|
|
Line 1: |
Line 1: |
| {{Adams | | {{Adams |
| | number = ML20216H346 | | | number = ML20199C486 |
| | issue date = 09/12/1997 | | | issue date = 11/18/1997 |
| | title = Insp Repts 50-313/97-14 & 50-368/97-14 on 970527-0807.No Violations Noted.Major Areas Inspected:Maint & Engineering | | | title = Discusses Predecisional Enforcement Conference Conducted to Discuss Violations Noted in Insp Repts 50-313/97-14 & 50-368/97-14.Violations Noted:Failure to Comply w/10CFR50, App B,Criterion IX & Failure to Remove Unit 2 from Svc |
| | author name = | | | author name = Merschoff E |
| | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) | | | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| | addressee name = | | | addressee name = Hutchinson C |
| | addressee affiliation = | | | addressee affiliation = ENTERGY OPERATIONS, INC. |
| | docket = 05000313, 05000368 | | | docket = 05000313, 05000368 |
| | license number = | | | license number = |
| | contact person = | | | contact person = |
| | document report number = 50-313-97-14, 50-368-97-14, NUDOCS 9709160206 | | | document report number = 50-313-97-14, 50-368-97-14, EA-97-382, NUDOCS 9711200032 |
| | package number = ML20216H336 | | | package number = ML20199C490 |
| | document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | | | document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE |
| | page count = 27 | | | page count = 5 |
| }} | | }} |
|
| |
|
Line 19: |
Line 19: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:. _ _ _ . _ . _ ._-- _ . _ _ _ _ _ . . _ . . _ _ . _ _ _ _ __ __ ___ --. - | | {{#Wiki_filter:. . .m . _. _ |
| _ _ ...__ __
| | '' * , |
| d
| | e t h' "8 004 UNIT ED ST ATES |
| .
| |
| . l LNCLASML1 U.S. NUCLEAR REGULATORY COMMISSION -! | |
| i
| |
| . REGION IV :
| |
| - Docket Nos.: 50 313 [
| |
| 50 368 j License Nos.: DPR 51 i
| |
| -
| |
| NPF 6 ;
| |
| | |
| :
| |
| Report No.: 50 313/97 14 '
| |
| 50 368/97 14
| |
| "
| |
| Licensee: Entergy Operations, In ;
| |
| ,
| |
| Facility:- Arkansas Nuclear One, Units 1 and 2-t Location: Junction of Hwy. 64W and Hwy 333 South -
| |
| .
| |
| Russellville, Arkansas y
| |
| '
| |
| Dates: May 27 through August 7,1997 inspector: 1. Barnes, Technical Assistant
| |
| ,
| |
| Accompanied By: Dr. C. V. Dodd. NRC Consultant ,
| |
| Approved By: Arthur T. Howell lit, Director Division of Reactor Safety i
| |
| ,
| |
| v i
| |
| i iTTACHMENTS:
| |
| - Attachment 1: Supplemental Information . -
| |
| Attachment 2: Entergy Operations Arkansas Nuclaar One Unit 2 Position Paper on
| |
| "
| |
| " Missed" Indications I b
| |
| ,
| |
| O k
| |
| | |
| -
| |
| '
| |
| l P
| |
| ..
| |
| .
| |
| n .
| |
| m
| |
| '
| |
| , e.n,,,-w+>--.r- s,.w,.-- r. , v v yr-- y,- y- .e- %=vw ,.-rv~ w y v- --%-+ y y e-+ww-y '-wwy g y- 7- y-y- v y -w-r-, y-----+y--- --
| |
| g*-ru+w-~7*
| |
| | |
| . . . . ._- -- -- . -
| |
| .
| |
| . 2-EMC1)TIVE SUMMARY Arkansas Nuclear One, Units 1 and 2 NRC Inspection Report 50 313/97 14;50 368/97-14 This inspection was performed to assess Unit 2 steam generator degradation status and the circurnstances pertaining to the issue on April 11,1997, of a Notice of Enforcement Discretion (and subsequent Amendment 180 to Facility Operating License DPR 51) for the Unit 1 steam generator Technical Specification requirements, h1R!n19nangg
| |
| * No significant disagreement with either production analyst " calls" or resolution determinations occurred during review by the NRC consultant of a sample of Refueling Outage 2R12 eddy current data, it was considered that some of the axial tube flaws present at eggcrate locations would not be routinely identified if analysis was cursory, with analysis difficulty increased by the presence of mix residuals of the order of 1 volt and some of the data being noisy (Section M1.1).
| |
| | |
| Ena!DEliDG
| |
| * The licensee use of a majority of data from Crystal River 3 in the Unit 1 bobbin coil sizing technique qualification, which had a different intergranular attack morphology and tubing degradation location to that observed at Arkansas Nuclear One, Unit 1, was considered technically questionable. An apparent violation was identified in regard to the f ailure to comply with the data sample and morphology requirements of the selected qualification method (Section E1.1).
| |
| | |
| * The licensee f ailure to correct regression analysis predictions of flaw depths (to reflect effects of data scatter) was considered nonconservative and a direct contributor to under prediction of true depth of intergranular attack (Section E1.1).
| |
| | |
| * The acceptance limit in Appendix H of Electric Power Research Institute Report TR 106589-V1 of 25 percent root mean square error (for flaw depth sizing techniques) was considered a potential contributor to errors in eddy current depth siring predictions (Section E1.1).
| |
| | |
| * The overall Unit 2 Refueling Outage 2R12 examination scope was considered to be comprehensive and appropriately responsive to emerging degradation modes in Combustion Engineering steam generators Section E1.2).
| |
| | |
| * The degradation data differences between the Unit 2 steam generators indicated that differences in thermal hydraulic characteristics potentially existed (Section E1,2).
| |
| | |
| __
| |
| | |
| . _ ______________ __________ -
| |
| .
| |
| .
| |
| 3-
| |
| * An apparent violation was identified pertaining to the f ailure to comply with the requirements of the Unit 2 Technical Specifications 4.4.5.4a 7 and 4.4.5.4b during Outage 2F961. A total of 210 tubes were documented during a Refueling Outage 2R12 review of Outage 2F961 eddy current data history, as containing flaw indications in excess of the 40 percent plugging or repair limit of the Technical Specifications (Section E1.3).
| |
| | |
| * The licensee appropriately implemented Unit 2 oddy current program commitments made in a March 25,1997, meeting with the NRC staff. Of particular note was the use of an *Entergy Review Group" for improved oversight of eddy current data analysis activities, measures implemented to track and trend analyst performance, and actions taken to improve training and testing of analysts (Section E1.3),
| |
| * The licensee responded promptly to the notification of the identification of eggerate erosion damage in a Combustion Engineering steam generator at another licensee's f acility. No evidence of similar degradation was found in the Unit 2 steam generators (Section E1.4 * An apparent violation was identified pertaining to the failure to take prompt corrective action (i.e., acceptance for continued service, without a demonstrated technical basis) of sleeved tubes exhibiting weld zone eddy current indications. An inspection followup item was also identified regarding the future examination criteria adopted for two sleeved tubes, for which a nonconformance report disposition had identified the potential for early primary water stress corrosion cracking (Section E1.5).
| |
| | |
| * The microstructure present in both Tubes R16C56 and R70C98 (i.e., fine grain size and intragranular carbides) was characteristic of that resulting from use of a low final annealing temperature. This type of microstructure no;mally results in enhanced susceptibility to inte granular stress corrosion cracking, compared to a coarser grain structure with predominantly intergranular carbides present (Section E8.1).
| |
| | |
| * Lead and sulfur (as sulfate) were present on the crack surfaces of Tubes R16C56 and R70C98, indicating their invulvement in the degradation process (Section E8.1).
| |
| | |
| * Pitting or erosion of grain faces was observed at tube depths at least up to mid wall in Tube R70C98. The grain pitting appeared to be indicative of a crack that had been in existence for a considerable period. The metallurgicalinformation suggested that tube through wall degradation in excess of tube mid wall was
| |
| | |
| necessary at eggerate support locations prior to bobbin coil tube examinations being capable of its detection (Section E8.1).
| |
| | |
| * Burst test values of 3200 psig and 3250 psig, respectively, were obtained for Tubes R16C56 and R70C98. These values exceeded the main steam line break accident analysis pressure of 2950 psig, but did not meet the Regulatory Guide 1.121 structural integrity criteria of 4750 psig (Section E81).
| |
| | |
| .
| |
| . 4 figport Qq1p l Sgnrnar.y of Pla_n1Slalug Unit 1 was at 100 percent power dunng the onsito portion of the inspection. Unit 2 was in Refueling Outage 2R12 at the beginning of the inspection penod, with Mode 1 reached following the outage on June 9,199 Il<Molnignmitt M1 Conduct c' Maintenance M 1.1 fluview of Tube Examination Data jnpoection Sqqng The inspector selected a limited sample of Unit 2 eddy current data from Refueling Outogo 2R12 for independent assessment by the NRC consultant. Included in the assessment scope were tubes that had been selected for in situ pressure testing; tubes containing tungsten inert gas and kinetic welded sleeves; defectivo calls by the primary and/or secondary analyst, which were overruled by the resolution j
| |
| '
| |
| analysts; and calls initially identified by the Entergy Review Grou Observations and Findinns The NRC consultant had no significant disagreement with either production analyst
| |
| " calls" or the resolution process, as a result of his review of a sample of Refueling Outage 2R12 eddy current data. During review of kinetic welded sleeve data, the NRC consultant noted that the data indicated that significant probe lift off occurred during the examinations and questioned the inherent ability of the selected plus point probe to fit down into the wold region ond give a sensitive examination. The NRC consultant also considered that the plus point coil examination techniques (used for both the kinetic welded and tungsten inert gas wolded sleeves) had been optimized for detection of tubo outside diameter indications, For tungsten inert gas wolded sleeves, the inspector did not consider this examination approach to be the most sensitivo, in that the preponderance of defects would be expected to initiate at the interf ace between the sleeve and the tube inside diameter surf ac DurMg the inspection, the NRC consultant performed a limited review of three coil (i.e.,0.115 inch pancake, axial sensitivo, and circumferential sensitive) motorized rotating pancake coil and bobbin coil tube data from eggerate support location In this review, which included a small amount of data from both Refueling '
| |
| Outage 2R12 and Outage 2F961, it was noted _that some of the axial tube flaws present at eggcrato locations would not be routinely identified by cursory analysi The NRC consultant noted that a mix was set up using a drilled support plate to eliminate the effects of the eggcrate support. Af ter mixing, however, a residual of the order of 1 volt remained, which made defect identification more dif ficult. Some of the motorized rotating pancake coil data was also noisy, therefore, tending to mask smaller signal _ _ _ _ _ _ - _
| |
| | |
| .
| |
| , 5- C_0ECLVH19El No significant disagreement with either production analyst " calls" or resolution determinations occurred as a result of review by the NRC consultant of a sample of Refuehng Outage 2R12 eddy current data, it was considered that some of the axial tube flaws present at eggerate locations would not be routinely identified if analysis was cursory, with analysis difficulty increased by the presence of mix residuals of the order of 1 volt and some of the data being nois AllJnainM1h19 E1 Conduct of Engineering E 1.1 Unit 1 Notice of Enfo1 cement Discretion During Refueling Outage 1R13 in 1996, an oddy current technique was used to depth size intergranular attack that was present in Unit 1 steam generator tubing in the tube region located within the upper tube sheet. The licensee utilized the criteria contained in Supplement H2, " Qualification Requirements for Examination of Steam Generator Tubing," to Appendix H of Electric Power Research Institute Report TR 106589 V1, "PWR Steam Generator Examination Guidelines: Revision 4,"
| |
| Volume 1, for performance of the sizing molification. On April 8,1997, the licensee discovered that the destructive examination results (i.e., metallography)
| |
| from three tube samples removed during Refueling Outage 1R13 indicated nonconservative bias of up to 50 percent through wall in the eddy current measurements of 1 packages of intergranular attack, thus, creating the possibility that tubes could have been left in service with flaws which exceeded the plugging or repair limit of the Technical Specification The staff made a determination on April 11,1997, to exercise discretion not to enforce compliance with Technical Specification 4.18.5 b until May 7,1997, or the date of issuance of an amendment to Technical Specification 4.18.5.b. The amendment, which was issued on May 7,1997, authorized operation to the next refueling outage with tubes containing intergranular corrosion indications in excess of the Technical Specifications plugging limit, [pJngetion Scone (50002)
| |
| The inspector reviewed the sizing qualification report, including the methodology used for the sizing qualification process; the criteria contained in Appendix H to Electric Power Research Instituto Report TR 106589 V1; the circumstances described in the request for Notice of Enforcement discretion; and the information contained in Licensee Event Report 50 313/97-001 01 and its supplement dated July 15,199 _ _ _ _ - - -
| |
| | |
| _ _ _ .. _ _ _ _ _ - _ _ _ _ _ _ _ . _ _ _ . _ . . _ _ _ _ . _ _ _ _ _ _ -
| |
| ,
| |
| * | | * |
| ,
| | ,, t NUCLEAR REGULATORY COMMISSION e REGlo4 tv g |
| , 6- | | g 611 RYAN PLAZA DRIVE, SUITE 400 |
| .
| | %. 4, * * * * * ,0 AR LINoTON, T EXAS 760108064 November 18, 1997 EA 97-382 C. Randy Hutchinson, Vice President Operations Arkansas Nuclear One Entergy Operations, Inc. |
| I Qbservations and Findinas During review of Supplement H2 to Appendix H of Electric Power Research Instituto Report TR 106589 V1, the inspector ascertained that Report TR 106589 V1 (which has not been formally endorsed by the NRC) considered a technique qualified for sizing if a root mean square error of less than or equal to 25 percent through wall was demonstrated. The inspector considered that if this root mean aquare error acceptance limit was approached in technique qualification, it would create the potential for errors in eddy current depth sizing predictions. Of particular concern to the inspector was information provided by licensee personnel, which indicated that the regression equation (derived from the relationship between eddy current flaw ,
| |
| depth predictions and metallographic measurements of flaw depths) was applied without correction for the offects of observed eddy current data scatter on the flaw ,
| |
| depth sizing prediction. The inspector viewed this information as a probable direct contributor to the observed under prediction of actual flaw depths in the removed tube samples.
| |
|
| |
|
| "
| | 1448 S.R. 333 Russellville, Arkansas 72801-0967 SUBJECT: NOTICE OF VIOLATION (NRC INSPECTION REPORT 50-313/97-14; 50-368/97-14) |
| Review by the inspector of the data used to establish the sizing qualification identified that the majority of the data originated from Crystal River 3. The inspector questioned the basic validity of this approach, in that it did not appear technically appropriate to use tubing free span intergranular attack, from a plant with a different operating history to Arkansas Nuclear One, Unit 1, to predict depths of intergranular attack in tubing in the upper tube sheet. The NRC consultant also informed the inspector regarding British work that was performed in the 1980s by McIntyre on conductivity of intergranular attack. This work, which was published as Electric Power Research Institute Report NP-4478, identified that tubing intergranular attack located in a tube sheet could exhibit significantly greater conductivity (i.e., up to a factor of six) than intergranular attack located in the free span of tubing. This information both illustrated a potential pitfallin the licensee sizing approach and suggested to the inspector that eddy current detection capabikties would be lower for intergranular attack located in a crevice region compared to degradation present in the tubing free spa ,
| |
| Review of the root cause analysis contained in the supplement to Licensee Event Report 50 313/97 001 identified two root causes for the under prediction of flaw depths. The first identified root cause was that the data sets used in the sizing qualification did not fully conform to the requirements of Appendix H to Electric Power Research Institute Report TR 106589-V1, with respect to data distribution (i.e., only 7 of 53 data points were greater than or equal to 60 percent flaw depth versus a requirement for at least two thirds of the data points to be in that range).
| |
|
| |
|
| The second identified root cause was that there were morphology differences between the flaws used to develop the sizing technique and the flaws in tubes removed during Refueling Outage 1R13 for laboratory examiution. The intergranular attack in the Crystal Rivet 3 tube samples was symmetrical and pit like in configuration, whereas the intergranular attack in the licensee pulled tube samples (removed from the steam generators in 1982 and 1984) was patch like
| | ==Dear Mr. Hutchinson:== |
| '
| | This is in reference to the predecisional enforcement conference held in the NRC's Arlington, Texas office on September 26,1997. The conference was conducted to discuss three apparent violations of NRC requirements related to inspections of steam generator tubes at Arkansas Nuclear One, Units 1 and 2. The results of our inspection were discussed with your staff during a telephonic exit briefing conducted on August 8,1997, and were documented in the subject NRC Inspection Report, which was issued September 12,1997. |
| with no consistent physical shape. The licensee reported that the intergranular attack in the 1996 pulled tube samples was also patch like in configuration, but contained sho3t deep intergranular penetrations. The licensee concluded that the
| |
| _ - _ _ _ _ _ _ , _ _- __ _ ,_ . _ _ _ _
| |
|
| |
|
| l
| | After the September 26 conference, ANO provided further information and clarification during a telephone discussion held on October 2,1997. |
| .
| |
| '!
| |
| , -7 ;
| |
| morphology differences between the data sets used for the sizing qualification and
| |
| - the observed flaws in the tube samples removed during 1996, resulted in the data sets not being appropriate for sizing the 1996 morphology flaws. The inspector did x
| |
| not fully concur with the licensee position, in that the bobbin coil examination technicue inherently averages the depth of the flaw present. Accordingly, patch-like flaws without consistent physical shape would inherently tend to be sized with potential vanance from true maximum dept The performance nf the sizing qualification, using data that did not conform to the requirements of the selected qualification method, was identified to licensee personnel on August 8,1997, as an apparent violation of Criterion IX of Appendix B te 10 CFR Part 50 (50 313/9714-01).
| |
|
| |
|
| }
| | The three apparent violations described in the inspection report involved: (1) a failure to comply with 10 CFR Part 50 Appendix B Criterion IX in the use of data in bobbin coil sizing qualification criteria that did not conform to the requirements of the selected qualification method; (2) the failure to remove Unit 2 tubes from service which contained flaws that exceeded the plugging limit of the Technical Specifications; and (3) the lac!; of prompt corrective action in November 1995 prior to-retuming potentially defective sleeved tubes to service. |
| - _QR DI R M L7
| |
| .
| |
| .
| |
| The licenseb use in the sizing qualification of a majority of data from Crystal River 3, whLi had a different morphology and tubing location to the observed Unit 1 intergranular attack, was considered questionable. The ac eptance limit in Appendix H of Electric Pawer R9 search Institute Report TR-106589 V1 of 25 percen;ioot mean square error (for flaw depth sizing techniques) was considered a potential contributor to errois in eddy current depth sizing predictions. The licensee f ailure to correct regression analysis predictions of flaw depths, to reflect effects of data scatter, was considered a direct contributor to the observed under prediction of trae depth of intergranular attack in the pulled tube samples. An opparent violation was identified in regard to the f ailure to compiy with the data sample and-morphoiogy requirements of the selected qualification nietho E1.2 Review of Refuelina Outaae 2R12 Tube Examination Scoce and Results Insoection Scone (GOOO2J The inspector reviewed the tube examinatin scope and methods that were csed in Refueling Outage 2R12 with respect to Technical Specification requirements, | |
| *
| |
| industry guidance, and as a result of emerging degradation modes. A review was also performed of tube plugging and repair history for the Unit 2 steam generators, the specific plugging data for R:!1eling Outage 2R12, and the reasons for tube
| |
| _ plugging, i Observations and Findinas The licensee Unit 2 final examination scope for Refueling Outage 2R12 was: a full
| |
| '
| |
| length bobbin coil examination of_ all active tetos in both steam generators, with the exception of tne aiea below sleeves (where 20 percent of the tubes were examined with the'cobbin coil); a rotating pancake coil examination of all active hot-leg side expansion transition regions in both steam generators; a rotating pancake coil examination of 20 percent of the cold-leg side expansion transition regions in the
| |
|
| |
|
| , . _-__ --____-____ - _ ..
| | During the September 26 conference, Enter'g y expressed disagreement with the first apparent violation, contending that Entergy's sizing qualification for intergranular attack (IGA) met the provisions of appropriate industry qualification processes (EPRI Appendix "H"), and therefore Entergy was in compliance with 10 CFR Part 50 Appendix B. With regard to the second _ |
| .
| | apparent violation, which involved Unit 2 steam generators, Entergy agreed that the violation occurred but contended that the violation warranted treatment as a non-cited violation. |
| .
| |
|
| |
|
| area bounded by the sludge pile; a plus point probe examination of all active Babcock and WHcox kinetic welded sleeves: a plus point probe examination of a 20 percent sample (plus 28 previously identified indications) of active Combustion Engineering tungsten inert gas welded sleeves (which was expanded to 100 percent
| | Entergy also disputed the third apparent violation, which is discussed below. |
| ) of the sleeves in Steam Generator B); a rotating pancake coil examination of a 40 g percent sample of Row 1 low radius U bends; and rotating pancake coil examination of a 20 percent sample of tubes at hot leg side drilled support locations (which was expanded to 100 percent for Steam i3cnerator B). The referenced rotating pancake coil probe examinations utilized a 0.115-inch diameter pancake coil, an axici
| |
| ] direction sensitive coil. and a circumferential direction sensitive coil. This probe is commonly referred to as a Delta probe. The licensee has been able to demonstrate, by in-situ pressure testing, that its selection of the Delta probe (over the more sensitive plus point probe) for examination of hot-leg side expansion transitions has not resulted in a decrease in demonstrated structural integrity margin with respect to Regulatory Guide 1.121 requirements. The inspector considered the overall examination scope to be comprehensive and appropriately responsive to emerging degradation modes in Combustion Engineering steam generators, b Table 1 lists the cumulative repair history for Unit 2 Steam Generators A and B.
| |
|
| |
|
| 3 A total of 350 tubes and 366 tubes, respectively, were removed from service in f Steam Generators A and B, which resulted in respective reductions of 78 and 18 y( sleeved tubes in service. Both steam generators remained below current approved
| | The NRC considered Entergy's arguments and has concluded that vic;dions of NRC requirements occurred. |
| *~ effective repair limits, f Tables 2 and 3 show, respectively, a cumulative compilation for Steam Generators A and B of degradation by repair location. These tabulations con"nue to exhibit what was first observed, during the initial steam generator tube integrity inspection performed in 1992, regarding differences between the steam generators in incidence of degradatio As of Refueling Outage 2R12, the total repairs in Steam Generator A related to the top of tube sheet and sludge pile regions (i.e., degiadation helieved to be related prirnarily to outside diameter stress corrosion cracking in the tube expansion transition region and sludge pile intergranular attack) were 1542 tubes, versus a corresanding total of 520 tubes in Steam Generator B. Attempts to explain this difference in incidence of degradation between the steam generators focused on the observec' greater historical sludge quantities in Steam Generator A and the original return of the partial domineralizer output to the condensate train for Steam Generator B. The corresponding degradation rates at eggerate support locations (believed to be primarily axial stress corrosion cracking) were, however, reversed, with a total of 280 tubes repaired in Steam Generator A, versus a total of 604 tubes in Steam Generator B. Wear at batwing locatiuns in the tube bundle also varied significantly between the steam generator 3 as of Refueling Outage 2R12, with respective repair totals of 19 tubes in Steam Generator A and 121 tubes in Steam Generator I 1,
| |
| - _ _ _ _ _ _ _ _ _ _ _ _ _ _-- _
| |
|
| |
|
| .__
| | Based on the discussions The at the violations conference, weare cited have in the modified the enclosed first violationNotice to stateof Violation (No that 7 Entergy's inservice inspections of Unit 1 steam generators were not appropriately controlled g |
| *
| | , and accumplished to identify tube defects that exceeded the technical specification plugging (f 9711200032 971118 i |
| ;
| | . |
| .
| |
| | |
| Table 1 UNIT 2 GTEAM GENERATORS (SGs) A AND 8 TUBE REPAIR HISTORY Time of Repair Operational . SG A Repairs SG B Repairs Outage N Time (EFPYs')
| |
| Plugs SI' Plugs Sl2 Pre Commercial 0.00 15 0 29 0 2R1 (1981) 0.89 0 0 0 0 2R2 (1982) 1.69 0 0 1 0 2R3 (1982) 2.?3 0 0 0 0 2R4 (1985) 3.31 0 0 0 0 2RS (1986) 4.16 0 0 0 0 2R6 (1988) 5.38 0 0 6 0 2R7 (1989) 6.52 0 0 0 0 2R8 (1991) 7.67 0 0 73 0 2F92 (1992) 8.51 29 392 11 56 2R9 (1992) 8.85 67 ( 4) 132 0 2P93 (1993) 9.36 47 0 3 0 ,
| |
| ,
| |
| 2R10 (1994) 10.20 170 (-1 ) 77 (-1 ) ___
| |
| 2P95 (1995) 10.85 215 0 85 0
| |
| -
| |
| 2R11 (1995) 11.49 200 412 150 175 2F961 (1996) 12.43 73 ( 6) 144 (-7)
| |
| '
| |
| 2R12 (1997) 12.80 350 (-78) 366 (- 18) d
| |
| * '
| |
| \ M
| |
| ' ' '
| |
| '
| |
| _
| |
| Total Repairs 1881 1282 ]
| |
| Effective Repairs (Plug Equivabnt)' 1192 1083
| |
| % Repairs (Effective. Total) 14.18, 22.36 12.88,15.24 1 - Effective full power years of operation 2 - Sleeves 3 - Based on 18 sleeves for kinetic welded and 44 sleeves for tungsten inert gas welded being considered thermally equivalent to 1 plug | |
|
| |
|
| ,
| |
| *
| |
| I
| |
| . | | . |
|
| |
|
| Table 2
| | Entergy Operations, Inc. -2-limit. The second violation involves Unit 2 steam generator tubes, with defects greater than |
| _
| | ' the technical specification plugging limit, that were improperly considered operable. |
| UNIT 2 STEAM GENERATOR A REPAIR HISTORY BY REPAIR LOCATION Outage EFPYs' RW2 TSP) SP' TTS5 FS5 OT'
| |
| t Pre Commercial 0.00 0 0 0 ,
| |
| 0 15 2R1 (1981) 0.89 0 0 0 0 0 0 2R2 (1982) 1.69 0 0 0 0 0 0 2R3 (1983) 2.33 0 0 0 0 0 0 2R4 (1985) 3.31 0 0 0 0 0 0 2RS (1986) 4.16 0 0 0 0 0 0 2R6 (1988) 5.38 0 0 0 0 0 0 2R7 !1989) 6.52 0 0 0 0 0 0 2R8 (1991) 7.07 0 0 0 0 0 0 l
| |
| ,
| |
| 2F92 (3/1992) 8.51 1 1 166 253 0 0 2R9 (10/1992: 8.85 9 30 11 17 0 0 2P93 (1993) 9.36 0 0 2 45 0 0
| |
| ., | |
| 2R10 (1994) 10.16 3 7 13 147 0 0 2P95 (1/1995) 10.86 0 0 12 203 0 0 2R11 (9/1995) 11.46 2 43 38 519 16 21 2F96-1 (1996) 12.43 0 29 6 13 4 21 2R12 (1997) 12.80 4 170 31 66 38 41 m + m Total Repairs 19 280 279 1263 58 98 1 Effective fu:1 power years of operation 2 - Batwing 3 - Tube support plate 4 - Sludge pile 5 - Top of ~;ube sheet ti Free span 7 - Other
| |
| !
| |
| -
| |
| | |
| 0-11
| |
| .
| |
| Table 3 UNIT 2 STEAM GENERATOR B REPAIR HISTORY BY REPAIR LOCATION Outage EFPYs' BW' TSP' SP* TTS' FS* OT'
| |
| Pre-Commercial O.00 O O O O O 29 2R1 (1981) 0.89 0 0 0 0 0 0 2R2 (1982) 1,69 1 O O O O O 2R3 (1983) 2.33 0 0 0 0 0 0
| |
| _
| |
| 2R4 (1985) 3.31 O O O O O O 2H5 (1986) 4.10 0 0 0 0 0 0 2R6 (1988) 5.38 6 0 0 0 0 0 2R7 (1989) 6.52 O O O O O O 2R8 (1991) 7.67 16 52 5 0 0 0 2F92 (3/1992) 8.51 2 3 4 58 0 0 2R9 (10/1992) 8.85 25 94 5 8 O O 2P93 (1993) 9.36 0 0 0 3 O O'
| |
| 2R10 (1994) 10.16 17 32 5 23 0 1 2P95 (1/1995) 10.86 0 0 5 80 0 0 2R11 (9/1995) 11.46 19 73 19 215 0 4 2F961 (1996) 12.43 11 100 3 13 0 17 2R12 (1997) 12.80 24 250 21 53 4 14 m
| |
| Total Repairs 121 604 67 453 4 65 1 - Effective full power years of operation 2 - BatwinD 3 - Tube support plate 4 - Sludge pile 5 Top of tube sheet 6 - Free span 7 - Other
| |
| -_
| |
|
| |
|
| l
| | Although these violations involve steam generators of different designs and different IGA flaw sizing methodologies, the violations are similar in that your methods for sizing lGA and tubing degradation in both Units 1 and 2 steam generators were shown to have been inadequate. |
|
| |
|
| .
| | The actual safety consequence of these violations is low because no significant leakage occurred as a re: ult of tube leaks. These issues are of regulatory significance because of the programmatic nature of the problem; that is, you relied on methods which were incapable of detecting flaws greater than the plugging limit, and as a result, operated Units 1 and 2 with steam generators with tubes containing flaws that exceeded the technical specification plugging limit of 40 percent through wall (TW). Therefore, these two violations have been categorized in the aggregate in accordance with the " General Statement of Policy and Procedure for WRC Enforcement Actions"(Enforcement Policy), NUREG-1600, as a Severity Level lli problem. |
| .
| |
| -12-
| |
| :. Conclusions The overall Unit 2 Refueling Outage 2R12 examination scope was comprehensive and appropriately responsive to emerging degradation modes in Combustion Engineering steam generator The degradation data differences between the A and B steam generators indicated that differences in thermal-hydraulic characteristics potentially existe E1.3 Review of Steara Generator Tube Examination Proaram Reauirements inspectipfi Scoce (50002)
| |
| The inspector compared the steam generator eddy current examination program requirements or Refueling Outage 2R12 against specific commitments made by the licensee during a March 25,1997, meeting with the NRC's Office of Nuclear Reactor Regulatio The inspector reviewed: Document ANO 2-OTH ESP-SGMAN, " Arkansas Nuclear One Unit 2 Steam Generator Eddy Current Training Manual," Revision 1; Engineering Standard HES-14, " Steam Generator Integrity Program Unit 2," Revision 1:
| |
| Engineering Standard HES-26, " Steam Generator Tube Plug Defect Management,"
| |
| Revision 7: Engineering Standard HES-28, "ANO-2 Steam Generator ECT Data Analysis Guidelines," Revision 6 and Standard Change Notice (s) 01 through 04; Engineering Standard HES-29, "ANO-2 Steam Generator ECT Perfo; nance Demonstration Guidelines," Revision 3; Engineering Standard HES 31, "ANO 2 ECT Data Management and Control," Revision 0; and HES-42, " Steam Generator Tube Sleeve Management," Revision 1. In addition, the inspector observed in-process eddy current analysis activities, discussed program criteria with licensee and contractor staff, and made a specific review of the Refueling Outage 2R12 program requirements with respect to the March 25,1997, meeting commitments, Observations and Findinas The inspector determined that the licensee had appropriately implemented the commitments made in the March 25,1997, meeting. The inspector considered the most significant process change maae was the creation of an "Entergy Review Group" to strengthen the oversight of eddy current analysis activities. This group, which was made up of experienced Level 111 eddy current analysts, reported directly to licensee supervision and provided an independent evaluation of the analysis process, included in the group charter was a review of data discarded by the resolution process, sampling of tubes with "no_ detectable degradation" calls by
| |
| :
| |
|
| |
|
| ___- _____-
| | In accordance with the Enforcement Policy, a base civil penalty in the amount of $55,000 is considered for a Severity Level 111 problem. Because your facility has been the subject of escalated enforcement actions within the last 2 years', the NRC considered whether credit was warranted for Identification and Corrective Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. Since Entergy identified both issues, the NRC has determined that Entergy is deserving of identification credit. |
| .
| |
| .
| |
| -13-primary and secondary analysts, review of historical comparisons performed by the resolution analysis team, review of guideline change forms, and ensuring that all repairable codes that had been dispositioned as non-repairable by resolution analysis had been. appropriately reviewed and documente The inspector also noted that the data analysis guidelines had been simplified to make them more user friendly, training of analysts was more structured with direct licensee personnel involvement, and testing of analysts was more focused on known Unit 2 degradation modes. All bobbin coil indications at eggerate supports, which were confirmed by a rotating pancake probe, and bobbin coil indications in the free span that exhibited growth, were required to be plugged. The licensee also implemented, for the first time, use of new Zetec analyst efficiency tracking software. This toftware provided the capability u track and trend individuai analyst 4 performance. 't he performance improvements that resulted from use of this process were notabl Review of the program criteria identified that acquisition of eddy current data was required to conform to the Electric Power Research Institute PWR Examination Guidelines. The inspector noted, however, no similar expectation with respect to performance of eddy current data analysis. The inspector found this status to be somewhat unusua While observing in process eddy current analysis activities, the inspector noted that eddy current analysts had documented (for indications that were confirmed s
| |
| to be flaws during Refueling Outage 2R12 and accordingly removed from service)
| |
| that re-review of Outage 2F96-1 data for these tubes indicated that flaw indications in excess of the 40 percent plugging or repair limit (Unit 2 Technical Specifications 4.4.5Aa.7) were present at that time. The inspector ascertained that a total of 78 tubes in Steam Generator A and 132 tubes in Steam Generator B were documented, from the re-review of Outage 2F961 data, to contain flaw indications in excess of the Technical E pecification limit. Unit 2 Technical Specification 4.4.5.4b requires steam generator tubes, which exceed the plugging or repair limit, to be plugged or repaired before determining the steam generator to be operable. The failure to plug or repair the 210 affected tubes in Outage 2F96-1, prior to return of the Unit 2 steam generators to service, was identified to licensee personnel on August 8,1997, as an apparent violation of the Unit 2 Technical Specifications 4.4.5.4a 7 and 4.4.5.4b (50 368/9714 02).
| |
|
| |
|
| The licensee subsequently provided a position papes regarding these indications,
| | Further, the NRC has determined that Entergy is deserving of Corrective Action credit. |
| -which has been included as Attachment 2 to the inspection report. The inspector concurred with the licensee's comments that the leaker outage, Outage 2F96-1, was not caused by a missed call, that the flaws detected in Refueling Outage 2R12 were typically of small amplitude, that the in situ pressure test results were adequate, and that the enhanced performance resulted from the program improvements made prior to Refueling Outage 2R12. The position paper did not
| |
| - address, however, that it was the licensee's records, which indicated that the
| |
| _ _ _ _ _ _ - _ _ _ _ _ _ - - _ _ _ _ .
| |
|
| |
|
| _ - - - -
| | Entergy's corrective actions for the first violation included declaring the steam generators inoperable, notifying the NRC, requesting (and being granted) a Notice of Enforcement Discretion from the NRC, administratively precluding IGA sizing, and investigating the cause of the problem. Entergy's corrective actions to the second violation included reviewing more |
| _ - . - - - . - -- -
| | "no detectable degradation" determinations to verify the quality of the 2R12 analyses, performing enhanced inservice inspectio".s of indications, repairing the eggerate cracks, enhancing training, and providing additional oversight of the inservice inspections. |
| o
| |
| .
| |
| -14-Unit 2 Technical Specification plugging or repair limit was exceeded for the tubes in Outage 2F961, without plugging or repair being accomplished. The inspector also noted that bobbin amplitudes for indications were similar between Outage 2F961 and Refueling Outage 2R12, and accordingly concluded that the defect indications were present in Outage 2F901. The difference in result appeared to the inspector to relate primarily to the increased use in Refueling Outage 2R12 of confirmatory rotating probe examination , Conclusions An apparent violation was identified pertaining to the failure to comply with the requirements of the Unit 2 Technical Specifications 4.4.5.4a.7 and 4.4.5.4b during Outage 2F961. A total of 210 tubes were documented during a Refueling Outage 2R12 review of Outage 2F961 eddy current data history, as containing flaw indications in excess of the 40 percent plugging or repair limit of the Technical Specifications. The licensee appropriately implemented Unit 2 eddy current program cornmitments made in a March 25,1997, meeting with the NRC staff. Of particular note was the use of an "Entergy Review Group" for improved oversight of eddy current data analysis activities, measures implemented to track and trend analyst performance, and actions taken to improve training and testing of analyst E1.4 VisualInsoeetion of Unit 2 Steam Generator Eancrate Supports n Lnmaction Scone (50002)
| |
| The inspector revwwed video tape of remote visual examinations that were performed, using a Welch-Allen camera, of the periphery of the eggerate supports in Steam Generator A. Visual examination of the eggcrate peripheries in the Unit 2 steam generators was performed by the licensee following notification of the identification of erosion damage at this location in the San Onofre Nuclear Generating Station, Unit 3, steam generators. The specific tape segment reviewed was obtained from the West side of the steam generator hot leg, Qbservations and Findinas The video tape reviewed indicated that the peripheral qgcrate isttice in all of the supports was in excellent condition, with no visual evidence of any erosion damage or thinning, Conclugigng The licensee responded promptly to the notification of the identification of eggerate erosion damage in a Combustion Engineering steam generator at another licensee's facility. No evidence of similar degradation was found in the Unit 2 steam generator __ _ . _ . - - . -
| |
| _ _ _ _ __ _ _ _ _ . . _ _
| |
| .
| |
| .
| |
| -15-E1.5 Review of Unit 2 ABB-Combustion Enaineerina Tunasten inert Gas Welded Sleeves Inspection Scone (50002)
| |
| A followup was performed of problems associated with the installation of welded sleeves by ABB Combustion Engineering during Unit 2 Refueling Outage 2R11 in November 1995. The inspection scope included a review of sleeve welding procedure and welding operator qualifications, procedural requirements, and sleeve nonconformance c.ieory and disposition Observations and Findinas The inspector reviewed automatic gas tungsten arc Welding Procedure Specification GTAA-43.43.992, Revision 0, and its supporting procedure qualification record, GTAA 43.43 01503, Revision 01. In addition, a review was performed of the qualifications of six welding operators who had performed automatic gas tungsten arc sleeve welds during Refueling Outage 2R11. No discrepancies were noted during this revie The inspector reviewed the technical requirements contained in contract procedures for sleeve installation pertaining to tube cleanliness preparation; sleeve insertion, expansion, and welding; post-weld heat treatment; and nondestructive examination of the siseve, tube parent metal, and weld, included in this review were:
| |
| (1) Procedure STD-400157, " Visual Examination of Steam Generator Tubs Sleeve and Sleeve Plug Welds," Revision 0; Procedure 00000 NSS-062, " Procedure for Ultrasonic Examination of Steam Generator Tube to Sleeve Upper Welds, Revision 10; Procedure STD 400-151, "Linearity Verification of Ultrasonic Instrumentation Used for the Examination of Sleeved Tube Welds, Revision 0; Procedure STD-400158, " Technical Operating Procedure for the Transition Zone Sleeve Rolling Tool and Controls for Steam Generator Tubes With 3/3 inch .043 inch or .048 inch Wall Thickness Using the Computer Controlled Sleeving | |
| - System," Revision 0; Procedure STD 500-020, " Technical Operating Procedure for the Post Weld Heat Treatment and Heater Dryout of Above the Tubesheet (ATS)
| |
| Welds Using Resistance Heating for Steam Generator Tubes With .048 inch Wall Thickness," approved July 31,1992; and Traveler 2005331-001, " Installing Tube Sleeves in the Steam Generator Tubesheet Expansion Transition Zone Using the Genesis 2000 Control System," Revision O. The inspector observed during this review, that no provisions for inspection of the cleaned tube surface (prior to insertion of the sleeve) had been included in the process controls. This procedural guidance omission was subsequently corrected by ABB Combustion Engineering following the identification of the presence of oxide inclusions in Prairie Island pulled tube sample L . _ - . _ _ _ _
| |
| .
| |
| .
| |
| - 16-The inspector ascertained from review of Contract NAS00176, effective date-August 14,1995, that the contractor's quality assurance program was invoked for the sleeving activity and that the licensee had required its review and approval of disposition recommendations for nonconforming conditions. The inspector examined the nonconfortronce report log for the contract (ABB Project 2005331)
| |
| and noted a total of 28 reports existed. The inspector selected a sample of nonconformance reports for review, including those issued for sleeve over expansions and eddy current exarnination indications. All nonconformances reviewed were noted to have been signed by either one or two licensee employee During review of Nonconformance Reports 2005331-14, 2005331-17, and 2005331 18, the inspector noted that ABB Combustion Engineering dispositioned tubes, which exhibited eddy current wald zone indications and suspected geometric anomalies, as "use-as is" based solely on review of ultrasonic examination and visual inspection results. Samples of sleeved tubes exhibiting eddy current indications viere not removed from the steam generators for laboratory examination to allow identification of the reason (s) for the eddy current signals and verification that other nondestructive examination methods could be relied upon for assurance of structural integrity. Laboratory examination in 1996 of samples of Prairie Island sleeved steam generator tubes, exhibiting similar eddy current indications, identified that the eddy current indications were the result of circumferentially oriented oxide inclusions and/or weld suckback on the sleeve outside diameter. These indications t were contrary to the sleeve license amendment, in that the minimum weld height was not being met and the oxide inclusions provided potential leak path The inspector questioned the licensee regarding the technical bases for disregarding the results of the eddy current examinations, in that it had not been demonstrated that ultrasonic examination and visual inspection could be relied upon to determine whether the eddy current indications were (or were not) indicative of the presence of flews Licensee personnel indicated in response that eddy current examin. ?iens were considered, at the time of installation of the Unit 2 sleeves, to not be applicable to acceptance of the sleeve installation. Rather, the examinations were performed as a baseline for subsequent inservice eddy current examinations. The inspector reviewed the topical report (marked as containing proprietary information) {
| |
| that had been approved by NRC staff, " Arkansas Nuclear One Unit 2 Steam f Generator Tube Repair Using Leak Tight Sleeves," dated July 1992, in order to establish governing requirements. Section 3.0, Acceptance Criteria, indicated with respect to Criterior 6 (pertaining to nondestructive examination for levels of detectability requirad to show structura! adequacy) that an eddy current technique had been developed that could detect a 20 percent through-wall defec Section 5.0, Sleeve Examination Programs, stated that a multi-frequency eddy current method would be used to provide a volumetric examination of the pressure boundary,_ including the sleeve, parent tube, and weld. The inspector also noted in Section 5.1.2 that the only defined expectation for ultrasonic examinations was to confirm the presence of satisfactory sleeve to tube weld fusio i
| |
|
| |
|
| ._ _____ _ - ___ - -
| | Therefore, to encourage prompt identification and comprehensive correction of violations, I have been authorized not to propcse a civil penalty in this case. However, significant violations in the future could result in a civi! penalty. |
|
| |
|
| I
| | The third violation cited in the enclosed notice involves Entergy's failure to take prompt corrective actions prior to retuming to service potentially defective sleeved tubes exhibiting weld zone eddy current indications. At the conference, Entergy representatives expressed disagreement with this violation, contending that sufficient technical basis existed (including |
| .
| | ' A Severity Level 111 problem and a $50,000 civil penalty was issued on April 9,1997, involving fire protection inadequacies and the plant staff's inadequate response to ind:ations of a fire (EA 96-512). |
| -17-The inspector did not agree with the licensee historical understanding of the expected role of eddy current examination, after review of the wording of the applicable topical report. The inspector also considered the view to not be relevant to the central concern; i.e., acceptance for continued service despite the lack of a demonstrated technical basis for disregarding the presence of eddy current examination weld zone indications. The failure to take prompt corrective action, prior to returning potentially defective sleeved tubes to service, was identified to licensee personnel on August 8,1997, as an apparent violation of Criterion XVI of Appendix B to 10 CFR Part 50 (50 368/9714-03).
| |
|
| |
|
| During review of Nonconformance Report 2005331-25, which pertained to the sleeve lower joint torque roll exceeding the maximum specified limit in two tubes, the inspector noted that the basis for a "use-as-is" dispositica included a statement (addresning the potential for development of primary water stress corrosion cracking) that a minimum life of more than 1.5 effective full power years was expected. The inspector considered that future examination requirements for these two tubes should have been evaluated as part of the acceptance of a "use-as-is" disposition, based on the potential for a short remaining life. Review of licensee actions in this regard are considered an inspection followup item (50-368/9714-04).
| | . .. - . - -. . |
| | * , |
| | t ; |
| | ' ~ |
| | Entergy Operations, Inc. 3 ultrasonic testing and visual testing) to justify use of the sleeves as-is. Further, Entergy representatives noted thet, after re-inspection, only one of 28 indications was identified for repair based on new acceptance criteria and that all 28 indications met structural and leakage requirements; Notwithstanding Entergy's argument that the structural integrity of the sleeves was maintained, the NRC's position is that Entergy had an indication of a condition adverse _to quality, and it took inadequate corrective actions. Specifically, your actions (in October 1995, during 2R11) were not adequate to evaluate defects that existed at the time, nor did you perform an operabrity assessment at the time. Therefore, in accordance with the its Enforcement Policy, the NRC has classified this violation et Severity Leve,l IV.- |
| | You are required to respond to this letter and should follow the instructions specified in the - |
| | encksed Notice when preparing your response in your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. The NRC will use your response, in part, to deteimine whether further enforcement action is necessary to ensure compliance with regulatory requirements, in accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its enclosure, arid your response will be placed in the NRC Public Document Room (POR). |
|
| |
|
| The results of staff review of weld defects in ABB Combustion Engineering designed steam generator tube sleeves were transmitted to the licensee in a letter cated March 14,1997. This letter reflected a commitment by the licensee to perform a 100 percent plus point probe examination of welded sleeves with eddy current indications from the previous inspection, plus an examination of a 20 percent sample of the remaining previously installed welded ABB Combustion engineering sleeves, Conclusions An apparent violation was identified pertaining to the acceptance for continued service, without an appropriate demonstrated technical basis, of sleeved tubes exhibiting weld zone eddy current indications. An inspection followup item was also identified rcgarding the future examination criteria adopted for two sleeved tubes, for which a nonconformance report disposition had identified the potential for early primary water stress corrosion crackin E8 Miscellaneous Engineering issues E (Closed) inspection Followuo item 50-368/9628-03: Review of Laboratory Examination Results for Outaae 2F96-1 Tube Samolet Insoector Followuo The inspector reviewed the laboratory examination results for samples of two Steam Generator A tubes, R16C56 and R70C98, which were documented in ABB
| | Sincerely, |
| _ _ _ _ _ _ _ _ _ _ _ _ .
| | 'f |
|
| |
|
| .=
| | Ellic W. Mers Regional Adm'iistrator Dockets: 50-313-50-368 Licenses: DPR 51 NPF-6 |
| ,
| | ' |
| .
| | cc w/ Enclosure: |
| - 18-Combustion Engineering Final Report A-PENG-TR Oll, " Examination of Steam Generator Tubes Removed From ANO 2 in 1997," June 1997. Field motorized rotating pancake coil examination during Outage 2F96-1 identified the presence of single axial indications in these it.bes at the 01H eggcrate support. For Tube R16C56, the estimated through wall depth was 78 percent, the amplitude 8.42 volts, and the length 1.13 inches. For Tube R70C98, the estimated through wall depth was 85 percent, the amplitude 11.14 volts, and the length 1.15 inche The inspector noted from review of metallographic data for Tube R16C56 that approximately 10 other axialintergranular stress corrosion cracks were observed to be present in addition to the main crack. The average depth of the main crack was approximately 55 percent tnrough wall, with a maximum depth of 100 percent through wall. The largest of the other intergranular stress corrosion cracks was about 77 percent through wall. The inspector concluded from review of the metallographic data that the extent of degradation present was significantly under predicted by the Outage 2F961 eddy current examinations. The inspector also noted that significant intergranular attack surrounded the stress corrosion cracks, suggesting that environment and material susceptibility were more significant contributors to the observed degradatior then membrane stres The average depth of the main crack in Tube R70C98 was found by metallographic examination to be approximately 78 percent through wall, with the maximum depth ascertained to be (like Tube R16C56) 100 percent through wall. Approximately eight axialintergranular cracks were found to be present in addition to the main crack, with a maximum depth of 100 percent through wall. The cracking parallel to the main crack in Tube R70C98 was noted to be more severe than the corresponding parallel cracks in Tube R16C56. The inspector noted that, like Tube R16C56, the metallographic examination results for Tube R70C98 indicated that the extent of degradation was significantly under predicted by the eddy current examination The inspector noted the following additional information during review of ABB Combustion Engineering Report A PENG TR-Oll:
| | Executive Vice President |
| * All cracking in the two tubes initiated from the outside surface of the tubes, which is consistent with the degradation observed in previous tube pull | | & Chief Operating Officer |
| , | | , |
| sample * Cracking was found to be present only at the points on the tube where the carbon steel eggcrate strips contacted or were closest to the tub * The carbon content of both Tube R16C56 and Tube R70C98 exceeded the 0.05 weight percent maximum of the tubing specification (i.e.,0.059 percent and 0.055 percent, respectively).
| | Entergy Operations, Inc. |
|
| |
|
| .
| | ' |
| . | | P.O. Box 31995 L Jackson, Mississippi 39286-1995 |
| 19-
| |
| * The microstructure of both tubes was fine grained (ASTM 7-11), with the distril ution of carbides predominantly intragranular and little present on the grairi joundaries. Relatively high yield strength values were measured for the h Des (Tube R16C56,55 ksi; R70C98,53 ksi). The observed micra ;tructure and yield strength properties suggested that the final anneal of the tubes was performed at a low temperature (i.e.,less than 1800 *F).
| |
|
| |
|
| This inconel 600 microstructure normally results in enhanced susceptibility to intergranular stress corrosion cracking, compared to a coarser grain structure with predominantly intergranular carbides presen * Significant quantities of sulfur, as sulfate, were present on the fracture and outside diameter surfaces (i.e., fracture, up to 4.1 atomic percent; outside diameter, up to 1.6 atomic percent). The inspector concurred with the report assessment that the presence of such sulfur quantities suggested its involvement in the degradation process. It was unclear to the inspector on the intent of a report statement that conclusive evidence was lacking regarding whether the sulfur compounds accelerated the corrosion proces * Lead was also found associated with the fracture and outside diameter surfaces (i.e., fracture, up to 0.9 atomic percent; outside diameter, up to 0.25 atomic percent). The report made the same conclusion for the role of lead, as it did for sulfur compounds, regarding the lack of conclusive evidence regarding whether the lead accelerated the corrosion process. The inspector considered the quantities of lead present to be significantly higher than had been seen on other tube pull samples, and concluded that its presence was a fairly strong indicator of it being a contributor to the degradation process. No insights were developed from discussions with licensee personnel regarding the origins of the lead associated with the cracking in the steam generator tube * Scanning electron microscopy of intergranular cracking in Tube R70C98 showed evidence of pitting or erosion on grain faces at tube depths at least up to mid wall. The grain f acets near the crack tip did not show similar features. The inspector considered that the grain pitting was indicative of a crack that had been in existence for a considerable period.The inspector also viewed the metallurgical information as indicating thi: Nbe through-wall degradation in excess of tube mid wall was necessary at eggerate support locations prior to bobbin coil tube examinations being capable of its detectio * Burst testing of tube axial fiaws, which initiated at the 01H eggcrate support in Tube R16C56, developed a 0.8 inch long crack at the 340 location at a pressure of 2500 psig, which progressed to a 1.30 inch long fishmouth failure at the burst pressure of 3200 psig. The burst failure exceeded the main steam line break accident analysis pressure of 2950 psig, but did not meet the Regulatory Guide 1.121 structural integrity criteria of 4750 psi __-_______ _ __ _ _ _
| |
| .. . .. ..
| |
| ,
| |
| *
| |
| i
| |
| .
| |
| -20-Burst testing of tube axial flaws, which initiated at the 01H eggerate support in Tube R70C98, developed a 0.7 inch long crack at the 190 location at a pressure of 2000 psig. The crack was essentially through wall after pressurization to 2950 psig, with the crack held together by nine ligaments of non-corroded material. The tube burst at 3250 psig, with two axial cracks present, which had a total length of 1.65 inche Qgnglusion_g The micrc structure present in both Tubes R16C56 and R70C98 gi.e., fine grain size and intratranular carbides) was characteristic of that resulting from use of a low final annealing temperature. This type nf microstructure normally results in enhanced susceptibility to intergranular stress corrosion cracking, compared to a coarser grain structure with predominantly intergranular carbides present. Lead and sulfur (as sulfate) were present on the crack surfaces, indicating their involvement in the degradation proces *
| |
| Pitting or erosion of grain faces was observed at tube depths at least up to mid wall in Tube R70C98. The grain pitting was indicative of a crack that had been in existence for a considerable perio The metallurgical information indicated that tube through-wall degradation in excess of tube mid-wall was necessary at eggerate support locations prior to bobbin coil tube examinations being capable of its detectio Burst test values of 3200 psig and 3250 psig, respectively, were obtained for Tubes R16C56 and R70C98. These values exceeded the main steam line break accident analysis pressure of 2950 psig, but eu not meet the Regulatory Guide 1.121 structural integrity criteria of 4750 psi E8.2 Licensee Event Report 50-313/97-001: Inconsistencies Between Unit 1 Destructive Examination Results and Eddv Current Deoth Sizina Predictions (for Tube Flaws in Three Pulled Tube Samoles From the Uoper Tube Sheet Reaion) Created the Possibility of Flaws Beinn Present in Inservice Tubes Which Exceeded the Throuah-Wall Limit in Technical Specifica_1. ions The results of inspection followup of this licensee event report are documented in Section E1.1 abov E8.3 LClosedl Licensee Event Reoort 50-368/97-001: Steam Generator Tube Surveillances Were Performed Durina Refuelina Outaae 2R11 (September-November 1995) Which Failed to identifv Existina Tube Flaws for Further Evaluation as Reovired by the Steam Geerator Tube Surveillance Procram This licensee event report pertained to the identification during Outage 2F96-1 (November December 1996) that Refueling Outage 2R11 (September-November i
| |
| l | | l |
| , . _ , . ..
| |
| _ _ - _ _ _ _ _ _ _ . _ - _ _ . _ _ _ _ _ _ . _
| |
|
| |
| _
| |
| o
| |
| .
| |
| -21-1995) analysis of bobbin coil examination results for Steam Generator A Tubes R16C56 and R70C98 had f ailed to identify the presence of distorted support indications, if identified, the surveillance program requirements would have required the performance of a motorized rotating pancake coil examination to f acilitate determination of whether the distorted support indications were indicative of the presence of an axial tube flaw. The circumstances pertaining to Steam Generator Tube R16C56 were evaluated and documented in NRC Inspection Report 50 313;368/96-28. A noncited violation was identified as a result of this evaluation. The inspector concluded that the identification of a second
| |
| -
| |
| tube with a missed distorted support indication would not have changed the determination to issue a noncited violatio V. Mananement Meetinas X1 Exit Meeting Summary The inspector presented the results of the onsite inspection to members of licensee management at the conclusion of the onsite inspection on June 20,1997. The licensee acknowledged the findings presented. Documents were reviewed during the inspection, which had been identified by ABB-Combustion Engineering as containing proprietary information. No information was included in the inspection report that was considered proprietary. Additional in-office review of the inspection findings was performed subsequent to the onsite inspection, with evaluations continuing until August 7,1997. A second exit meeting was conducted telephonically on August 8,1997, to inform the licensee that three apparent violations were identified during the additional revie '
| |
| -
| |
| ,
| |
| .s
| |
| ..
| |
|
| |
|
| 1 ATTACHMENT 1 SUPPLEMENTAL INFORMATION! -
| | r^- 1 |
| ~
| |
| PARTIAL LIST OF PERSONS CO_NTACTED
| |
| -
| |
| _
| |
| - . Licensee C. Anderson, Unit 2 Plant Manager G. Ashley, Supervisor, Licensing M. Cooper, Licensing Specialist
| |
| - D. Denton, Director, Support j D. Harrison, Supervisor, Engineering Programs
| |
| - R. Hutchinson, Vice President, Nuclear Operations B.-James, Unit 2 Outage Manager i R; Jones, Engineer, Engineering Programs R. Lane, Director, Design Engineering
| |
| = D.'Meatheany, Engineer, Engineering Programs W. McKelvy, Chemistry Superintendent ,
| |
| D. Mims,' Director, Nuclear Support S. Pyle,- Licensing Specialist .
| |
| R. Scheide,' Licensing Specialist j M. Smith, Manager, Engineering Programs C. Zimmerman, Unit 1 Plant Manager NBC
| |
| - K. Kennedy, Senior Resident inspector -
| |
| S. Burton, Resident inspector J. Melfi, Resident inspector j
| |
| .,
| |
| ' INSPECTION PROCEDURES USED IP 50002 Steam Generators | |
| '
| |
| ITEMS OPENED AND CLOSED Ooened ,
| |
| '50-313/9714 01 APV Apparent failure to comply with requirements of criteria used -
| |
| to qualify bobbin coil eddy current examination sizingi 1 _ technique
| |
| _
| |
| 50-368/9714-0 . APV.:' Apparent failure during Outage 2F961 to remove tubes fro service, which contained flaws that exceeded Technical-
| |
| :. Specification plugging or repair limit = j
| |
| .i r
| |
| j
| |
| . ?
| |
| .
| |
|
| |
|
| ______
| | .. |
| e l'
| | Entergy Operations, Inc. -4- |
| I
| | . |
| . | | Vice President Operations Support Entergy Operations, Inc. |
| | |
| 50-368/9714-03 APV Lack of prompt corrective action in Novembu 199C 7rior to '
| |
| retuming potentially defective sleeved tubes to sers ee 50 368/9714 04 IFl Review of examination provisions for two sleeved tubes with identified potential for limited service life before initiation of primary water stress corrosion cracking C191251 50-368/9628-03 IFl Review of laboratory examination results for Outage 2F96-1 tube samples 50 313/97-001 LER inconsistencies between Unit 1 destructive examination results and eddy current depth sizing predictions for tube flaws in upper tube sheet region 50-368/97-001 LER Refueling Outage 2R11 steam generator tube surveillances failed to identify presence in two tubes of distorted support indications, thus precluding further evaluation (motorized rotating pancake coil) required by the surveillance program DOCUMENTS REVIEWFD Procedures / Documents HES-14, " Steam Generator integrity Program Unit 2," Revision 1 HES-26, " Steam Generator Tubo Plug Defect Management," Revision 7 HES-28, "ANO 2 Str.m Generator ECT Data Analysis Guidelines," Revision 6 and Standard Change Notice (s) 01 through 04 HES-29, "ANO-2 Steam Generator ECT Performance Demonstration Guidelines," Revision 3 HES 31, "ANO 2 ECT Data Management and Control," Revision 0 HES 42, " Steam Generator Tube Sleeve Management," Revision 1 ANO-2-OTH-ESP-SGMAN, " Arkansas Nuclear One - Unit 2 Steam Generator Eddy Current Training Manual," Revision 1 Acquisition Technique Specification Sheet ACTS 1, " Bobbin Examination," Revision 1 Acquisition Technique Specification Sheet ACTS 2, "O.115 Pancake, Axial and Cir Directed Coils," Revision 4 i
| |
|
| |
|
| o
| | P.O. Box 31995 Jackson, Mississippl 39286-Manager, Washington Nuclear Operations ABB Combustion Engineering Nuclear |
| | ! |
| | Power 12300 Twinbrook Parkway, Suite 330 < |
| | Rockville, Maryland 20852 County Judge of Pope County Pope County Courthouse Russellville, Arkansas : 72801 Winston & Strawn - |
| | 1400 L Street, N.W. |
|
| |
|
| | WasNngton, D.C. 20005 3502 David D. Snellings, Jr., Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street, Mail Slot 30 Little Rock, Arkansas 72205-3867 Manager Rockville Nuclear Licensing ' |
| | Framatome Technologies 1700 Rockville Pike, Suite 525 Rockville, Maryland 20852 |
| . | | . |
| 3-Acquisition Technique Specification Sheet ACTS 3, "Plus-Point for Welded Sleeves,"
| | - ,- -- |
| Revision 5 Acquisition Technique Specification Sheet ACTS 4, *
| | , |
| Analysis Technique Specification Sheet ANTS 1, " Bobbin Coil Examination of Steam Generator Tubing," Revision 1 Analysis Technique Specification Sheet ANTS 2, "3 Coil RPC (0.115 Pancake, Axial &
| |
| Circ. Coils)," Revision 1 Analysis Technique Specification Sheet ANTS 3, "Plus Point for Kinetic Sleeves,"
| |
| Revision 1 Analysis Technique Specification Sheet ANTS 4, " Single Coil O.115 Pancake," Revision 3 5120.501, " Steam Generator integrity Program - Unit 2," Revision 6 5120.506, "C.E. Unit 2 Steam Generator Secondary Side inspection 2E24 A&B,"
| |
| Revision O ABB Combustion Engineering Report A-PENG-TR-Oll, " Examination of Steam Generator Tubes Removed From ANO-2 in 1997," June 1997 Procedure STD-400-157, " Visual Examination of Steam Generator Tube Sleeve and Sleeve Plug Welds," Revision O Procedure 00000-NSS 062, " Procedure for Ultrasonic Examination of Steam Generator Tube to Sleeve Upper Welds, Revision 10 Procedure STD 400-151, "Linearity Verification of Ultrasonic Instrumentation Used for the Examination of Sleeved Tube Welds, Revision 0 Procedure STD 400-158, " Technical Operating Procedure for the Transition Zone Sleeve Rolling Tool and Controls for Steam Generator Tubes With 3/3 Inch O.D. 043 Inch or .048 Inch Wall Thickness Using the Computer Controlled Sleeving System," Revision 0 Procedure STD-500 020, " Technical Operating Procedure for the Post-Weld Heat Treatment and Heater Dryout of Above the Tubesheet (ATS) Welds Using Resistance Heating for Steam Generator Tubes With .048 inch Wall Thickness," approved July 31, 1992 Traveler 2005331-001, " Installing Tube Sleeves in the Steam Generator Tubesheet Expansion Transition Zone Using the Genesis 2000 Control System," Revision 0
| |
|
| |
|
| | . _ . _ _ _ _ .- - . _ _ . -. _ _ _ . . |
| | , ., |
| | 9-Entergy OperatPpW inc, . -5- |
| | . |
| | bec w/ Enclosure (s): - |
| | PDR-LPDR hNUDOCS [ |
| | SECY EC's: Rl,Ril,Rlli CA PA (Os2G4) |
| | EDO (017G21) OiG (T 5D28) |
| | DEDO (0-17G21) OE (0 7H5) |
| | OE:EAFile (0 7H5) 01 (0 3E4) |
| | ^ |
| _ | | _ |
| . | | OGC (0-15818) OGC (015B18) |
| C
| | NRR (012G18) NRR/ADP (012G18) |
| .
| | NRR PROJECT MANAGER OC/DAF (T-9E10) |
| | | OC/LFDCS (T 9E10) AEOD (T-4D18) |
| . | | RA Reading File GSanborn-EAFile RWise (if app.) RWise-DOL File (if app.) |
|
| |
|
| Welding Procedure Specification GTAA-43.43.992, Revision 0, and its supporting procedure qualification record, GTAA-43.43 01503, Revision 1
| | RIV Files MIS Coordinator ECollins, RIV E Mall DlHTRIBUTION: |
| | | OEMAIL: JDyer (JED2) |
| C e,
| | TPGwynn (TPG) WBrown (WLB) |
| .
| | GSanborn (GFS) GMVasquez (GMV) |
| o ATTACHMENT 2 Entergy Operations Arkansas Nuclear One Unit Two Position Paper on Steam Generator " Missed" indications in the Fall of 1996, ANO 2 shutdown due to a steam generator tube leak. During the subsequent inspection, tubes were identified as having indications previously " missed" during the eddy current testing (ECT) analysis process in the prior outage. The definition of " missed" is difficult and presents a dilemma when one must make a reasonable determination of the condition of a tube in previous times. The tube which caused the t aker outage was not due to a " missed" call and thus the leaker was not preventable. An a 1 was initiated, and a non-cited violation was issued as a result of the " missed" calls, Because of the experience of the leaker outage, several changes were implemented as part of the ANO corrective actions from the root cause analysis which was performed. These changes included:
| | BHenderson (BWH) MHammond (MFH2) |
| Entergy Level 111 Review Team Repair of all Eggcrate Cracks RPC of allindications Tertiary review using Computerized Data Screening Review of a sample of NDD calls by Resolution Analysts or Entergy Review Team Enhanced Analyst Site Performance Testing These changes resulted in significant improvements to the ECT Program and a high quality inspection in 2R12, our most recent outage. The changes were made by ANO as a result of the corrective actions from 2F96 based on the performance of the inspection in 2R1 In 2R12, none of the indications found f ailed the Regulatory Guide 1.121 requirements. All structural and leakage requirements were met. A review of the 2R12 indications disclosed a number of examples where there is some evidence of the flaw existing in the previous inspection. Review of flaw history is routinely performed to evaluate the growth rate of the degradation. The growth ra;c is used to make a "forwardlooking" evaluation, or operational assessment, of the unit and determine the safe operating interval. It is common for flaws to have not been called in the previous inspection, especially when the
| | CHackney (CAH) DKunihiro (DMK1) |
| " signal noise" ratio is small. It is a very qualitative assessment that determines if a given flaw was " missed," or if it was reasonable to assume the ECT signal was too difficult to have been expected to be calle The 2F96 indications of interest are for the most part " questionable" as to whether or not they should have been called. There were none similar to the tubes pulled in 2F96, which were more obvious misses. The process used in 2F96 was much improved over 2R1 There were, however, no real program changes made to increase analyst sensitivity, other than the knowledge of the cause of the leaker. Shortly after the inspection began, the ECT results for the leaking tube were reported to the analysts, and insight gained from that information improved the quality of the analysi The flaws detected in 2R12 were typically small amplitude, and posed no structural or leakage concerns. This was demonstrated by in situ pressure testing of the largest flaws, with all exceeding the RG 1.121 requirements. The indications detected in 2R12 that have been classed as " missed" are really not indicative of a " missed" indication. The changes
| | Art Howell (ATH) DChamberlain (DDC) |
| | | KPerkins (KEP) ECollins (EEC) |
| e e
| | SENIOR RESIDENT- KBrockman (KEB) |
| .
| | EMerschoff (EWM) |
| e implemented in 2R12 provided for a drastic improvement in the quality of the inspectio That is evident by the results of the independent review of 10% of all NDD calls, which indicated that the analysts were making conservative calls, and were not " Missing" indications. This, along with the other changes implemented, yielded a high quality inspection. These changes were made following the review of the 2F96 circumstances, and this is the first opportunity we have had to implement them. A non cited violation was issued for the 2R11 " missed" indications identified in 2F96, and 2R12 was the first inspection af ter implementing corrective action During 2F96, two tubes were removed for analysis that contained axial cracks at the eggerates. The analysis provided information on the structural integrity of this type of flaw. The flaws burst at pressures above accident conditions but below Reg. Guide limits. It was at this time that ANO decided to remove all confirmed crack-like indications in the eggcrate region and make programmatic changes to enhance the probability of detection during 2R12. Tubes tested during 2R12 indicated the structural b'.egrity of the tubing is not challenged. This was proven by in situ testing the largest 11aws. A total of six tubes were tested and all had zero leakage at peak accident pressures and all were pressurized to 3 times the normal delta pressure limit specified in Reg. Guide 1.121. It is therefore believed that none of the 2R12 indications in question would have prevented the SGs from fulfilling their safety function, nor challenge the steam generator structural integrit A critical review of the ANO-2 program occurred after 2F96 and a comprehensive corrective action plan was implemented prior to 2R12. This corrective action plan was presented to the NRC during the March 1997 meeting. The benefits from those corrective actions were apparent in the number of indications identified, in summary, based on the following information:
| | DOCUMENT NAME: G:\EA\ DRAFT \EA97382.DFT To receive, copy opument, Indicats in box: "C" = Copy wthout oncesures "E" = Copy with enclosures *N" = No copy |
| A non-cited violation was issued following 2F96 Corrective actions were developed prior to 2R12 Tube pull destructive examination results were available just prior to 2R12 Comprehensive corrective actions administered during 2R12 Largest flaws passed Reg. Guide 1.121 with zero leakage at peak accident pressures ANO does not believe the recent review of 2F96 data identified tubes that were improperly dispositioned and should have been plugged during 2F96. Also, no additional corrective actions would be necessary if a violation was issued for the condition identified during 2R12.
| | ~ ~ |
| | ES //7AV EO O TA ,,,p DRP() , D:DRS@0 , |
| | MVasquez GSapid IBarnes f/'D EC611TfiP AHovTell ['' |
| | 11k//97 11/d/97 114r/97 _ 11/'//97 11/4/97 |
| | -_ |
| | RC f4 DRA RA g/ |
| | W8rown R '' JDyer @ EMy.f56tioff 11/G/97 11/f//97 11//fl97 OFFIC!AL RECORD COPY |
| }} | | }} |
|
---|
Category:CORRESPONDENCE-LETTERS
MONTHYEAR1CAN109906, Forwards Framatome Technologies,Inc non-proprietary TR BAW-10235P, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheet of Once-Through Sgs, Rev 11999-10-19019 October 1999 Forwards Framatome Technologies,Inc non-proprietary TR BAW-10235P, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheet of Once-Through Sgs, Rev 1 ML20217J4971999-10-18018 October 1999 Requests Addl Info Re Results of Util Most Recent Steam Generator Insp at ANO-2 & Util Methodology Used to Predict Future Performance of SG Tubes ML20217J3871999-10-15015 October 1999 Informs That Topical Rept BAW-10235P, Management Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through SG, Rev 0 Marked as Proprietary Will Be Withheld from Public Disclosure 2CAN109902, Submits Withdrawal of Code Case N-593 for ANO-2 Replacement SGs1999-10-15015 October 1999 Submits Withdrawal of Code Case N-593 for ANO-2 Replacement SGs ML20217J3601999-10-15015 October 1999 Informs That Topical Rept BAW-10235P, Management Program for Volumetric Outer Diameter Integranular Attack in Tubesheets of Once-Through SG, Rev 1 Marked as Proprietary Will Be Withheld from Public Disclosure 2CAN109903, Forwards Response to RAI Re Proposed Tech Specs Change for Special SG Insp1999-10-14014 October 1999 Forwards Response to RAI Re Proposed Tech Specs Change for Special SG Insp ML20217D1721999-10-0808 October 1999 Forwards RAI Re 990729 Request for Amend to TSs Allowing Special SG Insp for Plant,Unit 2.Questions Re Proposed Insp Scope for Axial Cracking Degradation in Eggcrate Support Region Submitted.Response Requested by 991015 1CAN109905, Discusses Insp of Once Through SG Tubing Surveillance Performed During 1R15 Scheduled RFO on 990910.Category C-3 Results,Included1999-10-0404 October 1999 Discusses Insp of Once Through SG Tubing Surveillance Performed During 1R15 Scheduled RFO on 990910.Category C-3 Results,Included ML20212L0621999-10-0101 October 1999 Forwards Safety Evaluation & Exemption from Certain Requirements of 10CFR50,App R,Section III.G.2, Fire Protection of Safe Shutdown Capability 1CAN099908, Withdraws 990919 Exigent TS Change Request to Allow Continued Installation of re-rolls for One Cycle of Operation Through End of Cycle 16 in Conjunction with Addl Insp Criteria1999-09-30030 September 1999 Withdraws 990919 Exigent TS Change Request to Allow Continued Installation of re-rolls for One Cycle of Operation Through End of Cycle 16 in Conjunction with Addl Insp Criteria 2CAN099902, Requests That NRC Assign CENPD-132,Suppl 4-P, Calculative Methods for Abb Cenp Large Break LOCA Evaluation Model, Review Priority So That Approval Will Be Granted No Later than Oct 31,20001999-09-29029 September 1999 Requests That NRC Assign CENPD-132,Suppl 4-P, Calculative Methods for Abb Cenp Large Break LOCA Evaluation Model, Review Priority So That Approval Will Be Granted No Later than Oct 31,2000 1CAN099903, Forwards Rev 0 to COLR for ANO-1 Cycle 16, IAW TS 6.12.31999-09-27027 September 1999 Forwards Rev 0 to COLR for ANO-1 Cycle 16, IAW TS 6.12.3 1CAN099907, Requests That Alternative Be Allowed in Accordance with 10CFR50.55a(a)(3)(i) & (II) as Discussed in Encl 1.Encl 2 & 3 Stress Analysis & Flaw Evaluation Summaries Ref in Encl Alternative1999-09-26026 September 1999 Requests That Alternative Be Allowed in Accordance with 10CFR50.55a(a)(3)(i) & (II) as Discussed in Encl 1.Encl 2 & 3 Stress Analysis & Flaw Evaluation Summaries Ref in Encl Alternative 1CAN099906, Forwards 1R15 Growth Data Obtained & Analyzed Through 990922 & Includes Plus Point Voltages,Axial Extent & Circumferential Extent Patches,As Well as Preliminary Growth Conclusions Based on Analysis of Data1999-09-24024 September 1999 Forwards 1R15 Growth Data Obtained & Analyzed Through 990922 & Includes Plus Point Voltages,Axial Extent & Circumferential Extent Patches,As Well as Preliminary Growth Conclusions Based on Analysis of Data 2CAN099901, Informs That G Kendrick,License SOP-43658,no Longer Has Need to Maintain Operating License on Ano,Unit 2.Entergy Requests That License for Individual Be Withdrawn,Due to Resignation, Effective 9908271999-09-24024 September 1999 Informs That G Kendrick,License SOP-43658,no Longer Has Need to Maintain Operating License on Ano,Unit 2.Entergy Requests That License for Individual Be Withdrawn,Due to Resignation, Effective 990827 2CAN099904, Forwards Ano,Unit 2 10CFR50.59 Rept for Time Period Ending 990225.Rept Contains Brief Description of Changes in Procedures & in Facility as Described in Sar,Tests & Experiments Conducted & Other Changes to SAR1999-09-23023 September 1999 Forwards Ano,Unit 2 10CFR50.59 Rept for Time Period Ending 990225.Rept Contains Brief Description of Changes in Procedures & in Facility as Described in Sar,Tests & Experiments Conducted & Other Changes to SAR ML20212F5031999-09-22022 September 1999 Forwards SER Granting Relief Requests 1-98-001 & 1-98-002 Which Would Require Design Mods to Comply with Code Requirements,Which Would Impose Significant Burden Pursuant to 10CFR50.55a(g)(6)(i) 1CAN099905, Submits Supplemental Info in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria.Proposed TS Rev & Info Related to Use of Alternate Repair Discussed in Attachments1999-09-17017 September 1999 Submits Supplemental Info in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria.Proposed TS Rev & Info Related to Use of Alternate Repair Discussed in Attachments ML20212D9961999-09-16016 September 1999 Informs That on 990818,NRC Completed Midcycle PPR of Arkansas Nuclear One.Nrc Plan to Conduct Core Insps at Facility Over Next 7 Months.Details of Insp Plan Through March 2000 Encl 1CAN099902, Forwards Proprietary Rev 1 to Topical Rept BAW-10235P, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs, in Response to 990831 Rai.Proprietary Encl Withheld1999-09-15015 September 1999 Forwards Proprietary Rev 1 to Topical Rept BAW-10235P, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs, in Response to 990831 Rai.Proprietary Encl Withheld 2CAN099905, Informs That Jk Caery,License OP-42589 & as Bates,License OP-42506,no Longer Need to Maintain Operating License at Ano,Unit 2.Withdrawal of Licenses Is Requested1999-09-0909 September 1999 Informs That Jk Caery,License OP-42589 & as Bates,License OP-42506,no Longer Need to Maintain Operating License at Ano,Unit 2.Withdrawal of Licenses Is Requested 1CAN099901, Forwards Responses to 990831 RAI Containing follow-up Questions Discussed on 990823-26,in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria. Revs to Proposed TSs Included in Attachments1999-09-0707 September 1999 Forwards Responses to 990831 RAI Containing follow-up Questions Discussed on 990823-26,in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria. Revs to Proposed TSs Included in Attachments ML20211P4121999-09-0707 September 1999 Requests NRC Staff Review & Approval of Integrated Nuclear Security Plan (Insp) & Integrated Security Training & Qualification Plan (Ist&Q), for Use by All Entergy Operations,Inc.Encl Withheld,Per 10CFR2.790(d) 0CAN099906, Forwards Comments on Ano,Units 1 & 2 Specific Reactor Vessel Info Contained in Rvid,Version 2,in Response to NRC 990708 & 0715 Ltrs1999-09-0101 September 1999 Forwards Comments on Ano,Units 1 & 2 Specific Reactor Vessel Info Contained in Rvid,Version 2,in Response to NRC 990708 & 0715 Ltrs ML20211L4901999-09-0101 September 1999 Forwards Insp Repts 50-313/99-12 & 50-368/99-12 on 990711- 0821.No Violations Noted ML20211J2351999-08-31031 August 1999 Forwards Request for Addl Info Re SG Outer Diameter Intergranular Attack Alternate Repair Criteria for Plant, Unit 1 ML20211E6161999-08-25025 August 1999 Forwards Amend 15 to ANO Unit 2,USAR,per 10CFR50.71(e) & 10CFR50.4(b)(6).Summary of 10CFR50.59 Evaluations Associated with Amend 15 of ANO Unit 2 SAR Will Be Provided Under Separate Cover Ltr with 30 Days 0CAN089905, Forwards Arkansas Nuclear One Units 1 & 2 FFD Program Performance Data for Period Jan-June 19991999-08-25025 August 1999 Forwards Arkansas Nuclear One Units 1 & 2 FFD Program Performance Data for Period Jan-June 1999 ML20211F4181999-08-25025 August 1999 Forwards SE Accepting Licensee 980603 & 990517 Requests for Approval of risk-informed Alternative to 1992 Edition of ASME BPV Code Section Xi,Insp Requirements for Class 1, Category B-J Piping Welds ML20211G0731999-08-19019 August 1999 Forwards Applications for Renewal of Operating License for Kw Canitz & Aj South.Without Encls 1CAN089904, Forwards Addl Info in Support of SG Tube End Cracking Alternate Repair Criteria,In Response to NRC 990728 Rai. Proposed TS Changes Encl1999-08-19019 August 1999 Forwards Addl Info in Support of SG Tube End Cracking Alternate Repair Criteria,In Response to NRC 990728 Rai. Proposed TS Changes Encl ML20211C5101999-08-19019 August 1999 Forwards Certified Copies of Liability Insurance Policy Endorsements Issued in First Half of 1999 for Each Entergy Operations,Inc Nuclear Unit,Per 10CFR140.15 ML20211A9501999-08-12012 August 1999 Discusses 990720-21 Workshop Conducted in Region IV Ofc,Re Exchange of Info in Area of Use of Risk Insights in Regulatory Activities.List of Attendees,Summary of Topic & Issues,Agenda & Copies of Handouts Encl 0CAN089903, Submits Addl Response to NRC Second RAI Re GL 95-07, Pressure Locking & Thermal Binding of Gate Valves1999-08-12012 August 1999 Submits Addl Response to NRC Second RAI Re GL 95-07, Pressure Locking & Thermal Binding of Gate Valves IR 05000368/19990111999-08-12012 August 1999 Forwards Insp Repts 50-313//99-11 & 50-368/99-11 on 990719-23.No Violations Noted.Insp Focused on Review of Licensed Operator Requalification Program & Observation of Requalification Exam Activities at Unit 1 2CAN089901, Forwards Description of Planned Scope & Expansion Criteria for Special SG Tube Insp,In Support of Proposed ANO-2 TS Amend for 2P99 Special SG Insp Submitted on 9907291999-08-0606 August 1999 Forwards Description of Planned Scope & Expansion Criteria for Special SG Tube Insp,In Support of Proposed ANO-2 TS Amend for 2P99 Special SG Insp Submitted on 990729 1CAN089902, Requests NRC Input on Encl Proposed Draft Format for ANO-1 License Renewal Application,Which Will Provide Option to Continue Operating Plant for Addl Twenty Years Beyond End of Current Operating License1999-08-0505 August 1999 Requests NRC Input on Encl Proposed Draft Format for ANO-1 License Renewal Application,Which Will Provide Option to Continue Operating Plant for Addl Twenty Years Beyond End of Current Operating License 2CAN089902, Informs That Tl Russell,License SOP-43587-1 & Jk Fancher, License OP-42300-1,no Longer Have Need to Maintain Operating License at ANO-2.Withdrawal of Licenses Requested1999-08-0404 August 1999 Informs That Tl Russell,License SOP-43587-1 & Jk Fancher, License OP-42300-1,no Longer Have Need to Maintain Operating License at ANO-2.Withdrawal of Licenses Requested ML20210L1461999-08-0303 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Requests Submittal of Ltr Identifying Individuals Taking Exam,Personnel Allowed Access to Exams & Mailing Address for Exams 0CAN089902, Submits 60 Day Response to GL 99-02, Laboratory Testing of Nuclear Grade Activated Charcoal. Proposed Actions That Will Be Taken on ANO Unit 1 RB Purge Filtration Sys & Unit 2 Containment Purge & Exhaust Sys,Clarified1999-08-0202 August 1999 Submits 60 Day Response to GL 99-02, Laboratory Testing of Nuclear Grade Activated Charcoal. Proposed Actions That Will Be Taken on ANO Unit 1 RB Purge Filtration Sys & Unit 2 Containment Purge & Exhaust Sys,Clarified 0CAN089901, Forwards Info Re Estimate of licensee-originated Licensing Actions for ANO-1 & ANO-2,in Response to Administrative Ltr 99-02,dtd 9906031999-08-0202 August 1999 Forwards Info Re Estimate of licensee-originated Licensing Actions for ANO-1 & ANO-2,in Response to Administrative Ltr 99-02,dtd 990603 ML20210L3581999-07-29029 July 1999 Ltr Contract,Task Order 43, Arkansas Nuclear One Safety System Engineering Insp (Ssei), Under Contract NRC-03-98-021 1CAN079903, Forwards non-proprietary Addendum to Rev 0 of Topical Rept BAW-2346P,in Support of Proposed TS Changes Revising SG Tubing Surveillance Requirements to Provide Alternate Repair Criteria for Tube End Cracks1999-07-29029 July 1999 Forwards non-proprietary Addendum to Rev 0 of Topical Rept BAW-2346P,in Support of Proposed TS Changes Revising SG Tubing Surveillance Requirements to Provide Alternate Repair Criteria for Tube End Cracks ML20216D8131999-07-28028 July 1999 Forwards Request for Addl Info Re SG Tube End Cracking Alternate Repair Criteria for Plant,Unit 1 ML20216D3561999-07-23023 July 1999 Discusses non-cited Violation Identified in Insp Rept 50-313/98-21,involving Failure to Have Acceptable Alternative Shutdown Capability for ANO-1 ML20210C2191999-07-21021 July 1999 Forwards Insp Repts 50-313/99-08 & 50-368/99-08 on 990530-0710 at Arkansas Nuclear One,Units 1 & 2,reactor Facility.No Violations Noted.Conduct of Activities at Plant Generally Characterized by safety-conscious Operations ML20209H5251999-07-15015 July 1999 Informs That as Result of NRC Review of Licensee 980701 & 990311 Responses to GL 92-01,rev 1 & Suppl 1 & Suppl 1 RAI, Staff Revised Info in Reactor Vessel Integrity Database & Releasing Database as Rvid Version 2 1CAN079901, Forwards Proposed Changes to Current Util 990409 Submittal Re Rev to RB Structural Integrity Requirements Contained in Plant Ts.Proposed Revs Affect ACs & Applicable Bases Re ISI Reporting for Containment Structures,Tendons & Anchorages1999-07-14014 July 1999 Forwards Proposed Changes to Current Util 990409 Submittal Re Rev to RB Structural Integrity Requirements Contained in Plant Ts.Proposed Revs Affect ACs & Applicable Bases Re ISI Reporting for Containment Structures,Tendons & Anchorages 0CAN079902, Responds to NRC Telcon RAI Re Proposed Administrative Controls TS Changes.Revised TS Pages Which Replaces Pages Previously Provided in 981124 Submittal,Encl1999-07-14014 July 1999 Responds to NRC Telcon RAI Re Proposed Administrative Controls TS Changes.Revised TS Pages Which Replaces Pages Previously Provided in 981124 Submittal,Encl ML20209E5551999-07-0808 July 1999 Informs That as Result of NRC Review of Util Responses to GL 92-01,rev 1,suppl 1,staff Revised Info in Rv Integrity Database & Releasing Database as Rvid Version 2 1999-09-09
[Table view] Category:OUTGOING CORRESPONDENCE
MONTHYEARML20217J4971999-10-18018 October 1999 Requests Addl Info Re Results of Util Most Recent Steam Generator Insp at ANO-2 & Util Methodology Used to Predict Future Performance of SG Tubes ML20217J3871999-10-15015 October 1999 Informs That Topical Rept BAW-10235P, Management Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through SG, Rev 0 Marked as Proprietary Will Be Withheld from Public Disclosure ML20217J3601999-10-15015 October 1999 Informs That Topical Rept BAW-10235P, Management Program for Volumetric Outer Diameter Integranular Attack in Tubesheets of Once-Through SG, Rev 1 Marked as Proprietary Will Be Withheld from Public Disclosure ML20217D1721999-10-0808 October 1999 Forwards RAI Re 990729 Request for Amend to TSs Allowing Special SG Insp for Plant,Unit 2.Questions Re Proposed Insp Scope for Axial Cracking Degradation in Eggcrate Support Region Submitted.Response Requested by 991015 ML20212L0621999-10-0101 October 1999 Forwards Safety Evaluation & Exemption from Certain Requirements of 10CFR50,App R,Section III.G.2, Fire Protection of Safe Shutdown Capability ML20212F5031999-09-22022 September 1999 Forwards SER Granting Relief Requests 1-98-001 & 1-98-002 Which Would Require Design Mods to Comply with Code Requirements,Which Would Impose Significant Burden Pursuant to 10CFR50.55a(g)(6)(i) ML20212D9961999-09-16016 September 1999 Informs That on 990818,NRC Completed Midcycle PPR of Arkansas Nuclear One.Nrc Plan to Conduct Core Insps at Facility Over Next 7 Months.Details of Insp Plan Through March 2000 Encl ML20211L4901999-09-0101 September 1999 Forwards Insp Repts 50-313/99-12 & 50-368/99-12 on 990711- 0821.No Violations Noted ML20211J2351999-08-31031 August 1999 Forwards Request for Addl Info Re SG Outer Diameter Intergranular Attack Alternate Repair Criteria for Plant, Unit 1 ML20211F4181999-08-25025 August 1999 Forwards SE Accepting Licensee 980603 & 990517 Requests for Approval of risk-informed Alternative to 1992 Edition of ASME BPV Code Section Xi,Insp Requirements for Class 1, Category B-J Piping Welds IR 05000368/19990111999-08-12012 August 1999 Forwards Insp Repts 50-313//99-11 & 50-368/99-11 on 990719-23.No Violations Noted.Insp Focused on Review of Licensed Operator Requalification Program & Observation of Requalification Exam Activities at Unit 1 ML20211A9501999-08-12012 August 1999 Discusses 990720-21 Workshop Conducted in Region IV Ofc,Re Exchange of Info in Area of Use of Risk Insights in Regulatory Activities.List of Attendees,Summary of Topic & Issues,Agenda & Copies of Handouts Encl ML20210L1461999-08-0303 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Requests Submittal of Ltr Identifying Individuals Taking Exam,Personnel Allowed Access to Exams & Mailing Address for Exams ML20210L3581999-07-29029 July 1999 Ltr Contract,Task Order 43, Arkansas Nuclear One Safety System Engineering Insp (Ssei), Under Contract NRC-03-98-021 ML20216D8131999-07-28028 July 1999 Forwards Request for Addl Info Re SG Tube End Cracking Alternate Repair Criteria for Plant,Unit 1 ML20210C2191999-07-21021 July 1999 Forwards Insp Repts 50-313/99-08 & 50-368/99-08 on 990530-0710 at Arkansas Nuclear One,Units 1 & 2,reactor Facility.No Violations Noted.Conduct of Activities at Plant Generally Characterized by safety-conscious Operations ML20209H5251999-07-15015 July 1999 Informs That as Result of NRC Review of Licensee 980701 & 990311 Responses to GL 92-01,rev 1 & Suppl 1 & Suppl 1 RAI, Staff Revised Info in Reactor Vessel Integrity Database & Releasing Database as Rvid Version 2 ML20209E5551999-07-0808 July 1999 Informs That as Result of NRC Review of Util Responses to GL 92-01,rev 1,suppl 1,staff Revised Info in Rv Integrity Database & Releasing Database as Rvid Version 2 ML20209D8521999-07-0707 July 1999 Responds to Util 990706 Request That NRC Exercise Discretion Not to Enforce Compliance with Actions Required by TS 3.7.2, Auxiliary Electrical Sys. NOED Warranted & Approval Granted for Extension of Allowed Outage Time to 14 Days ML20209A8561999-06-25025 June 1999 Refers to Investigation Rept A4-1998-042 Re Potential Falsification of Training Record by Senior Licensed Operator at Arkansas Nuclear One Facility.Nrc Concluded That Training Attendance Record Falsified IR 05000313/19990071999-06-21021 June 1999 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-313/99-07 & 50-368/99-07 Issued on 990514.Adequacy of Min Staffing Levels May Be Reviewed During Future Insps ML20196D4241999-06-21021 June 1999 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp of License SOP-43716 Issued on 990325.Believes That NRC Concerns Have Been Adequately Addressed at Present ML20207H3551999-06-10010 June 1999 Forwards Insp Repts 50-313/99-05 & 50-368/99-05 on 990411-0529.No Violations Noted ML20195G3481999-06-0909 June 1999 Ack Receipt of ,Transmitting Changes to Facility Emergency Plan,Rev 25,under Provisions of 10CFR50,App E, Section V IR 05000313/19993011999-06-0909 June 1999 Discusses Arrangements for Administration of Licensing Exam During Wk of 991213,per Telcon of 990602.As Agreed,Exams Repts 50-313/99-301 & 50-368/99-301 Will Be Prepared Based on Guidelines in Rev 8 of NUREG-1021 ML20195F1631999-06-0808 June 1999 Forwards Insp Repts 50-313/99-06 & 50-368/99-06 on 990524-28.Violation Identified & Being Treated as Noncited Violation ML20207G3111999-06-0707 June 1999 Ack Receipt of Changes to ANO EP Implementing Prcoedure 1903.010,Emergency Action Level Classification,Rev 34 PC-2, Received on 981218,under 10CFR50,App E,Section V Provisions. No Violations Identified ML20207G7951999-06-0707 June 1999 Forwards Notice of Violation Re Investigation Rept A4-1998-042 Re Apparent Violation Involving Initialing Record to Indicate Attendance at Required Reactor Simulator Training Session Not Attended ML20207E7131999-06-0202 June 1999 Discusses EOI 990401 Request for Alternative to Requirements of Iwl for Arkansas Nuclear One,Pursuant to 10CFR50.55a(g)(6)(ii)(B) & ASME BPV Code Section XI & Forwards Safety Evaluation Accepting Proposed Alternative ML20207B9521999-05-26026 May 1999 Discusses GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment. Staff Will Conduct Limited Survey in to Identify Sampling ML20207B4171999-05-24024 May 1999 Forwards Corrected Cover Ltr to Insp Repts 50-313/99-07 & 50-368/99-07 Issued 990514 with Incorrect Insp Closing Date ML20207A7771999-05-24024 May 1999 Forwards Insp Repts 50-313/98-21 & 50-368/98-21 on 981116-990406.One Violation of NRC Requirements Occurred & Being Treated as Noncited Violation,Consistent with App C of Enforcement Policy ML20206U4541999-05-17017 May 1999 Discusses Util & Suppl Re Changes to License NPF-06,App a TSs Bases Section.Staff Offers No Objection to These Bases Changes.Affected Bases Pages,B 202, B 2-4,B 2-7,B 3/4 2-1,B 3/4 2-3 & B 3/4 6-4,encl ML20206S4721999-05-14014 May 1999 Forwards Insp Repts 50-313/99-07 & 50-368/99-07 on 990426- 30.No Violations Noted.However,Nrc Requests That Util Provide Evaluation of Licensee Provisions to Maintain Adequate Level of Response Force Personnel on-site ML20207B4271999-05-14014 May 1999 Corrected Ltr Forwarding Insp Repts 50-313/99-07 & 50-368/99-07 on 990426-30.No Violations Noted.Areas Examined During Insp Included Portions of Physical Security Program ML20206R4741999-05-13013 May 1999 Informs That Staff Reviewed Draft Operation Insp Rept for Farley Nuclear Station Cooling Water Pond Dam & Concurs with FERC Findings.Any Significant Changes Made Prior to Issuance of Final Rept Should Be Discussed with NRC ML20206N7011999-05-12012 May 1999 Informs That NRC Office of Nuclear Reactor Regulation Reorganized Effective 990328.As Part of Reorganization, Division of Licensing Project Management Created ML20206M7581999-05-11011 May 1999 Forwards SE Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 ML20206S1761999-05-11011 May 1999 Responds to Informing of Changes in Medical Condition & Recommending License Restriction for Senior Reactor Operator.No Change Was Determined in Current License Conditions for Individual ML20206N4161999-05-11011 May 1999 Discusses Completion of Licensing Action for GL 96-01, Testing of Safety-related Logic Circuits, for Plant,Units 1 & 2 ML20206S4211999-05-10010 May 1999 Forwards Insp Repts 50-313/99-04 & 50-368/99-04 on 990228- 0410.Four Violations of NRC Requirements Identified & Being Treated as Noncited Violations Consistent with App C of Enforcement Policy ML20206H1031999-05-0606 May 1999 Forwards Results of Gfes of Written Operator Licensing Exam, Administered on 990407,to Nominated Employees of Facility. Requests That Training Dept Forward Individual Answer Sheet & Results to Appropriate Individuals.Without Encl ML20206F0611999-04-29029 April 1999 Forwards SE Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205R6331999-04-20020 April 1999 Ack Receipt of Which Transmitted Rev 39 to ANO Industrial Security Plan,Submitted Under Provisions of 10CFR50.54(p).No NRC Approval Is Required,Since Util Determined Changes Do Not Decrease Effectiveness of Plan ML20205P4641999-04-15015 April 1999 Forwards for Review & Comment Draft Info Notice That Describes Unanticipated Reactor Water Draindown at Quad Cities Nuclear Power Station Unit 2,Arkansas Nuclear One Unit 2 & Ja Fitzpatrick NPP ML20205N7251999-04-13013 April 1999 Forwards Summary of 990408 Meeting with EOI in Jackson, Mississippi Re EOI Annual Performance Assessment of Facilities & Other Issues of Mutual Interest.List of Meeting Attendees & Licensee Presentation Slides Encl ML20205M6881999-04-12012 April 1999 Forwards Safety Evaluation on Second 10-year Interval Inservice Insp Request Relief 96-005 ML20205L7711999-04-0909 April 1999 Forwards Insp Repts 50-313/99-03 & 50-368/99-03 on 990202- 17.No Violations Noted ML20205K7681999-04-0606 April 1999 Forwards RAI Re risk-informed Alternative to Certain Requirements of ASME Code 11,table IWB-2500-1 ML20205G8871999-04-0202 April 1999 Forwards RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs, for Plant, Units 1 & 2.Response Requested within 60 Days of Date of Ltr 1999-09-22
[Table view] |
Text
. . .m . _. _
* ,
e t h' "8 004 UNIT ED ST ATES
,, t NUCLEAR REGULATORY COMMISSION e REGlo4 tv g
g 611 RYAN PLAZA DRIVE, SUITE 400
%. 4, * * * * * ,0 AR LINoTON, T EXAS 760108064 November 18, 1997 EA 97-382 C. Randy Hutchinson, Vice President Operations Arkansas Nuclear One Entergy Operations, Inc.
1448 S.R. 333 Russellville, Arkansas 72801-0967 SUBJECT: NOTICE OF VIOLATION (NRC INSPECTION REPORT 50-313/97-14; 50-368/97-14)
Dear Mr. Hutchinson:
This is in reference to the predecisional enforcement conference held in the NRC's Arlington, Texas office on September 26,1997. The conference was conducted to discuss three apparent violations of NRC requirements related to inspections of steam generator tubes at Arkansas Nuclear One, Units 1 and 2. The results of our inspection were discussed with your staff during a telephonic exit briefing conducted on August 8,1997, and were documented in the subject NRC Inspection Report, which was issued September 12,1997.
After the September 26 conference, ANO provided further information and clarification during a telephone discussion held on October 2,1997.
The three apparent violations described in the inspection report involved: (1) a failure to comply with 10 CFR Part 50 Appendix B Criterion IX in the use of data in bobbin coil sizing qualification criteria that did not conform to the requirements of the selected qualification method; (2) the failure to remove Unit 2 tubes from service which contained flaws that exceeded the plugging limit of the Technical Specifications; and (3) the lac!; of prompt corrective action in November 1995 prior to-retuming potentially defective sleeved tubes to service.
During the September 26 conference, Enter'g y expressed disagreement with the first apparent violation, contending that Entergy's sizing qualification for intergranular attack (IGA) met the provisions of appropriate industry qualification processes (EPRI Appendix "H"), and therefore Entergy was in compliance with 10 CFR Part 50 Appendix B. With regard to the second _
apparent violation, which involved Unit 2 steam generators, Entergy agreed that the violation occurred but contended that the violation warranted treatment as a non-cited violation.
Entergy also disputed the third apparent violation, which is discussed below.
The NRC considered Entergy's arguments and has concluded that vic;dions of NRC requirements occurred.
Based on the discussions The at the violations conference, weare cited have in the modified the enclosed first violationNotice to stateof Violation (No that 7 Entergy's inservice inspections of Unit 1 steam generators were not appropriately controlled g
, and accumplished to identify tube defects that exceeded the technical specification plugging (f 9711200032 971118 i
.
.
Entergy Operations, Inc. -2-limit. The second violation involves Unit 2 steam generator tubes, with defects greater than
' the technical specification plugging limit, that were improperly considered operable.
Although these violations involve steam generators of different designs and different IGA flaw sizing methodologies, the violations are similar in that your methods for sizing lGA and tubing degradation in both Units 1 and 2 steam generators were shown to have been inadequate.
The actual safety consequence of these violations is low because no significant leakage occurred as a re: ult of tube leaks. These issues are of regulatory significance because of the programmatic nature of the problem; that is, you relied on methods which were incapable of detecting flaws greater than the plugging limit, and as a result, operated Units 1 and 2 with steam generators with tubes containing flaws that exceeded the technical specification plugging limit of 40 percent through wall (TW). Therefore, these two violations have been categorized in the aggregate in accordance with the " General Statement of Policy and Procedure for WRC Enforcement Actions"(Enforcement Policy), NUREG-1600, as a Severity Level lli problem.
In accordance with the Enforcement Policy, a base civil penalty in the amount of $55,000 is considered for a Severity Level 111 problem. Because your facility has been the subject of escalated enforcement actions within the last 2 years', the NRC considered whether credit was warranted for Identification and Corrective Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. Since Entergy identified both issues, the NRC has determined that Entergy is deserving of identification credit.
Further, the NRC has determined that Entergy is deserving of Corrective Action credit.
Entergy's corrective actions for the first violation included declaring the steam generators inoperable, notifying the NRC, requesting (and being granted) a Notice of Enforcement Discretion from the NRC, administratively precluding IGA sizing, and investigating the cause of the problem. Entergy's corrective actions to the second violation included reviewing more
"no detectable degradation" determinations to verify the quality of the 2R12 analyses, performing enhanced inservice inspectio".s of indications, repairing the eggerate cracks, enhancing training, and providing additional oversight of the inservice inspections.
Therefore, to encourage prompt identification and comprehensive correction of violations, I have been authorized not to propcse a civil penalty in this case. However, significant violations in the future could result in a civi! penalty.
The third violation cited in the enclosed notice involves Entergy's failure to take prompt corrective actions prior to retuming to service potentially defective sleeved tubes exhibiting weld zone eddy current indications. At the conference, Entergy representatives expressed disagreement with this violation, contending that sufficient technical basis existed (including
' A Severity Level 111 problem and a $50,000 civil penalty was issued on April 9,1997, involving fire protection inadequacies and the plant staff's inadequate response to ind:ations of a fire (EA 96-512).
. .. - . - -. .
t ;
' ~
Entergy Operations, Inc. 3 ultrasonic testing and visual testing) to justify use of the sleeves as-is. Further, Entergy representatives noted thet, after re-inspection, only one of 28 indications was identified for repair based on new acceptance criteria and that all 28 indications met structural and leakage requirements; Notwithstanding Entergy's argument that the structural integrity of the sleeves was maintained, the NRC's position is that Entergy had an indication of a condition adverse _to quality, and it took inadequate corrective actions. Specifically, your actions (in October 1995, during 2R11) were not adequate to evaluate defects that existed at the time, nor did you perform an operabrity assessment at the time. Therefore, in accordance with the its Enforcement Policy, the NRC has classified this violation et Severity Leve,l IV.-
You are required to respond to this letter and should follow the instructions specified in the -
encksed Notice when preparing your response in your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. The NRC will use your response, in part, to deteimine whether further enforcement action is necessary to ensure compliance with regulatory requirements, in accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its enclosure, arid your response will be placed in the NRC Public Document Room (POR).
Sincerely,
'f
Ellic W. Mers Regional Adm'iistrator Dockets: 50-313-50-368 Licenses: DPR 51 NPF-6
'
cc w/ Enclosure:
Executive Vice President
& Chief Operating Officer
,
Entergy Operations, Inc.
'
P.O. Box 31995 L Jackson, Mississippi 39286-1995
l
r^- 1
..
Entergy Operations, Inc. -4-
.
Vice President Operations Support Entergy Operations, Inc.
P.O. Box 31995 Jackson, Mississippl 39286-Manager, Washington Nuclear Operations ABB Combustion Engineering Nuclear
!
Power 12300 Twinbrook Parkway, Suite 330 <
Rockville, Maryland 20852 County Judge of Pope County Pope County Courthouse Russellville, Arkansas : 72801 Winston & Strawn -
1400 L Street, N.W.
WasNngton, D.C. 20005 3502 David D. Snellings, Jr., Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street, Mail Slot 30 Little Rock, Arkansas 72205-3867 Manager Rockville Nuclear Licensing '
Framatome Technologies 1700 Rockville Pike, Suite 525 Rockville, Maryland 20852
.
- ,- --
,
. _ . _ _ _ _ .- - . _ _ . -. _ _ _ . .
, .,
9-Entergy OperatPpW inc, . -5-
.
bec w/ Enclosure (s): -
PDR-LPDR hNUDOCS [
SECY EC's: Rl,Ril,Rlli CA PA (Os2G4)
EDO (017G21) OiG (T 5D28)
DEDO (0-17G21) OE (0 7H5)
OE:EAFile (0 7H5) 01 (0 3E4)
^
_
OGC (0-15818) OGC (015B18)
NRR (012G18) NRR/ADP (012G18)
NRR PROJECT MANAGER OC/DAF (T-9E10)
OC/LFDCS (T 9E10) AEOD (T-4D18)
RA Reading File GSanborn-EAFile RWise (if app.) RWise-DOL File (if app.)
RIV Files MIS Coordinator ECollins, RIV E Mall DlHTRIBUTION:
OEMAIL: JDyer (JED2)
TPGwynn (TPG) WBrown (WLB)
GSanborn (GFS) GMVasquez (GMV)
BHenderson (BWH) MHammond (MFH2)
CHackney (CAH) DKunihiro (DMK1)
Art Howell (ATH) DChamberlain (DDC)
KPerkins (KEP) ECollins (EEC)
SENIOR RESIDENT- KBrockman (KEB)
EMerschoff (EWM)
DOCUMENT NAME: G:\EA\ DRAFT \EA97382.DFT To receive, copy opument, Indicats in box: "C" = Copy wthout oncesures "E" = Copy with enclosures *N" = No copy
~ ~
ES //7AV EO O TA ,,,p DRP() , D:DRS@0 ,
MVasquez GSapid IBarnes f/'D EC611TfiP AHovTell [
11k//97 11/d/97 114r/97 _ 11/'//97 11/4/97
-_
RC f4 DRA RA g/
W8rown R JDyer @ EMy.f56tioff 11/G/97 11/f//97 11//fl97 OFFIC!AL RECORD COPY