ML20198F534
| ML20198F534 | |
| Person / Time | |
|---|---|
| Issue date: | 12/22/1998 |
| From: | Tim Reed NRC (Affiliation Not Assigned) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20198F542 | List: |
| References | |
| NUDOCS 9812280148 | |
| Download: ML20198F534 (48) | |
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k UNITED STATES g
,j NUCLEAR REGULATORY COMMISSION
' t' WASHINGTON, D.C. 20A%4001
- ,o December 22,1998 MEMORANDUM TO: Document Control Desk FROM:
Timothy A. Reed Materials and Chemical En ering Branch Division of Engineering Office of Nuclear Reactor Regulation
SUBJECT:
DOCUMENTS RELATING TO DIFFERING PROFESSIONAL OPINION (DPO): DPO CONSIDERATION DOCUMENT AND SEPTEMBER 25,1998 HOPENFELD MEMORANDUM Attached is a copy of (1) the DPO consideration document and (2) a memorandum dated September 25,1998 to the Commission from Joram Hopenfeld entitled "J.Hopenfeld's Differing Professional Opinion Concerning Voltage-Based Repair Criteria for Steam Generator Tubes:
Release of DPO Document for Public Comment." The DPO consideration document provides an integrated report of how the DPO issues (from a DPO associated with steam generator (SG) tube integrity) were considered during the development of the regulatory approach for addressing SG tube integrity. The September 25,1998 Hopenfeld memorandum to the Commission documents the DPO author's continuing concerns regarding the staffs regulatory approach for SG tube integrity. The DPO consideration was devcloped prior to the Hopenfeld memorandum and was not subsequently updated to incorporate or address the information in the September 25 memorandum. In SECY-98-248, the staff recommended to the Commission that the two attached documents should be released for public comment along with draft regulatory guide DG-1074 " Steam Generator Tube Integrity." In staff requirements memorandum dated December 21,1998, the Commission did not object to the staff recommendation. Accordingly, both documents are being made publicly available as part of the staffs effort to solicit public comments on draft regulatory guide DG-1074. The two attached documents povide a basis for the discussion of technicalissues that relate to the DG-1074 guidance. By making the September 25,1998 Hopenfeld memorandum publicly available, the NRC is not endorsing the memorandum, nor its contents.
Please place the attached documents in the public document room.
Vx Attachments: 1. DPO Consideration Document
\\
- 2. Memorandum to the Commission from Joram Hopenfeld, "J. Hopenfeld's Differing Professional Opinion Concerning Voltage-based Repair Criteria for Steam Generator Tubes: Release of DPO Consideration Document for Public Comment," dated September 25,1998
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5000 nor is there data for the SGTR spiking conditions to suggest that even with a factor of 10 increase in the spiking factor, the spiking factor will be greater than 5000. Therefore, the staff concluded that adequate margins exist with present spiking factor of 500 to maintain doses below Part 100 guidelines in the event of a MSLB even at lower reactor coolant activity levels.
Table 2 Spiking Factom Which Would Result in Releases Equal to Those of the MSLB SRP Case' Spiking Factors **
RCS Activity Level fuCi/o Primary to Secondary Leak 0-2 Releases 0-8 hour Releases Dose Eauivalent l)
Rate (com total) 0.5 100 8.18 35 12.5 26.6 10 86.3 98.6 1
0.1 100 37 57 35 133 150 10 502 511 0.05 100 91 118 35 283
~~
305 10 1020 1030 0.01 100 526 606 35 1490 1540 10 5180 5150 0.005 100 1070 1220 35 2999 3090 10 10400 10300
- SRP Case based upon reactor coolant activity level of 1 pCi/g of dose equivalenti, a spiking factor of 500 and 1 gpm total primary to secondary leak rate.
- Spiking factors would need to be multiplied by a factor of 10 in order to equate to a dose of 300 rem thyroid.
- Maximum allowable leak rate would be limited to approximately 60 gpm.
22 i
l Response to Issue 5. " Severe Accident losues":
The assessment of the CBRF due to thermally induced failures of flawed tubes, referenced in the response to issue 3 above, is based on a probabilistic treatment of the creep behavior of the surge line, hot legs, and steam generator tubes with various flaw sizes. These analyses yielded the probabilities of whether axially cracked tubes would fail before other portions of the RCPB under the temperature and pressure conditions predicted in different SBO core damage sequences. Those results were then combined with estimated flaw size frequency distributions to calculate the probability that one or more tubes would fail before another part of the RCPB for each accident sequence, in these analyses, the staff recognized that through-wall flaws which are too short to burst at normal temperatures may burst later when temperatures increase. The staff also assumed that the effects of impingement of a hot steam Jet from one burst tube may lead to rapid failure of an adjacent tube due to erosion. So, if a tube was calculated to be the first RCPB failure, then no credit was taken for later depressurization of the RCS by a subsequent failure of some other component.
Imoact of Small Cracks At tube temperatures before surge line failure, critical flaw lengths for axial cracks (i.e., those flaws that would burst) were usually in the range of 0.4 to 0.6 inch. In the staff analysis, flaws as short as 0.25 inch that propagated through-wall during the challenge were assumed to leak so severely at high temperatures that they would be equivalent to ruptures. Staff estimates of the opening of a 0.25 inch long crack at elevated temperatures resulting from erosion and creep or deformation could lead to adjacent tube effects in a short time (minutes to an hour). However, the degree of erosion of small cracks due to the high velocity passage of superheated steam has not been directly assessed. Further, the nature of other contributing factors, such as duration of peak temperature conditions, and possible effects of differential thermal expansion, make an explicit determination of crack opening rate difficult.
The staff acknowledges that under certain conditions of tube degradation and sustained high RCS temperature and pressure conditions, crack opening could be a concern. As indicated above, the model used in the staff's risk assessment assumed that through-wall cracks of.25 inches in length would lead to tube failure. The.25 inch length was based on consideration of corrosion cracking mechanisms and typical aspect ratios observed for cracking mechanisms. In order to assess the potential for the existence of shorter through-wall cracks, the staff reviewed available data from destructive examination of tubes removed from service due to a variety of typical degradation mechanisms. The evaluation of depth / length ratios of cracks, provided in the appendix, concludes that one type of tube degradation (e.g., primary water stress corrosion cracking in the hard roll transition at the top of the tube sheet) presented a concern that very small through-wall cracks could be present at a limited number of plants. The staff does not have data to conclusively demonstrate the behavior of very small through-wall cracks under core damage conditions. However, based on evaluation of erosion data at similar conditions for similar types of materials, the staff has assumed that such defects could induce tube failure.
Therefore, consideration of alternate repair criteria will need to include assessment of the severe accident risk contribution from all degradation in a plant, including possible crack opening and l
leakage effects.
i i
- ~ _ - - _
LRahage Effects on SG Inlet Plenum Mixina At the March 5,1997 meeting of the ACRS Materials and Metallurgy / Severe Accidents Subcommittee, the DPO author made a presentation that disputed conclusions reached in NUREG-1570, " Risk Assessment of Severe Accident induced Steam Generator Tube Rupture."
A basis of the discussion was that the impact of tube leakage on steam generator inlet plenum mixing during the course of a high pressure core damage event was not considered.
The criteria used in the staff analysis reasonably accounts for the risk that may be contributed by tubes leaking during severe accidents in NUREG-1570, tube rupture was considered the failure criterion contributing to risk, since a large release of fission products could then be anticipated.
Tube leakage was n t considered to have a significant impact since leakage of sufficient magnitude to disrupt natural circulation flows would not be expected unless a tube rupture had occurred. The basis for this can be seen by examining the thermal-hydraulic analyses documented in NUREG/CR 5214, " Analyses of Natural Circulation During a Surry Station Blackout Using SCDAP/RELAPS." The results indicate that hot leg flows of approximately 3
3 kg/sec can be expected. Using a tube bundle to hot leg flow ratio of 2 (approximation from test results), a tube bundle flow (rn) of about 6 kg/sec can be anticipated. To disrupt the tube bundle natural circulation flow pattern, a leaking tube would need to divert a sizeable fraction of this flow.
To estimate the magnitude of leak flow that could be anticipated during the accident, an anticipated crack size was first determined. Under steam generator degradation specific management (SGDSM), accident-induced leak rates of about 100 gpm are conceivable during design basis accidents (DBAs) that result in elevated tube differential pressure. Equations from Appendix ! of NUREG/CR-4483, " Reactor Pressure Vessel Failure Probability Following Through-wall Cracks due to Pressurized Thermal Shock Events" (see equations for sub-cooled conditions, below) were used to find the crack size that would result in a leak rate of 100 gpm under elevated differential pressure conditions associated with a DBA. This leakage modelis the same as that used n NUREG-1477," Voltage-Based Interim Plugging Criteria for Steam Generator Tubes." Using the model, a crack size of approximately 0.12 id was found to correspond to 100 gpm leakage.
i During the core degradation phase of the SBO progression, superheated steam will be entering j
the steam generator from the hot leg. In the analyses forNUREG-1570, tube differential pressures resulting from secondary system depressurization were considered. Thus, the flow through a crack in a tube can be computed assuming choked orifice flow (the choked flow
^
assumption holds since secondary side pressure is below the critical pressure of approximately i
1300 psi).
Using the equation for superheated conditions (see below), leak rates for superheated steam under the conditions calculated by SCDAP/RELAP5 can be estimated for various crack sizes.
The table below gives estimated leak rates for different crack sizes corresponding to: (1) a 0.25 in long,0.01 in wide opening, (2) a crack corresponding to a 100 gpm leak under design basis 4
accident (DBA) coriditions, and (3) a crack equalin size to the open end of a tube (7/8 in. diam.).
l 1
i 24 i
Leak Crack Size (in )
2 Rate 0.0025 0.12 0.47 kg/sec 0.027 1.3 5.13 lbm/sec 0.06 -
2.8 11.31
% of mi 0.5 22 86 Equations for sub-cooled conditions:
m = c A /(2 g, p(P, - P,,,))
P, = P,, - P, L (rn/A)*
P = 0.5 f r
D,,
g,p D
f=(2in(
) + 1.74)~8 Equation for super-heated conditions:
f
'("*')(*I g
2 2-k m = c A P"* \\RT k+1, where A
= crack opening crea c
= coefficient of discharge (0.6)
Dn a hydraulic diameter f
a friction factor g,
a conversion (32.2 lbm-ft/lbf-sed) k
= specific heat ratio (q/c,) = 1.26 L
= crack length m
o mass flow rate P.
= saturation pressure at RCS conditions l
P.,
a RCS pressure P,
a frictional pressure drop R
= gas constant (85.8 ft-Ibf/lbm-R)
T
= steam temperature in Rankine 5
= roughness height in crack (2x10r'in. from NUREG 4483) l
25 p
= liquid density @ P.
The first crack size shown in the table,0.0025 irf, would yield about 2 gpm leakage under DBA conditions. In the analysis, this is the smallest size crack that the staff assumed to fail by rupture under the predicted conditions during a severe accident. The steam leakage through this crack is a very small fraction of the total tube bundle flow,0.5% of rg and is not expected to have an impact on inlet plenum mixing. As mentioned previously, small through wall cracks are expected to be present in only a few cases involving one type of degradation.
The 0.12 in' crack (1 inch long, 0.12 inch wide), which corresponds to 100 gpm leakage under DBA conditions, would leak under severe accident conditions at a significant rate compared to the total tube bundle flow. This size flaw is considered equivalent to a tube burst if it is located in a single tube. It is 4 times as long and more than 10 times the area of the crack the staff assumed would burst under severe accident conditions. The largest tube opening shown in the table would divert a very significant portion of the tube bundle flow. These two flaws, which would both be assumed to rupture in the staff's analysis, could be expected to change the mixing conditions. Only the largest leak rate of 11.31 lbm/sec associated with this size opening is in the range of leakage cited in the DPO presentation (10 - 250 lbm/sec). Under the criteria used by the staff in the NUREG-1570 analysis, a tube would be considered failed upon rupture, which could result in crack openings on the order of a tube diameter. Once failed, further analysis of the thermal-hydraulic effects of this magnitude of leakage is not required to assess the risk impact, Operating experience has shown that the occurrence of tube leakage is often associated with a series of flaws; that there is not only a single flawed tube (e.g., ANO Unit 2,1992; McGuire, Unit 1,1992; Maine Yankee,1990 and 1994). The accident leakage that would be permitted under SGDSM is an aggregate value from all flaws that would leak. Thus, it would be reasonable to expect that in a severe accident situation, the leakage from flaws would be distributed throughout various locations in the tube bundle, and not necessarily confined to one tube or location. Therefore, the leakage effects on inlet plenum flow patterns, and mixing which is expected to occur, should be impacted to an even smaller degree than is indicated by the comparison of leak flow to bundle flow shown in the table above.
Uncertain factors in this discussion are the potential for cracks to open while they leak under severe accident conditions, and for leaking tubes to cause cascading failures. Preliminary staff estimates conclude that through-wall cracks could open at a significant rate if the high temperature and differential pressure conditions are sustained. However, the duration of very high tube temperatures is on the order of minutes. Also, the rate of cascading failure predicted in the DPO seems overestimated based on preliminary staff estimates which show only a modest erosion rate for tubes adjacent to a leak.
For analysis of severe accident risk, the staff chose 0.25 inch as the crack length at which to assume tube failure during severe accident thermal challenge. Based on a subsequent staff assessment of data from destructive examination of degraded tubes removed from operating plants as well as its assessment of leakage effects from cracks, this criterion is reasonable. The staff believes that there is a potential for through-wall cracks shorter than the staff's threshold of
26 0.25 inch to exist in a small number of plants. However, the risk significance of such defects requires further understanding of the frequency and nature of the plant specific sequences that could challenge tube integrity. The staff plans to investigate sequences in conjunction with the 1
1 IPE follow up program.
Fission Product Deposition and Heatina By letter dated May 20,1998, the DPO author forwarded to the ACRS slides from a presentation given by a Japan Atomic Energy Research Institute (JAERI) representative at the May 1998 Cooperative Severe Accident Research Program (CSARP) meeting. The JAERI analysis addressed the additional heating of steam generator tubes, during a station blackout sequence with secondary side depressurization, caused by the deposition of fission products on the inner tube surface.
The JAERI analysis, while it predicted the steam generator tubes would survive the transient, with a small margin, also predicted substantial fission product heating of the steam generator tubes due to the deposition of fission products. This is in contradiction to the NRC conclusion, based on analysis, that the effect of deposited fission products in tube heating is trivial. The staff, therefore, evaluated the JAERI analysis to understand why it produced substantially different results concerning fission product deposition, from corresponding NRC calculations.
i A review of the JAERI analysis revealed several differences from the NRC evaluation. First, it was learned that the JAERI analysis assumed the temperature of the steam entering the tube bundle was equal to the temperature of the unmixed vapor in the steam generator inlet plenum (from the hot leg). Assuming the temperature of the steam entering the tube bundle is equal to that of unmixed vapor from the hot leg is a severe conservatism; in contrast, the NRC analysis includes mixing of steam in the inlet plenum based on experimental data. (Use of unmixed vapor temperatures would produce excessively high tube temperatures irrespective of fission product deposition.) Use of the temperature of the unmixed vapor in the steam generator inlet plenum as the temperature of the vapor in the tube in the JAERI analysis apparently resulted in a temperature difference between the vapor and the steam generator tube wall of up to 250K inside the first section of the tube. It is this high temperature difference that is responsible for a large thermophoretic fission product deposition in the JAERI calculation. It is unrealistic to assume such a large temperature difference in this region without water on the secondary side of the steam generator given the heat transfer across the thin tube wall. The NRC's analysis showed a temperature difference of about 15K in this region, which results in minimal deposition by thermophoresis.
The NRC's VICTORIA code analysis of this sequence showed that the dominant mechanism for deposition was settling inside of the steam generator u-tubes onto the upward facing surfaces in the bends of the u-tubes. This is reasonable given the recirculatory flow patterns in the RCS and the lack of other driving forces for deposition other than settling. The JAERI analysis did not appear to model settling on upward facing surfaces inside of the steam generator u-tubes.
Finally, in another area, the JAERI analysis did not explicitly consider the primary source of heat (i.e., cladding oxidation)in their calculation of heat up of the steam generator tube wall. The relative importance of deposited fission product heating was determined by JAERI by comparison against the heat carried to the tubes by superheated steam heated only by decay
27 heat in the core. Because of their thermal hydraulic assumptions and analytical treatment, the staff concludes that the finding of the JAERI analysis, relative to the significance of fission product heating, is not relevant to our evaluation of this matter. As a result, the staff's previous conclusions regarding the severe accident risk implications of degraded steam generator tubes remain unchanged.
Conclusion The staff recognizes that alternate repair criteria could be proposed that might lead to the presence of short through-wall cracks. As mentioned previously in this document, the staff is encouraging licensees who propose alternate repair criteria to follow risk-informed approaches which may mean that licensees will need to address the issues discussed in this section as appropriate dependent on the nature of the proposal. Currently, with the exception of accident leakage rates associated with GL 95-05 repair criteria (which represent a different risk concern i
due to the confined nature of the cracks), accident leak rates for all other forms of degradation are limited to current licensing basis values (typically 1 gpm or less) which is well below the leakage rates contemplated above (i.e., on the order of 100 gpm). The staff will not approve any proposed repair criteria as an alternative to compliance with existing deterministic requirements if it unacceptably increases risk.
l a
Appendix STEAM GENERATOR TUBE FLAW GEOMETRY 2 i
The assessment of steam generator tube degradation identified during eddy current examinations has historically focussed on tubes with indications that could potentially challenge the structural and leakage integrity of the tube under postulated secondary side depressurization events. However, only a limited amount ofinformation is available on the number of tubes that contain degradation smaller in size relative to the larger, more significant flaws that are readily detected during eddy current examinations. Under severe accident conditions, elevated pressures and temperatures may induce steam generator tube failure by a time-dependent failure mechanism of short, through-wall flaws that exhibit no or marginal signal response during nondestructive examinations (NDE). Because of the limitations in inspection technology, it has been postulated that the existence of these flaws could lead to steam generator tube failure under severe accident conditions. In order to have the potential for failure, these undetected flaws must have a through-wall or near through-wall morphology to introduce a primary-to-secondary leak path that will eventually lead to ablation of the flaw surfaces and expand the dimensions of the defect. This could then lead to increased primary-to-secondary leakage of hot gases through the defective tube and subsequent failure of surrounding tubes by jet impingement.
The risk from the failure of longer, through-wall steam generator tube defects (length >0.25 inches) under severe accident conditions has been examined in previous work and is conservatively included in the risk assessment model. In order to address the issue of short crack failures under severe accident conditions, a scoping study was completed that included a i
review of metallurgical data obtained from steam generator tube destructive examinations and correlations developed relating flaw depth and length. The focus of the study was to attempt to determine if there is a significant population of tube defects in steam generator tubes characterized by short lengths (<0.25 inches) with through-wall or near through-wall depths that could develop primary-to-secondary leaks in the freespan tube areas under severe accident conditions.
Backaround Utilities occasionally submit to the NRC distributions of steam generator tube flaws as a function of some NDE parameter. These can be used to explain the overall significance of the degradation identified during an inspection in terms of the total number of degraded tubes identified and the nature of the indications in terms of NDE measurements (e.g., voltage, estimated length, depth). Figure 1 represents the typical form for an end-of-cycle voltage distribution of outside diameter stress corrosion cracking (ODSCC) indications at tube support plate (TSP) intersections. As seen in the figure, the number in some cases, the figures discussed in this report contained proprietary information and have therefore been removed from this report to enable this document to be made publicly available.
4 4
+w
2 of indications is low within each voltage bin for low voltages but increases rapidly for slightly larger voltages. At higher voltages the number of indications gradually tails off toward zero.
Figure 1: Voltage Distribution u.
4g
...__ 1l l
3 l
j.. ;
i.. !
a 5
l i
I b
ei.: n......, s.., u o u i. is i. o i.
vena.
These characteristics are consistent with voltage distributions reported to the NRC for indications detected at drilled tube support plate intersections. In addition, this observation is consistent with other reported distributions such as that depicted in Figure 2. This figure shows the number of flaws as a function of the estimated average crack depth. Again, the number of indications increases, peaks at some value, and then decreases toward the tail of the distribution.
Figure 2: Crack Distribution
)
25 l
20 l
t 3
315 q I
l f10 f
i 5M O-i i
0 10 20 30 40 50 60 70 80 90 100 Averag. Crack D.pth (%TW)
Although these distributions could lead one to conclude that flaws with low voltages or limited through-wall depth occur less frequently than those with slightly higher voltages or depths, this would not be a valid conclusion for several reasons. First, the above figures are representative data derived from eddy current inspections. Although sensitive to small flaws, eddy current technology has limitations. Steam generator tube defects that exhibit lower voltages will, by i
i l
3 definition, have only a smallinfluence on the eddy current signal. Since degradation generally occurs in crevice locations or in areas where the tube geometry is changing, other signals may overwhelm the defect response and mask its detection. Data analysts will also influence the i
threshold of detection for tube degradation. Although the inspection system may display a distortion due to the presence of a flaw, an analyst may not be as sensitive to these small signal changes compared with the signals from more obvious indications. A true flaw size distribution cannot be obtained from eddy current inspection data for smaller flaws (length, depth, voltage) because of the limitations of the analysts and the sensitivity of the technology used for steam generator tube examinations.
In order to better understand the distribution of cracks below the threshold of detection for eddy current inspection techniques, examining data in figures similar to that in Figures 1 and 2 is of little value due to the limitations of NDE. An estimate of the number of flaws below the NDE threshold could be obtained by extrapolating the distribution to lower voltages and depths.
However, extrapolating the data may introduce additional errors and incorrectly model the actual population of flaws over the range of the extrapolation. Although it may be reasonable to conclude that there is a larger number of much smaller tube defects than shown in flaw distributions obtained from eddy current inspection data, based on NDE data, little can be said for the morphology of these flaws.
To determine if a significant population of short, through-wall flaws exist within operating steam generators, metallographic examination data from tubes removed from inservice steam generators (i.e., pulled tubes) were reviewed. The pulled tube data assessment included a review of data for several different modes of steam generator tube degradation: (1) axial primary-water stress corrosion cracking (PWSCC) at dented tube support plate (TSP) intersections, (2) outside-diameter stress corrosion cracking (ODSCC) at TSP intersections and tubesheet expansion-transitions, (3) PWSCC at tubesheet expansion-transitions, and (4) pit-like intergranular attack (IGA) indications. Considering that the focus of most tube steam generator pulls is obtaining information on larger, more potentially significant defects, the amount of metallurgical data on short flaws is limited and sometimes biased. Nevertheless, some data have been reported to the NRC that include details on the shorter, less significant steam generator tube degradation.
One set of data examined in this study included lengths for PWSCC flaws at tubesheet roll transitions. The focus of the report from which the data were obtained was to quantify the leakage through axial PWSCC flaws [1]. Therefore, the data presented in the report only included flaw lengths associated with through-wall cracks that leaked during testing. It is likely that a number of short, part through-wall flaws existed in the tube sections examined in the PWSCC leakage study. However, because these cracks did not leak, these data would not normally be reported in such a study.
l
4 Discussion The threshold of detection for eddy current inspection methods is particularly sensitive to the volume of material affected [2]. That is, if a sample of flaws has similar electrical and geometric properties (i.e., orientation), then the eddy current response willlargely be a function of the relative size of each indication. Results from actualinservice inspections of steam generator tubes indicate that eddy current inspection techniques are very capable of detecting part through-wall defects in steam generator tubes with lengths on the order of 0.25 inches.
Although eddy current techniques can detect shorter flaws, it is assumed for this investigation that all through-wall flaws exceeding 0.25 inches in length would have a high degree of detedability.
Surface flaws can take on an infinite number of geometries with varying lengths and depths.
One approach used in fracture mechanics to simplify the description of flaw geometry is to assume a semi-elliptical flaw shape [3). With this assumption, a surface breaking flaw can be described by two parameters, the length along the surface and the through-wall depth. Given that the longest flaw length considered here is 0.25 inches and assuming a maximum tube wall thickness of 0.050 inches, the bounding flaw aspect ratio (length-to-depth ratio)is five. Since the focus of this assessment is on short, through-wall cracks, such flaws would have aspect ratios less than this bound.
Many data were available from destructive examinations of tubes containing ODSCC indications i
at TSP intersections. Although this mode of degradation is confined within the TSP and leakage from these flaws could not impinge on neighboring tubes, this mode of degradation is similar that found in other areas of the tube. ODSCC at TSPs and cracking that develops above the tubesheet in the sludge pile both develop in low stress, tight crevice tube areas. The morphology of the flaws that develop in these areas is expected to be similar. Therefore, it is assumed that conclusions can be drawn for sludge pile cracking based on an examination of the data that exists for cracking within TSPs.
Evaluation of Pulled Tube Data Data shown in Figure 3 represent the results obtained from destructive examinations of tubes removed from Diablo Canyon, Unit 1, in 1995 (4). The tubes contained OD and ID axial and OD circumferential defects located at dented and non-dented drilled tube support plate intersections.
l The tube pulls targeted the larger, more significant indications of degradation. However, not til of the flaws presented in Figure 3 were detected by eddy current inspection techniques prior to pulling. With one exception, all of the flaws had lengths on the order of 0.2 inches and maximum crack depths significantly less than through-wall. The only flaw that penetrated through the entire tube wall was the long axial indication that was easily identified with bobbin coil and rotating pancake coil probes.
4
5 i
Figure 3. Diablo Canyon Tube Pull Summary
[contains proprietary information]
j The licensee characterized the data in Figure 3 as the macrocrack analysis results. Additional l
flaws were identified in the destructive examinations; however, the level of detail was less than that provided for the ab 'e data. Some useful information was provided on the overall number and depth of less significant flaws idJntified from the destructive examination. For instance, at one TSP intersection an estimated 33 OD cracks were observed at the mid-TSP level.
Additional cracks were seen at other locations as well. The maximum and average crack depths determined from transverse sectioning of the tube specimens were 24 and 11-percent through-wall, respectively. Therefore, although a sizable number of flaws were seen in the metallographic analyses, all of these were shallow. A similar observation regarding flaw depths was made at one other TSP intersection. Four other TSP intersections were either defect free or found to have only a small number of shallow OD defects.
The Steam Generator Degradation Specific Management (SGDSM) database contains a substantial amount of pulled tube data on ODSCC at TSP intersections. The data in this database are classified by tube diameter, either 3/4-inch or 7/8-inch. Figure 4 contains the information available in the database on flaws that leaked during leak rate testing [5]. The ID/OD flaw ratio represents the totallength of crack on the ID surface divided by the OD length.
Since the flaws originated from the OD, these values are generally less than unity. As seen from the figure, many of the data were obtained from laboratory grown cracks from model boiler tests.
In addition, these flaws appear to have more extensive through-wall degradation than that observed in the pulled tube samples. As seen from the data no through-wall cracks were identified with OD lengths less than approximately 0.25-inches. Therefore, the 7/8-inch pulled tube data in the SGDSM database does not support the assertion that short through-wall flaws exist in large numbers. In fact, based on the available data, no short through-wall flaws were present. One note, the data in this figure represents only those tubes that leaked during testing.
Therefore, the samples used in the SGDSM database may contain through-wall cracks that did not leak at elevated pressures.
F.gure 4 Geometry of ODSCC TSP Flaws (7/8" Diameter Tubes)
(contains proprietary information]
Figure 5 represents a similar set of data for 3/4-inch diameter tubes [5]. The Si4-inch database is more extensive than that for 7/8-inch tubing. However, the conclusions that can be drawn from the data are similar. Once again through-wall cracks were only observed for those flaws that were longer in length, and with one exception for the pulled tube samples, leakage was not observed except for axial defects exceeding 0.45 inches in length.
Figure 5 Geometry of ODSCC TSP Flaws (3/4" Diameter Tubes)
[contains proprietary information]
/
l l
l 6
Pulled tube destructive examinations of ODSCC samples removed from the Trojan steam generators included a detailed assessment of microcracks in a number of the samples. Figure 6 shows some of the length-depth flaw data reported for these tubes [6). Although Trojan data are included in the SGDSM database, data shown in Figure 6 are not included in Figure 4. One observation from Figure 6 is that the metallographic examinations identified numerous short, part-through-wall cracks. Despite that large number of flaws present, no cracks penetrated through to the ID surface.
Figure 6 Trojan Microcrack Data
[contains proprietary information]
The examination of Trojan microcrack data is a conservative approach to verify the likelihood of occurrence of short, through-wall flaws. Specifically, since these microcracks are generally part of a larger network of cracks, these flaw networks are much more detectable by NDE techniques. If the flaws shown in Figure 6 are representative of the population of TSP cracks and other flaws that develop in low stress crevice locations, then it appears unlikely that there is a significant number of short (i.e., < 0.25 inch) flaws that have the potential for leakage.
Figure 7 includes the same data as Figure 6 with the through-wall depth shown in absolute dimensions rather than as a percentage of the wall thickness. What becomes more apparent in examining Figure 7 is that most of the flaws had an aspect ratio greater than one. This suggests that ODSCC at TSPs is more likely to grow in length than in depth. Therefore, complete through-wall flaw penetration is not expected except for longer (i.e., >0.25 inch) defects.
Figure 7 Trojan Microcrack Data
[contains proprietary information]
Most of the data presented to this point have come from degradation located at TSP intersections. Although these flaws may not be a significant consideration during severe accident conditions due to the restraint provided by the surrounding TSP, it is assumed that the torphologies observed for these defects is representative of degradation in crevice locations with no elevated stresses. Such degradation could exist in the sludge pile region above the tubesheet secondary face remote from the expansion-transition or within the crevice of a partially expanded tubesheet. Based on data shown in Figures 4,5, and 6, it appears unlikely that short, through-wall flaws exist in TSP crevices. Given the similarity between the flaws develop in TSP crevices and within the sludge pile region, there is a low likelihood that short, through-wall cracks exist in the sludge pile area of steam generator tubes.
IGA degradation in the freespan areas has recently become an issue of increased attention.
Although a relatively small number of tubes have been identified with these indications, pulled tube data suggest that many of these indications may go undetected with current NDE technology. Tubes were removed from the Crystal River, Unit 3 (CR-3), steam generators to i
assess the burst and leakage integrity of IGA degradation indications located above the lower tubesheet. Some of these data are shown in Figure 8 [7]. Tube defects were identified in the destructive examinations by visualinspection of the OD tube surface. Therefore, all potentially significant degraded areas should have been located in the metallographic examinations. Most l
l
7 of these flaws were not identified by eddy current inspections completed prior to or after the.ube pulls.
Figure 8 Patch IGA Indication Geometry Crystal River 3 Pulled Tube Data
[contains proprietary information)
One observation from the data in Figure 8 is that IGA defect length and depth appear related.
By extrapolating a regression to these data, a through-wall defect would not be expected until the length of the defect is approximately 0.16 inches. Although this value falls below the 0.25 inch threshold established for this investigation, what does remain clear is that the depth of IGA is related to length. Hence, one would not expect to identify a sizable population of similar IGA degradation exhibiting short lengths and extensive through-wall penetration. Since IGA is essentially three-dimensional degradation (i.e., volumetric versus crack-like), the detectability of these indications will increase significantly with increasing length, depth, and width. Some of the indications detected by eddy current during the CR-3 inspections were found to have axial lengths as low as 0.02 inches. Despite the short lengths of some of the degraded areas, the bobbin coil probe was able to detect some of these indications. In addition, the large volume associated with deeper defects facilitates their detection during eddy current examinations.
Based on previous experiences with IGA degradation, this mode of degradation appears more resistant to devd.opment of a leak path than other forms of tube wall damage (e.g., ODSCC).
Although the licensee for CR-3 has identified a several hundred of these IGA patches in its steam generators, primary-to-secondary leakage has not been attributed to these indications. In addition, a number of tubes that recently underwent in situ pressure testing did not leak. Pulled tube specimens obtained from CR-3 were burst tested to assess their structural and leakage integrity. The tubes were able to retain internal pressures on the order of 10000 psi without any measurable leakage prior to burst. Therefore, the IGA degradation in the tubes removed from CR-3 would not challenge the structuralintegrity of these tubes during design basis depressurization events.
Figure 9 presents data obtained from a report by the Electric Power Research Institute (EPRI) on axially-oriented PWSCC degradation located in the roll transition zone (RTZ)[1]. The data only include through-wall cracks from pulled tube samples to assess leak rates. The ID flaw lengths ranged from 0.06" up to 0.3". However, most of the data are located in the interval between 0.1 to 0.2 inches. Given that this is the area of interest for this assessment, the PWSCC data from the EPRI report is a useful resource. The vertical axis in the figure represents the ratio of the OD to ID crack length as measured from destructive examinations.
Figure 9 PWSCC Flaws at Roll Transitions (contains proprietary information]
One notable difference between the data presented in Figure 9 and that shown in the previous figures is that it is widely scattered with no apparent relationship between the crack length measured on the OD of the tube to that on the tube ID. In the previous figures, the flaw depths generally increased as the crack length on the surface increased. In general, this behavior is typical for surface initiated flaws. However, the data in Figure 9 do not follow this convention. If
8 flaw depth is dependent on the length, or vice-versa, one would expect to find an increasing flaw length ratio with increasing ID length, but this is not the case with the PWSCC data reported by EPRI. Since the data were obtained from only nine pulled tube samples, it is reasonable to conclude that there is the potential for short, through-wall cra' ks to exist in the RTZ. Also, given c
that leakage was observed from each of the pulled tube samp'es, one cannot rule out the possibility that the leakage originated from some of the shorter cracks.
Crackina in Exoansion-Transition Zones Some steam generator designs secure the tubes in the tubesheet with a partial depth roll expansion. This is a hard roll that begins at the primary face of the tubesheet and extends severalinches along the tube before terminating within the tubesheet bore. B&W once-through steam generators (OTSGs) and a smo" number of Westinghouse models have partial depth expanded tubes. Because the hard rC) terminates within the tubesheet, primary jet impingement of hot gases on adjacent tubing during severe accident conditions is precluded.
Incidently, axial cracking at RTZs is a significant mode of tube degradation for Westinghouse steam generators with partial depth expansions. Inspections of roll expansions in B&W OTSGs and pul'ed tube destructive examination results have identified only a small number of cracks in the RTZ to date.
The geometry of RTZ cracking appears different from the ODSCC at TSPs and IGA degradation discussed previously. Further discussion on the differences between the cracking in these areas is warranted to address the potential existence of short, through-wali flaws. Steam generator tube RTZs are highly stressed areas of the tube. Analytical calculations and
)
experimental measurements have concluded that the stresses in these areas are greater than in other stressed areas of the tube such as dents and tight radius U-bends. The residual stress field in the transition zone introduced from the rolling process produces stress levels much greater than that from primary-to-secondary differential pressures.
In order to eliminate the crevice areas of partial depth roll expanded tubes, steam generators later utilized a full depth hard roll expansion. The primary difference between these two types of expansion joints is that the RTZ exists at the top of the tubesheet in full roll designs. Despite the J
elimination of the crevice region, a dominant mode of tube degradation in RTZs continues to be axial cracking. EPRI completed one investigation into RTZ cracking for tubes removed from the Ringhals 2 steam generators [1]. The Ringhals 2 steam generators were the Westinghouse Model 51 design with a partial depth hard roll expansion. Destructive examinations of pulled tubes concluded that most of the axial PWSCC was located within the transition zone extending up toward the upper transition region. Since the RTZ for full depth expanded plants exists at the top of the tubesheet, it is reasonable to conclude that RTZ cracking is similar to cracking in the freespan area of the tube with regard to primary to-secondary leakage. Therefore, the existence of many short, through-wall axial cracks that could be a factor under severe accident conditions is possible for steam generators with full depth hard roll expansions.
There are currently nine plants in operation (33 steam generators) with full depth hard roll expansions. However, the utilities for five of these plants (19 steam generators) will replace the steam generators within the next four years with designs that incorporate more corrosion resistant materials and tubes with expansion-transitions that are less susceptible to stress
i 9
l corrosion cracking. Although roll transitions have the most severe state of stress with respect to the developn,ent SCC for all methods of expanding tubes into tubesheets, a closer investigation of the mechanisms of cracking within explosive and hydraulic expansions is warranted.
Axial cracking in explosively-expanded steam generator tubesheet joints has been identified by a number of utilities. This applies to steam generators designed by Westinghouse and J
Combustion Engineering. Limited data are available on the location of axial cracking relative to the top of the tubesheet for these steam generators. However, a recent assessment of inspection findings in the Salem, Unit 1, steam generators concluded that all but nine of 177 axial flaws (ID and OD) were located below the top of the tubesheet [8). Although this conclusion is based on NDE inspection results and not pulled tube destructive examinations, it is clear that the axial cracking in the explosively expanded tubes at Salem favors an area of the 1
expansion-transition that is located below the tubesheet secondary face. Therefore, if short, through-wall cracking similar to that found in RTZs develops in explosively expanded tubesheet joints, then it is most likely to occur in an area of the tube where primary-to-secor'dary leakage would not impinge on neighboring tubes.
The current method of forming the tubesheet expansion joint is to hydraulically expand the tubing into an interference fit within the tubesheet bore. The majority of steam generators that entered into service since the early 1980's were fabricated with hydraulic tubesheet expansion joints. Cracking experience in hydraulically expanded tubing has been limited to date. There l
are no operating plants with these expansion joints that have identified significant levels (i.e.,
large numbers) of flaws in the transition zone. This is believed to be a reflection of the lower much stresses introduced from the hydraulic expansion process [9]. Nevertheless, given enough time, these tubesheet expansion joints could develop cracking similar to degradation presently observed in steam generators with rolled or explosively-expanded tubesheets. By design, however, the expansion-transition in hydraulically expanded tube terminates below the top of the tubesheet. Therefore, primary-to-secondary leakage from short, through-wall cracks in a hydraulic expansion-transition should impinge on the bore of the tubesheet and not an adjacent tube.
10 Circumferentially-oriented SCC in tubesheet expansion-transitions affects a number of steam generators. Because this mode of degradation cannot be adequately detected using bobbin coil eddy current probes, utilities employ more sensitive inspection probes to assess the condition of tubesheet expansions. In addition, numerous tubes have been removed from steam generators in order to assess structural and leakage integrity as well as eddy current sizing capabihties.
Entergy, the licensee for Arkansas Nuclear One, Unit 2, (ANO-2) submitted a report that included a correlation relating the cracked percentage of tube wall (percent degraded area
{
(PDA)) to the nominal through-wall crack length [10]. Although the NRC staff has not evaluated the validity of the data supporting the correlation, the relationship indicates that a through-wall crack would be expected when PDA is on the order of 10-percent. Using this data along with the dimensions of the tubes at ANO-2, and assuming a single, semi-elliptical flaw geometry, a through-wall crack would be anticipated when the overall crack length exceeds approximately 0.28-inches. Circumferential flaws with such lengths should be detected with present day inspection technology.
It has been noted in many instances that circumferentially-oriented cracks in expansion-transitions are actually a series of shorter length flaws separated by smallligaments. However, due to limitations in eddy current inspection technology and for additional conservatism in assessing structuralintegrity, smallligaments are often ignored. If these ligaments are accounted for in determining the actual crack lengths, the nominal flaw aspect ratio corresponding to a through-wall crack should decrease. Destructive examination results of pulled tubes containing circumferential cracking do not show evidence of a large number of short and deep cracks, but these examinations have identified networks of circumferentially oriented flaws. The presence of multiple flaws at the same axial tube location improves the chances of detecting this mode of degradation. Therefore, it is unlikely that numerous tubes are left in service with undetected, short, through-wall, circumferential flaws.
Ctgking in Other High Stress Locations Cracking in tubesheet expansion joints has been responsible for a large number of tubes removed from service due to corrosive degradation. This is primarily due to the higher susceptibility to SCC at this location because of the elevated residual tube stresses and l
temperatures. Tube small radius U-bends and TSP denting are other locations where high stresses favor the initiation of SCC [9]. In fact, early operating experience indicates that cracking in short radius U-bends and RTZ cracking appear after a similar length of service. This would indicate that stress levels (although not necessarily stress distribution) in these two regions are comparable. Several studies have concluded that residual stresses in the U-bend l
region are similar in magnitude to those in expansion-transitions. Although dented locations also exhibit elevated residual stresses, the potential existence of short, through-wall flaws at these locations will not be addressed here because primary-to-secondary leakage from these flaws would impinge on the adjacent structure partially responsible for the development of the dent.
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11 Cracking in low radius U-bends has been primarily associated with Westinghouse designed steam generators with low temperature mill annealed (LTMA) alloy 600 tubing. This mode of cracking has caused some utilities to take preventative measures in an attempt to mitigate the problems from U-bend degradation. Preventative plugging of low row tubing and additional heat treatments are two examples of these measures. Historically, U-bend cracking has been a difficult mode of degradation to detect. Despite the problems associated with managing this mode of degradation, only limited pulled data is available. Because cracking in the U-bend cannot be extracted by removing the tube from the primary inlet plenum (i.e., tube pull),
specimens must be " harvested" from the secondary side of the steam generator. Due to the complexity of this operation, the removal of U-bend tube sections has been attempted only on a few occasions.
Portland General Electric Company, the licensee for the Trojan Nuclear Plant, completed an effort to evaluate U-bend cracking in the early 1980's. Twenty-six row one tubes and three row two tubes were removed for metallographic analysis. The destructive examination of the row 2 tubes did not reveal the presence of any cracking. Although the Trojat. study did not include the lengths and depths of U-bend flaws, it did discuss some of the findings on crack geometry identified from tubes removed from Surry 1 and Turkey Point 4. PWSCC found in tubes removed from these units were characterized as "intergranular and extended halfway through the tube wall (aspect ratio of ~4)." This observation leads one to the conclusion that, on average, through-wall cracking would not be expected until the flaws have lengths on the order of 0.2 inches. Therefore, U-bend flaw aspect ratios would seem to be greater than cracks observed at other high stress locations such as roll transitions. Thus, the potential for short, through-wall cracks in tube U-bends appears unlikely based on these findings.
One additional note on low radius tube U-bend cracking is that only a limited number of plants are currently affected by this mode of degradation. Recirculating steam generators fabricated by
)
Combustion Engineering have not experienced U-bend cracking similar to that observed in Westinghouse designed steam generators. With one exception, U-bend cracks have only been found in LTMA alloy 600 tubing. Once-through steam generators are a single pass design that do not have U tubes. Therefore, this mode of degradation cannot occur in these steam generators. These observations further narrow the population of plants that are affected by U-bend cracking.
Other Forms of Steam Generator Tube Crackina Recently, indications of freespan cracking have been identified in both Combustion Engineering and Babcock & Wilcox designed steam generators. The industry has generally described the mode of degradation as long indications of IGA. Pulled tube destructive examinations have confirmed that the depths of detected indications are generally less than 50-pucunt of the total wall thickness. In order to assess the structural and leakage integrity of freespan cracking, licensees have removed several tubes with indications and completed burst and leakage testing.
i The data indicate that short, through-wall freespan cracks are not present for this mode of l
degradation. In addition, burst test results have demonstrated that these tubes have considerable margins for tube structuralintegrity. Burst pressures for axial freespan cracks in tubes removed from the Calvert Cliffs steam generators (CE) were measured on the order of
(
l
12 undegraded tubing (11). Leakage has not been attributed to these flaws during operation, during in-situ pressure testing, or observed in testing conducted after tube removal.
Conditions Necessarv for Short. Throuah-wall Cracks This assessment identified one location susceptible to short, deep steam generator tube flaws.
Tubes with hard roll expansions have the conditions suitable for the development of this mode of cracking. The unique geometry of RTZ cracking is likely a consequence of the elevated stresses introduced into the tube by the rolling process. No other locations in a steam tube will have stress levels resulting from the steam generator fabrication processes equal to or greater than those in RT7.s. Althcugh the development of RTZ flaws is result of a combination of susceptible materials, environmental conditions, and residual tube stresses, the uniqueness of the morphology of the cracking originates primarily from the state of stress in the material. It may be reasonable to assume that extremes in the other factors for stress corrosion cracking could also lead to flaws with atypical geometries.
With respect to material susceptibility to stress corrosion cracking, steam generator tube materials are relatively homogeneous. It is unlikely that extremes in tube material chemistry would be isolated to areas less than a fraction of an inch giving rise to short, through-wall cracks.
Crevices, favorable for the development of stress corrosion cracking, can exist at several locations along a steam generator tube: (1) tubesheet crevices, (2) TSPs, (3) U bend supports, (4) in the preheater region for certain steam generator designs, and (5) above the tubesheet in the sludge pile. Deposit buildup on tube surfaces has also created freespan crevices that led to cracking, but the number of reported occurrences for this type of degradation has been limited to date. Crevice conditions could potentially exist over a very short length of tubing. However, in general, each of the tube crevice examples given above have length scales larger than the 0.25 inch threshold considered in this study. In addition, crevices primarily result from the tube lying adjacent to a support structure on the secondary side of the steam generator. Potentialleakage through any defects originating within the crevice should impact the structure responsible for creating the crevice rather than a neighboring tube. Therefore, it is likely that short, deep flaws with the potential for freespan leakage will only initiate in areas of the tubing where elevated residual stresses exist.
Conclusion in order to evaluate the plausibility of the existence of a significant population of short (i.e., <
0.25 inch), through-wall steam generator tube cracks destructive examination data were reviewed to assess flaw geometries. The purpose of this assessment was to determine if undetected cracking could lead to primary-to-secondary leakage and subsequent hot gas jet impingement on adjacent undefected tubes under the high temperature and pressure conditions postulated for severe accident conditions. Pulled tube destructive examination results were reviewed for: (1) axial ODSCC at TSP intersections, (2) circumferential cracking at expansion-transitions, (3) pit-like IGA indications in the freespan region, (4) PWSCC, and (5) axial freespan cracking in CE and B&W designed units.
Based on the data presented herein, the existence of short, through-wall cracks that could become significant during severe accident conditions does not appear likely for all modes of
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13 steam generator tube degradation with the exception of PWSCC in roll transitions. However, the presence of the tubesheet in partial depth roll expanded tubing and the scheduled replacement of a number of steam generators with rolled tubes susceptible to these cracks limit the number of tubes that could lead to primary-to-secondary leakage affecting surrounding tubes under severe accident conditions. Furthermore, current practice is to remove defects of this type from service upon detection and the detection sensitivity has increased significantly in the last few years.
l e
ng enn
14 References 1.
"PWR Steam Generator Tube Repair Limits: Technical Support Documentation For Expansion Zone PWSCC in Roll Transitions - Revision 2," EPRI NP-6864-L, August 'i993.
2.
"PWR Steam Generator Examination Guidelines: Revision 3," EPRI NP-6201, November 1992.
3.
Ernst, H. A., Rush, P. J., McCabe, D. E., " Resistance Curve Analysis of Surface Cracks,"
Fracture Mechanics: Twenty-Fourth Symposium, ASTM STP 1207, J. A. M. Boulet, D. E.
McCabe, and J. D. Landes, Eds., American Society of Testing and Materials, Philadelphia, 1994.
4.
"Diablo Canyon-1 '95 Pulled Tubes - Destructive Examination Results," Presented at NRC/PG&E Meeting on Februa'y 23,1996.
5.
" Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates Database for Alternate Repair Limits - 1996 Database Update," EPRI NP-7480-L.
Addendum 1, November 1996.
6.
" Examination of Trojan Steam Generator Tubes - Volume 1: Examination Reptilts," EPRI TR-101427, November 1992.
7.
. Letter from P.M. Beard Jr. (Florida Power C"ooration) to NRC, " Refuel 9 Inspection Plan for Once Thrcuph Steam Generators," D(,u No. 50-302, April 19,1994.
8.
Submittal from PSE&G to NRC, " Reg. Guide.13 21 Assessment of Indications at Salem Unit 2," Docket No. 50-311, August 19,1996.
9.
" Steam Generator Reference Book, Revision 1," EPRI TR-103824, Volume 1 December 1994.
- 10. Letter from Entergy Operations, Inc. to NRC, " Repair Limits for Circumferential Cracks,"
Docket No. 50-368, August 28,1995.
- 11. Letter from P.E. Katz (Baltimore Gas & Electric Company) to NRC, " Steam Generator Tube inspection Results," Docket No. 50-317., August 30,1996.
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ATTACHMENT 2.
SEPTEMBER 25,1998 HOPENFELD MEMORANDUM 4
a
pwo uq\\
UMTED STATES p
g j
NUCLEAR REGULATORY COMMISSION WASHINGTON D.C. 20086 4001 o,g +
September 25, 1998 l
MEMORANDUM TO: Chairman Jackson
[
Commissioner Diaz l
Commissioner McGaff' n FROM ne ask Manager p Generic Safetyt, sues Branch Division of Reguiawry Research t
l Office of Nuclear Regulatory Research
SUBJECT:
J. HOPENFELD'S DIFFERING PROFESSIONAL OPINION CONCERNING VOLTAGE-BASED REPAIR CRITERIA FOR STEAM GENERATOR TUBES: RELEASE OF DPO CONSIDERATION DOCUMENT FOR PUBLIC COMMENT l
l l
The attached document " Differing Professional Opinion (DPO) Regarding NRC Approach to Steam Generator Aging" addresses the issue of primary to secondary leakage of aging steam l
generators during design basis and severe accidents. This issue was originally raised in 1991. It is still unresolved. The documents which are being released for public comment, Regulatory l
' Guide DG-1074 and staff's responses to the DPO, are based on flawed premises. Until the staff can demonstrate that methods for measuring the safety performance of degraded steam generator exist, the present 40% plugging rule should not be relaxed.
. The attached document was originally prepared as an appendix to GL-98-xx which was scheduled to be released in late 1997. The driving forces for its preparation were: (1) to inform the public that GL-98-xx and GL-95-05 allow plar,ts to exceed Commission safety guidelines by wide margins and (2), provide the public with a perspective on NRC's readiness to institute risk-informed, performance-based regulations for steam generators.
i Like its predecessor SG Tube Integrity Rulemaking, GL-98-xx has now been withdrawn; GL 05 is still in effect despite the fact that major safety issues remain unresolved. The information document, DG-1074 and the responses to the DPO are being issued instead. The Advisory Committee on Reactor Safety (ACRS) would not endorse the responses to the DPO without the staff providing furtherjustifications.
The attached document provides an insight as to how the staff has addressed steam generator related issues during the past 6 years. The staff's approach was not based on science; key assumptions were neither stated norjustified, sensitivity studies were limited to a narrow range, justification for selection of models was not provided even though the models contradict known L
physical phenomena, opposing technical views (staff, ACRS )were ignored. A systematic l
adaptation of models to attain a desired outcome was used.
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ATTACHMENT 3
m 2
i The move from prescriptive regulation towards risk-informed, performance-based regulation has the potential of significantly improving the cost effectiveness of nuclear power. However, if not accompanied by adequate controls to insure that proper data is used in risk assessments, this move may compromise public safety.
A true test of the agency's readiness to institute performance based regulation is the existence of a methodology for measuring safety performance of major components. In case of aging steam generators, there is very little data to allow such measurements. The attached document indicates that data is needed to baseline the level of safety of degraded steam generator tubes.
The recent reduction in tube ruptures during normcl reactor operations is unrelated to accident j
leakage.
l The operation of major reactor systems with degraded components raises very difficult issues because of lack of data. Steam line breaks occur very infrequently and therefore it is virtually l
impossible to demonstrate the benefits of a risk informed approach over a relatively short
(
period of time. Innovative solutions to this problem are badly needed. The staff cannot be faulted for failing to produce specific ruidan::e to the industry on how to deal with aging steam generators after 6 years of extensive efforts. What is disturbing, however, is the complete lack l
of planning for collecting data in this regard.
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Attachment:
As stated.
t cc: EDO J. Mitchell i
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