ML20236C137
| ML20236C137 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/09/1987 |
| From: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| CAL, NUDOCS 8710270028 | |
| Download: ML20236C137 (2) | |
Text
{{#Wiki_filter:__ 4 34.i-e f eY 0 'OCT 0 9 1997 j i I Virginia Electric and Power Company . ATTN: Mr. W. L. Stewart,'Vice. President. g . Nuclear.0perations a P. 0. Box 26666- '. R. c-g :M _ Richmond, VA 23261 a-o Gentlemen: ? > ta
SUBJECT:
VIRGINIA' ELECTRIC-ANDPOWER' COMPANY (VEPCO)LETTERDATED .g ) SEPTEMBER 22, 1987' REQUESTING APPROVAL FOR NORTH ANNA-POWER a STATION )JNIT 1 TO STARTUP AND OPERATE AT 50 PERCENT POWER N j f DOCKET NO. 50-338 This confirms.a: telephone -conversation on October 9,1987, concerning our j ' Confirmation of Action letter 'to.you dated. July 22, 1987, which required NRC 1 . concurrence prior-to restart of North Anna Unit 1. 1 We'have evaluated the information presented in the subject letter (See enclosed . Safety Evaluation' Report :(SER) dated October 8,1987), and 'your Standing Order { No. 155 dated October 2,;1987,-Unit I and Unit 2 primary to secondary leakage. Based on this evaluation and our confirmation that the commitments described in your letter and Standing Order No.155 have been satisfactorily met, we concur in your plans to restart and operate Unit 1 at 50 percent of licensed power, j arovided the.N-16 monitor you described in your presentation to the NRC in i 3ethesda, Maryland on September 10, 1987, is operable above 30 percent licensed j power. NRC will review the results of operation at 50 percent power as well as your l evaluation of steam generator tube' performance. Prior to authorizing operation j at greater than 50 percent power, the staff will evaluate the operating results -and other information, and issue a final SER. After we have completed our evaluations, you will be advised by separate. letter concerning full power operation. If your understanding of our discussion is different from the above, please inform this office promptly. Sincerely, 8710270028 871009 PDR ADOCK 05000338 G PDR J. Nelson Grace Regional Administrator
Enclosure:
Safety Evaluation Report -cc w/ encl: E. W. Harrell, Station Manager N. E. Hardwick, Manager - Nuclear Programs and Licensing \\ bec.w/ encl: (See page 2) y
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~ t 's a I887 Ykginia! Electric'andPower' Company . 2 ': 9 1 J-j bec'W/encli. P NRC Resident Inspector. c d Document Control-Desk Commonwealth of Virginia l u t-q t-4 l 'l RI 4 I RII'F2 y { ,RII. L 7'T FCantrell 'V8roknlee' peyps n ME st 10/f//87 10/f/87-30/h/87 10/ 87 1 /. 87 j I 1i .i 'l o-
___._________._..._..____l_._,___.______
) ) 0CT.08 '87.10:50 HRC. MESSAGE CENTER BETHESDA MD P.002 x ENCLOSURE
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w u, UNITED STAT 5s NUCLEAR REGULATORY COMMISSION naamnetow,o,c seses J out.ober 8, 1967 I e...e Docket No. 50-338 NEMORANDUM T0: Luis A. Reyes, Director Division of Projects, Region !! { FROMP Gus C.. Lainas. Assistant Director for Region !! Reactors Division of Reactor Projects-!/!!
SUBJECT:
SAFETY L.~ UATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELAft0 TO RESTART AND OPERATf0N OF NORTH ANNA,UNITNO.1(NA-1),AT50PERCENTPOWER l' The sub. ject NA-13afety Evaluation (SE) dated October 5,1987 is enclosed in accordance with the revised TIA dated September 24, 1987 regarding the NA-1 steam generator tute rupture event. As stated-in the NA-1 SE, the NRR staff finds that interim operation at reduced power (less than or..pqual to 50 percent power) is acceptable. Region !! concermace on W subject SE was received on October 2,1987 in a telecon between F. Centrol", Region !!, and the NRR Project Manager, L. Engle. NRR will review, in conjunction with R-!!, the results of operation at 50 percent power as well as the licensee's evaluation of $8 tube perfomance. Prior to euthorisation for operation at greater than 50 percent power, the staff will l evaluate the operating results and other infomation and issue a fina.1 safety i evaluation report. { 0 s C. L inas. Assistant Director for Region !! Reactors Division of Reactor Projects-!/II
Enclosure:
As stated ec w/ enclosure T. Murley J. Sniesek R. Starostocki F. Miraglia J. Richardson A. Thadeni &kj 5h155-{lDffC)-
OCT.08'8710:50 H K MESSAGE CENTER BETHE5DA'MD P.003 SAFETY EMAJJA""0N BY TH: 0FFICE OF NUC. EAR R %CTOR REGULATION E A ,1 TU TEsr NT W vruRATIQN AT GE POWER E UILTNv mi ~ M L,I5E--- NU. M f=4 T IWII LA i.ES IG W MMER tt-ANY LD i nyIll MI Le i RIC ---5 ERAT ;VE NU In H RA vus 3 i lIAURi uRIT led. 1 un.u ut. mo aas l INTMDWCTION By letter dated September 22, 1987, the Virginia Electric and Power Company l (the licensee) requested that the North Anna power Station, Unit No. 1 (NA-1) be permitted to start up and operate at 50 percent of full power. Following the July 15, 1987 M-1~ Steam Generator Tube Rupture (SGTR) event, the licensee agreed to obtain concurrence from the NRC prior to NA-1 restart (Mode 2). This agreement was specified in the NRC Confirmatory Action Letter (CAL) issued July 22, IN7. The licenses has completed the evaluation of the SGTR event and has submitted by letters dated September 15 and 25 IM7 the evaluation of the $GTR event, including the 4GTR failure mechanism an,d modifications to be made prior to restart. T m NRC review of these matters may extend beyond early October,1987 when M-1 is scheduled to be ready for restart. Therefore, as noted above tion of M-1 at 50 percent of full power pending NRC authorization for fullthe lice power operations. In order to place these matters in pro r perspective, a brief description of the NA-1 SGTA event and the licensee s investigation of this event is provided below. Prior to 0630 hours on July 15 1987, NA-1 was operating at 1004 power. At 0630 hours, the Mein staan Line, "C" radiation monitor registered a Hi Hi alars and the Control Operator (CRO noted pressurizer (PZR) level and pressure ere,theCR0inc)reasedcharmingflowtotheReactorcoolant decreast The System ( $). The unit was manually tripped at 0635 hours and approximately 20 sec later e Lo-Le pressure safety injection signal actuated automatic trip. At 0639 s's Notification of Unusual Event was declared and at 0650 hours feeduster f to 50 "C" was isolated. Mowever, the level of SG "C" was identified to be reasing indicating an SG tube rupture or break (SGTR). At 0654hoursenalentwasdeclaredandat0705hoursSafetyInjection($I)was tereinated. At 0$0 hours emergency procedures were initiated for post-5GTR sooldown using backfill. The Technical Support Center was activated at 0757 hours and the locdl emergency operations facility activated at 0915 hours. The unit entered Mode 4 (Hot shutdown) at 1108 hours and at 1218 hours the RHR system was placed ' n service. The unit entered Mode 5 (Cold Shutdown) at 1330 i hours and t w eye s was terminated at 1335 hours. ) No automatic actudion of primary or secondary safety relief valves occurred. ) Total radioactive t release was less than 15 of Technical Specification (TS) limits. The tube leakafe rate (as determined later) was in the range of 560-637 gallonsperminute(GPM). Offsite environmental monitoring teams detected no increase in radiogetivity above normal background levels. The SGTR event was 1
m. l. l OCT.08 '87 10:51 HRC-MESSAGE CENTER BETHESDA MD P,004 dotarained to be bounded by the Updated Final Safety Evaluation Report (UFSAR). The maximus leak Pete (660-637 ppe) was less than the UFSAR value of 710 gpm. Core safety Itmits were not cha longed and shutdown and thermal margins were asintained. Once access to the NA-1 $Gs A, 8,(and C was gained, the licensee's evaluation of the 50TR event was oriented to:
- 1) Detamining the root cause of the failure; (2) Ascertaining the condition of the SGs, particularly with respect to the failure mechanism; and (3) Performing the necessary corrective actions to preclude the futu,re occurrence of a tube rupture event.
On July 21, 1987, VEPC0 identified a ruptured tube in SG C. The tube location was Row 9, Column CSI (R9 C81) on the cold leg side at the seventh support level. A fiber optic examination identified the failed tube to be the classical double-endedpuillotinebreak. On August 12, 1987 VEPC0 successfully completed the l remove of tube R9 C51 on the cold leg side up to and including the break at l the seventh support level. The tube was immediately sent to Westinghouse for an extensive nondestructive / destructive examination to determine the frec-ture sorphology and the failure propagation mechanise. The results of these examinations and the determination of the tube failure mechanism are provided l in.the licensee's final report dated September 15, 1987, and are discussed below. InordertoprovidejustificationforfuturerestartofNA-1,VEPCOhascon-1 ducted an extensies inspection of all three $Gs. The inspection has been the most extensive ed@-current. testing pre' a undertaken at a U.S. domestic facility with emptosis on detecting cire m ierential defects. Eddy current testing (ICT) ion method involves the insertion of a test coilis the principal m inspections. This Inspect inside the tube that traverses the tube length. The test coil is excited by an alternating current which creates a magnet'c field that induces oddy currents in the tebe wall. Eisturbancesoftheeddycurrentscausedbyflawsinthe the test coil terginals. ponding changes in the electrical im>edance 'as seen at tube wall produces corres Instruments are used to translate tiese changes in ' test coil lapedante inte output voltages which can be monitored by the test operater. Tie depth of the flew can be detensined by the observed phase angle response. The teet equipment is calibrated using tube specimens containing artificially induced flaws of known depth. The ECT testing poogtes has Tncluded the ins,pection of every tube support , junction and straight tube sections in all tiree SGs with an 8x1 pancake array probe. This probe (tal) has the sensitivity to detect all inner-diameter defects, either axial or circumferential, and with defects 20% or deeper and with a length of 3/16 of an inch or longer. Also, the 8x1 probe is able to detect auter-diameter cracks and intergranular attack on either the inner-or outer-diameter. In addition all indications detected by the 8x1 probe have been tested with the Rotating Pancake (RPC) probe. Finally, pro-filometry has been conducted on selected intersections. A Westinghouse Intelligent Eddy Current Data Analysis system (IEDA) has been used as an aid in flagging suspect bobin coil indications, which are then dispositioned by data analysts. The data free each tube has been independently reviewed by two different analysts. One analyst has used the Westinghouse IEDA system and the other analyst has used the 2etec Digital Date Analysis System. 2
OCT.08 '8710:53 HRC-MESSAGE-CENTER BETHESDA MD P.005 All data analysts are certified at least Level II in accordance with American Society of Nondestructive Testing (ASNT) requirements. The analysts have been given additional training by Westinghouse and required to pass a test that covers the specific data analysis being used for the present NA-1 eddy current l
- tests, l
Finally, it is noted that an NAC Aupented Inspection Team (AIT) was dispatched to the NA facility. The AIT was charged with determining whether the licensee's actions in response to the July 15, 1987 $4TA were adequate to protect the health and safety of the public and that appropriate action was being initiated to determine the seuse of the event. In addition, the procedures followed by the licensee relative to the SGTA were evaluated to assess the adequacy of in place procedures to cope with serious events of this type. The NAC AIT Report was issued August 28,1987. The Report, in part, concluded that, "The overall results achieved we m outstanding in that the operator tripped the plant, isolated the leek and brought the plant to cold shutdown in seven hours without using the S/G power-operated reitef valves. This contributed to a neOligible release to the environment. Our discussion and evaluation of these matters with respect to restart of NA-1 for operations not to exceed 505 ef full power am provided below, DISCV$$10N 8 teas Generator lesnection As noted above, tly licensee conducted an extensive $4 ECT inspection of the NA-1 $0s A, B and C. Identified indicatio.v. were either present in the April 1987 refueling ou with no discernable changs indicated or in previously uninspected porti 6 of each 50. Additionally, a review of the data from the last outage using he 1 resent analysis rules revealed several tubes t' hat should have been plugged at tw previous outage. This apparent, though not actual. change in the SG ndition is due to the change in the analysis rules and in. creased awareness the analysts of North Anna specific ICT signals. A review and cogarisen of $4 C hot les data desenstrates that the m is essentially no change in tube sendition from the April 1987 mfueling outage to July 1987 (when the event ogsurred). Of significant importance was the fact that there were no indications of cfrcumferential nature found at any tube support plate locations, includtig the seventh tube support plate. The number of tubes inspected is shown 5elow. Each steam generator contains 3388 tees. HoweWer, a number of tubes have been plugged from previous SG I inspections. The insiber of non plugged tubes are: 50 A - 3179; SG B - 3210; I and $G C - 3117. Tht number of tubes to be removed free service based on the SGT inspection by l indication type am indicated in the following table, i 3
- 0CT.08 '87 10894 NRC MESSAGE CENTER BETHESDA MD P.006 L
s- $ TEAM QENERATOR TU8ES TO BC P.UGGED AS RESULT OF SGTR EVENT ~ (gy Ind calion Type) Total' l l Tubes tubes Clear 1 Distorted 8 Sheet 8x1 Possible4 to be l jfGIndications Indications Indications Indications Others Pluaaed A 0 6 6 11 2 25 8 0 3 5 12 1 21 ~ C 2 2 20 11 4 39 1 Clear Indications (defective) - bobbin indications of greater than 40 percent "thru-wall" depth, sDistorted Indications - bobbin indications of undetermined "thru-wall" depth at tube support plates. sTubesheet indications - bobbin indications of undetermined "thru-wall" depth at tubesheet. 48x1 Possible Indications - indications identified by 8x1 probe.. '8 Tubes with broken probes or which would not pass 8x1 probe - includes failed tube. ' Plugging summary is as of 9/14/87 based on ECT results - does not include tubes to be plugged as a preventative asasure based on fatigue considerations or other concerns. Tube Failure Mechaniss UponarrivalatWestinghouse, tuber 9C51(thetubewiththecircumferential break) was immedi ly subject to a series of non-destructive / destructive tests to determine the failure. mechanism. Visual examinations and macroscopic examinations of the tube fracture surface were conducted to determine crack origins and crack propagation paths. Scanning Electron Microscope (SEM) and Transmission Electron Microscope (TEM) fractographic examinations were also perfomed to confirm tube crack origins and crack propagation paths. Mechanical properties of the tube were determined and found to agree closely with the 1971 tube certification data applicable to NA-1. Microstructure was typical of mill annealed Alloy 600 for AA-1. Grain size was small, ASTM 9.5. Based on the above, the cause of the failure was detensined to be fatigue. No evidence of any significant intergranular corrosion was observed on or immedi-stely adjacent to the fracture surfaces. High cycle fatigue striations were present and were measured to obtain the stress intensity which led to initiation of the fatigue crack and crack propagation. The mode of crack propagation 4
I 0CT.0887 10:54 HRC MESSAGE CENTER BETHESDA MD' P.007 ..W r concluded that leakage occurred between the time of total through-wall develop-l sent of the crack front and the final circueferential break. Theorientationandspacingofthestriationssupporttheconclusionthat normal design operational oedings were not sufficient to lead to the fatigue failure. Therefore, some other loading mechanism was acting on the tube to produce the failure. Measurements of the striation spacing provided necessary t data to determine the range of loadings that led to eventual fatigue of the tube. Adverse flew mechanisms were evaluated, such as turbulence, vortex shedding, and fluid elastic excitation. Review of the data supports the con-clusion that fluid elastic excitation was the most probable mechanism that could provide sufficient loadings or alternating stresses to induce fatigue. An additional method was utilized to determine these loadings and verify the striation spacing sensurements and resultant loading conditions. This method used tube dont data (obtained th profilometry and physical measurements) and finite element analysis to es lish mean stress data through the dent. This mean stress data, the dented configuration and fatigue curve were then used to detenmine the alternating stress intensity required to initiate a fatipue crack. This calculated range of stress intensity supported the similar conc usion determined free striation spacing sessurements that tube failurs was induced by fatigue.- A fluid elastic stability ratio was defined for failed tube 49 C51. The sta-bility ratio represents a measure of the potential for tube vibration due to instability during service.- Values greater than unity (1,0) indicate fluid elastic instabili %. The fTuid elastic stability ratio is defined as the effective velocity divided by the critical velocity. The calculated flow ratio was determined for current NA 1 flew parameters. Calculations determined that the tube would be more susceptibl6 to fluid elastic instability due to lower damping caused by denting. Simulated shakar tests supported the conclusion that in this regies of low desping, tube R9 C51 would be fluid elastica 11y unstable. As discussed above, the results of the present 50 inspection indicated no addf current indications of a circumferential nature at any seventh support plate location. This is consistent with the fatigue mechanise described above. The enjority of the f$1gue process lies in the cyclic loading (via alternating stress) to initiate a crack (or cracks) in the tube. Once the fatigue crack initiates, the tig required to propagate the crack is comparatively small. Antivibration Bara (AVBs) limit the high vibration amplitudes needed to achieve the alternating stress necessary for fatigue crack initiation. The depths of AVB penetration 1 40 the SG tube bundle can be estimated from addy current indications that enn then be translated to a SG inspection map which provides an indication of non-unifore AVS insertion depths. A large number ed AVB indications were identified during the curreni. SG inspec-tien. This is not unusual in a Series 51 Westinghouse 50. However, a few indications were identified as far down as Row 4. Therefore, extensive eddy current testing was performed to identify AVB indications. The inspection revealed that the majority of the Row 9, 10 and 11 tubes were supported by AVBs. However, failed tube R9 C51 was not supported by an AVB. 5
P.008 OCT.08 '8710:56 'NRC MESSAGE CENTER BETHESDA MD Correlation with the known deflections uired to provide sufficient stress to initiate fatigue show that the AVSs it the tube motion to below the required deflect on limit. This data provided further support to the con-clusion that the loading mechanism for Rg C61 was; fluid elastic excitation. the licensee concludes that the tube failure was due to high cycle c#, In summaryYhe fatigue mechanism was determined to be a combination of stresses -fatigue. i imposed by tube denting at the seventh support plate and vibration due to fluid elastic instability. 4 Corrective Actions The licenses has implemented a series of c'e'erec$ive actions and modifications NM' to preclude similar tube failures at NA-1. These include the installation of s a downsamer flow resistance plate in order to reduce the loadings experienced Preventive SG tuba plug 0 ng is being toplemented to e i by susceptible tubes. further reduce the probability of tube rupture.' In addition, an enhanced moni-toring pro 0 ram is being implemented to provide sufficient notification of tube f', leakage in order to shut down NA-1 prior to :s tube rupture. These matters are //,,g discussed below. "\\ (1) Downcener Flew Resistance Plates 6 The NA-1 SGs A, 8 and C are being modified to include a downcomer flow resist-y The DFRP wil reduce the steam generator recirculation flow anceplate(DFRP). and is expected to result in.the improvement in tube " stability ratio" needed to preclude further ttM failures of active tubes b the fluid elastic insta- /[I At w ted above, " stability ratio is a relative measure bility mechanism. of the potential for W oe vibration due to fluid elastic instability. Evalua tiene by the lic ee have concluded that a 15 improvement in stability ratio t shou 16 provide t necessarlt reduction in fati . usage (reduced amplitude of vibration) to p lude furtwe tes failures this echanism over thel remain-l ing life of the s as generators. The, install tion of the DFRPo,will be cem,0' f plated prior to N4-1 restart. (. I' the Final Safety Anal ObP(reducedmassflow)ysis For operations at greater than 5g percent in the Report (FSAR) will be revised to include tA reanalysis of the $$1R event with the 0F o SGTRaccidentana]ysis. I suited in a calculated offsite dose which is greater than reported in the UFSAR. The incretse in dose consequences for the SGTR event occurs only for rated thermal power l'evels above approximately 55, and the consequences are' still well within established acceptance criteria as defined in the UFSAR and the bases for NA-1 TS 3/4.4.8. (2) Preventive pluccino / s Preventive pluggieg will take place on the potentiald susceptible tubes in l 0 Row $ through 11 The essential criterion for identHybg specific tubes for All preventivepluggiegisthattheynotbesupportedbyatleastofwAVB. such tubes will be plugged. On the cold leg side, each tube meeting'this plugging criteria will be plugged with a sentinel plug. The sentinel' plug will pers't internal pressurization of the tube and low level leakage in the event a through-wall crack develops in the plugged tube. Thfs will serve as an early warn bg detection method for occurrence of a similar circumferential break of 6 1
4. n- ..V 1 apluggedtube,thusprecludingimpactwithanyadjacenttubeswhicharein I (service. This plugging will be completed prior to testart for NA-1.- ) ( l .It is noted that the cambined effect of the OFRP and the >reventive plugging l discussed above represents a reduction of 24' percent in tie stability ratio 'i referred to above </ (3)< $ 11aation of Failed Tube h 3 / ,f s e g The failed tube-R9 C51~ will be stEbh,lred to preunt further damage to adjacent tubeC In additten, its neighboring tubes will he preventative 1y plugged with j
- g\\hehtinelplugs.
x \\ (4) ' Tube pluanine-Eddy Current Testina y" < Based on the extensive $0 eddy current testing program,.all tubes having ECT . indications tiill he plugged to.neet established plugging criteria. These tubes will be:plugjad prior to restart for NA-1. (5) Ausmented Sutyeillance Prearau a[ An a nted surveillance program for monitoring steam generator primary-to-see ry leakage is being implemented. This program is based on the use of , O) N18 several radiation monitors and easpling (including the installation of a new M 3aans detection system) to quantify primary-to-secondary leakage. The E h,/ progree is designed te detect leakage during thi early st i I_ failure so that an orderly shutjfown can be accoupitthed. Administrative con-I trols addressing leak rate-limits ' operator actions and monitoring equipment /, J n' l operability are being prepared. the ht' gamma detection system will bec operational prior to power ascension. greater than 30 percent of full power. All 5 other aspei/'s of surveillance p ugram will be in place prior to restert for NA-1. WO d O. T STABILITYRATIOAT50PERCEl([,1.0AD(POWER) ~ The steam. flow rate frem a.stena ponerator is hi As the load decreases the ' steam f ow decreases. ghly dependent on themal load. a' In additien, s decrease in load is accom Both of those effects work toreducet%pentedby'.vdecreaseinvoidfraction.At 65 lead, the effect on stability tube stabflity ratios. ratios from steam flow (velocity and density) is a 26% reduction in stability ratin.reletive te, full load and prior to any of the modifications noted above. In ede4tioti, at'AW load there.is a reduction in void fraction of about 10% which increases t ee' damping. This reduction in void fraction, if considered, would result in a further reduction in stability ratios over<the 26% noted "saboyy,
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Studies perfomed for tN liceases by Westinghouse have detemined that a $G tube similar to R9 C51 itquires a 105 reduction in the stability ratio in order s to reduce the fatigue usage factor (cycles to failutt par year) to a value substantially less than unity (1.0) for the remaining design life of the SGs. Thus, at 505 load, the required reduction in stability ratio is achieved with atlea(tafactorof2.6forsafety. Therefore, it is highly unlikely that a fatigue' crack would initiate due to this mechanism at 50% load and prior to any I modifications. I %p \\ f8 l 7 ~ g l r j
y ~ L j- .i k \\ ..c O g \\ [v I g 3 5, i In addition, the licensee is modifying the steam generators ta install 0FRP's l \\ and preventiveldplay'*kertdin tabes which are conslenred most susceptible to firfdelasticinstab.yity.jThelicenseehasdeterminedthatthenoteffectof aparatioT at 5(R00a4 fa,1cer6fnetton with the DFRP and prwantattwnplupctng repreuwu a. reduction in' stability rettos of 605 relativiita stability ratios!, 3 at 1910 fand and prior to any modifications. I ruductim in stat.ili*.y ratio is achieved with a facter of E.0 for safeW(Thus) Finall tions,y, the licenserha>s deteraiced that at SWE Tood, and with these atmffca s l no tube reaafahic in dervice will have a stabHity ratiu greater ti.s Y unity. cannot initiate due to the fluid elastic instability.anchtnise at 5M p y 4 j SNIRR7} 'I 3 \\ . ;t 1 The staff concurs with the licensee'that the cause ut h.hs $G C, AS C51 tube failure 3es fatigue. The staff is continuing to evaluate the exact machanian of the' cube failure and the Ifeensos's overall program of multiple actions designed to prevent a simih r tube egture event. Newever the staff finds thera G reasonable assurance that the adverse fluid fisw c,ondiths which lw, to excessive tube vibration that caused the ra not t* present at 50 percent power operation. pfd 9Papagation fracture will Thar6 fore interim operation'at i, ' q g reduced puwer (less than or equal to 50 ptet. ant power) is, acceptabl6; f 1 The staff's final evaluation of the tube 1411ure anchanism and multip e actioni Y designed tq prevent a similar tsta ruptur6 will be completed and issund prior 3 x i to any eutaorization for opeM. ion at gr$ater than 505 power. gelamentatingand Verificgtjon of Modifications PrfM,To testart (Mode 2): PriortorestartofNA-1(Mohe2),NRC,RegionIItillverifythatthefollow-ing modifications and procedures have been completed or folloued as specified ,\\ b1w: s i
- s s
(1) The e$rivacy of thh licensee's operating procedures for 3G 1efdge rats i' s i i surystilance,, i, 1, 1 s (2) ThaVNGtubeR9C51hasbeenstabilizedinconf4rshncewithvendor(W) , recessendations, o' (3) That flow restrictor plates have been installed fa conformance with vendor (W) recommendations, und 1 s (4) That applicah% precedures have been followel for losse parts accountability. e \\ In addition, prior' to any powr ascension ereeMr than 30 percent Athe o t ity of the newly installed 9 8 monitor shall be verified to be ag rable.perabil- ? ggt 3g: cesober 5, 1987 'll l i 3 i Princine( Contributor: , Leon Engle t \\ n ,x i g g *g i dIb ) \\
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[Y I OCT 14 e47 '-Carolina Power and Light Company ATTN: Mr. E. E, Utley Senior Executive Vice' President Power' Supply and Engineering Nana Construction 1 P.30. Box 1551. Raleigh,7NC:'27602 ' Gentlemen: E
SUBJECT:
REPORT N0. 50-400/87-31 Thank.you ~ for.'your responses of September 23 and October 1,1987, to our . Notice of Violation, issued on September'8,~1987, concerning activities conducted dt' your Shearon Harris facility. We hape evaluated your responses and found
- j. 0)g that they' meet the.~ requirements : of 10 CFR 2.201.
We will examine the implementation of your corrective. actions during future inspections. E We appreciate your cooperation in this matter. Sincerely, mHM *tmm & DAVID R Vrenty i David M. Verre111, Chief . Reactor Projects Branch 1 Division of Reactor Projects i m cc: R. A.-Watson, Vice President Harris Nuclear Project D. L. Tibbitts,. Director of Regulatory Compliance (. ~J. L. Willis, Plant General ' Manager l 1 p4 bec: C. Barth, 0GC NRC Resident. Inspector A. Upchurch, Chairman, Triangle J 4 Council of Governments Document Control Desk State of North Carolina 4N .08 SVias:cdg hPFredrickson l 10/g/87 10/jp/87 1 t ____f___-____}}