ML20090J521
| ML20090J521 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/22/1985 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML082380335 | List:
|
| References | |
| FOIA-91-106 NUDOCS 8510250552 | |
| Download: ML20090J521 (12) | |
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l ENCLOSURE SAFETY EVALUATION REPORT REVIEW 0F SUPPLEMENT 1 TO WCAP-10698, EVALUATION OF 0FFSITE RADIATION DOSES FOR A STEAM GENERATOR TUBE RUPTURE ACCIDENT I
INTRODUCTION In a May 24, 1985 letter to the NRC, the Steam Generator Tube Rupture (SGTR)
SubgroupoftheWestinghouseOwnersGroup(WOG)submittedSupplementIto WCAP-10698, Evaluation of Offsite Radiation Doses for an SGTR Accident, to support the resolution of the licensing issues associated with an SGTR accident. Th_is Safety Evaluation Report documents the s' ff review of the results and methodology presented in Supplment I to WCAP-10598.
As a result of the January 1982 SGTR at the R. E. Ginna Plant, the NRC has questicned the assumptions used in the safety analysis of a design basis SGTR, including the operator action time assured in terminating leakage from the primary to the secondary coolant systems, and the qualification of the equipment assumed to be used in the SGTR recovery, in response to these concerns, a subgroup of utilities in the WOG was formed to address the licene g issues associated with an SGTR event on a generic basis.
In December of 1984, the subgroup submitted WCAP-10698, SGTR Analysis Methodology To Determine the Margin to Steam Generatnr Overfill, which presented the N
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developt.ent of a design basis TGTR analysis methodology.
Supplement 1 to WCAP-10698 presents the evaluation of potential of fsite doses for a design basis SGTR in the absence of steam generator overfill. The subgroup also plans to submit by November of 1985 an evaluation of the consequences of steam generator overfill resulting from an SGTR.
WCAP-10698 presented results from the following tasks in the development nf a design basis SGTR analysis rnethodology:
(1) development of LOFTTR1, an analytical model which is a modified version of LOFTRAN, that incorporates improved models for break flow and the steam generator secondary side, and an improved capability to simulate the operator actions for SGTR recovery; (2) detennination of operator action times for sign basis application based on the guidelines of Revision 1 of the WOG Emergency Response Guidelines issued in September 1983; (3) sensitivity studies to identify conservative values of plant p.arameters; (4) single failu ~ analysis of the design basis equipment; and (5) application of the methodology to a reference plant.
The evaluation of offsite doses presented in Supplement I to WCAP-10698 used steam release rates to the environment and thermal and hydraulic parameters for the primary and secondary sides which were calculated using the LOFTTR1 computer code.,and operator action times developed in WCAP-10698.
In addition, the single failure analysis and sensitivity studies of Supplement I relied heavily upon the corresponding results of WCAP-10698, it should also be noted that staff review of the subgroup's evaluation of the consequencec of steam generator overfill could potentially lead to changes in tne analysis bssumptions used in evaluating the radiological consequences of a design basis SSTR accident. Thus, the results and conclusions of this SER will be modified i
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as appropriate if staff review identifies the need for significant changes in the~ design basis SGTR analysis methodology presented in WCAP-10698. WCAP-10698 is' currently under review by the staff with SER issuance for WCAP-10698 and for the_ evaluation of overfill consequences projected for the second quarter of
- FY1986, it should be noted, however, that the review of the dose analysis methodology (Section 5,0) presented in Supplement I with its assumptions and models of coolant activity levels and iodine transport processes is not dependent upon the results of the review of these other submittals.
Supplement 1 to WCAP-10693 presents the results of the following tasks:
selection of a reference plant and site; single failure an61ysis to detemine the worst single failure with respect to offsite doses; calculation of the mass releases to the enviroment using the results of the L0fTTR1 analyses fren WCAP-10698 for mass releases prior to temination of the primary to secondary
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leakage, and the results of an analysis based on a continuation of the SGTR recovery aclions in the WOG Emergency Response Guidelines for mass releases during the period between leakage termination and the end of the accident; and the development of the dose analysis methodology.
DISCUSSION The evaluation of offsite doses in Supplement 1 to WCAP-10698, was performed for a reference plant ant, site. Atmospheric dispersion factors which were representative for typical Westinghouse plants were used in the dose calculations. The reference plant, as described in Section 4.1 of WCAP-10698, was selected on the basis of a preliminary analysis which provided estimates of the relative time to overfill for several representative Westinghouse plant
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types. The calculations to determine the relative time to overfill compared the secondary side steam volume to the equilibrium break flow rate, defined as the break flow rate at the primary pressure at whi:h o.tgoing break flow is balanced by incoming safety injection flow. The calculations did not consider accident system response and operator actions.
The staff notes that the selection of a reference plant based on the above estimates of the relative time to overfill does not assure the selection of the most conservative plant design with respect to potential offsite doses.
Operator action time and system response tine, whith depend on plant specific equipment, operating proceduras and individual plant design and parameters, must be considered in determining the duration and severity of the accident and the amount of radioactivity released to the atmosphere.
The evaluation presented in Supplement I to WCAP-10698 was based on a reference plant with representative atmospheric dispersion factors, instead of a conservative plant design with,b,ounding atmospheric oispersion factors. Th' staff concludes that the offsite dose calculations presented in Supplement I constitute representative examples of the application of the proposed design basis SGTR analysis methodology to a' reference plant and site, but are not bounding cases, Plant specific analyses will be necessary to demonstrate that the radiological consega nces of_a postulated SGTR accident at an individual plant meet the acceptanca criteria of Section 15.6.3 of the Standard Review Plan (NUREG-0800, Rev. 2 July 1981).
The single failure analysis to determine the worst single _ failure with respect to offsite doses and sensitivity studies to identify conservative (with respect to offsite doses) plant conditions, parameters, and other analysis assumptions l
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S presented in Supplement I relied heavily upon the results of the single failure analysis and sensitivity studies in WCAP-10698 which were used to identify conservative assumptions with respect to margin to overfill.
(The margin to overfill is defined as the steam space volume remaining below the steam generator outlet no::le when the primary to secondary leakage is terminated).
As stated in Supplement 1, it is expected that most of the conservative assumption > and initial conditions which were used in the evaluation of the margin to overfill would also be conservative with respect to offsite doses.
This is based on the fact that both offsite doses and the potential for overfill are primarily dependent upon the amount of primary to secondary leakage and the amount of steam released from the ruptured steam generator.
The staff agrees that, in general, conditions and assumptions which are conservative with respect to overfill would also be conservative for offsite doses. The decrease in the margin to overfill as a result of a postulated single failure or a conservative analysis assumption is due to the increased operator ac'. ion time and system response time required to complete the recovery action. The increased operator action time and system response time would prolong the accident and generally lead to increases in the release of radioactivity to the environment.
As discussed in Supplement 1, however, a decrease in the margin to overfill represents the additional net accumulation of water in the secondary side of the ruptured steam generator. Net accumulation of water increases with increases in the amount of primary to srcondary leakage, but decreases with increases in the amount of steam released from the ruptured steam generator.
(This follows from mass continuity considerations if one neglects 4
6 interdependency ef fects.) For those cases in which the amount of steam released to the atmosphere does not change, conservative conditions with respect to overfill would also be conservative with respect to offsite doses, in these cases the decrease in the margin to overfill is a result of an increase in the amount of primary to secondary leakage due to increased operator action time and system response time.
This prolongs the accident and results in increas i releases of radioactivity to the environment.
The single failure analysis presented in Supplecent I has identified and examined those cases which result in ircreases in the amount of steam released from the ruptured steam generator.
Ir, addition, the analysis identified an estimated proprietary hydraulic parameter which was conservative with respect to offsite doses, but was not conservative with respect to margin to overfill.
This assumption is discussed in Section 5.2 of Supplement I and was investigated I
in various case comparisons, including a comparison of calculated doses for Cases 1 and 5.
l Based on the above findings, the staff concludes that the single failure analysis and sensitivity studies in Supplement I have ideritified the worst single failure and the analysis assumptions which are conservative with respect to offsite doses. This conclusic,n is based upon the following:
staff review of the sensitivity studies and equipment failure evaluation in WCAP 10690 to assure that conservative plant conditions, parameters, and analysis assumptions and the worst single failure with respect to margin to overfill have been properly identified; the generic applicability of the single failure analysis in WCAD-10698; and the use of the assumption which was identified in f.ection 5.2 of Supplement 1 to be conservative with respect to offsite doses but not i
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7 with respect to margin to overfill in subsequent applications of this methodology for the evaluation of offsite doses from an SGTR accident.
The results and conclusions of this SER will be modified as appropriate if the review of WCAP-10693 identifies the need for significant changes in the results of the sensitivity studies and equipment failure evaluation presented in WCAP-10698.
In addition, the single failure analysis presented in WCAP-10698 is based on the WOG Emergency Response Guidelines which are applicable to nearly all Westinghcuse plants, and a design basis equipment list which identifies sufficient principal equipment to teminate primary to secondary leakage for all Westinghouse plants.
The generic applicability of the analysis may be limited, however, based on plant specific dif ferences which would affect changes in operator action times and system response times required to complete the recovery operation as a result of a postulated single o
failure.
For example, the staf f notes that the resuits of the sir.gle failure analysisHn WCAP-10698 may not apply ts two loop Westinghouse plants.
If, is a result of the staff review of WCAP-10698, it is detemined that the single failure analysis is not generically applicable, then plant specific analysis to detemine the worst single failure with respect to of fsite doses m'iy be required.
The staff has re.iewed the evaluation of offsite doses for the single failure cases considered in Supplerent I to WCAP-10698. Mass releases from the ruptured and intact steam generators to the atmosphere were determired from LOFTTRI analyses (described in WCAP-10698) for the period from accident initiation to the temination of primary to secondary leakage.
Mass releases for the period from leakage temination to the end of the accident, assumed to be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, J
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8 were determined from an analysis based on SGTR recovery operations in the WOG Emergency Response Guidelines. Revision 1 of the Emergency Response Guidelines provides for three alternate means of perforning the post - SGTR cooldown.
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method using steam dum,, Guideline ES-3.3, was selected for evaluation of the mass releases since it results in conservative results for the offsite dose evaluation. The ES-3.3 guideline specifies the actions required to bring the Reactor Coolant System down to Residual Heat Removal System ten.perature and pressure levels. This is accomplished by using steam dump to the condenser, or using tha power operated relief values of the intact and ruptured steam generatnrs if the condenser is unavailable.
The dose analysis rethodology as presented in Supplement I to WCAp-10598 uses assumptions for the initial primary and secondary coolant activity concentrations, the radiological consequences of iodine spiking, a ecolant todine spiking model for the accident initiated iodine spike case, and primary to secondary-system leakage in the intact steam generators which are consistent with those in Section 15.6.3 of the Standard Review Plan, in the determination of ictine transport to the atmosphere, the methodology presented in Supplement 1 discusses the volatilization of iodine in the primary coolant due to flashing and atomization, and the scrubbing of iodine contained in the steam phise and atomized droplets for release points which are below the steam generator water level.
It does not, however, explicitly describe the models and assumptions used in the detennination of iodine transport in the faulted generator.
Thus, no staff review of the iodine transport models was possible, and independent staff verification using the iodine transport models referenced in the Standard Review Plan will be necessary on a case-by-case
- basis, it is the staff's position that plant specific analyses should u.
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4 provide a detailed description of, or reference, the explicit iodine transport models used in the analyses.
l The staff concludes that the dose analysis methodology presented in Supplement 1 to WCAP-10698 is generally consistent with Section 15.6.3 of the SRP and, thus, is acceptable with the exception of the iodine transport models which will be reviewed on a case-by-case basis, i
CONCLUSIONS The staff has reviewed the methodology and results presentad in the evaluation of offsite doses for an SGTR accident in Supplerent 1 to WCAP-10598. The staff concludes that the dose analysis methodology used in the evaluation is acceptable with the exception of the determination of iodine transport to the atmosphere for which explicit rodels and assumptions were not provided.
independent staff verification using the iodiae transport models referenced in the SRP will be r, aces.ary on a case by-case basis.
The staff notes that the offsite dose calculations presented in Supplement I were based on a reference olant and reference site and, thus, did not constitute bounding cases for all reactors and sites.
Plant specific analyses will be necessary to demonstrate that the radiological consequences of a postulated SGTR accident at an individual plant meet the acceptance criteria of Section 15.6.3 of the SRP.
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6 10 The results and conclusions of this SER will be modified as appropriate
.f staff review of WCAP-10698 and of the subgroup's evaluation of the consequences of steam generator overfill identifies th: need for significant changes in the design basis SGTR analysis r.cthodology presented in WCAP-10698.
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IMPLEMENTATION l
t As discussed above, plant specific evaluations of offsite doses using l
i appropriate plant specific mass releases and thermal and hydraulic parameters for the primary and secondary systems will be necessary for individual plants. The evaluation should consider the worst single failure and plant conditions, parametcrs, and assumptions which are conservative with respect to offsite doses. The results of the single failure analysis and sensitivity studies in Supplement 1 are acceptable, provided the single failure analysis in WCAP-10698 is generically applicable _ and the staff review of WCAP-10698 does not identify the need for significant changes.
If, as a result of the staff review of WCAP-10698, it is determined that the single failure analysis is not generically applicable, then plant specific single failure analyses to cetermine the worst single failure with respect to offsite doses may be required.
In addition, the plant specific evaluations cf offsite doset should use the ar;,ysis Assumption which was identified in Section 6.2 i
of Supplement 1 to be conservative with respect to offsite doses but
11 which was not conservative with respect to margin to overfill.
The plant _ specific analysis should provide suf ficient information for staff review, including the following information as a function of time during an SGTR -to allow an independent evaluation to be made by the staff of the radiological consequences:
(1) Total mass releases and mass release rates from the ruptured steam generator to the atmosphere.
(2) Total mass releases and mass release rates from the intact steam generator (s) to the atmosphere.
(3) Pr'imary to secondary system leakage flow rate in the faulter) ge_nerator (break flo'w rate),
.(4) Pressure differential between the RCS and the ruptured steam generator, (5) Water level above the break location in the ruptured steam generator.
(6) Mass of water in ruptured steam gen?rator, (7) Pressure in tbc.uptured steam generator, and d
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(8) RCS hot leg and cold leg temperatures in the ruptured
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-In addition, it is the staff's position that plant specific analyses should include i detailed description of, or reference, the explicit iodine transport models used in the analyses,
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UNITED sT ATts 7-E NUCLEAR REGULATORY COMMISSION
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Q WASHINGTON. n r %6 AUG 11 W67
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MEMORANDUM FOR:
C, Y. Cheng. Acting Chief-Materials Engineering Branch Division of Engineering and Systems Technology i
THRU:
Keith Wichman,-Section Leader Materials Engineering Branch Division of Engineering and Systems Technology FROM:
Herbert F. Conrad Materials Enoineering Branch Division of fngineering and Systems Technology
SUBJECT:
TRIP REPORT - NORTH ANNA 1 STEAM GENERATOR TUBE RUPTURE INVESTIGATION, JULY 22, 23, 29 AND 30, 1987-Suma ry e
The 360' circumferential double ended break at the top of the uppermost cold leg tube support' plate is now believed by the licensee to be related to both stress corrosion cracking and fatigue with the origin (ID or OD) not yet known.
The fracture location was last inspected in 1981; no crack indications were found at that time. The Utility has comitted to a comprehensive full
_ tube length, all steam generators inspection t at will be the most extensive h
~ and sensitive eddy current inspection program conducted on a U. S. Nuclear Plant to date. Every effort will be made to remove a sample of the fractured tube, bUt its location within the bundle at the top near the U-bend (row 9, column 51) makes removal and stsbilization of the remaining tube end difficult.
The plant was shut down in:an orderly manner after the rupture with all safety limits and. thermal margins maintained. The licensee's analysis indicates that the event was bounded by.the steam generator tube rupture event calculations in the Plant Final-Safety Analysis Report. Radioactive releases via the condenser air ejector were less than 1% of the Technical Specification Limit
_and well within 10 CFR 100 limits.
Introduction I traveled to the North Anna Nuclear Power Plant on July 22, 1987 and joined with Dr. C. V. Dodd, Oak Ridge National Laboratory, to participate on the Augmented Inspection Team (AIT) which was-led by Floyd S. Cantrell of Region II.
Dr. 06dd is under a technical assistance contract with the Materials Engineering Branch for on-call consultation in the-area of. eddy currer,t testing. He also does research for the Office of Nuclear Regulatory Research. We participated in the AIT Tem activities on July 22 and 23 and returned to North Anna on July'29 for the meeting-between North Anna management and J. Nelson Grace, Region II Administrator and' members of NRR management. On July 30 we completed our input to the AIT Inspection Report covering eddy current testing.
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Steam Generator Tube Rupture On Wednesday, July 15,1987, 6t approximately 6:30 a.m., Unit 1 of the North Anna Power Staticn experienced a tube rupture in steem generator C, cf tube R9C51 at the top of the seventh support plate in the cold leg. This accident occurred only about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter the reactor returned to 100t power after the Spring 1987 refueling outage. The ex&ct chronology of this tube rupture event is given in Attachment 1.
The operators at the plant were able to bring the reactor to a cold shutdown mode without further damage to the plant or any significant radiation release to the environment due to appropriate execution t
of the operating procedures.. A review of the event, run on the training simulator by the operators for the NRC staff. demonstrated the shutdown process after the tube rupture. In addition to the training simulater, a model power plant in a see-through glass case was shown. A simulated tube rupture showed the loss of coolant in one steam generator and the problems and effects of this on the plant. The model had all the. major components of a nuclear power plant.
including two steam generators, one once-through and one recirculating steam generator. The water level and boiling in.the various. components could Oc seen, t
At the' time of the initial meetings with the power station personnel, the exact nature of the defective tube was not known. Dr. Dodd and I were present in the Westinghouse trailer on Tuesday afternoon (July 21) when the eddy current tapes of the leaking tube were analyzed for the first tirre.
They showed an indication at the top of the seventh tube support plate so large that it saturated out the electrenics. The analysist insisted that it had the signatut'e of a tube end. The utility at that time, however, reported it as a i" to.1" long longitudinal crack even though calculations indicated that such a short crack could not account for the observed leak rate (560-637 gpm).- It was not until the video fiber-optics examination Tuesday night that the tube was confirmed to be a 360' guillotine break with the ends approximately f" to t.
1" apart. Detailed examinction of_the videotape examination of the fiber optics scan of the tube by the VEPC0 Metallurgist is given in attachment. My own observations agree with his.
The full length of the tube was inspected in 1979 and again in 1981 by a bobbin probe. It was inspected during the April 1987 refueling outage, only to the seventh support plate on the hot leg side, not the full length or around to the seventh cold leg support plate. The review of the 1981 inspection tapes revealed nothing. These tapas are analog and the present inspection equipment (MIZ18) can give a far superior inspection. A_n investi-gation of the background of the previous eddy-current inspections will_be
.psrformed.
Backcround of Eddy-Current Tests The generators were modified before operation by explosively expanding the tubes ird the tubesheet region, eliminating the crevice region that had been a l
source of tube leaks at other plants in the 1970s. This, huwever, mnved the expansion Yegion up the tube near the top of'the tubesheet, which has caused some eddy-current inspection problems. The history of eddy-current inspections
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3 and repairs is sumarized in Attachment 3.
The inspections in 1979 with the 1
bobbin type coils revealed that denting had occurred. This denting has considerably complicated the subsequent eddy-current tests of the tubes at tne intersection of the tubes and the tube supports and made the detection and measurement of other modes of degradation much more difficult. Profilometry data performed in subsequent inspection has revealed no growth in the denting, but leaks have revealed the continued degradation of the generator. The early eddy-current inspections were performed with single-frequency eqvipment and recorded on analog tape.
More accurate inspections, performed with three-frequtncy instruments using digital data reduction and analysis techniques, revealed what was referred to as " distorted tube support plate indications." The>e were first observed in 1984, and attempts to resolve these indications led to the use of the 8 x 1 probe and the rotating pancake coil (RPC). described in Attachment 4 Inspections with these probes resolved the distorted tubasheet signals into axial cracks for some of ~ the tubes, with the others found to have nn iiefects.
In addition to these, circumferential defects were located in the tubesheet expansion region. These defects were detected by the 8 x 1 probe and verified and mapped by the RPC. About 150 tube support junctions in steam generators A and B were also inspected with the RPC. These intersettions had not revealed
- any indications with the bobbin coil inspection, and they did not reveal any indications in the RPC inspection. Tube pull data from the 1985 and 1987 outages revealed that there was intergranular cracking (IGC) at the top of the
-first tube support plate, on the outer diameter of the tube, up tu 28%- deep.
In addiMon, there were circumferential cracks on the tube inner diameter at the top of the tubesheet, associated with the explosive expansion dnd oxial cracks at the tube supports, associated with the dents. Tube burst tests on t pulled tube having an 84% defect,150* around the tube, showed that 10,700 psi was required to fail the tube.
Eddv-Current Inspection Plan After the failure of tube R9C51 in a circumferential manner at the top. of-the seventh tube support in the cold leg, an extensive eddy-current testing _
program was planned with emphasis on detecting circumferential defects. This program is lis ted in Attachment 5 and includes the inspsction of every_ tube-support junction (and the straight tube sections in betweent in all three
--steam generators with an 8 x 1 pancake array probe. This is the most 1
- extensive, sene,tive, and ambitious inspection program attempted to date fur steam generator inspect'lon. It will strain the availability of pmbes and data analysts in the industry. This probe (8 X 1) hes the sensittYity 10 detect all inner diameter defects, either axial or circumferential, 20% or deeper,.with a length of 3/16' in, or longer.
In addition, it thould also be able to-detect outer diameter cracks and intergranular attack on either the inner or outer diameter. All indicatior.s detected by the 8 x 1 probe will also l
be' tested using the RPC probe. The tubing standard-used for the pancake coils l
has a range of outer diameter circumferential electrodischarged machined l
notches ranging-from 20 to 100%. The standard scans showed good depth separation between the outer diameter notches of different depths at 400 kHz, i
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and a gnod separation between the tube support signals and defect signals at 200 kHz, although some of the depth measurement ability was lost at-this lower frequency. Although no notch st6ndard was available for inner diameter defects, they could certainly be detected and estimated from an interpretstion between no defect and 100t defect.
Evaluation of Procedures and Analysis We obtained a copy of the North Anna 1
" Analysis Rules-Steam Generator inspection Procedure Package" dated July 1987 and Dr. Dodd, ORNL, the NRC's eddy current consultant reviewed the written procedures as well as observing the actual eddy current data analysis in the Westinghouse Trailer at the North Anna Site. He provided the following evaluation:
"The written data analysis methods are clear and detailed, with more than adequate examples for all three types of eddy current inspections.
The Senior data analysts are very experienced with the facility, the equipment, and the general types of tube degradation that has occurred at all other Westinghouse f acilit.ies end with the methods of detecting tube degradation. The Intelligent Eddy Current Data Analysis System (IEDA) is being used as an aid in flagging suspect bot. bin coil indications which are then dispositioned by the data analyst. The data from each tube is independently reviewed by two different analysts, witn one using the Westinghouse IEDA system and the other using a Zetec. Digital Date Analysis System (DDA4). All the data aralysts are-at least certivled '.evel 11, An'erican Society of Nondestructive Testing (ASNT) in accordance with ASNT requirecents. This includes industry experience, class room training, a technical education, and testing on both general eddy current knowledge and specific eddy current knowledge for steam generator inspection. The analysts are given additional traiair.g by WeGtin0 house and are required to pass a test that covers the specific data analysis used for *he three eddy current tests at North Anna 1."
Current schedules call for return to power on September 30, finish of inspection on September 5 and for the removal of R9-C51 to begin on August 11 by shrinking the tube with a longitudinal wald bead. The Utility plans to itsue daily inspection status reports, the la' *st of which is included as Attacnment 6.
I will keep you informed of all new develognents.
73 4 Herbert F. Conrad Materials Engineering Branch Division of Engineering and Systems Techriology 4
cc: See next page r
AUG.t1 elf
Enclosures:
As stated cc:
R. Starostecki L. Shao
- 1. Varga J. Richardson G. Lainas L. Rubenstein L. Enale K. Wichman E. Murphy L. Frank W. Fazelton H. Conrad DISTRIBU110N:
Docket File EMTB RF EMTB PF DE DES F)fTB d
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08/ e t/87 0,gf/87 0FFICIAL RECORD COPY (5520 document name: Conrad-11)
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jft CHRONOLOGT STEAM i hiERAT0il TJBE RUPTURE EVDfT NoliTH ANNA POVER STATION UNIT I JULY 15, 198f To support the investigation of the Steam Generator Tube Rupture the following chronology was reconstructed f rom the print outs of the alarm typewriter attached to the Control Room P-250 process computer, the Sequence of Events Recorder (Dranetz) driven by the Hathavay annunciator system, the data printoups extracted f rom the record kept by the ERF Computer in the Technical Support Center, RO and SRO logs and interviews, and strip charts f rom Control Room recorders.
Selected data was transmitted f rom the various records based on the significance of each datum as it identified a sub-event or demonstrated.
explicitly or implicitly, a sub-event in the sequence. The intent is that this chronology can be integrated with other analyais to deter-aire the timeliness, accuracy and effectiveness of the measures applied to mitigate the accident.
Once the data was transcribed, a review was performed to identify the synchronism for time of the various data sources. The principal item selected for synchronism was the Automatic Pzt Lo-Lo SI.
The SI action incorporates several actions including feedvater isolation and nottal charging isolation that make it readily comparable over all records. The Sequence of Events
_ Recorder logged SI at 06:35:24:805; the alarm typewriter en the P-250 logged SI at 0639. However the earlier Reactor Manual Trip has caused the P-250 to alter its scan rates. The P-250 Post Trip review logged SI at 06:35:24 plus 1012 cycles, which equates to 06: 35:40.86. The ERFC data set collected at 06:34: 14 records full normal charging flow and full power feed flow to the steam generators, approximately 16 seconds efter the reactor trip had been manually initiated. By 06: 34: 21, the ERFC dan set charging flow is reduced to 82.568 gpm and feed flows are about 600KL3H
- a 800KLBH. By 06:34:34, all flows had reached a stable but luv level, it oppears that SI occurred at or slightly before 06:34:14.
This chronology vill use 06:34:14.
For automatic initiation of Safety Injection, the clock comparisons are as follows:
, RECORDER TIME Sequence of Events Recorder (SER) 06:35:24: 805 P-250 Computer (Alarm Typewriter) 06:35:41 ERF Computer (ERFC) 06:34:14 For Reactor Manual Trip the clock comparisons are as follows:
RECORDER TIME Sequence of Events Recorder (SER) 06:35:04:548 P-250 Computer (Alarm Typewriter) 06:35:24 ERF Computer (ERFC) 06:33:56
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It is concluded that the P-250 led, the SER vas wit.hin 20 secondt. of the P-250 and the ERTC was about one minute behind the P-250.
The chronology that follovs is annotated by clock time based on the P-250.
All the events that occurred within each minute are listed in order of occurrence as could best be dete n.ined.
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C[lRONOLOGY STEAM CfMERATOR TUBE RUPTURE EVITT hORTH ANNA POVfR STATION UNIT 1 JULY 15, 198J, July 14 1987 2230 Air Ejector Radiation Monitor (RM-RMS-121) was declared inoperable due to erratic operation.
July 15! 1987 0630 An alarm was received on the Unit i annunciator panel for Main Steam (High Range) Radiation Monitor.
Vhen checked by backboard operator.
"A" and "B" monitors were in " Alert" and "C" monitor var in "High" alarm.
0631 Unit 1 CR0 observed the pressuriser level decreasing rapidly.
0632 U-2 SRO recalled the Shif t Supervisor and U-l Sao to the Control Room.
U-l CEO took manual control of charging and set TCV-ll22 to f ull open.
Received Pressuriser low pressure alarm at 2135 psig (Alarm Typewriter).
0633 Shift Supervisor entered the Control Room and directed letdovn isolation.
CR0 initiated realignment of charging pump suction to RWST and a 2% per minute turbine ramp down. Third CR0 (Backboards
' Operator) assumed BOP duties on Unit 1 Control Beard to assist Unit 1 CRO.
(Alarm Typewriter: Make-up commenced. VCT lov level alarm 20.31-)
0634 STA arrivsd in the Control Room.
Alarm Typewriter: Pressuriser Pressure 2109 poig.
Superintendent of Operations was notified and directed the Unit to be onnually tripped.
0635 At direct:'on of Shif t Supervisor, U-l CR0 manually tripped the reactor and turbine and in'ltiated EP-0.
CRO observed pressurizer level at approximately 45% and pressuriser pressure at approximately 2100 psig
-at the time of the manual trip.
POST TRIP REVIEV:
Initial Event 06:35:24 Rx Manual trip (2) 0.00 sec.
Turbine Trip and P7 0.16 High Flux Rate Trip 0.33 Rx Manual trip (1) 0.60 Par - Lo Press Trip 2,80 Stm Cen B Lo-Lo Trip 4.2 Stm Cen C Lo-Lo Trip 4.5 PAGE 1
-. ~.
r Stm can A Lo-Lo Trip 4.5 P*r Lo*Lo $1 16.66 i
Manual S1 Train " A" (1) 43.38 Manual SI Train "B" (2) 44.97 seconds Alarm typewriter:
Auxiliary Teed Water Pumps Start. VCT level 22.1%
increasir-g (indicatas that charging pnsp suction shif t to RWST is completed.)
SER: Main Teedvater Pumps Trip (06:35:25) 0636 e Unit 1 CR0 noted pressurizer pressure less than 1700 psig and pressuri:er level less than $1.
Alarm Typewriter Pressuriger level 2.7%.
Main Teed pump Breakers tripped.
"B" Char ging Pump S tar t.
"A" and "B" LHS1 pumps start.
0637 Alara Typewriter: 0-12 breaker open.
A Notification of Unusual Event was declared. Unit 2 SRO assumed 0639 duties as Interim Station Emergency Manager and initiated the EPIP's.
(Step 21 of EP-0) 0640*
Entered EP-3 f rom Step 23 of EP-0.
0641 Alarm Typewriter: T less than 543*F. P-12 interlock set.
gyn
' 0642 Alara Typewriter: "C" Steam Generator level increasing above 18%
narrow range.
0644 Alarm TypevTiter:
Si and Phase A reset. LHS1 pumps "A" and "B" shutdown.
(Steps 9, 10. and 13, respectively of EP-3) 0645 Alarm Typewriter: "C" Steam Generator at 25% narrow range and increasing.
0646*
Auxiliary Feed Vater to "C" Steam Generator isolated.
(Shift Supervisor confitned Steam Generator Tute Rupture in "C" Steam Generator based on "C" Steam Cenerator level continuing to rise.)
0646 ERFC:
"C" Main Steam Trip Valve elesed.
0647 Alaru Typewriter:
"A" Steam Generator at 23% (Narrow Range) and fneressing.
Steam supply f rom "C" Steam Generator to 1-TV-?-2 (Terry Turbine)
'06486 isolated.
(Step 4 of EP-3) 0648 ERFC: Craphs of pressurizer level and RCS pressure reveal increasing level and pressure.
(Also noted on the strip _ chart in the Control Room.)
Pressurizer to Press / Steam Line High Flow SI circuit 0649 Alarm Typewriter blocked. (The Note prior to Step 15 of EP-3) Commenced rapid cooldown on "A" and "B" steam dump valves.
(Step 15 of EP-3) Alarm Typewriter: "B" Steam Cenerator level at 25% (Harrow Range) and increasing. Alarm Typewriter: Pressurizer level 8.5% and decreasing.
P&PP 1 m.-..
a......
,- - -,. ~
_., -. - -.... _... _ -.~.
.-.,,..m-,,...
..,_~,.:.,__--.
e 4
0 1 CR0 noted pressuriser level off scale lov.
0650*
Unit Initial notificatiens made to the State / Local Governmenta (EPIP 2.01) 0651 and NRC (EP!P 2.02).
0652 Alars Typtvtiter:
"A" T* - 509.5'P "B" 7* - 509.5'T "C" T* -
523.5'P classification to Interim Station Emergency Fanager upgraded event 0654
" ALERT".
(Step 40 of EP-3)
"B" Hain Peed Pump breakers racked to test and
, Alare Typewriter:
closed (To provide a flow path for condensate pumps to feed "A" and "B" Steam Generators) Steam Generator "A" and "B" pressures at 589 psig.
Initiated EPIP-3.01. 5.03, and 5.04 (Call Out, Accountability, and 065.4 Access Control).
RCS Temperature being maintained at 480'P (Step 15 of 0657 Strip Ciiart EP-3).
Alarm Typewriter: Pressurizer Spray control at 100% demand.
Valves " A" a_nd "B" open (step 18 of EP-3).
n 0658*
Unit 1 CR0 noted pressuriser level on scale and increasing.
Sourre Range Nuclaar Instruments manus 11y 0659
/larm Typevittert re-energized. (Intermediat t Range Nuclest Instruments were undercompensated.)
Pressuriter icv level hester cut of f cleared (IcVel 0700 Alarn Typewriter t at 15% and increasing). Pressurizer heaters energized (83$ KW).
0701
.Alaru Typewriter: Unit 1 CR0 manually de-energizes pressuriser beaters.
Notifications made to the State / local Governments and NRC of upgraded 0702 alert classification.
Opened one Pressuri:er PORV to reduce pressure 0704 Alaru Typev11 tert (Step 19 of EP-3).
SK0 observed pressure reduction of approximately 18 40 psig and instructed CR0 to close PORV and spray valves (Steps 1
and 19 of EP-3).
Alarm Typewriter:
Pressurizer Relief Tank pressure 15 psig.
SAO noted "C" Steam Cec.erator level increase stopped.
SI redaction criteria met.(Step 21 of EP-3).
"B" Charging Pump secured (Step 22 of EP-3).
Initiated the isolation of B1T flovpatn (Step 24 of EP-3) and establi.shed the normal charging flovpath (Step 25 of EP-3).
0706**
"A" and "B" pressurizer spray valves closed.
Alarm Typewriter: Non Regenerative Heat Exchanger outlet flov 47 gpm.
0709
.(Evaluation of this entry indicates that notaal 1ctdown t.ad been restored in accordance with Step 29 of EP-3.
o cr S s.,
' emmi m
i t
0710 Alarm Typewriter: fressuriter heater t reakets closed (Step )! of EP-3).
0711 Alara Typewritert Presruriser heater breakers closed (Step 31 of j
EP-3).
l 0713 Alarm Typewriter Sscured "C" and "B" RCP's (Step 38 of EP-3).
0714 Alarm Typewriter:
Spray demand 761. (Fica this time forvard RCS pressure is maintained by manus 1 control of spray and heaters.)
i Superintendent of Operation and SRO-On-Call arrived in the Control l
071$
e Room.
0718 Transitioned to ES 3.1 " POST-STEAM CENERATOR TUBE RUPTVRE COOLDOVN USINC BACKTILL" (Step 42 of EP-3).
0720 Station Manager arrived iti the Conttol Room.
0721 Alarm Typewriter ATV Feed Pump 3A to "C" Steam Gefierator secured.
0722 Alarm Typewriter: Pressurteer level 73% and decreasing.
0723 Alarm Typewriter: ATV Feed Pump 3B to "B" Steam Generator secured.
(Subsequently ATV pumps are run interuittently to support Steam Generator feed requirements <)-
072$
Alarm Typewriter:
Start ed "B" Condensat e pump (Both "A" and "B" Condensate pumps now running).
0727**
Began RCS cooldown in accordance with ES 3.1.
I 0730 Assistant Station Manager arrives in Control Room and initiates transition of EPIPs and communications from Control Room to TSC.
0739 Station Manager assumes St ation Emergency Manager position.
0745 Alarm Typr nitert Turbine on the turning gear.
t 0756 Condenser Air Ejector manually diverted to containment.
0757 Technical Support Center activated.
0810 Alarm Typeeriter: Secured "B" condensate pump, 0820 Corporate Emergency Response Center activeted.
i OR4 L Started "B" RHR pump for system warm-up (Step 9 of ES 3.1).
i 0853' Alarm Typewriter: Closed "A" MPP breakers.
(Breakers in test to
[
permit opening of pump discharge valve to use Condensate Pumps for feed to Steam Generators.)
l 1
0857 Alarm Typewriter: Open "B" MFP Breakers (To permit isolation of "B" MTP to stop spraying f ron "B" powp suction relief valve.
.a,
..-.~.-. _.--..-.- -
-.~-..~,.
-- -..~.
.-~n l
2 l
0900*
toose Parts Monitoring System niaru on "C" Steam Centrator.
0915 1.ocal Emergency of f aite Facility activated. Commenced using auxiliary
~
spray to supplement P.CS depressurization.
- 0949 Containment partial pressure exceeded allowable set point due to Air t
Ejector exhaust diversion to containeent, 1040 Pressuriser PORV Key Switches to "AUT0" f or t;DTT proteccion.
(SRO Log) 1108 Entered Mode 4 o
- 11$3 Cycled reactor trip breaker to re-enable auto:tatic Saf ety injection.
Placed "A" and "B" charging pumps and "5" LHSI pump in " Pull-to-Lock" 1200 in accordance with 1-0P-3.3.
1219 Placed RNR System in service to cont Anue kCS cooldown (Step 9 of ES i
3.1).-
1 1221 Secured "A" Reactor Coolant Pump.
(SRO Lop) 1254 Main Steam Systen secured in accordance with 1-OP-28.1.
1312 Restored Air Ejs.ctor exhaust to normal alignment.
1330 Entered Mode $
1335 Station Emergency Manager terminated the emergency.
Notified Nuclear Regulatory Commission State and Local Governments of 1336 termination of emergency status.
1336 Implemented Recovery Organizatien.
d
- Approximate time based on computer or strip chart dara.
hh
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07/.'2/57 P_ hell _MINARY EVALUATION NORTH ANNA UNIT 12E,SfG TOBE FAILURE AT SUPPORT PLATE 7, COLD LEG An evaluation was made of the vidno tapes generated by Vestinghouse, using fiber optics, of the R9 C51 tube failure. The following observstions were made:
r 1.
The tube failed over 360 degrees of circumference and the severed ends displaced in the axial direction approximately }"
The tube f ailed just above support plate #7 on the cold leg side.
2.
As viewed f rom the cold leg side upward at the brook location, an area of 60 degress or less is noted to be angled to the tube O.D.
This may represent a final failure location in tensile overload or cyclic bending.
3.
The f racture surf ace is generally rough and granular in appearance.
t 4.
As viewed from the side, from the tube ID, the fracture edge is friegular and appears to be circumferential in orientation with little or no axial orientation of the elements of the track.
5.
Where several small axial cracks, or t ears, do appear, they seem to be associated with a small thin zone of final rupture. They do not appear to be individual axial cracks.
4.
The f racture surf ace does not appear to show a zone of flat fracture which zeight be associatt.d in an initial fatigue crack.
Although some cyclic bending may have been associated with the final rupture, possible crack initiation no indication of fatigue is obvious as a point.
The rough irregular nature of the edge of the fracture is similar to 7.
edge fessures produced by stress corrosion cracking.
8.
There is no clear indication that the f racture initiated f rom the ID rather than the OD. The outside OD edge of the fracture c utnot be viewed by the fiber optics probe.
9.
There are indications f rom the video tape that what may be a small parallel gone of irregular circumferential cracking is visible in the 90 degree angle tape. This small zone of cracking appearn to be at the level of l
just below the priury f racture which would place it i
the tcip of the No. 7 support plate.
l 10.
The ise of video tapes, without further laboratory work. is not censidered suf ficient to clearly identify the cause and nature of the failure.
1
(
J H. McAsoy l
l l
I i
. - - -.. - _. ~ -
. 6YM b
INSPECTION AMD KEPAlR HIST 0kY
' Unit Start up in 1978 7
\\
f
- 19 79 Refueliny Out age Tubes inspected te ajk an Tuben Plu3gd None S/G k 94
,5 / G A - 4 4 0 S / G B - I T)
None S/G B - 94 i
S/G C 480 2 leaks in S/G C S/G C - 96 Comments: Resin intrusion during cycle. Row I's preventively plugged. 2 other tubes plugged due to denting.
Denting first obsersed, Botic acid treatment initiated. Le akege rat e barely detect able.
- 1982 Re f ueling Outage i
Tubes inspected Mak age T,ubes Plugged S/G A + 107 None None
S/G B - 1165 None None S/G C 243 None None Comine n t s : Partial tobe end repair due to split pin damagt in 5/G's x and C.
- 1934 Forced Outage Tubet. Inspected Le ak age Tubes l_ lugged S/G B
- 579 3 Seeks in S/G B S/G B 4
S/G C - 552 2 leaks in 5/G C S/G C 5
~
Comments: No progression in tube t enting observed. Row I leaking explosive i
plugs repaired.
Partial tube end repair performed.
Distorted indications at support plater, first noticed. Leakage rate 396 GPD.
m
~~
- --- - - __~_ _ _ _ __._.__ _ __
- _. ___.__~.
_____._=_ - -
Page :
i 07/21/87 s
+1984. Refueling Outage Tubes Inspected Leake.g Tubes Plugged
$/G A 100*,available None S/G A 10 S/G B - 100Y available S/G B 1
S/C L 100% available S/G C - 5 P ofilometry in all 3 S/G's.
/
Comes,t s : Partial tube end repair performed. Attempted tube removal in A S/G.
Di.itorted indications observed.
Poreign object located and removed in S/G C.
2 tubes plugged preventively.
Leakage rates 2.3 GFD in A, and 10.6 GPD in C.
9:7.<%dr bu l J (%,.cl, 4
- no
/D w t-y
./
y, jp
&%$ (Am Y-e
/
,Py
- 1985 Outage Tubes Inspected Le ak an Tubes Plugged S/G A - 830 3 leaking tubes S/G A 13 Comments: Distorted indications observed. Leakage rete 213 GPD.
~
- 1985 Refueling Outage Tubes Ie.spected Le a k a r,e Tubes Plugged S/G A - 100% available^
None S/G A - 9 S/G B - 100% available" 2 leaks in B S/G B - 17 S/G C - 100% available 4 leaks in C S/G C - 47 Coment s : Two tubes removed with 4 support plate intersections 30 tubes f rom a
the three steam generators were plugged due to " strong" distorted indications. Sample specialized NDE-applied in S/G C.
Leakage rate 90 GPD.
M M6f, avoid /s = c,V feof p(v f d, M t. o f
,m.,-y
,.., ~,, -. -.. _
m 4
Psge 3 07/21/81 4
o
- 0ther Events During 1986 ti.ru March 1987
~%xtensive examinstion of tubing and mater sais with EPRI and Vestinghouse.
4 Preparation and submission of VCAP to NRC.
-Requested and held meeting with NRO staf f in March,1987.
-Developed eddy current r le base for April 1987 Refueling, e
- 1987 Ref ueling Out age Leskate Tubes Plugge,d Tubes Inspected S/G A 83 S/G A - 100% available#
None
+
S/G B - 100% available 2 tubes in B S/G B - 62 4 tubes in C S/G C - 118 S/G C - 100% available Extensive additional NDE performed included:
Comments:
-Profilometry of more than 100 tubes in each S/G.
-8 X 1 probing of nearly 100% of available tubes.., M # i d %
~
indications and
-Rotating pancake probing of all identified tubesheet a sample of support plate intersections.
-AVB indications first noted, primarily in B S/G.
All indications less than 40% and no tubes plugged.
Tube end repair completed.. U-bend stress relief performed on all available Row 2 tubes in all 3 steam generators.
Support plate stress Two tubes removed from S/G A relief demonstration performed in S/0 containing 2__ tube sh_e e t indications end one support plate intersection.
11.5 GPD in B and 14.6 GPD in C.
u
- v. LJ dn.,i f~ua,& / M b y Leakage rate:
4 %4 ~m.J;a~ n R~.
ynf n
. L.,
J 6: X.
,, a 4>1. g. :.
,u 4,
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I o o'). = a.\\\\ t,A plu3e4 9 6 -4o Y s
)
_-._.._.______.___,..__...._,_....._..____..~.m.._-_____..___
w Page 4 07/21/87 NORTH ANNA UNIT 1 TUBE PLUGGING
SUMMARY
l TOTAL OUTAGE DATE STEAH GENERATOR TUBES l
A B
C l
SEPTEMBER '79 94 94 96 284 i
o JANUARY '84 0
4 5
9 MAY '84 10 1
5 16 AUGUST '85 13 0
0 13 NOVEMBER '85 9
17 47 73 APRIL '87 8.?
62 118 263 i
1 TOTAL 2C9 178 271 658 (6.2%)(5.3%) (8.0%)
(6.5%)
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M1TACdA*F<r'7 3 hc" Revised 07/21/87 4
JTEAM GENERATOR INSPECTION AND HA1NTENANCE "C" STEAM GENERATOR 1987 REFUELING OUTAGF, I
Eddy Current inspection:
- $16 tubes inspected full length (16%)
247 tubes inspected through the hot leg to the #7 support plate, cold leg s ide ( 7. 6',).
- 2472 tubes inspected through the #7 support plate, hot leg side (76.4%).
All available tubes inspected using 8 x 1.
Inspection encorrpassed all available tubes on the hot leg side (tube sheet area).
- RPC inspection performed on 41 tubes at top of tubesheet, hot leg side.
Plugging of 118 tubes due to support plate or tubesheet indications.
One (1) tube out of total plugged due to error. (No indication in tube).
- Profilometry inspection of 121 tubes through the (i7 TSP llot leg.
~
Other Maintenance and Inspection Activities:
- Row 2 U-Bend Stress Relief. (75 returned to service).
- Inspection of J tubes (8 sampled). Also, visual examination of steam
-drum.
- Sludge Lance. Thirty (30) passes removed 1610 pounds of sludge.
- Annulus inspection of steam generator. Both hot and cold leg side.
Tiowslot photography Removed-and re-installed tube lane blocking devices.
p g,, 'i T A g Md.eg
D e<b49 3(CO
-)
'C'
$/G DATA SUKKARY ASOF7/24/(T~~~~
~ ~ ~
The IS sample selected for the
'C' S/G inspection is based on the followingt Satisfy 15 T.S. sample plan Sample shall include:
- 1) all previously identified degraded tubes (degraded defined as any callable indication)
+
2) tubes identified by 3x3 grid for rows 10-46 ant a 3x4 grid for rows 2-9 (tube will be excluded if previously plugged) 3)
the 8 tubts surrcunding the failed tube To date the standard bnbbin coil inspection has been performsd frun the hot leg on a total of 366 tubes (Westinghouse analysis is complete on all 366) out of a total of 374 Tubes in rows 10-46 were inspected from tubesheet to tubesheet.
Rows 2-9 were inspected to the 7th support plate on the cold leg side.
Of the 366 tubas analyzed there have been five distorted indictations (D1's) _ identified and one clear indication.
-The following summariser chese indications and provides a review of the spring refueling outage dats for these tubes.
Row Column Spring Data Ju'_y Data Explanation 16 10 Not identified D1 D1 is located at the 6th support plate on the hot leg.
Indication was missed in spring inspection.
Signal appears the same now as in
~
sprieg.
9 32 Not tested D1 DI indication just above the 7th support plate on the cold leg. This area was not inspected during the spring outage.
31 49 Not identified 7C%
Indication is located approx.
1/2 in, above the tubesheet.
Indication was missed in spring outage.
19 19 No flaw apparent DI Di located just above the 6th support plate on the cold leg. Signal appears to have changed.
34 49 No flaw apparent D1 D1 located just above the 1st support piste on the hot leg.
Signal appears to have changed.
25 58 No flaw apparent D1 D1 located just below the 2nd support plate on the hot leg.
Signal appears to have changed.
P
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.e In addition to the standard bobbin coil inspection, an 8xl inspection has begun on 'C' S/G on the hot leg side to just past the 7th support plate.
The initial 8xl inspection plan consisted of 150 tubes in the columns around column St.
Of the tubes inspected (107). 19 have been analyzed by Vestinghouse.
The results of these analysis show two possible indications. These indications have not been verified with RFC. Neither of these tubes were inspected beyond the hot leg tubesheet region during the spring refueling outage.
The indications are summarized belor Rov Column Indication Location 46 49 3rd and 4th support plate hot leg 46 50 ist support plate hot leg h
l t
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ATTACHMENT 4 i
i PROBE TYPES USED FOR STEAM GENERATOR TESTS DIFFERENTIAL B0BBIN PROBE Coils are coaxial with the tube, about 0.050 in, long and about 0.050 in, apart.
Tney are usually 0.720 in, in diameter and are operated With the MlZld eddy-current in an absolute and dif ferential bridge mode.
instrument, they are driven at four multiplexed f requencies (10, 200. 400, and 600 AHz).
The eddy-current pattern in the tube is also coaxial to the of eddy tube,' and any tube property that interrupts or changes the flow These tube property currents will cause a change in the coil impedance.
variations include tubesheets, tube supports, dents, magnetite on the tube
)
Only or in the crevice, defects in the tube, and intergranular attack.
the axial component of defects will interrupt the circumferential flow of eddy currents produced by the bobbin coil so that circumf erential esfects, with very little axial component, produce very low amplitude signals, l
These signals can be easily lost among signals from other property variations.
8 = 1 PROBE f
This probe consista of eignt independent pancate coils operated in an absolute mode, being driven at 200 and 400 kHz.
These probes are typically 3/16 in. in diameter, a n in. long, and contoured to fit the The eight totis are st ronged in two rings of four curvature of the tube.
and overlapped in a manner such that every point on the tube coi,ls each, least one coil.
Each coil is individually spring loaaed passes under at against the tube to minimize distance between the coil and tube wall, or Ine eddy current flow pattern f rom these coils is circular,
" lift-off."
around tne coil axis, and a cract of any orientation will interrupt the The coil is smaller than the bobbin coil ard main flow of eddy currents.
so a small defect causes a larger change has a more concentrated field, more sensitive to the variations in in signal.
The coil is, however, coil-to conductor spacing or lif t-of f than the larger bobbin coil.
While the spring loading against the tube well helps, irregular and sharp dents lif t-of f signal.
Since information at only two will give a substantia) f requencies are recorded (200 and 400 kHz), this coil type does not have as much data available as the bobbin or rotating pancake coil.
ROTATING PANCAXE C0ll (RPC)
This probe is similar to the individual 8 = 1 coils, but is smaller It has a still smaller focus, which gives better (typically 0.125 in.).
resolution to small defects, sees less of the tube outer diameter artif acts, and is more sensitive to lif t-of f.
The probe head, containing Data the coil, is rotated and the coil is sprung againtet the tube wall.
recorded at three frequencies (at least), and a very fine and The spring are time-consuming scan is made of a " suspected arca" of a tube.that it rides the surfa loading and si'.' of this prob ( are such well, and a thc:e-dimensional plot of tne cata gives a good contour of any defects.
. _. ~. - -... _, _. -.
m. _ _... _ _ _ _ _ _, _ _ _., _.,.
ATTACethE9T S 1600 7/2? TO 0800 7/23 Inspection Plen 1.
Complete is sample on hot leg.
2.
Perform endoscope inspection from hot leg.
3.
Start initial 8 x 1 inspection Hot leg side te U-bend Rows 2-12 Column 48 Entire colunins 49-51 Rows 3-13 Column 52 4
Verify 8 x 1 data with RPC as needed.
BEYOND 0800 7/23 1.
100% 8 x 1 Hot leg through the 7th support plate.
2.
RPC verification of 8 x 1 indications.
3.
Profilometry of verified indications.
4 Retest as required.
5.
Plug as required.
6.
Renove SM-10 fixture.
7.
Set-up in cold leg.
8.
Complete 15 inspection.
Perform standard bobbin on portions not inspected in spring outage 9.
- 10. 100% 8 x 1 inspection cold leg through 7th support plate,
- 11. RPC and profilometry verification ai required.
- 12. Plug as reqaired.
- 13. Remove SM-10 fixture.
\\
a 1
g s
14 Y A c u v % WT
(
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I 4
6 t ?
STEAN CEmt1ATOE INSFhCTION $TATUS Dart os/07/s7__
Standard 3ebbin.
8x1 UC Profilometry T.otal :o Insp./////////////////////////////////////////////////
A 268$
3179 11 TBD i
3 2662 3210 9
T&D C
374(H) 3117 36 30 2390(C)
No.Inerected//////////////////////////////////////////////////
A 2683 Se9(c) 0 n
2667 1018(C) o 370(R) 1366(E) 604(C) 30 No. Analyzed by V//////////////////////////////////////////////
A 7to
.o....
3 1703 0
370(R) 104(B)
C 0(C) 11(c) 11 30 8:1 Cleared Verified Clear Number Tc.
5/G DI's FI's By U C By UC Indications se Plugged 0
A 11
+
0 5
9 C
21 13 10 1
1
........ - _............ ~.. _.. _..
8x1 testing is comp?.ete. All of first. shif t (total of 7) passed, othere are being graded. U C testing will be done on a liafted basis (approximately 6).
D
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