ML20066K066

To Design Rept for Recirculation Line End Cap Repair,Monticello Nuclear Generating Plant

ML20066K066
Person / Time
Site:	Monticello
Issue date:	10/31/1982
From:	Charnley J, Eng N, Rich Smith NUTECH ENGINEERS, INC.
To:
Shared Package
ML20066J960	List: ML20066J958 ML20066K002 ML20066K066
References
IEB-82-03, IEB-82-3, NSP-81-103, NSP-81-103-R01, NSP-81-103-R1, NUDOCS 8211290357
Download: ML20066K066 (57)

Text

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Attachment (4)

Director I&E, NRC November 22, 1982 NSP-81-103 Revision 1 October 1982 30.1281.0103 DESIGN REPORT FOR RECIRCULATION LINE END CAP REPAIR MONTICELLO NUCLEAR GENERATING PLANT Prepared for:

Northern States Power Company Prepared by:

NUTECH Engineers, Inc.

San Jose, California l

Prepared by:

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REVISION CONTROL SHEET l

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REVISION CONTROL SHEET (CONTINUATION)

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CERTIFICATION BY REGISTERED PROFESSIONAL ENGINEER I hereby certify that this document and the calculations contained herein were prepared under my direct supervision, reviewed by me, and to the best of my knowledge are correct and complete.

I am a duly Registered Professional Engineer under the laws of the States of Minnesota and California and am competent to review this document.

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1 TABLE OF CONTENTS l

l Page LIST OF TABLES iv LIST OF FIGURES v

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1.0 INTRODUCTION

1 2.0 REPAIR DESCRIPTION 4

3.0 EVALUATION CRITERIA 6

3.1 Strength Evaluation 7

3.2 Fatigue Evaluation 7

3.3 Crack Growth Evaluation 8

4 4.0 LOADS 10 4.1 Mechanical and Internal Pressure Loads 10 4.2 Thermal Loads 11

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5.0 EVALUATION METHODS AND RESULTS 12 5.1 Code Stress Analysis 12 l

5.2 Fracture Mechanics Evaluation 14 5.2.1 Allowable Crack Depth 15 i

5.2.2 Crack Growth 17 4

5.2.3 Tearing Modulus 20 l

6.0

SUMMARY

AND CONCLUSIONS 34 i-

7.0 REFERENCES

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LIST OF TABig Number Title Page 5.1 Thermal Stress Results 22 5.2 Code Stress Allowable 22" End Cap 23 5.3 Crack Growth Cases 24 NSP-81-103 vi Revision 1 nutech EPeGWWEE385

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LIST OF FIGURES l

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1.1 Conceptual Drawing of Recirculation Manifold 3

i 2.1 Schematic of Weld overlay.

5 5.1 ANSYS itodel of 22" End Cap tield Overlay 25 5.2 Applied Stress Profile Through Limiting 26 Section 22" End Cap

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5.3-Weld Overlay Thermal flodel 27 l

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i 5.5 Crack Growth Residual Stress 22" End Cap 29 i

5.6 Stress Intensity Factor Versus Crack Depth 30 5.7 Crack Growth 22" End Cap 31 i

5.8 Allowable Crack Depth 22" End Cap-32 5.9 Tearing Modulus 22" End Cap 33 4

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1.0 INTRODUCTION

This report summarizes e' aluations performed by NUTECH to assess a weld overlay repair of the end cap to Loop A recirculation manifold weld at Northern States Power Company's Monticello Nuclear Generating Plant.

The weld overlay has been applied to address ultrasonic and radiographic examination results believed to be indicative of intergranular stress corrosion cracking l

(IGSCC) in the vicinity of the weld.

The purpose of the overlay is to arrest any further propagation of the cracking, and to restore original design safety margins to the weld.

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The required design life of the weld overlay repair is at least one fuel cycle.

The amount that the actual design life exceeds one fuel cycle will be established by a combination of future analysis and testing.

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Three crack indications have been found in the end cap weld heat affected zone.

Figure 1.1 shows the recirculation manifold in relation to the reactor pressure vessel (RPV) and other portions of the recirculation system.

All three crack indications are located in the 12 o' clock position adjacent to the weld NS P 10 3 1

Revision 1 nutggh

and all are axial.

The largest crack indication is approximately 11 percent of the wall thickness and 1 inch long.

The existing pipe material is ASTM A358, Class 1, Type 304.

The existing cap material is ASTM A403, Grade WP304.

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6.0

SUMMARY

AND CONCLUSIONS i

i The evaluation of the repairs to the recirculation end cap reported herein shows that the resulting stress levels are acceptable for all design conditions.

The stress levels have been assessed from the standpoint of load capacity of the components, f atigue, and resistance I

to crack growth.

J Acceptance criteria for the analysis have been established in Section 3.0 of this report which demonstrate that:

1.

There is no loss of design safety margin over those provided by both the original Construction Code for the piping system (B31.1) or the current Code of Construction for Class 1 piping (ASME Section III),

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2.

During the design lifetime of the repair, the observed cracks will not grow to the point where i

the above safety margins would be exceeded, i

Analyses have been performed and results are presented which demonstrate that the repaired weld satisfies these criteria by a large margin, and that the design life of the repair is at least five years.

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7.0 REFERENCES

1.

ASME Boiler and Pressure Vessel Code Section III, Subsection NB, 1977 Edition with Addenda through Summer 1978.

2.

USA Standard Code for Pressure Piping, " Power Piping", USAS B31.1.0 - 1967.

3.

ASME Boiler and Pressure Vessel Code Section XI, Article IWB-3640 (Proposed), " Acceptance Criteria for Flaws in Austenitic Stainless Steel Piping" (Presented to Section XI Subgroup on Evaluation Standards in September 1982).

4.

" Design Report Recirculation System Monticello Nuclear Power Station", General Electric Document Number 22A2603 Rev. 1.

l 1

5.

"NUTECH Reanalysis of the Reactor Recirculation Piping System," Letter to S.

J.

Hammer from G. A.

Wiederstein, GAW-82-014, File Number 30.2354.0003.

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Purchase Specification for Monticello Reactor l

Pressure Vessel, General Electric Document Number 21A1112, Revision 6.

7.

Telecon between NUTECH (J.

E. Charnley and N. Eng) and NSP (S. J. Hammer), " Weld Overlay Repair Program Technical Issues," dated October 20, 1982, File 30.1281.0001.

8.

ASME Boiler and Pressure Vessel Code Section XI, 4

1977 Edition with Addenda through Summer 1978.

9.

ANSYS Computer Program, Swanson Analysis Systems, Revision 3.

10.

Schneider, P.J. " Temperature Response Charts", John Wiley and Sons, 1963.

11.

NUTCRAK Computer Program, Revision 0, April 1978, File Number 08.039.0005.

12.

EPRI-2423-LD " Stress Corrosion Cracking of Type 304 Stainless Steel in High Purity Water - a Compilation of Crack Growth Rates", June 1982.

NSP-81-103 36 Revision 1 nutggh

13.

EPRI-NP-2472, "The Growth and Stability of Stress l-Corrosion Cracks in Large-Diameter BWR Piping,"

4 July 1982.

14.

NUREG-0744 Vol. 1 for Comment, " Resolution of the Reactor Materials Toughness Safety Issue".

15.

EPRI-NP-2261, " Application of Tearing Modulus Stability Concepts to Nuclear Piping", February 1982.

16.

EPRI-NP-1413, " Measurement of Residual Stresses in i

Type 304 Stainless Steel Piping Butt Weldments",

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June 1980.

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NSP-81-103 37 Revision 1 nutagh

novis fa; Attcchment (5)

Dirsctor I&E, NRC ovatber 22, 1982 11 42)

ENGENEE54B 6835 VIA DELORo e SAN JOSE. CALIFORNIA 93119 e PHONE (408) 629-9800

• TELEX 352062 November 18, 1982 NSP-81-022 Mr. Steve J. Hammer Northern States Power Company Monticello Nuclear Generating Plant Post Office Box 600 Monticello, MN 55362

Subject:

Leak-Before-Break Considerations for Recirculation System Stress Corrosion Cracks at the Monticello Nuclear Generating Plant

Reference:

See Attachment F

Dear Mr. Hammer:

The purpose of this letter is to document the leak-before-break aspects of boiling water reactor (BWR) piping system intergranular stress corrosion cracking (IGSCC) which have led virtually all investigators to conclude that it represents an availability rather than a safety issue.

The subject is investigated with particular reference to the recent IGSCC occurrences at Monticello, and it is determined that these occurrences do not alter this conclusion, nor do they reduce the design basis safety margin or increase the probability of an accident at the plant.

1.0 INTRODUCTION

Ultrasonic (UT) examination of recirculation piping system welds at the Monticello Nuclear Generating Plant have resulted in the detection of several intergranular stress corrosion cracks or crack-like indications in a number of welds.

All welds containing such indications have been repaired using the weld overlay process as described and documented in References 1 l

and 2, and therefore restored to at least their original safety margin.

However, co'nsidering uncertainties in the UT examination procedure, there is a reasonable chance that other, similar cracks may have gone undetected during the examination.

This possibility raises the question of the potential effect of such cracking on the continued, safe operation of the plant.

IGSCC has occurred in numerous stainless steel piping welds in operating boiling water reactors since 1974.

These occurrences were initially observed in 4-inch diameter recirculation bypass 5-1 l

l

Mr. S. J. Hammer November 18, 1982 Northorn States Power Company NSP-81-022 lines and 10-inch diameter core spray lines.

Since that time, however, IGSCC has occurred'in increasingly larger diameter piping systems, both in the U.S.

and overseas, up through and including 12-inch and 28-inch main recirculation system piping.

More than 400 cracked welds have been observed to date and recent cracking experience at Nine Mile Point, Unit 1 and Monticello will significantly increase this figure.

The c'acking has been the subject of investigations by the r

Nuclear Regulatory Commission (References 3, 4, and 5), General Electric Company (Reference 6), and numerous studies sponsored by EPRI and a BWR Owners Group on the subject (References 7 and 8).

The unanimous conclusion of all of these studies is that IGSCC in BWR piping, while undesirable from a plant reliability stand-point, does not represent a significant hazard to public health and safety.

This conclusion is based primarily on the exceptional toughness and crack resistance of the austenitic stainless steel from which BWR piping is fabricated, and the distinct tendency for cracks which develop in such material to develop into small, detectable leaks before any significant reduction in the structural integrity of the piping (leak-before-break).

The purpose of this letter is to summarize the bases for this conclusion from References 3 through 8, and to confirm its validity and applicability to the current cracking situation at Monticello.

This letter specifically addresses the potential for undetected cracks in welds other than those which have been repaired.

2.0 EFFECT OF IGSCC ON PIPING SYSTEM STRUCTURAL INTEGRITY 2.1 Net Section Collapse The simplest way to determine the ef fect of IGSCC on the structural integrity of piping is through the use of a simple

" strength of materials" approach to assess the load carrying capacity of a piping section after the cracked portion has been removed.

Studies have shown (References 7 and 8) that this approach gives a conservative, lower-bound estimate of the loads which would cause unstable fracture of the cracked section.

Typical results of such an analysis are indicated in Attachment A (from Reference 7).

This figure defines the locus of limiting crack depths and lengths for circumferential cracks which are predicted to cause failure by the net section collapse method.

Curves are presented for both typical piping system stresses and stress levels equal to ASME Code limits.

Note that a very large nutech 5-2

Mr. S. J. Himmer Novcmbar 18, 1982 Northern Statoo Powar Compcny NSP-81-022 percentage of pipe wall can be cracked before reaching these limits (40% to 60% of circumference for through-wall cracks, and 65% to 85% of wall thickness for 360* part-through cracks).

Also shown in Attachment A is a sampling of cracks which have been detected in servicd, either through UT examination or leakage.

In each case there has been a comfortable margin between the size crack that was observed and that which would be predicted to cause failure under service loading conditions.

Also, as discussed below, there is still considerable margin between these net section collapse limits and the actual cracks which would cause instability.

2.2 Tearing Modulus Analysis Elastic-plastic fracture mechanics analyses are presented in Reference 8 which give a more accurate representation of the crack tolerance capacity of stainless steel piping than the net section collapse approach described above.

Attachments B and C graphically depict the results of such an analysis from Reference 8.

Through-wall circumferential defects of arc-length equal to 60* through 300' were assumed at various cross sections of a typical BWR recirculation system.

Loads were applied to these sections of sufficient magnitude to produce net section limit load, and the resulting values of tearing modulus were compared to that required to cause unstable fracture (Attach-ment B).

Note that in all cases there is substantial margin, indicating that the net section collapse limits of the previous section are not really failure limits.

Attachment C summarizes the results of all such analyses performed for 60' through-wall cracks in terms of margin on tearing modulus for stability.

The margin in all cases is substantial.

2.3 Leak Versus Break Flaw Configuration 1

Of perhaps more significance to the leak-before-break argument is the flaw configuration depicted in Attachment D.

This configura-tion addresses the concerns raised by the occurrence of part-through flaws growing, with respect to the pipe circumference, before breaking through the outside surface to cause leakage.

Attachment D presents typical size limitations on such flaws based on the conservative, net section collapse method of Section 2.1.

Note that very large crack sizes are predicted.

Also shown on this figure are typical detectability limits for short, through-wall flaws (which are amenable to leak detection) and long part-tarough flaws (which are amenable to detection by UT).

The margins between the detectability limits, and the conservative, net section collapse failure limits are l

nut.e_qh 5-3

Mr. S. J. Htmmar November 18, 1982 Northern States Powar Company NSP-81-022 substantial.

It is noteworthy that the likelihood of flaws developing which are characterized by the vertical axis shown in Attachment D (full 360' circumferential with no through-wall component) is so remote as to be considered impossible.

Material and stress asymmetries always tend to propagate one portion of the crack faster than the bulk of the crack front, which will eventually result in " leak-before-break".

This observation is born out by extensive field experience with BWR IGSCC.

2.4 Axial Cracks The recent IGSCC occurrences at Monticello were predominantly short, axial cracks which grew through the wall but remained very short in the axial direction.

This behavior is consistent with expectations for axial IGSCC since the presence of a sensitized weld heat-affected zone is necessary, and this heat-affected zone is limited to approximately 0.25 inch on either side of the weld.

Since the major loadings in the above net section collapse analysis are bending moments on the cross section due to seismic loadings, and since these loads do not exist in the circumferen-tial direction, the above leak-before-break arguments are even more persuasive for axially oriented cracks.

There is no known mechanism for axial cracks to lengthen before growing through-wall and leaking, and the potential rupture loading on axial cracks is less than that on circumferential cracks.

2.5 Multiple Cracks Recent analyses performed for EPRI (Reference 9) indicate that the occurrence of rultiple cracks in a weld, or cracking in multiple welds in a single piping line do not invalidate the l

leak-before-break arguments discussed above.

l 3.0 CRACK DETECTION CAPABILITY IGSCC in BWR piping is detected through two means:

non-destructive examination (NDE) and leakage detection.

Although

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neither is perfect, the two means complement one another well.

This detection capability combined with the exceptional inherent toughness of stainless steel, results in essentially 100%

probability that IGSCC would be detected before it significantly degraded the structural integrity of a BWR piping system.

3.1 Non-Destructive Examination The primary means of non-destructive examination for IGSCC in BWR l

piping is ultrasonics (UT).

This method has been the subject of considerable research and development in recent years, and nutech 5-4 l

Mr. S. J. Hemmer Nov:mbar 18, 1982 Northorn States Power Compcny NSP-81-022 significant improvements in its ability to detect IGSCC have been achieved.

Nevertheless, recent UT experience at Monticello and elsewhere indicate that there is still considerable room for improvement, especially in the ability to distinguish cracks or crack-like indications from innocuous geometric conditions.

Attachment D, however, illustrates a significant aspect of UT datection capability with respect to leak-before-break.

The types of cracking most likely to go undetected by UT are relatively short circumferential or axial cracks which are most amenable to detection by leakage.

Conversely, as part-through cracks lengthen, and thus become more of a concern with respect to leak-before-break, they become readily detectable by UT, and are less likely to be misinterpreted as geometric conditions.

This argument is further enhanced by the usual practice of supplementing the UT inspection with radiography (RT) when large, UT indications are observed.

If a long UT indication is truly a geometric condition, it will be observable as density differences on the radiograph.

If, on the other hand, no significant RT density differences are observed in the vicinity of the UT indication (or if the density differences are abrupt and crack-like), the observed indication is usually diagnosed as IGSCC.

3.2 Leakage Detection Typical leakage detection capability for BWR reactor coolant system piping is through sump level and drywell activity monitoring.

These systems have sensitivities on the order of 1.0 gallons per minute (GPM) of unidentified leakage (i.e. not from known sources such as valve packing or pump seals).

Plant technical specification limits typically require investigation / corrective action at 5.0 GPM unidentified leakage.

Attachment E provides a tabulation of typical flaw sizes to cause 5.0 GPM leakage in various size piping (from Reference 7).

Also shown on this table are the critical crack lengths for through wall cracks based on the ' net section collapse method of analysis discussed above.

For conservatism, the leakage values are based j

on pressure stress only, while the critical crack lengths are l

based on the sum of all combined loads, including seismic.

(Considering other normal ~ operating loads in the leakage analysis would result in higher rates of leakage for a given crack size.)

Note that there is considerable margin between the crack length to produce 5.0 GPM leakage and the critical crack length, and that this margin increases with increasing pipe size.

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nut.e_ch 5-5 i

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Mr. S. J. Hemmar Novcmbor 18, 1982 Northorn States Powar Company NSP-81-022 t,

3.3 Historical Experience The above theories regarding crack detectability have been born out by experience.

Indeed, of the approximately 400 IGSCC incidents to date in BWR piping, all have been detected by~either UT or leakage, and none have even come close to violating the structural integrity of the piping.

4.0 CONCLUSION

On the basis of a large body of analytica] and experimental work, (References 3 through 8) which is brieflylaummarized in this letter, it is concluded that the recent IGSCC experienced in the reactor recirculation system at Monticello~does not increase the probability of a design basis pipe rupture at the plant.

This conclusion expressly considers the nature of the cracking which x

has been repaired at Monticello, and the likelihood that other, similar cracking may have gone undetected.

The conclusion is based primarily on the extremely high inherent toughness and ductility of the stainless steel piping material; the tendency of cracks in such piping to grow through-wall and leak before affecting its structural load carrying capacity (which indeed was the case in the defects observed at Monticello), and; the fact s

that as cracks lengthen and are less likely to " leak-before-break," they become more amenable to detection by other NDE techniques such as UT and RT.

very truly yours, O []f6(At&

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P.

C.

Riccardella, P.E.

d Senior Director l

l PCR:lak Attachments cc:

G.

H. Neils (Nicollet)

D.

M. Musolf (Nicollet)

B.

D.

Day (Monticello)

G.

T.

Krause (Monticello)

D. M. Vincent (Monticello) nutgch 5-6 l

November 18, 1982 NSP-81-022 ATTACHMENT A l

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.5 FIGURE 2-4 TYPICAL PIPE CRACK FAILURE LOCUS FOR 0 PART-THROUGH COMBINED THROUGH-WALL PLUS 360 CRACK 5-10

November 18, 1982 ATTACHMENT E NSP-81-022 TADLE 3-1 EFFECT OF PIPE SIZE ON THE RATIO OF THE CRACK LENGTH FOR 5 GPM LEAK RATE AND THE CRITICAL CRACX LENGTH (Assumed Stress a = 5,/2)

Nominal Crack Length Critical Crack Pipe Size for 5 GPM Leak (in.) Length I in, c

c 4-in. Sch 80 4.50 6.54 0.688 10 -in. Sch 80 4.86 15.95 0.305 24-in. Sch 80 4.97 35.79 0.139 5-11

November 18, 1982 NSP-81-022 ATTACHMENT F LIST OF REFERENCES 1.

" Design Report for Recirculation Line End Cap Repair, Monticello Nuclear Generating Plant," NUTECH Report NSP-81-103, Revision 1, November 1982.

2.

" Design Report for Recirculation Line Safe End and Elbow Repairs, Monticello Nuclear Generating Plant," NUTECH Report NSP-81-105, Revision 0, November 1982.'

3.

" Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants," U. S.

Nuclear Regulatory Commission, May 1979 (NUREG-0531).

4.

" Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," U. S.

Nuclear Regulatory Commission, July 1977 (NUREG-0313).

5.

H.

Tada, P.

Paris, and R.

Gamble, " Stability Analysis of Circumferential Cracks in Reactor Piping Systems," U.

S.

Nuclear Regulatory Commission, February 1979 (NUREG/CR0838).

6.

H.

H.

Klepfer, et.

al.,

"Cause of Cracking in Austenitic Stainless Steel Piping," General Electric Company, 1975 (NEDO-2100).

7.

EPRI-NP-2472-SY, "The Growth and Stability of Stress Corrosion Cracks in Large-Diameter BWR Piping," General Electric Company, July 1982.

8.

EPRI-NP-2261, " Application of Tearing Modulus Stability Concepts to Nuclear Piping," K.

H.

Cotter, et.

al.,

November 1981.

9.

Presentation by EPRI and BWR Owners Group to U.S. Nuclear Regulatory Commission, " Status of BWR IGSCC Development Program," October 15, 1982.

nutech 5-12

1 Attachment (6)

Director I&E, NRC November 22, 1982 Monticello Nuclear Generating Plant Determination of Reactor Coolant Leakage The Monticello Nuclear Generating Plant is provided with redundant and diverse methods of detecting reactor coolant system pressure boundary leakage. These methods include:

1. Equipment and floor drain sump pump timers.

An alarm is sounded when sump filling time is less than a preset time.

2. Equipment and floor drain sump level transmitters.

Sump level is displayed and recorded on the control board. The plant process computer computes sump level rate of change and a computer alarm is generated when the preset setpoint is exceeded.

These computer points provide rapid response to changes in leak rates.

3. Equipment and floor drain sump flow totalizers and flow recorders

4. Drywell pressure (13 - 17 psia narrow range)

5. Drywell temperature (seven points on multipoint recorder)

6. Drywell particulate monitoring and sampling system.

A moving particulate filter and a beta scintillation detector provide an extremely sensitive and rapid means of detecting reactor coolant leakage.

Leakage at very small rates can be detected. This is generally the earliest indicator of leakage.

Operating experience indicates the sump level monitoring system is capable of measuring leakage in the range of 0.1 gallons / minute.

This system is also responsive to changes in leakage rate of about 0.1 gallons / minute or better.

The containment particulate radiorctivicy monitoring system is extremely sensitive. The syste,cesponds to leakage before such leakage can be physically identified.

This system cannot easily quantify drywell leakage, but it provides a very early indication of changes in leakage. Response time is dependent on the leakage rate. Large leakage increases will l

cause the system to respond within minutes.

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