ML20137V330

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Forwards Request for Enforcement Discretion Re Steam Generator Tube Surveillance
ML20137V330
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 04/09/1997
From: Mims D
ENTERGY OPERATIONS, INC.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20137V318 List:
References
1CAN049702, 1CAN49702, NUDOCS 9704170264
Download: ML20137V330 (27)
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Subject:

ArkansasNuclear One-Unit 1 l Docket No. 50-313 l Tle==3 No. DPR-51 Raquest ForEnforcement Discretion l Gentlemen: i i This letter 6 aaats the Arkansas Nuclear One, Unit-1 (ANO-1) position (enclosed) discussed j on April 9,1997, with members of the Nuclear P=u1='ay Commission staff. During this i discussion, ANO formally requested enforcement discretion firom tlw requirements of Technical j S,dretion 4.18 regarding surveillance of Once Through Steam Generator tubes. This enforcement discretion was requested in order to allow sumcient time for the subrmttal and NRC l review and approval of an exigent change roguest for a one time exception to the requirements of Section 4.18.5.b. This exception will allow tubes with Intergranular Attack indications within the upper tube sheet with potential through wall depths of reeter than the plugging limit to remain in S l service for the remainder of cycle 14. A markup of the proposed change is attached. 4 The attmehd request was reviewed and approved by the ANO Plant Safety Committee at 4 i appr@a'aly 0800 CDT on April 9,1997. Verbal approval of this enforcement discretion request was received at 1535 CDT on April 9,1997. This discretion wiE '.>e in effect until May 7, j 1997, or until the Staff acts on the proposed technical specification change request, whichever occurs first. i Very truly yours, O 2 Dwight C. Mims Director, Nuclear Safety i DQd/rbs 9704170264 970409 ATTACHMENT 1 POR ADOCK 05000313 G PDR g

~ ~ ~~SKO-710.@ bioc a:01~ ~ ^ ^ ~ ~ ~ ~ ~~ ^ sc1 'utes ~~ ~ ~ ~ ~ ~ ics' r. cam 66:39 ~ ~ - ~ ~ ~ T-U. S. NRC Apr0 9,1997 ICAN049702 PAGE 2 - cc: Mr. EEs W. Marschoff Ragions: Adminisuator U. 3.Nucisar Regulatory Comminston RegionIV 611 Ryan PlazaDrive, Suite 400 Arlington, TX 760118064 NRC Senior Resident faMor Adansas Nuclear One P.O. Box 310 f =daa. AR 72847 Mr. Geoqp Kalman NRR Prcdect Manager Ragion IV/ANO-1 & 2 U. 8. Nuclear Regulatory Commission NRR Mail Stop 13-N-3 One White FlintNorth 11555 Rockville Pike Rocicvdle, MD 20852 4 e

- gg p,gpg g i i REQU3ST FOR ENFORCEMENT DISCRE110N REGARDING ONCE THROUGH gTEAM GENERATOR TUBE SURVEILLANCE l Denedntion of CMPla=hamira-ts fbr wkleh umferemetDiagrationABeaussted l l Intergranular Attack (IGA)is known to be present above the 15th Tube Support Plate (TSP) within the ANO 1 Once Through Steam Oensrators (OTSGs) as veri 5ed by destructive l examination (DE) flrom 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 l during the enriy years ofplant operation. The cantaminant causing IGA of the ANO-1 tubing is sulfbr as a result of thermal decomposition ofion exchange resins. The ANO-1 IGA can be i . categorized as volumetric, or " patch-like", with no specific orientsuon. Since discovery, there has l been no evidence ofleakage fromIGA Saws at ANO-1. ? During the IR13 ro6mling outage, an oddy current (EC) technique was employed to depth sim the IGA. This technique had been quali5ed per Appendix "H" of the EPRI"PWR Steam i Generator Tube **= a.a chm.u " This technique was used to depth size allIGA flaws j within the upper tubenheet (UTS). During this ins =% 25% of a!! indications detected within i j the UTS region by bobbin coil were examined using Rotatirng Pancake Coil (RPC) to characterize these Aaws. AH IGA indications between the 15th TSP and the aaaaaday fhoe of the UTS were j removed from service by plugging. All UTS IGA indications with a depth size of 240% 1 through-wall (TW), as determined by the quali5ed sizing technique, were also removed from ) service by plugging. i j Three tube samples containing IGA flaws were removed firam the "B" OTSG for future development of an Alternate Repair Criteria (ARC) and to fbrther support the qualified EC sizing ] technique employed during the IR13 refbeting outage. ( j Preliminary DE resuhs ofIGA flaws contained within tube sarnples removed from "B" OTSG indicate that the Saw depths do not correlate well with the depths sized using the qualified EC technique. The entire data set, including the resuhs of the most recent outage, still satis 8es the qualihtan requirements of Appendix H. The inservice inspection of the ANO-1 steam generators is conducted in accordance with ANO-1 Technical 8,=# Mons 4.18. Sp**= ion 4.18.2 states: "Inservics bsqpecrton qfmeam gerwrakw tubig dallinclude non destructin stamiration by edQ. current testiq or other spholent archnipes." spa % 4.18.3 requires that a minimum sample size be examined in eooordance with speci6 cation 4.18.5. Speci5estion 4.18.5.b. notes. "7ha mean generstorWall be determinedoperable q$er ecoqpisthng the correqpondig actions plug or sleen all tubes

  • "**diq du p%ggtq ltatt avut all tubes contatnkng throughmall crackr) reptred by Tkble 4.18-2." Table 4.18 2==W the expansion criteria for sampling ofthe steam tubes and requires "defecttw" tubes to be plugged or sleeved. Speci6 cation 4.18.5 Dekctas i61\\ows: *an in\\ perfection ofsuch newrity akat ir exceeds thepluggiq limit except where the

umunewmunwinn ummuu sarvwmr7nna U. E. NEC e April 9,1997 ICAN049702 PAGE 2 inperfection har been Jpanned by the installatiors ofa sleen. A tube conminirs a defect in its pressure basembry is dvecttw." Nying Timfr is Mnari in the same specification as follows: "the inperpction depth at or beyondwhich the tube shall be restored to serviceability by the trasullackm ofa sleen or i--M.)han arvloe because it may become acuerviceable prior so l the next inspeeden; it is equal no #% qf the noseinal tube well skicknest.' l The Bases fbr spacMendan 4.18 states #7he surwfllance requirensentsfor inyection of the sesam gersmawr tubes ensure ahat the sinocturalintegrity ofthisportion of the RCS will be maintained theprogram,tbr inservice inpaction qfsnnan sensranar tubes is basedon a apodttfondon qfRegulatory Gadde 1.83, Ransion 1." l Criterion !X, " Control of Special Processes," contained in Appendix B to 10 CPR Part 50 states, j in part, that "Adessures shall be==tahli=hari to assure that special processes, including j nondestruedve testing, are controlled and accomplished by qad6ed personnet using qualised procedures." l Criterion XI, " Test Control," requires, in part, that a test program be established to assure that s!! testing required to demonstrate that structures, systems, and components will perform i satisfhetonly in service is identi6ed and perfonned in accordance with written test procedures I which incorporate the requir= nanta and acceptance limits contained in applicable design dan, manta, i To demonstrate the eddy current bobbin coil's ability to adequately depth size IGA patches within the upper tubesheet, AND nuaHGad a technique in accordance with Appaadiv "H" ofEPRI"PWR Steam Generator Tube Rumminatian Md-H==." Revision 4 dated June 1996. Compliance with the EPRI ="Idalia* was considered an aa~; table method to qualif'y non-destructive memminatian l (NDE) techniques ibt the detection and sizing of damage mechanisms. This was the only l qualification Wians available at that time. i The??H84 data set demonstrating the capability of the inspection process consisted of j eervice-degraded tube specimens (i.e., tube samples removed from the ANO-1 and Crystal River i steam generators). The nondestructive examination parameter responses for the Crystal River j tubes were fldly ounaistent with the nondestructive examination parameter responses of the i ANO 1 flaws. During 1R13, three tubes with bobbin indications within the upper tubosheet were removed from the steam generator. Two of the three tubes contained flaws that would have required repair. 3 The third tube was near the repair limit and may have been preventively repaired. The tubes were enlacted on the basis of their containing multiple indications with depths representative of the sverage indicadon depths as sized by EC AAer burstin6 the tubes in the Ishorntory, the Asws were esamined and sized. If a flaw was not opened by the burst of the tube it was bent open for destrucdvs examinadon. The DE results are not consistent Mth the previous qnmHAendon data of the bobbin ooil for sizing IGA Saws in the upper tubesheet. % reason for the inconsistency in 1 sizing IGA in the upper tubenheet is still under review. As a result af this condition, it is possible 1 4 i

04-HFU( SU @ Urse e t1U U W1gDIUd ~~ ~ ~~ 9-4ts' P.CC/15 JoL-IP U. g. NRC April 9,1997 ICAN049702 PAGE 3 that tubes were leR in service with through-wall defisets greater than the technical =r=?= dons plugging limk. I Theredbre, ANO is requesdag enfbroomers discretion regarding technical a,W% 4.18 to ) allow sufEcient time for submittal and NRC review and approval of a proposed techmcal speelflestion change requent fhr a one-time exception to the requirements of Section 4.18.5.b which will eBow tubes with IGA delbots within the upper tube sheet with potential through-wall depths greater then the plugging link to remain in service for the remainder of cycle 14. j Since the subject flaws do not represent a structural or leakage concern, ANO has concluded that i the presence ofinnervice upper tubesheet IGA defbets with through wall extents that may escoed the technical speci6 cation plugging limit does not poes a concern relative to safe operation of the plant or the beehh and safety of the public. The basis for this conclusion la presented below. Initiation of a plant shutdown to p. Gero the subject survelHance action would increase the i potential for a plant upset. Although this increased potanhal has not been raamad, it is j considered by ANO to represent a greater risk than continuing to operate with the =%3 tube i flaws for the remainder of the current cycle. Therefore, NOED eriterion B.1.(a) is considered j applicable to this request. l k i Compensatory Measures l Extensive measures have been previously taken by ANO to enhance the operstars ability to detect and respond to steam generator tube leakage. Additionally, ANO has previously implemented ) more restricdvs shutdown Mmits based upon primary-to-segundary leakage than those required by the technicalupecifications Since these measures were already in place, no additional i cosapensatory measurus were deter-land necessary to addreas the surveulence de6eiency. A i summary of ANO l's detectiontmanittring capability, shuutawn limits, operator guidance, and i trainingis 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 calculata leakage. Additionally ANO-1 has a procedurallimit oro.1 spm (144 apd) that is more conservative than the 0.347 gym (500 spd) j Hmk avowed by T.shnical spwinemlon (Ts) 3.1.6.3.b. Manaewnant has previously established a conservative administrative limit of 0.069 gym (100 gpd) at which, upon confirmation, a plant shut downis to beintslated. ) Operations personnel trend information Som the steem, condenser off-gas and OTSG process monhor syneen to duermine indiendon of an orso sub. l k. st line are h ed by radiation monitors and N-16 gamma detectors that provide chemists and operators with the capatnuty orgie yW detecting primary 4ogecondary leakage. The amount ofN-16 present 'm the secondary system is 'm8uenced by the size of the leak, loostion, and the power level. ANO 1 utilises saintilation type detectors as N-16 monitors.

umrtate e r V M m Job-m U. S. NRC April 9,1997 ICAN049702 PAGE 4 These monitors are normally selected to measure gross activity fhun the OTSG but are selected to snonhor N.16 in accordanc: _vith#mormal Operating Procedure (AOP) guidanos for small OTSO tube leaks. The monitors pmvide input to control room annunciators naaelatad with OTSO sube leakage These N-16 monkors 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 Isaks", to correlate an N 16 reading of 1x10PA CPM as being indicative of tube leakags of k 0.1 spm sad a 500 CPM change being indientive of a 0.01 apm change in prbnary-i te :n?

  • yleakage at 100% power.

ANO 1 instaued high sensitivhy N-16 detectors in January of1997 to enhance detection of small changes in primary to secondary tube leakage at various power levels A madm=tian was made to the plant computer to allow manharing of both the original and newly instaBed N-16 monitors { to provide a readHy visible irdattari of changes in count rate due to changes in leakage The plant computer input has an clarm'that can be used to actuate a comrol room annunciator panet to i alarm at a value set by operators. Ouldance fbr the use of the new N-16 detectors for monitoring primary-to-secondary leakage is given in Operations Infbrmation Notice (OIN) #44. The OIN yid.ds shut down guidanos for a step change in N-16 count rate ofgreater than 500 opm in the "B" OTSG in less than one hour. This c(.,ii j, ands to a 60 spd leak that might be indicative of a } tube beginning to fhil. However, this correlation has not been done for the "A" OTSG since there j has been noleakage to trcad. 1 % aaa4===r adr-gas monitor is an in-line detector on the oombined maation line of the l aandaa=r vacuum pumps. It is a gamma sensitive scintillation detector that provides a means to i measure the gaseous activity brvels 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 ofthe controlroom. I 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) mass inventory. This program allows detection of changes in the make up tank level and determination o(leak rate changes based on the time interval selected, i j The SPDS is also available fbr use by operators. This system has a screen dedicated for use during suspected or actual primany-to-secondary leakage events. The " Steam Generator Tube j Rupture" screen contains N-16 readings (fkom the original detectors), condenser off gas, RCS j Avg. Tenp (Loop A/B), OTSG TubwShen deka T (OTSG A/B) and T-Sat for the OTSGs. In j eddition, the SPDS graphics display is outlined in red and flashing when a parameter on the i graphics displayisin alarm. i

  • Ihe N*y Department roudnely analyzes and trends samples Dom the RCS and secondary water systems to identify and quantify primary-to-secondary leakage. Ofr-gas samples taken from j

the condenser vacuum pump discharge are analyzed for Argon-41 activity. Liquid condensate i 4 I 1 1 ~. .i

i* U. S. NRC l Apru 9,1997 ] ICAN049702 PAGE 5 j ~ l samples are analyzed for tritium to quantify activity levels.in the secondary system. Argon-41 1 levels yield a better measure afinstantaneous levels of primary-to-secondary leakage. Tritium j levals in the mamadary system increase linearly over time during a primary-tc ::x=4'y leak. A { primary to secondary leak rate can also be determined kom the tritium analysis. Secondary liquid er.+k are also routinely analysed for fission product activity using gamma spectroscopy. An AOP directs special sampling by the Mi*y Department untu primary-to-secondary leakage is reduced below 0.1 SPm or the reactoris tripped I ne operations and Chemistry Departmems udlize available infbrmation to detect changes in primary tc :::==' y leakage and to initiate actions to plam the unit in a safe condition. j Proceduree are provided such as Emergency Operating Procedure (EOP) 1202.06, " Steam Generator Tube Rupturt," AOP 1203.023, "Small Steam Generator Tube Imaka," and the I203.012 series fbr annunciator corrective actions are utilland when the monitors, indicators, l trends, or annunciators exhibit changes indicative of the developmem of, or change in, primary-to-secondary leakage The Gyations department uses these procedures to place the plant in a stable canditian and to mitigate the consequences of an OTSO tube leak. t Finaliy, ANO maintains thorough W.;.

  • g oflicensed operators by using the plant simulator for i

primary to secondary tube leaks and ruptures. 'Ihis insures fhmiliarity with the s@T4 and indications of this event to enable timely diagnosis and action for placing the unit in a safe condition. l l Evaluation ofEafstr Einnificance The subject addy current alzing technique was employed for IGA defects within the UTS. All UTS IGA indioetions w' h a depth size of 240% TW were removed from service by plugging the n af5ected tubes. l The three UTS IGA tube samples removed during 1R13 were sutdocted to room temperature i burst testing Burst testing was performed separately within the flawed and unflawed regions of 1 1 the tube samples. No simulated tubesheet was employed during the tests. The tests were i performed using bladders in the flawed region. No foils or lateral restraint systems were used. l The burst pressures fbr the flawed regions were between 10,000 and 11,000 psig. The unflawed regions burst at pressures between 10,700 and 11,200 peig. For ANO-1 OTSGs, structural integrity is conservatively demonstrated by pressurizing the steam generator tubing to three times normal operadng diffbrantial pressure. This pressure ibr ANO-1 is 3765 psig. The burst testing i resuhs indicate that substantial strucmral margin exista. In 1996, to support ANO's study of1GA, burst testing ofpm. defected tubes was completed by Framstome Technologies Inn. (PTI).1he burst testing consisted of nine tubes containing through wall drilled holes up to 0.5 inches in diameter and one tube==lalaa no defbets placed i within a simulated enhahad. Nine of the specimens burst at pressures it 10,941 psig. Each tube burst outside the tubenheet within the non-defected portion of the tubes One tube reached a 1 l i m y ,..y

aywunumr wounun aume nanAWGMDFWr l U.M.NRC April 9,1997 ICAN049702 PAGE 6 pressure of 9,577 psig but did not burst due to bladder leakage. These test rekilts la%te that the tubenheet provides sufEcient support to preclude tube rupture within the tubenheet The tube samples removed Dom ANO-1 in 1996 included eleven IOA indications in the UTS. ] Since it was con 6rmed that the inservice IGA indications are volumetric, boh amplitude (vohage) was used as a Wag parameter. The eddy current responses Rom these flaws were compared with the population ofinnervice IG A indications to determine how representative the i Saws were of those remaining in service. The 600 KIh bobbin coil signal amplitude offlaws in tubes that were pulled during 1996 ranged Anm 0.46 to 2.69 vohs. Of the 470 inserwoe IOA l indM=a. all are bounded by the 169 value. AdditionaDy, a K,n.ysil6on ofRPC data was performed to fluther substantiate that the pulled tube i flaws bound those indications remaining in service The RPC data collected for the tube puH l-samples resuhed in a maximum flaw extent oro.16 inches RPC signal information was callacM l on 118 irvliendaan within the UTS. Ten of the largest RPC voltage indications were examined to i determine the length-by-width extent by RPC. The largest RPC extent for those IGA indications len in service was 0.14 inches. Therere, k is concluded that the inservice IGA indications are bounded by those tube samples th:2 were destructively aramhwi. i Structural integrity of the tubing within the tubesheet is assumd based upon comonstration of the following: l A. The actual tube samples removed kom ANO-1 during 1R13 exhibited burst pressures that substantially exceeded the required strucharal limit. 4 l B. The structural support provided by the tubeeheet precludes tube rupture. i C. The inservice IGA indications are bounded by those flaws contained in the tube samples that were pulled. The IGA patches destructively====*=1 were not through-wall; therefore, normal operating pressures did not result in through-wallieskage. This was evident during inservice inspection of the tubing in which no iniications of residual leakage was noted. A comparison of 1R12 ar.d IR13 reibeling outage TL algnatures indicates that the IGA exhibits little or no growth. Also, comparison oflaspartian data prior to the IR12 r** Hag outage supports this conclusion. AdditionaDy, during May 1996, "B" OTSO tubing was subjected to a differential pressure of sporoximately 2100 paid ibt several hours as a result of a lhedwater transient. No immediate b.xsase in pdmary to secondary leakrate was noted during the event or ibuowing startup. The primary-to-secondary leskrate did increase by appravimataly la spd three days ibliowing staitup; however, none of the leskage detected daring the 1R13 reibeling outage was hom 1GA naws. It is concluded that leakage through IGA fiswa in the UTS is highly unlikely at Main Steam Line Brealt (MSLB) pressures due to the flaw morphology and the near MSLB diftkrantial pressure that occurrod in May 1996 with no resultant leakage.

m~ awauuu r-ugrvww m-uw i U. S. NRC 3. j Aptil 9,1997 ICAN049702 PAGE 7 4 1 1 Conditional core damage probability is the increase in core damage toquency due to a given } ar-I% other than that assumed for the base PRA. The PRA assumed that the tube integrity is ) i such that no steam generator tube rupture would be inducul due to transient conditions. N Hmhing licensing basis transiem which could most adversely affbet the tubes by creating a high l differential pressure across the tubes is a MSLB Accident This accident could produce a tube differemial pressure of up to 2500 paid. The tube sample burst pressures were well above pressures which would be seen in a limiting MSLB accident. Thus, the likelihood of tubes suptuting is not increased because of the larger than =q=*M flaw sizes due to IGA in the UTS. This situation has been=="%ely assessed and the aandidanal core damage probabluty ibr this condhion is satimated to be F+-- 7= dal l j The limiting licensing basis accident with respect to dose consequences Dom induced tube leakage la the MSLB accident. This accident assumes a total leakage of I spm with 1% lhiled ibel in the l core. However, steam genwator tube leakage is procedurally umited to 0.1 spm during nonnal i operation. Even though leakage is not expected to occur, MSLB induced tube leakage has been i conservatively estimated to be 0.53 apm on the affected steam generator. The fbliowing rr, M are made concerning the number of saws and associated leakage: 1 l !) Since steam generator A has the largest number ofIGA patch ladicatiana (285) k was chosen j as the affected generator bounding steam sencrator B with only (185) indicati^ns.

2) Halfof the indioetions are assumed to neek under MSIR oonditions.

l

3) Representative leakage values for axial flaw lengths were utilized to bound the leakage expected AomIGApatches.

[

4) Applied the longest IGA length calculated kom RPC data to the 50% W don assumed to l

leak. I l

5) Assumed the flaws grew in length an additional 25% over the cycle
6) 50% of the flaw 1:agth will be 100% TW in depth.

Since there are 285 indications, half of this value will be 143. The longest length in the axial plar.e j was 0.14 inehas When increased by 25% this yields a flew length of: i j 0.14 inches * (1.25) = 0.175 inches If 50% of the length is assumed to be 100% TW: 0.175 inches * (0.5)= 0.0875 inches Using leakage curves developed for OTSO's for axial flaws, the leakage kom a single flaw i (0.0875 inch,100% TW) is determined to be 0.0025 spm. To compensate for normal operating temperature the value is multiplied by 1.47 to yield a finalleakage of 0.003675 spm por flaw. l 1 I i .m u

_g, j U. g. NRC l April 9,1997 ICAN049702 PAGE B l ' Ibis value is than = man by the number of potential leaking flaws to give a total leakage of: l 143 flaws

  • 0.003675 gym / flaw = 0.53 gym When the estimated leakage in the afBected steam generator is added to that which is allowed by procedure, the total leakage rate is expected to be no greater than 0.63 spm. Sissce the assumed l

lesknas rate is greater than slw conservative + -" the current licensing basis assumption of I spm r==ains bounding. i' The subject flaws do not represent a structural or leakage concom. Thereibre, the presence of inservice upper tubenhest IGA delbots with through-wall extents that may exceed the technical aparinaadaa plugging limit does not pose a conomen relative to the heahh and safety of the public. 3 1 Enstneerlan Eraluation and Basis For No BinalReant Hazard Consideration An evaluation of tim proposed NOED request has been prbrmed in nooordance with 10CFR50.91(a)(1) regarding no significant hazards cormuration using the standards in 10CFR50.92(c). A discussion of those standards as they islate to this request follows: Crherion 1 - Doss Not Involve a C=-?= Increase in the Probability or Consequences of an Aooident Previously Evaluated. The steam generators are used to remove heat firom the reactor coolant systaru during normal operation and during accident nanditiania The steam generator tubing fbrms a substantial portion of the reactor coolant pressure boundary. A steam generator tube failure is a violation of the ruector coolant pressure boundary and is a specific sooident analyzed in the ANO 1 Safety Analysis Repost. The purpose of the periodic surveillance performed on the steam generators in accordance with ANO-1 Technical SWdaa 4.18 is to ensure that the structural integrity of this portion of the RCS will be maintai-d. The techrical specification plugging lindt of 40% of the nominal tube wall thickness requires tubes to be repaired or removed flrom service because the tube may hanarm unserviceable prior to the next la==+1aa Unseniceable is defined in the technical =paalAaadaa= as the manditian af a tube ifit leaks or aaa' alas a defbet large enough to afBect its stmetural integrity in the event of an operating basis eaWaah a loss-of-coolant accident, or a steam line break. Of these.,,u.nea, the most severe aanditian with respect to IGA degradation within the UTS is the MSLB. During this event the differential pressure across the tube could be as high as 2500 paid. The rupture of a tube during this event could permit the flow of resator coolant into the secondary system thus bypassing tin containment. From testing perfbnned on simulated flaws within the tubenhest it has been shown that the patch IOA indications within the upper tubenheet left in service during 1R13 with potential depths greater than 40% do not represent structurally signi5 cant flaws which would increase

ADW UiYtrVimwum i U. S. NRC April 9,1997 1CAN049702PAGE9 the probability of a tube fhilure beyond that currently assumed in the ~ R8 Port. surst tests were eaad ad on tubing whh airnulated saws wkhin the tubestem. In these tests, through-wall holes of varying elses up to 0.5 inch in diameter were specimens. The flawed specimen tubes was then laserted into a sim pressurised. In all esses the tube burst away nom i within the upper tubenheet during 1R13. These tests demonstrate 1br naw patch 1GA fbund in the ANO 1 upper tuhmakaar that the tubes will no j i under accident anaditiana. l The done consequences of a MSLB anddent are analyzed in the ANO-1 4 l This analysis assumes the unit is operating with a 1 spm steam generator l the unit has been operating with 1% defective lhel. 4 Inoressed neskage during a postulated MSLB sooident resuhing torn the I l service in the upper tubesheet is not expected. IGA has been present in th generators fbr many years with no known lask i i l pressure conditions. Therefore, the NOED which allows continued operation with IGA flaw whh potential through-wall extents greater than the technical aparm { does not result in a sip %=^ increase in the pmbability or cery*= of an ~ l I previously evaluated for ANO-L e Critarion 2 - Does Not Create the Possibility of anew or Different Kind of.Aaald ( l tom anyPreviouslyEW'ad l The steam generators are passive components. The intent of the tec surveillance requirements is being met by this change in that adequ l leakage integrhy will be maintained. The proposed change intro j j plant operation. i Thorstbre, the NOED does not create the possibility of a new or diffb e l eom any previously evaluated, i BA etaainthe Margin of Safety. l Gdsadena - Does Not Involve a Significant The ANO-1 Technical Specification Bases specify that the survatti l menudse the P usging limits) are to ensure the structural integrity l pressure boundary. The technical specification plugging limit o l j well thickness requires tubes to be repaired or removed 60m service l become unserviceable prior to the next inspection. Unserviceable i I l l I w ,y .,, - y

awumts m m g g u r U.S.NRC April 9,1997 ICAN049702 PAGE 10 i speci5 cations as tlw condition of a tube ifit leaks or contains a defect large enough to affect its struaaral integrity in the event of an operating basis earthquake, a loss-of-coolant accident, or a MSLB. Of these accidents the most severs M% with respect to IGA within the Ur8 is theMSLB. Tests of tubes with representative IGA Saws removed from ANO-1 OTSOs during 1R13 showed that flawed tubes are capshle of withstanding differential pressure in excess of 10,000 paid without the presence of the tubesheet. Testing of simulated through-wall flaws afup to 0.5 inch in diameter within a tubesheet abowed that the tubes always fhiled outside of the tubasheet. Thus the structural esquirement of the bases of the surveillance 5=4A4 is satis $ed considering this NOED, I4akage under accident conditions would be lindted due to the small size and morphology of the flaws and would be low enough to ensure OShite dose limits are not exceeded Therefore, the NOED does not involve a signifk.c4 radim in the margin of safety. This condidon was also evaluated in accordance with 10CPR50.59. He evaluation concluded l that the condition did not represent an Unreviewed Safety QMbn. 1 i t Baals for No Envirenamental Ceasequences l His request fbr enforcement discretion does not have a significant edbet, impact, or change to i the quality of the human environment at ANO. This request, when :+;-lamented, does not impact the ANO Environmental Report Operating IJcense. Therefore, it does not involve any i environmetal consequences. Imaraved Technical Snaciftentiaan Innlications The condition fbr which enforcement discretion is being requested would not have been prevented iflmproved Technical Speci6cadcas were implemented fbr ANO-1. 4 .n

~ U. 5. NRC April 9,1997 i i ICAN049702 PAGE 11 1 J i 1 l 4 f i 4 i 4 i 1 I 1 1 k 1 a 4 ,i I h MAREUP OF ANO-1 TRNMICAL SPyrTFICATION 4.18.5.b I i i I 4 ( 1 4 e s A l 4 4 J

g 8. Un nrviceablo dancrib2s ch3 conditisn of a tub 3 if it 1saks l er centeino o defect larga cntugh ta offcet its ctrustural ~ integrity in the event of an Operating Basis Earthquake, a loss-of-coolent eccident, or a eteen line er feedwater line break as specified in specification 4.18.4.c. 9. Tune Inspection means an inspection of the steam generator tune grasa the point of satry ooseletely to the point of exit. I b. The steam generator shall be determined opezable after completing the corresponding actions (plug or sleeve all tubes saceeding the plugging limit and all tubes containing through-wall cracks) { required by Table 4.18-2, with the followins excentient 4 l P2han witT intararanular attaek indientlans within the unner ethe a nest uit i the notantini of th rou eh-=== 11 deaths amantar than taa eluseina A wi t anav r--in in service for the rammind=r of cycle 14, j i 1 l 5 1 l I I 1 1 i ) i h 1 4 l l l Amentnant No. M,M,M,M6,M4, 110ml

1@26 MWW5 UXUXWheW TEL:301-508-5369 Apr 09'97 19:23 Apr 09 Apr 09 ~ Transmit Journal No. Remote Location Mode Start Time Pages Result Note 003 914147556233 Norm 09.08:15 12'34 19 OK Tx 004 918477463337 Nerm 09.09:49 04'13 05 OK Tx 005 916103375324 Norm 09 11:57 11'35 20 OK Tx 006 4156359 Norm 09.12:23 06'32 09 OK Tx 007 913607531496 Norm 09.14:13 04'04 07 OK Tx 008 916023935442 Norm 09.15:42 09'41 23 OK Tx Receive Journal I No. Remote Location Mode Start Time Pages Result Note 005 604 832 3663 Norm 09.08:31 04'03 07 OK 006 610 337 5349 Norm 09.08:37 02'07 03 OK 007 6107747540 Norm 09.08:56 02'02 03 OK 008 G3 Norm 09.09:35 01'23 02 OK 009 5018584685 Norm 09.09:55 01'53 04 OK 010 2077984220 Norm 09 10:00 02'54 03 OK 011 14711105 Ncrm 09.11:27 01'36 02 OK 012 5018584685 Norm 09.12:37 08'55 15 OK ~ 013 203 443 5893 Norm 09.12:52 03'16 04 OK ~ 014 6093395435 Norm 09.12:56 03'55 08 OK 085 508 830 E575 Norm 09.13:55 01'54 03 OK 016 402 533 7291 Norm 09.14:18 05'34 10 OK 017 402 533 7291 Norm 09.14:34 16'27 26 OK 088 615 365 8000 Norm 09.14:53 01'47 03 OK 019 402 533 7291 Norm 09.15:05 19'17 29 OK 020 860 440 2091 Norm 09.15:27 01'39 03 OK 021 G3 Norn 09.16:02 01'53 03 OK 022 6107747540 Norm 09 17:19 05'15 09 OK 023 5018584685 Norm 09.19:13 09'12 15 OK

.-.. ~ - _ APA-09-97 12:30 From:ANO GB 1 5018504683 T-400 P 02/15 Job-918 j ^ i i a i l April 9,1997 i ICAN049702 i U. S. Nuclear Regulatory Commission i Document controlDesk l Mail Station PI-137 Washington, DC 20555 l

Subject:

Arkansas NuclearOne-Unit 1 i Docket No. 50-313 l License No. DPR-51 i Request For Enforcement Discretion Gentlemen: This letter documents the Arkansas Nuclear One, Unit 1 (ANO-1) position (enclosed) diamined on April 9,1997, with members of tha Nuclear Regulatory Commission staff and fbnnally requests enforcement discretion f>om the requiresnents of Technical SpeciSantion 4.18 regarding i surveiDance of Once Through Stasm Generator tubes. His enforcement discretion is requested in l~ oaler to suow suf$cient time for the submittal and NRC review and approval of a proposed change request fbr a one time exception to the requirements of Section 4.18.5.b which win allow i tubes with Intergranular Attack indications within the upper tube sheet with potential through-waB depths of greater than the plugging Ilmit to remain in service for the remainder of cyc!e 14. A markup of the proposed change is attached. i l The attached request was reviewed and approved by the ANO Plant Safety Committee at xxxx on i April 9,1997. Verbal approval of this enforcement discretion request was received at xxxx on Apr5 9,1997. l Verytruly yours, i Dwight C. Mims Director, Nuclear Sainy 4 i i i I l j ATTACHMENT 2

,~ -App.gg.~g7"jji30 ~~ FrosiAii0 dB 1 ~ ~ ^ ' ' ~ ~ ~ ' ~ ^ ^ 501858I68i T-liOT03/li Jcb 818 2 i U. C. L%h%e April 9,1997 i ICAN049702 PAGE 2 i cc: Mr. Ellis W. Merschoff Regional Administrator i U. S. Nuclear Regulatory Commiulon RegionIV 611 Ryan Plaza Drive, Suite 400 I, ArEnston, TX 76011-8064 NRC Senior Resident Inspector j Arkansas Nuclear One l P.O. Box 310 l London, AR 72847 Mr. Doorge Kalman l NRR Project Manager Region IV/ANO-1 & 2 U. 8. Nuclear Regulatory Commission l NRR Mail Stop 13.H 3 One White Flint North i 11555 Rockville Pike j Rockville, MD 20852 I 3 i t 9

.s. l REQUEST FOR ENFORCEMENT DISCRETION REGARDDiG ONCE THROUGB STEAM GENERATOR TUBE SURVEIILANCE Descriotion of Condition /Sgquintments for which Enforcement Discretion is Reauested Intergranular Attack (IGA) is imown to be present above the 15th Tube Suppcrt Plate (TSP) j within the ANO-1 Ones hrough Steam Generators (OTSGs) as variSed by destructive j examinaths. (DE) from previous tube pulls. IGA is a damage mechanism caused by corrosion of the material grain boundaries. The corroaton resulted Gem contaminants introduced on the subing i during the early years of plant operation. The contaminant causing IGA of the ANO-1 tubing la sulfbr as a result nf thermal decomposition ofion exchange resins. The ANO 1 IGA can be l categorised as volumetric, or " patch-like", with no speci8c orientation. Since discovery, there has been no evidence ofleakage kom IGA flaws at ANO-1. l During the IR13 refbeling outage. an eddy current (EC) tWeie was employed to depth size j the IGA. His technique had been quall6ed por Appendix "H" of the EPRI"PWR Steam l Generator Tube Examination Guidelines." his technique was used to depth size all IGA Saws i within the upper tubesheet (UTS). During this inspection,20% of all indications detected within i the UTS region by bobbin coll were examined using Rotating Pancake Coil (RPC) to characterize i these flaws. AllIGAindications between the 15th TSP and the secondary fhce of the UTS was removed kom service by plugging. All UTS IGA indications with a depth size of > 40% i through-wall (1W), as determined by the quae 6ed slains technique, were also removed 8 rom servicebyplugg* s. m Mres tube samples containing IGA Saws were removed firom the "B" OTSG fbr thturs l 66r-r of an Ahemate Repair Criteria (ARC) and to fbrther support the quali6ed EC sizing j technique employed during the IR13 refueling outage. l Preliminary DE results ofIGA Saws contained within tube samples removed Rom *B" OTSO indicate that the flaw depths do not correlate we!! with the depths stand using the qua&d EC j technique. He entire data set, including the results of the most recent outage, still satis 8es the l qualification requirernents of Ap,M H. j i l The inservice inspection of the ANO-1 steam generators is conducted in accordance with ANO 1 Technical SpectScations 4.ls. Speciccation 4.18.2 states: "Imerwor kupeuthm gaseen 1 sernerator Arbut shall haciudt noruitstructtn sunnination by eMycurrent tandq or other i equinaient aschniquer? Speci6 cation 4.18.3 requires that a minimum sample size be examined in j eoeordamse with 7 '" " r 4.1e.5. SpeelScation 4.18.5.b. notes: "7hr.esamm gunsresor Aall l &a dreriadned opera 6le q#sr conqpledig she cow g -Og oc#as (pag or sirew a# subes i exoneeng mepkgpg naar outau arbes connintry thnesh-wall onnend rep,orway Taur i 4.18-2." Table 4.18-2 specifies the expansion criteria for samping of the steam generator tubes j and requires "defectiw" tubes to be plugged or sleeved. S, a%n 4.18.5 deSnes effet as l f6tiows: *ar inperfoetion gmh merity that it moede the phesshe linrit etspt where the } ) i

_ _ _ _M-{,_{Q_jr_os *.. - ~._ _-- - ---------- --! 0 GS8 1 501:5s4ss5 T-400 P.05/is Job-sis 1 April 9,1997 t ICAN049702 PAGE 2 Inqperfretion has been aponed by the insnellation ofa steen. A tube oantainhw a dqfact in its presmre bowabry is defectiw " Pharfar Limit is defined in the same speci8 cation as follows-l "the imperfection drpth at or beyond whicle the tube shall be restored so serviceablitry by the } installarian da sleen or remondham service because it mqy become umerviceableprior to the next impaction; it is spal to #% of the nominal tube wall thickneas." The Bases for Speci6 cation 4.18 states: "The surveillance repirementsfor htquection 4the ) steam gencretor tubes erasure that the stucturalinnegrity of thisportion qf the RCS will be maintained Ihrpmeramforimevionknqpectionofsneamasnaratortubesisbasedona l mod $ cation ofRegulatory Guldt 1.8.9, Revision 1." ) Criterion IX, " Control of Special Processes," contained in Appendix B to 10 CFR Part 50 states, i in part, that " Measures shall be established to assure that special processes, including ) nondestructive testing, are controlled and accomp!!shed by quallned personnel using quali8ed i procedures." i l Criterion XI, " Test Control." requires, in part, that a test program be established to assure that all l testing required to demonstrate that structures, systems, and components will perform satisfhetorily in service is identi8ed and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. l i To demonstate the eddy current bobbin coil's ability to adequately depth size 10A patches within the upper tubesheet, ANO q'="A-' a technique in accordance with Appendix "H" ofEPRI"PWR ] Steam Generator Tube Examination Guidelines," Revision 4 dated June 1996. Compliance with the EPRI guideline was considered an acceptable method to qualify notwiestructive amanninstion (NDB) techniques ibr the detection and sizing ofdamage mechanisms. This was the only qualification technique available at that time, m quali6 cation data set demonstrating the capability of the ir,spection process consisted of service degraded tube specimens (i.e., tube samples removed 'Irom the ANO 1 and Crystal River steam generators). The nondestructive examination parameter responses for the Crystal River tubes were th!!y consistent with the nondestructive examination parameter responses of the ANO-1 Saws. During IR13, three tubes with bobbin indications within the upper tubesheet were removed firom the steam senerator. Two of the thros tubes contained Saws shes would have required mpair. The third tube was near the repair limit and may have been praedi rspaired. The tubes were selected on the basis of their containing multiple 1=Eaa+iaaa with depths representadve of the avernen inhelan depths as sized by BC. After bursting the tubes in the laboratory, the daws were examined and sized. If a Saw was not opened by the burst of the tube it was bent open for destructive examination. The DE results are not consistent with tlw previous qi=NAA data of the bobbin coil lbr sizing IOA flaws in the upper tubesheet. The reason for the '=+i=y in sizing IGA in the upper tubesheet is still under review. As a result of this condition, it is possible

- ~ -. -.. ---..~ - _ - l APF-09-97 12:32 From ANC 03s 1 E010!84C8 T-4C0 P.C uls M -sle

u. a. mw i

April 9,1997 ~ j ICAN049702 PAGE 3 } that tubes were let in servios with through-wall defoots gruter than the technical speel6eations I phagginglimit. Mrefore, ANO is requesting enforcement discretion regarding technical specification 4.18 to aHow suf5cient time br submittal and NRC review and approval of a proposed technical 1 speci8 cation change request for a one-time exception to the requirements of Section 4.18.5.b which will allow tubes with IGA defects within the upper tube sheet with potential thmugh wall depths arenter than the plugging Emit to remain in service fbr the remainder ofeycle 14. 'Ihis request assis6es Criterion B.I.(b) of the NOED guidance. Cesanessatory Measures l h methodology br monitoring the secondary systan for laakage includes the une of process monitors to check radiation levels in the condenser offgas, N.16 gamma levels from the OTSGs. chemistry samples, and RCS mass balances to calculate leakage. Additionally ANO-1 has a t procedural limit of 0.1 spm (144 gpd) that is more conservadve than the 0.347 spm (500 spd) l Ilmit aBowed by Technical Speci5 cation Oli) 3.1.6.3.b. Management has previously established a conservative administrative limit of 0.069 apm (100 spd) at which, upon confirmation, a plant shut down is to beinitiated. j Operations personnel trend information tom the steam, condenser otr-gas and OTSG process j monitor systems to determine indication of an OTSG tube leak. Steam lines are monitored by radia% unonitors and N-16 ganana detectors that provide chemises and opemtors with the capability of prompey detecting primary to ncondary leakass. l The amount of N-16 present in the secondary systerm is influenced by the sias of the leak, l loossion, and the powerlevel. ANO 1 utilizes Geiger-MueRar type detectors as N46 monitors. These monitors are nonnaDy selected to measure gross activity tom the OTSO but are selected to i monitor N 16 in accordance with Abnormal Operating Procedure (ADP) guidance for small OTSG tubo leaks. h monitors provide input to control room annunciators annociated with j OTSG tube leakage, hse N-16 monitors have only a single po* t correlation ofleakage ti., an m l N-16 reading based on 100% power level. Ouldance is given in AOP 1203.023,"Smals Steam Generator Tube Leaks", to correlate an N 16 reading of 1x10E4 CPM as being indicative of tube leakage of 2 0.1 gym sad a 500 CPM change being indicative of a 0.01 gpm change la primary-i e n ='- y !=kase at 100% power. ANO-1 lastalled high aansttivity N-16 detectors in January of 1997 to enhanos detoottom of smal! changes in primary-to-namadary tube leakage at various power levels. A modi 6 cation was made l to the plant computer to slow monitoring ofboth the original and newly instaued N-16 monitors j to provide a readily vienile laA4 archasses

  • count rete due to changes in leakage. The m

plant computer input has an alann that can be used lo actuate a control room annunciator panel to alarm as a value set by operators. outdance Ibr the use of the newN-16 detectore hr menhorias primary-to-secondary leakage is given in Operations Infbrmation Notice (OIN) #44. This document provides a correlation of count rate versus leakage. The O1N provuses shut down j i 4 i

m-es-srJr;ss Fr ahe ose 1 seist:4c:: 7 4sg p gpfis 2,3.gi, April 91997 0 l 1CAN049702 PAGE 4 4 guidanos for a step change in N-16 count rate ofgreater than 500 cpm in less than one hour. This l corresponds to a 60 spd leak that might be indicative of a tube beginning to fhlt. I The condenser oSgas monitor is an in-line detector on the combined suction line of the j condenser vacuum pumps. h is a gamma sensitive scintillation detector that provides a means to i 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. ne main steem high range radiation monitors are Geiger-Mueller type detectors. Dese detectors provido input to the Safety Parameter Display System (SPDS) for display in the ikne of the l control room. i l The plant computer leak rate program provides operators the ability to validate indications of ) primary * ---" y leakage by observing changes in the Rasetor Coolant System (RCS) mass j inventory. This program allows detection of changes in the make up tank level and determination 1 orleak rate changes based on the time interval selected. The SPDS is also available for use by operators. His system has a screen dedicated for use during suspected or actual primary-to-secondaryleakage events. The" Steam GeneratorTube Rupture" screen contains N-16 readings (firom the original detectws), condenser off gas, RCS Avg. Temp (Loop NB), OTSO Tube-to She!! detta T (OTSO A/B) and T-Sat fbr the OTSOs In amaa, the SPDS graphics display is outlined in red and flashing when a parameter on the graphics displayisin alarm. The Chemistry Department routinely analyzes and trends samples Aom the RCS and secondary water systems to identify and quantify primary-to secondary leakage. 05sas samples taken kom the condenser vacuum pump discharge are analyzed fbr Argon 41 activity. Uguld condensate samples are analysed for tritium to quantify activity levels in the secondary system. Argon 41 levels yloid a better measure ofinstantaneous levels ofprimary to-secondary leakage Tritium levels in the seconday system increase linearly over time during a primary-to-secondan leak. A primary to secondary leak rate can also be determined kom the tritium analysis. Secondary liquid samples are also routinely analyzed Ibr fission product activhy using gamma q,nwpy. An AOP directs special sampling by the Chemistry E@du..r.: until primary-to-secondary leakage is reduced below 0.1 spm or the reactor is tripped. The Operations and Chemistry Departments utilize available information to detect changes in primary to maaandary leakage nut so initiate actions to place the unit in a safir condition. Procedures are provided such as Er i

y Operating Procedure (EOP) 1202.06," Steam Generator hbe Rupture," AOP 1203.023, "Sman Steam Generator Tube Leaks," and the 1203.012 series for annuneister aorrective actions are utDized when the monitors, indicators, trends, or annunciators achible changes indicative of the development o( or change in, primary-to-essendaryIsaksgs. no operstlans department um these proeulures to place ele pleas la e stable condition and to mitigste the consequences of an OTSO tube leak.

=

APR-08-8T 12:33 Fros:ANO G$s I 5018584665 T-400 P.0 U15 Job-918 w..... April 9,1997 i 1CAN049702 PAGE 5 i f FinaDy, ANO maintains thorough training oflicensed operators by using the plant simulator fbr primary to-secondary tube leaks and ruptures. This insures ihmiliarity with the symptoms and l indications of this event to enable timely diagnos:s and action fbr piscing the unit in a safe l condition. I l Evaluation of Safety Sinalficance l The subject addy current sizing technique was employed ihr 10A defbets within the 17T3. All UTS IGA indications with a depth nine of 240% TW were removed Dom service by plugging the l affboted tubes. I l The three UTS IGA tube samples removed during 1R13 were subjected to room temperature l burst testing. Burst testing was performed separately within the flawed and un8 awed regions of j the tube samples. No simulated tubesheet was employed during the tests. The tests were l performed using bladders in the Sawed region. No foils or lateral restraint systems were used. 1he burst pressures for the flawed regions were between 10,000 and 11,000 pais. m unsawed l regions burst at pressures between 10,700 and 11,200 psig. For ANO-1 OTS0s, structural integrity la conservatively demonstrated by pressurizing the steam generator tubing to thrse times j normal operating differential pressure. This pressure for ANO-1 is 3765 psig. The burst testing resuka indicate that substantial structural margin exists. i l In 1996, to support ANO's study oflGA. burst testing ofpre-defected tubes was completed by l Framatome Technologies Inc. (FTI). The burst testing consisted of nine tubes containing i through-waB drSled holes up to 0.5 inches b diameter and one tube containing nn doisets placed { within a almulated tubesheet. Nine ofthe specimens burst at pressures 210,941 psig. Each tube l burst outside the tubesbest within the nonalefbeted portion of the tubes. One tube reached a j pressure o(9,577 peig but did not burst due to bladder leakage. These test results indicate that the tubesheet provides sufBeient support to preclude tube rupture within the tubesheet. The tube samples removed tom ANO-1 in 1996 included eleven IGA indications in the UTS. Since it was confirmed that the inservice IGA indications are volumstric, bobbin amplitude (vokage) was used as a bounding parameter. The eddy current responses tom these Saws were compared with the population ofinservice IGA indications to determine how i y. ::Mdve the Saws were of those remaining in iervice. The 600 KHz bobbin coil signal amptitude of eaws in tubes that were puUed during 1996 ranged from 0.46 to 2.69 vohs. Of the 470 inservice IGA irwHe** lana, au are bounded by the 2.69 value. Additiona9y, a comparison of RPC data was performed to ibrther substantiate that the puHed tube Saws bound those indications remaining in servios. The RPC data couacted for the tube pull namples rundted in a maximum Saw extent of 0.16 inches. RPC signal infbrmation was ooDected i on 113 indications within the UTS. Ten of the largest RPC voltage indications were ernminal to 4 i decennine the length 4y width estest by RPC. The largest RPC esteet ihr thoss IGA indications left in service was 0.16 inches. Therefore, it is concluded that the insarvice IGA indications are l bounded by those tube samples that were destructively examined. i 4 ---r -,y ,.-,,m e --y 7

n 7p,.w. { i -g e,,7- --- -- - -gg----gggggy- - l April 9,1997 i 1CAN049702 PAGB 6 i Structural integrity of the tubing within the tubesheet is assured based upon demonstration of the following: j A. 'the actual tube samples removed from ANO 1 dudng IR13 exhibited burst l pressures that substantially snoseded the required structural limit. the structural support provided by the tubesheet precludes tube rupture. j B. i j C. The inservios IGA indications are bounded by those flaws contained in the tube i aamples that were punod. i l The IGA patches destructively examined were not through wall; therefore, normal operating j pressures did not resuk in tirough-wall ieskage. 'Ihis was evident during inservice ' spection of a the tubing in whloh no indioetions of residual leakage was noted. A comparison of 1R12 and IRl3 redbeting outsee EC signatures indicates that the IGA exhibits little or no growth. Also, comparison ofinspection data prior to the IR12 reibdng outage supports this conclusion. l Additionally, during May 1996. "B" OTSG tubing was subjected to a diffbrendal pressure of l approximately 2100 paid for several hours as a resuk of a feedwater transient. No immediate increase ir; primary-to. secondary leskrate was noted during the event or following startup. Le l p y-t -seoor,t.r.sy lenkrete did increase by approximately 18 gpd three days following startup; i however, none of the leakage detected during the IR13 refbeling outage was from IGA flaws. It j is concluded that leakage through IGA flaws in the UTS is highly unlikely at Main Steam Line i Break (MSLB) pressures due to the Saw morphology and the near MSLB differential pressure j that occurred in May 1996 with no resultant leakaps. I 5 conditional core dange probability is the increau in core damass esquency due to a siven sendition other than that assumed for the base FRA. The PRA assunned that the tube intserity is such that no steam generator tube rupture would be induced due to transient conditions. De limiting licensing basis transient widch could most adversely affbet the tubes by creating a high didrerential pressure across the tubes is a MSLB Accident. This accident could produce a tube di8hrential pressure ofup to 2500 paid. The tube sample burst pressures were wet above pressures which would be seen in a umiting M5LB accident. Stru the likenhood attubes j rupturing is not increased because ofthe larger than expected Saw sizes due to IGA in the UTS, j the condaional core damage probabuity ibr this condidon is zero. 4 ] 'Ihs limiting licensing basis accident with respect to dose consequences Oom induced tube leakage la the MSLB accidsat. This accident assumes a total leakage of I spm with 1% iWied Aael in the care. However, steem generator tube leakage is proosduraHy limited to 0.1 spm during normal l operstloa. Byen though leakage is not expected to occur, MSLB induced tube leakage has been =-- --J4, a*3= aim 8 so be 0.$3 gym on the a$beted steam generator. 'Ihe ibHowirig r =ghs are made concernirig the saunbar of flaws and associated leakage: i

1) Since steem generator A has the largest number oflGA patch indications (285) k was chosen i

as the affected generator bounding steam generator a with only (In) indications. I b i

N @W Free:AND G3s 1 5018!04685 T 4t0 P.10/15 Job-818 U. S. NRC April 9,1997 ~ ~~ ICAN049702 PAGE 7

2) Half of the Indications are assumed to leak under MSLB conditions.
3) Representative leakage values for axla! flaw lengths were utilized to bound the leakage expected toenIOA patches
4) Applied the longest IGA length calculated Som RPC data to the 50% population assumed to j

leek. j

5) Assunned the Saws grew in length an additional 25% over the oyola.

i

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

i Since there are 2s5 indications, halfof this value wul be 143. The longest length in the axial plane l was 0.14 inches. When increased by 25% this yleids a daw length of. j 0.14 inches * (.25) = 0.175 inches 4 l If 50% of tbs longth is neaumed to be 100% TW: 1 0.175 inches * (0.5) = 0.0875 inches i l Using leakage curves developed br OTSG's for axial Saws, the laskass Rom a single flaw (0.0875 inch,100% TW) is determined to be 0.0025 spm. To enmpensate for norinal operating temperature the value is multiplied by 1.47 to yield a Snel leakage af o.003675 spm per flaw. This value is then multiplied by the number of potential leaking *aws to give a total leakage ce { 143 Saws

  • 0.003675 gym /Saw = 0.53 spm 1

When the estimated leakage in the affected steam generator is added to that which is a!! owed by procedure, the total leakage rate is expected to be no greater than 0.63 spm. Since the assumed l leakage ruta is greater than the conservative calculation, the current licensing basis assumption of I spm remains bounding. The sutdoct flaws do not represent a structural or leakage concern Thersibre, the presence of i inservios upper tubeshoot IOA defoots with through-wall estants that may auseed the technical W plugging limit does not pose a concem relative to the hashh and safety of the public. } Enainearian Evalsatten andlasis For No Binnincant Hazard Consideration l An ameussion of the proposed NOED request has been p.e%.--- M in menordance with j 10CFR50.91(a)(1) ryardirng no significent hazards consideration using the standards in j 10CFR50.92(c). A d,scuamon of t ione standards as they reise to this request hilow.: Critarina 1 - Does Not Involve a SigniScant Increase in the Probability or Consequences l of an Accident Pievioudy Evatusted. i 1 l h

E -W -lN 12:16 From:ANO fot8 1 5018584615 T-400 P.11/15 Job-918 U. 8. NRC .. s April 9,1997 1CAN049702 PAGE 8 N steam senwetors are used to remove heat firom the reactor coolant system during normal operation and during accident conditions. The steam generator tubing forms a substantial portion of the reactor coolant pressure boundary. A steam genwator tube failure is a violation of the reactor coolant pressure boundary and is a speciSc accident analysed in the ANO 1 Safety Analysis Report. The purpose of the periladio surveillance perfbrmed on the steam generators in accordance with ANO 1 Technical speciscation 4.ls is to ensure that the structural Integrity of this portion of the RCS will be maintained. The technical spe@'+ Ion plugging limit of 40% of the nominal tube waH thickness requires tubes to be repaired or removed tom service because the tube may become unserviceable prior to the next inspection. Unserviceable is denned in the technical spectScadons as the condition of a tube ifit leaks or contains a defect large enough to aGleat its structural integrity in the event of an operating basis earthquake, a loss of-coolant accident, or a steam line break. Of these accidents, the most severe condition with respect to IGA degradation within the UTS is the MSLB. During this event the dlShrential pressure across the tube could be as high as 2500 paid. % rupture of a tube during this event could permit the flow of reactor coolant into the secondary system i thus bypassing tlw containment. j l From testing performed on simulated Saws within the tubesheet it has been shown that the i patch IGA indications within the upper tubesheet lea in servios during 1R13 with potentla! depths greater than 40% do not represent structurally signi8 cant Saws which would increase j the probability of a tube fbilure b pad that currently assumed in the ANO-1 Safluy Analysis j. Report. i Burst tests were conduded on tubing with simuisted Saws within the mbakaat In these j tests, through-wsE holes of varying sizes up to 0.5 inch in diameter were ditled in test specimens. The flawed specimen tubes were then inserted into a simulated tubesheet aad l pressurized. In all cases the tube burst away tom the Saw in that portion of tube that was i outside the tubesheet. The size of these simulated flaws bound the indications lea in service i within the upper tuboehest during lit!3. These tests demonstrate for Saws similar to the peseh IGA hund in the ANO-1 upper tubesheet that the tubes wlB not fhil at this location 4 i under accident conditions. I h dose consequences of a MSIR accident are analyzed in the ANO-1 accident analysis. l This analysis assumes the unit is operating with a 1 spm steam generator tube leak and that j the unit has been operating with 1% defective ibel. Increased leakage durtig a postulated MSLB acoklemt resulting 90s the patch IGA id in servios in the upper tubesheet is not expected. IGA has been present in tlw ANO 1 steam i generators for many years with no known leakage attributed to this damage mechanism. l Because of ha loestised nature and morphology, the flaw does not open under accident Pressure conditions. l t j

AP H I-If 12:36 From:AND CtB 1 51158468! T-400 P.12/15 Job-818 .g U. M. NRL; April 9,1997 iCAN049702 PAGE 9 l Therefbre, the NOED which allows continued operation with IGA Saws within the UTS with potential through-wall extents greater than the technical speci6 cations plugging limit does not result in a signi8 cant increase in the probability or consequences of an accident previously evaluated for ANO 1. Criterion 2 - Does Not Create the Possibility of s New or Different Kind of Accident I tom any Previously Evaluated. The steem generators am passive components. h intent of the technical specification surveillance requirements is being met by this change in that adequate structural and leakage integrity wiH be maintained. The proposed change introduces no new modes of plant operation. Thorstbre, the NOED does not create the possibility of a new or different kind of accident Dom any previously evaluated. Criterion 3 - Does Not Involve a Signiscant Reduction in the Margin of Safety. The ANO-1 Technical Speci8 cation Bases specify that the survelliance requirements (which includes the plussing limits) are to ensure the structural integrity of this portion of the RCS pressure boundary. The technical speci6 cation plugging limit of 40% of the nominal tube 3 wou thickness requires tubes to be repaired or removed &am service because the tube may become unserviceable prior to the next inspection. Unserviceable is defined in the technical I specifications as the condition of a tube ifit leaks or contains a defbcs lares enough to effeet its structural integrity in the event of an operating basis earthquake, a loss ofcoolant WM or a MSLB. Of these accidents the most severe condition with respect to IOA within the UTS is the M5LB. Tests of tubes with representative IGA Saws removed kom ANO 1 OTSOs during 1R13 showed that Sawed tubes are capable of withstanding differential pressure in excess of 10,000 paid without the presence of the tubesheet. Testing of simulated through-wall flaws i ofup to 0.5 inch in diameter within a tubenheet showed that the tubes always fhiled outside ~ of the tubesheet. Thus the structural requiremea of the bases of the surveillance specificesion is satis 8ed considering this NOED. IAakage under accident conditions would be limited due to the sma5 size and morphology of the Saws and would be low enough to ensure offbito dose limits are not a w Thersibre, the NOED does not involve a signif% sat reduction in the margin of safbty. This condition was also evaluated in accordance with 10CFR50.59. The evaluation conduded that the condition did not represent an Unreviewed Safhty Question. Basis for No Enviressental Consecuences u-

i o APF-os-er 12 3r Fr :Aho ass i 5018!I4205 T-4C0 P.13/15 Job-sis g g, g f,.. April 9,1997 1CAN049702 PAGE 10 1 This request for enforcement discretion does not have a significant effect, impoet, or ehenge to l the quality of the human environment at ANO. This request, when implemented, does not impact j the ANO Environmental Report-Operating Lloense. Therefore, it does not involve any environmemalconsequences. l Imareved Tes)nical snerifications knalications i j ne condition for which ensbreement dieersion le beins requested would not have been prevented j if!rnproved Technical Speci$ cations were implemested Ibr ANO-1. 1 5}}

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