1CAN049703, Application for Amend to License DPR-51,requesting Exigent TS Change Re Steam Generator Tube Insp Surveillance Requirements

Application for Amend to License DPR-51,requesting Exigent TS Change Re Steam Generator Tube Insp Surveillance Requirements

ML20140E156
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/11/1997
From:	Hutchinson C ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20140E159	List: 1CAN049703, Application for Amend to License DPR-51,requesting Exigent TS Change Re Steam Generator Tube Insp Surveillance Requirements ML20140E162
References
1CAN049703, 1CAN49703, NUDOCS 9704250102
Download: ML20140E156 (17)

Text

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. .y . Ent rgy oper tion:,Inc.

J/#00 RuwAde. AR 72801 Tel 501M8 4888 C. Randy Hutchinson Vce Prmm April 11,1997 #

1CAN049703 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station PI-137 Washington, DC 20555 l l

Subject:

Arkanst.s Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 1 Exigent Technical Specification Change Regarding Steam Generator Tube Inspection Surveillance Requirements Gentlemen:

As described in our letter dated April 9,1997 (ICAN049702), and discussed during a telephone conversation on April 9, 1997, with members of the Nuclear Regulatory Commission staff, Arkansas Nuclear One, Unit 1 (ANO-1) is requesting an exigent technical specification (TS) change to specification 4.18.5.b. This specification contains the '

requirements for the repair or removal from service for those tubes with indications exceeding the plugging limit. The attached TS change will allow tubes with intergranular attack flaws within the upper tube sheet with potential through-wall depths of greater than the plugging limit to remain in service for the remainder of the current operating cycle (e 1514). TWnd i of the current operating cycle is scheduled for the spring of 1998. Attached is the proposed revision to TS 4.18.5.b and the detailed justification for this exigent TS change request.

The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the attached submittal.

Entergy Operations requests that the effective date for this change be upon NRC issuance and prior to the expiration of the notice of enforcement discretion received on April 9,1997, for this specification.

We request that this proposed change be considered under exigent circumstances as described in 10 CFR 50.91(a)(6) in that failure to act quickly could result in the shutdown of ANO-1.

As required by 10 CFR 50.91(a)(6), attached is a statement of the exigent circumstances surrounding this request.

9704250102 970411 " ,i PDR ADOCK 050003135 ll P PDRg .

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U. S. NRC April 11,1997 1CAN049703 Page 2 Very truly yo s, W '

rde Attachments To the best of my knowledge and belief, the statements contained in this submittal are true.

SUBSCRIBED AND SWORN TO before me, a Notary Public in and for rw County and the State of Arkansas, this ll day of ll&ll I

,1997.0

.lta rv% s > JA

// mensen Notary Pupic ,

My Commission Expires //-8'A'Od 8 0 j.L % l JOHNSON COUNTY I W Commeseen Expree:11 8 2000 l

l 1

U. S. NRC .

April 11,1997

, ICAN049703 Page 3 i

cc: Mr. Ellis W. Merschoff

. Regional Administrator i

U. S. Nuclear Regulatory Commission Region IV

, 611 Ryan Plaza Drive, Suite 400

Arlington, TX 76011-8064

NRC Senior Resident Inspector i Arkansas Nuclear One P.O. Boy. 310 London, AR 72847 Mr. George Kalman -

NRR Project Manager Region IV/ANO-1 & 2 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-H-3 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. David D. Snellings Director, Division ofRadiation Control and Emergency Management Arkansas Department ofHealth 4815 West Markham Street i Little Rock, AR 72205 1

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. ATTACHMENT  ;

TD 1CAN049703,

.i PROPOSED TECHNICAL SPECIFICATION AND I RESPECTIVE SAFETY ANALYSES IN THE MATTER OF AMENDING LICENSE NO. DPR-51 4

ENTERGY OPERATIONS. INC.

4 ARKANSAS NUCLEAR ONE. UNIT ONE DOCKET NO. 50-313 I

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DESCRIPTION OF PROPOSED CHANGES Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specification (TS) surveillance requirement i '4.18.5.b has been modified to allow tubes with intergranular attack indications within the l upper tubesheet with potential through-wall depths of greater than the plugging limit to remain in service for the remainder of cycle 14.

j' BACKGROUND i

! The inservice inspection of the ANO-1 steam generators is conducted in accordance with i j- ANO-1 Technical Specification (TS) 4.18. Specification 4.18.2 states: " Inservice inspection ofsteam generator tubing shallinclude 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. notes: "Ihe steam generator shall be determined operable after completing the corresponding actions (plug or ,

j sleeve all tubes exceeding the plugging limit and all tubes containing through-wall cracks) l

required by Table 4.18-2." Table 4.18-2 specifies the expansion criteria for sampling of the I

i steam generator tubes and requires " defective" tubes to be plugged or sleeved. Specification l 4.18.5 defines Defect as follows: "an imperfection of such severity that it exceeds the

] plugging limit except where the imperfection has been spanned by the installation of a sleeve. A tube containing a defect in itspressure boundary is defective." Pluzzinz Limit is l

defined in the same specification as follows: "the imperfection depth at or beyond which the l

tube shall be restored to serviceability by the installation of a sleeve or removedfrom service

! because it may become unserviceable prior to the next in.spection; it is equal to 40% of the nominaltube wallthickness."

The bases for specification 4.18 states: "The surveillance requirementsfor inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modyication ofRegulatory Guide 1.83, Revision 1."

Intergranular attack (IGA) is known to be present above the 15th Tube Support Plate (TSP) j within the ANO-1 Once Through Steam Generators (OTSGs) as verified by destmetive  !

examination from previous tube pulls. IGA is a damage mechanism caused by corrosion of the material grain boundaries. The corrosion resulted from contaminants introduced on the tubing during the early years of plant operation.' The contaminant causing IGA of the ANO-1 tubing is sulfur as a result of 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 refheling outa8e, an eddy current technique was employed to depth size the IGA. This technique was qualified per Appendix H of the EPRI "PWR Steam Generator Tube Examination Guidelines." Compliance with the EPRI guideline is considered an acceptable method to qualify non-destmetive examination (NDE) techniques for the detection

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' and sizing of damage mechanisms. This was the only qualification technique available at that time. This technique was used to depth size all IGA flaws within the upper tubesheet (UTS).

During this inspection, 25% of all indications detected within the UTS region by bobbin were examined using rotating pancake coil (RPC) to characterize these flaws. All UTS IGA indications with a depth size of >40% through-wall (TW), as determined by the qualified sizing technique, were removed from service by plugging during this inspection.

During 1R13, three tubes with bobbin indications within the UTS were removed from the steam generator. Two of the three tubes contained flaws that would have required repair.

The third tube was near the repair limit and may have been preventively repaired. The 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 destructive examination. The DE results are not consistent with the previous qualification data of the bobbin coil for sizing IGA flaws in the UTS. The reason for the inconsistency in sizing IGA in the UTS is still under review. 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.

When non-compliance was determined on April 8,1997 at 2012 CDT, the time clock for TS 4.0.3 was entered allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to seek regulatory relief. ANO verbally requested notification of enforcement discretion at 1400 CDT on April 9,1997. Verbal approval of this enforcement discretion request was received at 1535 CDT on April 9,1997. This discretion will be in effect until May 7,1997, or until the Staff acts on the proposed technical specification change request, whichever occurs first.

Therefore, ANO is requesting an exigent TS change to allow a one time exception to the

- surveillance requirements of Section 4.18.5.b. This exception will allow tubes with IGA indications within the upper tube sheet with potential through-wall depths greater than the plugging limit to remain in service for the remainder of cycle 14.

JUSTWICATION OF CHANGE The three UTS IGA tube samples removed during 1R13 were subjected to room temperature burst testing. Burst testing was performed separately within the flawed and unflawed regions of 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 ll,000 psig. The unflawed regions burst at pressures between 10,700 and 11,200 psig. For ANO-1 OTSGs, structural integrity is conservatively demonstrated by pressuiizing the steam generator tubing to three times normal operating differential pressure. This pressure for ANO-1 is 3765 psig.

The burst testing results indicate that substantial structural margin exists.

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. ICAN049703 4 Page 3 of11 In 1996, to support ANO's study ofIGA, burst testing of pre-defected tubes was completed by Framatome Technologies Inc. (FTI). The burst testing 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 simulated tubesheet. Nine of the specimens burst at pressures ;t 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 i tubesheet.

l The tube samples removed from ANO-1 in 1996 included eleven IGA indications in the UTS.-

i Since it was confirmed that the inservice IGA indications are volumetric, bobbin amplitude i (voltage) was used as a bounding parameter. The eddy current responses from these flaws

, were compared with the _ population of inservice IGA indications to determine how l 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.

l 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 j pull 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 e examined to determine the length-by-width extent by RPC. The largest RPC extent for those IGA indications left 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 structurallimit.

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

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

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 IR13 refueling outage EC signatures indicates that the IGA exhibits little or no growth.'

Also, comparison of. inspection data prior to the IR12 refueling outage supports this conclusion. Additionally, during May 1996, "B" OTSG tubing was subjected to a differential pressure of approximately 2100 psid for several hours as a result of a feedwater transient. No immediate increase in primary-to-secondary leak rate was noted during the event or following startup. The primary-to-secondary leak rate did increase by approximately 18 gpd three days

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, Page4 ofil i - following startup; however, none of the leakage detected during the IR13 refueling outage l

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 and the near

MSLB differential pressure that occurred in May 1996 with no resultant leakage.

4 Conditional core damage probability is the increase in core damage frequency due to a given

. condition other than that assumed for the base PRA. The PRA assumed that the tube integrity I is such that no steam generator tube rupture 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 j 1

! tube differential pressure of up to 2500 psid. The tube sample burst pressures were well l i above pressures which would be seen in a limiting MSLB accident. Thus, the likelihood of - l l tubes rupturing is not increased because of the larger than expected flaw sizes due to IGA in j

! the UTS. This situation has been qualitatively assessed and the conditional core damage l I

. probability for this condition is estimated to be inconsequential.

h The limiting licensing basis accident with respect to dose consequences from induced tube

leakage is the MSLB accident. This accident assumes a totalleakage of I gpm with 1% failed fuel in the core. However, steam generator 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 affected steam generator.

The following assumptions are made concerning the number of flaws and associated leakage

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$ 1) Since steam generator "A" has the largest number ofIGA patch indications (285) it was j chosen as the affected generator bounding steam generator "B" with only (185) 'l l

indications. l q

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2) Half of the indications are assumed to leak under MSLB conditions.

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

4) Applied the longest IGA length calculated from RPC data to the 50% population assumed to leak.

5) Assumed the flaws grew in length an additional 25% over the cycle.

6) 50% of the flaw length will be 100% TW in depth.

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

0.14 inches . (1.25) = 0.175 inches If 50 % of the length is assumed to be 100 % TW:

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Page 5 of11 l 0.175 inches * (0.5) = 0.0875 inches a

Using a leakage curve developed for OTSGs for axial flaws that is included in this submittal as Attachment 1, the leakage from a single flaw (0.0875 inch,100 % TW) is determined to be 0.0025 gpm. The leak rate curve established for the OTSG leakage evaluation is determined

by using the leak rate calculations from an EPRI program on burst strength and leakage .

_ prciperties of steam generator tub i ng w i th corros i on degradation. The crack opening is e cahulated based on material properties, tubing geometry, pressure, and temperature. . .The l

PICEP two phase flow model is then used to compute flow rates through the cracks as a

' function of crack opening area, temperature, pressure, and through wall length. A check of

. the validity of the leak rate equations was made by comparing the calculated leak values to the  ;

, measured leak values. The results revealed that the calculated values bound the axial stress 5

corrosion crack data. Insufficient data exists to develop curves for patch IGA flaws. The  :

axial crack data is considered conservative with respect to IGA which has been demonstrated

through industry experience. 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

1 s

1 143 flaws e 0.003675 gpm/ flaw = 0.53 gpm When the estimated leakage in the affected steam generator is added to that which is allowed  ;

1 by procedure (0.1 gpm), the total leakage rate is expected to be no greater than 0.63 gpm.

Since the assumed leakage rate is greater than the conservative calculation, the current licensing basis 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

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specification plugging limit does not pose a concern relative to the health and safety of the public.

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COMPENSATORY MEASURES i

Extensive measures have been previously taken by ANO to enhance the operator's ability to detect and respond to steam generator tube leakage. Additionally, ANO has previously implemented more restrictive shutdown limits based upon primany-to-secondary leakage than those required by the technical specifications. Since these measures were already in place, no additional compensatory measures were determined necessary to address the surveillance i deficiency. A summary of ANO-l's detection and monitoring capability, shutdown limits, operator guidance, and training is provided below.

The methodology for monitoring the secondary system for leakage includes the use of process '

monitors to check radiation levels in the condenser off gas, N-16 gamma levels from the OTSGs, chemistry samples, and RCS mass balances to calculate leakage. Additionally ANO-4 0

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• I has a procedural limit of 0.1 gpm (144 gpd) that is more conservative than the 0.347 gpm .

(500 gpd) limit allowed by Technical Specification (TS) 3.1.6.3.b. Management has (

previously established a conservative administrative limit of 0.069 gpm (100 gpd) at which, upon confirmation, a plant shutdown is to be initiated. m Operations personnel trend information from the steam, condenser off-gas and OTSG process .

monitor systems to determine indication of an OTSG tube leak. _ Steam lines are monitored by radiation monitors and N-16 gamma detectors that provide chemists and operators with the capability of promptly detecting primary-to-secondary leakage.

l The amount of N-16 present in the secondary system is influenced by the size of the leak, location, and the power level. ANO-1 utilizes scintillation type detectors as N-16 monitors. ,

l These monitors are normally selected to measure gross activity from the OTSG but are  :

selected to monitor N-16 in accordance with Abnormal Operating Procedure (AOP) guidance j l for small OTSG tube leaks. The monitors provide input to control room annunciators >

associated with OTSG tube leakage. These N-16 monitors have only a single point correlation ofleakage to an N-16 reading based on 100% power level. Guidance is given in AOP 1203.023, "Small Steam Generator Tube Leaks", to correlate an N-16 reading of 1x10E4 CPM as being indicative of tube leakage of 2 0.1 gpm and a 500 CPM change being indicative of a 0.01 gpm change in primary-to-secondary leakage at 100% power. I ANO-1 installed high sensitivity N-16 detectors in January 1997 to enhance detection of small changes in primary-to-secondary tube leakage at various power levels. A modification was made to the plant computer to allow monitoring of both the original and newly installed N-16 monitors to provide a readily visible indication of changes in count rate due to changes in leakage. The plant computer input has an alarm that can be used to actuate a control room annunciator panel to alarm at a value set by operators. Guidance for the use of the new N-16 detectors for monitoring primary-to-secondary leakage _ is given in Operations Information Notice (OIN) #44.

The condenser off-gas monitor is an in-line detector on the combined suction line of the condenser vacuum pumps. It is a gamma sensitive scintillation detector that provides a means to measure the gaseous activity levels released to the system vent. The monitor provides displays and an alarm in the control room to alert operators of a possible OTSG tube leak.

The main steam high range radiation monitors are Geiger-Mueller type detectors. These detectors provide input to the Safety Parameter Display System (SPDS) for display in the front of the control room.

The plant computer leak rate program provides operators the ability to validate indications of primary-to-secondary leakage by observing changes in the Reactor Coolant System (RCS) j mass inventory. This program allows detection of changes in the make up tank level and L determination ofleak rate changes based on the time interval selected.

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. . The SPDS is also available for use by operators. This system has a screen dedicated for use during suspected or actual primary-to-secondary leakage events. The " Steam Generator Tube Rupture" screen contains N-16 readings (from the original detectors), condenser off gas, RCS Avg. Temp (Loop A/B), OTSG Tube-to-Shell delta T (OTSG A/B) and T-Sat for the OTSGs. In addition, the SPDS graphics display is outlined in red and flashing when a parameter on the graphics display is in alarm.

The' Chemistry Department routinely analyzes and trends samples from the RCS and secondary water systems to identify and quantify primary-to-secondary leakage. Off-gas .

samples taken from the condenser vacuum pump discharge are analyzed for Argon-41 activity. i Liquid condensate samples are analyzed for tritium to quantify activity levels in the secondary system. Argon-41 levels yield a better measure of instantaneous levels of primary-to-secondary leakage. Tritium levels in the secondary system increase linearly over time during a primary-to-secondary leak. A primary-to-secondary leak rate can also be determined from the l

- tritium analysis. Secondary liquid samples are also routinely analyzed for fission product activity using gamma spectroscopy. An AOP directs special sampling by the Chemistry l Department until primary-to-secondary leakage is reduced below 0.1 gpm or the reactor is l tripped.

The Operations and Chemistry Departments utilize available information to detect changes in primary-to-secondary leakage and to initiate actions to place the unit in a safe condition.

Procedures such as Emergency Operating Procedure 1202.06, " Steam Generator Tube Rupture," AOP 1203.023, "Small Steam Generator Tube Leaks," and the 1203.012 series for j annunciator corrective actions are utilized when the monitors, indicators, trends, or i annunciators exhibit changes indicative of the development of, or change in, primary-to-secondary leakage. The Operations Department uses these procedures to place the plant in a stable condition and to mitigate the consequences of an OTSG tube leak.

Finally, ANO maintains thorough training oflicensed operators by using the plant simulator for primary-to-secondary tube leaks and ruptures. This insures familiarity with the symptoms and indications of this event to enable timely diagnosis and action for placing the unit in a safe condition.

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" ' DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION 4

i An evaluation of the proposed change has _ been performed in accordance with '

L 10 CFR50.91(a)(1) regarding no significant hazards considerations using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this amendment request follows:

i Criterion 1 - Does Not Involve a Significant Increase in the Probability or  :

Consequences of an Accident Previously Evaluated.

1 i

The steam generators are used to remove heat from the reactor coolant system during normal l

operation and during accident conditions. . The steam generator tubing forms a substantial

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portion of the reactor coolant pressure boundary. A steam generator tube failure is a violation

. of the reactor coolant pressure boundary and is a specific accident analyzed in the ANO-1 Safety Analysis Report.

l The purpose of the periodic surveillance performed on the steam generators in accordance with ANO-1 Technical Specification 4,18, is to ensure that the structural integrity of this

portion of the reactor coolant system (RCS) will be maintained. The technical specification )

i (TS) plugging limit of 40 % of the nominal tube wall thickness requires tubes to be repaired or j i

removed from service because the tube may become unserviceable prior to the next l

i inspection. Unserviceable is defined in the TS as the condition of a tube ifit leaks or contains a defect large enough to affect its stmetural integrity in the event of an operating basis earthquake, a loss-of-coolant accident, or a steam line break. Of these accidents, the most j severe condition with respect to patch intergranular attack (IGA) degradation within the l

upper tube sheet is the main steam line break (MSLB). During this event the differential pressure across the tube could be as high as 2500 psid. The rupture of a tube during this event could permit the flow of reactor coolant into the secondary coolant system thus bypassing the containment.

t  :

From testing performed on simulated flaws within the tubesheet it has been shown that the  ;

patch IGA indications within the upper tubesheet left in service during 1R13 with potential depths greater than the plugging limit, do not represent structurally significant flaws which  !

would increase the probability of a tube failure beyond that current.ly assumed in the ANO-1 i Safety Analysis Report.

Burst tests were conducted on tubing with simulated flaws within the tubesheet. In these tests, through-wall holes of varying sizes up to 0.5 inch in diameter were drilled in test specimens._ The flawed specimen tubes were then inserted into a simulated tubesheet and pressurized. In all cases the tube burst away from the flaw in that portion of tube that was outside the tubesheet. The size of these simulated flaws bound the indications left in service within the upper tubesheet during 1R13. These tests demonstrate for flaws similar to the patch IGA found in the ANO-1 upper tubesheet that the tubes will not fail at this location under accident conditions.

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f The dose consequences of a MSLB accident are analyzed in the ANO-1 accident analysis.

This analysis assumes the unit is operating with a 1 gpm steam generator tube leak and that

) the unit has been operating with 1% defective fuel.

4 j Increased leakage during a postulated MSLB accident resulting from the patch IGA left in service in the upper tubesheet is not expected. IGA has been present in the ANO-1 steam -

generators for many years with no known leakage attributed to this damage mechanism. ,

Because of its localized nature and morphology, the flaw does not open under accident  ;

pressure conditions.  ;

i l This change allows continued operation with IGA indications within the upper tube sheet with i F the potential of through-wall depths greater than the technical specification plugging limit.  ;

Continued operation with these flaws present does not result in a significant increase in the

probability or consequences of an accident previously evaluated for ANO-1.  !

Therefore, this change does' no.t o involve a significant increase in the probability or ,

consequences of any accident previously evaluated. j

, Criterion 2 - Does Not Create the Possibility of a New or Different Kind of Accident i from any Previously Evaluated. l 4

The steam generators are passive components. The intent of the technical specification

surveillance requirements are being met by this change in that adequate structural and leakage integrity will be maintained. Additionally, the proposed change does not introduce any new modes ofplant operation.

Therefore, this change does p_o1 create the possibility of a new or different kind of accident from any previously evaluated. ]

Criterion 3 - Does Not Involve a Significant Reduction in the Margin of Safety.

The ANO-1 Technical Specification Bases specify that the surveillance requirements (which -

includes the plugging limits) are to ensure the stmetural integrity of this portion of the RCS pressure boundary. The technical specification plugging limit of 40% of the nominal tube wall thickness requires tubes to be repaired or removed from service because the tube may become ,

unserviceable prior to the next inspection. Unserviceable is defined in the technical i specification as the condition of a tube ifit leaks or contains a defect large enough to affect its structural integrity in the event of an operating basis earthquake, a loss-of-coolant accident, or i a MSLB. Of these accidents the most severe condition with respect to IGA within the upper 3

tube sheet is the MSLB.

Testing of tubes with representative IGA flaws removed from ANO-1 OTSGs during IR13, showed the flawed tubes to be capable of withstanding differential pressures in excess of a

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10,000 psid without the presence of the tubesheet. Testing of simulated through-wall flaws of q up to 0.5 inch in diameter within a tubesheet showed that the tubes always failed outside of the tubesheet. Thus the structural requirements listed in the bases of the technical specification is satisfied considering this change.

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[ Leakage under accident conditions would be limited due to the small size and morphology of i the flaws and would be low enough to ensure offsite dose limits are not exceeded.

Therefore, this change does g involve a significant reduction in the margin of safety.

{ In conclusion, based upon the reasoning presented above and the previous discussion of the j- amendment request, Entergy Operations has determined that the requested change does g 1 i involve a significant hazards consideration.

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• - STATEMENT OF EXIGENT CIRCUMSTANCES i 10 CFR 50.91(a)(6) states that whenever an exigent condition exists, a licensee requesting an

amendment must explain why this exigent situation occurred and why it could not be avoided.

! i During the IR13 refueling outage, an eddy current technique was used for the satisfactory

[ completion of the ANO-1 steam generator inspection surveillance. The technique used had I been qualified per Appendix H of the EPRI "PWR Steam Generator Tube Examination I Guidelines." This technique was used to depth size all intergranular attack flaws within the 2

upper tubesheet. As required by the technical specifications, all upper tube sheet IGA i indications with a depth size of greater than the plugging limit as determined by the qualified sizmg technique, were also removed from service by plugging.

i

During the steam generator inspections, three tube samples containing upper tubesheet IGA

! flaws were removed from the "B" OTSG and sent offsite to be antlyzed for future

! development of an alternate repair criteria and to further support the quali6ed eddy current sizing technique employed during refueling outages. The preliminary destructive examination i results were recently received by the ANO staff. This data arrived approximately 5 months

after the resumption of operation following the steam generator inspections that occurred j during IR13. These results indicate that the flaw depths do not correlate well with the depths

! sized using the qualified eddy current technique. Upon further review, ANO has determined  ;

i that the application of the sizing criterion is no longer valid. With the qualified sizing j

technique invalidated, there is a potential that tubes could have been left in service with  !

indications that have through-wall depths greater than the plugging limit specified in the

- technical specifications. This would be considered a condition that is not allowed by the ,

- technical specifications. Prior to the receipt of the preliminary destructive examination results, ANO had no reason to question the adequacy of the steam generator inspections that occurred  !

i during IR13.

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Attachment to

- ICAN049703 Page11of11

'In order to continue plant operation in non-compliance with technical specification 4.18 enforcement discretion was verbally requested and received from the NRC on April 9,1997.

Enforcement discretion was requested for a period of time necessary for the NRC to process this technical specification change which will allow continued operation in the current configuration for the remainder of the operating cycle. Enforcement discretion was granted for this purpose but was limited to May 7,1997, after which time, the actions of TS 3.0.3 are 1 required to be followed. This change must be processed expeditiously to meet the May 7, 1997 deadline. Therefore, Entergy Operations requests that this proposed technical specification change be considered under exigent circumstances as described -in .

10 CFR 50.91(a)(6).

i 1

Attachment 1  !'

Leak Rate vs. Crack Length at SLB,

' ANO OTSG Tubing .

100 l I I I

i

/

, /

f

/ <

10 f

/

/

/

s. ,/

.) ,

a 1 ,

a

/

E i CL /

cm e

$ /

i e j 5 .

s 0.1 ,

l

,/

4

/

/

0.01 ,

/

/

4-I

/ .

0 i ( \

/ 1 0.001 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 i crack length, inches

et 4 e i

1 l

MARKUP OF CURRENT ANO-1 TECHNICAL SPECIFICATIONS (FOR INFO ONLY) i i

1

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