ULNRC-03790, Forwards 90-day Response to GL 97-06, Degradation of SG Internals, Dtd 971230.Response Is Not Intended to Preclude Subsequent Changes Following Normal Administrative Procedures.Ltr Contains No New Commitments
| ML20217N860 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/30/1998 |
| From: | Passwater A UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-06, GL-97-6, ULNRC-03790, ULNRC-3790, NUDOCS 9804090230 | |
| Download: ML20217N860 (16) | |
Text
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U;ian El:ctric One Ameren Plaza
- 1901 Chouteau Avenue PO flox 66149 St. louis, MO 631664149 314 621.3222 March 30, 1998 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station: P1-137 Washington, D.C. 20555-0001 Gentlemen:
ULNRC-03790 g f,,
DOCKET NUMBER 50-483 f
CALLAWAY PLANT g #18 2 /1 UNION ELECTRIC COMPANY UE RESPONSE TO GENERIC LETTER 97-06
Reference:
- DEGRADATION OF STEAM GENERATOR INTERNALS," dated December 30,1997 The reference letter requested licensees to provide a written report that includes a discussion of any program in place to detect degradation of steam
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generator intemals and a description of the inspection plans, including the inspection scope, frequency, methods, and equipment. provides Union Electric's 90-day response to Generic Letter 97-06 and describes programs in place as of the date of this letter, it is not intended to preclude subsequent changes following normal administrative procedures or to require NRC notification or consent for such changes other than those currently required. This letter contains no new commitments.
Should you have any questions or need additional information conceming this matter, please contact us.
Very truly yours, W. -z'cn Alan C. Passwater Manager Licensing and Fuels JMC/
Attachment
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9804090230 980330 PDR ADOCK 05000483 I
P PDR l
a subsufiery of Amoren Corporation
l STATE OF MISSOURI )
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Alan C.
Passwater, of lawful age, being first duly sworn upon oath says that he is Manager, Licensing and Fuels (Nuclear) for Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his j
knowledge, information and belief.
l hh By Alan C.
Passwater Manager, Licensing and Fuels Nuclear l
SUBSCRIBED ar3d sworn to before me this Y
day of
- /7W7 1998.
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i PATNCIAL REYNOLDS mwnrusuo-amuwpusam sT.LDUS COUNTY BffOMA80810NEXFWW OELE M i
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cc:
M. H.
Fletcher Professional Nuclear Consulting, Inc.
19041 Raines Drive Derwood, MD 20855-2432 Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive Suite 400 Arlington, TX 76011-8064 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Barry C. Westreich (2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E16 11555 Rockville Pike Rockville, MD 20852-2738 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 Ron Kuccra Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Denny Buschbaum TU Electric P.O. Box 1002 Glen Rose, TX 76043 Pat Nugent Pacific Gas & Electric Regulatory Services l
P.O. Box 56 l
Avila Beach, CA 93424 l
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i bec: J. Brandt/A160.761
/QA Record ~(CA-758)
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~J. V. Laux-j G..L.
Randolph j
R. J.
Irwin P. M. Barrett J. D. Blosser.
A. C.
Passwater D. E. Shafer-W..E.
Kahl-S. Wideman (WCNOC)
A. J. DiPerna,.(Bechtel)
H. D. Bono NSRB (Patty Reynolds)
J. M. Chapman T. E. Herrmann A160.412 (97-06) i-l i
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ULNRC- 03790 4
I ATTACHMENT ONE RESPONSE TO GENERIC LETTER 97-06
Response to Generic Letter 97-06. Dearadation of Steam Generator internals Introduction Generic Letter 97-06 (GL), " Degradation of Steam Generator Intemals" was issued to: (1) again alert addressees to the previously communicated findings of damage to steam generator intemals, namely, tube support plates and tube bundle wrappers, at foreign PWR facilities; (2) alert addressees to recent findings of damage to steam generator tube support plates at a U.S. PWR facility; (3) emphasize to addressees the importance of performing comprehensive examinations of steam generatorintemals to ensure steam generator tube structuralintegrity is maintained in accordance with the requirements of Appendix B to 10 CFR Part 50; and (4) require all addressees to submit information that will enable the NRC staff to verify whether addressees' steam generator intemals comply with and conform to the current licensing bases for their respective facilities.
This response provides information for Callaway Plant requested by the GL. The information requested includes:
(1) A discussion of any program in place to detect degradation of steam generatorintemals and descriptive inspection plans, including the inspection scope, frequency, methods and equipment. The GL requires discussions to include the foliowing information for each facility:
(a) Whether inspection records at the facility have been reviewed for indications of tube support plate signal anomalies from eddy current testing of the steam generator tubes that may be indicative of support plate damage or ligament cracking (b) Whe'.her visual or video camera inspections on the secondary side of the steam gerarators have been performed at the facility to gain information on the condition of the steam generatorintemals (e.g., support plates, tube bundle wrappers, or other components).
(c) Whether degradation of steam generatorintemals has been detected at the facility, and how the degradation was assessed and dispositioned.
(2) If the addressee currently has no program in place to detect degradation of steam generator intemals, discussion and justification of the plans and schedule for establishing such a program, or why no program is needed.
Although Callaway has Westinghouse Model F steam generators, the following background information is provided. Prior to issuance of the GL, the Westinghouse Owners Group, the Electric Power Research Institute and the Nuclear Energy Institute (NEI) developed an action plan to assess the susceptibility to secondary-side degradation. Callaway Plant intends to follow the industry action plan. Included in the action plan is a requirement to understand the causal factors involved in the degradation experienced in the French Units. This information is captured in EPRI report GC-109558, " Steam Generator Intemals Degradation: Modes of Degradation Detected in EdF Units". This report was submitted to the NRC via NEl letter, dated December 19,1997.
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The Westinghouse Owners Group has reviewed EPRI GC-109558 relative to the design of Series 51 steam generators and determined limited potential susceptibility. For plants with Series 51 steam generators, this conclusion is documented in report WCAP-15002, Rev.1,
- Evaluation of EDF Steam Generator Intemals Degradation -Impact of Causal Factors on Westinghouse 51 Series Steam Generators", December 1997. The 51 Series SGs include Westinghouse model designations 51,51M,51F, and 54F. Because of a similarity in design, this report also covers any implication of susceptibility for two replacement steam generator i
resigns, the Delta 47 and Delta 75. The 51 Series Designs are the most similar to the EdF i
units.
WCAP-15002, Rev.1, documents visualinspections of the plants. it is concluded that the number of plants that have been inspected and the inspection results demonstrate that the l
causal factors identified by EDF do not jeopardize the continued operability of Westinghouse j
Series 51 steam generators. Eddy current inspection of the tubes would detect any detrimental effects on the tubing due to wear caused by tube support plate (TSP) ligament degradation, loose parts, and secondary side flow distribution changes. Foreign object search and retrieval (FOSAR) efforts are conducted to discover loose parts.
A similar detailed evaluation is planned for the remaining types of steam generators (Model 44F, F D3, D4, DS and E1/E2), to be the subject of later reports.
Safety Assessment The following safety concems have been postulated relative to the French steam generator, intemals degradation experience. These are:
Loss of tube support leading to steam generator tube wear and possible primary-to-secondary leakage or inadequate burst margins.
More significant tube support plate deformation during a postulated LOCA +SSE event resulting in unacceptable steam generator tube collapse or secondary-to-primary in-leakage.
The generation of a loose object in the secondary side of a steam generator which may e
result in tube wear or impacting and possibly primary-to-secondary leakage.
Based on a review of Table 1.0, the only degradation types that may occur domestically that might result in the loss of tube support plate integrity are: TSP flow hole / ligaments erosion-corrosion, TSP ligament cracking near the patch plates, and TSP ligament cracking in random areas. There are no observations of post chemical cleaning inspections discovering any significant materiallosses. There are no observations of any wrapper having dropped. There are no observations of TSP ligament cracking or thinning that is progressive and continuing.
TSP ligament cracking or missing pieces of ligaments have been observed, but only in units with carbon steel TSPs with drilled round holes and flow holes. All utilities with Model D and E steam generators with carbon steel support plates inspect a significant percentage of steam generator tubes every : utage with a bobbin probe, eddy current examination. If sections of the tube support plate are missing, this would be readily detectable due to a lack of eddy curren' response at the tube support plate elevation.
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It is expected that there is no increased susceptibility to ligament cracking near the wedge supports in the Model D, E,44F and F steam generator designs. Existing calculations evaluating the effects of LOCA + SSE loadings on the tube bundle continue to apply in determining whether certain tubes should be excluded from application of the voltage-based plugging criteria or whether certain tubes should be removed from service in plants which do not currently apply such a criteria but which may have steam generator tubes experiencing cracking at the tube support plate intersections.
Another occurrence resulting from steam generator intemals degradation that may affect a steam generator in performing its intended safety function is the potential for tube wear and primary-to-secondary leakage due to the generation of a loose object on the secondary side of the steam generator. This may occur due to erosion-corrosion of the moisture separators, tube suppt:t plate flow holes, preheater water box erosion / corrosion or the occarrence of tube support plate ligament cracking. If primary-to-secondary leakage should occur due to tube wear from a loose object, the expected consequences would be bounded by a single tube rupture event and, therefore, would remain within the current licensing bases of a plant.
Background
As discussed in WCAP-15002, Rev.1, surveys were sent to all WOG utilities requesting the results of all steam generator, secondary side inspections and relevant tube inspections for tube support plate conditions. Completed surveys were received for 37 of 49 plants. For the Model D, E,44F and F steam generators, responses were received for 12 plants. Of these,11 responded as having inspected or reviewed inspection data for TSP ligament indications and 8 having performed SG secondary side entries that give confidence of not having wrapper drop.
TSP ligament indications were not found in steam generators with either carbon or stainless steel support plates.
The modes of degradation detected include many cases of flow-assisted corrosion, or erosion-corrosion, and of premature cracking that results from either surface fatigue or from corrosion cracking that is associated with surface conditions such as pitting or geometric concentrations.
For the most part, however, the surveys do not report os 9ction of several modes of I
degradation experi?nced in the EdF units. There is no ei,ance of post chemical cleaning inspections discovering any significant materiallosses. There is no evidence of any wrapper having dropped. There is no evidence of TSP ligament cracking or thinning that is progressive and continuing. TSP ligament cracking or missing pieces of ligaments have been observed, I
but only in units with carbon steel support plates with drilled round tubn holes and flow holes.
These conditions are generally traceable to initial inspections and are not progressing based on sequential inspection data. Many of the conditions are probably related to original TCP drilling alignment. There are cases of indications in TSPs that have been linked to patch plate welds.
l Plants with significant hour glassing of the tube support plates as a result of the denting process have exhibited ligament cracking throughout the thickness of the support plate between the flow holes in the plate or the flow holes in the tube lane. If denting remained uncontrolled, as subsequent support plate corrosion occurs, the potential exists for fragments
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of the support plate material to become completely free of the main TSP structure. However, 3
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these plate segments generally remain locked in place because of the in-plane forces that give rise to denting, as well as the deformation that contains the individual piece. Operating plants I
with active denting are under periodic monitoring by the utility and have long-standing criteria and review by the NRC. In addition, the EdF experiences reported are not related to support plate degradation that has progressed to the tube denting stage. These plants are therefore not included in this response to GL 97-06.
The secondary side intemal degradation types found in Westinghouse steam generators are identified in Table 1.0.
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4 Table 1.0 Secondary Side Internal Degradation Types in Westinghouse Design SGs SG Feed Ring Preheat Feed Ring Preheat Category:
Carbon Steel Carbon Steel Stainless Steel Stainless Steel TSPs TSPs TSPs TSPs Degradation Type Erosion-Corrosion:
1 Moisture separator X-S X
S Water Box NA XW NA S
TSP Flow S
S NA NA Hole / Ligaments Feed Ring /J-Tubes X
NA X()
NA Cracking TSP X
S L
L W')
Ligaments Wrapper Near L
L L
L Supports
- Transition X
L-X()
L Cone Girth W eld Other Wrag)er L
L L
L Drop l
X= Observedin some steam generators S = Susceptible L = Low Susceptibility NA = Not Applicable (1) Various indications of possible tube degradation may be artifacts of manufacturing anomalies related to patch plate welds and drilling alignment
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(2) Various Westinghouse design features are beneficialrelative to some steam generator design features of foreign manufacturers.
(3) In SG replacements with the originalshelland/orupperintemals not replaced (4) This mechanism does not apply to the Model D3 because of the Alloy 600 inlet manifold design used.
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Discussion of Inspection'Results
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As a detailed evaluation has not been completed for the Model F steam generators installed in Callaway. Plant, inspection recommendations, although planned, have been defined on an '
interim basis.1 For each of the steam generator types, the more detailed evaluation should be completed by the end of May of 1998. A response to item 1 for Callaway Plant's steam generators, along with a discussion about a tentative inspection plan is provided below. Item 2 of the GL does not apply based on our interim program and plans to update it following the -
detailed evaluations discusced above.
Callaway Plant is equipped with Westinghouse Model F steam generators. Each steam r
' generator has 5626,11/16" tubes. Support plates are stainless steel with quatrefoil tube :
holes. The tubes were fully expanded in the tubesheet using a hydraulic expansion method. It is not expected that the detailed evaluation described above will add any significant findings for Callaway Plant's steam generators. However, when the evaluation is issued, we will adjust.
our secondary inspection program accordingly to address any concems that are identified.
it is Union Electric's position that loose objects should be removed from the steam generator, whenever possible. Tubes observed to have visible damage should be eddy current inspected
~and plugged if found to be defective.
For the types of steam generator intemals degradation that could occur at Callaway Plant, it is expected that degradation would be limited in extent such that the tubes will remain capable of sustaining the conditions of normal operation, including operational transients, design basis accidents, extemal events, and natural phenomena permitting the affected steam generator to perform its intended safety function. Eddy current inspection, foreign object search and retrieval (FOSAR) activities (during each refueling outage) and loose parts monitors help to ensure the maintenance of tube integrity during subsequent plant operation A steam generator inspection summary for Callaway Plant is included in Table 2.0 below.
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i Table 2.0 SG Inspection Summary for Callaway Plant S/G Date Type ofInspection Extent and Results ABCD May 1986 FOSAR Secondary Tubesheet and Blowdown Lane - No objects or degradation found.
ABCD Apr 1987 FOSAR Secondary Tubesheet and 1
Blowdown Lane-Small foreign object (nut / bolt) found in B and removed.
ABCD Apr 1989 FOSAR Secondary Tubesheet and Blowdown Lane - Small nail found in A and removed.
ABCD Oct 1990 FOSAR Secondary Tubesheet and Blowdown Lane-Small wire brush handle found in D and removed.
A Oct 1990 In-Bundle Inspection In-bundle video examination of hot and coldlegs at tubesheet - No sludge pile identified.
A Oct 1990 Upper Internals and J-Nozzles Examined general condition of (UT) upper bundle and performed UT on every 4* J-nozzle - Good general condition. No wear at J-nozzles.
A Oct 1990 Girth Weld Inspection Examined upper transition girth weld - No pitting in weld identified.
ABCD Apr 1992 FOSAR Secondary Tubesheet and Blowdown Lane-Roll pin found in D and removed.
BC Apr 1992 In-Bundle Inspection in-bundle video examination of hot and coldlegs at tubesheet - No sludge pile identified.
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I S/G Date Type ofInspecthm Extent and Results
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i BC Apr 1992 Tube Support Plate Insp.
To bottom of 3"' support plate -
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No problems identified.
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D Apr 1992 Upper Internals Inspection General cotxtition and J-nozzles l
(UT)- No problems identified.
i ABCD Oct 1993 FOSAR Secondary tubesheet and I
blowdownlane - Piece of gasket ring removed from D, piece of weld slag removed from A.
A Oct 1993 Tube Support Plate Insp.
To bottom of 5* support plate -
Identified heavy fouling on upper sections of tubes. No other problems identified.
ABCD Apr 1993 FOSAR Secondary tubesheet and blowdown lane-Small piece of flat steel found and removedin D.
D Apr 1995 In-process inspection during Examined 7* support plate prior j
chemical cleaning, to start of cleaning, and after both j
iron and copper steps - Chemical cleaning appeared to be very i
effective.
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-ABCD Apr 1995 FOSAR Secondary tubesixet and blowdownlane-Piece of check valve seal ring found and removed in B. Small piece found between tubes in C. Unable to remove.
i Processed evaluation to allow to I
remain in place. Small nail removed from D.
j ABCD Apr 1995 Tube Support Plate Insp.
Up to bottom of 7* support plate -
All S/Gs very clean.
ABCD Apr 1995 Upper Internals Inspection General condition-No problems identified.
BC Apr 1995 J-Nozzle Video Insp.
5 J-nozzles in each-Nonje to feed ring weldin good condition.._
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1 S/G j.
Date Type ofInspection Extent and Results ABCD Oct 1996 FOSAR Secondary tubesheet and blowdownlanc-Found and removed relativelylarge split pins (6"in length)in A and C. Misc, other smallitems found and removed in all S/Gs except D.
AC Oct 1996 Tube Support Plate Insp.
Bottom of 7* support plate-S/Gs still very clean from chemical cleaning. No other concerns-(
identified.
C Oct 1996 Upper Girth Weld Insp.
Visual examination of upper girth weld. General condition of upper intanals - Upper internals and girth weld in excellent conditions.
No problems identified.
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in Service Inspection Plan To Address item 1(a), (b), And (c)
Based on the above, the following inspection plan has been implemented at Callaway Plant.
Except where noted, these inspections will be completed each refueling outage. Inspection scope and frequency may be adjusted as necessary based on site specific experience and evaluation of industry results of these inspections.
Item 1(a)
Tube Support Plate Lloament Erosion / Corrosion and Crackina This is considered a low susceptibility event since the tube support plates in the Model F steam generators in Callaway Plant are stainless steel. A number of studies have reported that the
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chromium content in carbon steel has a significant effect on resistance to a erosion / corrosion 1
mechanism. The chromium content of stainless steel, support plates is expected to preclude the occurrence of this degradation mechanism. Eddy current inspection is not applicable for j
Callaway Plant, which has support plates with quatrefoil broached holes. Therefore, a j
historical review of eddy current inspection records has not been performed.
As Callaway Plant has early Model F design steam generators, a sample inspection of the patch plate plug weld regions will be made. In addition, a sample inspection of the top support plate tube lane region (where flow holes are provided instead of elongated slots) will be conducted. These inspections will be rotated between steam generators each outage to ensure all are examined within a reasonable period of time. Flow holes are used to provide strengthening of the top tube support plate for u-bend support. If initial drilling produced a separated ligament, an evaluation of the effect on the u-bend will be made.
Wrapper Droo:
1.
It will be verified that the sludge lance equipment can be inserted without interference.
Sludge lancing is performed at Callaway Plant each outage.
2.
If interference with the sludge lance equipment is detected, the lower wrapper support blocks will be visually inspected.
Wrapper Crackina:
No inspection is planned unless evidence of wrapper mis-position or tube damage in the periphery of the first tube support plate is detected. If degradation is detected, a visual inspection of the lower wrapper support blocks will be conducted.
Upcer Packaae:
Primary and secondary moisture separators, feed ring, (J-tube, carbon steel feed ring adjacent to J-tubes, T section, reducer, backing ring and thermal sleeve). The significance to tube integrity as a result of tube degradation of these components is primarily a loose part.
Inspections will be performed to verify the condition and integrity of the upper package components. These inspections will be performed on a rotating bases between steam 10
generators each outage to ensure all components are examined within a reasonable period of time.'
Transition Cone Girth Weld:
This area is ultrasonically inspected in accordance with the steam generator shell,Section XI Inservice Inspection requirements. In addition, Callaway typically performs a visual inspection of the transition girth weld. This will be performed on a rotating bases between steam generators each outage to ensure the welds are examined within a reasonable period of time.
Feedwater Nozzle Degradation of the themial sleeve may affect the feedwater nozzle. Loose parts monitoring, which is an on-going process, and in service inspection requirements for the feedwater nozzle will be completed.
Item 1(b)
See Table 2.0 for inspection summary of Callaway Plant's Steam Generators.
Item 1(c)
To date, no secondary side intemal degradation has been identified at Callaway Plant.
Conclusion it is expected that there is no increased susceptibility to ligament cracking near the wedge supports in the Model F steam generator design used at Callaway Plant. Existing calculations evaluating the effects of LOCA + SSE loadings on the tube bundle continue to apply in determining whether certain tubes should be excluded from application of the vc!tage-based plugging criteria or whether certain tubes should be removed from service in plants which do not currently apply such criteria but which may have steam generator tubes experiencing cracking at the tube support plate intersection.
Callaway Plant will continue to monitor industry developments and will tailor the secondary side inspection program to ensure any future intemals degradation is promptly identified.
Reference 1.
WCAP-15002, Rev.1, ' Evaluation of EDF Steam Generator Intemals Degradation -
Impact of Causal Factors on Westinghouse Series 51 Steam Generators" 11
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