ML20086A624
| ML20086A624 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/27/1995 |
| From: | Mccoy C GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-95-03, GL-95-3, LCV-0626, LCV-626, NUDOCS 9507030316 | |
| Download: ML20086A624 (21) | |
Text
,
Georg'a Power Company i
40 Invemess Center Parkway
'Pom o'tce Box 1295 Ermingham. Alabama 35201 T(6 phone 205 877 7122 I
GeorgiaPower l
C.K.McCoy-Vce Presdent, Nuclear -
_Vogne Protect the southem clex ric system June 27, 1995 LCV-0626 Docket Nos.: 50-424, 50-425 t
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:
I VOGTLE ELECTRIC GENERATING PLANT RESPONSE TO NRC GENERIC LETTER 95-03 Enclosed please find the response to NRC Generic Letter 95-03, "Circumferential Cracking of Steam Generator Tubes", for Georgia Power Company, Vogtle Electric Generating Plant, Units 1 and 2. Contained therein is information which is provided in response to the NRC's request for information as delineated in Items 1 and 2 on page 4 of the subject Generic Letter.
Mr. C. K. McCoy states that he is a Vice President of Georgia Power Company and is authorized to execute this oath on behalf of Georgia Power Company and that, to the best l
of his knowledge and belief, the facts set forth in this letter are true.
Georgia Power Company 5
By:
C. K. McCoy Sworn to and subscribed before me this A7b ay of b 1995.
d U
W$
Notary Public t
MYCOMMISSION EXPlRES
. W R.1998 l
v y
9507030316 950627
}I i
PDR ADDCK 05000424 i
P PDR
1
?
i
[-
- l MeOrgiaIt)Wer U. S. Nuclear Regulatory Commission e
LCV-0626 PageTwo CKM/JAE/jae.
Enclosures:
1.
Response to NRC Generic Letter 95-03
- 2. Inspection Summary for VEGP-1 Outages 1R4 and IRS and VEGP-2 Outages 2R3 and 2R4
.l xc: Georgia Power Company Mr. J. B. Beasley, Jr. (w/encls.)
Mr. W. L. Burmeister (w/encls.)
Mr. M. Sheibani (w/encls.)
. NORMS (w/encls.)
U. S. Nuclear Regulatory Commission Mr.. S, D. Ebneter, Regional Administrator (w/encls.)
)
~Mr. D S. Hood, Licensing Project Manager, NRR (w/encls.1 i
Mr. B. R. Bonser, Senior Resident Inspector, Vogtle (w/encls.)
l 1
r
)
700775 -
ENCLOSURE 1 TO GEORGIA POWER COMP /NY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" Vogtle Electric Generating Plant, Units 1 and 2 NRC Docket Nos. 50-424,50-425 United States Nuclear Regulatory Commission (NRC) Generic Letter 95-03, "Circumferential Cracking of Steam Generator Tubes", was issued on April 28,1995 to all holders of operating licenses or construction permits of pressurized water reactors (PWRs). The NRC issued the Generic Letter to (1) provide notification of recent steam generator tube inspection fmdings at Maine Yankee Atomic Power Station and the safety significance of those fmdings, (2) request that actions as described in the Generic Letter be implemented, and (3) require the submittal of a written response to the Generic Letter regarding implementation of the requested actions. In addition, Generic Letter-95-03 emphasized the importance of performing comprehensive inspections of steam -
generator tubes using techniques and equipment capable of reliably detecting degradation to which steam generator tubes may be susceptible.
Actions specifically requested of all holders of operating licenses and construction permits for PWRs by NRC Generic Letter 95-03 included the following:
(1) Evaluate recent operating experience with regard to the detection and sizing of circumferential indications to determine the applicability to one's plant, (2) On the basis of the evaluation in Item 1 above, past inspection scope and results, susceptibility to circumferential cracking, threshold of detection, expected or inferred crack growth rates, and other relevant factors, develop a safety assessment justifying continued operation until the next scheduled steam generator tube inspections are performed, and (3) Develop plans for the next steam generator tube inspections as they pertain to the detection of circumferential cracking. The inspection plans should address, but not be limited to, scope (including sample expansion criteria, if applicable), methods, J
equipment, and criteria (including personnel training and qualification).
Upon completion of the foregoing actions, the following information was required to be submitted to the NRC:
El-1
4
'.r e
ENCLOSURE 1 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued)
(1) A safety assessment justifying continued operation that is based on the evaluations performed in accordance with Items 1 and 2 above, and (2) A summary of the inspections plans developed in accordance with Item 3 above and a schedule for the next planned inspection.
Georgia Power Company (GPC) has completed the required actions described as Items 1, 2, and 3 in NRC Generic Letter 95-03 for Vogtle Electric Generating Plant, Units 1 and 2 (VEGP-1 and 2) and offers the following in response to NRC Requested Informaiion Items 1 and 2:
The Westinghouse Model F steam generators utilized at VEGP-1 and 2 contain 11/16-inch diameter steam generator tubes which are manufactured from Inconel-600. Each of the four steam generators per VEGP unit contains 5,626 tubes. These tubes are thermally-treated and are hydraulically-expanded the full depth of the tube sheet. The Effective Full Power Years (EFPY) experienced are 6.9 and 5.2 for VEGP-1 and 2, respectively. The Temperature-Hot (b) for both VEGP units is 618 degrees Fahrenheit.
Based on information provided by Westinghouse, Model F steam generators that are installed at various PWRs, including VEGP-1 and 2, have not experienced any circumferential cracking of tubes manufactured from thermally-treated Inconel-600 material. Since there have not been any reported occurrences of circumferential cracking in plants having thermally-treated, hydraulically-expanded Inconel-600 steam generator tubes, no rotating pancake coil (RPC) eddy current inspections have been performed at VEGP-1 and 2 to inspect for circumferential cracking. RPC has been used on a limited basis to confirm indications discovered by bobbin coil eddy current inspection. Steam generator tubing inspections at VEGP-1 and 2 are performed in accordance with a program recommended by the Electric Power Research Institute (EPRI) in their document NP-6201 which provides guidelines for inservice inspection of steam generator tubes. In the VEGP program, fifly percent (50%) of the tube population is inspected from two steam generators typically each maintenance / refueling outage rather than twenty percent (20%) of the tube population from four steam generators each cycle such that after four cycles of plant operation, each of the tubes will have been tested. Our inspection program offers advantages which include shorter inspection time in order to achieve complete steam generator inspection and improved sensitivity to degradation that might be initiating El-2
ENCLOSURE 1 TO GEORGIA POWER COMPANY LETTER LCV-0626, j
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued) within a particular steam generator. In addition, the test program includes recommendations for additional testing for any areas of concern as specified in the EPRI guidelines. Results of the bobbin coil eddy current inspections performed to date have previously been provided to the NRC as required by plant Technical Specifications 4.4.5.5.a and 4.4.5.5.b. In each of the reports submitted, especially those submitted pursuant to plant Technical Specification 4.4.5.5.b, information is included which addresses, but is not limited to, the number and extent of tubes inspected, location and percent of wall-thickness penetration for each indication of an imperfection, and identification of plugged tubes. Reports submitted pursuant to plant Technical Specification 4.4.5.5.a typically only identify the specific tube (s) which may have been plugged during a maintenance / refueling outage.
Although RPC testing has never been performed at VEGP-1 and 2 except to confirm indications discovered by bobbin coil cddy current inspection, continued plant operation until the next scheduled maintenance / refueling outage for each VEGP unit can bejustified based on the following which was developed with input from Westinghouse Electric Corporation:
Inconel-600 Thermally-Treated Tube Material Plants Thermally-treated Inconel-600 tubing represents an intermediate step in the evolution of progressively optimized corrosion-resistant tubing materials. EPRI report NP-3501,
" Optimization of Metallurgical Variables to Improve Corrosion Resistance on Inconel Alloy 600", identifies the distinct advantages of thermally-treated Inconel-600 over mill-annealed Inconel-600. Data contained in that EPRI repon indicates minimal stress corrosion cracking in thermally-treated Inconel-600 c-rings at 600 degrees Fahrenheit in caustic solutions (ten percent sodium hydroxide). Crack depths were generally 2.5 to 4.5 times less than mill-annealed Inconel-600. Also, no stress corrosion cracking was detected in thermally-treated Inconel-600 small radius U-bends tested at 680 degrees Fahrenheit. The EPRI report also showed a dependence upon residual stress level and crack growth rate and initiation times. Westinghouse data has shown that the stress levels in hydraulically expanded tubing are less than the associated levels in either explosively or mechanically-expanded tubes. Some plants utilizing thermally-treated, hydraulically-expanded Inconel-600 tubes operate at lower temperatures than the plants with mill-El-3
l ENCLOSURE 1 TO GEORGIA POWER COMPANY I-LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued) annealed, hydraulically-expanded Inconel-600 tubes. Since the stress corrosion cracking rate is temperature dependent, a lesser potential for rapid stress corrosion cracking would be expected.
Pulled Tube Examination Results In 1990, two thermally-treated, hydraulically-expanded Inconel-600 tubes were pulled from the replacement steam generators at Virginia Power Company's Surry Unit 1. Field nondestructive examination (NDE) data suggested the presence of circumferentially oriented degradation. Upon further review, it was concluded that the poorly defined RPC signal for one of the tubes was similar to that of a " ding" or mechanical deformation.
Upon destructive testing of the subject tube, no stress corrosion cracking on either the inside diameter (ID) or outside diameter (OD) was detected. The source of the NDE indications was determined to be attributed to eddy current probe lifloffin the expansion transition and mechanical imperfections in the tube resulting from the tube installation process. The maximum diameter of the second tube occurred approximately 0.6 inch above the top of the tubesheet. A 70 degree " groove" which was " mechanical" in nature was found on the OD of the tube and was attributed to the interaction of the tube with the edge of the tubesheet during the expansion process. It has been concluded that this second tube was improperly expanded resulting in an over-expansion of the tube above the tubesheet. The hydraulic expansion process used was designed to locate the transition slightly below the top of the tubesheet.
End-of-Cvele (EOC) Structural Limit Crack Anele Calculations No detectable degradation has been observed at plants utilizing thermally-treated Inconel-600 tube material. Based on the extended plant operational periods to date, which extend from 1980, it is unlikely that rapid tube degradation in the expansion transitions in plants utilizing the thermally-treated Inconel-600 tubes would occur prior to the next scheduled tube inspection or on a cycle-to-cycle basis in the near future. EOC structural limits for 7/8-inch OD Inconel-600 thermally-treated tubing would be consistent with the available data developed initially for Westinghouse's WEXTEX-tubed plants. Despite the fact that no circumferential degradation has been detected in these units having Inconel-600 thermally-treated steam generator tubes, EOC structuralihmtt are provided below for El-4 l
l
)
i
4 ENCLOSURE 1 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued) completeness.
To permit a rapid scoping assessment for tube burst capability of circumferential indications, a burst correlation was developed for through-wall circumferential indications.
The burst correlation was then applied to define the structural limit on through-wall crack angles that satisfy NRC Regulatory Guide 1.121 burst margin for 3 times normal operating pressure difTerential. If measured RPC crack angles, afler reduction for coil lead-in and lead-out effects (about 30 degrees) for through-wall indications are less than the structural limit, it can be readily concluded that the indications satisfy burst margin guidelines. If the, measured RPC angles exceed the assumed through-wall structural integrity limit, additional inspection such as ultrasonic inspection or structural analysis are needed to assess structural integrity.
The development of the through-wall crack angle structural limit is described as follows.
Utilizing the burst correlations developed from electrical discharge machining (EDM) notch data and analytical models, the structural limits for through-wall circumferential indications were developed as given for the crack models in the following tables. The burst pressure data were adjusted to account for lower tolerance limit material properties.
7/8 Inch Tubing EOC Structural Limits for Circumferentially Oriented Degradation Single Through-Single Through-wall Segmented wall Crack with 50%
Through-wall Crack Model Degraded Ligament Crack Model 3 delta P = 4500 psi 210 degrees 210 degrees 264 degrees 3 delta P = 4300 psi 226 degrees 226 degrees -
269 degrees SLB delta P=2560 psi 321 degrees 283 degrees 318 degrees El-5
J '.
ENCLOSURE 1 TO J
GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued) 11/16 Inch Tubing, EOC Structural Limits for Circumferentially Oriented Degradation Single Through-wall Crack with 50% Degraded Ligament 3 delta P = 3750 psi 247 degrees SLB delta P = 2560 psi 283 degrees The single through-wall crack model is applicable for both ID or OD degradation. The segmented model is typical of primary water stress corrosion cracking (PWSCC). The through-wall plus fifly percent (50%) deep model was developed to represent 360 degree -
indications frequently found for outer diameter stress corrosion cracking (ODSCC).
Therefore, a single, uniform through-wall crack of 247 degrees with 50% deep OD degradation existing over the remaining 113 degrees of tube arc would satisfy the NRC Regulatory Guide 1.121 three delta P (3 delta P) burst recommendations while a 283 degree through-wall crack with 50% deep OD degradation existing over the remaining arc would have a predicted burst pressure of 2560 pounds per square inch (psi) for 11/16-inch 7
OD tubing. Most likely, based on the industry-accepted detection thresholds for circumferential cracking of this type ofindication, if encountered in the field, it would produce an RPC crack angle of nearly 360 degrees.
t In addition, it is expected that through-wall cracks of that size would experience primary-to-secondary leakage, alerting plant operators of the tube condition. Such was the case for the kinetically welded sleeve leakage events at Duke Power Company's McGuire plant. For the McGuire events, the leakage was able to be trended and was readily detectable. The morphology of the cracks at McGuire was more that of single uniform cracks since the morphology was driven more by high residual stress than by intergranular corrosion.
Based on development work done for 7/8-inch tubing, the corresponding crack lengths for single uniform through-wall cracks without additional degradation in the remaining ligament would be consistent at the 3 delta P condition and greater than 283 degrees for 11/16-inch OD tubing at a steam line break (SLB) delta P of 2560 psi. Based on the similarity of the limiting crack angles for the two tubing sizes, the limiting single through-El-6
J '.
i ENCLOSURE 1 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued) wall crack angle which would support a burst capability of 2560 psi would be approximately 310 to 320 degrees.
Inspection Methodologies and Adequateness Steam generator tubing inspections at VEGP-1 and 2 are performed in accordance with a program as described herein. During several recent maintenance / refueling outages, the steam generator tubing inspection programs were expanded due to power uprate and Westinghouse recommendations concerning anti-vibration bar (AVB) wear.
This portion of thejustification for continued operation will address the previous two 4
maintenance / refueling outages on each VEGP unit; specifically, VEGP-1 Outages IR4 and IRS and VEGP-2 Outages 2R3 and 2R4. Steam generator tubing inspections consisting primarily of bobbin coil eddy current inspections during those outages were performed by Westinghouse using their procedure MRS 2.4.2 GPC-3, " Eddy Current Inspection of Preservice and Inservice Heat Exchanger Tubing" The methods and techniques described in that procedure are applicable for either preservice or inservice inspections, and for use on bobbin coil,8 x 1 surface riding,8 coil profilometry, double crosswound, and RPC probes. RPC was utilized during VEGP-2 Outage 2R3 on a single tube. No other RPC inspections were performed in these outages. Bobbin coil eddy current inspections were performed using Echoram-manufactured eddy current inspection probes typically 0.520, 0.540, and 0.560-inch in diameter. Withdrawal rate during these inspections has been 24 inches per second and the test frequencies were 630 kilohertz (KHz),320 KHz,160 KHz, and 10 KHz. Primary and secondary analysis of the eddy current data was performed by Westinghouse personnel utilizing their analysis software, ANSER. Data management was accomplished using SUPERTUBIN sof1 ware. GPC-site specific data analysis guidelines were used to perform the analysis. Each analyst was required to complete a training session and then be performance tested. Analyst training meets the requirements of EPRI document NP-6201, Appendices G and H. Previously, Westinghouse prepared a program which included training tapes, test tapes, and a training manual for GPC. Analysts were required to pass the testing prior to analyzing any new data. Each analyst was trained and tested in bobbin analysis. Typically, a portion of the analysis group also received the RPC training and testing.
E l-7
ENCLOSURE 1 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued)
Details on the inspections discussed above, including steam generator tubing which required plugging, have been previously made available to the NRC through submittal of written reports required by plant Technical Specifications 4.4.5.5.a and b. These reports typically were provided as part of the " Owner's Reports for Inservice Inspection" The reports, particularly those submitted under Technical Specification 4.4.5.5.b, address, but are not limited to, the number and extent of tubes inspected, location and percent of wall-thickness penetration for each indication of an imperfection, and identification of plugged tubes. Provided as Enclosure 2 to this letter is a summary of bobbin-coil eddy current inspections conducted at VEGP-1 and 2 and the results thereof for VEGP-1 Outages IR4 and 1R5 and VEGP-2 Outages 2R3 and 2R4.
While RPC has not typically been used at VEGP-1 and 2 for the inspection of circumferential cracking, industry experience has not identified any problems with circumferential cracking in Model F steam generators having thermally-treated Inconel-600 steam generator tubes. As a result, no RPC has been performed for the express purpose of circumferential crack detection in tubes. In addition, the existing procedure and techniques utilized at VEGP-1 and 2 are deemed adequate based on the identification of pluggable indications on both units. To date, twenty-seven (27) tubes have been plugged in the four steam generators at VEGP-1 while eighteen (18) tubes have been plugged in the VEGP-2 steam generators. The number of tubes plugged on each VEGP unit includes tubes plugged preservice, i.e., plugged at the manufacturing facility, and since startup of each VEGP unit. In certain cases, tubes have been plugged administratively to preclude any possible future problem.
Individual Plant Tube Inteerity Assessment The past two inspection programs at VEGP-1 and 2 have been performed consistent with the EPRI and industry guidelines regarding calling criteria guidelines and initial sample inspection size such that any structurally significant indications would have been identified.
There have been no instances of degradation of any sort, axial or circumferential, in the expansion transitions of thermally-treated tubing. For the sake of completeness, postulated EOC circumferential crack angles will be considered. Postulated EOC crack angles would be projected to be well below the EOC structural limits for single or single through-wall cracks with 50% degradation in the remaining ligament, as listed herem.
l El-8
i
,=
s ENCLOSURE 1 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03"
?
(continued)
Since there are no domestic operating experiences of circumferential indications in full I
depth hydraulically-expanded plants with Inconel-600 thermally-treated tubing, it is reasonable to assume that growth rates of postulated circumferential indications are negligible, and would be considered quite conservative to assume growth rates of any value, based on currently available data. When considering these negligible growth rates and the industry-accepted detection thresholds for through-wall circumferential j
degradation, no indications would be expected at VEGP-1 and 2 which would challenge tube integrity at the end of the current operating cycle. Similarly, tube structural integrity l
would be expected to be maintained during all future operating cycles considering the historical performance of these hydraulically-expanded tubes, and also assuming that plant
]
operating parameters will not be significantly altered from current conditions.
Ilistorical 1,cakage Events in Westinehouse Steam Generator Tubes Addressed in this section are circumferentially oriented tube degradation in Westinghouse PWR plants with non-sleeved tubes such as VEGP-1 and 2.
Several leakage events have been attributed to WEXTEX-tube region cracking and to U-bend tangent point leakage. Low level (<20 gallons per day [gpd]) leakage was detected during the operational cycle prior to the 1987 inspection outage at Virginia Power's Nonh Anna Unit 1. Other low level leakage events may have occurred in some 3/4-inch full-depth hardroll expansion plants. Several small (<20 gpd) events may have occurred in steam generator outer row U-bends, e g., McGuire Unit 1. Also, a Row I leaker of approximately 200 gpd was found at Duke Power's Catawba Unit 1 in 1988.
Both of these events had dent signals associated with the cracking.
It should be noted that in each case described above, the tubing material was different from that utilized at VEGP-1 and 2 in that Inconel-600 mill-annealed tubing typically was involved at those plants experiencing leakage.
Defense in Depth Westinghouse plants with thermally-treated laconel-600 tubing utilize full depth hydraulic expansion. Plants with thermally-treated Inconel-600 tubing have been operating since 1
I El-9
i 6
ENCLOSURE 1 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued) 1980 with no reports of corrosion degradation. There are no outside driving factors which suggest that rapid corrosion degradation of thermally-treated Inconel-600 tubing would be experienced, either up to the end of the current operating cycles for these units, or during any cycle in the near future.
Should through-wall cracking occur for some unforeseen reason, the emergency operating procedures (EOPs) used at VEGP-1 and 2 are specifically designed to respond to single and multiple tube rupture scenarios. The NRC has performed additional analysis efforts as outlined in Draft NRC NUREG-1477 and NUREG-0844 which indicate that the refueling water storage tank would not become depleted during response to multiple tube ruptures and, as such, does not represem a large core damage frequency probability.
In an effort to minimize sludge buildup, sludge lancing has been performed on each steam generator at VEGP-1 and 2 each maintenance / refueling outage to remove any aggressive chemicals from the steam generators before they create sludge piles and attack the tubes.
In addition, Pressure Pulse Cleaning (PPC) has been performed on each steam generator.
PPC is currently scheduled to be performed every five operating cycles. The average amount of sludge removed during each maintenance / refueling outage has been approximately 27 pounds per steam generator. The amount of sludge removed from the steam generators has increased when PPC is performed. The following tables identify the total sludge removed from each steam generator during the two most recent maintenance / refueling outages for each VEGP unit, VEGP-1 Outages 1R4 and 1R5 and VEGP-2 Outages 2R3 and 2R4:
VEGP-1 Outage 1R4 - Total Sludge Removed with Sludge Lancing Steam Generator 1
2 3
4 Total Number Sludge j
Removed in 10 15.5 12.5 15 53 Pounds
)
I l
El-10 i
.X i
ENCLOSURE 1 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued)
VEGP-1 Outage IRS - Total Sludge Removed with Sludge Lancing and PPC Steam Generator.
1 2
3 4
. Total Number -
Sludge Removed in 72 48 56.5 64 240.5 l
Pounds VEGP-2 Outage 2R3 - Total Sludge Removed with Sludge Lancing Steam Generator 1
2 3
4 Total Number Sludge Removed in 13 12.5 13 19 57.5 i
Pounds i
VEGP-2 Outage 2R4 - Total Sludge Removed with Sludge Lancing and PPC Steam f
Generator 1
2 3
4 Total Number Sludge Removed in 21 17.5 20 18.5 77 l
Pounds Westinghouse has performed sludge height analysis on the VEGP-1 steam generators in an effort to quantify the amount of sludge on top of the tubesheet. The eddy current inspection data from VEGP-1 Outage IR5 was utilized for the sludge height analysis.
There was no measurable sludge detected on the bafIle plate or support plate structures.
l Measurable sludge was detected only at the tubesheet on a total of 129 steam generator tubes for all four steam generators. The sludge piles are small, localized areas near the center of the tubesheet, close to the tube lane. These locations are consistent with El-11 l
l l
l_
l ENCLOSURE 1 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued) observat!ws of other steam generators. The sludge height measurements ranged from a height of 0.52 inches to 4.42 inches with the average sludge height being 0.95 inches for these 129 tubes. Visual inspections of the tubesheet are performed each i
maintenance / refueling outage in order to quantify the amount of sludge in the " kidney" areas of each steam generator. These visual inspections have not identified any significant amounts of sludge on top of the tiibesheet or any excess tube fouling on the steam generator tubes. In addition, visual inspections of the upper suppon plates, i.e., the first support plate through the sevei.th support plate, are performed on one steam generator each maintenance / refueling outage. No abnorma! amounts of sladge or tube fouling have been observed during these visual inspections.
Vogtle procedure 31055-C, " Chemistry Parameters Trending and Data Correlation", is used to trend or identify any excursions in plant chemistry. When chemical excursions or anomalies are discovered, the anomalies are resolved as soon as possible utilizing procedural controls. It should be noted that no extended, abnormal chemistry excursions have occurred at VEGP-1 and 2. However, two abnormal chemistry excursions occurred l
on VEGP-2 due to resin intrusions. The total duration of the two events was ten hours.
We do not believe the steam generator tubing was adversely affected based on the shon duration of the events, subsequent steam generator hideout return results, and eddy current inspections.
Summary of Future Inspection Plans at VEGP Because of concerns with circumferential cracking of steam generator tubes as discussed in NRC Generic Letter 95-03, GPC commits '.o performing an enhanced inspection at V1 J'P 1 and 2 during the next scheduled maintenance / refueling outages for each unit whi, Vgin in Spring 1996 and Fall 1996 for VEGP-1 and 2, respectively. Specifically, GPC ;ommits to inspecting 300 tubes per steam generator during those particular outages from the hot-leg side in each of the steam generators scheduled for eddy current inspection for the express purpose of circumferential cracking detection. Typically each j
maintenance / refueling cutage, selected tubes in two steam generators are inspected using an inspection program as discussed herein. The tube inspections shall include the area 3 inches below the top of the tubesheet to 3 inches above the top of the tubesheet. In the El-12
~.
ENCLOSURE 1 l
TO GEORGIA POWER COMPANY LETTER LCV-0626, l
" RESPONSE TO NRC GENERIC LETTER 95-03" i
(continued) enhanced steam generator tubing inspection program for the detection of any circumferential cracking, should one or more tubes have a definite circumferential " crack",
the affected tube (s) would be plugged using the current tube plugging criteria. The inspection scope would then be expanded, based on a pre-approved plan that will encompass available EPRI criteria and will take into consideration stmetural integrity _
evaluations and the nature and number of flaws discovered. The inspections and any scope expansion thereof will be performed using an EPRI NP-6201 (Revision 3 or subsequent revision)-qualified eddy current probe that enhances inspections for circumferential cracking. Personnel training and qualification for these inspections will continue to be performed in accordance with Appendices G and H of EPRI document NP-6201.
l It shall be noted that for those maintenance / refueling outages scheduled subsequcnt to those for 1996, we have not yet formalized our future inspection plans for the detection of circumferential cracking of steam generator tubes. GPC will evaluate any guidance provided by EPRI specifically for the inspection of thermally-treated, Inconel-600 tube material for signs of circumferential cracking and results from the upcoming VEGP-1 and 2 inspections in 1996 for use in the development of a future inspection sampling program.
1 El-13 t
1 ENCLOSURE 2 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" Vogtle Electric Generating Plant, Units 1 and 2 NRC Docket Nos. 50-424,50-425 The following information is a summary of bobbin-coil cddy current inspections conducted at VEGP-1 and 2 during the two most recent maintenance / refueling outages, VEGP-1 Outages 1R4 and 1R5 and VEGP-2 Outages 2R3 and 2R4.
VEGP-1 OUTAGE 1R4 La_spection Plan Summary Exignt ofInsnection Steam Generator (S/G) No.
S/G No. 3 2
Full Length of Tube 2,951 tuber 2,934 tubes Straight Length of Tube 1 tube _.___
None The above actually meet or exceed the original test program for VEGP-1 Outage IR4. In some cases, either the required minimum extent was exceeded or additional tubes were tested.
lntdications Observed r
Indications S/G No 2 S/G No 3 m
m 40% and greater wall None 1 tube penetration 20-39% wall thickness 14 tubes 27 tubes penetration Less than 20% wall 22 tubes 25 tubes thickness penetration The above table is based on the largest indication per tube.
Corrective Actions i
There was one tube which was observed to have a pluggable indication during VEGP-1 E2-1
.3
.c.
ENCLOSURE 2 TO-l
. GEORGIA POWER COMPANY.
LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued) p VEGP-1 OUTAGE 1R4 (continued)
~!
1 1.
Corrective Actions (continued)
Outage IR4 and was mechanically plugged. In addition, three tubes were mechanically plugged administratively during the outage to prevent any future possible problems.
VEGP-1 OUTAGE 1R5 Lnsp_cction Plan Summary Extent of S/G No. I S/G No. 2 S/G No. 3 S/G No. 4 InJrection Full Length of 4,211 tubes 4,208 tubes 4,208 tubes 4,209 tubes Tube Straight Length None None
~
None N6he ofTube l
U-Bends of None -
None None None Tubes The above e.ctual extents met or exceeded the original test program for VEGP-1 Outage IRS. In some cases, either the minimum extent was met or exceeded by testing additional tubes.
E2-2
1-ENCLOSURE 2 TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued)
VEGP-1 OUTAGE 1RS (continued)
Lndications Observed Indications S/G No.1 S/G No.2 S/G NoJ S/G. Nod 40% and None 2 tubes 1 tube 2 tubes greater wall thickness 20-39% wall lo tubes 13 tubes 25 tubes 9 tubes thickness penetration Less than 20%
21 tubes 39 tubes 25 tubes 22 tubes wall thickness penetration The above table is based on the largest indication observed per tube.
Corrective Actions Five tubes were observed to have pluggable indications during VEGP-1 Outage IRS and were mechanically plugged. In addition, seven tubes were mechanically plugged administratively during the outage to prevent any future possible problems.
VEGP-2 OUTAGE 2R3 inspection Plan Summary Extent _ofInspection S/G No 1 S/G No. 4 m
Full Length of Tube 2,984 tubes 3,008 tubes Straight Length of Tube None None The above actuals met or exceeded the origmal test program for VEGP-2 Outage 2R3. In some cases, either the required minimum extent was exceeded or additional tubes were tested. In addition, one tube on S/G No. 4 was tested using RPC to confum an indication.
E2-3 l
l I
i.
I v.-
1 ENCLOSURE 2 -
TO GEORGIA POWER COMPANY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued)
VEGP-2 OUTAGE 2R3 (continued)
Indications Observed Inslications S/G_NoJ S/G._ Nod 40% and greater wall None None thickness penetration 4
20-39% wall thickness 7 tubes 12 tubes jenetration Less than 20% wall 14 tubes 26 tubes thickness penetration The above table is based on the largest indication per tube.
Corrective Actions There were no pluggable indications identified during VEGP-2 Outage 2R3.
l VEGP-2 OUTAGE 2R4 Inspection Plan Summary Extent ofInspection S/G No 2 S/G No_J Full Length of Tube 4,382 tubes 4,576 tubes Straight Length of Tube None None The above actuals met or exceeded the original test program for VEGP-2 Outage 2R4, In some cases, either the required minin um extent was exceeded or additional tubes were tested.
I E2-4
,.1 z ENCLOSURE 2 TO GEORGIA POWER COMP /d TY LETTER LCV-0626,
" RESPONSE TO NRC GENERIC LETTER 95-03" g
(continued)
VEGP-2 OUTAGE 2R4 (continued) p L
Indicctions Observed Indication 3 S/G_No 2 S/G_NoJ n
m 40% and greater wall 2 tubes None thickness penetration
(
20-39% wall thickness 5 tubes 1 tube L
penetration Less than 20% wall 19 tubes 18 tubes thickness penetration The above table is based on the largest indication per tube.
Corrective Actions l
Two tubes were eserved to have pluggable indications during VEGP-2 Outage 2R4 and-were mechanically plugged, in addition, one tube was mechanically plugged administratively to prevent any future possible problems.
For any additional details on the past two steam generator tubing inspections performed at l
VEGP-1 and 2, please refer to the follow'mg correspondence which has been submitted to the NRC:
l Unit /Outagg GPC Letter No /Date to the NRC VEGP-1 Outage IR4 MSV-01465, July 21,1993 VEGP-1 Outage IR5 LCV-0524, January 10,.5995 VEGP-2 Outage 2R3 LCV-0227, Januanf 11,1994 VEGP-2 Outage 2R4 LCV-0549, June 26,1995 As noted in Enclosure 1 to this letter, to date,27 tubes have been plugged in the four steam generators at VEGP-1 and 18 tubes have been plugged at VEGP-2. The number of tubes plugged on each VEGP unit included tubes plugged preservice, i.e., plugged at the manufacturing facility, and since startup of each VEGP unit. A sumn_ary of plugged tubes E2-5
_-___.__.____--_-_m.
i ENCLOSURE 2 TO GEORGIA POWER COMPANY LETTER LCV-0626, r -
" RESPONSE TO NRC GENERIC LETTER 95-03" (continued) follows and also includes those tubes which were plugged administratively:
S/G Nc.1 S/G No. 2 S/G No. 3 S/G No. 4 Total Tubes -
Plunged VEGP-1 1 tube 8 tubes 11 tubes 7 tubes 27 tubes.
VEGP-2 2 tubes 7 tubes Itube 8 tubes 18 tubes I.
?
\\
I
- b 1
E2-6 I