05000313/LER-1997-001, :on 970408,inconsistencies Between Once Through SG Tube Destructive Examination Were Noted.Caused by Morphology Differences Between Flaws Used to Develop Sizing Technique.Discretion Was Verbally Approved

:on 970408,inconsistencies Between Once Through SG Tube Destructive Examination Were Noted.Caused by Morphology Differences Between Flaws Used to Develop Sizing Technique.Discretion Was Verbally Approved

ML20217A200
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/18/1998
From:	Mims D, Tracy Scott ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1CAN039804, 1CAN39804, LER-97-001, LER-97-1, NUDOCS 9803240310
Download: ML20217A200 (13)

LER-1997-001, on 970408,inconsistencies Between Once Through SG Tube Destructive Examination Were Noted.Caused by Morphology Differences Between Flaws Used to Develop Sizing Technique.Discretion Was Verbally Approved

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(1) 10 CFR 50.73(a)(2)(x)
3131997001R00 - NRC Website

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k mthiuc AR 72E01 Td 50182-5000 March 18,1998 ICAN039804 U. S. Nuclear Regulatory Commission Document Control Desk

' Mail Station OPI-17 Washington, DC 20555

Subject:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR Licensee Event Report 50-313/97-001-02 Gentlemen-In accordance with 10CFR50.73(aX2XiXB), enclosed is a supplement to the subject report concerning Once Through Steam Generator tube surveillance requirements. This supplement provides revised information regarding the root cause.

V truly yours, N[

Dw' t C. Mims Dif or, Nuclear Safety DCM/tfs

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l enclosure v

9v 9903240310 900318 PDR ADOCK 05000313.

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U. S. NRC March 18,1998 1CAN039804 PAGE 2 cc:

Mr. Ellis W. Merschoff Regional Administrator U. S. Nuclear Regulatory Commission RegionIV 611 R) m Plaza Drive, Suite 400 Arlington, TX 76011-8064 Institute of Nuclear Power Operations 700 Galleria Parkway Atlanta, GA 30339-5957 NRC Senior Resident inspector Arkansas Nuclear One P.O. Box 310 London, AR72847

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FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCN (MNSS 7714), U.S. NUCLEAR hfGULATORY Cols4XSSION, WASNINGTON, DC 20555 0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104),

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IArkonsesNucleerOneUnIt1 05000313 1 0F 11 1 TITLE (4)

Incenelstencise Between once Through Steen Generator Tube Destructive Examination Resulte and Intergranuter Attack Flow Sizing Guellficetten Data caused The Possibility That Tubee Were Left In Service With Flows Exceeding The Technical Specification Llelt l

EVENT DATE (5) l LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8) l MONTH l

SEGUENTIAL REVISION MONTN DAY YEAR l YEAR l YEAR DAY NUMBER NUMBER 04 08 97 97 001 02 03 18 90 1 OPERATING l l TNIS REP 0NT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR (Check one or soc O L11)

MODE (9) lN l

20.402(b) 20.405(c) 50.73(a)(2)(lv) 73.71(b)

POWER l

20.405(e)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

EVEL M ffe)(1)(lI) 50.36(e)(2) 50.73(a)(2)(vi()

OTNER 20.405(e)(1)(l1I)

X 50.73(a)(2)(1) 50.73(a)(2)(vifI)(A)

SpecIfy in 20.405(e)(1)(fv) 50.73(a)(2)(li) 50.73(a)(2)(vill)(B)_ Abstract teleu 20.405(a)(1)(v) 50.73(a)(2)(fil) 50.73(a)(2)(x) erd in Text LICENSEE CONTACT FOR TNIS LER (12) lNAME TELEPHONE NUMBER (include Area Cede)

Themse F. Scott, Nucteer Safety and Licensing Specialist 501 854-4623 COMPLETE ONE LINE FOR EACH COMP 0NENT FAILURE DESCal5ED IN TNIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTURER

CAUSE

SYSTEM COMPONENT MANUFACTURER SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR I

YES NO SUOMllSION (If yes, complete EXPECTED SUOMIS$10N DATE)

X DATE (15)

ABSTRACT (Lielt to 1400 apaces, i.e., approximately 15 eingle speced typewritten lines) (16) on April 8, 1997, ANO-1 discovered that inconsistencies between destructive examination resu?.ts of three Once Through Steam Generator (OTSG) pulled tubes and field oddy current sizing of intergranular attack (IGA) flaws in the upper tubesheets created the possibility that tubes had been left in service during a refueling outage in the fall of 1996 with defects exceeding the through-wall limit contained in Technical fpecifications (TS).

The sizing technique had been qualified in accordance with EPRI guidelines. Technical Specification 3.0.3 was entered, with actions delayed per 4.0.3, until enforcement discretion was obtained from the NRC. A request for an exigent TS change was submitted to allow operation for the remainder of the current cycle with IGA flaws in the upper tubeshest exceeding the TS limit. The TS change was issued as Amendment No. 189 on May 7, 1997. ANO has not been able to determine a root cause for this condition, but several potential contributing causes have been identified. Tubes that could remain in service with flaws in excess of the limit were determined not to represent a structural or leakage concern.j i

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02 TEXT t i f =ca

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A.

Plant Status At the time this condition was discovered, Arkansas Nuclear One Unit 1 (ANO-1) was operating in normal steady-state conditions at 100 percent power.

B.

Event Description

On April 8,1997, ANO-1 discovered that inconsistencies between destructive examination results of three Once Through Steam Generator (OTSG) [AB] tubes pulled during 1996 and previous oddy current sizing ofintergranular attack (IGA) flaws in the upper tubesheets created the possibility that tubes had been left in service during a refueling outage in the fall of 1996 with defects exc@g the through-wall limit contained in Technical Speci6 cations (TS).

Intergranular Attack (IGA) is known to be present above the 15th tube support plate (TSP) within the ANO-1 OTSGs as verified by destructive examination (DE) from previous tube pulls.

IGA is s damage mechanism caused by corrosion of the material grain boundaries. The corrosion resulted from contaminants introduced on the tubing during the early years of pir.n* operska.

The contaminant causing IGA of the ANO-1 tubing is sulfur resulting from thermal decomposition ofion exchange resins. The ANO-1 IGA can be categorized as volumetric, or

" patch-like", with no specific orientation. Since discovery, there has been no evidence ofleakage from IGA flaws at ANO-1.

During the IR13 refueling outage in September and October of 1996, an eddy current (EC) technique was employed to depth size the IGA. This technique had been qualified per Appendix "H" of the EPRI "PWR Steam Generator Tube Examination Guidelines," Revision 4 dated June 1996. Compliance with the EPRI guidelines was considered an acceptable method to quahfy non-destructive examination (NDE) techniques for the detection and sizing of damage mechanisms, and the guidelines were the only qualification technique available at that time. This technique was used to depth size all IGA flaws within the upper tubesheet (UTS) using the bobbin coil. During this inspection, 25 percent of all indications detected within the UTS region by bobbin coil were examined using a rotating pancake coil (RPC) to characteriae these flaws. All IGA indications between the 15th TSP and the secondary face of the UTS were removed from service by plugging All UTS IGA indications with a depth size of greater than or equal to 40 percent through-wall ('IW), as determined by the qualified sizing tMmique, were also removed from service by plugging.

NRC 70AM 366A (5 92)

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VEAR sEeLENTdAL REVISION 1ASIBER IANGER Arkansas Nuclear one Unit 1 05000313 3 F 11 97 m

02 TEXT fif mara -- = in e - 'ead u== -w' e t " **='a= af aner sar= ' ""1 (17)

During IR13, three tubes with bobbin indications within the UTS were removed from the "B" OTSG for future development of an Alternate Repair Criteria (ARC) and to further suppolt the qualified EC sizing technique. Two of the three tubes contained flaws that would have required repair. The third tube was near the repair limit. h tubes were selected on the basis of their containing multiple indications with depths representative of the average indication depths as sized by EC. After bursting the tubes in the laboratory, the flaws were examined and sized. If a flaw was not opened by the burst of the tube it was bent open for DE. The DE results were not consistent with the previous qualification data of the bobbin coil for sizing IGA flaws in the UTS.

The flaw depths did not correlate well with the depths sized by using the qualified EC technique.

- As a result of this condition, it is possible that tubes were left in service with through-wall defects greater than the technical specification plugging limit.

C.

Root Cause As documented in supplement 1 (ANO letter ICAN079701 dated July 15,1997) the initial evaluation identified two root causes for the unexpected correlation of the EC depth sizing with DE depth sizing. h first was that the data sets used to qualify the EC technique for sizing IGA were not exactly in accordance with recontmendations of the EPRI guidelines for data i

distribution. The second was that there were morphology diff' rences between the flaws used to e

develop the sizing technique and the flaws in tubes removed from ANO-1 during 1996.

Subsequent to the original root cause evaluation, information has developed indicating that neither can be conclusively identified as a root cause of the condition.

The regression model for sizing IGA was developed based upon EC and DE data from tube pulls in 1992 and 1994 at Crystal River 3 (CR-3) and from tube pulls in 1982 and 1984 at ANO-1. The regression combined multiple EC measurements in a linear equation that provided improved EC depth estimation for data from the pulled tube evaluations. W EC data included voltage and

)

phase angle measurements for 600,400, and 200 kHz frequencies. The data sets included 53 data points (47 from CR-3 and 6 from ANO-1) ranging in depth from 15 to 100 percent TW. The 53 dats points did not meet the recommendations of the EPRI PWR Steam Generator Tube Examination Guidelines, Revision 4, for data distribution based upon flaw depths Only 7 of the 53 data points were greater than or equal to 60 percent TW. The guidelines indicated that at least two-thirds should be in that range with a uniform distribution over the range. Information provided by the technique developer, Framatome Technologies Incorporated (FTI), indicated that other detection / sizing techniques qualified in accordance with the EPRI guidelines have not met this criterion. Additional consideration was given to acceptance of this deviation. The limited number of real tube volumetric IGA defects in the depth range of 60 to 100 percent TW required the use of a more conservative detection range of 40 to 100 percent TW. There were a sufficient oc PORN mea (5 92>

t.

INIC ruum M4A U.5. NUCLEAR NEGULAsuun immI5510N APPROWED sY OM No. 3150 0106 (542)

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YEAg SEeUENTIAL REVISI0li IEDGER insWER Arkansas leuclear One - Unit 1 05000313 97 001 02 TENT 'f8 = ara ----- ta e--"t rase tana==8d8

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-' ' ' s (17) number of detections in this expanded range to serve as a valid App-h "H" detection data set.

The expanded range contained 38 flaws and 35 detections for a probability of 0.83 at the 90 percent lower confidence level. Appendix "H" specifies that the sizing capability be demandrated on the detection data set. The regression sizing technique was demonstrated on the detection data set containing 34 detected flaws (62 percent of the demonstration data set) greater than or equal to 40 percent TW and 21 flaws less than 40 percent TW. One defect did not contain the necessary frequency data; therefore, it could not be sized with regression. Elimination of four defects below 40 percent TW from the demonstration data set would have produced two-thirds of the data set above 40 percent TW. Dev~ations from the EPRI guidelines are acceptable with j

adequatejustification. The past experience provided by FTI regarding other technique qualifications and the additional consideration was felt to provide justification for allowing the deviation for this, technique. As part of the qualification, a review was coM+i by a peer group. At the beginning of the peer review, the Appendix "H" guidehnes were reviewed and the data point deviation explained to the peer review group members. The deviation was accepted by the peer review group, one of whose members was an EPRI employee. Subsequent to the original root cause evaluation, the EPRI qualification guidelines were revised. Revi:, ion 5 clarifies that the data distribution need only be as uniform as reasonably achievable. Additionally, Revision 5 requires the sizing data set to include the detection data set and does not allow the exclusion of data. Comparison of the qualification data set with current guidelines indicates that not having used a larger data set cannot conclusively be identified as a root cause; however, it cannot be ruled out as a potential contributing cause of the sizing discrepancy. The data distribution was typical of what was known of OTSG volumetric IGA prior to information about depth range that was obtained from 1996 tube pull results.

Variations within the flaw morphology were observed when comparing the 1996 tube pull DE results with the tube pull results used in the qualification of the technique. The 1982 and 1984 ANO-1 IGA possessed a " patch-like" shape, i.e., the flaws were volumetric but had no consistent physical shape. Each of these flaws also contained sulfur deposits within the flawed area. The IGA in the CR-3 tube samples also contained sulfur deposits within the flawed area but ws.s in a more symmetric, " pit-like" configuration. The flaws in the 1996 ANO-1 tube samples were

" patch-like" but contained short, deep IGA penetrations within the flawed area with no evidence of sulfur deposits. The 1996 ANO-1 IGA appears to have formed from closely spaced IGA cells that merged to create an irregular depth contour. The physical differences between pre-1996 ANO-1 and CR-3 IGA were known at the time of the regression development. These differences were not considered to be detrimental to rizing of the IGA. The morphology comparison concluded that the pre-1996 morphology for both CR-3 and ANO-1 were reasonably similar and acceptable for support of the sizing development. Based on a review of the 1996 tube pull DE results, the data sets were originally determined to be inappropriate t=== it was beheved that the 1996 morphology differed from the technique qualification basis. However, further unc FORN 3sdA (5 92)

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YEAg sEeuENTIAL REVIs10N IpWER IMSER Arkanses Nucteer one - Unit 1 05000313 5 0F 11 9r M

02 TEXT fff =ar. -- ^^ i r - se.d u

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- t av une s.,. -e e n (try comparisons in support of Alternate Repair Criteria (ARC) for IGA indicate that the morphology is essentially the same for all plants from which tubes were pulled. Also, a comparison of ec'dy current data from plant to plant shows that the IGA eddy current responses from different plants and locations are similar. The morphology comparison in the ARC report does not contain enough detailed information to eliminate the IGA differences as a potential cause of the sizing error, but it does not reveal evidence sufficient to conclude that morphology is a root cause.

. A potential contribution to the poor correlation was differences in the DE techniques for locating and inessuring flaw depths. The 1996 ANO-1 tube samples were burst pressure tested to determine the structural integrity of the tubing. The burst pressure test resulted in a " fish mouth" failure of the tube's most degraded area and swelling of other flawed areas. Any flaws that were not " fish mouth" were then fatigued open by bending. Each flaw was depth sized visually using a scanning electron microscope (SEM) along the fracture face. This method is considered to be the most accurate in measuring the flaw depth. h burrt and swell locations represent the weakest area of the flawed region. No burst testing was performed on the pre-1996 tube samples. The previous DE technique employed was transverse metallography. This technique requires sectioning of the tubing to measure flaw depth using the~ SEM. The tubing axial face is polished multiple times with flaw depth measurements made between each polishing. This method's accuracy is limited by the control ofmaterial removed during the polishing step. Short, deep penetrations can be removed during the polishing process. Flaw removal during the polishing task can result in incorrectly sizing the flaw too shallow. This potential cause has been supported by information from the report of the IR13 tube pull examination. The fractography depth measurements taken in the lab for tubes pulled during 1R13 have an estimated +/- 5 percent TW accuracy. Thic error, coupled with the error from the metallography on the flaws used to quahfy the regression, could be significant enough to account for some of the sizing error.

1 The qualification data set contained four flaws that were 100 percent TW. The amplitude l

response from these flaws could have had an adverse effect on the development of the sizing regression formula. This formula uses both voltage amplitude and phase measurements. The

~

phase accuracy for the 100 percent TW flaws is expected to be better than for shallow flaws.

However, the voltage amplitude could fluctuate significantly depending on the penetration area.

i Voltage amplitude sizing works well when considering flaws less than 100 percent TW where both the actual depth of penetration and volumetric degradation (amplitude) can be correlated.

When the flaw reaches 100 percent TW, the depth constituent of the relationship is cw=at, but j

the voltage amplitude can continue to increase, thus causing the accuracy of the amplitude-depth J

relationship to be in question ' For example, the 600 kHz diffe'rential amplitude for the 100 percent TW flaws ranged from 0.82 - 6.87 volts. Even though there were a limited number of 4

100 percent TW flaws in the qualification data set, the impact of this factor cannot be eliminated as a potential cause.

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YEAg steUENTIAL REVISION InsWER lasWER Arkansas Nucteer One - Unit 1 05000313 97 001 02 TENT 'If = ara --- - in e= d ead

= = =M'*8 - ' - 8== as mac s.

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At the time of the original root cause evaluation, ANO believed that another potential contributing cause was the EC bobbin coil inherently averaging acquired signals over the coil's field of view. The field of view for the 0.510 inch differential bobbin coil is approximately 0.24 inches. The largest IGA flaw length measured in 1996 tube samples was 0.25 inches. This inherent averaging was thought to support the fact that the bobbin coil measured the flaw average depth with values closely correlated to the depth measured by DE. Upon further review, it was determined that this factor was considered during development of the correlation. While certain aspects oflGA morphology could potentially affect oddy current response, a review of the ARC report indicated that there is no evidence that this was a cause of the sizing error.

Another potential cause that was eliminated was the effect of the tubesheet on the eddy current signals Recent work completed by ITI utilizing laboratory IGA indicates that the net effect on the regression formula due to the tubesheet is minimal. It was concluded that tubesheet bobbin voltages are 80 percent of the freespan voltages when comparing the 200,400, and 600 kHz ciannels. Comparing phase angles from the same three ch==-4 yields a minimal, insigificant difference. However, since the regression technique uses a combination of voltage and phase angle data, the effect of an amplitude change caused by the presence of the tubesheet is minimized A comparison of eddy current data for tube samples pulled during IR13 revealed that signal traces were similar for the various frequencies between tests with the tube inside or outside of the tubesheet It has been concluded that the difference in Anw location (tubesheet vs.

freespan) had no impact on the regression formula.

In summary, based on the available information, including numerous uncertainties associated with DE measurements, eddy current variability, qualification requirements, and specific morphology aspects of flaws, ANO has not been able to conclusively determine a root cause for the error in sizing IGA tube flaws at ANO-1. Potential contributing causes are the adequacy of requirements for technique qualification in Revision 4 of the EPRI guidelines, differences in flaw morphology at a level that cannot be accurately compared, differences between destructive examinations and the measurement uncertainties associated with lab measurement techniques, having used 100 percent TW flaw amplitude parameters as inputs to the formula development, and potentially inadequate data distribution of the flaws used in qualification of the sizing technique.

D.

Corrective Actions

At 2012 on April 8,1997, both OTSGs were administratively declared inoperable and TS 3.0.3 entered with actions delayed as allowed by TS 4.0.3. A request for enforcement discretion was i

verbally approved by the Nuclear Regulatory Commission (NRC) at 1535 on April 9,1997, at j

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YEAg sEGUENTIAL AEVis10It IRSIBER InsesER Arkansas Nucteer One - Unit 1 05000313 97 001 02 TEXT 'sf = ara

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.# meso,. men (173 which time the TS was exited. A letter (ICAN049702) documenting the enforcement discretion i

request was submitted on April 9,.1997. W duratio'i of the discretion was until May 7,1997, or i

until the NRC Staff acted upon a proposed exigent TS change request, whichever occurred first.

On April 11,1997, ANO-1 submitted a letter (1CAN049703) requesting an exigent TS change to allow tubes with IGA flaws greater than 40 percent TW in the OTSG UTS to remain in service i

for the, remainder of the current operating cycle. Following a telephone conversation between ANO and the NRC staff on May 2,1997, a revision was submitted to the exigent TS change to

~ add a more retstrictive limit for allowable leakage through tubes of one OTSG. This limit is consistent with procedural controls that were already in place for the unit and with guidance provided to other facilities for similar conditions. The NRC approved the proposed TS change as Amendment No. I89 in a letter dated May 7,1997.

ANO-1 had previously implemented shutdown limits for primary-to-secondary leakage that were more restrictive than TS limits. A review of the extensive measures previously taken by ANO-1 to enhance the operators' abilities to detect and respond to OTSG tube leakage indicated that no additional compensstory measures were necessary to address the surveillance deficiency.

Administrative action has been completed to preclude future use of the IGA sizing technique at ANO.

The OTSG tubes with IGA indications in the UTS area will be dispositioned prior to staltup from the next scheduled refueling outage. The next refueling outage is currently scheduled to start in March of1998.

E.

Safety Significance

The subject EC sizing technique was employed for IGA defects within the UTS. All tubes containing UTS IGA indications with a depth size of greater than or equal to 40 percent TW were removed from service.

The three UTS IGA tube samples removed during IR13 were subjected to room temperature burst testing Burst. testing was performed separately within the flawed and unflawed regions of i

the tube samples No simulated tubesheet was employed during the tests. The tests were performed using bladders in the flawed region. No foils or lateral restraint systems were used.

The burst pressures for the flawed regions were between 10,000 and 11,000 psig. The unflawed regions burst at pressures between 10,700 and 11,200 psig. For ANO-1 OTSGs, structural integrity is conselvatively demonstrated by pressurizing the tubing to three times normal operating NRC FORM 3MA (5 92)

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.# nanc roem '" n (17) differential pressure. This pressure for ANO-1 is 3,765 psig. The burst testing results indicate that substantial structural margin exists.

In 1996, to support ANO's study ofIGA, burst testing of pre-defected tubes was completed by FTL The burst testirg consisted of nine tubes containing through-wall drilled holes up to 0.5 inches in diameter and one tube containing no defects placed within a sim61sted tubesheet. Nine of the specimens burst at pressures greater than or equal to 10,941 psig. Each tube burst outside the tubesheet within the non-defected portion of the tubes. One tube reached a pressure of 9,577 psig but did not burst due to bladder leakage. These test results indicate that the tubesheet provides sufficient support to preclude tube rupture within the tubesheet The tube samples removed from ANO-1 in 1996 included eleven IGA indications in the UTS.

Since it'was confirmed that the inservice IGA indications are volumetric, bobbin amplitude (voltage) was used as a bounding parameter. The eddy current responses from these flaws were compared with the population ofinservice IGA indications to determine how representative the flaws were of those remaining in service. The 600 KHz bobbin coil signal amplitude of flaws in tubes that were pulled during 1996 ranged from 0.46 to 2.69 volts. Of the 470 inservice IGA indications, all are bounded by the 2.69 volt value.

Additionally, a comparison ofRPC data was performed to further substantiate that the pulled tube

{

flaws bound those indications remaining in service. The RPC data collected for the tube pull I

samples resulted in a maximum flaw extent of 0.16 inches. RPC signal information was collected on 118 indications within the UTS. Ten of the largest RPC voltage indications were examined to determine the length-by-width extent by RPC. The largest RPC extent for those IGA indications lea in service was 0.14 inches Therefore, it is concluded that the inservice IGA indications are bounded by those tube samples that were destructively examined.

Structural integrity of the tubing within the tubesheet is assured based upon demonstration of the following:

A. The actual tube samples removed from ANO-1 during IR13 exhibited burst pressures that substantially exceeded the required structural limit.

B. The structural support provided by the tubesheet precludes tube rupture.

C. The inselvice IGA indications are bounded by those flaws contained in the tube samples that were pulled.

NRC PORN 366A (5 92) l a

unc rusus 366A U.5. NUCLEAR utuuLMunT unmI55 ION APPROWED sf GNs No. 3150-0106 (5-92)

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THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION E,,,Q,gs Y0f01 NENb iO5 ANo REDUCTION PROJECT (31$00104), 0FFICE OF MANAGEMENT AND BUDGET. WAsNINGTON. DC 20503.

FACILITY NAIE (1)

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PAGE (3)

YEAt SEQUENTIAL REvlsIGII IESIsER WUISER Arkanses Nuclear One - Unit 1 05000313 97 001 02 TEXT fff =aea - - - - - l a r==" f rad una aM8ttaaat eaataa of une rare ms (17)

The IGA patches destructively examined were not through-wall; therefore, normal operating pressures did not result in through-wall leakage. This was evident during inservice inspection of the tubing in which no indications of residual leakage was noted. A comparison of IR12 and 1R13 refueling outage EC signatures indicates that the IGA exhibited little or no growth.

Comparison ofinspection data' prior to the 1R12 refueling outage supports this conclusion.

Additionally, during May 1995, "B" OTSG tubing was subjected to a differential pressure of approximately 2,100 psi. for several hours as a result of a feedwater transient. No immediate d

increase in primary-to-secondary leakrate was noted during the event or following startup. The primary-to-secondary leakrate did increase by approximately 18 gpd three days following staltup; however, none of the leakage detected during the IR13 refueling outage was from IGA flaws. It is concluded that leakage through IGA flaws in the UTS is highly unlikely at Main Steam Line Break (MSLB) pressures due to the flaw morphology. The near MSLB differential piessure that occurred in May 1996 caused no resultant leakage.

Conditional core damage probability is the increase in core damage frequency due to a given condition other than that assumed for the base Probabilistic Risk Assemament (PRA). The PRA assumed that the tube integrity is such that no OTSG tube mpture would be induced due to transient conditions. The limiting licensing basis transient which could most adversely affect the tubes by creating a high differential pressure across the tubes is a MSLB accident. This accident could produce a tube differential pressure of up to 2,500 paid. The tube sample burst pressures were well above pressures which would be seen in a limiting MSLB accident. Thus, the likelihood of tubes rupturing is not increased because of the larger than expected flaw sizes due to IGA in the UTS. This situation has been qualitatively assessed and the conditional core damage probability for this condition is estimated to be inconsequential.

The limiting licensing basis accident with respect to dose consequences from induced tube leakage is the MSLB accident. This accident assumes a total leakage of I gpm with I percent failed fuel in the core. However, OTSG tube leakage is procedurally limited to 0.1 gpm during normal operation. Even though leakage is not expected to occur, MSLB induced tube leakage has been conservatively estimated to be 0.53 gpm on the limiting steam generator. The following assumptions were made concerning the number of flaws and associated leakage:

~ ince OTSG "A" has the largest number oflGA patch indications (285), it was 1)

S chosen as the more limiting generator bounding OTSG "B" with only (185) indications.

2)

Half of the indications were assumed to leak under MSLB conditions.

NRC PORN 3664 (5 92)

Nuc russi 3eeA U.5. NUCLEAR REGULAvun ww4353GN APPROW4!0 SV OMO No. 3150-0106 (5 92)

EMPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITN TNIS !NFORMAtloN (XILLECTION REQUEST: 50.0 NES.

FORWARD CosqMENTS REGAR0 LNG BURDEN ESTINATE TO LICENSEE EVENT REPORT (LER)

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@,,Q,y.S EARREGULA}0{

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PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER Arkansas Nucteer One Unit 1 05000313 10 of 11 97 ua..adr e s-a. t as anac sam *"* s (17)

TEXT f i f mara --

.t r-s e.d 3)

Representative leakage values for axial flaw lengths were utilized to bound the leakage expected from IGA patches.

4)

The longest IGA length calculated from RPC data was applied to the 50 percent population assumed to leak.

5)

The flaws were assumed to grow in length an additional 25 percent over the cycle.

6)

Fifty percent of the flaw length was assumed to be 100 percent TW in depth.

Since there are 285 indications, half of this value is approximately 143. The longest length in the axial plane was 0.14 inches. When increased by 25 percent, this yields a flaw length of:

0.14 inches * (1.'25) = 0.175 inches If 50 percent of the length is assumed to be 100 percent TW:

0.175 inches * (0.5) = 0.0875 inches Using leakage curves developed for OTSGs for axial flaws, the leakage from a single flaw (0.0875 inches,100 percent TW) is determined to be 0.0025 gpm. To compensate for normal operating temperature the value is multiplied by 1.47 to yield a final leakage of 0.003675 gpm per flaw.

This value is then multiplied by the number of potential leaking flaws to give a total leakage of:

143 flaws

• 0.003675 gpm/ flaw = 0.53 gpm When the estimated leakage in the limiting OTSG is added to that which is allowed by procedure, the total leakrate is expected to be no greater than 0.63 gpm. Since the assumed leakrate is greater than the conservative calculation, the current licensing basil, assumption of I gpm remains bounding.

The subject flaws do not represent a structural or leakage concern Therefore, the presence of inservice upper tubesheet IGA defects with through-wall extents that may exceed the technical specification plugging limit does not pose a concern relative to the health and safety of the public.

MAC FORM 366A (5 92)

i NRC ruse Jean U.3. RUcLEAR REGULArun wed5510R APPROWED BY 05 NO. 3150 0104 (5 92) gxPIRES 5/31/95 f

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TME INFORMATION AND MCORDS MANAGEMENT BRANCN TEXT CONTINUATION E,Q Q.s;gyNuCLEAR,REeULAT cpc gp O M REDUCTION PROJECT (31b-0104),' 0FFICE OF MANAGEMENT AIS BLDGET, W4sNileGTON, DC 20503 FACILITY NAM (1)

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YEAR SEQUENT!kL REVl810N NUMBER 1RSSER Arkanses Nucteer One - Unit 1 05000313 97 001 02 TEXT tif = ara - - a la raauscad u

.adattanat cant n# une rare saaas (17}

F.

Basis for Reportability The inservice inspection of ANO-1 OTSGs is conducted in accordance with Technical Specification 4.18. Specification 4.18.2 states, " Inservice inspection of steam generator tubing shall include non-destructive examination by eddy-current testing or other equivalent techniques."

Specification 4.18.3 requires that a minimum sample size be examined in accordance with specification 4.18.5. Specification 4.18.5.b states, "The steam generator shall be determined operable after completing the corresponding actions (plug or sleeve all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.18-2." Table 4.18-2 requires " defective" tubes to be plugged or sleeved. Specification 4.18.5 contains definitions of" defect" and " plugging limit" that require tubes containing imperfections of a depth greater than or equal to 40 percent of the nominal tube wall thickness to be sleeved or removed from service Since tubes containing IGA defects exceeding that value were possibly left in service, this constitutes an operation prohibited by Technical Specifications that is reported in accordance with 10CFR50.73(aX2XiXB).

Both OTSGs were administratively declared inoperable and TS 3.0.3 was entered at 2012 on April 8,1997. Actions were deferred as allowed by TS 4.0.3 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in order to process a request for enforcement discretion. Entry into TS 3.0.3 is also reportable in accordance with 3

10CFR50.73(aX2XiXB) as an operation prohibited by Technical Specifications.

G.

AdditionalInformation There have been no similar conditions reported as Licensee Event Reports by ANO.

Energy Industry Identification System (Ells) codes are identified in the text as [XX].

i