ML20062J167

Transmits Summary of Meeting Between EPRI SG Reliability Project & NRC on 920310 in Rockville,Md.During Meeting RES Staff Presented Questions Re Proposed voltage-based Alternate Plugging Limits.Response Requested by 920828

ML20062J167
Person / Time
Issue date:	06/05/1992
From:	Richardson J Office of Nuclear Reactor Regulation
To:	Mcdonald R SOUTHERN NUCLEAR OPERATING CO.
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ML20062J170	List: ML20062J167
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FOIA-93-66 NUDOCS 9206160161
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{{#Wiki_filter:+ JUN 0 5 ES2 R. P. Mcdonald, Chairman Utility Steering Committee EPRI Steam Generator Reliability Project c/o Southern _ Nuclear Operating Company P.O. Box 1295 Birmingham, Alabama 35201

SUBJECT:

ALTERNATE TUBE PLUGGING LIMITS

Dear Mr. Mcdonald:

This letter transmits a summary of the meeting bef. ween the EPRI Steam Generator Reliability Project (SGRP) and the NRC staff on March 10, 1992, at NRC Headquarters in Rockville, Maryland (Enclosure 1). During that meeting, staff from the NRC Office of Nuclear Regulatory Research (RES) presented a number of questions pertaining to the proposed voltage-based alternate w plugging limits. These questions are provided in Attachment B of Enclosure 1. The bases for these questions are provided in Enclosure 2 to this letter. These questions relate to the generic methodology used to develop the proposed voltage-based limits and, therefore, we request that the EPRI/SGRP coordinate the industry response to these questions. A response by August 28, 1992 is requested. Sincerely, , g__. 9 James E. Richardson, Director Division of Engineering Technology Office of Nuclear Reactor Regulation

Enclosures:

DISTRIBUTIM: As stated Central files JTWiggins JGPartlow EMCB RF/PF DET RF HConrad EMurphy KWichman JMuscara, RES GJohnson KKarwoski WTRussell

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N i p Krog 'i._f b f o UNITED STATES a' o 3 .E NUCLEAR REGULATORY COMMISSiUN [ WASHINGTON D.C. 20656 f

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%w* * * # / yy 0 71932 MEMORANDUM FOR: James E. Richardson, Director Division of Engineering Technology FROM: James T. Wiggins, Acting Chief Materials and Chemical Engineering Branch Division of Engineering Technology

SUBJECT:

MEETING

SUMMARY

OF MARCH 10, 1992 A meeting was held on March 10, 1992, with representatives from the Nuclear Regulatory Commission (NRC), Electric Power Research Institute (EPRI) and several utilities to discuss on-going work in the Steam Generator ~ Reliability Project (SGRP). The material presented by EPRI is included as Attachment A. - t Steam Generator Programs managed by EPRI have existed since 1977. The- ,i i programs macaged by EPRI provide member PWR owners with information on the. status of current research in the areas of steam generator (SG) degradation and methods to minimize this degradation. EPR!s SGRP, which has existed.since t 1987, has both U.S. and foreign utility participation... The major objectives of the work in the area of SG reliability include: 1) reducing lost capacity due to SGs, 2) reducing repair and maintenance efforts. associated with SGs, 3) reducing radiation exposure from work on SGs, 4) maximizing SG operational safety, and 5) extending SG life. Evaluation of industry trends have shown that average capacity factor losses due to steam generator problems have decreased and that the average age of the i SG at the. time of replacement has increased. Additionally, SG. tube plugging rates have remained steady although the dominant degradation mechanisms have changed. Based on the current rate of SG tube plugging, it is expected that f the average service life of SGs in PWRs will be 20 to 25 years. A number of these improvements are due to.research by both U.S. and foreign groups. Research activities, including non-destructive examination methods.for e? detecting and sizing of flaws, are currently on-going in the following areas M of SG degradation: 1) primary water stress corrosion cracking (PWSCC) at roll i 2?I . transitions; 2) PWSCC of hot worked alloy 600; 3)intergranular attack (IGA). jd and stress corrosion cracking (SCC) at tube support plates (TSPs), in sludge piles, and in the free span of the tubes; 4) circumferential SCC; 5) anti - j.O 1 vibration bar (AVB) wear and fatigue; and 6) denting. S gi In order to mitigate the rate of SG degradation, research is also being j E performed in the area of secondary side water chemistry. The research = ~; f 6)ii potential monitoring, 3) high hydrazine water chemistry, 4). IGA / S M' includes the following programs: 1) ' alternate alloy tests,-2) electrochemical j I O W testing, 5) alternate amines for AVT, and'6) CREV-SIM and MULTEQ computer l y 9 $ k codes. A draft revision to the EPRI PWR Secondary Water Chemistry Guidelines u ~; 7.' will be issued early in 1993 to reflect laboratory work and field experience $ =Ii d in this area. i . n.:.

I t J. Richardson Since tne dominant SG tube degradation mechanism has changed from wastage to SCC, utilities have begun to question whether or not the current depth based SG tube plugging limit is appropriate for all forms of degradation. Alternate tube repair limits are currently being proposed that utilize the relationship between SG tube degradation mechanism, structural integrity, remedial measures, operational leakage, and in-service inspection capability to define defect-specific repair criteria that maintain the safety margins contained.in Regulatory Guide 1.121 and ASME Code Section III. The industry is continuing its development of the SG Inspection Performance Demonstration Program in order to ensure a current knowledge base and demonstrated skill level by the analysts and to establish overall NDE system performance goals. Staff from the NRC Office of Nuclear Regulatory Research presented a number of questions and concerns (Attachmen* m regarding the details of the proposed' alternate plugging criteria. A ri m n se to these concerns was requested. lsi James T. Wiggins, Acting Chief Materials and Chemical Engineering Branch Division of Engineering Technology Office of Nuclear Reactor Regulation Attachments: A. NRC and SGRP handout B. Alternate Plugging Criteria Concerns cc: W. Russell J. Partlow R. Bosnak R.P. Mcdonald B.D. Liaw FOR DISTRIBUTION SEE ATTACHED WP FILENAME: g:EPRIGRP.MTG gp DET:ENCB DET:EMCB DET:EMCB ODET:EMCB w. 48M -

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M ..;j 1 ^ ATTACIDE.NT A ' INFORMATIONAL MEETING NRC AND STEAM GENERATOR-RELIABILITY PROJECT i ) .I March 10,1992 ROCKVILLE, MARYLAND

-f4ny 'o. ~y ; . "_.kqf'l s'$h$ ,+ 411AD, M. ~ 3 q >ugg epp, 4 ' Q(!fp? ::n sq:.. 1F+3@f?.BURST PRESSURE PARAMETER CURVES y: { EDM Slot Specimens (7/8 x 0.050 inch Tubing) 1 y ug;Q; q 1; YA 0.9 - 10 's ~. 0.8 - 4 e 0.7 - o Q. v-F<i - 0.6 - h= = = = m- =+/t = 0.5 h EQ 0.5 - =+=**=+n o.~w 20 - s 0+ 3--c Woc i 0 20 40 60 80 100 Measured Crack Depth, % Figure 10 - Results from NRC' Steam Generator Group Project Reponted in NUREG/CR-2336 and NUREG/CR-5117

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I IGSCC Round Robin Team MD - Bobbin 2.5 2.0 - .5 4b 5 1.5-a Ve E Z E 1.0 - = mw l--w 0.5 - 00 0 O 0.0 * = N =, 0.0 0.5 1.0 1.5 2.0 2.5 Measured Crack Length, in. Figure 7 - Results from NRC Steam Generator Group Project Reported in NUREG/CR-2336 and NUREG/CR-5117

o r t IGSCC Round Robin Team MD - Bobbin 100 80-o O i C1. c> 60 - O t S es = ,E 40-M O w &w 20 - I O$ 3--o-a>rmoc => 0 20 40 60 80 100 Measured Crack Depth, % Figure 6 - Results from NRC Steam Generator Group Project Reported in NUREG/CR-2336 and NUREG/CR-5117 x

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i TABLE I. 00 SCC Flaw Dimensions and Bobbin Voltages

• l Specimen Maximum OD Surface Haw Bobbin Renormalized Number Depth, %

Length, In. Voltage, Volts Voltage, Volts B-6348 26 1.41 0.32 1.16 B-46-02 31 1.06 1.00 3.61 F-10 37 0.25 1.62 5.85 F-15 38 0.25 2.38 8.59 B-30-10 38 0.53 1.37 4.95 B-6107 42 0.25 0.43 1.55 B-07 43 0.66 1.48 5.34 B-63-01 44 1.14 0.46 1.66 B-62-08 42 1.43 2.04 7.36 B-61-03 47 0.69 0.62 2.24 E-11-05 50 0.64 1.31 4.73 E-0747 58 0.45 3.44 12.42 B-6242 61 0.50 1.71 6.17 B-46-04 58 0.70 0.37 1.34 B-55-04 59 0.91 1.92 6.93 B-63-06 59 1.11 1.82 6.57 B-5947 76 0.81 2.22 8.01 E-1103 86 0.44 4.57 16.50 B-1048 99 1.09 7.24 26.14 Data from NRC Steam Generator Tube Integrity Program-Phase II 1 6 7 1

w M Points of interest are that longer (above 0.8") through-wall flaws exhibit low burst pressures and even relatively short but deeper flaws also exhibit low burst pressures. Figure 17 shows for example that a tube with 0.4-inch long through-wall flaw can withstand about 3,500 PSI pr:::tre but : 0.8-inch lona through-wall flew would withstand only about 1200 PSI pressure. A tuDe with a 0.5-inch long flaw at 50 percent through-wall penetration would withstand about 6,500 PSI pressure at 80 percent through-wall, 4,000 PSI at 90 percent through-wall, 3,400 PSI and the 0.5-inch long flaw,100 percent through-wall, would withstand approximately 2,500 PSI. Because voltage does not directly relate to crack length and depth for crack morphologies of interest, a good correlation between voltage and burst pressure is not expected. For example, short tight cracks (deep or through-wall) would produce a low voltage and a high burst pressure, however, a long tight crack (deep or through-wall) could produce a low voltage but also a low burst pressure. Similarly, a series of short tight cracks, axially aligned, with short ligaments in between would produce low voltage and low burst w pressure. Figure 18 shows a plot of voltage vs burst pressure used by industry in support of alternate tube plugging criteria. The 95 percent lower tolerance limit on the data is shown. Besides the large variability of a factor of two in burst pressure and one to two orders of magnitude in the voltage, the plot lacks data for the tight long flaws or tight short flaws, axially aligned that could produce low voltages and low burst pressure. 6 )

W Leak Rate / Crack Size Predictions and Variabilities Assuming for a moment that a) through wall cracks can be reliably detected and tneir length accurately sized b) the amount of crack growth, any changes in cracking mechanism and morphology, and growth outside of original zones can be reliably predicted and c) particular cracks will not approach critical sizes during the next operating cycle and are left in service, then these cracks must be monitored during operation to assure that they will not approach critical size. To accomplish this monitoring, leak rate measurements and specifications are established. Unfortunately, for the types of cracks of interest, correlations of leak rates to crack sizes and measured leak rates vs. predicted leak rates (from fluid flow and fracture mechanics models) show approximately two orders of magnitude variabilities, Figures 14 and 15. PLEASE NOTE THE PROPRIETARY NATURE OF FIGURE 15. The variability is due to several unknown or uncontrollable factors. The length of cracks varies from the inside surface to the outside surface, these lengths are not always known or easily measured in service; the leak path for IGSCC is variable and highly tortuous; cracks can be very tight and of variable tightness. Further, in service, cracks can become fouled with small particles and/or corrosion products and may be surrounded by support structures and corrosion products. Under these conditions it is difficult to relate leak rate to crack length (to assure it is below critical length). Furthermore, through-wall cracked tubes in steam generators leak very little inservice whether the cracks are short or long because of tightness, fouling, and constriction by corrosion products or support structures. So, critical-cracks cannot be distinguished from subcritical ones based on observed leak rates. Furthermore, the approach of cracks growing to critical sizes cannot be determined since very small changes in leak rates are expected in service. Finally, when hundreds or even thousands of tubes may be leaking a very small amount, how does one distinguish those tubes that have cracks of, or approaching critical size under MSLB conditions? How many are there? Small leakage does not necessarily mean short cracks. A recent draft EPRI report attempts to shoe correlations between leak rate and EC probe voltage, Figure 16. These correlations are used in submittals to support alternate tube plugging criteria. The log - log plot of Figure 16 shows very little correlation of voltage with leak rate. Five orders of magnitude variability is shown for leak rates at a given voltage and one to two orders of magnitude variability in voltage for a given leak rate. The correlation coefficient for this plot is reported as 0.73 which also indicates very poor correlation. This plot also lacks data for cracked SG tubes which produce low voltages. Burst Pressure vs. Dearadation The NRC tube integrity results indicate that tubes with short flaws exhibit more strength than tubes with longer flaws of the same depth, also tubes with shallow flaws can withstand considerably more pressure. Figure 17 shows plots from an empirical equation derived from the data for EDM notches and validated by testing of stress corrosion cracks. It is not surprising that tubes removed from service have exhibited high burst pressures; this can be predicted for short through-wall flaws or for other reasonably deep fi__,. 5

A. ? Uniform intergranular attack has been experienced which essentially produces a thinning from an integrity point-of-view. Other uniform intergranular attack progresses to a given depth, then is accompanied by cracking through additional depth in the tube. Still another form, callular intergranular attack, is a network of axial intergranular cracking connected by circumferential cracking. At different plants and within the same generators, axial and circumferential cracks have been found. In the tube support plate region, cracks have occurred in tubes within the thickness of the tube support plate but have also grown beyond the support plate region and axial cracks have also been accompanied by circumferential cracks. Although, at first, cracks may be noticed only within the tube support plate region, cracks may grow beyond the support plate region in time. Research studies have shown that even for materials that are difficult to crack, once cracks are initiated their growth is sustained and the crack growth rates are similar to those for materials that are more susceptible to crack initiation. To varying degrees, the crack itself may act as a crevice and growth is sustained. Therefore, cracks that initiated in the tube support plate crevice region could grow beyond the original crevice. After the different cracks are observed, the modes of cracking are recognized; however these cracking phenomena have not been predicted nor their occurrence easily controlled. The various mechanisms, causative factors and synergism between important parameters are not well understood. Even if the effect of some of the important parameters such as chemistry were well understood, their control in crucial locations such as in crevices is difficult if not impossible to achieve. Concentration factors, of different species under different conditions, from the bulk to the crevice as high as 10' to 10' can be expected. To summarize the above discussion, several modes of cracking have been experienced in U.S. steam generators and several modes can be experienced in a given unit. The cracking modes can change with time and cracking that might have initiated in a given region, tube support plate for example, can extend beyond that region. The mechanisms, interactions and causative factors are not well understood or controllable and the cracking phenomena were not predicted a priori. Laboratory testing of the same heat of material under the same environment and loading conditions produce crack growth rates differing by an order of magnitude. Conditions in an operating plant are not so well known or controlled and even higher variability in growth rates can be expected. Thus, changes in mechanisms, growth of cracks beyond given regions and crack growth rates cannot be reliably predicted. Therefore no assurance can be provided for cracks found during an inspection (even if accurately sized) that they will not reach critical size during the next operating cycle. Recent proposals'try to use a " voltage growth rate" obtained from consecutive eddy current ISIS as a measure of crack growth rate. As discussed previously, for _the cracks and crack morphologies of interest, there is no unique correlation between the voltage and the crack length or depth (parameters of interest to structural integrity) therefore the voltage growth rate cannot be used as a measure of crack growth rate. i 4

phase angle for detection and sizing of flaws. Recently a parameter has been l emerging, the voltage (or amplitude), as a measure of tube integrity. This parameter does not uniquely measure the length or depth of flaws, the critical parameters from a structural integrity point of view.' Table I shows data from laboratory produced part-through-wall stress corrosion cracks. For various crack morphologies of interest the voltage is not expected to relate to tube integrity for the following reasons:

1) For flaws of given width and depth a correlation exists of increasing voltage with length up to a flaw length of approximately 0.5 inch.

Longer flaws will not produce a larger voltage than this saturation level; this saturation of voltage for approximately 0.5 inch long flaws and longer is based on the coil design. 2) The voltage produced l can be related to the tightness of the cracks; if the cracks are tight enough, and conductivity paths exist, low voltage response is expected whether the cracks are short or long. Of course from a structural point of view the larger flaws are more important and the voltage parameter would not distinguish between them. 3) The voltage produced is insensitive to critical crack morphologies. For example a number of short, tight cracks (deep or.

• i through wall) axially aligned with short ligaments between them would produce a small voltage indicative of the tight short segments of the cracking.

From a structural point of view such cracking would behave like a long crack 1.e. tubes would have low pressure holding capability; under pressure the ligaments would join to produce critical length cracks and high leak rates. The voltage response from such cracking would not predict the structural integrity. Crackino Mechanisms and Growth Rate 4ariabilities The discussion on crack detection and sizing reliability indicates that I important cracks can be easily missed and those that are detected cannot be i adequately sized.. Even if.important flaws were adequately detected and-sized, the crack growth rates, both in terms of depth and length, are required in j order to estimate the crack sizes at the end of the operating cycle, before .:1 the next inspection, to assure that accepted cracks remain below the critical size by a reasonable margin. Research results show that variabilities of' one l order of magnitude can be easily expected.in crack initiation times and growth 2 rates for environmentally assisted cracking even under test conditions where j samples of the same material are exposed to the same environment, temperatures, stresses, etc. Much variation in the operating environment of steam generators exists for the' power plants in the U.S. Conditions of water l chemistries, temperatures and thermal hydraulics can differ from plant-to-j plant; geometries, crevice conditions, heat of material, temperatures, water chemistries, stresses, etc. can vary even within the same steam generator. As a consequence, ma,ny different types of cracks have been experienced-at different U.S. steam generators and.even within-the same steam generator.. Primary and secondary side cracking has been experienced. Cracks in tubes at-various locations has occurred such as in the tube sheet crevice,.at top of tube sheet, in free span zones, within the tube support plate regions, at l U-bends etc. Fatigue cracks, intergranular corrosion cracks, intergranular attack and crevice corrosion cracks have been experienced. Some of the intergranular cracking is associated with stress such as at dented regions, 1 other intergranular cracking is not associated with any significant stress such as at crevices in undented regions. Several forms of intergranular J attack'and combinations of intergranular attack and cracking have occurred. j 1 3 i

to ensure maintenance of structural integrity in cracked steam generator tubes, cracks must not exceed certain sizes during operation and tubes with cracks above these sizes must be removed from service. Cracks present in tu'oes during a M e n in:pc:. tion must not reach critical sizes during operation before the next ISI. This requires that cracks must be reliably detected and accurately sized, that the sizing errors are known and that crack growth rates (both in depth and length) are known for the wide spectrum of conditions and mechanisms that occur in steam generators. Furthermore, if cracks are accepted they must be monitored during operation to ensure that their sizes do not approach critical sizes which would place the tubes at risk of large leak or rupture during a HSLB. The information from the monitoring must relate directly to crack size. No single factor mentioned above by itself can assure maintenance of structural integrity, these must be applied together. Reliable crack detection and sizing is required,along with accurate estimates of crack growth rates and reliable leak rate / crack size correlations for monitoring crack evolution and stability during operation. Discussions related to these w capabilities follow. Crack Detection and Sizino Uncertainty Some of the most extensive research conducted to evaluate flaw detection probability (as a function of flaw size) and flaw sizing accuracy was the inspection of the Surry generator removed from service. Figures 1-4 show the flaw detection probability as a function of flaw size and flaw sizing accuracy obtained from EC ISI teams. Plots for all the teams and for the best team are shown. These data are for flaws found in the Surry generator i.e. wastage and combinations of wastage and pitting. These flaws are considered to be large volume flaws and easier to detect and size than small volume flaws such as cracks. It is expected that the performance for cracks would be even less reliable. To supplement the data from Surry, a round robin was conducted on a 18-tube box containing laboratory produced stress corrosien cracks. Sixteen tubes contained cracks of various lengths and depths. The depth of four cracks ranged from 25 to 40 percent through-wall, the remaining 12 ranged from 40 percent to through wall. The lengths varied up to 1.5 inches long. Although the total number of flaws in this test is relatively small, some trends are evident from the results. Four organizations inspected these tubes using several techniques a) the standard field practice techniques that met code and regulatory guide requirements and b) alternate techniques to represent the organization's best effoits and techniques. Figures 5-13 show some typical results. The probability M detection for these flaws ranged from 0.2 to 0.75 and on the average was ap,nroximately 0.5 for either conventional or aJternate techniques; the conventional technique used the bobbin coil while the alternate techniques included rotating pancake coil, array coil and an alternate bobbin coil design. Sizing accuracy was poor. The through wall flaw was sometimes missed and other times reported as a shallow flaw. Of particular interest was the poor length sizing ability even with the alternate techniques, where flaws up to 1.5 inches long were missed or sized at 0.2 to 0.5 inches. Note that some of these cracks are of critical length or longer and the EC would classify them shorter than critical length. The above discussion on flaw detection and sizing is based on techniques and parameters in common use. Multifrequency procedures, using amplitude and 2

s Enclosure (2) STEAM GENERATOR TUBE INSPECTION, INTEGRITY AND PLUGGING ISSUES The following discussion addr$tsses issues related to operation of steam generators with through-wall cracked (leaking) tubes. Two general areas are discussed; 1) engineering design philosophy and the policy of defense-in-depth and 2) technical issues related to assurance of maintaining tube integrity of cracked steam generator tubes during reactor operation. Enaineerina Desian Philosophy and Defense-in-Deoth General Design Criteria (GDC) of Appendix A to 10CFR50 require that the reactor coolant pressure boundary (RCPB) have an extremely low probability of abnormal leakage, of rapidly propagating failure and of gross rupture. Further, the RCPB is to be designed to permit periodic inspection and testing to assess the structural and leak-tight integrity. Using materials that exhibit leak-before-break behavior, maintaining leak-tightness of the RCPB and conducting inservice inspection (ISI) to assess structural and leak-tight integrity are important elements of defense-in-depth for maintaining safety and are not meant to allow operation with a leaking RCPB. The GDC indicate and the NRC staff has interpreted that through wall cracks in the RCPB are not acceptable. Several recent actions attest to this interpretation:

1) GDC 4 on exclusion of dynamic effects from ruptured pipes does not apply to materials susceptible to degradation; 2) ASME and NRC ruTis for evaluation of cracked stainless steel pipe do not allow operation with pipes containing through-wall cracks even though these pipes may exhibit leak-before-break.

Pipes with cracks deeper than 75 percent through-wall must be repaired; 3) NRC guidance for leak monitoring of RCPB allows for a small amount of unidentified leakage, however, if leakage is from a through wall crack, the component must be repaired; 4) NRR comments from review of a proposed revision to Regulatory Guide 1.121 required the guide to state that through-wall flaws of any type and identified cracks of any size are unacceptable. Since the steam generator tubes comprise over 50 percent of the RCP0 surface area and hundreds, even thousands of tubes could be leaking with an alternate tube plugging criteria, it is important to adhere to the policy of non-penetration of the RCPB. TECHNICAL ISSUES If it is decided that it is acceptable to operate a nuclear power plant with i the primary pressure boundary violated, i.e. with through-wall cracked steam generator tubes f.or the situation under discussion here, then a strong engineering case needs to be made and actions taken to assure maintenance of structural integrity. The important parameters relating to the structural l integrity of steam generator tubes are the crack length for through-wall cracked tubes' and the crack length and depth for other cracks. Cracked tubes can exhibit no leakage, small leakage or large leakage and burst behavior under normal operating and accident conditions. For through-wall cracked tubes, with axial cracks, the crack-length at which large leakage or burst occurs (critical crack 'ength) under MSLB condition is approximately one-inch. Various combinations of crack lengths and depths for part-through-wall flaws can lead to burst under normal operating or MSLB conditions. Therefore, 1

m m. N> 4 ~ -i 6. How well 'do we understand the various. mechanisms of cracking?. What are the causative. factors and synergisms? What assurance is there that cracking mechanisms will not change during. operating cycles? Why j wouldn't existing cracks grow beyond the initial: locations, i.e. outsioe of support plates for crevice corrosion cracking? What are the crack j growth rates to be applied to estimate crack length (or' depth) at the end of operating cycle? 7. What reliable correlations exist between crack length and measured or

• S predicted leak rates? How do leak rates measured inservice relate to crack length considering corrosion products, fouling, residual stresses, etc. which tend to restrict leakage? What changes in leak rates are expected and can be measured as cracks' approach critical lengths?'

t 8. In the Monte Carlo evaluations used to predict expected leak rates under. ~ normal operating and accident conditions, how are non-detections of through-wall cracked tubes considered? i [ -i f d 1 I-i s

ATTACHMENT B CONCERNS REGARDING THE ALTERNATE PLUGGING CRITERIA If 1; is decided that it is acceptable to operate SGs with through-wall cracked tubing, thereby eliminating the leak tight integrity of the Reactor Coolant Pressure Boundary (RCPB) as a very important element in defense-in-depth for maintaining safety, then a strong engineering case needs to be made to assure maintenance of structural integrity during operation. To maintain structural integrity, flaw length must remain below a critical length. Key issues in assuring structural integrity are knowledge of: a) the through-wall flaws present b) the crack length and accuracy of measurement c) the cracking mechanism and crack growth rate d) the crack size and progression from leak rate monitoring w Questions and comments related to these issues are as follows: 1. What is the probability of detection (P00) for deep and through-wall cracks as a function of crack length? Past experience and results indicate a low P00, 2. What is the accuracy of length and depth sizing? 3. If voltage is used as a measure of tube integrity, how is voltage related to length (and depth for deep flaws)? Voltage saturates as a function of length below the critical length. What is the voltage response for tight cracks - even if long? What voltage response and variation is expected for effectively long cracks made up from a series of short cracks axially aligned with small ligaments in between? 4. Tubes with short cracks, even if through-wall, will exhibit high burst pressures. However, tubes with deep partially through-wall flaws (85 percent and greater) and through-wall flaws approximately 0.6 inches long will exhibit burst pressures below the differential pressure experienced during a Main Steam Line Break (MSLB). Again, from the voltage, what is the flaw size for these flaws? In the burst pressure versus voltage correlation, were effectively long, tight cracks (expected of producing low voltage and low burst pressure) considered? 5. Regarding voltage versus leak rate, considering five orders of magnitude scatter in the data and correlation factors of 0.7 to 0.8, is it considered that a reasonable correlation exists? e

l i EPRllNPD) Sampies - Survey of Existing Samples - Review of Database of Pulled Tubes - Development of Sample Matrix ~ - Specification of Fabrication -- NDE Verifiable, Direct or Indirect Method (Short term, October 1992) 3 - Pitting, Thinning, Wear, Impingement -- Requiring Metallurgical or Qualified Sizing Technique (Longer Term,1993) - PWSCC, IGA / ODSCC EPRI/ NPDj Status of implementation Program Development - Appendices G & H Completed 9/91, approved by SGRP Tech. Advisory Group 11/91 - Implementation approach developed 11/91 and efforts at the NDE Center started 12/91 - Survey of available SG tube samples completed 1/92 - Development of IDB is underway - Development of training materials is underway - Development of interactive software shell for PD program Integration is under contract negotiation azi

EPRl/NPD) What's a QDA? o Analyst trained in all NSSS Vendor Plant Experience & data acquisition / analysis techniques Text Material - Laboratory ECT data sets o Demonstrated Analysis Sidils - Tested on a!! damage mechanisms o TNnning o Waar o (GA/ SCC o impingement damage o Pitting oPWSCC lNDEC 2/92" EPRt/NPO) APPENDIX G - ANALYST QUALIFICATION User Logan / \\ Frame *wer ECTAnalysis Training Wlodow Window _ Ratest Score t 60%on 50 Randomly Selected Quesdons No Ratest Yes M f ECT Data x s.e. j ea som., lPaes all? l Training Complete Updata Records I 'NoeC2/s2:

-5 EPRI/NPDj S.G. Performance Demonstration Schedule E SU NNU M UMN W M laneyessun WYHW md o b 3M beanse cessanus Commiereams Osmaammese auss%# = . m. e amuse assung Osseasumme Casement aus esas emannes Esammee EsuAng tsummun a Ramese Sammig seemed P we mee senese sens, assumes WW esame unem wummme i m h- %# h isha stelGene etsmusus Eamme hush to tasme tegems Laures aus $se M b true d Asuasass Summam tessummons 4 abusen er M Femeens w-m 'SGRPh W b h i b i O k I I I h t

y s EPv:.7-Defect Mechanism Specific Tube Repair Limits are not new to U.S. Industry ( Site specific applications airsady exist at some plants P* 1* F* Crtterlon. Sorne relaxation !! cracks are at a apacified distance below top of tube sheet. M pitthe deorsdation. Justifies a plugging Ilmit of 83% allowable wall loss for tubes arperiencing pitting. Wertsee. Just!fles a 47% plugging repair limit for steam generator tube thinnhg. J l i !unea Oapsac Me47 LPRknP j Defect Mechanism Specific Tube Repair Limits are Thoroughly Investigated and App!!ed in Europe in general, crteris7ndy, depending on defect mechan!am and SG e defect locatkm, invoke: RsRance origha# AtgM fot **%y based on " leak before break"(LBB) Resence on addtionalconstraint afforded by surrounding tubesheet or tube support plata. Reduced Tech Spee leek Emit Inspect for steam generator reliabt!!ty and plug tubes at a defect length that might be a problem durbg next cycle (La., account for growth rate) - Speelal hopection procedures and techniques may M used. Demonstrate that tube burst capability le not compromised including degradation of >40% thru-wall. Now moving towards simply a crack length based limit or even through war cracka -no absolute reliance on L85. !wac:pa

O EP ALNP j Alternate Tube Repair Limits INDUSTRY RESPONSE Utillze the rotationahlp between tube degradation mec hanism, structural Integrity, remedial rneasures, operational Isakage, and in service inspection capability to define defect-specific repair critaria that rnalntain Reg. Guide 1.121 and ASME Section m margins. !Nac sw m ms EPRLNP i Plant O arational Advantages of Defect Mec1anism Specific Tube Repair - Limits for Steam Generators l Avobd premsturWunnecessary tube plugging 4 Optimize steam gmerator evaltability ALARA principle consideratione Maintah flerbi!!ty for long-term repair options 1 Maximtze the available heat transfer ares Optimizing the cost-effectivonass of the steam generator repair and inspection program ! mac sw

~ _m_.--- \\ Status of Generic Defect Mechanism Specific Tube Repair Limit Documents PWR Steam Generator Tube-Plugging Limits: Technical Support Document for Expansion Zone PWSCC in Roll Tranaltions - issued December 1990, deals weh multiple axial cracks Updated (Rev 1) to include the possibility that circumferential cracking can occur issued December 1991. w Altsmate Repair Limits for OD SCC at TSPs, deals predominant!y eth axial cracking, to be issued by March 18,1992 a -,. EPRLNP j Alternate Stearn Generator Tube Repair Umits for Certain Defect Mechanisms Focus on True Measure of Structuralintegrity For Example: Eddy current voltage based plugging criterion for 00 SCC at tube support plates rnandates action be taken in a voltage range which can be accuratory rnsasured and depth is not a consideration . Voltage range is dictated by burst pressurs espability of the tube and Isakage considsrations with Reg. Guide 1.121 and 3 ASME Section IE Consorystism incorporated Crack length bened repair limit also ignorse depth parameter and ) deals with that which directly influences tube burst )

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- [ .A. - I 6 EPRtNPj U.S. Utility li.dustry Takes initiative to Define Generic Defect Mechanism Specific Tube Repair Limits Prepare industry reviewed and supported documents published by EPRI for Alternate Repair Lim!!s for specific defect mechanlems in susceptible PWR plants Document would be used with site.epecific submittal to NRC ' with appropriate modifications where needed (Individual utlittles will submit shortly) They are "living documents" being reviewed and updated by industry Ut!Ilty group willin the near future formally submit a request for a Tech. Spec. change to allow application of defect specific ' repair criteria as reviewed and approved by NRC Methodology for development of defect specific repair. criteria presented in Tech. Spec.. Coordination for document properation and updating performed by the Steam Generator Project Office of EPRI 1 !wacwe ce = =. i I , EPRl/NP i Industry "Ad-Hoc" Committee for Alternate Repair Limits for EZ PWSCC j ...m..... 4. . t ...'..=..u..... l ......!l::*. .o. W.....

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'[spavup3 Alternate Repair Limit for Tube Support Intersections Incurring Predominantly Axial OD Stress Corrosion Cracking (ODSCC) Tube Repair Limit Tubes with bobbin coli indications exceeding 4.5 volts will be repaired SLD t.sakage Criterion Prodleted SLS leak ratas from tubes left in service must be less than 55 gpm for sach S/G, including considerations for NDE unesttainties and CoSCC growth rates Elther a deterministic or probabittstic leak rate analysis may be performed ! wac sn EPRWP i Alternate Repair Umit for Tube Support Intersections incurring Predominantly Axial CD Stress Corrosion Cracking (Cont) Inspection Requkements A 100% bobbh cell inspection shall be performed for all + hot les tsp Intersections and a11 cold leg htersectione downlo the lowest cold leg TSP with CoSCC bdications All tubes w!Lh bobbin coilindications >1.5 voRs at TSP + Intersections shall be inspected using RPC probes. The RPC resuRs shall be evaluated to su dominant degradation mechanism. pport oDSCC as the !wnesie i 4

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!Nac$$3 --.n.e EPRLMP i EZ.PWSCC Axisi Crack Configurations That Should be Repaired Per The Guidance in This document h4eine.ede amat cracts tawI =he,e the da.el 6ength te gfeeter them the ,g ain. epee ru tregtei e4. As 6 i T tivisepw en at crocks i.1st spacmg a a12 a tviseen edewent es.at crats less sham 1:1 mm ad 4 em I tcp to qui and amai leegin greater than I/2 ihe allowathy c, . 'O ama etwk leegik lag 1e1Ai ! NaC $$3 Daldonc4 68 9 as

EPRLNPj Alternate Repair Limit for Tube Support intersections Incurring Predominantly Axial OD Stress Corrosion Cracking (Cont.) Operating Leakage Limits Plant shutdown will be implemented if normal operating leakage exceeds 150 spd per S/G Exclusions from Tube Repair Limit Tubes with RPC Indications not attributable to oDSCC and circumferentialIndications shall be evaluated for tube repair based on a 40% depth limit. ! Nac sta Attomate Tube Repair Limits Established for Expansion Zone Primary Water Strses Corrosion Cracking at Roll Transitions (EZ PWSCC) EPRt Report NP-ee6M, Rev.1 (December 1981) Repair t.imit based on crack length (not depth) Based on pubitshed tube burst data Azied cracka only Repak N eksuelecendal crack is indicated Restrict spacing between adjacent stial cracks Enhanoodinspecton 100% rotating pancake co# (RPC)ln affected areas Allowsnoe lor RPC uncertainty Anowanos for crack growth between inspeedons Reduced prtmary to secondary lesk rate ilmM (0,1 spm) to enhance lenk-before-brook Umit number of cracked tubes in servies to Emit leakage for postulated occh$ents Developed for fulkiepth tou transidons CorrecWon for tuboeheel constraint Appuceblo, but over-conservative, for partial depth ros transitions Appscadon of NRC Reg. Guide 1.121 safety tactors ! NRC ist

i i i EPRLNP j [ Future Actions Complete 3/4 inch tube tosting and data report (June 1992) 'l Only 3/4 inch data Combined 3/4 and 7/8 inch dets Complete 3/4 loch generic support document (June 1992) Not yet planned . - Initiate work for other mechanism specific degradation forms - (September 1992) +4 ; IGA witt. s?me SCC i Cracking at dented intersections Free span cracking (e.g., in sludge pile) i i ! NRC M2 ~~m ) J I i j ] l t -I 1 'j i 1 l a'l l 1 -) i 4 i ~ m m

yy 9 y i Program Objectives - Establish an indu"e.y-wide data analyst and training and qualisation program to ensure a current knowledge base and demonstrated skill level - Establish NDE system performance e.g.,- + procedure (technique), instrumentation, and Individual analyst - using a performance - demonstration methodology.- for alldamage mechanisms .w 1 EMllINPO* Development of Capability To implement - SG ECT Performance Demons' tration Appendices G & H - The industry is developing source materials and implementation protocol through EPRI at the EPRI NDE Center. - This capability (source materials ) will be transportable and intended for program implementation by utilities 'or other third partyL providers of qualification services. ( j

D ' EPRI I NPO,a 1. T Status of Industry SG Inspection Performance Demonstration Program i Mohamad Bshraveth Steam Generator Reliability Project Annual Meeting With The NRC NRC Offices, Rockville, Maryland March 10,1992 \\- 1SGRP ' EPRi / NPD "a

Background

The industry program for SG NDE Personnel Qualification and Performance Demonstration was last presented to the NRC on October 1,1991 The program was developed by an industry group representing utilities, NSSS, and ISI vendors and it is described in two Appendices to the SG ISI Guidelines; - Appendix G, Personnel Qualification - Appendix H, Performance Demonstration 'SGRP

PRI / NPD

• I implementation Task Groups Membership i

J. Benson, NEU G. Boyer, FP&L B. Curtis, ANA H. Houserman, Zetec R. Ingraham, W D. Malinowski, W R. Marlow, CONAM R. Maurer, ABB CE D. Mayes, Duke S. Redner, NSP T. Richards, B&W K. Wachter, RG&E s Chuck Welty, EPRI Gary Henry, EPRI NDE Center Steve Brown, EPRI Consultant Doug Harris, EPRI Consultant Mohamad Behravesh, EPRI Project Manager jSGRP) EPRl/NPOl i Industry Data Base o CentraRzed Data Base - Purpose is to prtmde uniform training and examination source material br empbyer implemented data analyst qualification and technique ax***vi using protocols established in Appendices G & H. o 108 kv5alized using EPRI developed source material with continued updates e.g.,105 makitenance, on a yearty baats o User Responseilities - Implementation (training & qualirmation) - Cartfication - Record keeping (norsentralized) NoEC 2/92l

'i l f EPRI/ NPD-h SG Eddy Current inspection ) Performance Demonstration t TRAINING DATA MATERIALS SETS N ~ INTEGRATION h SAMPLES {SGRP' EPRl/ NPDh ( Utility Oversight and Technical Input / Review - Overall guidance and oversight provided by SGRP Technical Advisory Group's Subcommittee 4 Chairmen - Technical input and review are given by individual industry experts through participation in various Task Groups and contracted efforts 'SGRP

i EPRI/ NPO) interim System - Computer-Based Analysis -ManualTraining (Book) -- Design and Operating Experiences (CE, E & B&W ) -- Acquisition & Analysis Techniques (CE, E & B&W) -- Pulled Tube Experiences aes-EPRUNPO) i Shell Software p .i o Integration - Text meterial t - Wrttlen questions - ECT data sets o UserFriener - Menu drtwen system access I O seculty - umns access a wneten questions and practical exarrenation data sets i No.c

P I EPRl/NPDj P industry Data Base o System built around Zetec HP/UX Eddynet system with optical l networking & storage i i o EPRI Prepared Source Material . Descriptive test for training M; - Written Questions . ECT data sets for training & Optical Disk r qualification lNDEC 2/92) i h rEPRI/HPD-3 Overview of Appendix G Implementation Goal: To Be in Place By October 1992 Annroach for Industry Database (IDBh Interactive Computer Based Training System (HP 8/or IMB) Fully Integrated by 1993 - Training Data Sets: CE, W, & B&W -6 - Qualification Data Sets: CE,W, & B&W - Retest Data Sets: CE,_W, & B&W t 'SGRP

r EPF<LNPi Alternate Steam Generator Tube Repair Limits An Industry initiative David A.Steininger Technical Advloor Steam Generator Project Office Electric Power Research institute i ! wac sw EPRbHPi Steam Generator Tubes are Experiencing Multiple Defect Mechanisms Primary water stress corroe6on crseking (e.g., within the tubesheet reglon, LA. EZ PWSCC), astal and circumferettal Secondary side stress corrosion cracking (ODSCC), arlal and circumferential @ st tube support plate intersections, and @ tube sheet region Intergranular attack (IGA), a volumetric form of attack (La., not crack Eke) Fretting and wear Secondary side wastage Denting Fatigue ! wnc sw !>4*8teMDt2

'::waai l j U.S. Plugging Practice To Date l in General Consists Of: I L Plug @ >40% thru wall degradation (Ls., depth) as indicated by eddy current using " bobbin coll" phase relationship Originated during the days of steam generator " tube wastage"as the tube defect mechantam Plug all" crack-Ilka" Indications as indicated by RPC Interrogation M

PC Inspection " triggered" by a bobbin coli distorted phase signal No special snandated Inspection technique other than bobbin coR a 3% sampio plan la indicated

. No change in " Tech Spec" leak limii ~~, _j Nec see EPRLNPi ) Multiple Defact Mechanisms Have Resulted in NDE "Falso Calls" and Excessive Tube Repair i k PROBLEM: For some tube degradation on mechantams (e.g., short, arlal

i challenges current HDE technology Moreover, tube burst

j strea corrosion cracka, IGA), the 40% repair criterton tests Irwe tte that, for these forms of tubing degradstion, IntegrRy t argins are maintained will beyond the 40% thru-weg limit I Can lead to extended Inspection campaigns and to tubes being removed from sarvlee unnecessarity (e.g., recent experience of Portland General Electric at their Trojan pant) !nacsse

i j EPP8HP3 3 f Alternate Tubing Alloy Comparison Update - i Results Almost + 100 references have been identified. Most references dealt with SCC testing in 10-50% caustic, pure / primary or AVT water, and pitting environrnents. >11 of the references had comparative results for alloy 690 TT vs. 800 mod. for 12 of the 19 corrosion issues. >18 of the references had enmparative results for 690 MA or Tr 3, vs, r,00 MA or Mod. for 15 of the 19 corroelon issues. Corrosion issues of pitting in chlorides and wastage in sulphates previously thought to be unsupported by data were found to have some references. ' NRC.3/92 t 4 i Retative Rrektu of the CorToedon Feststance of Aflows see,691300 and Steinless Steel s ' ~ seeer ese aser ese emer ses e3 F e*mmans * * \\ es e TT ff ne OS 4 Erges Ca6 ErnstJun i. LEWrtff / J &AM s i I I J (8) R4BS <5pq Waal I I I (II (JB wp'Isr 4TT) 1 I I 3 5 3 CaesIIe " s psame es / r J 3 I 3 5 NNW _/ e / i 4 3 3 3 5 t J. wour t e W. A AW 4 3J 3 4 i esitt I. (j a Ftspe edq)3 s 4 a J I (3) 5 4 5edster esegesas i Adg agm J &S I J 5 4W m i I I (ID 43D j 44$ FWSWWW 6 Am&M compagne 3 -(J) J J 45) "] 5. u.s 3 . Agt 4 44 5 J (4) PW {w &TT) 3-4 (3 I 5 . Amissue 34 (J 45 5 5. " w a. Cortestet i 5. Act J-4 3..l 33 (4.y) is.yp j 3 Am msg 4 13 i 5 4 if) U. Ptlema e m 34 J-d 3 J 4 D. We RW 'i I. Pteestates (J) U f31 43) (J) 3 3essges J J 3 (J) (II P W 3.Onel 8 Einseu 9 h EH88MM8 IRA e150 - TI o Thumumaby Tummes 4 i

J EPRLNPDl T I s Planned Secondary Chemistry Guldelines Revision 1 Laboratory work and flaid experience warrants another revision lisms to consider include: + Use of advance &altamate amines for iron minim!zation Anion 40-estbn ratlo as a crevice pH Indicator - Use of plant chemistry mods!!ng to develop site-specifk limits . Orygon vs. ECP of isodwater - Other monitoring Improvements Elevated hydrszine operation w Opt!mtzstion of hideout return processes Sodlum phosphate chemistry treatment poss1ble AVT phosphate conversion IAodifiestbos to boric acid treatment practice Draft should be sysliable by 1/93 INRC412

j EPRl/NPDl Crevice pH vs Number of Outages [ (Cumutenve Prompt Hideout Resum) ,o. 5 0 <3 3-4 4-55-66-7 7-8 8-99-80 10 cre,,ee e

NRC492 j EPRLNPD E f

Status of Alternate inhibitor Work for OD IGA / SCC Boric Acid laboratory work is finished Fleid effecthroness is a matter of major concern 11tanlurn com nd studies have passed the screening tests for hhlbRion of C, and are now being tested for solubility, stablDty and will shortly be tosted in heet transfer devices other Inhibitors: are still undergolog screening teste e . Modes of Applicatlon Being Considered. . Off.Ilne oceking with more concentrated solutions . On line addklons . Colloidaldispersions ,NRC492 i

jEPRUNPDE First Model Boller Test of Titania-Silica Get Compound 25 day test, csunt!c with titania-silica inhibitor P . white deposits in asch crevice and in sludge . distrlbution and compos!!!on of crevice deposits Weight % Top 1rJ Middle Bottom 1/3 T1 0.68 0.34 0.29 S10 2 3.2 2.8 3.2 Na 0.18 0.84 0.18 . Titania has sufficient solubility to penetrate packed or eccentric crevices. . no deersdation observed on tube surface , NRC 112 _ m,, i j EPRLHPD l f Alternate Tubing Alloy Comparison - Update - Approach Revlawrecord au personal and EPF5 referencee Reviewtecord aR KWU Laborefec and Clemat provided documentat6on Creets new corrosion issue categories on table !! required Assign I!!srature referonce numbers to each of 19 corrosion Isopes/ancy categories Confirm whether each category is or is not supro ted by references, and then attempt objective ranking of siloys [ NRC.112

.? 1 EMLit."D I Cumulative Concentration in Crevice SG A(Lee 1) / / I. , / 1 1 u U sans g.- .. n -- a w :... .-e. r-- - _:. i.fi. ..m = e ' NRC.342 I ,J EPRVNPD l Crevice pH f SG 1 u Le = g l u - a n I u 1 n u t u 1 u Iu u Et - .. m. ..__.m SWW iWW H/W 81/ES f t/e9 94/ 94 95/ I6 44/38 ba86 8 em WPt e et Ers e saTse a a gre a a fys

9 j EPRLNPD 3 ( CREV-SIM & MULTEQ Cycle Chemistry PROGRAM OBJECTIVE: Develop an approach for making real tirne crevice chemistry predictlone from routinety measured olowdown chemistry parameters, leading to improved prodletion of corrosjon mode and providin damage.g the basis for correlating blowdown chemistry to corrosion TECHNICAL APPROACH Assume crevice hideout is controlled by local bolling rate in crevices Total crevice hideout rate is known from tracer injection tests and la prcportional to blowdown concentration. Distribute total hideout to crevices based on available superheat Integrate crevice inventory over time Calcutste and trend best pH using MULTIQ '. Correct inputs based on observed hideout return ,NRC112 t j EPRLNPD 3 ( BLOWDOWN CHEMISTRY o PLANT CREV-SIM MULTEQ DESIGN e w PARAMETERS o CREVICE CHEMISTRY PREDICTIONS j NRC.112 )

~ l 1 EPRLNPD Alternate Amines for AVT l Program Objective . To reduce iron transport and erosion corrosion in single and two-phase flow + To provide laboratory and loop test data showing the beneficial and side sffects of advanced AVT using low-votatility amines Results Achieved . Morphothe.AVTin wide sprecd use. Has reduced Iron transport by 2 3 t!mes in U.S. PWR*s. . Five bop tested smhss appear to be acceptable substitutes for ammonia or morpholine . Ethanolambe best of the five aval!able commercially w 1,2 d!ambo ethane best for copper-free cycles but tends to strip modlum from condensate polishers status Umited fleid trials arpocted late 1992 or 1993 [ NRC.112 } EPRLMPD j f Summary of Amine Properties from Loop Tests Amine Mot.Wt PPM @ 9.1 Statmt.lg % Decomp natie h loop @ 2ss c.tr ETA 81 2 -0.34 -2 DAE 60 0.9 -0.41 -30 AMP 88 1.1 - 1.2 -80 3CHQ 127 2.4 -0.2 -7 MPA as 1.2 -1.8 -4 MORPH 87 5.2 - 1.2 -9 NH3 17 0.8 -3.5 0 e.

h g } EPRt/N >O l i ( l } ... e. ....u . e. .o... m ..u s... c.. .n... ee no se se se se se te se .is.. ... i......... s i u. e..... ...,....... o u. s.i c... ' NRC492 l 2 t t l i j EPRLNPDE i I 1 Boric Acid Effectiveness 'I Soric acid continues to be recommended to mitigate secondary side { caustic induced IGSCC and denting In order to maximks boric acid effectiveness a to minimke and lower 30 corrosion potential are attil r i Elerste feedwater hydrazine to 50-100 pp6 ff possbis i ) P Ify remove all copper components and chemically clean r I copper from secondary system - . Males feedester pH ss high as practical to minimtre transport of ({ homethe and magnetite . Tube's should be removed from the SG's to regularfy aseees tube. . bundle condition.' t! I ,' NRC.3/92 1

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o j EPRLNPDl ( PWR Steam Generators Secondary Side Chemistry, Corrosion & Materials issues J P N Paine A R McIlree w T. O. Passell C. R. Wood P. J. Millett EPRI-NRC Meeting March 10,1992 >NRC412 i j EPRLNPDI Secondary Side Topics Boric Acid Effectiveneas. survey tosuits Altomate ambs treatment update CREV SIM and MULTEQ Experience Status of work on ARemats inhibitors Attemate tubbg afloy studies . Planned secondary chemistry guidelinea revision y y C.ts2

F 4 - i EPRLNPDf . SECONDARY SIDE CHEMISTRY ISSUES - IGA / SCC Inhibitor Testing. Inhibitors tosting le following EPRI and U.S. leed. ' ~ Boric acid and sodium phosphate being tested. Model boilers are being readied for use in Sweden and Spain data that has been used to validate high) temperature compu codes. i 2 --, EPRLNPDi

SUMMARY

EPRra on going world-wide Information exchanges provide considerable insight into 50 PWR operatione. I 2 -- 1

I 1 O EPRLNPO j SECONDARY SIDE CHEMISTRY ISSUES Electrochemical Potential Monitoring Test programs in Germany, Sweden and Finland. PWR Power plant installations at Bibile, Ringhals, Lovisa, and at many BWR*s. Conciusions: Probe Ilfe about one cycle. Valuable insights into oxidizlng transients. Can be much more affective than O. monitors. ECP rnessurements suggest need for higher hydrazlne operatiott VTT probe combines ECP, pH and high temperature conductivity. RI-Utility tests are planned for late 1992. ! NRC/3/92 LPRLhPOi SECONDARY SIDE CHEMISTRY ISSUES High Hydrazine Water Chemistry KWU recommends 70 - 100 ppb Hydrazine in blowdown based on fleid and laboratory ECP messurements. Studevt recommeMe high hydrazine, but level can be plant speelfic. Ringhais 3 and 4 operate with 70 and 50 ppb hydrazine in feedwater respectively. Akneraz operates with about 100 ppb hydrazine in feedwater. Japanese Units: 400 500 ppb hydrazine in feedwater at Oh! Taxahams, Genkal sites. KWU and SSPB recommend high hydrazine.high ammonia to protect against erosion corrosion. Much simpler than aminee chemistry and c!almed as effective. se au >I NRC/3/92

EPRLNPD j AVB WEAR / Fatigue Not a serious problem in Europe, but has led to AYB replacement in Behlum and plans for replacement in one Spanish unit. EDF has measured average wear growth rates of 2% of wall per year. In one tube growth rate was measured at 14% AV8 wear is serious concem for Japanese units. After M!hama 2 avant all units have had or w!!! shortty have AYB's replaced with a new MM Westinghouse daalgn. g Tube fatigue is primary cause for reptscement at Mtharna 2. Tube fatigue is a msjor concem for Ontarlo Hydro at the Candu Bruce station. Fatigue cracks grow slower at Bruce than in PWR S.G.s. Raason not understood. JP9 Att 'I NRC/3/92 EPRLNPD i OCCURRENCE OF AND REMEDfAL MEASURES FOR: DOfTtHQ Dentktg is contbuing in several Japanese, Korten, Seiglan and Spenlan units. 10 or 00 Crackbg at dented intersections has been very minimal PWSCC in alloy 600 TT in dented tut >ss has been observed in new French 1300 MW unks. se me 'I NRC/3/92 .. a

EPFwr PO i IGA / SCC at TSP's,in SLUDGE PILES, on FREE SPAN TUBES (CONT) Romedial actions: In Japan, Spain and Sweden: Use of botic acid, increased hydrazine levels and snodifications to chemistry control systema. Effsettvenesa not apparent h some cases. Chemical cleaning has been used and studge lancing is used. Effectiveness not apparent in some cases. In France: Tighter blowdown limits on sodlum, more efficient sludge lancing. Full height chemicalcleaning and use of boric seid are being evaluates. In Belgium: 11ghter I!mits on rityingresa and periodic sludge lancing. Chemical clean og Tlhange t several years ago and of Doel 4 planned. Use sodium phosphate is behg consloered. Japanese still believe that SG's with ear er sure to sodlum phosphate are much less suscepuble to SCC. W I NRCrN92 EPRLhPO ' 0D CIRCUMFERENTIAl. SCC Affecting two units in Belgium, three units in Spain and several unfts in France encountared in the roll transition but ateo in the free span Iley !!mit SG life in some units. ISI Activities b Europe: Serious concern with IGA / SCC in sludge pDe. Extenalve and frequent inspections. Belgium and France ushg absolute rnode methods since detection is very difficult with dmerent!al mode bobbin coil or RPC. Altsmata plugging critoria are in place, based of correlations between absolute mode voltages and burst strength , NAC/W92 I

P 'I EPRONPOg 1GA/ SCC at TSP's, in Studge Piles, on Free Span Tubee Growing problem at many units b Spain, France, Belgium, Sweden and Japan. Affecting 17of 20 French units with LTRAA tubbg and drilled hole supports. At supports: a major problem only for drified hole unha, fleplacements: Primary cause of planned Cankal t, Chl1 & Takahama 2 replacements. 9 Primary cause of St. Laurent 81 posalble replacement. Prfrnary cause of T1hange 1 posalble replacement. Significant cause in Almara21 & 2 and Asco 1 & 2 planned replacements. Signifleant cause h Ringhala 3 possible replacement. or nr INA0/312 w EPRVNPOi IGA / SCC at TSP's, in Studge Pfles, on Free Span Tubee ISI Activltlee in Europe: Extensive inspections, no safety problems at supports. Revloed 91 criteria based on an allowed depth of 70% and a thre bobbin coil voltage H Sweden. In France and Se6 glum plugging crRorie bemed on bobbin cou voltages and pulled tube burst strength. Take creds for TSP. Cause of attadt not known in a majority of cases Assumed caustic in Ff ance, with lead irrvolvement h several cassa, Hideout return chemistry data is not sysllable for prior years examination. Lead Imrofvement shown at Ontario Hydro At Doel 4, local chemistry still questloned but lead le present. In Spain, tube examinations have shown avidence for acid or akall at different units. Japanese assume caustic for initiation in all cases. I. NAC/312

9 EPRL,NPDf PWSCC at roll tranaftlons (cont.) 1 Remedial Actions: - Roto and shot poening performed after Initial start up is believed to have slowed occurrence with alloy 600 LTMA but not to have halted it. Due to cracking with kise rolled alloy 600 TT tubing, EDF la shot eening their nawer SGs. hot Reduction at Ringhals 3. Growth rate reduced. Detection rate apparently not affected. . For part depth roll units: Japanese hydraulic + roil re expansion and shot peoned has been sucesssful(several cycles) For part depth roll units: At Doel 2, rewrrpansion followed by nickel plating on 10 have been satisfactory too date. (one cycle) Nickel plating appears successful to prevent initiation and growth. (severalcycles) Sleeving la seldom used in Europe: expenstva, interferes with inspections, can't repair above the sleeve. In Japan sleeving is common at several elevations in each sleeved tube. Stress reflef heat treatment of u-bands appears successful to date. 1 NRCTS92 mu EPRt/NPD i i PWSCC of hot worked alloy 600 PWSCC of hot worted alley 600 has occurred h Mechank:al tube plugs Presourtter nozzles Controlrod drive mechanisms Studies are beIng initiated to srplora crack growth rates. Rank susceptibility of different heats Finita element analyals (CROM cases) I NRC/3/92 eu

n* .t Y ~. i EPhthPQ j SECONDARY SIDE CHEMISTRYISSUES l s Afternate alloys tests Electrochemical potential monttoring i High hydrazine water chemistry V IGA / SCC Inhlbhor testing I e I, NRC/3/92 l r LPRLNPD i PWSCC at roll transiflons Major problem at many unks in France, Belgium, Sweden and Japan Especia#y a ivcLT. for plants with kiso rolle. - Severaf plants operate with 20% or more of tubes with roll transitione cracked through well Repiscemente: - Primary cause of Demplerte 1 replacement. j - Primary cause of Doel 3 and Bugey 5 planned replacemente. - Major cause of Ringhale 3 possible replacement. - Signficant ceues m Almaraz 1 & 2 and Amos 1 & 2 planned roptooemente. ISI ActMtleein Europe: - Extenehreand RPC inspectione using detect le plugging criteria., no esfety probleme, a me -i NRCfktt

b EPRLhPD i Recent Steam Generator Experience Europe-Canada-Asia Year End: 1991 NRC EPRIStatus Meeting March 10,1992 i w as 'tNRCm92 1 EPRLNPD i SUBJECTS COVERED OCCURRENCE OF AND REMEDIAL MEASURES FOR: PWSCC at roIItransitions PWSCC of hot worked alloy 600 IGA / SCC at TSP's, in sludge p!!es, on free span tubes j . Circumferential SCC. AVB wear /Fetigue.,. ) Denting maa -fNRC&92 'l

i i 't CFPLNPD i i Mafor Tasks and Key Contacts 1 .j -I EPM carames Task Desertation laggfsmag i R. Jones (4154ss-27901 Stretag Deression R. P. IdeDonald

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G. Srthh (4154664104 ThenneW4ydrouges R. Pearmen (012-351131) ene vibranen %) IL Sehrsveen (4154ss 23es) troervice me.ecelen . J.Ernan(7164es.27eep i D. Staharger(41mte A m AW P. h(91m77sep [ enn ns.ieser,.r. j 5GRP NRC [ -_-1., g i l -L b 6 .I i e h f

F.PRLNPD i Assessment Based on current plugging rates, the aversgo service life of original equipment staam generators in PWRe worldwide is projected to be 20 to 25 years. The average ago of the original equiprnent unita in U.S. plants is now -13 years - Most original equipment sisam generators will have to M be replaced within the plant life. Many U.S. units will have to be repisced within the next 10 to 15 years. N ] __w

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EPRLhPD i Planned R&D Response . To mhlmize future Industry costs, additional R&D is planned aimed at the development and det!very of technology to Further artend the !!fe and increase the reliability of original equipment stsam generators Reduce the duration of replacement outages and the associated rsdlatlori exposures j Assure that replacement units can meet ut!!!tles'ilfe and rollability goals 1 l 1 l SGRP NRC( l 1

EPRtlNPD i Percentage of Steam Generator Tubes Plugged Due to Denting in U.S. PWRs t Percentage of Steam Gsnerator Tubes Plugged Due to Denting in U.S. PWRs ir.rcenIT*tD serva. . se t t anoonnnnn ,, 1 EPRbHPDi Percentage of Steam Generator Tubes Plugged Due to OD SCC /lGA at Support Plates in U.S. PWRs 4PerceaY*t in s.<v= l ..e a . R3 = .a-I s,:: QOn 75 's 77 7. to.. si sa..... .s .e '_ Year l SGRP NRC I i j

i l t EPRLNPD j ) Steam Generator Experience Trends l Average capacity factor losses have decreased from 5.3% (19791984) to 2.5% (19851990) and the average age of units that have had to be replaced has increased from 10.4 years (replacements thru' 1964) to 13.4 years (replacements since 1985) Stsam generator tube plugging rates have remained steady for several years but the dominant degradation mechanisms have continued to change - Some problems (e.g., denting) have been solved but other probleme (e.g., stresa corrosion cracking at tube support plate Intersections) have emerged and are rapidly increasing in importance. l SGRp NRC( l [ EPRl/NPD j Percentage of Steam Generator Tubes Plugged in U.S. PWRs T.m ri.ee.s 1,.... ., 1... s O F5 < FS ?4 77 70 79 80 44 43 83 84-SS OS 87 - 80 St De YEAll l SGRP.NRCl .j m cm.. I J

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EPRtHPD j -) SGRP Performance Measures f+arnbar of tubes plugged / year 74Jmber of tube leak outages / year 50 related espacity loss / year a Number of tube ruptureatyear 50 ege at replacemert . Replacement outsgo duration and radiation exposure Service !!fe of replacement generators l

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i Forelan Utility Particination Technical Exchance Aareements i

1 111111 1 CEGB CRIEPl EdF Electronucleaire - ~' Ontario Hydro Spanish Utilities Swedish State Power Board. j Q ) 1 son,.une r 1 unowro, i SGRP-NRC MEETINGS . Annual Meetings on Program Status (5/13/91). . Special Topical Meetings j - ISI Performance Demonstration (10/1/91)- - Tube Repair Criteria - ACRS (11/6/91) 1

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5 i EPRWPDg U.S. Utility Participants in SGRP All EPRI. mornber PWR Owners (33) - Current EPRI non-rnembers are: American Electric Power Consumers Power Southern Califomia Edison Virginia Power -{T55.N m-EPRLNPo j Executive Groun Grant Bamon Nonheast Utilities R. P. Mcdonald (Chairman) Atabama Power Steve Rosen Houston Ughting & Power Donald Schnos Union Electre Company 1 Robert Smith (Vee Chairman) Rochester Gas & Electre l Bart Withers Wolf Creek Nuclear Operating Corp. James Zach Wisconsin Electnc Power l SGRPMRCl w =

l / EPENPDj - - - ' ' -T I j Budget j Steam Generator Reliability Project 'j ~ i. Year Budget S Millions 1987 4.4 1 1988 4.4 l 1989 5.2 1990 4.7 1991 4.3 1992 4.4 ) i SGRP NRC[ 1 LPRL%90 i Stoem Generator Re!!abl!!!y Project 8 " *****r seem o wmer >=>.: om 1> O o Yaehreens v Or** O Asememessas .i ..J?"L iSGRP44RC[

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' EPRVr4PC 1 1 Steam Generator Reliability Project Introduction -w l R. P. Mcdonald Y

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I EPRLNPDi EPRI Managed 1 Steam Generator Programs 5000l SGOGN EPM - EPN EPRI i Period 77'82

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Steam Generator Reliability Project Annual Meeting with NRC NRCOffices One White Fiint North 11555 Rockville Pike, Rockville, Maryland March 10,1992 AGENDA 1:00 PM Introductory Remarks R.P. Mcdonald /NRC ' w 1:10 Summary of SGRP Scope and Plans ' R.L Jones 1:20 Ch,erview of NRC Plans for SG-Related Work NRC 1:30 Secondary Chemistry Modeling and J.P.N. Paine Improvement 2:00 Status of Generic Damage-Form Specific D.A. Steininger Alternate Tube Repair Criteria Documents 2:30 Comments on Generic Alternate Tube Repair NRC Criteria 2:50-Status of Industry SG Inspection Performance M.M.' Behravesh Demonstration Program 3:20 Comments on Industry Program and Status of - NRC SG Mockup and ISI Reg. Guide 3:S0 Closing Remarks - R.P. Mcdonald /NRC 4:00 ' Adjourn- .}}