ML19270E988
| ML19270E988 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 12/29/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19270E985 | List: |
| References | |
| NUDOCS 7901100451 | |
| Download: ML19270E988 (6) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 46 TO LICENSE NO. DPR-32 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NO. 1 DOCKET NO. 50-280 INTRODUCTION By letter dated December 12, 1978, as supplemented December 20, and 29,1978, Virginia Electric and Power Company (VEPCO) submitted the inspection program and the results of the steam generator tube inspection perfomed at Surry Unit No. I during Decembe, 1978, including the plugging criteria implemented for the three steam generators.
Based on these inspection results, the implemen-ted plugging patterns and previously submitted ECCS analysis, VEPC0 concluded that the f acility can be returned to operation for another six (6) equivalent months. VEPC0 had originally requested nine months but, following discussions with NRC staff the licensee's final submittal, dated Decemt;er 29, 1978, concluded that six months of operatien was justified.
Surry Unit No. I has been operating under a June 23, 1978, NRC Order for Modification of Facility Operation License No. DRP-32.
That Order required VEPCO to inspect the steam generators on or before the Order's expiration date and to obtain NRC approval prior to resuming power operation. Another condition of that Order stipulated a maximum pemissable reactor coolant system to secondary system leakage rate of 0.3 gpm through the steam generator tubes. Due to the development of a leakge rate exceeding the leakage limit, VEPC0 was required to shutdown the unit approximately three weeks prior to the end of the authorized period of operation.
DISCUSSION Insoection Procram The steam generator tube inspection perfomed during this shutdown included programs to assess the conditions associated with both the denting and " wastage" problems. For denting, tube gauging was done in all three steam generators in order to assess the extent and pattern 7 9 0110 0 VS/
. of tube denting. On the hot leg side, all tubes near the tube lane which are predicted to be bounded by the 15% hoop strain contour were gauged. Based on previous leaker history at Surry Unit No. I and at similar units, as well as previous gauging results, the gauging program also included wedge, patch plate, and other regions in the tube bundle. Additionally, when a restricted tube was found close to the inspection boundary, the inspection was expanded in that area. Gauging was also performed on cold leg tubes in all three steam generators in conjunction with the U-bend inspection program conducted from the cold leg side.
An inspection for wastage was perfomed in order to identify tubes with significant or suspected thir.ning. This inspection was prompted by current inspection results, discussed later, and was primarily based on experience at Surry Unit No.1.
A large number of tubes in the middle of the hotleg side of the tube bundle
(" kidney" region) were inspected for all three steam generators.
Handhole inspections of the visible tube support plates using photographs were perfomed in all three steam generators in order to assess the support plate conditions. Also, wrapper to shell annulus measurements were made in steam generator C to monitor the support plate growth due to continuing formation of magnetite at tube / tube support plate crevices.
Results of Inspection and Corrective Action The primary to secondary leakage rate, which made it necessary to shut-down the facility, was approximately 0.31 gpm. Following shutdown, the secondary side of each steam generator was pressurized and the tubesheet inspected for evidence of leakage. The results of this hydrotest inspection are discussed below.
In steam generator A, eight plugs were wet indicating potential leakage.
However, there were no indications of leaking tubes. The eight leaky plugs were weld repaired.
Indicated leakage through these plugs was very small and, since the pressure differential during nomal operation would tend to seat the plugs, VEPCO has concluded that no significant leakage occurred through these plugs during operation.
- There were no indications of any tube or plug leakage in steam generator B.
Five tubes in steain generator C indicated obvious and possible leakages. The tube at location RSC69 released a continuous stream of water and has been identified as the principal contributor to the leakage which required the facility shutdown. The tube at location R3C76 was wet, but not believed to have contributed significantly to the leak rate prior to shutdown. A slow drip was indicated from the tube at location R17C8 and is not believed to have been a significant contributor to the overall leakage rate. Tubes at locations R25C64 and R29C61 both had a slow drip and also are not believed to have contributed significantly to the cumulative leakage rate. Inspection of these two tubes with a 0.720" probe show 00 wastage immediately above the tubesheet of approximately 801. The tube at location R25C64 was inspected in May,1978, and was found to have approximately 30% OD wastage.
VEPC0 states that accurate detennination of the extent of wastage was difficult due to the presence of masking signals which have been attributed to the presence of minor denting. These five tubes, discussed above, have all been plugged including all immediately surrounding tubes, if not previously plugged.
Gauging results indicated that no tube in any of the three steam generators restricted passage of the 0.540" probe. Activity was noted in wedge and/or patch plate areas in all steam generators.
The growth of magnetite and tube denting in these regions appears consistent with previous experience at other units. On the coldleg side, all tubes met the gauging criteria with a 0.610" probe and all tubes, with one exception, were satisfactorily U-bend inspected. One tube in steam generator A, location R4Cl2, would not allow passage of the 0.610" probe over the U-bend.
This tube was plugged, although it did allcw passage of a 0.510" probe. Dented tubes were plugged in accordance with the plugging criteria discussed later.
Results of the wastage inspection indicate that 16 of the 918 inspected tubes, besides the two that showed some leakage during the hydrotest, were judged as near or exceeding the 40% plugging limit, and in each case these were tubes which were close to the limit, i.e.,35-397, wastage, at the previous inspection. However, VEPCO states that m,iny of the eddy current signals were distorted due to the presence of masking signals attributed to the presence of minor denting. '4here the signals could not be interpreted with I
. reasonable co.1fidence, the tubes were plugged in accordance with the plugging criteria discussed later. The majority,of tubes plugged were tubes indicating minor wastage and masked eddy current signals.
The handhole inspections revealed that the conditions of the visible support plates had not changed since previous inspections and no new phenomena were observed.
The implementation of the plugging criteria discussed below combined with previous plugging for various causes, resulted in a total of approximately 24.3% of the tubes being plugged.
Plugging Criteria The plugging criteria implemented by the licensee are essentially the same as those used at other units with similarly degraded steam generator conditions. As in the previously accepted plugging criteria; e.g., those discused in the SER attached to the Order of December 3, 1977, VEPCO has performed preventive plugging in order to preclude the development of tube leaks resulting from the progression of denting. Additionally, those tubes inspected for wastage in the
" kidney" regions that exhibited either significant wastage or some wastage in proximity to a dent of sufficient size to mask the defect were plugged.
EVALUATION Surry Unit No.1 is on' Of the six lead PWR facilities that have suffered moderate to extensive tube denting and that have been under close monitoring by the NRC staff following the September 15, 1976, tube failure at Surry Unit No. 2.
The tube inspection program just completed is the fifth such program for tnis unit. A discussion on the technical background and safety evaluation of the denting related phenomenon was presented in an SER dated February 11, 1977, in support of the NRC Order for Modification of Facility Operating License No. DPR-32 dated February 8,1977, and in an SER attached to a later NRC Order affecting Surry Unit do. I dated May 6,1977. The background information contained in the February 11, 1977 and May 6,1977, SERs remain valid and is incorporated in this Safety Evaluation by reference. The information discussed above represents an update on the condition of steam generators at Surry Unit No.1 The eight wet plugs in steam generator A were all weld repaired and are of no safety significance during normal operation or under postulatec accident conditions. The three plugs in steam generator
. C at locations RSC69, R3C76, and R17C8 are located in the area of the generator where denting is anticipated and has been observed in the past. This area is well within the gauging boundary.
All three tubes have been plugged. The two tubes at locations R25C64 and R29C61 in steam generator C are in the region where wastage has historically been observed. These tubes have also been plugged.
The staff would have some concern if a tube were to increase from 30% 00 wastage to thru-wall in less than six months as would appear to be the case at tube location R25C64 in generator C.
- However, we concur with VEPCO's statement that the presence of denting can mask an eddy current signal. The signal can also be distorted by the presence of sludge, the tubesheet, or a support plate. The inspection and preventive plugging implemented by VEPC0 to address potential wastage problems, discussed earlier and evaluated below, provides reasonable assurance that unacceptable rates of tube thinning will not occur during the next operating intarval.
Although an inspection for Mastage was not required, VEPC0 perfomed a fairly comprehensive inspection in the kidney regions and particularly around the two leaking tubes found in generator C.
We believe that this inspection, although not required, was certainly desirable. An inspection in accordance with Regulatory Guide 1.83 was done in May of 1978. Therefore, we consider the wastage inspection just perfomed adequate in establishing the current extent and magnitude of wastage in the Surry Unit i steam generators.
The steam generator tube gauging was perfomed in accordance with a program that is consistent with previously implemented programs at Surry Unit No.1 and other units. Therefore, we consider this inspection is adequate in the establishment of the condition of steam generators at Surry Unit No. I and provides a satisfactory assessment of the extent and magnitude of denting.
The gauging program performed at Surry Unit 1 was essentially the same as the programs perfomed at Surry Unit 2 and Turkey Point Units 3 and 4.
As in the gauging program perfomed during May,1978, the 15% tube hoop strain contour was used to define the gauging boundary.
These gauging programs have been developed over the course of time in consultation with the NRC staff and have been determined to be acceptable. The inspection of the Unit No. I steam generators has demonstrated that the tube degradation which has occurred to cate follows the pattern experienced at Surry Unit No. 2 and Turkey Point Unit Nos. 3 and 4.
Furthemore, the results of this inspection at Surry Unit No.1 indicates that no unexpected degradation is occurring and no new phenomenon was uncovered.
. The preventive plugging implemented by VEPC0 to alleviate concern over the potential wastage related problems is at least as conservative as that found acceptable by Regulatory Guide 1.121.
The preventive plugging pattern for denting bounds those tubes which may be anticipated to attain the level of strain which could lead to stress corrosion cracking during the next period of operation.
The preventive plugging conduct 2d by the licensee during the current outage justities operation of the Surry Unit No. I steam generators for an additional six (6) equivalent months.
Environmental Consideration We have determined that this amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made-this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR 151.5(d)(4) that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
CONCLUSION We have concluded based on the considerations discussed above, that (1) Surry Unit No.1 may be operated for an additional six (6) equivalent
- months under the restrictions delineated in the Amendment to which this SER is attached. At the end of this period the facility is to be shut down, the steam generators are to be reprobed to determine the extent and pattern of additional tube denting and the results of this gauging program are to be submitted to the NRC staff for review and, evaluation prior to the resumption of power operation, (2) because the attached Amendment does not involved a significant increase in the probability or consequences of accidents previously c:,nsidered and does not involve a significant decrease in a safety margin, the Amendment does not involve a significant hazards consideration, (3) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed mannsr, and (4) sucn activities will be corducted in compliance with the Commission's regulations and the issuance of this Amendment will not be inimical to the common defense and sacurity or to the health and safety of the public.
- For purposes of this SER, equivalent operation means operation of the facility with a primary coolant temperature exceeding 350*F Date: December 29, 1978