ML19254E457

Forwards Response to NRC Questions Re Fatigue Crack Growth & Repair Options Applicable to Feedwater Sys Piping Cracks. Discusses Program to Assess Feedwater Pipe Cracking Phenomena

ML19254E457
Person / Time
Site:	Millstone
Issue date:	10/24/1979
From:	Counsil W NORTHEAST UTILITIES
To:	Reid R Office of Nuclear Reactor Regulation
References
TAC-11793, NUDOCS 7911010135
Download: ML19254E457 (41)

Text

<.3 IMHtTHIIAST trrILITIIIS 3 H ARTFoRo. CONNECTICUT,06101 (203) 666-6911 L L J

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October 24, 1979 Docket No. 50-336 Director of Nuclear Reactor Regulation Attn: Mr. R. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

References:

(1) E. L. Conner telecopy dated October 16, 1979.

(2) W. G. Counsil letter to R. Reid dated September 28, 1979.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Feedwater System Piping Please find enclosed as Appendix #1, Northeast Nuclear Energy Company's (NNECO) response to questions raised by NRC relative to fatigue crack growth and repair options applicable to the feedwater system piping cracks. This information is essentially comprised of that which was presented to the NRC Staff orally in Bethesda, on October 19, 1979.

During that meeting, a question was raised by the Staf f relative to the probable cause for crack initiation and the rationala for its present depth of approximately 100 mils. i.s a member of the Westinghouse Owner's Group on Feedwater Pipe Cracking, we are currentl3 involved in an extensive program aimed at assessing this cracking phenomena. Feedwater piping strain, temperature, and acceleration have been measured at several plant sites including Millstone Unit No. 2. This data has nct revealed ang significant dynamic strain which could be construed to have caused the crack initiation. Significant thermal stratification in the first horizontr.1 run of the feedwater piping has been measured at all plants during plant startup, coincident with low feedwater flow rates. Stresses calculated from this steady-state stratified thermal profile do not reveal levels sufficient to cause crack initiation or appreciable crack growth. It is our belief, however, that this existence of a zone of temperature instability is the cause of crack initiation and crack growth.

NNECO has had considerable experience with two-temperature fluid regimes in feedwater systems as a result of inspections performed on the Millstone Unit No.1 RPV feedwater nozzles beginning in 1974. This inspection and subsequent thermocouple data revealed a thermal instability phenomena of sufficient magnitude to initiate and grow tracks to a certain depth. At that time, ti - General Electric Company initiated a ecmprehensive program to rasolve these crackit.g occurrences.

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This program culminated in the publication of two (2) CE reports, NEDE-21480 and NEDE-21821, which document in detail, experimentation and analysis to support continued safe operation of these BWR plants.

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PWR feedwater piping temperature measurements have not confirmed the presence of high cycle ( sul hz) water temperature fluctuations during these periods of thermal stratification. However, tests performed LA GE at their two-temperature test facility and the Moss Landing Facility demonstrate a trend tovatd thermal fluctuations in the vicinity of the two-temperature flui ! .erface. Assuming that fluid temperature fluctuations on the order of 300*F are present during low flow conditions (Millstone Unit No. 2 measurements reveal a potential of 350*F), it can be seen from Table 3-1 of NEDE-21821 that cracking can initiate af ter approximately 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> of hot standby operation. It is estimated that the thermal skin stresses for carbon steel are approximauely 2/3 of tuose for stainless steel. The depth to which these skin effects penetrate is repre-sented by Figures 3-2 through 3-5. It is noteworthy that very little aetal response to wasar temperature fluctuation is felt beyond the 0.1 inch depth, pal ticularly for the higher frequency casec.

There is no evidence of a driving force which could cause appreciable crack growth beyond the 100 mil depth to occur in the Millstone Unit No. 2 feedwater piping system. This is supported by analytical and experimental ob;ervations of the Millstone Unit No. 2 system. Conditions which have caused Aarger crack growth at other plants could be:

(1) The aetallurgical condition of the material.

(2) The residual stress distribution.

(3) Counterbore geometry.

(4) External loadf ug conditions.

Regardless of the cause, it is important to note that none of the above conditions change with time. That is, the state of stress intensity factor at the feedwater pipe crack tips can be described by information obtained from the installed instrumentation and the ultrasonic examinations throughout future plant operation.

The reexamination of these crack indications is scheduled to be performed during the upcoming October, 1979 outage. This and the other planned interim examina-tions will confirm that the crack growth is within the predicted crack growth rates.

As stated in Reference (2), if any increase in crack depth is evidenced, vithin the tolerance of the ultrasonic examinations, the cracks will be repaired. The results of this inspection will b plotted and compared directly agair.sc those taken in August, 1979. It is judged that any appreciable change in crack front geometry or depth wLil be readily apparent.

With regard to operational philosophy of the auxiliary feedwater system during plant startup, NNECO will continue to operate in the continuous feed mode whenever possible. That is, once auxiliary feedwater flow is established, abrupt changes in flow rate will be avoided in order to minimize u e potential for crack propaga-tion.

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• A permanent repair is planned for the 1980 refueling outage. At that time, sufficient data will have been reduced and causative mechanisms identified such that positive solutions can be employed to minimize the potential for further cracking. This approach results in the significant advantages of minimizing personr.el exposure and reducing plant outage time. Performing a 19 pair at this time is further complicated by the recent sick-out organize 3 by the local craft unions. Resolution of this situation may not recur for several weeks. To effect a repair at this time may require solicitation of non-union personnel and qualification and indoctrination to the various repair procedures. It is anticipated that this could add substan-tially to the duration of the outage, if the repair were to be undertaken at this time. Def erring the replacement, therefore, results in substantial man-rem and economic benefits.

Should NNECO's anticipation of no detectable crack growth be realized, the 1980 replacement program is concluded to be technically defensible and appropriate.

An interim ultrasonic examination, prior co March 1, 1980 will be conducted to verify the absence of crack growth. The leak detection equipment will remain functional and be monitored without change from the current program until the issue is permanently resolved.

We trust you find the above information responsive to the Reference (1) requests and questions raised by the NRC Staff during our October 19, 1979 meeting.

Very truly yours, NORTHEAST NUCLE /I FNERGY COMPANY

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W. G. Counsil Vice President Attachments

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c; e DOCKET NO. 50-336 APPENDIX #1 MILLSTONE NUCLEAR POWER ST..IION, UNIT NO. 2 RESPONSES TO ADDITIONAL INFORMATION REQUIRED TO ASCERTAIN ACTIONS NECESSARY REGARDING THE FEEDWATER SYSTEM PIPING 12 8 i .'.T OCTOBER, 1979

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PURPOSE In response to the NRC request of October 16, 1979, the additional information required to ascertain actions necessary regarding the feedwater system, which was presented to the NRC on October 19, 1979, in Bethesda, Maryland, is hereby submitted.

Question 1 In your letter of September 28, 1979, you stated that thermal variations (stratification) was observed during low flow conditions. Address the potential for cracx propagation during these low flow conditions (low cycle f atigue).

Provide a quantitative analysis regarding crack growth rates during the thermal transient cycle.

Response

The inspections conducted in response to NRC I6E Bulletin No. 79-13 revealed linear circumferential indications adjacent to the steam generator feedwater nozzle safe-end to pipe and pipe to elbow welds in both feedwater piping loops. The largest observed linear indications near each weld were subsequently mapped by ultrasonic inspection in terms of through-wall depth and circumferential length. In addition, the arec of interest was instrumented in order to establish the mechanical and thermal feedwater piping loading conditions durin o startup and full-power operation. The results of the instrumentation data specifically applicable to Millstone Unit No. 2 are presented in Attachment #1.

Ynowing the crack sizes and orientation of the feedwater piping flaws, Westinghouse assessed the potential for crack growth in a conservative manner considering not only the original design basis transients but also the additional thermal loading derived from the instrumentation results at the subject unit. The assessment of crack growth for the linear indications in the feedwater piping was performed by Westinghouse and it is enclosed as Attachment #2.

The results of the conservative assessment indicate that the largest 11aw in the feedwater piping system will not grow significantly. Furthermore, the final flaw size for the worst location is a factor of five (5) smaller than the -

established critical flaw size.

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.- e Question 2 Assuming the analysis requested above predicts crack growth at a suf ficiently low rate to ensure adequate safety margins can be maintained until a permanent repair can be made at the June,1980 refueling outage, provide the details of an augmented inspection program which verifies that crack growth has not occurred at a rate faster than predicted by the analysis.

Response

The results of the crack growth assessment confirm that crack growth will occur at a sufficiently low rate. Therefore, adequate safety margins will be maintained until the June / July,1980 refueling outage.

The augmented inspection program for the observed feedwater pi 'ag flaws between Octobe) 31, 1979 and the next refueling outage was evaluated consistent with the present understanding of the feedwater piping cracking phenomena. Based on this and the results of the Millstone Unit No. 2 feedwater piping system instrumentation data, it is important to note that thermal loading conditions exist durf ag plant startup and plant shutdown operations which have the potential to induce further crack growth. Therefore, it is imperative that the inspection frequency be established consistent with the objective to minimize the potential for crack growth in the interim period.

Based on the above, it is proposed that the feedwater piping flaws be inspected as a minimum prior to March 1,1980 to verify that the actual crack growth is within the predicted crack growth. In addition, it is proposed to conduct additional inspections of the feedwater piping flaws at any plant cold shutdown of more than two (2) days duration between October 31, 1979 and the next refueling outage. However, the time interval between successive inspections shall be more than three (3) calendar months. All results of the additioral inspections conducted in the interim period will be submitted to the NRC within seven (7) days followf.ng the completion of the examination.

Question 3 In the proposed repair / replacement program you submitted, you stated that the removal of the shield wall section can be made within design bases limitations.

a. Provide the technical information supporting your conclusion.

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Response

The original shield wall model was modified to include a six (6) foot diameter by twelve (12) i.ch deep cutout. Imposing the design basis loading conditions indicates that the shield wall cutout will not degrade the load bearing capa-bility of the shield wall. The design basis loading conditions were derived from the subject unit Final Safety Analysis Report.

Question 3.b Provide assurance that the method of removing part of the concrete wall by drilling and chipping will not damage the concrete left in place and the existing reinforcing bars. Describe the quality assurance procedures which will be used during thc concrete removal operation.

Response

The shield wall removal by drilling and chip,.nc sill be performed in accordance with the work procedures submitted to the NRC on September 28, 1979. Industry experience indicates that the proposed removal process will not damage the concrete left in place and the existing reinforcing bars.

As part of the QA Category I work procedures referenced above, Quality Control will monitor and perform a final inspection of the shield wall cutout to ensure that the shield wall removal was accomplished within the specified dimensional requirements and that the minimum specified covers exist for the specified rebar.

Question 3.c Describe the procedure which will be used if the reinforcing bars must be removed.

Response

A grinding disk will be used to remove reinforcing bars.

Question 3.d If replacement on the removed shield wall segment is required, ad/.res; the following:

.

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.: e (1) Define the concrete mix which will be used to fill the recess in the shield wall.

(2) Describe the procedure for reinforcirg bar replacement.

(3) Define the method to be used to ensure compatibility of the new and old concrete, especially the measures planned to limit shrinkage of the new concrete. Discuss the degree of working together that can be expected from the new concrete and existing wall.

Response

Based on the fact that the steam generator feedwater nozzle / safe-end to pipe weld ' will be subject to future periodic inspection, it is concluded that the shield wall cutout will remain as is to provide the required accessibility.

Question 4 Provide the details for material removal as discussed in repair Option B.

Also, provide the detailed procedures for the weld repair on the ID of the pipe should the wall thickness be reduced from Code limits. Address the mock-up used to qualify the welding procedures, for training and to qualify the welders / welding machine operators.

Response

The Option B repair method consists of pipe removal outside the shield wall and repair of the steam generator feedwater nozzle safe-end to piping indications from the ID.

CE Chattanooga has developed a grinding fixture which will be positioned inside the pipe. Prior to positioning of the grinding fixture, a liquid penetrant examination will be performed to map the inside pipe / safe-end area containing the linear indications. A six-inch grinding wheel attached to the positioning fixture will be used to remove any linear indications. The width of the grinding wheel is approximately one inch. After the grinding operation, a final liquid penetrant examination will be performed to ensure complete removal of the linear ind ications.

The weld repair will be performed with the Jiametrics internal welding fixture in accordance with applicable code requirements. The weld repair criteria will 1281. ::'

y be based on fatigue considerations rather than minimum wall thickness limitations specified by the ASME Section III code. Therefore, it is our intent to weld repair all ground-out areas from the ID in order to eliminate all stress intensi fication locations. The final weld repair contour will be polished and blended uniformly to the adjacent base material.

All welders will be qualified to the approved weld procedure. The weld procedure will be qualified in accordance with the applicable code requirements.

Question 5 Describe the simulation for welder training and qualification to account for the limited access between the shield wall and steam generator in Option A.

Provide any details regarding the consideration of automated welding to make the nozzle to pipe repair.

Response

The Option A repair requires the specified shield wall removal in order to gain access to the steam generator nozzle safe-end to pipe welds. After shield wall removal, the feedwater pipe will be cut between the steam generator nozzle to safe-end weld and the safe-end to pipe weld as specified in the work procedures submitted to the NRC on September 28, 1979.

The weld procedure for the Option A repair method will be qualified in accordance with the applicable code requirements. All welders will be qualified to position 6G (QW-405 ASME Section IX) which essentially qualifies them to perform welding in any position. In addition, all weldcrs will be trained in the steam generator nozzle / shield wall mock-up to simulate the welding conditions in the limited access area.

Based on our investigation of four manufacturers supplying automated welding tools, it is concluded that no automated welding tools exist at the present time which will physically fit into the limited access area.

128 F ._ / '

.: e Question 6 State if a UT baseline examination will be performed for the nozzle to piping weld if Option A is used.

Response

A baseline volumet' examination will be performed for the repair nozzle to piping welds (Option A) . However, based on the requirements for future inspec-tions of the subject velds by radiography and the fact that radiography has been used to detect the feedwater piping indications, it is concluded that the baseline volumetric examination will be accomplished radiog aphically.

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• DOCKET No. 50-336 ATTACHMENT #1 MILLSTONE NUCLEAR POWER STATION, UNIT No. 2 FEEDWATER SYSTEM PIPING 1281 .':1

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SUMMARY

OF TEMPERATURE PROFILE OCCURRENC - . , ,

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I RANGE NUMBER '

PRCFILE 2 EVENT NUMBER OF OCCURRENCES 0F EVENTS(I) 0F PROFILE PROFILE -

1 y PLANT 4 5 3

1 2 L ,

276- 1242 276-1242

__

'

IIS

-_

5 9 _

6 6 6

-

HOT STANDBY I MILLSTONE i

- (LINE2). I N l i

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~

> 300"F) IS BASED ON AVAILADLE TEST D (I INf 0RMAT10f' .

,' NUMBER OF OCCURRENCES (WITil aTTOP/00TTOM Y tif orCllRRii.t I S I ?

(2) NUMBER OF EVENTS IS BASED ON PRESENTLY ATED AS: RAllGE = (

(3)X RANGE FOR TOTAL (Il EV NTS) X S, WilERE S = EVENT SIMILARITY FACTOR A NUMBER ND .555 OF OCCURRENCES O 5 1.5.

_

.: .

DOCKET NO. 50-336 ATTACHMENT #2 MILLS *IONE NUCLEAR POWER STATION, UNIT NO. 2 FEEDWATER SYSTEM PIPING 1281 .X'

.

.

OCTOBER, 1979

s:

ASSESSMENT OF GROWTH OF FEEDWATER LINE FLAWS MILLSTONE II 0 d W.H. Bamford

'

The purpose of this work is to estimate the future growth of a flaw located in the counterbore region near the feedwater nozzle safe-end-to-pip e vield. The flaw of interest has been confirced by UT to be apprcxirately 0.10 inches deep, and oriented circurferentially.

As a result of the location of this flaw, instrumentation was in-stalled to monitor the temperature fluctuations in one loop. Results shotted inat in a certain flow rate range the water stratifies, produ-cing significant stresses which are potentiall importcnt for crack growth.

The types of stratification produced were typical of those observed in other plants, but not as severe. The observed stratifications were classified under five different types, as shown in Figure 1. The tem-perature difference from top to bottom of the pipe for profile 1 was measured at about 350*F, whereas for other plants it has been found

'~~

^

-

to be as high as 450*F.

~'

.

.. .

A three dimensional finite element stress analysis has been completed for each of the five temperature profiles in Figure 1, and transient studies have shown that the five profiles represent limiting conditions compared viith the stress results obtained for any transient step in be-tvieen the profiles.

To acco ,plish a fctigue crack growth analysis, the system design tran- -

sients for corral, upset and test conditions were corbired t:ith the cycies ci stress from stratification, which occurs durir.g hot standby c craticn. As sho.:n in Figure 2, there are approximately nine cycles of varicus degrees wi'ich for the purpose of this coalysis, we will a r s.- ., occur er:" tire hot st:r.dby cccurs.

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A, tabulation of the cycle types used in the crack growth analysis, along with applicable stresses, is provided in Table 1. Tables 2* through 5 show the stresses at various locations around the pipe as a result of the stratification.

- The actual stresses from the three dimensional analysis were used for

>

'

the fatigue crack growth analysis, except in two cases, where compres-sive stresses far exceeded the yield stress in corression. The locat-ion is at the top of the pipe, and the condition occurs only when the pipe is nearly filled with cold water (profile 1) at low flow. For this case, tensile residual stress values were assumed to exist, equal to the yield strength. This is seen at locations 1 and 2 in Tables 2 and 4. This assumption is considered to be extremely conservative.

.

Crack growth was calculated at each of thirteen' locations around the pipe for periods of 1, 2, 3 and 4 years, assuming an initial flaw of 0.100

,_

inches deep, extending entirely around the inside of the pipe.

.

' A fatigue crac.k growth law which accounts for mean stress or R ratio i

(omin./ e max.) as - ell as the presence of the water environment was used. The law is shown in Figure 3.

Results of the crack growth analysis are shown in Table 6 for each of

,

the locations considered. These results show that the observe flaws will not grow significantly during the next years service. The final flaw size for the worst location is a factor of 5 smaller than the critical flew size for the pipe, us shor in figure 4.

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I APPENDIX A-NONMANDATORY Fq:. A-tM01 g

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E A* SURF ACE FL*WS J .40 -- (Water Reactor Environment) _.

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> i i i i 2 3 4 5 678910 20 50 40 50 60 70 EC 90 00 STRESS !!JTf t.'SITY Ft.CTOR R AfJGE 2, oK IKSI T I FATIGUE CRACK GROWTP DATA FOR SA.503, CliSS 2 ANb CLASS 3 AND SA 533, F4 #.E 2 GRADE B, CLASS 1 STE'ELS

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• m: s TABLE -1 TRE';51Ef;TS USED IN FATIGt'E CRAtr. GE C'.5 H A*: ALYS15

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• MILLSTONE II ,

I

.

.

Cycles Inside Surface (. tress Outside Surf..ce St-Min. Max.

l Description (40 years) Max. {

-

i Hot Standby 1 50 ) ~7 Hot Standby 2 1500 )

These stresses are dependent on circumferent' Hot Standby 3 500 ) position. See Tables 2 through 5.

Hot Standoy 4 2500 )

l 15000 10.47 7.71 8.49 7

-

Unit Load-Unload "

i Step Increase /0ecrease 2000 9.56 7.87 7.89 7 8.42 3.79 7.

Partial Loss of Flow 40 23.7 40

~

23.02 8.42 3.18 7.

Loss of Lo~ad 22.69 8.21 2.88 7.

Reactor Trip 400 11.23 0.0 10.04 0.

.

' Secondary Leak Test 200 l

27?

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I STRESS RESULTS - AXIAL DIRECTION ,

g CONDITION 5 HOT STANDBY f1 h

'

Inside Surface Outside Surface Min. Max. Min.

Location Max.

(Ksi) (Ksi) t 40.0 0.0 -40.0 0.0 1

40.0 0.0 -40.0 0.0

! 2 9.46 -23.0 8.56 8.29 3

35.12, 4.63 11.23 7.02

'4

.

68.97 - 1.43 13:20 4.66 5

66.05 - 7.28 12.61 1.71 6

46.17 - 8.06 10.83 -0.33

[ 7 7.27 7.34 3.69 8 24.37

24. 61 7.27 8.98 2.01

.9

'

2'3.67 - 2.47 6.47 -3.44 10 14.93 - 5.67 - 0.44 -7.05

. 11 9.30 - 5.44 - 6.03 -8.45 12

- 8.11 -8.69  ::::  : 2 1-13 7.62 - 4.95 . : li iA U' : :!:-

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TABLE 3 STRESS RESULTS - AXIAL DIRECTIO!4

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- C0!iDITI0fi 5 HDT STA!iDBY f2

.

Inside Surface Outside Surface

.

Min. Fa x. Min.

Location Max.

(Ksi) (Ksi) 9.78 9.62 3.24 1 13.49 9.68 9.40 3.25 2 12.57 9.34 8.56 3.24 3 9.46 4.63 3.21 7.02 4 8.76

-1.43 3.17 4.60

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7 7.69 7.27 3.22 3.63 8 7.60 7.71 8.97 3.37 9 24.61 8.02 6.47 3.63

-

10 23.67 8.44 -0.44 3.96 11 14.93 8.76 -6.03 4.20 12 9.30 7.61 8.86 -8.11 4.29 MI.: 1

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TABLE 4 l

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i - AXI AL DIRECTIOfi STRESS RESULTS

.

1. 3 CONDITION 2 + 1 - HOT STANDBY

. -

-

.

Outside Surface .

.

Inside Surface Max. Min.

Max. Min.

Location (Ksi)

(Ksi')

-40.0 0.0 40.0 0. 0 1

-40.0 0.0

~

40.0 .0.0 2

9 .31 8.30 16.19 23.71

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11.23 8.45 35.12 10.11 4 6.52 2.39 13.20 5 68.97 12.61 3.45 66.05 - 4.E5

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8 -8.39

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I STRESS RESULTS

- AXIAL DIRECTION f4 CONDITION 2 HOT STtrtDLY

. _

Outside. Surface Inside Surface Max. Min.

.

Max. Min. _

' Location (Ksi)

>

(Ksi) 9.73 3.24 21.51 9.78 1 .

9.60 3.24

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'

10.11 8.76

'

4 6.52 2.39 3.17

' 5 8.23 3.17 3.45 7.91 - 4.88 f 6 7.69 -11.04 3.18 -0.88 7

3.22 -5.64

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-14'.93 3.37

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J.63 -4.21 8.02 - 3.12 10 3.96

-

8.44 7.35 19.41 11 8.76 16.75 4.20 . . . ;j:: gi f

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36.44 12 13 41.23 8.85 19.41 4.28

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TABLE 6 P.ESU:.T5 OF TATIGUE CRACK GD.0'.:TH A*;f.'.YS: S

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INITIAL CRACK LENGTH = 0.100 INCHES L

Crack Depth After Year 2 3 4

.1

,. Location

.1008 .1016 .1024 .1032 1

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