IR 05000382/2006008

IR 05000382-06-008; Entergy Operations, Inc., 03/06-24/2006; Waterford Steam Electric Station, Unit 3; Biennial Baseline Inspection of the Identification and Resolution of Problems

ML061290021
Person / Time
Site:	Waterford
Issue date:	05/08/2006
From:	Smith L J Division of Reactor Safety IV
To:	Venable J E Entergy Operations
References
IR-06-008
Download: ML061290021 (35)

Text

May 8, 2006

Joseph E. Venable Vice Presi dent Operations Waterford Stea m Elec tric S tation Unit 3 Entergy Operation s, Inc.17265 Riv er Road Killona, Louisiana 70066-0751 SUBJE CT: W ATERFORD STEAM ELECTRIC STATI ON, UNIT 3 - NRC PROBLEM IDENTIFICATI ON AND RESOLUTION INSPECTI ON REPORT 05000382/2006 008 Dear Mr. V enable: On March 24 , 2006, the U.

S. Nuclear R egulatory Co mmission (NRC) completed a team inspe ction at yo ur Waterford St eam El ectric Stati on, Un it 3. The enc losed report docume nts the inspecti on findings, w hich we re discussed with y ou and other members of your staff during an exit mee ting on Ma rch 24, 2006.

This i nspect ion w as an exami natio n of acti viti es con ducted under your lice nse as they relat e to the identifica tion and reso lution of probl ems, compliance with the Commission'

s rules and regulations an d the conditi ons of your op erating licen se. The team rev iewed 237 conditi on reports , appa rent ca use an d root cause analy ses, as wel l as s upport ing do cuments. In addition, the team review ed crosscutting as pects of NRC- an d licensee-identified findi ngs and interv iew ed per sonne l regar ding th e safety consc ious work e nvir onment.On the basis o f the sample sel ected for revie w, there w ere no findings o f significance ide ntified during this i nspection. The team concluded that, in general , problems w ere properly identified, eval uated, and c orrecte d. The team co nclud ed tha t a pos itiv e safety-consc ious work environment existed a t your Waterford Steam Electric S tation, Unit 3. Several examples of minor problems were ide ntified, incl uding conditi ons adverse to quality that were no t identified and en tered i nto y our co rrecti ve ac tion p rogram.

Entergy Operation s, Inc.-2-In acco rdance wit h 10 C FR 2.3 90 of th e NRC's "Rul es of Pr actice ," a co py of t his l etter a nd its enclosure, an d your resp onse (if any)

will be avail able elec tronically for public i nspection i n the NRC Publ ic Document R oom or from the Publ icly Av ailable Records (PAR S) component of NRC's doc ument system (AD AMS). AD AMS is accessibl e from the NRC Web site at http://w ww.nrc.gov/readi ng-rm/ada ms.html (the P ubli c Ele ctroni c Read ing Ro om).Since rely , /RA/Linda Joy Smith, Chie f Plant Engine ering Branch Divi sion of Reac tor Safe ty Docket: 50-382 License: N PF-38 Enclo sure: NRC Inspecti on Report 050 00382/2006008 ATTACHME NT A: Suppl emental Information ATTAC HME NT B: Waterford 3 Pres suriz er Surge Line Tempera ture Ch ange Ra te ATTAC HME NT C: White P aper o n Effect o f Diese l Sump Pump I nopera bili ty on Ulti mate Heat S ink Ope rabil ity cc w/encl osure: Senior Vi ce President and Chief Operatin g Officer Entergy Operation s, Inc.P.O. Box 319 95 Jackson, MS 39286-1995 Vice Presi dent, Op eratio ns Sup port Entergy Operation s, Inc.P.O. Box 319 95 Jackson, MS 39286-1995 Wise, Carter, Child & Ca raway P.O. Box 651 Jackson, MS 39205 General Ma nager, Plant Operati ons W aterford 3 SES Entergy Operation s, Inc.17265 Riv er Road Killona, LA 70066-07 51 Entergy Operation s, Inc.-3-Manager - Li censing Ma nager W aterford 3 SES Entergy Operation s, Inc.17265 Riv er Road Killona, LA 70066-07 51 Chairman Louisiana Public Se rvice Commi ssion P.O. Box 911 54 Baton Rouge, LA 70821-9154 Director, Nuclear Safet y & Regul atory Affairs W aterford 3 SES Entergy Operation s, Inc.17265 Riv er Road Killona, LA 70066-07 51 Michael E. Henry, State Liais on Officer Depart ment of E nvir onmenta l Qual ity Permits Div ision P.O. Box 431 3 Baton Rouge, LA 70821-4313 Paris h Pres ident St. Charles P arish P.O. Box 302 Hahnvil le, LA 700 57 Winston & Strawn LLP 1 7 0 0 K S t r e e t , N.W.Washington, DC 20006-3817 Entergy Operation s, Inc.-4-Electronic distribution by RIV:

Region al Ad minis trator (BSM1)DRP Directo r (ATH)DRS D irecto r (DDC)DRS D eputy Direc tor (RJC1)Senio r Resi dent In specto r (MCH)Branch Chief, DRP/E (DNG)Senior Project Engineer, DRP/E (VGG)Team Leader, DRP/T SS (RLN1)RITS Co ordin ator (KEG)DRS STA (DAP)S. O'Co nnor, OE DO RIV Coord inato r (SCO)ROPrep orts W AT Site Se cretar y (AHY)ADAM S: / Yes G No Initi als: __ljs____ / Publi cly Av ailable G Non-Publicly Available G Sensitive

/ Non-Sensit ive DOCUMEN T: R\_WAT\2006\W T2006-08RP-ELC.w pd RI:DRP/E PE:DR P/B PE:DR P/A SOE:DR S/OE RI:DRP/E ELCro we/l mb DHOverland MABro wn MEMu rphy GFLarkin/RA/ T/RA//RA//RA//RA/ T 5/5/06 5/5/06 5/5/06 5/5/06 5/5/06 BC:DR P/E SRI:DRS/EB2 BC:DRS/EB2 DNGraves D. Proulx LJSmi th/RA//RA//RA/5/8/06 5/5/06 5/8/06 OFFICIAL RECORD COPY T=Telephone E=E-mai l F=Fax Enclo sure-1-ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket: 50-382 License: NPF-38 Report: 05000382/2006 008 Licensee: Entergy Operation s, Inc.Facility: Waterford Stea m Elec tric S tation , Unit 3 Location: Hwy. 18 Killona, Louisiana Dates: March 6-24 , 2006 Inspectors:

M. Brow n, Project Engine er, Projects Branch A E. Crowe, R esident Inspe ctor, Projects Branc h E G. Larkin, Reside nt Inspector, Projec ts Branch E M. Mu rphy, Seni or Operations E ngineer, Operation s Branch D. Overland, Project Engineer, Projects Branch B Approved by: L. J. Smith, Chi ef Engineering Bran ch 2 Divi sion of Reac tor Safe ty Enclo sure-2-SUMMA RY OF F INDING S IR 05000382/2006008; Enterg y Operations, Inc., 03/06-

24/2006; W aterford Steam Electric Statio n, Uni t 3; bi ennia l bas elin e ins pecti on of th e ide ntifica tion a nd res oluti on of pr oblems. The inspection was cond ucted by tw o resident i nspectors, one s enior operati ons engineer, a nd two project e ngineers. The NR C's program for ove rseeing the safe op eration of commercia l nuclear pow er reactors is described i n NUREG-1649 , "Reactor Oversi ght Process," Rev ision 3, dated July 2000.Identi ficatio n and Resol ution of Prob lems*The team revie wed 237 corrective action program docu ments, apparent an d root cause analyses, as well as supportin g documents to asse ss problem i dentification and resol ution activ ities. Base d on th is rev iew , the te am found the l icens ee's proces s to identify, pri oritize, e valuate, a nd correct probl ems was genera lly effectiv e; thresholds for identifying i ssues remained appropriatel y low and, in most cases, correctiv e actions were adequate to address co nditions ad verse to qual ity. Ho wever, a number of issues were ide ntified associ ated with the proper i dentification of degraded conditi ons in the plant. The team re viewe d corrective actions asso ciated wi th these degraded conditions and design i ssues at Waterford Steam Electric S tation, Unit 3, which had crosscuttin g aspects in th e area of proble m identificatio n and resolu tion.The team conclud ed that a posi tive safety-co nscious w ork environment exists at Waterf ord Steam Ele ctric Station, Unit 3, base d upon inte rviews conducted w ith plant personnel. The team determined that employe es and contracto rs feel free to raise safety concerns to their superv ision or b ring concerns to the employe e concerns program.Insp ecto r-Id enti fied and Sel f-Rev eal ing F ind ings None Enclo sure-3-REPORT D ETA ILS 4.OT HE R AC TI VI TI ES (O A)4OA2 Identi ficatio n and Resol ution of Prob lems a.Effectiveness of Probl em Identification (1)Inspection Sc ope The in specto rs rev iew ed ite ms sel ected across four of th e sev en cor nersto nes to determine if probl ems were be ing properly identified, c haracterized , and entered i nto the corrective a ction program for eva luation an d resolution. Specifical ly, the tea m's review included a selection of 237 condi tion reports, equi pment walkdow ns, review of operator logs, maintenance records, and s tation quarterly trend reports. The majority of the condition re ports were o pened and cl osed since the last NRC problem ide ntification an d resolution i nspection co mpleted on M ay 21, 200 4. The team also performed a histori cal revi ew o f condi tion r eports wri tten ov er the last 5 year s for the high p ressur e safety injection sy stem, main feedwa ter isolatio n valv es, main steam i solation v alves, ess ential chillers, a nd the emergency diesel gene rators. The team rev iewed a sample of lic ensee audits and s elf assessments, trend ing reports, sys tem health repo rts, and vari ous other reports and d ocumen ts rel ated to the pr oblem ident ificati on and resol ution program. The audit and self-assessment resul ts were compa red with the self-reveal ing and NRC-identified issues to determ ine the effectiveness of the audits and self assessments.

The team intervi ewed stati on personnel and eval uated correctiv e action doc umentation to determine the licensee'

s threshold for i dentifying probl ems and enteri ng them into the corrective a ction program. In a ddition, i n order to asse ss the lice nsee's hand ling of operat or ex perie nce, th e team r evie wed the li censee's ev aluat ion o f selec ted in dustry operating exp erience reports , includi ng licensee event reports, NRC generic l etters, NRC bull etins, and N RC information n otices, and gene ric vendo r notifications to assess if issues applicable to W aterford Steam Electric Station, Unit 3, were appropriately addressed.

A listing of spe cific documents re viewe d during the i nspection i s included in the attach ment to this r eport. (2)Assessment The team determined that, in general , problems w ere adequately identified a nd entered into the correc tive acti on program, as ev idenced by the relativ ely few findings identi fied during the asses sment period.

The licensee

's threshold for entering issue s into the corr ecti ve act ion pr ogr am was appr opri atel y low. Ho wever, the t eam f ound two examples o f ineffective proble m identificatio n during this inspection.

The license e also failed in some instances to identif y or document deficiencies, which resulted in NRC nonci ted v iola tions.

Enclo sure-4-Current Issues Example 1: The licens ee failed to i dentify multip le temperature ch anges of the pressurizer surge line, which exceeded the heatup and cooldown rate described in Section 5.4.3.1 of the station'

s Final S afety Analy sis Report. S pecifically , the inspecti on team discove red during a pl ant shutdow n in August 20 05 that the pres surizer surge line had experienced 19 changes in temperatur e, which exceeded this limit. This example is further d escrib ed in Secti on 4OA 2.e of th is rep ort.Example 2: The te am found the l icens ee's i denti ficatio n of adv erse tr ends t o be w eak. The inspection team review ed 17 condi tions reports, i n which the licen see documented inadequacies in the procure ment of replacement p arts for the station.

The license e had ident ified a trend of impro per pa rts pas sing th rough th e rece ipt i nspect ion, b ut fail ed to identify adv erse trends rel ated to lack of engin eering invo lvement, as required by th e procur ement p rocess; failu re to p erform pro fession al en gineer ing ev aluat ions for parts transferred into the system; and rec eipt inspec tion documents missing required attributes. These pr ocurement process weakne sses resulted in a nonseismically qualified sy nchronizati on switch being instal led in an otherwise operable e mergency diesel genera tor and a non conforming fuel oil nipple pa ssing receipt i nspection.

Example 3: Th e N RC id ent ifi ed tha t th e l ice nse e mi sse d se ve ral op por tun iti es t o i den tif y the c onta inme nt fan coo ler cond ensa te fl ow swi tche s tha t di d no t mee t the des ign requirements f or detecting a one gallon per minut e reactor coolant system leak (NRC Inspec tion R eport 0 50003 82/200 5005-0 1).Example 4: Control roo m operators missed several opportunities over a 32

.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> period to i dentify that a vacuum had been draw n on the reacto r coolant sy stem during refueling outage drai ndown con ditions (sel f-revealing, NRC Inspection Report 05000 382/20 05010-03).Historical Issue Example: The NRC id entified the l icensee fail ed to identi fy an inappro priate val ue of the unfiltered in-l eakage parameter used to calculate the control ro om operator dose for design basis accident con ditions i nvolvi ng radiological releases (N RC Inspectio n Report 05000 382/20 04006-01). b.Prioritiz ation and E valuatio n of Issues (1)Inspection Sc ope The team revie wed condi tion reports, en gineering operabi lity ev aluations, a nd operations op erability determination s to assess the licensee'

s ability to evalu ate the importance of the co nditions ad verse to qual ity. The team review ed a sample of condi tion r eports , failu re mode anal yses, appar ent ca use an d root cause analy ses, to ascertain w hether the li censee iden tified and con sidered the ful l extent o f conditions, Enclo sure-5-generic impli cations, common c auses, and pre vious occ urrences. The team also observed manag ement oversigh t of th e signif icant co nditions adverse t o qualit y, inc lud ing o ne C orre ctiv e Ac tion Rev iew Boa rd me etin g.In addition , the inspecto rs review ed license e evalua tions of selecte d industry operating experience reports, incl uding licen see event reports, NRC gene ric letters, NR C bulleti ns, NRC information notices, and generic ven dor notices to assess w hether issues appli cable to Waterford S team El ectric Stati on, Un it 3, w ere ap propri ately addre ssed. The tea m performe d a hi storic al rev iew of cond ition report s cov ering t he la st 5 y ears regarding the high p ressure safety i njection sy stem, the emergency diesel genera tors, main fee dwat er iso latio n val ves, essent ial c hill ers, an d the d ry co olin g towe r to determine if the l icensee had appropriatel y addressed long-standing i ssues and tho se that mi ght be a ge depe ndent.A listing of spe cific documents re viewe d during the i nspection i s included in the attach ment to this r eport. (2)Assessment The team concluded that pr oblems were generally prioritized and evaluated in accordance with the licensee's corrective action progr am guidance and NRC requirements. The team foun d that for the sampl e of root cause an alyses rev iewed, th at the license e was genera lly sel f critical and exhaustiv e in its re search into th e history of significant condi tions adv erse to quality. Howev er, the team found on e example of ineffecti ve pr oblem eval uatio n duri ng this insp ectio n. Current Issues Example 1: The inspectors discovere d the lice nsee had cate gorized the fai lure of a fuel oil p ipe n ippl e in t he Eme rgency Dies el Gen erator B in 2 002, a s a con ditio n adv erse to quali ty. Th e lic ensee follo wed Proced ure EN-LI-102 , "Corr ectiv e Acti on Pro cess,"Revisio n 4, in making the determination of significance. The inspectors foll owed the steps o f Proced ure EN-LI-102 and a rrive d at th e same leve l of si gnifica nce, h owev er, the procedure p rovides a provisi on for the Condi tion Revi ew Group to change the lev el of significance, as warranted by the conditi ons. The insp ectors determined that this w as a significant co ndition ad verse to qual ity becau se the failure rendered one emergency diese l ino perabl e. The Emergen cy Di esel Generat or A ex perie nced a failu re of it s corresponding fuel oil nip ple in 20 05. The lice nsee determined this failure was a significant condi tion adve rse to quality solely because of the re petitive nature of the failure. c.Effectiveness of Correc tive Acti ons (1)Inspection Sc ope The team revie wed 237 condition reports to veri fy that correctiv e actions rel ated to the issues wer e identif ied and imp lement ed in a time ly manner commen surate with safe ty, includin g corrective a ctions to add ress common cause or generic conce rns. The team Enclo sure-6-review ed correctiv e actions pl anned and i mplemented by the licen see and sampl ed specific techni cal issues to determine w hether adequate d ecisions re lated to structure

, syste m, and c ompone nt ope rabil ity w ere mad e. In addition , the team revi ewed a s ample of those co ndition rep orts written to address NRC inspe ction findings to ensure that the corrective actions adequate ly addres sed the issues as described in the inspection report writeups. T he team also reviewed a sample of corrective ac tions close d to other cond ition reports and programs, such a s work and engineering work req uests to ensure that the condit ion described was adequately addressed and corrected.

A listing of spe cific documents re viewe d during the i nspection i s included in the attach ment to this r eport. (2)Assessment The effectiveness of id entified correcti ve action s to address a dverse cond itions w as generally adequate. The tea m evaluated several occurrences where the licensee did not effect ivel y add ress co nditi ons ad verse to qual ity a nd cor rectiv e acti ons ta ken we re untimely a nd inapprop riate. These i ncluded fiv e examples , one identi fied by the te am and fou r by other NRC inspect ions, wher e the licen see fa iled to tak e promp t corre ctive acti ons to re sol ve l ong-s tand ing i ssue s. Th e tea m al so e val uate d ni ne o ther find ings identified by the NRC ba seline i nspection pro gram and other NRC inspection s at Waterf ord Steam Ele ctric Station, Unit 3, sin ce the last p roblem identi fication and resolut ion inspec tion tha t had cro sscutt ing aspe cts rela ted to pr ompt an d eff ective correc tive actio ns to r esolv e cond ition s adv erse to quali ty. Current Issues Example 1: The re actor c oolan t drai ndow n proc edure failed to id entify that te mporary vent rigs, required by procedur e to properly establish vent paths, included in-line ball valves in series with the vent path a nd also fail ed to direct th ose ball valves be opened to esta blis h the v ent pa th. The lice nsee w as aw are of a nd di d not fi x the proce dure to addres s the b all v alve s in 2 002 (N RC Ins pecti on Rep ort 050 00382/20050 10-02).Example 2: The NRC i dentified the l icensee fail ed to correct the condition which resulted in multiple cy cle timer failu res in the es sential chi ller (NRC Inspection Report 05000 382/20 05002-01).Example 3: The NRC identified the licensee failed to prevent recu rrence of throug h wall pipe leakage on the main stea m line Pi pe 2MS2-123. This defici ency resul ted in an unisolabl e steam leak requiri ng NRC approv al to dev iate from the America n Society of Mech anica l Engi neers Boil er and Press ure Co de Cas e N523-2 to p erform tem porary repai rs prev entin g a pla nt shu tdow n (NRC Inspe ction Repor t 0500 0382/2 00500 4-03).

Enclo sure-7-Historical Issues Example 1: The NRC i dentified the l icensee fail ed to correct a known deficie nt condition involv ing the failure to account for ins trument uncertainty to satisfy Techn ical Specification Surveil lance Require ment 4.7.6.5.a. This failure potenti ally affects the abil ity o f the con trol ro om env elope to per form its design functio n wi th resp ect to protecting operators from postulated de sign basis ac cidents resul ting in radio logical relea ses (N RC Ins pecti on Rep ort 050 00382/20040 06-03).Example 2: The NRC i dentified the l icensee fail ed to correct a known deficie nt condition involv ing multiple occasions of accumulator ov erpressure condi tions resulti ng from degraded hydraulic fluid adversely aff ecting the main feedwater isolation valve hydraulic actuator pressure re lief system. These over pressur e conditions potentially result in valve closure stroke ti mes outside de sign basis v alues (NRC Inspection Report 05000 382/20 04005-03).Example 3: The N RC id entifi ed the lice nsee fai led to prompt ly c orrect insta nces w here the main feedwater isolation valve actuator therm al relief valves failed to properly function. In one case, the li censee failed to properly address sy stem operabil ity and, for a 2-week peri od, actual v alve ope rability was unknow n (NRC Inspe ction Report 05000 382/20 04006-02).Example 4: The NRC i dentified the l icensee fail ed to correct de ficiencies i n the emergency dies el generator lo ading and fuel oil consumpti on analy sis. The li censee inappropriately closed a corrective action requiring the revisions, which subsequently resulted in the failure to mai ntain design control of the emergenc y diesel generator fuel oil s torage i nven tory r equire ments to ensur e a 7-d ay po stacci dent fue l oil inv entory (NRC I nspect ion R eport 0 50003 82/200 4002-0 5).Example 5: The NRC i dentified the l icensee fail ed to determine the cause an d precl uded r ecurre nce of ma in ste am iso latio n sol enoid-opera ted du mp val ve fai lures. The inspectors n oted that the l icensee's apparent cause did not pro vide an extent of condition a nalysis for the solenoi d-operated v alve fail ure (NRC Insp ection Report 05000 382/20 04004-03).Example 6: The N RC id entifi ed the lice nsee fai led to take ad equate correc tive actio n to ensure the torque a pplied to the flow con trol valv e for Accumulator B of main feedwate r isolation Valve 1 was sufficient to prevent an o-ring from extrudi ng, resulting in a loss-of-system hy draulic flui d and renderi ng the valv e inoperabl e (NRC Inspe ction Report 05000 382/20 04008-02).Example 7: The N RC id entifi ed the lice nsee fai led o n multi ple o ccasi ons to correc t a known defici ent co nditi on in volv ing the failu re to a ccount for ins trument uncer tainty to satisfy Techni cal S pecifi catio n Surv eill ance R equire ment 4.7.6.5.a. This fa ilure potentiall y affects the abil ity of the con trol room env elope to pe rform its design function with respect to protecting oper ators from pos tulated design basis accidents result ing in radio logica l rel eases (NRC I nspect ion R eport 0 50003 82/200 4006-0 3).

Enclo sure-8-Example 8: The licens ee failed to re place know n age-degraded o-rin gs affecting the main feedwater isolation valves in the Yea r 2000 resulti ng in o-ring failu re and inoperabil ity of the Train A feedwater i solation v alve on December 27, 20 03 (NRC I nspect ion R eport 0 50003 82/200 4002-0 1).Example 9: The NRC i dentified the l icensee fail ed to establi sh appropriate torque specification to ensure adequate o-ring compressio n that ultimate ly led to an o-ring failure and the inoperabi lity of the Trai n A main feedw ater isolati on valv e. The licen see had previ ously i dentified conce rns related to inadequate w ork instructions for performing maintenance ac tivities on the main feedwater isol ation val ves (NRC Inspection Report 05000 382/20 04002-02). d.Assessment of Safety-C onscious Work Environment (1)Inspection Sc ope The team intervi ewed 24 indivi duals from the li censee's staff, represen ting a cross sectio n of funct ional organi zati ons an d supe rvis ory a nd non superv isory perso nnel. These intervi ews assess ed whethe r conditions existed that woul d challen ge the establ ishmen t of a sa fety-co nscio us wo rk envi ronmen t. The t eam in tervi ewed the si te employ ee con cerns p rogram co ordin ator. (2)Assessment The team conclud ed that a posi tive safety-co nscious w ork environment exists at W aterford Steam Electric Station, Unit 3. Based on interviews, station personnel felt free to enter issues into th e corrective action program , raise safety concerns with their supervisi on, to the empl oyee conce rns program, and to th e NRC. The te am determined that the majority of safety concerns were addre ssed through the s ite's normal chain of command by the relative ly few sa fety concerns en tered into the employee c oncerns program and the sm all number of allegations made t o the NRC.

e.Specific Issues Identified Duri ng this Inspecti on (1)Inspection Sc ope During this assessment, the team perform ed the inspections scoped in Secti ons 4OA 2 a.(1), 4OA2 b.(1), 4 OA2 c.(1), and 4OA2 d.(1) ab ove. (2)Finding Details (i)Unresolve d Item: 05000382/2 006008-01 , "Failure to Maintai n Design Con trol of the Pressurizer Surge Line" Introduction. The team iden tified an unres olved i tem related to co mpliance w ith 10 CFR Part 50 , Appe ndix B, Cri terion III, "De sign C ontrol ," for the failu re to tr ansla te design-basi s heat up and cool down rates for the p ressur izer surge l ine i nto ap propri ate specifications , procedures, an d instruction s. As a resul t, Entergy Operatio ns, Inc., failed Enclo sure-9-to effectively control and e valuate p ressurizer s urge line tempera ture changes on numero us occ asion s. Description. Final Safety Analysi s Report (FSAR) Section 5.4

.3.1, "Reactor Co olant Pipi ng Desi gn Basi s," and Secti on 5.4.10.1, "Press uriz er Des ign Ba sis," s tates, in pa rt, that duri ng he atup and coo ldo wn of the pla nt, t he a llo wab le r ate o f temp erat ure c hang e for the s urge li ne is limi ted to 200°F/hr. Tec hnica l Requ iremen ts Ma nual (TRM), Section 3.4.8.2 , "Pressurize r Heatup/Cool down," spec ifies the li miting conditi on for operation, in part, as a max imum heatup rate of 200°F per hour and a maxi mum coold own rate of 1 35°F p er hou r.On Apri l 18, 2005, Entergy Condi tion R eport C R-WF3-2005-1 392 st ated th at a pressu rize r surge line temper ature t ransi ent oc curred wit h the s urge li ne temp erature dropping from 425°F to 140°F, a change of appr oximately 285°F with approximately 200°F occurr ing w ithin 8 minu tes. Tec hnica l Requ iremen ts Ma nual, Secti on 3.4.8.2 Actio n spec ifies, "With any o f the pre ssuriz er li mits i n exc ess of th e abov e, the operat ors must restore the affected pa rameter to wi thin the li mits withi n 30 minutes; perform an engineering ev aluation to determine the effects of the out-of-limit cond ition on th e structu ral i ntegrit y of th e pres suriz er; and enter TRM L CO 3.0.3." The team noted that Entergy Operation s, Inc., failed to restore pressuriz er/surge line limit s wi thin 3 0 minu tes an d perfor m an en gineer ing ev aluat ion to determ ine th e effects of the out-of-limit con dition on the structural i ntegrity of the pres surizer/surge l ine. The team review ed Entergy Opera tions, Inc.'s o perating procedure s for plant heatup and coold own activ ities , OP-01 0-005, "Plan t Shutd own," and OP-010-0 03, "Pl ant Sta rtup,"and did no t find procedure steps to limi t surge line te mperature changes to less than 200°F/hr, nor w ere there any procedure step s to assess w hether surge li ne stress or fatigue limits had been exc eeded. This a ppeared to be a viola tion of 10 CFR Part 50, Appendix B, Criterion III, "Design Contr ol," for the f ailure to translate design-bas is heatup and co oldown rates for the pressuri zer surge li ne into app ropriate speci fications, proced ures, a nd in structi ons. The des ign li mit in Repor t CEN-387-P was b ased, in pa rt, by temper ature gr adien ts greater than 200°F occurring less than 3.6 occurre nces per heatu p/cooldow n cycle for 500 heatup/coo ldown c ycles ov er the 40-ye ar life of the pl ant. Calcul ation CN-OA-04

-53 documented 19 i nstances w here pressuriz er insurges, in excess of the v olume of the surge line, occ urred with a temperature gradi ent greater than 20 0°F. These pressu rizer insurges occurred during five refuel ing outage heatup/c ooldown cycles (R efueling Outages 8-12)for an av erage of 3.8 temperature gradi ents greater than 20 0°F per heatup/cooldo wn cycl e.Entergy Operation s, Inc. disagreed and provi ded a paper (Attachment B), w hich documented thei r position.

While they acknowle dged that the FSA R was no t up to date, they stated th at the pressuri zer surge li ne temperature trans ient on Apri l 18, 2005, was bounded by Combustion E ngineering Own ers Group Report CEN-387-P, "Pre ssurizer Surge Line Flow Stratification Evaluation," subm itted to the NRC in response to NRC Bull etin 8 8-11, "Pressu rize r Surge Line Thermal Strati ficatio n." Report CEN-3 87-P concluded th at the pressuri zer surge li ne met all a pplicabl e design code s, FSAR, and Enclo sure-10-other regulatory commitments for the li censed life o f the plant consi dering the phenomenon of thermal stratification i n fatigue and stress evaluati ons. The team note d that this conclusion was based on operating the plant consist ent with the assumptions in the evalu ation (Report CEN-387-P). A dditional inspection is required to complete the review of Entergy Operatio ns, Inc.'s, positi on and determi ne whethe r the licens ee was operating their facility w ithin the a ssumptions of the a nalysis.Analysis. The significance of this issue depends on w hether or not the analysi s bounds past plant op eration. Enforcement. The p otenti al fai lure t o trans late t he des ign ba sis i nto ap propri ate specifications , procedures, an d instruction s to effectively control and evaluate surge line tempera ture ch anges, d uring p lant h eatup and co oldow n, that exce eded t hose l imits descri bed i n the F SAR a nd the TRM i s unre solv ed: (U RI 050 00382/20060 08-01);"Fail ure to Mai ntain Desi gn Cont rol of t he Pre ssuriz er Surge Line." (ii)Unresolve d Item 05000382

/2006008-02 , "Failure to Ensure that Written Procedures Adequately Incorporate Regul atory Require ments and Desi gn Basis"Introduction. The te am ide ntifie d an u nresol ved i tem rel ated to compl iance wit h Technical Sp ecification, S ection 6.8.1, for the failure to ens ure that wri tten procedures adequately i ncorporate regulato ry requirements a nd the design basis for the dry cooling tower die sel-drive n sump pumps.

Description. Waterford Safety Evaluation Report, Suppl ement 4, Sectio n 2.4.2.3, discusses the design basi s rainfall ev ent and combi nation of eve nts. This suppl ement commits the li censee to the probable max imum precipita tion even t. Because of the fact that the motor-driv en sumps are no t seismicall y qualified, the NRC requeste d the licensee a nalyze the effects of a standard p roject storm, whi ch consists o f 50 percent of the probable maximum precipation event concur rent with an operating basis earthquake. The results of the licensee's analysis showed the licensee was susceptible to ponding in the dry co oling towe r sumps, assuming the loss of all motor-driven pumps, which w ould endan ger the safety-relate d transformers and motor c ontrol centers located in the cool ing tow er area s. The li censee submi tted Am endmen t 34, d ated J anuary 1984, subse quent to Safety Evaluati on Report, Sup plement 4. Se ction 2.4.2.3.4 of this amendment su bmittal contains an analysi s showin g the probabil ity of standard project storm and operating basis earthquake i s 3.6E-08, w hich is co nsidered negl igible. Ho wever, th e licensee prop osed to pr ovide a 100 g pm po rta ble pu mp th at wou ld be s uff icien t to p ump d own the dry cooling tower sumps in the event of the standar d project storm.

The NRC determined that the portable p ump was sufficie nt (as evi denced in Safety Eval uation Report , Supp lement 4) pro vide d the p ump w as pl aced i n oper ation wit hin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. In 2000, after determini ng that more sump pumpi ng capacity was neede d, the licen see installed a diesel-dri ven sump pu mp, with 30 0 gpm capacity , in each d ry cooli ng tower sump. The Design Basis Cal culation E C-M99-010 analyz ed for a probabl e maximum precipation event, concu rrent with a loss-of-offsite powe r, and determine d that a higher capacity portable pump was needed. The calculation also analyzed for a rainfall equivalent to 60 percent o f the probable max imum precipati on event, co ncurrent wi th a loss o f all mo tor-dri ven s ump pum ps, an d dete rmined that a 300 gp m porta ble p ump Enclo sure-11-would b e sufficient. The li censee's Pro cedure OP-100-01 4, "Technical Specification s and Requirements Compliance,"

Revisio n 14, states tha t two motor-dri ven sump pu mps or one motor-driv en pump and o ne diesel-driven pu mp are required for ul timate heat sink operabil ity. This p rocedure impl ies that the d iesel dri ven sump pu mp can be out of ser vi ce i nde fin ite ly wi tho ut a ffec tin g op era bi li ty of t he ul tim ate he at s in k. Th e N RC sta ff believes this procedure does not adequately address the requirement of the portable sump pu mp in t he des ign ba sis of t he ul timate heat s ink, no r does the pr ocedu re requi re any compensa tory actions be taken in th e event the diesel-dri ven sump pu mp becomes inope rable. Als o, the staff bel ieve s the c ontrol s and locat ion o f the di esel-driv en sump pump ar e not a dequate ly a ddress ed by the l icens ee. Analysis. The significance of this issue h as not been d etermined.

Enforcement. The licens ee has prov ided a pos ition pape r (Attachment C) rel ated to the design basis requirements for the dry cooling tow er diesel-dri ven sump pu mps, which has no t been fully revi ewed by th e NRC. The p otenti al fai lure t o ensu re regul atory require ments for these pumps is un resol ved: (URI 0 50003 82/200 6008-0 2) "Fai lure t o Translate Desi gn Control in to Station Do cuments Regarding D iesel-driv en Dry Co oling Tower Sump P umps" 4OA6 Exit M eeting The team discusse d the findings of the Problem Identi fication and R esolution i nspection with Mr. J. Venable, Vice President Operations, and other me mbers of the licensee's staff on March 24, 2006. Licen see management did not identi fy any materi als exami ned during the inspe ction as proprie tary.The licensee acknowledged the f indings presented. The inspect ors noted that while propri etary informa tion w as rev iew ed, no ne w ould be in clude d in t his re port.ATTACHMENT A

Supplementa l Information ATTACHM ENT B: Waterford 3 Pr essuri zer S urge Li ne Tempe rature Change Rate ATTACHM ENT C: White Pape r on Effec t of Die sel S ump Pu mp Inop erabi lity on Ul timate Heat Sink Op erabi lity Attach ment A A-1 KEY POINTS OF CONTACT Licensee Pe rsonnel B. Baxter, C ontrol Room S upervisor C. DeDeaux Sr., Senior P roject Manager, L icensing R. Dodds, M anager, Operations R. Fletcher, Trai ning Mana ger C. Fugate, Assis tant Operations M anager J. Hall, Opera tions Trainin g Supervisor - Operator Requali fication J. Holman, M anager, Nuclear Engineering J. Laque, Man ager, Mainten ance R. Muril lo, Senior Staff Engineer R. Osbo rne, M anager, Engin eerin g Programs and C ompone nts A. Pilutti, Manager, Rad iation Prote ction O. Pipkins, Seni or Licensin g Engineer R. Porter, Superi ntendent, M echanical Maintena nce B. Proctor, Sy stems Engineerin g Manager J. Rachal, D esign Engineeri ng Superviso r J. Ridgel, M anager, Correctiv e Action Pro gram T. Tankersley, Actin g Director, Nucl ear Safety Assu rance K. Walsh, General Mana ger, Plant Operatio ns B. Williams, Engineering D irector J. Venable, Site Vice President, Waterford 3 NRC M. Hay , Senior Re sident Inspecto r Waterford 3 LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 05000382/2006 008-01 URI Fai lure to M ain tain Des ign C ontr ol o f the Pres suri zer Sur ge Line (Secti on 4OA 2 e.)05000382/2006 008-02 URI Fail ure to Transl ate De sign C ontrol into Statio n Docu ments Regarding Diese l-driven Dry Cool ing Tower Sump Pumps (Secti on 4OA 2 e.)

Attach ment A A-2 LIST OF DOCUMENTS REVIEWED Plant Proced ures NAME TITLE REVIS ION CEP-IS T-1 IST Bases Docume nt 3 EN-OP-115 Conduct of Operatio ns 0 LI-102 Corrective Action Proce ss 4 LI-19645 Quali ty Re lated Admin istrat ive Proced ure 2 MM-006-119 Yard Oil Se parator to CW Temporary Pumping S ystem 0 OI-042-000 Watch Station Procedures 1 OP-001-003 Reac tor Coola nt Sys tem Drain down 23 OP-005-004 Main S team 12 OP-009-008 Safety Injection System 18 OP-100-001 Operations Stan dards and M anagement Expe ctations 22 OP-100-009 Contro l of Va lves and B reakers 17 OP-100-0014 Techni cal S pecifi catio n and Techni cal R equire ments Compliance 13 UNT-005-004 Temporary Altera tion Control 16 Engine ering R eports ER-W3-2002-0055 ER-W3-2004-0537 ER-W3-2005-0426 ER-W3-00-0337 ER-W3-2003-0010 ER-W3-2005-0305 ER-W3-2002-0278 Calculati ons CN-OA-04-53 EC-M99-01 0 MN(Q)-6-27 Root Cause A nalysis Reports for CR-WF3-

2001-0317 2002-0339 2003-0062 2003-3891 2004-759 2004-1011 Attach ment A A-3 Condition Reports, CR-WF3-

1997-1227 2000-0441 2000-1347 2000-1455 2001-0596 2001-0673 2001-0782 2001-1284 2001-1367 2002-0468 2002-0470 2002-0588 2002-0678 2002-1410 2002-1842 2002-2799 2003-0147 2003-0577 2003-1192 2003-1202 2003-2758 2003-2759 2003-2991 2003-3088 2003-3649 2003-3891 2004-0251 2004-0304 2004-0309 2004-0326 2004-0420 2004-0464 2004-0483 2004-0494 2004-0508 2004-0634 2004-0651 2004-0701 2004-0703 2004-0721 2004-0759 2004-0821 2004-0835 2004-0865 2004-0903 2004-1011 2004-1047 2004-1190 2004-1208 2004-1312 2004-1340 2004-1446 2004-1480 2004-1518 2004-1553 2004-1572 2004-1593 2004-1621 2004-1645 2004-1646 2004-1668 2004-1679 2004-1684 2004-1716 2004-1751 2004-1753 2004-1763 2004-1810 2004-1850 2004-1854 2004-1855 2004-1863 2004-1880 2004-1942 2004-2002 2004-2228 2004-2290 2004-2320 2004-2326 2004-2382 2004-2404 2004-2487 2004-2496 2004-2517 2004-2520 2004-2522 2004-2545 2004-2547 2004-2549 2004-2638 2004-2690 2004-2722 2004-2734 2004-2766 2004-2884 2004-2890 2004-2928 2004-2973 2004-2995 2004-3066 2004-3130 2004-3200 2004-3219 2004-3244 2004-3413 2004-3460 2004-3464 2004-3695 2004-3720 2004-3725 2004-3753 2004-3853 2004-3881 2004-3924 2004-3944 2004-3949 2004-4000 2005-0033 2005-0081 2005-0098 2005-0109 2005-0132 2005-0134 2005-0197 2005-0217 2005-0346 2005-0413 2005-0415 2005-0471 2005-0489 2005-0530 2005-0587 2005-0590 2005-0591 2005-0592 2005-0608 2005-0717 2005-0763 2005-0804 2005-0805 2005-0806 2005-0839 2005-0852 2005-0921 2005-0966 2005-0967 2005-1132 2005-1143 2005-1173 2005-1247 2005-1260 2005-1279 2005-1315 2005-1332 2005-1346 2005-1362 2005-1363 2005-1392 2005-1463 2005-1626 2005-1646 2005-1694 2005-1821 2005-1836 2005-2070 2005-2139 2005-2267 2005-2272 2005-2350 2005-2402 2005-2469 2005-2489 2005-2536 2005-2546 2005-2548 2005-2600 2005-2679 2005-2685 2005-2695 2005-2780 2005-2799 2005-2819 2005-2837 2005-2844 2005-2869 2005-2874 2005-2990 2005-3006 2005-3091 2005-3293 2005-3308 2005-3455 2005-3474 2005-3659 2005-3698 2005-3812 2005-3822 2005-3830 2005-3831 2005-3840 2005-3872 2005-3872 2005-3902 2005-3914 2005-3924 2005-3928 2005-3960 2005-3961 2005-3985 2005-4038 2005-4065 2005-4066 2005-4067 2005-4147 2005-4149 2005-4151 2005-4173 2005-4251 2005-4444 2005-4480 2005-4597 2005-4647 2005-4694 2005-4915 2005-4917 2005-4929 2005-5024 2006-0006 2006-0058 2006-0164 2006-0200 2006-0380 2006-0492 2006-0759 2006-0767 2006-0839 2006-0895 Attach ment A A-4 Learni ng Organi zati on Con ditio ns Rep orts LO-OPX-2004-0247 LO-OPX-2005-0100 LO-OPX-2005-0217 LO-OPX-2006-0011 LO-OPX-2005-0036 LO-OPX-2005-0103 LO-OPX-2005-0243 LO-OPX-2006-0034 LO-OPX-2005-0085 LO-OPX-2005-0132 LO-OPX-2005-0252 Work Orders 51697 51699 52824 52825 57759 62641 72604 72606 412565 4281801 4599901 5100331101 Mai ntenan ce Act ion It ems 420105 438981 Mis cell aneou s Docu ments Commercial Grade Evaluati on 01214 C-PAC-002 L-19645 L-23993 MMR Project 53465 PO WPY20583 2004 S econd Quarter Waterford Quart erly Trend R eport 2004 Th ird Qua rter Waterford Qu arterl y Tren d Repo rt 2004 F ourth Qu arter Waterford Quarter ly Tre nd Rep ort Quali ty As suranc e Audi t Repo rt QA-12-20050-WF3-1 Quality A ssurance Audi t Report QA-12-200 50-WF3-009 Quali ty As suranc e Audi t Repo rt QA-12-20050-WF3-1 PO 10083675 INITIAL MATERIAL REQUEST INITIAL INFORMA TION REQUEST FROM WATERFORD 3 FOR PI&R INSP ECTION (Report Numbe r 05000382/200 4006)The inspection will cover the period of October 2 002 to Ma rch 2004. The i nformation may be provided in either electronic or paper media or a combination ther eof. Infor mation provided in electronic med ia may be in the form of CDs, or 3-1/2 inch floppy disks. The agency

's text ed iting software is C orel WordPerfect 8, Presentations, and Quattro Pro; h owever, we hav e document viewi ng capabili ty for MS Word, Excel , Power Po int, and Ado be Acrobat (.pdf) tex t files.Please prov ide the follo wing informatio n to Peter Al ter by M arch 29, 2004 at the Resid ent Inspec tor Office at Waterford-3 Attach ment A A-5 All proced ures governin g or applyi ng to the correctiv e action program, i ncluding the processing of informatio n regarding generic communications and industry operating experience s Procedures and description s of any informal systems, used by engine ering, operation s, maintenance, security, tr aining, and emergency planning f or issues below the threshold of the formal correctiv e action program A sear chabl e tabl e of al l corr ectiv e acti on doc uments (condi tion r eports) that w ere initiated or closed du ring the period , include condition re port number, descri ption of issue and s ignificance cl assification Either annota te on the abo ve list or a separate l ist of all co ndition rep orts associated with: (1)Human performance i ssues (2)Emergency prepare dness issues (3)Respo nse to 10 CF R Part 21 rep orts A separate l ist of all co ndition rep orts closed to other programs, such a s maintenance action items/w ork orders, engineeri ng requests, etc.

A copy of eac h significant ev ent review team report and root cause ana lysis rep ort for the pe riod (not ne cessar ily the w hole condi tion r eport)Copies of cond ition reports (for the period) ass ociated w ith nonescal ated (no respon se required) or nonc ited vio lations for the period Copies of cond ition reports for the period as sociated w ith repetitiv e problems or issues Copies of condition reports f or the period associated with ineffec tive or untimely corrective a ctions List of all s elf assessments or quality ass urance assessmen ts/audits for the pe riod All correcti ve action program reports or metri cs used for tracking effective ness of the corrective a ction program for the pe riod All quali ty assurance audits and surveill ances, and functi onal self asse ssments of corrective a ction activ ities comple ted for the period Control room l ogs for the Year 2003 Security e vent logs for the year 2003 Radiation protection ev ent logs for the y ear 2003 List o f risk si gnifica nt sy stems fro m W3 PRA/PSA , base d on ri sk achi eveme nt wo rth (RAW) and "0% availabi lity CD F" Attach ment A A-6 Searchable list of all maintenance ac tion items/w ork orders for the perio d List of all S SC's plac ed in or remov ed from the maintenan ce rule a(1) c ategory for the period All correcti ve action documents rela ted to the follo wing indu stry operatin g experience generic communica tions: NRC Bull etins NRC Bull etin 2002-001 , "Reactor Pressu re Vessel H ead Degradation and Reactor Coolant Pressur e Boundar y Integ rity"NRC Information N otices NRC Information N otice 2004-00 1, "Auxil iary Feed water Pump R ecirculatio n Line Orifice Fouling - Po tential Common Cause Fai lure"NRC In formatio n Noti ce 200 3-019, "Unan alyz ed Con ditio n of Rea ctor Co olant Pump Seal Leakoff Li ne Dur ing Po stula ted Fi re Sce nario s or St ation Blac kout"NRC Information Notice 2003-

013, "Steam Generat or Tube Degradation at Diablo Canyon"NRC Information N otice 2003-01 1, "Leakage Found o n Bottom-Moun ted Instrumentation No zzles"NRC In formatio n Noti ce 200 3-008, "Poten tial Flood ing Thro ugh Uns ealed Concr ete Floor Cracks" NRC Information N otice 2003-00 5, "Failure to Detect Freesp an Cracks in P WR Steam Generator Tubes" NRC Information N otice 2003-00 2, "Recent Ex perience With Reactor Coo lant Syste m Leakage And Bori c Acid Corro sion"NRC Information N otice 2002-03 4, "Failure of Safety-Related Circuit Brea ker External Auxili ary Swi tches at Colu mbia Generating S tation" Attach ment A A-7 Information Reque st 1 - January 2006 Waterford PIR Ins pection (IP 711 52; Inspection Report 50-382/0 6-08)The inspection will cover the period of M arch 1, 2004 to March 1, 2006. All requested information shoul d be limite d to this peri od unless o therwise s pecified. The in formation may be provided in either e lectronic or paper media o f a combination of this media. Informatio n provided in electroni c media may be in the form of e-mai l attachment(s), CD s, or 3 1/2 inch floppy disks. The agency's tex t editing softwa re is Corel WordPerfect 10, Presentations, and Quattro Pro; ho wev er, w e hav e docu ment v iew ing cap abil ity fo r MS Word, Excel , Pow erPoi nt, and Adobe A crobat (pdf.) text fil es.Please prov ide the follo wing by February 8 , 2006, to:

U.S. Nuclear Regulatory C ommission Resident Inspector's Of fice - Attn. Gran t Larkin Waterford Stea m Elec tric S tation Unit 3 Entergy Operation s, Inc.17265 Riv er Road Killona, Louisiana 70066 Note: On summary li sts please i nclude a d escription o f problem, status, in itiating date, and owner organi zation 1.Summary lis t of all condi tion reports op ened during the period 2.Summary list of all open condi tion r eports wit h signi ficance of "B" o r greate r whi ch we re generated during the period 3.Summary lis t of all condi tion reports w ith significanc e of "B" or greater clo sed during the specified peri od 4.Summary list of all condition reports which were down-graded or up-g raded in significance du ring the period 5.A list of all corrective action documen ts that subsume o r "roll-up" on e or more small er issues for the pe riod 6.List of all ro ot cause anal yses comple ted during the p eriod 7.List of all a pparent cause analyses completed duri ng the period 8.List of root cause analyse s planned, b ut not complete at end of the peri od 9.List of plant sa fety issues rai sed or address ed by the employee c oncerns program during the peri od 10.List of action i tems generated or ad dressed by the plant safety review committees during the peri od Attach ment A A-8 11.Summary list of oper ator w ork-arou nds, e nginee ring re view request s and/or ope rabil ity evaluati ons, temporary mo difications, sa fety system defici encies, and control room deficiencies 12.All quali ty assurance audits and surveill ances of correctiv e action acti vities co mpleted during the peri od 13.A list of all quality a ssurance audi ts and surve illances scheduled for co mpletion duri ng the period, bu t which were not co mpleted 14.All corrective action activity reports, funct ional area self-assessmen ts, and non-NRC third party assessments compl eted during the period 15.Corrective action performance tre nding/tracking information generated during the period and broken dow n by function al organiza tion 16.Curren t proce dures/poli cies/gu idel ines for: 1.Condition Reporting 2.Corrective Action Program 3.Root Cause E valuatio n/Determination 4.Deficiency Reporting and R esolution 17.A listing of al l external events ev aluated for appl icabili ty at Waterford during the period 18.Cond ition Repor ts or othe r act ions gene rat ed f or ea ch of the it ems b elow [A DAMS access ion n umbers or othe r cross referen ce li sted for some]: 1.Part 21 Reports (2005-41-00; 20 05-38-00 [ml053 180299]; 2005-3 7-00;2005-33-01 [ml0 52860229]; 200 5-30-01 [ml0526 40220]; 2005-26

-01[ml052 91038 9]; 200 5-22-0 0; 200 5-20-0 0; 200 5-17-0 0 [ml05 11100 87];2005-16-00 [ml0 51100285]; 200 5-13-00 [ml0509 50428]; 2005-12

-01[ml052 08036 8]; 200 5-12-0 0 [ml05 06302 75]; 20 05-10-00 [ml0 50560 142];2005-0 7-00; 2 005-05-01 [ml 05110 0355]; 2005-01-00 [ml043 52007 7];2004-27-01 [ml0 43280541]; 200 4-24-00 [ml0424 70299]; 2004-22

-00[ml042 66017 5]; 200 4-21-0 0 [ml04 25200 48]; 20 04-17-00 [ml0 41900 058];2004-15-00; 200 4-14-00; 2004-10

-00 [ml0411403 35]; 2004-08-00

[ml041110893]; 2 004-02-01 [ml04 0420567]2.NRC Information N otices 05-32; 05-31; 05-30; 05

-29; 05-26; 05-25

05-24; 05-

23; 05-21; 05-19

05-16; 05-11; 0 5-09; 05-08; 05-0 6; 05-02;04-021

04-019; 04-

016;04-012; 04

-011;04-010; 04

-009;04-008; 04

-007;04-001 3.All LERs issued by Waterford during the period 4.NCVs and V iolations issued to Waterford during the peri od 19.Safeguards event l ogs for the period 20.Rad iati on p rote ctio n ev ent l ogs Attach ment A A-9 21.Current system health reports or similar i nformation 22.Current predicti ve performance su mmary reports or s imilar informati on 23.Corrective action effectiven ess review reports generated d uring the perio d

ATTAC HM ENT B Waterford 3 Pres surizer Surge Lin e Temperature Change R ate Waterford 3 Pressurizer Surge Line Temperature Change RatePage 1 of 4PurposeThis Paper is to document the Entergy position on the potential NCV of 10CFR50Appendix B, Criterion III, "Design Control" for not translating design basis criteriainto plant operating procedures. The design basis criteria in question is astatement in the FSAR (Section 5.4.3.1) which states:During heatup and cooldown of the plant, the allowable rate oftemperature change for the surge line is increased to 200°F/hr as a designrequirement specified in Subsection 3.9.1.1.BackgroundThe following is a time line of the Entergy response to NRC Bulletin No. 88-11.This concludes that the fatigue life of the Waterford 3 surge line is 40 years whichthe NRC concurred with.* The NRC issued NRC Bulletin No. 88-11, Pressurizer Surge Line ThermalStratification, on December 20, 1988. The purpose of the Bulletin was torequest that addressees establish and implement a program to confirmpressurizer surge line integrity in view of the occurrence of thermalstratification and to inform the staff of the actions taken to resolve thisissue.* CEN-387-P was transmitted to the NRC on July 27, 1989. Thisdocumented that the Waterford 3 surge line fatigue life is longer than 40years.* On August 28, 1989, Entergy sent a letter to the NRC stating that Bulletin88-11 item 1b, 1c and 1d were addressed in CEN-387-P and that item 1a(visual inspection of the pressurizer surge line) would be addressed duringthe next refueling outage.* On March 7, 1990, Entergy sent a letter to the NRC which addressed theresults of the visual inspections of the pressurizer surge line. The letterconcluded that the Waterford 3 surge line was structurally sound.* On August 15, 1990, the NRC issued a letter stating there was not enoughinformation in the CEN document to conclude that the pressurizer surgeline meets all appropriate Code limits for a 40 year plant life.* On December 20, 1991, CEN-387-P, Revision 1-P was sent to the NRC toaddress the concerns of the August 15, 1990 NRC evaluation of the CENdocument.* On May 5, 1992, Entergy sent a letter to the NRC documenting thesubmittal of the revised CEN document and stated the only remainingaction to complete the response to the Bulletin is for the Waterford 3 toupdate the pressurizer surge line design documentation. This wascommitted to be completed within 180 days of issuance of a favorableSER by the NRC.

Waterford 3 Pressurizer Surge Line Temperature Change RatePage 2 of 4* On June 22, 1993, the NRC issued an SER for CEN-387-P, Revision 1. Itwas concluded that the analysis in the CEN adequately demonstrates thatthe bounding surge line and nozzles meet ASME Code stress and fatiguerequirements for the 40 year design life of the facility considering thephenomenon of thermal stratification and thermal stripping. The staffrequested Entergy to provide a final status of the Waterford 3 activitiesrequired by NRC Bulletin 88-11.* On December 23, 1993, Entergy sent a letter to the NRC stating that alldesign documents had been updated and that all actions required by NRCBulletin 88-11 had been completed.CEN-387-P, Revision 1, is the Combustion Engineering response to NRCBulletin 88-11. This document addresses pressurizer surge line flowstratification. The document provides a detailed fatigue analysis of stress due tostratified temperature profiles of the fluid in the pressurizer surge line. Note thatthis document indicates that thermal stratification is assumed for all surge flow asthe velocities will always be low. This document also specifically indicates thatthe stratified temperature analysis envelopes high velocity flow and thermalshock.The following paragraphs are excerpted from the Thermal Striping Analysis forthe pressurizer surge line in CEN-387-P, Revision 1. The conclusion is the"effect of thermal striping is negligible and will not affect the fatigue life of thepressurizer surge line".The term "striping" refers to the thermal oscillations that occur at the hot-cold interface.The period of oscillations was chosen to be 1 second and 4 seconds forthe surge line analysis. Test data was measured or was empiricallydetermined to be in the range of 1 second to 10 seconds. For the largetemperature differences and high heat transfer coefficient used in thisanalysis, the period is closer to 1 second than 4 seconds. A longer periodwould yield a lower heat transfer coefficient, and therefore smallerchanges in metal temperatures. However, to be conservative, the sameheat transfer coefficient was used for all cases.The stresses due to each gradient as a function of time were calculatedusing formulas in ASME Code Section III, NB-3653.2. Table 3.5.3-2 liststhe alternating stress calculated for each of the four transients used forevaluating fatigue. As can be seen from this table only one of the fourtransients contributes anything to fatigue. That transient is number four(4) with an alternating stress of 15,780 psi and a number of allowablecycles of 1.42E7.

Waterford 3 Pressurizer Surge Line Temperature Change RatePage 3 of 4Waterford 3 Design Specification 9270-PE-140 is the project specification forreactor coolant pipe and fittings. This document provides a summary of thedesign analysis for surge line temperature transients. It includes text sectionsand 2 tables as they apply to the surge line and surge line nozzle. The tablesaddress temperature differences anticipated as a result of thermal stratification.Table 4.5.15.3.1 lists expected occurrences of temperature differences betweenthe pressurizer and the RCS hot leg and provides the number of expectedoccurrences. Table 4.5.15.3.2 lists expected occurrences of temperaturedifferences between the top and bottom of the surge line piping. Thesetemperatures differences are for the pressurizer surge line piping and not thefluid temperature in the piping. The number of occurrences is the expectednumber for the life of the plant.Entergy PositionThe Entergy position is that pressurizer surge line temperature is not required tobe specifically monitored per procedure to ensure the design limits aremaintained, and that FSAR Section 5.4.3.1 should have been revised in 1993when the Waterford 3 stress and fatigue analyses and design specifications wererevised per NRC Bulletin 88-11 to reflect the results of CEN-387-P, Revision 1.This section of the FSAR has not been revised since the initial FSAR. CR-WF3-2006-0839 was initiated to revise the FSAR. The reasoning for Entergy'sposition is documented in the paragraphs below.The pressurizer surge line temperatures during heatup and cooldown aremaintained by ensuring the heatup and cooldown limits in the RCS andpressurizer are maintained. The RCS limits are located in the TS and thepressurizer limits are located in the TRM. Temperature changes in the surge linecan be greater than 200°F due to thermal stratification and thermal stripping.CEN-387-P, Revision 1 documented that the pressurizer surge line meets Codestress and fatigue requirements for the 40 year design life of the facilityconsidering the phenomenon of thermal stratification and thermal stripping.Analysis in the CEN has indicated that temperature differences of up to 340°Fhave been evaluated for.The data recorded by the temperature element in the surge line has shownperiods of temperature changes greater than 200°F/hr. Thermal stratification isapplicable to all of these recorded temperature changes. These temperaturechanges do not necessarily reflect the average temperature change of the surgeline but reflects a change in local fluid temperature at the temperature element.This recorded temperature changes over time are not the same deltatemperatures listed in the tables in 9270-PE-140.Therefore, the temperature difference in the pressurizer surge line is bounded bythe analysis performed in CEN-387-P, Revision 1 and monitoring pressurizer Waterford 3 Pressurizer Surge Line Temperature Change RatePage 4 of 4surge line temperature per procedure during heatup and cooldown is notnecessary.Additional InformationThe additional information specifically addresses the difference between thesurge line temperature increase seen during Refuel 13 and during the shutdownfor Hurricane Katrina, and the delta temperature values in 9720-PE-140. It alsoaddresses the reason Waterford 3 does not currently monitor surge linetemperature during heatups and cooldowns.There following information is clarification regarding cycles listed in DesignSpecification 9270-PE-140 and the temperatures recorded in PI with thetemperature element located in the surge line. The graphical data recording thesingle surge line temperature element over time for our Refuel 13 outage and theHurricane Katrina outage indicates periods of temperature changes greater than200 degrees within one hour. Thermal stratification is applicable to all of theserecorded temperature changes. Thermal stratification temperature changes wereaddressed by CEN-327-P (NRC accepted response to NRC Bulletin 88-11). Thissingle temperature element does not necessarily reflect the average temperaturechange of the surge line but reflects a change in local fluid temperature at thetemperature element. The recorded temperature changes of a single point overtime is not the same delta temperatures listed in the tables of the W3 Designspecification of RCS Piping and Fitting document (document #9270-PE-140).The table 4.5.15.3.1 lists expected occurrences of temperature differencesbetween two different locations; the pressurizer and the RCS hot leg andprovides the number of expected occurrences. Table 4.5.15.3.2 lists expectedoccurrences of temperature differences between the top and bottom of the surgeline piping. These tables clearly state this information at the end of theirrespective sections. Thus comparing a graph of temperature changes withrespect to time to these tables is not appropriate.The effects on the Pressurizer Surge Line due to thermal stratification andthermal stripping were evaluated in CEN-327-P, Revision 1. This was reviewedby the NRC and in the SER the Staff concluded that the surge line meets ASMECode stress and fatigue requirements for the 40-year design life. Waterford 3currently monitors heatups and cooldowns of the RCS and Pressurizer. Theeffects of these heatups and cooldowns on the pressurizer surge line have beenevaluated in CEN-327-P.

ATTAC HM ENT C White Paper on Effect of Diesel Sump Pump Inoperability on Ultimate H eat Sink Operab ility 11.0 PurposeThis paper provides an answer to the question, what is the original licensingbasis for flood protection of essential equipment in the Dry Cooling Tower Areas?The paper also provides the chronology of regulatory requirements and licensingbases that support the conclusion.2.0 Conclusion Regarding Licensing BasisThe original licensing basis for essential equipment in the Dry Cooling Towerareas is that essential equipment be protected from Standard Project Storm(SPS).The elements of the licensing basis are the following:§The SPS, with all installed sump pumps inoperative, was analyzed as anevent less severe than the probable maximum precipitation.§Provisions are required to be in place for emplacing the portable sump pumpwithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of an SPS event to ensure that the ponding level from SPSdoes not adversely affect essential equipment if installed pumps areinoperative.§The electric pumps are seismically designed but not seismically qualified;therefore they were assumed not to be available following an OBE.§The probability of the occurrence of an SPS and OBE is 3.6E-8 andnegligible.In essence, the original licensing basis required that the portable sump pump beemplaced and started within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the start of an SPS (sump high levelalarm) to ensure that essential equipment in the DCT areas is not flooded.On July 26, 1999, Condition Report CR-WF3-1999-0789 was initiated to identifythat the Dry Cooling Tower sump pump capacities were not sufficient to meet theoriginal licensing basis.A new discharge path for the DCT sump pumps was installed via DCP-3251.The DCP also replaced the 1 portable sump pump that had a capacity of 100gpm with 2 portable sump pumps having a capacity of 300 gpm each. Theinstalled sump pump's capacities were reduced from 325 gpm to 270 gpm due tothe new piping configuration. The revised time frame for starting the portablesump pump to ensure essential equipment is not flooded was re-established as 3hours from the start of an SPS (sump high level alarm). Procedure OP-901-521instructs Operations to operate the DCT Portable Sump Pumps in accordancewith OP-003-024, Sump Pump Operation within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of the sump level alarm.

23.0 ChronologyRegulatory Guide 1.70, Revision 2, September 1975Waterford 3 is committed to Regulatory Guide 1.70, Revision 2, as noted insection 1.8 of FSAR. Neither Regulatory Guide Section 2.4.2.3, "Effects of LocalIntense Precipitation," or Section 2.4.3.1, "Probable Maximum Precipitation(PMP)," have any requirement to consider OBE or SPS concurrently.Regulatory Guide Section 2.4.2.3 states:"Describe the effects of local probable maximum precipitation (see Section2.4.3.1) on adjacent drainage areas and site drainage systems, includingdrainage from the roofs of structures. Summarize the design criteria for sitedrainage facilities and provide analyses that demonstrate the capability of sitedrainage facilities to prevent flooding of safety related facilities resulting fromlocal probable maximum precipitation."The fundamental requirement in the Regulatory Guide is that the applicantensures that safety related equipment is not adversely impacted from maximumprecipitation.Regulatory Guide 1.59, Revision 2, August 1977Waterford 3 is committed to Regulatory Guide 1.59, Revision 2, as noted insection 1.8 of FSAR. Regulatory Guide 1.59, Revision 2, does not have aspecific requirement to consider OBE and SPS concurrently.Two important requirements are discussed in the Regulatory Guide.First, seismically induced floods are associated with land features specific toeach site such as streams, estuaries, dam failures, and landslides. Thisrequirement does not apply to flooding in the DCT sump areas.Second, the Regulatory Guide states that the most severe flood conditions maynot indicate potential threats to safety related systems that might result fromcombination of flood conditions thought to be less severe. The Regulatory Guidestates that reasonable combinations of less severe flood conditions should beconsidered to the extent needed. The Regulatory Guide states that suchcombinations should be evaluated in cases where theprobability of theirexisting at the same time and having significant consequences is at leasecomparable to that associated with the most severe hydro-meteorological orseismically induced flood. We judge that the requirement to consider the SPSoriginates from this requirement. Also, since the probability of a SPS and OBEconcurrent was later established to be negligible, we judge that not consideringthe SPS concurrent with the OBE is in conformance with the Regulatory Guide.

3Standard Review Plan 2.4.3, Revision 2 July 1981Standard Review Plan 2.4.3 does not have a specific requirement to considerOBE and SPS concurrently.Standard Review Plan 2.4.3,Section I, states:Included is a review of the details of site drainage-, including the roofs of safetyrelated structures, resulting from potential PMP probable maximumprecipitation-"Standard Review Plan 2.4.3,Section IV, states:"The local PMF resulting from the estimated local PMP was found not to causeflooding of safety related facilities, since the site drainage system will be capableof functioning adequately during such a storm."The fundamental requirement in the Standard Review Plan is that the applicantensures that safety related equipment is not adversely impacted form maximumprecipitation.NRC Safety Evaluation Report, July 1981The NRC evaluates the effects of a 6-hr duration PMP on the open cooling towerareas and adjacent roofs. The NRC concludes that, assuming one sump pumpin each area is inoperable and that the roof drainage system is clogged withdebris during the PMP, that the ponding could inundate the transformers andMCC's in the cooling tower areas.The Safety Evaluation Report makes no reference to SPS or OBE.FSAR Amendment 25, January 1982FSAR Section 2.4.2.3.4 was initially added to the FSAR; previously it did notexist. This FSAR Section is titled, "Effects of Standard Project Storm (SPS) onCooling Tower Areas".Two important aspects of the licensing basis are established in this FSARSection.First, a probability evaluation is documented establishing that the occurrence ofan SPS and OBE is 3.6 E-8 and negligible.Second, FSAR Section states that the SPS was still analyzed, assuminginoperability of all pumps, in order to determine the time available before levelsare reached that could affect essential equipment in the Cooling Tower Areas.

4Safety Evaluation Report, Supplement 4, October 1982The SER states the following:"An alternative combination which should be considered is an operating basisearthquake (OBE), which fails the sump pumps, coincident with a rainfall eventless than the PMP. This combination is considered appropriate since the pumpsare not seismically qualified1, and thus cannot be shown to be operable followinga seismic event. The staff therefore, requested that the applicant provide ananalysis of the effects of a standard project storm (SPS)2 assuming all fourpumps in the cooling tower areas are inoperable."The SER further states:"-the staff considered a SPS of 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> duration. This event would produce atotal rainfall of about 23 inches and would result in a ponding depth of about 1.9ft in the cooling tower areas assuming that all four pumps are inoperable. Sincethis is higher than the maximum allowable ponding depth of 1.71 feet, theapplicant has proposed to provide a portable pump with a pumping capacity of100 gpm and sufficient head to pump over the cooling tower wall. -a provisionwill be included for emplacing the portable pump within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of a seismicevent if the installed pumps fail."FSAR Amendment 33, September 1983FSAR Amendment 33 revises Section 2.4.2.3.4 to state the following:"The maximum height to which rainwater can rise in this area before essentialequipment is reached is 1.71 ft (see subsection 2.4.2.3.3d). As shown in Table2.4-6c, this level would not be reached for over seven hours into the SPS.""Furthermore, a portable pump is provided, with a pumping capacity of 100 gpmand sufficient head to pump over the cooling tower wall. Provisions are includedfor emplacing the portable pump within six hours of a seismic event if theinstalled pumps fail and heavy rains are expected."Thus, the FSAR Amendment 33 is in agreement with NRC SER Supplement 4 inthat the fundamental requirement is to protect essential equipment in the coolingtower areas in the event of a SPS. The specific requirement in FSARAmendment 33 is that provisions be made for emplacing the portable sumppump within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of a SPS event and that essential equipment be protected,by ensuring that the ponding level does not reach 1.71 ft. The seismic event is avehicle to postulate the installed pumps are not available; however, important tothe licensing basis is the condition that the electric sump pumps will not beavailable and that essential equipment needs to be protected prior to the pondinglevel reaching 1.71 ft.

5NRC Letter dated December 18 1984, Issuance of Five Percent Power License,The NRC issues five percent power license, and Section 2.B.2 of the licenseapproves operation as described in FSAR as supplemented and amendedthrough Amendment 36.NRC Letter dated March 16, 1985, Issuance of 100% Power LicenseThe NRC issues 100 percent power license, and Section 2.B.2 of the licenseapproves operation as described in FSAR as supplemented and amendedthrough Amendment 36.Design Change, July 26, 1999On July 26, 1999, Condition Report CR-WF3-1999-0789 was initiated to identifythat the Dry Cooling Tower sump pump capacities were not sufficient to meet theoriginal licensing basis.A new discharge path for the DCT sump pumps was installed via DCP-3251.The DCP also replaced the 1 portable sump pump that had a capacity of 100gpm with 2 portable pumps having a capacity of 300 gpm each. The installedsump pump's capacities were reduced from 325 gpm to 270 gpm due to the newpiping configuration. The revised time frame for ensuring essential equipment isnot flooded was re-established as 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from the start of SPS (sump high levelalarm). Procedure OP-901-521 instructs Operations to operate the DCT PortableSump Pumps in accordance with OP-003-024, Sump Pump Operation within 3hours of the sump level alarm.