,1979, for dissolved oxygen in the feedwater show that 95.235 of the readings show no oxygen at all, as illustrated in Table 2.

The auxiliary feedwater source (the emergency feedwater storage tank) is maintained with a residual level of d hydrazine which is checked on a weekly basis and is covered with a nitrogen blanket to keep the water ongen free. ,

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l TABLE 2 l -

Dissolved 02 Level. Feedwater, in ppm Readings Accurate to + 5 ppb Data Period - January,1974, tErough July 5,1979 .

1 002 Percent Occurrence 0.000 95.2

<0.005 98.1 10.010 99.3 20.015 99.8 -

3020 100.0 ,

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1.0 POTENTIAL FOR CRACK OCCURRENCE (Continued) 1.3 Piping and Nozzle Stress Levels The feedwater piping is designed to withstand the combination of pressure, dead weight, thenna1, and seismic loading. The stress analysis (computer printout enclosed) shdws that calcu-lated stress levels as defined in the USAS 8 31.7 code are well below the allcwable. No seismic or water hamer type events ,

have occurred since piping installation, and the allowable stress levels have never been exceeded.

1.4- Ther.nal Transfents The feedwater nozzles are not subject to rapid thermal transients during startup, shutdown, or load changes due to the gradualness of the condensate /feedwater system temperature changes. Even during a plant trip, feedwater temperatures will change slowly -

over'several, hours with temperatures reducing from a full load value of 440 F to ambient. Sudden introduction of cold auxi-

,11ary feedwater flow to a hot line will not produce rapid temperature transients in the feedwater nozzle, as the auxiliary feedwater connection to the main feedwater system is upstream -

of the feedwater control valves outside of the containment building. This means that the initial auxiliary feedwater flow must travel th.ough a 200 foot long section of a large, already hot Tine before reaching the steam gener;ators. Thus, there will be no severe thennal transients due to the intro-duction of eixiliary feedwater flow. The " worst case" heatup and cooldow, rates have been used in the enclosed crack growth analysis discussed in Section 2.6 below. -

1.5 Operating Histog ,

The Fort Calhoun St-tin he had a relative;

• stable operating history with the plant generally werating at or near rated load. The plant history includa., only 16 cold shutdowns and 38 hot shutdowns. There have been no water hammer or severe vibration problems on the main feedwater piping at Fort Calhoun. Severe water hamers or vibrations .

have been noted at several of the plants in dich feedwater nozzle cracks have been noted. It is important to note that one plant which has had a history of water hammer problems, Calvert Cliffs Unit 1, is another Combustion '

Engineering plant and has completed the feedwater nozzle -

examinations directed by the IE Bulletin 79-13. Their  !

examination revealed no cracks. The materials used in fabricating the nozzle and safe end of the Calvert Cliffs steam generators are identical to those used in the Fort .

Calhoun steam generators. The radiographs of the nozzle-to-pipe welds of the Fort Calhoun steam generators have been re-examined, finding only minor amounts of slag or porosity.

Several of the radiographs displayed "No Apparent Defects."

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1.0 POTENTIAL FOR CRACK OCCURRENCE (Continued) 1.6 Crack Growth Study Enclosed is A Study of Crack Growth Rate in the Fort Calhoun Feedwater Noiile., performed by Pacifica Technology. The study, using historic operating and design data from the Fort Calhoun Station, and methods consistent with the philosophy and practice of the ASME Boiler and Pressure Vessel Code, Sections III and XI, shows that the worst possible crack that could.have hypothetically occurred thus far,due to p'rfmary and secondary stresses at the Fort Calhoun nozzlewwould easily meet the Section XI requirements for continued operation without repair.

2.0 err-ci 0F FEEDWATER N0ZZLE CRACKING Cracking or rupture in the Fort Calhoun Station feestater nozzles

_ wili not preclude safe shutdown of the plant.

There are two possible adverse effects of a pipe ifne crack or rupture. One is the loss of the line's function. The other is dasage to essential equipnent caused by the ruptured ifne.

Analysis, reported in Reference 3. has been perfonned that shows the plant can be safely shut down following the postulated rupture -

of a feedwater line. Furthcnnore, the loss of main feedwater 11aes will not prevent supplying auxiliary feedwater to the steam genere- .

tors since each steam generator has a separate emergency feedwater nozzle that is connected to the auxiliary feedwater syst'm. e In the event of a feedwater line rupture, the affected steam generator would be isolated from the feedwater system while maintaining the ability to feed the intact steam generator..

Restraints and other protective devices have been provided to prevent damage to essential equipment following the postulated I rupture of a feedwater line.

Any leakage from the secondary system inside of containment will be detected by the containment humidity detectors and the con-tainment sunp level detectors. The information can be correlated with both priniary and secondary system inventories to determine if and from where a leak is occurring. .

3.0 REFERENCES

3.1 IE Bulletin 79-13. USNRC.

3.2 Minutes of USNRC public Meeting " Briefing on Feedwater Nozzle Cracks in Westinghouse Reactors." - .

I 3.3 Fort Calhoun Station Unit 1 Final Safety Analysis Report.

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A ST15Y W OtAut GROWTH RATE IN TE FORT CALHOUN FEEDWATER N0ZZLE

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Omaha Public Power District 1623 Harvey Street Omaha, Nebraska 68102 ATTN: Mr. Richard Kellogg FRON: i Alan S. Kushner Robert E. Nickell Robert S. Dunham Ernest P. Esztergar i

July 24, 1979 Revised: July 31, 1979

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INTRODUCTIO.1 As a result of Nuclear Regulatory Corraission (NRC) IE Bulletin 79-13, dated June 25, 1979, Omaha Public Power District (OPPD) is under instruction to examine within 90 days its Fort Calhoun pressurized water reactor (PWR).. ,

This is because of findings of significant corrosion assisted fatigue cracicing in Westinghouse PWRs. During this shutdown, a detailed volmeetric examination of the steam generator main feedwater nozzle and its adjacent piping' is to be performed. Pacifica Technology has performed a thermo-structura.1 analysis of the main feedwater nozzle in order to assist WPD in respending to NRC IE Bulletin 79-13. -

The potential for corrosion-assisted fatigue crack growth in or near the safe-end weld of the feedwater nozzle of the Fort Calhoun steam generators has been examined from two points of view, both consistent with the philosophy and practice of the ASME Boiler and Pressure Yessel CodeE13, Sections III and XI'. The first approach consists of assuming the largest conceivable undetected flaw from pre-service radiooraphy and magnetic particle inspection. This flaw, in accordance with evaluation procedures specified in Appendix A of Section XI, is oriented in the most unfavor,able position in the nozzle with respect to crack growth potential, and the extension of the flaw is estimated from the service history stresses and water chemistry during the first six ya:rs of operation. This growth is ,

found to be minimal.

Secondly, this flat, in its extended state, is now subject to am Appendix A evaluation for need of repair, specifically with respect to its potential for reaching a critical crack size as the result of a Level C

. Service Limit seismic event (g levels defined in Appendix F of FSAR). The assume'd large number of cycles for this event provides the most severe test for unstable crack growth potential. The fact that the flaw easily meetsSection XI requirements for continued operation without repair is indicative l of the quality' assurance for this component.

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A summary of the results of these analyses follows. Additionally, a discussion of the mthods and justification for the analyses is incNded.

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SUMMARY

A worst case, highly conservative analysis of the stresses in the Fort Calhoun steam generator main feedwater nozzle has been performed. Stresses due to 1) internal pressure (p), 2) pioing system deadweight (DW), 3) thermal (th) loadings, and 4) an operating basis earthquake (OBE) have been -

considered. Of these loadings, the piping system deadweight, thermal er;:zus-forr under maxt:nts operating ' temperature of 550 F and operating basis ear:frquake were supplied in the form of the architect engineer's (Gibbs and Hill) cceputer output. The internal pressure and thermal transient loadings were evaluated by PacTech using the TEXGAP computer code [23. The following conditions were used: . .

Internal pressure of 1,000 psi, corresponding to the maximum relief valve setting. -

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Thermal transient haat up rate of 1200 F/hr (a conservative assumption since the maximum h' eating rate generally does not exceeded 300 F/hr). 1 The pressure and transient thermal stresses were calculated using the same TEXGAP model shown in Figure 1. The maximum piping stress resultants are susmarized in Table 2. As an additional conservative assumption the maximum values from the analysis of Gibbs and Hill were added together regardless of the actual point on the pipe cross section where they occurred.

The thermal stresses due to the heat up transient were found negligibly ,

small (less than 200 psi in the nozzle safe end) even with the increased l heating rate. A cross check on this result using a differeni: approach j (direct integration of a heat balance equation) is included in Appendix A. l The maximum stresses were found to act in the axial direction with l

valuas of 10,715 psi at the outside surface and 8390 at the inside surface"of l the pipe. The corresponding maximum values of the membrane, and bending components of a, = 9560 pst and ab = 1165 psi (shown also in Table 1) were

used to evaluate the crack growth potential according to the ASHE Code procedures. The calculations verify that:

1. Any undetected pre-service flaws have not grown to an un-

. accaptable level.

2. A postulated quarter thickness flaw would not propagate to cause the pipe to rupture under worst case seismic condition.

e The maximum undetected pre-service flaw is assumed to be equal to .025 times the wa.11 thickness. For the sixteen thermal cycles experienced so far, '

and assisming complete unloading of stresses from all sources during each cycle, this crack is found to grow .37 X 10-6 inches. For the quarter - -

thickness flas undergoing 50 seismic cycles of complete stress unloading the crack growth is found to be 94.4 X 10-6 inches . Both of these crack growths are so small as to leave the net section stress resultants unaffected, implying zero reduction in the service life capaht11ty. Details of these calculations are given in Appendix B. _,

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. DISCUSSION Partial and through-wall cracks in primary and secondary heat removal systems of light-water-cooled nuclear power plants have been a souce of concern to the industry and the regulatory bodies for almost a decade.

Origina.ily, this concern was focused on feedwater nozzles and small-diameter stainless steel piping of boiling-water-cooled units (BM's),with extensive resear=r carried out by the vendor, the General Electric C E33; through trtility owner groups; the U.S. Nuclear Regulatory Consission  ; and the utility research arm, the Electric Power Research Institute . The charac-teristics of this cracking are by now moderately well understood, and can be attributed to at least three interacting factors:

. 1) the relatively high primary and secondary stresses in these components, due to the combination of pressure and thermal loading, especially during start-up transients (although relatively high, these stresses meet ASME Code requirements without difficalty except fo'r fatigue analysis);

2) the metallurgical effects of welding, ranging from partial sensi-tization of austenitic stainless steels to residual stresses in dissimilar metal welds; and -

3) the deleterious effect of adverse water chemistry, especially the oxygen content, on the conversion of fatigue crack growth from the transgranular (or cleavage) mode, where yield stress plays a dominant role, to the intergranular moda, where the energy required l

for crack extension is a small fraction of that needed for cleavage.

As the result of research in these three areas, it leas been possible to simulate the accelerated crack growth rate in the laboratory for BW i systems. This experimental exercise was instructive especial!y with regard I

to items 1) and 3), since a combination of primary and secondary stress range that exceeded the 3 S, limit of Section III of the ASME Boiler and Pressure l

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l r Vessel Code and an oxygen content in the coolant stream of the order of several thcusand parts per billion were needed to produce cracks. l On the other hand the PWR secondary side cracking problem is somewhat

. enipatic, at this time, for several reasons. First, the stresses due to nomal and upset loads seem to be well below those levels that would sustain rapid fatigue crack growth. Second, although some cracking has been observed in the heat-affected zone (HAZ) of the walds connecting piping to transition piece to nozzle, other cracks have been observed in the base metal at disram that are thermally remote from the weld region.

There are characteristics of some PWR systems where crackina was found, .

, ,_ ,that differ significantly from the Fort Calhoun Combustion Engineerins . _ _

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(CE), design. We suggest that these differen.ces may'be capab'le of cro5cina '._

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.. "the. accelerated crack growth rates seen in other PlIR ' design's while ' '.. '. _ .I.'_~

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,__ substantiating the low propagation rate'found from our anal.ysis'."Th'e _._ ..~ ~' ..

_( major ' difference' is the welded joint between the carbon steeliloina Z "_/

-]and the high-alloy steel of_ the steam ' generator shell and, noz'zie "found.2 ~~

in Westinghouse reactors. The affect of differing thermal expansion and thermal conductivity properties on the residual stresses and deformations have only been cursorily studied , primarily as the result of the liquid-metal-cooled fas.t breeder reactor (UfBR) steam generator program, but the history of premature failures in superheater transition welds of fossil-fired boilers is well documented .

. The other significant difference between other PWR designs and the Fort Calhoun unit is the elimination of the likelihood of unanticipated cyclic loading, such as waterhamner.

It is premature to speculate on the importance of these various factors in promoting intergranular crack growth without more definitive experiments.

Nevertheless, it is likely that a number of mechanisms are needed in combination to produce failure. It may be that either thermal or mechanical loadings, are higher than anticipated in the design specification; dissolved oxygen levels are significantly higher at the nozzle than at the point of

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i measdrement; or the field welding process may introduce adverse residual stresses and metallurgical changes. The possibility of additional thermal

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cycles caused by the introduction of cold make-up water into the nozzle could greatly accentuate the problem and oxygen may enter in the system at some

. undatected point.

Fcr the Fort Calhoun Plant unlike Westinghouse Plants, these factors can alt he aTiminated. The loading and thermal transients experienced by the plant have prnd nd very low stresses. Make up water is' chemically treated and added to the recirculating water at the heater, so it is stated that the heater brings the make-up water to 440 F, hence there is no possibility of a cold ficw at the nozzle during normal operation. Additionally, oxygen

, content readings are taken as the feedwater leaves the heater, with only a pipe run to the nozzle, eliminating any possibility of additional oxygen entering after the water chemistry is measured. -

The Undetected Flaw The Fort falIdh steam generator nozzle, safe end,'and coanecting

~..._ piping are. presently classified as Class 2 structures in accordance with the 1967 edition of the Code. The justification for this safety classification is their location outside the primary pressure boundary. As Class 2 structures these components are not required to undergo the rigorous volumetric pre-service inspection by ultrasonic methods that is demanded by Section XI for Class 1 Structures. However, two methods of inspection were used on the safe-end weld prior to service: a surface discontinuity detection method, magnetic particle, and a volumetric discontinuity detection method, radio-graphy. 3 i

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, Magnetic particle inspection is a nondestructive method of detecting the presence of cracks, seams, inclusions, segregations, porosity, lack of fusion and similar discontinuities in magnetic materials. The areas to be ,

inspected are covered by finely divided magnetic particles which react to the

, magnetic leakage field produced by the discontinuity. These magnetic I

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, particles form a pattern on the surface which is an indication of the approxi-mate shape of the surface projection of the discontinuity.

tio surface discontinuities were detected at the safe-end weld during tha magnetic particle inspection.

e Radicgraphic inspection involves the absorption path of short-wave-1enr/.h radiation through metal, wi,th the detection of defects due to the variaticn in absception from that of sound metal. The shorter propagation path in the defective structure is measured on a film sensitive to the radiaticn, which shows an image that represents a normal projection of the indicstiort Radiographic methods are assumed to produce films with a sensi -

tivity of 2 percent relative to a change in weld thickness, or propagation path. For example, a one-inch (2.5 cm) thick weld would have a reference filart: rage such- that a variatica in propagation path of .02 in is detectable.

The radiographic procedure for a particular structure, such as the safe-end welds in the Fort Calhoun steam generator feedwater nozzles, is

, calibrated th.ough the use of a AS4E Code gage device called a pene-trameter-a thin strip of metal equal in density to the weld metal, and less than or equal in thickness to 2 percent of the weld metal thickness. The sharpness of the outline of the penetrameter image against the fi.la background is an indication of the sensitivity of the film.

Pre-service radiography showed no indications of any kind in the .

. safe-end welds. Therefore, it is reasonable to assume that the undetectable l flaw is at the upper limit of the penetrameter calibration thickness-namely, 2 percent of the weld thickness. For the Fort Calhoun feedwater nozzle safe-end welds, the depth of the undetected flaw would have an upper limit of

.017 in. As a conse vative assumption an initial undetected flav of .025 in, was assumed'in the following crack growth analysis which attempts to show

that if the operating condition of the Fort Calhoun Plant were similar to those of a Westinghouse plant doulots would develop as to the structural integrity of the main feedwater nozzle.

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~ It should be pointed out that a closed planar flaw oriented parallel to the radiation path would not be detected by radiographic methods, since the projection in a plane perpendicular to the propagation path would be minimal for this configuration. However, the perfectly planar flaw is an analytical concept, designed to simplify the geometric treatment of the crack extension proc 2ss. In fact, crack extension, especially for intergranular growth of the type expected in oxygenated water, would take place along an irregular path with ample projection in a horizontal plane.

C.* Er=wth Rate Estimates The crack growth model will be based upon the expression developed by Paris  :

h=C(AK)" (1) 4 where a is the depth.of the crack, N is the loading cycle, K is the stress intensity range, and C and n are constants that reflect environmental conditions (temoerature, dissolved oxygen levels, etc.) and material structure. For loading that is not completely reversed or removed in a cycle, the Walker adjustment will be used:

h=C(AK,ff)" (2) i where

, AK,7f = Kmax U 'N) (3) - ,

R=K ainNaax -

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~ The constants C and n are determined from crack growth experiments of  !

the type described in [9,10] with standard partially-cracked specimens. The l data of Hale et al. is especially informative. Two types of specimens of A333 Gr 6 were tested, one with partial cracks in base metal and the other with partial cracks in the heat-affected zone (HAZ). The testing temperature was 550 F (288C) and the dissolved oxygen level was monitored and maintained at 200-400 ppb. Two cyclic frequencies -(for loading and unloading the spectren) were used,18 and 75 cycl'es per hour. Although the lower cyclic

- fra y clearly indicated an order of magnitude increase in crack growth j rata, probably due to intergranular attack at stress, the authors state that )

"no additional increase in crack growth rate is observed (for low alloy steels) for cycle frequencies less than roughly 10 to 20 cph". This .

. statement is unsubstantiated. (If these laboratory tests were carried out with hold periods, at peak stress intensity, the results would be more believable.) The effect of cyclic frequency holdtime is a rel:tively new consideration in crack growth studies. .

The A333.Gr 6 material has precisely the same chemistry as $41068.

. Therefore, an extrapolation of the effect of cycle frequency, in order to account for. the longer duration of exposure to the oxygenated water at stress can be attempted from the two frequencies in 23. Figure 2 shows that the mid-range of crack growth data for an intensity of about 25 kst/is and a cylic frequency of 75 cph is about a factor of thrce er four lower than for the same intensity at a frequency of 18 cph. If the startup transient is the event of concern and if the period of time at peak stress is on the order of one to three hours, then the cyclic frequency of the cycife event is about 1 cph. This period alsa coincides with the temporary rise in' dissolved oxygen l content found in the Fort Calhoun water chemistry readings.

l Since the intergranular corrosive attack is a diffusion-controlled process with an exponentially-decreasing rate, the extrapolation is carried out on a logarithmic scale. This would imply an acceleration of the crack growth rate by about another factor of four or five under actual conditions.

Therefore, for K,ff = 25 ksiE, the actual crack growth rate is probably 4

l near'10-3 in/ cycle, or even slightly higher. For conservatism we choose th crack growth rate to ce even higher h = 10-2 (3geff)* .(5) wMdr implies that the Section XI Yatigue crack growth rate in a water

.- ewircnment has been shifted upward by a factor of about 20.

If the minimum yield strength is given as 35 ksi, so that the 3 k limit for primary and secondary stress is 70 ksi, then simple estimates .of crack growth can be made by assuming an initial crack length and a stress  ;

state that just satisfies the design limit. For an initial crack length of 0.025 in, the stress intensity range would be on the order of 0.3 (3 I s), w  !

about 20 ksi dn. There is an posibility that th.is value could be as high as 30 ksiEor as low as 10 ksi dn. Taki.1g the upper limit, we find that

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h=.003in/ cycle , (6) so that, in twenty cycles, the crack would grow to a length of about 0.1 is.

The preceeding is a highly spe:ulative indication that nozzles designed to an operating stress level approximately 4 to S times higher than the Fort Calhoun Plant and experiencing oxygm levels two orde s of magnitude higher could experience low c le corrosito assisted fatigue crack. It is apparent from the data of both 93 and DO) that for stress intensity factors less than 10 ksi 6 , the ASME BPV Code design curve is extremely conservative regardless of environment or cycle frequency. Indeed, would imply that the 3 ksi Klevel used for operating life crack extensten in Appendix 8 is below the threshold for which no cyclic crack growth is found.

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' The region where significantly higher crack growth rate has been found is bounded by 15 ksi Eiiand 50 ksi 6as sho.m in Figure 6 reference 12, a l recent suanary study by W. Sainford.  ;

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, REFERENCES , ,

[1] AStiE Section XI, Appendix A, " Rules for Inservice Inspection of Nucisar Power Plant Components, Hun-Handatory Analysis of Flaw Indications", ASME Boiler and Pressure Vessel Code (July 1,1977 ,  !

EditionwithAddenda). ,

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[2] Rwkar, E.E. and ILS. Dunham, "Three Dimensional Finite Element" -

r7*=e Program Development" AFRPL-TR-78-86, February 1979. ' ' -

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[3] Danko, J.C., et. al., "A Pipe Test Method for Evaluating the Stress '

Corrosion Cracking Behavior of Welded Type-304 Stain 1ess Steel .l

. Pipes", in "Prcperties of Steel Weldments for Elevated Temperature ,

and Pressure" ASME, 1978. -

[4] Pipe Crr.k Study Group. " Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants", '

NUREG-0531_, '5."S' NucTear Regulatory Commission (February 1979).

i

[5] Harston, T.U., Editor, " Flaw Evaluation Procedures: ASME Section XI";

' EPRI-NP-719-SR, Electric Power Research Institute (August 1978). *

[6]. Mayfield, M.E., Rodabaugh, E.C. and Eiber, R.J., A Comparison of

, Fatigue Test Data on Piping with the ASHE Code Fatigue Evaluation

• Procedure Paper No. 79-PYP-92 ASME, New York (1979). , ,

[7] Burchheit, R.D., Herry, W.E., Strabel , G.R. and J.L. McCall, " Failure Analysis of a 304 Stainless Steel Schedule 8 Reducer" Final Report ' .

NP-20325, Battelle-Columbus Laboratories (August 26, 1974). -

[8] Gray, R.J. , King, J.F. , Leitnaker, J.M. , and G.M. Slaughter,'

" Examination of a Failed Transition Held Joint and the Associated ~

Base Mei:als", Report No. ORNL-5223, Oak Ridge ' National Laboratory (January 1977). '

e

,,,e-._- we.- ,w-+---ne ~~~- -~ ~ - - --- - - - * - - - - " = ~ ~ ~ ~ ~ " " " ' " ~ ~ ' ' " ' ' ' ' ' ' ' ' '

__ _ _ _ _ _ - . . . - -- =.--. .- . --

[9] Hale, D. A. , Jewett, C.W. and J.N. Kass, " Fatigue Crack Growth Behavior of Four Structural Alloys in High Temperature High Purity Oxygenated Water", Paper No. 79-PVP-104, ASME, New York (1979).

[10] Vanderglas, H.L. and B. Mukherjee, " Fatigue Threshold Stress Inter:sity and Life Estimation of ASTM-A1068 Pipe Steel", Paper No.

~ ' ' '

73-P'IP-85, ASME., New York (1979). .

[11] Walker, K., "The Effect of Stress RatJo During Crack Propagatios *

,' l ani Fatigue for 2024-T3 and 7075-T6 Aluminum, Effects of Envirc,~_ .,t .

and Complex Load History on Fitidue' Life, ASTM-STP-462, Ame~rican " -

Scciety for Testing and Ha'terials,1970, pp.1-14. -

[12] Eamford, W.H.. " Application of Corrosion Fatigue. Crack Growth Rate l Data to ' Integrity Analysis of Nuclear Reactor Vessels *, ASME Paper-No. 79-PVP-116, March 1979.

- ~

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=; '3; , ,,

Table 1

' Stresses (psi) for 1000 psi internal pressure Case 'h h h h I. 4412 2682 3550 860

2. 4985 3210 4100 890

3. 5730 5180 5460 275 4.- 10.715 8390 9560 1165 Case Sumary: .

- l'. pressure + dead weight

. 2. . pressure + dead weight + 0.B.E.

3. thermal stresses

4. pressure + dead weight + 0.8.E. + thermal Note: The stresses listed are calculated axial stresses to which an extremely conservative value equal of pr/2t = 4,750 has been added. -

4 w e

W

Table 2 Joint Forces f. ADL PIPE Z :

I = 1156 in4 PT.0 e - -

A = 40 in 2

~

r,= 8'

~

- (2.37,0,-8 SA 106-3

.\

ASTH-A106-G .

F F' "y " ## 0.D. = 16" .

'u " II '*

t = 0.843" - - -

e ,= .35 - 1 r i -

PT.1 F x F y F, 4 My R, Thermal ,

4,970 -1,942 19,231 -852,2865 '12,164 261,166

~

Seismic 1 ,41 5 392 4,942 -54,581 -64,253 25,520

. Deadweight 0 104 -2' 9,007 -353 -

245 8 = tan'I (2.37/8.87) = 14.96* ,

m FN = F, sin e + Fzcos 9 -

L J F5 = F, cos e - F, sin e ,

F, F 'F H M 5 y ,, 3 . M, ., .

Thermal 19,862 -163 -1,942 472,327 755,980 17,164 Seismic 5,140 92 392 11,531 -59.577 -64,253 Deadweight 2 0 104 2,088 8,638 . -353 m__ .

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.4 .

APPENDIX A As a check on the zero thermal stress state predicted by the TEXGAP j analysis a heat balance calculation was performed. A measure of the impc:tance of the thermal resistance within a solid body is the ratio of tt j

i:tter.al to the external thermal resistance, referred to as the Biot number a-y

~

(A-1) where h is the average unit surface conductance, L is a significant dimensii-obtained by dividing the volume of the body by its surface area, and K is ti thermal conductivity of the solid body. It has been found that for -

cylinders, the assumption of a uniform internal temperature is in error by less than .05 if 8 < .1. For the nozzle safe end 8 = .12 and the assumptl.on,,of.,a uniform internal temperature is valid if we can show that the cylinder is heating at the same rate as the stream temperature, i.e 1200F / hour.

To demonstrate this we define the heat balance equation

, -pcVh=hA(T-T=) (A-2) where p = .283 lb/in 3 c = .13 BTU /lb.-F "

V = 40 in 3 A = 45 in 2

}

T= = 120 F + 1200t F 2

h = .3 8TU/in -F -hr .

t = time in hours

_ . , _ , ____,,e_-

.q , ,

The solution to A-2 is

-9.2t T = (130 e - 10 + 1200t) 'F (A-3) 1 Eq'.ation (A-3) indicates that after a transient of less the .1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the j hearing rate is 1200F / hour. Since the actual heating rate very rarely hc exceeded 300F / hour, the assumption of uniform cross-sectional temperature seems valid.

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APPErlDIX B Tne crack growth results reported were calculated according to the proceedures outlined in Appendix A of Section XI of bl. Two cracks were pastulated, a .025 t and a .25 t, where t is the piping / nozzle wall '

thickness. The .025 t crack represents the maximum allowable undetected

. preservice crack, while the .25 t crack represents the maximum service life '

crade growth. The procedure is to first demonstrate that the .025t crack -

, would not have experienced any appreciable grodh during the known service Iffa trh^ wry. The crack growth law

=.3795X10-3(3geff) W (8-1) where , .

4 A K ,f f = ,Kg x (1-R)0.5 (8-2)

R = (9,4x ,

(B-3) shall be used. The service life cyclic history has so far involved 16

. thermal cycles during which the temperature never dropped below 120 F. For ,

conservatis:s we shall assume that all stresses for Case 4 of Table 1 cycle between 0 and their maximum values during a thermal transfent. Hence Keff' "

Enax *"# .

. K,,, = G,M , E /a/Q + Gy My [ /alQ (B S t

f DJ -

e l . ..

,. ,, ~. - - . - .-- , - - = = - ~*- '~* -~ " ' " ' ' '" ~ ' ~ '

M ,= 1.1 Mb = 1.07

.Q = 1.2

= 3 ksi /Tii

.K max (B-5)

, Usimg tMs value of K,, in (B-1) yields .

aa = .37 u in. (5-6) ,

a crack growth that has no effect on the net section Toad carrying capability.

The second step is to now demonstrate that the .25t crack will not

. cause a pipe rupture during a seismic event corresponding to an 0.8.E. For a = .25t, (B-4) yields h x = 9.8 ksi 6 -

47) l Reference requires the demonstration that the crack does not grow to 1/2 the critical crack size (defined as 3/4 of the wall thickness). The US NRC Standard Review Plan (Section 3.7.3) requires that 10 maximum stress cycles .

be assumed for an earthquake event. For additional' conservatism we asume 50 cycles for the seismic loading. As can be seen from Table 1, assumfag the seismic and pressure loads to cycle would have the stress decreasing a maximum of 50 percent, implying R = .5 and K,ff = (.7) K ,. For con- .

servatism, we shall use K,f f = K,,x . This yields, for 50 cycles

. C .

Aa = 94.4 y in.

(8-8)

O G

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