ML120030235

G20110740/EDATS: OEDO-2011-0680 - Redacted Version of 2.206 Petition (TAC ME7332)

ML120030235
Person / Time
Site:	Callaway
Issue date:	11/08/2011
From:	Criscione L S - No Known Affiliation
To:	Polickoski J T, Thadani M C Plant Licensing Branch IV
Thadani, M C, NRR/DORL/LP4, 415-1476
Shared Package
ML1120030071	List: ML120030235
References
G20110740, TAC ME7332, 2.206, OEDO-2011-0680
Download: ML120030235 (61)

Text

CC: "-0: l£e.Os I \..l I (0140 From: Lawrence Criscione IlscriscioneOholmaitcom]

Sent: Tuesdav. November 08, 2011 6:29 PM To: Thadanl, Mohan: Polickosld.

Jam"

Subject:

Please rnpond Attachmentll:

1OCFR2-206 Request on blocking LoTavg FWIS.pdf Mohan/Jim:

Please see the attached 10CFR2.206 petitfon which was copied to you on October 7, 2011. It took me until October 14,2011 to get a certlftcate so ItIat I CDuld electronically submit the attached request. Once 1 received this certifIcete an uploaded the petition, the NRC refused to accept It because some of the pages were scans. As a result, I re-uploaded the petltTon wIth the last 18 pages left: out (I.e. I uploaded the first 40 of 58 pages). The purpose of this email Is to ensure that you are processing my RJLL petllon. The petition should be 58 pages long (a one page cover letter and a 57 page enclosure).

If you have only processed the nrst 40 pages of my petition (i.e. the electronic submission from October l4th) then please add the last 18 pages of the attached document to my petition.

Please note, the attached document Is what was copied to you on October 7, 2011. If you have already processed the Oct. 7 document, then there Is nothing you need to do with this attachment. . Please respond to this email so that I know you have the entire petition.

I am sorry for any confusion this has caused, bUt some of the documents Intluded In lNeI'e only available to me as scans* . ",; Thanks, Larry Lawrence S. Criscione (573) 23D-3959 From: iscrisclonehotmall.com To!

jeanette.oxford@house.mo.gcw; mohan.thadanl@nrc.govi james.pollckosk/@nrc.gov FWIS at Wolf Creek end Callaway Plant 1 23:00:31 -0400 Mr. Borc:hardt:

In late Spring 2010 I read Revision 01 to LER 482-2009-009.

Because of work I spearheaded at callaway Plant, upon reading the Wolf Creek Ucensee Event Report I was concerned about the acxeptability of the Reactor Shutdown procedure at callaway Plant. Iii the summer of 2010 I wrote most of the attached lOCFR2.206 Request, which I did not submit Since I believed that there was some likelihood that either Region IV or CallaWay Plant would ada;Juately respond to the Wolf Creek lER by revising Table 3.3.2-1 fI. the callaway Plant Technical Spe.dflCBtlons In a sImilar manner as Wolf creek had applied to revise the same table In their TechnJeaI SpecHk:atlons.

However, It has now been 1 Va years and I no longer think It Is likely that Region IV or Callaway Plant are going to correct the Issues with the Reactor ShutdoWn procedure and callaway Plent will likely be again ,,!olatlng their Technical Spedncations by bypassing the M/564°F FWIS In MODE 1 as part of their reactor shutdown Plan for their upcoming refl!ellng

""I.ltage.

My request Is: . The US NRC prew.'ftt callaway Plant from bypassing the '-4/564 0 , a:"JfJIS In MODEs 1 thmugh 3 until their Technical Specifications are revised to allow thl$ practJce.

I would like a preliminary evaluation of the steps in callaway Plant's Reactor ShfJtdown pnxl'dure (which allow bypassing of the P-4/S64°F MIS) performed prior to callaway Plant using thet procedure tQ shut down the reactor for their upcomIng reruenng outage. The attached doclJrl'*f1t provides the Justltk:3tlon for thIs request as well as some less 1 ML \ 13j3A20b b S : EJ;:iJ-0 emlnent Issues whICh need to be looked Into. A Senior Reactor Operator (SRO) at callaway wrote a condition report (CARS 200703001)

In March 2007 questioning whether or not the plant's allowed the M/564°FFWIS to be bypassed In MODEs 1 & 2. Because this condition report was written just days prior to Refueling OUtage 15, plant management pressured the Senior Reactor Operator to withdraw his condition report. Because the CDndltlon report was deleted pr10r to it being sent to the plant's SCreenIng Committee, It does not appear In the callawaV Action Request System's database; however, I have a hard copy of the orfglnal report. Enclosed with the attached letter Is some badtground InformatIOn regardIng this Issue. Included In the enclosure are some Internal Celtawav Plant emalls CDncerning the pent's decision to allow bvpasslng the P-4/564°F FWIS In MODes 1 & 2. From the emals It Is apparent that there was much discussion regan:lIng the decision of whether or not the plant's Technical SpecI1'IcatIons allowed this safety function to be bypassed.

The decision by Ameren to not pursue a change to Callaway Plant's Technical SpecifICations was a conscious decision by Its LicenSing Department.

I am requesting that the us NRC review thIS decision and determIne whether or not It Is their opinion that a change to CalIawaV Plant's Tedmlcal Specifications Is needed. As a fonner submarine officer, I assume you are rammar with the following quote from Admiral RIckover:

A major flaw In our system of government, lind even In indUStry, Is the Jatitixle allowed to do less than Is necessary.

Too otten otrldals are willing to accept and IKilJpt to sit.Ulltions they know to be wrong. T1Ie tendency Is to downplay problems instead of actively tfyIng to correct them. I believe that If you choose to look Into this Issue, you will find that the tendency to "downplav problems Instead of actively trying to correct them" was not onlv present withIn Ameren when they consdouslV chose not to revise their Technical SpecIfICations prior to blocking P*4/564°F in MODE 1, but Is also present In our own Region IV where they have allowed Callaway Plant to CDndUd: practices, for which they cited Wolf Creek, because Ameren was able to get a less than adequate safely evaluation past NRR In the mkt-1990s (II safety evaluation which only addressed bypassing the p. 4/546°F FWIS In MODE 3 and was sIlent on MODEs 1 & 2). I've copied Missouri legislator Jeanette OXford 01'1 this email and the attached IOCfR2.206 Request. Representative Oxford has been assisting me wIth getting Safety Culture Issues addressed at callaway Plant, and she IS also concerned with ensuring the ratepayers in the State of Missourl are not unnecessarily burdened with operating expenses stemming from poor stewardship of generating facilitIes (altljough the Wolf Creek Nuclear operating Company Is In Kansas, there may be some Missourians In the Kansas CIty area who fa!llnto WCNOCs rate base since It Is partially owned bV Kansas CIty Power & light). The Citizen's Utility Ratepayer Board In Kansas may be interested in the outcome of this request since this Issue obviously concerns theIr ratepayers.

It Is my opinion that callawav Plant has not been meeting Technical Specification 3.3.2; however, if J am wrong about callaway Plant, then It Is mY opinion that Wolf Creek unnecessarily Incurred expenses responding to the errors of NRC InspectM In 2009 and 2010. These expenses Included protesting a nontlted violation (NCV 05000482/2009004-(4), writing and revising a Ucensee Event Report (lER -482-2009-009, reviSIOns 0 and 1), and processing a Technical SpecIfication amendment (LA 194). VIr. larry lawrence S. CrIsdone (573) 230-3959 Human exper/enal shows thilt people, not organlzattons or mi1fM{JtllTlf!lJ systems, get things done. From: Mohan.1"hadanl@nrc.gov To: IscrlsdoneOhotmail.CDm; James.PoIlCkoskl@nrc.gov Date: The, 6 Sep 2011 08:19:07 -()400 Subjett. RE: P-4/564°F FWIS at Wolt' Creek and callaway Plant Larry: I have not seen an amendment'request.

similar to the subject Wotf Creek Amendment, for Callaway Plant, Unit 1. 2 Mohan From: Lawrence Qlsdone [mallto:lscr1sdoneGIhotmall.com]

sent: friday, September 02, 2011 8:27 PM To: Thadanl, Mohan; PoIick05ki, James SUbJect: P-4/564°F fWIS at Wolf Oeek and callaway Plant Jim/Mohan, Please see the attached document (MLlI0SS0846) (Dll(BJ1lng the P-4/564oF FWIS at wor O'eek. Both Wolf Creek and Gallaway Plant have a ESFAS feature wherein a Feed Water Isolation Signal Is generated under the following aHldltlons:

1. The reactor trip breakers are open (1:'15 read by permlsslve P-4) with P-1 not reset AND 2. Reactcr Coolant Temperature less than 564°F (Lo-Tavg). On April 13, 2010 Wolf Creek Nuclear Operating Company a request (ML101100391) to amend ItS operating license that the P-4/S64°F FWIS was no longer required during MODE On March 30, 2011 we approved Wolt Creek's requested amendment To your knowledge, has callaway Plant submitted iii sfmllar amendment?

That Is, to your knowledge, do the Specifications at callaway Plant allow it to block the P-4/564Of FWlS (function 8.8) during MODE 31 Larry Lawrence S. Criscione (573) 230-3959 3 October 7.2011 1412 Dial Court Springfield.

IL 62704 Bill Borchardt Executive Director of Operations United States Nuclear Regulatory Commission Washington, DC 20555..()001

Dear Mr. Borchardt:

I am submitting the information contained below as a 10CFR2.206 request. The address above is my home address; however, I work in the Washington, DC area and make it home to Illinois infrequently.

Please send all correspondence to me electronically at either my personal email account (LSCriscione@hotmail.G9m) or my work ema! I. If you must send me a hard copy, please send it to me at Mail Stop CSBlC2 A7. The Reactor Shutdown procedure at Callaway Plant (OTG-ZZ"()OOO5) is not compliant with the planl's Technical Specifications.

In 2007. the Reactor Shutdown procedure was revised to allow the operators to shut the plant down by tripping the control rods. Tripping the control rods causes the P-4 permissive to energize.

One of the functions of the P-4 permissive Is to enable a Feedwater Isolation Signal (FWIS) to occur on a "Low Tavg-sIgnal (which occurs at 564°F). As part of the Reactor Shutdown procedure the Instrumentation

& Controls technicians bypass the P-4/564"F FWIS. Callaway Plant's Technical Specifications requIre the P-4 permissive and all "8 associated . functions to be operable (i.e. not bypassed) when the plant's average coolant temperature is above 350°F. By bypassing both trains of the P-4/564"F FWIS. the plant is not in compliance with its Technical Specifications.

Please process this letter and Its enclosure as a 10CFR2.206 request. Very respectfully, £'s.

Lawrence S. Crtsclone, PE (573) 230*3959 Enclosure (1) Cc: Jeanette Matt Oxford, Missouri House of Representatives lOCFR2.106 Request Regarding Blocking orthe P-4lLoTavg Feedwater isolation Signal (FWIS) at Callaway Plant §I. Background WolfCreek is a nllclear reactor plODt near Burlington.

Kansas which is operated by the Wolf Creek Nuclear Operating Company (WCNOC). Callaway Plant is a nuclear reactor plant in Callaway County, Missouri which is operated by Ameren Corporation.

The two nuclear reactor plants are Westinghouse 4-Loop Pressurized Water Reactors and are ofsimilar vintage and design. As part oftheir Engineered Safety Features Actuation System (ESF AS) both plants have a pcnnissive denoted P*4. P-4 is an electrical signal whieh is energized when the reactor trip breakers are open and which pemlits the completion of other electrical signals. One of the several elcctrical signals which arc pennitted by P-4 becoming energized is a Fccdwater Isolation Signal (FWIS) on low average reactor coolaDttemperature (LoTavg).

In 2007 (and possibly to this day) LoTavg at Callaway Plant was electrically set to S64°F. The Reactor Trip Breakers (RTBs) provide power to the reactor's control rods. The opening of the RTBs calise the mechanical eql1ipment holding the control rods OUI of the reactor to energize and thereby cause the control rods to fall into the reactor core. The control rods are made of a neutron absorbing material and their insertion into the reactor core dismpts the nuclear fISsion chain reaction, causing the reactor to shut down. Pressurized Water Reactors (PWRs) in the United States are inherently designed such that they will automatically shut down due to a sharply rising reactor coolant temperature.

The corollary to this design feature, however, is that a sharply lowering reactor coolant temperature can, in conjunction with certain equipment failures, cause II shutdown reactor to inadvertently restart. One Of the "defense-in-deplh" measures designed into WolfCreek and Callaway Plant to prevent a sharply lowering reactor coolant temperature on a. shutdown reactor is the Feedwater Isolation Signal which occurs when overage reaCtor coolant temperature drops below 564°F with the reactor trip breakers open (i.e. the P*4/S64°F FWJS). A Feedwater Isolation Signal causes the plant's normal feedwatcr path to isolote and the plant's auxiliary feed water system to activate.

Although this is II. desirable outcome during many reactor accident scenarios, operating 011 auxiliary feedwater does have its drawbacks: Controlling steam generator levels is more difficult for !he operators while operating on auxiliary fecdwatcr than while operating on normal fcedwatcr Enclosure, page 1 Auxiliary feedwater is typically cooler than nonnal fcedwater and can cause Ihcnnal stresses 10 the feedwaler piping. For the above two reasons, by the mid-l990s both Callaway Plant and Wol r Creek were in the habit or using electrical jumpers to bypass the P-4/564°F FWIS during certain plant evolutions.

Note thaI it is not my opinion that this was a bad practice.

It is my opinion that this was a good practice, however, the acceptability and control orthis activity may not have been adequately evaluated by the utilities and the US Nuclear Regulatory Commission to detennine i(the benefhs oftlle practice (e.g. less thermal stress and better operator control) outweighed the drawbacks (e.g. degraded "derense-in-depth" for certain unlikely combinations of equipment failures).

§1.1. Callaway Plant Operating Lieenst Amendment 126 In 1996. Callaway Plant originated an internal modification package (CMP 96-10 16A) to install bypass switches around the P-4IS64°F FWIS so that. when plant evolutions desired this signal to be bypass. the signal could be bypassed by use of installed switches instead of by installing jumpers. This was a wise modification.

Installing jumpers on engineered safcty features always involves some amount of risk of human error that is not present during the operation of an installed switch. As part or tile implementation of CMP 96-10 l6A. Union Electric (Ameren's predecessor) applied 10 the US NRC ror a change to their Technica.l Speciticarions on August 8, 1997. This application was later supplemented on November 10, 1997 and approved as License Amendment number 126 (LAI26) on Apri123, 1998. The approvallener is located in the Agencywide Documents Access and Management System (ADAMS) as ML021640348.

Enclosure 2 of til is leiter (the safety evaluation for LA 126 conducted by the Office of Nuclear Reaclor Rel:,l11lation) is included below 011 pp. 40-42. .During a highly.cx.pcrienced and 1horough US NRC licensed Senior Reaclor Operator (SRO) at Callaway Plant. Was involved in the rc:view and implementation of LAI26 and CMP 96-1016A.

§l.l. IOCFRSO.S9 Screening of the Callaway Plant Reactor Shutdown Procedure 1 was a US NRC IicellSed Senior Reactor Operator at Callaway Plant when. in January 2005, the Outage approached me about revising the Reactor Shutdowll Procedure (OTG-ZZ-OOOOS) such that it pemlitted manually tripping the reactor as of the nonnal Oleans of conducting 8 reactor shutdown.

Due to my involvement in revising the Plant Cool Down procedure (OTG-ZZ-00006) and assignments for the Steam Genera10r Replacement Outagc (RFl4) it look me nearly a year before I could act upon request. Enclosure, page 2 In early January 20061 originated Request For Resolution (RfR) :200600140 in the Callaway Plant Action Request Sys1em (i.c, Amcrcn's computcl;zed syslem for implementing the legally mandated Problem Idenlification and Resolution process al Callaway Plant). As originally writtcn, RFR 200600140 requested resolution of two questions: Do t:u,.,.ent Jicen,dng documellls require the P-4ILo T"vg FWIS to be operable in MODE I 01'1 (tile originatePI' qf ,his RFR cOllld lIotfin(1 reqlliremem)? Do any Operability Determinations on the P-4ILo FWIS being oWliioble?

On January J 0,2006 onc of the engineering supcrvisors_

requested that I rc-write RFR 200600140 such that it was more general (see' below, p, 39 oflhis enclosure).

I no longer have access to RFR 200600140, but the wording was something to the effcct of"make any licensing changes neeessaJ)'

for revision ofthe Reactor Shutdown procedure", RFR 200600140 WitS then rejected by the RFR Screening Comnlittee as being too general. In April 2006, the issue was re-submitted as RFR 200602749.

Although tbis document was only conccrned with blocking the FWIS permitted by P-4 and NOT with disabling pemlissivc P-4, RFR 200602749 was closed with a Lead Response which merely discusscd thc fact that P-4 is rcquired to be operable in MODEs I and 2 pcr Tablc 3.3.2-J oftbe plan.'s Technical Specifications (see below, p. 30 ortbis enclosure), Note that at the time (and quitc possible sli11) Callaway Plant did not requirc any closure review of Requcsts for Resolution to ensure they were adequately addressed.

On September 5, 2006 I wrote yet a third Request for Resolution on this topic (RFR 200607357) which made the follow requests (see below, p, 29 of this enclosure): either document that Ihere are no regu/olOry requlrement.l*

which prevel7t bypassing the Feed Water Isolalio" Sigl1al cam,etl by P-4 and La Tavg (S.tI6°F) priol' /0 tripping the reaclor.from MODE 1012, or make allY necessary amendments to lic'el7se dOClflll(;!l7ts to al/(nll bypaJ'.I'il1g the P-4ILo Tavg FWIS P'";OI' to trippil1g the reactor/rom MODE / or 2. In a September 7, 2006 cmail exchange regarding "Tripping the Control BCJI7k.t il1 ,00005" (see below, pp. 26*28 ofthis enclosUIe),_ofCallaway Plant's Licensing group wrote, in refercnce to bypassing thc P*4J564°F FWIS in MODEs I and 2. that: _ and his group [Safety Analysis]

dOll't support dOinE:ju.vt a Bases change. In a laler email that same morning._

Enclosure, page 3 I main/ain

[bypassing Ihe P*4/564°F FWiS in MODE 1 or 2] can be utme via a Bases change only; P-4 is operable whenever it receive.'

the required RTB pasition inputs and Ihe permissive is its logic OUlputs to SSPS. The enabled funclion Iha/ relie., (111 P-4. FWIS (JIl low r*avg, ;S 110t required by ESFAS Flinc/ion 5 1101' credited in on,l' occident analysis Ihal I'm aware of _ then responded in an email slating (see below. p. 26 oflhis enclosure):

The Safety An(liy.t;s group is naf opposed 10 Q TS Bases change. Ourpo;nt was Ihal we interpreled the TS such Ilral 0 TS change WQuld be required.

Huwever. we deforrecllo Lic'ensillg os Ihe TS Subjeci Maller E.r.perts. .vii[Licensing]

believes dlJd c'an documelll why a TS change ;s 1101 needed and Ihat on(v a TS Bases clllmge is t,'l.Ifficienl, we have no Laler, in a September 22,2006 email (see below, p. 25 of this enclosure),_ slatemenls which seem to endorse_ view Ihal a TS change was

...I'd recommend a /oolnote be added to Table 3,3.2* / Fllnc,ion 8a describing the circumstances behind our desire 10 block IhisparlicuJar enabled/unction/rom P-4 (i.e., allow illo be blocked during any of the Applicable MODES/or P-4 dflring Q plant shutdown only, to be restored prior to MODE 2 el7llY osee/lding).

[Note thar this is what WolfCreek was required to do in 2010 in order to conlinue to block the P*4/564°F FWIS during MODE 3.] Safety A I/olysls makes a valid painl when they say typical ITS ,,"Ies o/llsage do not allow the Bases to modify Ihe LCO Applicability . .. . llhil1k the besl way to resolve ,,/I of the above Is 10 submit a TS change and gel a very dear thumbs up or down/rom NRC. __position, however, changed by January 9, 2007 when he slated in an email (see below, p. 22 of this enclosure):

In the NRC's Sc!fery Evaluation for LA J 16 dOled 4*2J-98, they spec/flcall.y reviewed and lotlnd acceptable our bypass switch design ol1d OUI" using it to block the FWIS initialed by 'he coi/,ddence of P*4/1ow as 10l1g as its use was limited 10 'he }o/lowing plant cQndjtions:

Enclosure, page 4 with T-avg less II,all or equal to 564°F (YOli call be ill MODE I or 2. bull1ltlSI be ;S.S64°F) just prior 10 openillg Ihe RTBs (which sallsfie., the P*II portioll oftllisfeedwatel' isolaliun sigl1(1l's logic). NRC also wantecll/liS FWTS 1'e!l/cll'ed by diff!(flillg the bypass p,.ior 10 enlering MODE :1 ascending during startup ji'mn 077 outage. As IUl7g as these

(/tY! observed.

"£1 atuendl'l1elll is needed since it '" already beell reviewed alld approved by NRC. We CQuid hal'e bee/l doillg this sillce 4-23-98. By January 10. 2007. I had been Irying unsuccessfully for ovcr a year 10 get the Callaway Plant organization either to state in some type of Qualtty Assurance record that it was alright for Operations to bypass the P-4/S64°F FWIS when manually tripping thc reactor in MODE I or 2. or to apply 10 the US NRC to change Table 3.3.2*1 ofthe plant's Technical Specifications in OJ'der to permit the bypassing of the P-4/564°F FWIS in MODEs I find 2 just prior to manually tripping the reactor * .1t was at this point thut the issue had. aRer nearly two years, come full circle back. to who in 2005 originally requested that the Reactor modified to allow manually tripping the reactor) who, by 2007. was the ]11 a January 10.2007 email to_<scc below, pp. 21*22 of this enclosure)

I St8tOO: I would like 10 see it documel1led 011 a QA record tlUlI Accidel1t

.4.tJa(vs;s alld Licensing

('Ol/CUI' Iltattile P-41S64of FW/S ('on be bYP(/$Sed in MODE I beluw 564QF. In an etTon to meet the above request. it was decided in a January 10-11.2007 email exchange (see below. pp. 18*20 of Ihis enclosure) thai: ...the IOCFRSO.J9 screening e\lo///(/tioll jor-the p1"Oceuure clronge i,y Ihe appr<1priate place ICJ doclimellllhat Ihis change ;3 lVi/hill Callowoy S clIrrem licensing baStS, On February 28, 2007 _ of Licensing prepared the I OCFR50.59 Screening for Revision 00 to OTG-ZZ-ooOOS.

Addendum 01 (the ncw procedure which allowed bypassing the P-4/564°F FWIS just prior to manually tripping the control rods). _ of Safety Analysis reviewed this IOCFRSO.59 Screening on March 1. 2007: In the screening document if is stated (see below, p. 45 of this enclosure):

rlli.f lJoll-CI'iUcal enabled.limction.

FW/S 01/ P-4 coincident willi 10.... ReS T-(/vg. is not a rs required sse If it were a TS reql/ired sse, it wuuld be required 10 be listed os a SlIb-jimction under TS Tuble 3.3.1*1 Function 5.

/lat. FWIS 011 P-4 coincidel11 with Enclosure, page S low RCS T-avg doe:; not meetonyofthefoul' criteriafor rs inclusion in JOCFR50.36 (c)(2)(U).

At tbe lime (i.e. early March 2007), I failed 10 question Ihe above assertions since they supported my goal ofimplementing Ihe new Reactor Shuldown procedure prior to Ihe upcoming refueling outage (RF J 5 in Apri I 2007). §1.3. Callaway Action Request 200703001 On March 29, 2007 I gave a copy ofOTG-ZZ-00005, Addendum 01 to review. That night._was the Field Supervisor.

1 can no longer recal1 why 1 gave. _ a copy of the procedure.

It may have been that his crew was scheduled to shut down the reactor on April 1, 2007 to start Refueling Outage 15. Or it Illay have been that, as an experienced Senior Reactor Operator, 1 wanted his opinion on the new procedure.

_ had a significant concern with OTG-ZZ-OOOOS', Addendum 0 I. In his opinion, Ihis new procedure violaled Technical Specification 3.3.2. Unbeknownst 10 me until thai evening, _ had been involved wilh modification package CMP 96-10 16A and Operating License Amendmenl 126. From_involvement in these aClivities it was his opinion thai when ULNRC-0368I was written the phrase "in a shutdown evolution" meant Ihe rod banks were inserted and the operators were at the pomt of opening the reactor trip breakers (Le. the operators were perfonning the evolution of shutting down the planl and were already in MODE 3), It should be noted that when LA 1 26 was issued, the only way to shut down lhe reactor at Callaway Plant per the nomlal reactor Shllldown procedure was 10 manually drive the control rods into the reactor core. Therefore.

when the P-4IS64°P PWIS was to be bypassed prior to opening the reactor trip breakers, tbe conlrol rods were already fully inserted and the reactor unquestionably shut down and in MODE 3. However, with Ihe release of Addendum 01 of OTG-ZZ-00005 in March 2007 it was now going to be the practice 10 bypass the P*4/S64°F FWJS with the reactor in MODE I, with Ihe nuclear fission reaction still erilical, and wilh the control rods either fully or ncor fully withdrawn.

It was_view that Liec:nsc Amendment 126 did not apply to bypassing the fWIS under Ihe new conditions.

I was in no position to either refute _ concerns or to act upon them and change the plan for the upcoming reactor shutdown (which was to occur on tbe evening of April J. 2007). M.y advice to_was to'document his concerns in the Callaway Action Request System. which he did as CAR 200703001.

To ensure that CAR 200703001 got the level of, concern which it warranted, I intbnned all interested parties about il in an email in the early morning hours of March 30,2007 (see below, p. 16 ofthis enclosure).

Enclosure, page 6

_ concerns wcre not well recei ved (sec below, pp. 13-15 of this enclosure).

For my part, after spending Ihe better part of three years optimizing the Reactor Shutdown and Plant Cool Down procedures, I was accllsed by Opcrations of nttempting to sabotage the reftlcling outage. By the nine o'clock hour on March 30, 2007 I was tired of arguing with my superiors, and] was willing to accept and adapt to a situation I knew to be wrong and to focus on downplaying problems instead of octively trying to correct them. In an email to_ (sec bclow, p. 13 of this enclosure)

I olTercd to answer CAR 200703001 with wording similar to the lOCFR50.59 Screening of Revision 00 to OTG-ZZ-00005, Addendum OJ. Similarly broken of his desire to continue fighting, _ dclctcd CAR 20070300 I (sec below, p. J 3 of this enclosure).

Since CAR 20070300 I had never been sent to the daily Screening Committee mceting, there is no record of it at Callaway Plant, but somewhcre in a box in my vacant home in Jefferson City. Missouri 1 still have a hardcopy of it. It should be noted Ihal beating dowu_and his inconvenient safety concem was financially the correct course of nction for Ameren to take under the US NRC's Renctor Oversight Process (ROP). Under the Reactor Oversight Process. the safety significance of an issue determines the level of regulatory scrutiny and punishment which will be applied to it. As _ would assuredly acknowledge.

his concerns wcrc of low safety signific.1Dcc due to the fact that the P-4/564°F FWlS would be bypassed for a very short time and the likelibood of thc requisitc equipment failures occurring during that short time window is very slight.

• _ conceTllS were abolll doing Ihe right thing frol11 a regu latory standpoint.

However. in order to do the right thing, Callaway Plant would have Dceded to change its refueling outage schedule at the eleventh hour. At hundreds ofthollsands of dollars an hour, even sligh I changes to the outage schedule would have amouDtcd to significant expenses.

Bonuses at Callaway Plant are heavily dependent on meeting refueling outage schedules and costs. so there was monetarily much at risk by Callaway Plant's upper management.

Conversely, under the Reactor Ovcrsight Process. low risk* significant violations of plant licensing commitments typically amount to 110 monctary penaltics.

A \though "willful violations" of licensing cOI11D1itments are treated seriously, these are nearly impossible to objectively prove. The odds this issue would ever be brought b"torc the US NRC wcre slight (it is only by chance that r came across it again wh(.'n reviewing revision 01 to WolfCreek's LER 482.2009.009).

Now that it has been brougbl before the NRC, it is unlikely this IOCFR2.206 will be acted upon. .Even jfacted upon, it is still very unlikely that there will be anything greater thnn II nOll cited violation issued to Callaway Plant. By removing the subjective judgmellts of the regiollaJ leadership from the regulatory formula. the Reactor Oversight Process has made it nearly impossible to punish plants who are "gaming the regulations" until a risk significant incideni occurs; as a result. we are reactive in our rcgulation instead of proactive.

Enclosure, page 7

§1.4. Integrated Inspection Report at WolfCreek Nudear Operating Company On August 22, 2009 US NRC inspectors at Wolf Creek observed Instrumentation and Conlrols (J&.C) technicians install jumper wires to bypass the P-41564 D F FWJS while the plant was in MODE 3. In Integrated Inspection Report 05000482/2009004 (see below, pp. 47-51 of til is enclosure) the inspectors stated: The inspectors and the NRR tec:llflil.'a/

speciflcatiOil branch.loulld this 10 he comrary to the Updated Safe(y Report, Ihe tet'hnical spedjicalions.

the technit'al specification base.y. and the NRC softly evnlUQlions supportit78 Ihe let'Imien/

specifications.

Based on the above observation, the inspectors issued a noncited violation (NCV "05000482/2009004-04) for "Failure 10 Implement Engineered Safety Features Actuation System Technical Specifications Results in a Missed Mode Change.II §1.5. Licensee Event Report 482*2009-009 As a result ofNCV 050oo482/2009Q04.Q4.

WolfCreek Nuclear Operating Company submitted Licensee Event Report 482-2009-009 in December 2009 and submitlcd a revision (LER 2009-009-01) 011 March 22, 2010. IteIllI.C.S of the TMI Action Plan requires Ihallicensees shall: ... prepare procedures to assure thai opemtlng Illforlnation pertinelll to plallt safet}' origillating both within and outside tile uttlity organizatiOlI is contimtal(v supplied 10 .opera/ors alld o/he/' personnel and t... incorporated inlo training and retraining programs.

It is unclear how Callaway Plant meets the above requirement.

One would think that a Licensee Event Report from Callaway Plant's "sisler plant" concerning an entry into Technical Specification 3 .0.3 would be "operating illfonnation pertinent 10 plant sofeIY". It is unclear why, fOllowic"S the release of LER 482-2009-009.

no one at Callaway Plant questioned the plant's practice of bypassing the P-4/564°P FWIS not only in MODE 3 but also in MODEs 1 and 2. f1.6. WolfCreek Nuclear Operating Company Operating License Amendment On April 13, 20 I 0 Wolf Creek Nuclear Operating Company applied to the US NRC for an amendment to their Technical Specifications to allow bypassing the P-4/564°P FWIS during MODE 3 (see below, pp. 54-56 of this enclosure).

On March 30. 20)1 the US NRC approved Enclosure, page 8 I amcndment 194 to WolfCreek's Technical Specitications.

Amendment 194 added II footnote to TS Table 3.3.2-1 similar to what_proposed for Callaway Plant in 8n email on September 22.2006 (see below, p. 25 offhis enclosure).

11.7. Willrul Violations From January 2006 through February 2007 there was a fair amount of discussion Dnd dcbate at Callaway Plant regarding whether or not an amendment to the plant's Technical Specificalions was required to allow bypassing the P*4IS64"F FWIS in MODEs I and 2 just prior to manua'lly tripping the reactor. In the end, it is my opinion that_(of Licensing) and_ (of Safety Analysis) reached the wrong conclusion in their IOCFRSO.S9 screening of Revision 00 to OTG-Zz..OOOO5, Addendum 01. However. it is also my opinion that neither one oflhese men were gUilty of either willful violation or incompetence; their error was nothing more tban an honest mistake. made while trying to understand complex licensing (e.g. TS Table 3.3.2-1 clearly requires P-4 to be operable, sillce Although I do not believe there was willful violation committed during the I OCFRSO.S9 review process, 1 do believe that either willful violations or instances of gross incompetence have occurred 8t Can away Plant since February 2007 with regard to bypassing the P41564°F FWTS. Specifically:

• II is my opinion that the two bulleted items above dcmonstrale tbot Callaway Plant has either 8 grossly incompetent management team or has a culture which willfully chooses (0 ignore inconvenient licensing issues. Enclosure, page 9 There are some at the NRC who will allempt to prevent the above items from being transparently addressed through the IOCFR2.206 process. If the concerns above are remitted to another process (e.g. the anti.transparent Allegation Process) I believe that it is inappropriate ror inspectors from Region IV to be assigned to investigate these concerns.

Region IV has consistently validated Callaway Plant's Problem Identification and Resolution (PI&R) proccss and Safety Culture as satisfactory and therefore has a vested interest in downplaying the problems which exist in Ameren's corporate culture. It is my opinion that any allegation investigation should be perfomled by NRC Headquarters or another regional office. Requests per Please treat the requests in the sections below per IOCFR2.206.

§2.1. Immediate Action Request r request that the US Nuclear Regulatory Conlmission take the following action in a timely manner to ensure that Callaway Plant does not inadvertently enter Technical Specification 3.0.3 by violating Technical Spccitication 3.3.2: Prohibit Callaway Planl from bypassing the P*4/564°F Feedwater Isolation SigDal until the practice has been reviewed by the US NRC and detennined to be in compliance with Technical Specification 3.3.2. §2.2. Requests for Evaluation 1 request the appropriate staff at the US NRC Office of Nuclcar Reactor Regulation perfoml the following actions in order to evaluate my concerns: The US NRC review the Green NODcited Violation ofTccbnical Specification 3.0.3 from August 22. 2009 contained 011 poges 3, 4. 19,:20 and 21 of the enclosure to Integrated Inspection Report OS00048212009004 (ML093140803) and detemline if a similar violation applics to Callaway Plant (tbese palles are provided below as pp. 47-5 J). The US NRC review LER 482*2009-009-01 (Mt;I00890421) and detennine ira similar LER is required by Callaway Plant to report any violations ofTS 3.0.3 as a result of their bypassing of the P-41564°F FWIS d\lring MODEs I or 2. Tbe US NRC review the Green Nondted Violation of Technical Specification 3.0.3 from August 22, 2009 contained on page 10 of the enclosure to Integrated Inspection Report 0500048212009005 (ML I 00430713) and detemline if a simi lar violation applies to Callaway Plant (this page is provided below as p. S2). Enclosure, page 10 The US NRC review Amendment 12610 Callaway Plant's operating license (ML021640348) and determine if they believe there is anything in this license amendment which allows the utility to block the P*4/564°F Feedwater Isolation Signal during MODEs I or 2 just prior to shutting down the reactor by manually tripping the control rods. Please commcnt specifically on paragraph 2.4 of Enclosurc 2 (found Of! page 24 ofML021640348 and provided below on p. 41). The US NRC revicw Callaway l>>lant's Reactor Shutdown Procedure (OTO-ZZ-OOOOS) incl uding the IOCFRSO.S9 screening paperwork for OTG-ZZ-OOOO5.

Addendum 0 I. Revision 00 (included below as pp. 43-46) which was (212812007)'

and_(3/1/2007), and detennine if the US NRC agrees with the utility'S answer to screening question S. Please comment specifically on the statement:

This n(m-crirical enabled/unction.

FWIS on P-4 coincident with low ReS T-avg. is 1101 a TS reqllired SSC. /.fil were a TS required sse, il wollld be required 10 be lisled Cl:f a SUh.jUllctioll under TS Table 3.3.2-/ Function 5. 11 is nOI. FW/S 01' P-4 coilU.:i(/ent with 10111 ReS T -tng does nol meel ally of the fOUl-criteria for TS illc/usioll in IOCFR50.36(c)(1}(ii).

§2.3. Requests for Actfon Based on the determinations made by the NRC staft'for items Iin section §2.2, please take action per either section §2.3.1 or §2,3.2 as appropriate.

§2.3.1. Actions regarding Callaway Plant If the NRC detemlines that by-passing the P-4/564"'F Feedwater lsolation Signal dluing MODEs I ond 2 at Callaway Plant is a violation of the plant's Technical Specifications, then I request that the actions below be taken. Note that, due to Region IV's past involvement with this issue. 1 suggest the actions below be handled by inspcctors andlor investigators from either headquarters or a different regional office: Issue a violation to Callaway Plant for every ioadvenent entry into TS 3.0.3 which lUIS occurred as a result of by-passing the P*41564°F FWJS during MODEs I and 2. Detcmline what deficiencies in Callaway Plant's lOCFRSO.S9 Screening Process allowed a procedure change to be made which violated the plant's Technical Specifications. Review the email trail included in this enclosure (pp. '3-39) and investigate what failed in the Safety Culture at Callaway Plant that caused the concerns raised in Callaway Action Request 200703001 to go unaddressed.

10. Determine irthere are Illly deficiencies in Callaway Plant's ability to process and learn from industry Operating Experience (OpE) in light of the fact that apparently no aClion was taken by Ameren in response to LER 482-2009-009 revisions 00 and 0 I. Enclosure, page 11

11. Detennine if there are any deficiencies in Callaway Plant's ability to work with industry peers in light of the fact that their "sister plant" submitted a License Amendment (Wolf Creek's LA 194) which. although somewhat applicable to Callaway Plant, was not addressed by Callaway Plant. J 2. Determine why the US NRC did not look at Callaway Plant's practices regarding blocking the P*4/564°F FWIS once it was noted in August 2009 that WolfCreek's practices (Callaway's "sister planl") did not meet her Technical Specifications.

§2.J.2. Actionl regarding Wolf Creek Nuclear Operating Company If the NRC determines that by-passing the P-4IS64°P feedwater Isolation Signal during MODEs I and 2 at Callaway Plant is not a violation of the plant's Technical Specifications, then I request that the actions below be taken. Note that, due to Region IV's past involvement with this issue, I suggest the actions below be handled by inspectors and/or investigators from either headquarters or a different regional office: 13. Review NCV 05000482/2009004*04 in light of the Callaway Plant detemlination and, if appropriate.

withdraw this nonciled violation (see below, p. 51 of this enclosure).

14. Review the NCV fromlR 0500048212009005 regarding LER 482*2009*009-00 in light of the Callaway Plant detenninalion and, ifappropriale.

withdraw this noncited violation (see below, p. 52 of this enclosure).

15. Review Licensee Event Reports 482*2009*009-00 and 482*2009*009*0 I in light of the Callaway Plant determination and, if appropriate, have Wolf Creek either withdraw the LERs or subnlit a new revision which correctly discusses how her Technical Specifications were not met. 16. If appropriate, reimburse Wolf Creek Nuclear Operating Company for any expenses unnecessarily incurred in submitting and processing LER 482*2009*009 revisions 00 & o I and Anumdmcnt 194 to the plant's Technical Specifications so that the nuclear ratc payers of the State of Kansas are not unfairly burdened by errors made by the staff of Ihe US Nuclear Regulatory Commission.

§3. Supportlna Documents The remainder of this enclosure is supporting documentation:

Alneren emails ..............................

• ...........................................

pp. NRR Safety Evaluation for LA 126..................................................

IOCFR50.59 Screening for OTG-ZZ*OOOO5.

Addendum 01, Revision 00....... pp. Select pages from ML093140803, MLl00430713, MLI01100391, ML1105S0846.

and MLII1661877

...........................

pp. Enclosure, page 12 From: CriscIOne, Larry S. Issue resolved, From: Criscione.

Lany Sent: March 30, 2007 9:58 Dave. ss switches of the P-4/564"F FWIS an d 1. Licensing has documented in the 10CFR50.59 Screening of OTG-ZZ*OOO05.

Addendum 01 that the P*4/564"F FWIS may be used in MODe 1 or 2 2, _ email below indicates licensing's basis for Ihe 10CFRSO.59 Screening addressed In the CAR system I can add It to the Refuel 15 critique.

II was obvious today that not appreciate being questioned at the "eleventh hour". I apologize for this. I did my best to were identified and resolved ealiier in the cycle but unfortunately did not succeed. Vir. Larry -I found OL 126. See atlachmen!.

Section 2.4 used the same wording as ULNR(}()3681: "during startup and shutdown evolutions with Tavg:!: 564-F just prior to opening the reactor trip breaker *.* Although this wording does not speclflcaHy state the function can be used In MODE 1 or 2. the wording neither specifically slates the I'vnction can only be used In MODE 3 or below. I am growing tired of this issue too. but don't let that discourage you if you suil have 8 concern, With our current level of regulatory scrutiny on design basis. we need to make sure we have all our commitments addressed.

My opinion Is the wording of "during startup and shutdown evolutions with Tavg $

the reactor lrip breakers' can be Interpreted to apply in MOCE 1 or 2 as well as In MOCE 3. From_email below. Licensing is obviously defending this. Please look at OL 126 (it's short) and let me know if you have any further concems. If you want this issue documented in a CARS response vice an email, I cen answer the CARS. Thanks. Larry :z Enclosure , page l3 LA 126 only requires that T-avg be less than or equal to 564: use of the block switch Isn't tied to any MODE. Restoration Is tied to MODE 2 entry. Neither our ULNRC submittals nor LA126 were contingent on the Insertion of oontrollshutdown banks, or lack thereof. Rod position was never discussed.

NRC wanted the us. of this block switch to be limited to startups and shutdowns at or near the time the FWIS coincidence is made UD for this function.

If you don't believe me. rescind OTG*ZZ-00005 Addendum 1. I grow tired of these never-encling email daisy chains. From: criSCIone, Larry S. Sent: March 2007 7:41 PM To: -I re*reviewed the 50.59 Screening last night when. raifl'ed his I have never had any concern with the TIS bases for P-4. I have always agreed with the 50.59 Screening with regard characterization of P-4I564-F.

Additionally, I have never had any title of email _ was involved with the initial modification which Installed the P-4/564°F bypass switches. is familiar with our commitments in ULNRC-03681.

He explained to me last night that when ULNRC*03681 was written. the statement of "during startup and shutdown evolutions with Tavg < 564'F just prior to opening the reactor trip breakers" meant we had Inserted rods. we were In MODE 3 and we were now 10 open lhe reaclor tri Instead ofULNRC-03681.---..oL Amendment 126 in the 50.59 Screen when stating "the planl can be in MODE 1 or 2. bul OL Amendment 126 supercedes ULNRC*03681 by two years. I do not know what the exact wording in OL Amendment 126 is. If this amendment specifically stales (or reasonably ImpHes.l..:!e planl can be in MODE 1 or 2" when bypassing the P-4f564'F FWIS. then it is my belle' (I do nol wiSh to speak for_lhat.

concerns are resolved.

_ CAR was written on the mld*watch when licensing was unavailable for consultation.

Tonight. I will attempt to locate and review Ot Amendment 126. I believe. will void his CARlfOL Amendment 126 specifica"y references MODE 1 or 2. Vir. Larry Criscione J Enclosure, page 14 Call me at home about this tonight. I Thanks, I call this the *Crlsclone Tralr In honor of the 20+ page ematl that accompanied this issue until it was answered by my 50.59 screen for OTG-ZZ-00005 Addendum 1 Revision O. See attached; asked and answered.

CAR I 200703001 should be volded.1 Now maybe I can get back to my hobby as PWROG Licensing Committee chairman and Westinghouse TSTF rep, since I have to work on those activities on my days off and alter Enclosure, page 15 From: CrIscione, Larry Sent: Friday, March 30, 2007 4:50 To: Dl CAl. CARS In ULNRC*03681, AmerenUE commilled to use the P41l0 Tavg FWIS bypass only under the following conditions:

(1) RCS Tavg < 564-F (2) I The plant is in a shutdown evolutlonland at the point before the Rx trip breakers are opened (3) The P41Lo Tavg FWIS will be restored prior to reaching Mode 2 during a startup evolution.

The concern in CARS 200703001 Is that we may not be legalistically meeting our commitment (2) In ULNRC-03681 and literal compliance with TIS 3.3.2 Function 8.a. . It is not clear what "in a shutdown evolution" entails. In his answer CA0139 comment (see attachment to Ihe r assumes "in a shutdown evolution" to mean the are Intending to shutdown the plant. The concern is that when ULNRC-Q3681 was written, "in a shutdown evolution" meant the rod banks were inserted and the operators were al the pOint or opening the reactor trip breakers (i.e. the operators were performing the evolution of shutling dowri the plant and were In MODE 3). Refer to RFR HOl5A. If this Is how the NRC will Interpret "In a shutdown evolution" then they may question whether OTG-ZZ-00005, Addendum 01 meets our commitments to them. It should be noted that CARS 200703001 does not contend the performance of OTG-ZZ-00005, Addendum 01 is unsafe. The contention is that It may not meet the NRC's interprelation of our UlNRC*0368l commitments.

Based on our current regulatory level of scrutiny, these concerns must be addressed prior to conducting the shutdown.

Please ensure CARS 200703001 is screened appropriately.

Thank you, Larry Criscione 11:56 AM -.ILlst wonted to let yOll know I scnt out a mcssage to Ihe design supervisors to let me know iff can the change from I\n RFR 10 lin ACNO. As soon as I hear from them. T will be happy to change the CAR Iype. 2 Enclosure, page 16 6:48AM CAL CARS Screening 200607357 Then pleo.se make the change. Thanks, There is no need to send this buck to screening as iL is nol being considered for cl1angc from or 10 <In adverse condition.

r COil make this CARS lype chnnge with Ihe concurrence orlbe appropriate dcsign engineering supervisor.

Thanks.

Sent: fr ay, anulry 2007 6:06 AM I To: DL CAL CARS

Subject:

Re-Screen CAR Fellow Screeners.

Please rc-scrcen CAR 200607357 from an RFR to on Action Notice. I believe that there Is sufficient justification in the string of c-malls below. Lorry has provided the following which should be included In the Screenin9 Notu os the justification There are no regulatory requirements which need to be addressed to allow using the P-4/564°F FWIS bypass feature in MODE 1 below 564-F. therefore a Request for Resolution Is NOT require(j.

The concurrence from Licensing and Accident Analysis of bypassing the P-4/S64"F Feed Water Isolation Signal In MODE 1 will be documenled In the 10CFRSO.59 Screening for OTG-ZZ.()(l()C)5.

RFR 200607357 will remain open as an AcUon Notice to lrack this Issue to ensure Ills spedlk:ally documented In the 10CFR50.59 Screening.

Thcmks,-3 Enclosure, page l7

---

Ed gs: Please have RFR 200607357 re-screened as an Action Notice. The basis is: There are no regulatory requirements which need to be a!l:\ressed 10 allow using the P-4f564-F FWIS bypass feature in MODE 1 below 564"F. therefore a Request for Resolution 18 NOT required.

The concurrence from Licensing end Accident Analysis of bypassing the P-4/564*F Feed Water Isolation Signal in MODE 1 will be documented In the 'OCFRSO.S9 Screening ror OTG-ZZ-00005.

RFR 200607357 wUI remain open as In Action Notice to track this issue to ensure it is specifically documented in the 10CFR50.59 Screening.

If the above reason is not acceptable, then you should be able 10 find something else from the emaillrail below. Larry Criscione The RFR has been reassigned back 10 Larry. I'm available to provide whatever support is needed to flll out the CA251OJ2511/2512 forms for the procedure change to OTG-ZZ-00005.

IIIbelOW and plealle 'e....i9n lhe RFR . ..... Sent from my BlackBerry WIreless Handheld From: O1sclone, Larry Sent: 11,20076:28 --il re-screened as an Action Notice and That wOrks for me. Please hay ssign RFR 200607357 to me. I will use it to suggestion.

Based on a dlsclJsslon I flad I think there Is e good chlflce we will not pursue thii(01"G.ZZ-oOOO5 revision) in cycle 16 -although I int""nlifln Is to pursue thiS now. Because of 'I Enclosure.

page 19 the uncertainty, I would like 10 capture lome 01 the resolutions from the email trail below In an active Action Notice so information is available in the event a delay is Larry From: Sent: W , nuary 10, 20076:00 PM To: Criscione, Larry S. cc* I'sub.tect:

fW: 200607357

&. O'1'G-ZZ-00005 Larry, Here is one opinion. How would Ihls work for you? " OTG-ZZ-OOOOS Based on_response, I believe that the 10CFRSO.59 screening evaluation for the procedure change Is the approprlat';"'pliCe to document lhat this change is within Callaway's current licensIng bases. The purpose of 10CFR50.59 Is to determine if prior NRC approval is required.

It appears that the NRC has appoved thia for Callaway beck In 1998* *From: Olscione, Larry S.

January 10, 2007 1:26 I just spoke With_conceming some research she did earlier In the year on tripping the control rods Into the core -during 8 normal She Informed me she did some benchmarking for you and It is her opinion this 18 something we should pursue. . I spoke with*****ear6er In the day and his understanding Is there arB stili Chemistry concerns regarding this issue. I first got irlVolved with this issue two years ago when your predecessor

• • • • • • asked me 10 pursue It In January 2005. I have gotten nowhere In two years. 5 Enclosure, page 19 Now that we halle the -go-ahead" from licensing.

Operations stltl does not want to Incorporate the option of shutting by tripping the control rods into OTG*ZZ*OOOO5 due to Chemistry Operations Is currently upgrading the OTGs. Ir Ihe upgrade of OTG*ZZ'{)OO05 does not include the option of down by tripping the control rods, it is unlikely this option will be available to us in What Chemistry concerns exist regarding tripping the control rods during the shutdown?

Halle the posed in the email OTG 200601261129 (several pages below) been Larry larry. John Will add this option to OTG*ZZ-00005.

Thanks for pursuing an RFR to get it on paper. Rick From: Criscione, larry S. sent: January 10, 20079:48 AM -requested I send you a CARS on OTG*ZZ*OOOOS.

See Action Notice 200700222

• .--does not wish to pursue this currently due to outstanding industry concerns (mainly with Chemistry).

_ this wi!h youand_. . If we decide not to pursue these changes in the current upgrade of OTG-ZZ-OOOO5 but wish to pursue them during cycle 16, I would suggest aSSigning Action to Chemistry (and any other concerned departments) to perform the necessary ellaluations EARLY in cycle 16 (due dates within 60 days of the end of RF1S) to ensure suffiCient time is allaUable In cycle 1Sfor Operations to perform an upgrade. larry Criscione Please see that we halle the proper amount of technical rigor documented in our response to address the issue that larry describes below. I do not know what type of dOCt..lnlenl is most appropriate, I suspect that an RFR is, built we halle some other licensing or safety analysis way of doing this that Is searchable and retrievable, let me know, Thanks, 6 Enclosure, page 20 Sent: 10,20079:28 AM To: Criscione.

Larry S.

Subject:

RE: RFR 200607357,.

OTG-ZZ-00005 Come talk to me From; QlscIone, Larry S. In January 2005 (cycle 14), we aiscussed shutting down the Reactor by tripping the Control Banks InlO the core. Altha time. we agreed we would not make this change for RF14 due to the amount of plant resources needed 10 address all the other issues associated with RF14. As long as OTG-ZZ-00005 was being upgraded,l wanted to provide the option, regardless of if we would be taking advantage of II In RF14. This was acceptable to you. OTG-ZZ.oo005 was not upgraded during cycle 14. In January 2006 (cycle 15).1 discussea shutting !he down the Reactor by lripplng the Control Banks Into the core with _.RFR 200600140 was written to very specificalfy address blocking the P-41564"F FWIS. At the reque&t RFR 200600140 Was re-written to be more "general-and Ihen was rejecled as not being specific enough to answer. The issue was re-submitted as RFR 200602749 in April 2006 and was rejected by Accident Analysis based on a misunderstlllnding of what was being requested.

In September 2006, the Issue was re-submltted as RFR 200607357.

A decision had been made by Operations in August that there was not enough time remaining in the cycle to allow all the necessary Issues (major reviSion to an OTG, possible license amendment.

license Operator training concems) aSSOCiated with shutting down the Reactor by tripping the Control Banks Into the core to be addressed prior to RF1S. The current slatus of RFR 200607357 is as follows: licensing has stated (see email trail below) no regulatory concems exist with blocking the P-41564 D F FWIS in MODE 1 as long as T8Vg is below 564°F. Licensing has stated they are not qualified to answer RFRs Accident Analysis has stated that since no concerns exist with blocking the P-41564 D F FWIS prior to tripping the control banks in MODE 1, then RFR 200607357 should be rejected.

I wOUld like 10 see It dOCumented on a QA record that Accident Analysis and licensing concur thet the P-4/564 DF can be bypassed in MODE 1 below 564*F. RFR 200601351 is the moat convenient way to accomplish this. Since licensing amendments are necessary.

it is my opinion RFR 200607357 can be re*screened as an Actlon Nolice. I documentation somewhere other than an email In my Inbox -an Action Notice Is sufficient for me If that is easier for your groups. If your groups will not provide this documentation, then I cen settle for \heir signatures On a ClOSS Discipline Review of the procedure

-however I cannot speak for the other Operations personnel who will be involved with writing. reviewing and validating OTG*ZZ-00005.

Documenting your apPfoval in an Action Notice or RFR will greatly simplify the process of changing the procedure.

Based on my experience with writing. validating and training the crews on OTG*ZZ*OOOO6 In cycle 14, I we move forward with the necessary changes to support shulling down the reactor by tripping the control core during the upgrade ot OTG-ZZ*OOOOS.

It ill my understanding Is being reralned to upgrade Enclosure t page 21 the remaining OTGs, NOTE that just because we have the option 0' shulling down the reactor by tripping the control banks does not mean we need to do it. of OTG*ZZ-OOOO5Is completed before RF15, it need not change any RF15 plans. 'h'IiiOses the risk that the necessary changes to OTG*ZZ-oOOO5 WIll not be made . If contract runs out prior to the OTG*ZZ-OOOO5 changes being incorporated.

it is will be ava a to properly revise, validate and lrain on OTG*ZZ* 00005 prior to RF16. In cycle 14. we trained on a major revision to OTG-ZZ*OOOO6 while it was slill being written and validated.

We do not want to repeat that again. Vir. larry Criscione Gentlemen, In Ihe NRC's Evaluation for LA126 dated 4-23-98, they specifically reviewed and fouoo acceptable our bypass switch design and our using It to block the FWIS iniUaled by the coincidence of P-4l1ow T-avg as long as its use was limited to the following plant conditiOns: with T-avg Jess than or equal 10 564°F (you can be In MODE 1 or 2. but must be ! 564°F) jusl prior to opening lhe RTBs (whidl &SUsfies the P-4 portion of Ihis feedwater Isolation signal's logic). NRC also wanted this FWIS restored by defeating the bypass prior to entering MODE 2 ascending during startup from an outage. As long as these Umitatlons al'9 observed.

no amendment Is needed since It's already been revieWed and approved by NRC. We could have been doing this since 4-23-98. From: CrISCione, Larry S. . 09, 2007 7:27 AM -RFR 200607357 r6qU8ats one of two Ihlngs be done: It is requested the Lead for this RFR perform one of the following: Document in the Lead Response to this RFR that there are no regulatory requirements which prevent bypassing the Feed Water Isolation Signal caused by P-4 and Lo Tavg (S46°F) prior to Iripping the reactor from MODE 1 or 2. Make any necessary amendments to license documents to allow bypassing the: P4/Lo Tavg FWIS prior 10 tripping the reactor from MODE I or 2. With reference to RFR 200601357, does Bert believe ililm 1 can be done. J recognize.

does not believe HE can do item 1, but does hit believe there Is justificetion for it. Enclosure, page 22 If there Is Justification for Item '. then RFR 200601351 can be answered by Accident Analysis.

U we must go the route of item 2. then RFR 200607357 must be owned by Licensing.

Larry Criscione lrlV"lnn..

larry S.

200602749 and 200600140

• said he could provide a response to Larry's email question, but Is not qualified to answer RFRs. If that Is required.

we will need to get an Engineering person to do It. -r Email Sen

• nuary , 2007 5:55 PM From-=:. Sub] :: 7357,200602749 and .is* our plan? and 200600140 about this on Monday.I plan to tall< -frDnl'___

20073:32 PM To:

Subject:

I 602749 and 200600140 Are we on target to address this issue? 9 Enclosure, page 23 Should we have a meeting to ensure we all agree wilh our directioti and What is needed to support each Let me -from: Crlsdone, Larry S. RFR 200607357 Is still In Evaluate.

It was due on 10/1912006.

The email trail below tells the saga. We have been leeking a resolullon since January 9, 2006. From email DTG 200609221821.

without any license document changes we can bypass the P-4/564°f FWIS in MODE 1 once we are below 564°F. This should correspond 10 around 20% reactor power which would be acceplable tome. I believe RFR 200607357 should be dlsposilloned as follows: Per item 1 of the RFR. LicensIng should affirm we can bypass the P*4/564 OF FWIS In MODE 1 once T-avg is below 564°F. 2. Per Item 2 of the RFR. Licensing should affirm no licensing amendments Sre necessary.

F rom.email OTG 200609071016 "the time line the! needs to be used on this issue" is one tMt will support Implementation iri"J!refueI16.

From my experience whh OTG-ZZ.QOOO6 changes in cycle 14. we need an answer prior to RF15 If we are to make major changes to OTG-ZZ-OOOO5 by RF16. I do not We need consensus between Operations and licensing on when RFR 200607357 mu'Si"be"a,;sweriif.

My opinion is a due date of no later than March 31. How do you want to proceed on this Larry 25, 2006 6:21 AM -Rjck supplied me the responses from the other plants earlier. II does not appear any of them have the capability to block the P-4/Lo Tevg FWIS (this was not part of the original design at Callaway and apparently none of the other plants made the same mod). We do not need 10 block P-4/lo Tavg FWIS to trip the plant. Since we have the capability.

it just works better that way; Easier on plant eqUipment

-no AUlC Feed Water transient thermally shocking the Steam Generators Easier on piant operations

-no Aux Feed Water transient challenging SG Water Level Control Being an outlier is not necessarily bad. By deHniti0iAif you are In the top tier, you are an outJler. It is my understanding we spent the money onlmod package CMP 96*101 Ito achieve the two bullets above when opening the trip breakers in MOOE 3. let's take advantage of the fact that at one time these plants were willing to spend money to Improve their' Enclosure, page 24 designs. The mod package was Instal/ed yean; ago: I just want authorization to use it advantageoUsly.

now that our operational needs have.changed (opening trip breakers in MOCe 1 vice MOCe 3). Larry From '= ...... I r 22, 2006 6:21 PM Larry. To get the full background behind Commitment (COMN) 43387. you need to read our docketed letter (ULNRc.tI3681 dated 11/10197.

Alt. 1. page 5 of 8) and the NRC Safety Evaluation for LA126 (see attached, Seclfon 2.4. page 2). We told NRC we would use the FWIS bypass switch dUMg slartup and shutdown evolutions with T*avg 564 41 F just prior to opening the RTBs and we would restore the FWIS prior to MOCE 2 (during the ensuing startup).

In order to abide by that NRC SEI we would not use the bypass switches until T*avg 564 41 F which probably doesn't gat you very far into MODE 1. We could do that now with no licensing document or COMN changes.

• NRC has seen TSTF-444.

T via a North Anna amendment.

Dominion only used the part of TSTF-444-T that supported their desire to do a P*4 TADaT every 18 months (not after /lIVery trip breaker cycle), but Nodh Anna submitted the entire TSTF to NRC In one of their RAI responses.

TSTF-444-T needs to be revised generically through the Tech Spec Task Force to discuss blocking specifIC functions enabled by P-4. We could not block turbine trip on P-4 since it's modeled in several accident analyses, but most of the other Poo4 enabled functions could be blocked since they're not credited or modeled In aoodent analyses.

North Anna adopted the Bases portIOn of TSTF-444*T where only MODES 1 and 2 were listed 8S being required for turblna trip and FWIS on p. 4110w T-avg.

• Although TSTF-444*T uses the Bases to modify the LCO Applicability of the functions enabled by P-4, TSTFoo444*

T does not say any of those enabled functions can be made Inoperable throughout the P*4 Applicability of MODES 1*3.

• Safety Analysis makes a valid point when they say typicsllTS rules of usage do not allow the Bases to modify the .LCO Applicability.

• Thera is a program (RITSTF 8a) Just getting underway thet would ITy to relocate all RTS and ESFAS permlssives out of the Tech Specs, but thai program has a 2*year timaline to gel NRC approval.

11 Enclosure, page 2S In OTG-ZZ-OOOOS In the attached email you state Therefore.

I see the desire to lise this switch in MODES 1 and 2 as triggering the commitment change process of APA*ZZ*00540 Step 9. The desired change would Impact COMN 43387. The text for COMN 43387 states; The administrative controls governing slartups will ensure the P*4n..o*Tavg bypass switch is manually defeated and U'e isolation function is restored prior to entering Mode 2. I do not wish to change anything in the above commitment.

During the Heal Up/Start Up II is still Important that Operations procadures "ensure the P-4/Lo-Tavg bypass swilch is manually defeated and the jsolation function is restored prior to entering Mode 2: What I wish to do Is to bypass the P-4/Lo-Tavg FWIS in MODE 1 or MODe 2 Cf\Jrlng shotdown.

This does not violate COMN 43387 as What Is my next step? Is there 8 different commitment to which I need 10 request a Sen.t: pDtember 07, 2006 2:30 PM U I TO: OiscJone, Larry S. Sub ec:t: : pplng the COntrol Banks In OTG-ZZ-oOOOS There you go Larry. begin wilh APA-ZZ-00540 Larry. The Safely Analysis group is not opposed to a TS eases change. Our point was that we interpreted the T$ such that a TS change would be required.

However. we deferred 10 Licensing as Ihe TS Subject Malter Experts. If.believes and can document why a TS change Is nol needed and !hat only a TS Bases change is sufficient.

we have nO"'i'Uoog objections.

... 12 Enclosure, page 26

• • From: Criscione, Larry S. Sent: To: In order to meet a Refuel 16 limeline, all license amendments must be implemented prior to Issuing OTG*ZZ*OOO05 for training In cycle 16. What is an acceptable timeline for evaluating (i.e. deciding a course of action, assigning Actions for evaluating and implementing license changes, and taking the RFR 10 InProcess)

RFR 2006073571 My e)(perience is that If we do not sel earlier milestones, we will be pressing people at the last minute. to meet later milestones.

I was unaware a final decision for RF15 had been made prior to requesting an October 6 due date for RFR 200607357.

Regardless, I think 30 days is sufficient time to evaluate this issue and set out a plan for RF16. From: E Sent: Thursday, september 07 To:

This will not happen for Refuel 15. The Operations Refuel meeting August 21 decided to pursue this for Refuel 16. That Is the time line that needs to be used on this issue.-From: Criscione, Larry S * ..., Below is the email trail on the RFR we discussed this morning. During our discussion I stated the plan was to remove P-4 from TIS via the WOG. This is a stand alone issue regardless of RFR 200607357.

It is related to the RFR in thai once il is done, there will be no question as to whether the P-4/564°F FWIS can be bypassed al power. From. email (two below) this morning, It appears RFR200607357 can be resolved with a TIS Bases change. this could occur prior to RF15 if we wanl to drive II. This would allow us to Irip the reactor after taking the lurbine off*llne without any feea water tranSient.

Is thiS something we want to pursue for the RF15 ahutaown?

Larry From: Criscione, Larry 5, 07, 20068:59 AM RE: The red highlighting in _ response was added by me. I think it is the crux of the malter. Do you disagree with it or know of any accident analYSIS which relies on a post trip FWIS below 546°F? Can you support a TIS Bases change? Larry 13 Enclosure.

page 27 from: __ sent:

To: rri.,..i".,.. Ffom: Qlsdone, Larry S. 8:44AM I was lold in January that _ was doing the APA-ZZ-00540 paperwork.

Is this is something I need to do, please let me know. 07, 2006 8:38 AM l"rI!lrton,,,,.

Larry S. the Control Banks In OTG-ZZ-OOOOS Yes. I sent. all of the plant survey responses that I received on blocking this FWlS func::lion.

All plants thet trip the reactor In MOm deal WIth the FWIS, reset It and normal feed all qUlcl\. flO th"y can. None blcxok the FWIS on P-4l1ow avg or have the block switch design installed.

Analysis Is the that pltlvlded the less-then-adequate RFR respoflse, not I don'llhink this RFR should be assigned to Licensing.

The answer to the first question is 'Yes: We made this commitment to NRC in a docketed letter (UlNRCsare part of our licensing basis) during the 011189 review and no one Ihat wants to change that commitment ha$ fOllowed APA*ZZ-00540 yet. APA-ZZ procedures apply to everyone.

The 2nd question deals with ODs that might credit this funcllon and we don't write ODs or even know about all the ODs thai may have been written. 14 Enclosure, page 29 S@Dh!!mihf!r 06, 2006 7:34 AM the Control Banks In OTG*ZZ-oOOOS

_1& this a continuation or the exchanges you have had with ****? Let's discuss this and where you think it Is gOing. r OS, 2006 2:07 PM the Control Banks In OTG*Zl*OOOOS the Control Banks in OTG-ZZ-OOOOS Please send RFR 200607357 to and This Is a licensing Issue and. as demonstrated by the response to RFR 200602749, E We h'IVe been seeking an IInswer to this Issue since January 9. RFR 200602749 was Originally submitted ps RFR 200600140 and then reworded at the request 0' The new. worded RFR 200600140 was \hen rejected.

RFR 200602749 was then submitted with 200600140.

RFR 200602749 was dOSed without addressing the question.

From the attached and _ appear to be prepared to answer this issue. The History List on RFR 200602749 AcciTen'i'Analysis and not I.icensing.

OTG..zZ-OOOO5 must be upgraded by October 31. 2006 to meet a Refueling Milestone.

We cannol afford to avoid answering this issue any longer. Thirty dafSMnot be much Urns to evaluate \hIs Issue. but In Operations we have been awaiting an answer for nine months. If is not the right individual for this RFR, please ensure It Is appropriately assigned at Screening tomorrow to an iviclual who can provide an evaluation by October 6. 2006. Note we do not need licensing amendments by October 6; we only need an evaluation as to what license amendments (if any) musl be initiated and their feasibility.

Thank you for your assistance.

Larry CriSCiOne IS EnClosure, page 29

moor 05, 2006 11:26 AM To: Criscione, Larry S.

Subject:

RE: Tripping the Control Banks iii OTG-ZZ*OOOOS Please write 8 new RFR -From: Criscione, Larry S * .. Do you still want the option to trip the control banks In OTG-ZZ-OOOO5?

RFR 200602749.

Evaluate ByPBsslng P-4/564°F FWIS during MODE 1 (15%) or 2 Manual Trip, was noIanswered correctly.

The two questions posed still remain unanswered; Do current Ycensing documents require the P-4flo Tavg FWIS to be operable in MODE 1 or 2 (the Originator of this RFR couid not find any requirement)? Do any Operability Determinations rely on the P-4/Lo Tavg FWIS being available?

The Lead Response to the RF R answered the RFR by evaluating If the P-4 function Is required.

W. do not wish to dl,able the P-4 function; we only want to bypass the FWlS which is caused by the simultaneous inputs or P-4 and temperatura les. than or equal to 564*F. The last word I got was we are no longer considering trtpplng the control banks due.to cnemlslry concerns, so I chosen not to pursue an appropriate closure of RFR Please let me know If I need to continue to pursue Larry Sent: 22, 2006 5:45 PM To: Criscione, L.arrV S.

Subject:

FW: survey on feedwater isolation from P-4 and low T-avg FYI .. see plant survey responses. Enclosure, page 30 22, 2006 11:40 AM Yes. 22, 2006 11 :39 AM on fee<lwater lsolatlon

'rom P-,. and low T-avg _ all that equal 2 years from today?-22, 2006 11:24 AM -Any TIS change we pursue would need to be coordinated with a revision to TSTF444* T to reftect the NRC*approved North Anna amendment and this TSTF revision must first be processed through the PWROG. E-bar eUmination took 5 'leers to work Its way through the PWROG process from Initial conception to final TSTF-490 submittal to NRC (2001* 2005) and our 01..#1257 amendment submittal (5-9-06).

This shouldn't take as long as E-bar took. but I would guess 12 months to get the PWROG (which now Includes CE end B&W) to agree on a revised TSTF-444-T. I will woJ1l. on a Callaway-specific amendment in patanel with the PWROG. but NRC would likely take a year to review a submiltal if it came in today and we are a ways off from making an amendment submittal.

22, 2006 9:21 AM on feedwater Isolation from P-4 and low T-8vg you pursue the TIS change? Based on your response It will take some time and probably will not happen in time for RF 15. 00 you have a estimaled completion date?* l7 Enclosure, page 31 we start the 81 Inputs 'rom the breaker position switches and the permissive outputs are proViding the required logic Inputs the SSPS 10 enable the various end device functions.

I said that I don't believe there should be any requirement fOr the 3.3.2 Bases to discuss non-credlted.

end device functions.

Our TS Bases {pages B 3.3.2-36 and -37) could be modified to make this type of statement and to qualiry the teld on the enabled functions to either their Enclosure, page 32 *avg III,scussed this with

• and It appears based upon our previous discussions that would be the best since \hIs issue deals with the Tech. SpecsJTech.

Spec. Bases and applicablliy of thlt 'U, ".L1l11'" what is credited in the Safety Analysis.

We are Willing to say, as discussed in the CARSlRFR 200602149, this function is not credited for accident mitigation in the Safely Analysis.

Safely Analysis can support this by performing a CDR review of the change, as desired. is your group willing to support a TIS Bases change? The drop dead data for RF 15 is 11117106.* IS, 2006 10:26 AM feedwi!ler

!I!!""and law T-avg We discussed TSTF-444-T and P-4 during the PWROG T8 Working Group meeting on 611 and 812. I don't recommend adopting TSTF-444-T. as wrlIIen, besides NRC would not approve a license amendment by the data It would be needed foe-ne>et March's outage. Ha_r. all ofWorklng Group meeting rdt"nelees

_ in ag....ment (at I__t I h **rd no spoken disagreement) that a T8 Bases change could be pursued under 50.59 to accomplish what you want. Our MODe 3 restriction on using the P-4/1ow T-avg block sWitch is but thIS commitment is not captured In the TS or Basel. I recommend that you check what Is required to revise a commilment under Westinghouse to develop an argument that none of the RTS or ESFAS perm issives satlsfy any of the 4 criteria under 50.36 for TS inclusion and Justify their relocalior\

from the T8 -however that program has a 2*year timeline* Here's whall've received back so far on the survey responses.

Moal people live with the FWIS when they trip at power. I also heard thai North Anna received NRC approval for a portion of TSTF0444*T. They cited TSTF-444-T to justify an 1 a. month TADOT frequency for P-4 rather than the STS NUREG-1431 TADOT frequency of checking P-4 after every RTB cycle. Their P-4 Applicability remains Modes 1*3 In T8 Table 3.3.2-1. but they have varying appllcablrdy requirements In the Bases for the functions enabled by P-4 (their NRC-approved Bases say FWIS on P-4 and low T-avg 18 required only In Modes 1 and 2). II Here is the BraidwOOdlBi.ro.nire.s.eo.ns.e.

If you need additlonallnformatlon you can

• He provided the response or give me a call 1. Does your plant design Include FWIS on P-4 coincident with low RCS T-avg? Yes 2. If yes, do you allow normal feedwater to be isolated when RCS T-avg reaches its setpoint (564"F at Callaway)?

If so, does your Operations staff share the above concems expressed at Callaway?

No. We generally trip the reactor from 8 to 10 % power after the turbine has been tripped. There is usually a 5 minute delay between turbine trlp and reactor trip which allows us to stabilize a bit. FW temp is about 275 or so at the time of the reactor trip.. We expeditiously reset the FW Isolation and establish low flow feed. We have a startup FW pump so AF is not used. 3. If you block FWIS on P-4 with low T-avg during shutdown, how would reconcile that against the TS Table 3.3.2-' requirement that P-4 (and the TS Bases inference that enabled functions from P-4) be Operable In Modes 1*3'1 We do not block P-4. '152006

Subject:

RE; survey on feedwaler isolation from P-4 and low T*avg Responses:

01: Yes, same setpolnlS as at Callaway.

02: Yes, Vagtle verifies proper FWlln the controlling procedure (Vogtle would not enter EOP E-O). Currenlly we have not, to my knowledge.

experienced any significant increase in dose rates. . Enclosure, page 33 03: No. we do nol blt:.l<* FWI. One thing I noticed In the benChmarking data Is the Rx power at whidt VEGP would initiate a manuallrip for the shutdown Is stated as *5%-, Thllillow.

W. would initiate the !rip around 18%*20% Rx power. 20063:08 PM 'edwater isolation rrom P-4 and low T-avg-Could you take

• lOok It***questiOns below? L -Answers Inserted belOw. Please answer the follOWing three questions: Does your plant design include FWIS on P-4 COInclent with low RCS T .avg? VB 2. If yes, do you allow normal faedwatar to be isolated when ReS T -avg reaches Ita aetpoint (564'F al Callaway)?

Ves. If so, does your OperatiOns staff share the above concerns ".xpressed at Callaway?

NO. CPSES has procedurallzed tha ldionl required to conllolth.

resulting sKOndary Iystem tl'Wllient (i **** mlnlnize the potentlallor

';'1... of AFW. etc.). For mora Info re: procedures cont.ct If you blook FWIS on P-4 wilt! low T-eYfi during shutdown, how would reoonc:llalhat against the T5 Table 3.3.2-1 raqulremenllhat P-4 (and the TS BaseslnflK8nce lhat enabled funcliol1s from P-4) be Operable In Modes 1-3? CPSES de81gn doe8 not InClUda block capability for FWtS .

• 1ft responses below. Seems Ihls condition has been discussed by engineering reeenily and may be the SUbject of future dlsaJsslons regarding the acceptability of tripping and creating FWI. Thanks, I '7 BncloBure, page 34 Please answer the following three quest/ons: ,

• Does your plant design include FWIS on P-4. colncient with low RCS T -avg? Yes If yes, do you allow normal feedwater to be isolated when ReS T..avg reaches Its setpoint (564"F at Callaway)?

If so. does your Operations staff share the above concerns expressed at Callaway? Not identified by operations at this time. However, aystem engineering has noted a higher than usual maintenance need for FW relief valves and the pressure seen during these trips could be the cause. If you block FWIS on P-4 with low T-avg during shutdown, how would reconeile that against the TS Table 3.3.2-1 requirement that P-4 (and the TS Bases inference that enabled functions from P-4) be Operable in Modes 1*37 NlA ,rom*.Sent: ust 14, 20062:11 PM SUD ec : survey on feedwater J5Clatlon from P-4 and low r-lIIg .you received any feed back?-Callaway Plant would like your help responding to a shOrt sUIVey. I brought up an Issue at lest week's PVYROO Licensing Subcommittee I TS Working Group meeting regarding our desire to manually trip the reactor during MODE 11n our March 2007 refueling outage. The following Information provtdes some background on What our Operallons staff would like to do: Callaway would like to Incorporate a planned 'hal d Shutdown" which means a manual reactor trip after the generator breakel'$

are open and the turbine Is secured. The reason behind this Initiative Is 10 save approximately 3-4 hours of critical path time and align Callaway with the industry excellent plants in the area of refuel outage duration.

The concem with the feedwater isolation signal (FWIS) that occurs on the coincidence of P-4 and low T-avg (564°F al Callaway)

Is the disruption of feedwater from the normal feedw8ter system which would cause an overpresSlJre situation due to feed heating. resulting in relief valves lifting. trip Ihe main feed pumps. and the potential need to use AFW for Inventory control. Enclosure.

page 35 The lollowingare the iniLial Reactor Power -10 to 20 Main Generalor Breakers Turbine tripped. Condenser Steam Dump System armed in Steam Pressure MOde maintaining RCS temperature at approximately 560 RCS pressure at 2235 At last week's meeting It became clear that no plant had adopted TSTF-444*

T which requires P-4 to be Operable Modes 1-4, but has differing Applicability requirements In the Bases for enabled functions (FWIS on P-4 and low required in Modes 1 and 2 per TSTF-444-T).

Based on benchmarking data given to me (see attached files).

Byron, COok. McGuire. Vogtle. Farley (last outage), Wolf Creek, Comanche Peak. Sequoyah, Watts Bar. Salem. Valley. Calvert Cliffs, Glnna, and Palo Verde trip the reactor in Mode 1 (varying between 6% and 25o/D RTP) refueling outage Please answer the following three Does your plant design include FWI$ on P*4 c:oincient with low RCS T*avg? If yes, do you allow normal feedwater to be Isolated ...men RCS T*avg reaches its setpelnt (564"F at Callaway)?

If so, does your Operations staff share the above concems expressed at Callaway? If you block FWI$ on P-4 with low T -avg during shutdown, how would reconcile that against the T8 Table 3,3.2*1 requirement that P-4 (and the T5 Bases Inference that enabled functions from P-4) be Operable In Modes 1*3? Thanks.

Sent: 25,2006 9:47 PM TO: ____lone, Larry Suti : RF "e not requesting to block P.... Our request is to disable one of the many of P-4,-From: criscione, Larry S. Sent: 25, 20067:27 AM *Pleaee add the following to RFR 200600140:

Identify and Initiate required FSAR and Tech Spec changes to allow blocking the P-41564°F Feed Water Isolation I Signal In MODE 1 with the turbine amine. . See RFR Larry Enclosure, page 36 i sent: 26, 2006 12:25 PM TO:--,LarS. Su' * : . pass switch use In MODES land C(luld you give. more informl\liOIl of bow the Shlll<!<.lwn would O!;cur'1 Aren't we still !loil18 to conduct II plant shllldown 10 Ihe swap-over o'i"ihe feed reg vnlvell and 81'110 (10/. rower before we lri,,', Undrrstnnd (J()O()!I 10 penni! opening the Reactor Trip brenkers in 1 after removlI'Ig main gencmttol' from service fram YOllr I have _looking into the commitment change procC'ss thut has adVised me Rbout. ' . II was not clear 10 me what you were asking for in the request belo\y. It is my uuderstanding thllt Outages is looking into the possibility of tripping the plant from approltimately 30% power. I wOllld like to provide some justification for not doing this so that questions can bc nsked when other plants thaI have experience in tripping their plant during ShlllO\)wn.

J know STP has recently several refuelings with very large particulate releases Ihal have caused hiuller than predict.:<i dose rates. One reason for this particuhlte rcleusc may be duc to running all 4 .RCPs (high now rales may be causing the crud release).

The shutdown datil tor Co-58, Co-60. Cr-SI nkkeland iron daw should bc calleetc*d for thc cntire shutdown cleanup (both particulate and soluble dala). This should include datil from before the trip and shutdown.

The plont's history of dose rate and contamination trends should also be reviewed.

From a chemistry cleanup ond dose Prosl)Cctive, the best type of slmtdown is a smooth one (with no ubmpt challges, trips. water hammers. rod drop tests, elc. that could cause a large pmtieutale enid release... Operating history shows that plants j)crfonning rod drop tests durillg shutdown have experienced large particulate crud relcases early during thc shutdown (with pm1iculates dropping out in Ollt of core 3reas causing high dose rates). These same plants hoye experienced high activity particulates during steam generator tube inspeclions causing high dose lind conmmintlliCln i:i!lllcs.

Since Callnway will be completing the baseline gcn\}ralor inspections during RFI5, it would not be II good refhel to have high dose nltes. nor do we wnollO deposiL high flClivity crud in our new steam generatal'S during the tirst shutdown following replacemcnl.

Chemistry has ulready esti 1l)31cd the peak Co-58 release for RFI 5 Dnd it is c:-(pected to be the largcst release for Callaway due fO the SGR (increased release rates for the first rew cycles of operation 11l1d 45%. increase in tubing surf;tce area). However, Ihis release should be soluble and easy 10 deanUI), Actually by injecting zinc during the cycle, we e"pect our pcnk Co-58 to be reduccd by a facfor of :!-3. The RFJ 5 cstimntcd cleanup time required prior to stopping the Reps is estimated to be 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br />. For Refuel 16 Dnd beyond (after the oxide layer is fonned) this peak should be reduced COllsiderably with almost no cleanup time required (just a lew hours), Recent information presented by EPR I inditatcs Ihal it i3 no 10llger critical to cool d\1Wn slowly or to SillY in acid-reducing conditions for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during shutdown so 1here may be other ways 10 I't:,duce the time to ollload 23 Enclosure.

page 37

--1'\1('1. I realize that short rr;fueling are Ihe beSllliillg tor CalillwllY (I'm all ror <20 days) but we need 10 be careful and consider all the consequences.

1 and 2 *In !'e,ponse to your question on allowing the use of Ihe reedvlaler isolation bypass switch in MODE I or MODE 2, I uncovered the following.

We cllleaorized the bypass switcb design change u an Unreviewed Safety Question in OL#1189 (ULNRC-Ol628 dated 818197) under the ok! 50.59 rule lind limited its use to prior to MODE 2 enlry in ULNRC-0368I dated 11110197.

NRC's Safety Evalualion for Amendmenl 126 dated (4123198 (boltom of page 2) also IimilC'd the use of the byplIss switch 10 "prior to enterina I MODE 2." However. nOlning was added 10 the Tech Spet!s, the 'j'ech Spec Bases. or fSAR on this MODE 2 limitation and currenl *Bllses page 8 3.3.2-)7 has no such reslriction.

Therefore.

I sec Ihe desire 10 usc this switch in MODES I and 2 liS lria;gering Inc commitment change procl'lSS of APA*ZZ*Q0S40 Slep 9. The desired <:bange would impact COMN 4]387. Jam nol involved with the C.omnlitmenl Tracking Process, nor have I ever been, bUI reading through Ihe pro<:cduTe J would IIdvise lhe fDllowins:

1. Pt\rson requesting Ihe enung(l fllls out fornUi CA IS71lUld CAll58 per step 9 of APA-ZZ-00540.

2. forms should be sent of2/1/06, Commitment TraCking transfer; from Mart Reidml'yer's old RRA group 10**** Most of the CAlJS8 form questions be answmd "No" based on the discussion in the OUt 189submitt.1l (ULNRC* 03628 discusses how FWIS on 1"-4 wilh eoillcilientlow T.av. is nOl credited in IIny accident analyses);

however. question 4 on the form needs 10 be answcred "ysince the <:ommihnCnl is in an NRC SE. J think thj, change will be allowed, but must be described in the "nexi commitment update report", whatever that is. Maybe MARean provide more bo.clreround, but Oall or Pat will be the ones actually I'C'sponsible for this report. I have nO further knowledge of the commitment cbange bullhe above sllould lei the process sloned. From:__ Sent: w;r"ne;aay,'january 11, 2006 8:22 To: Criscione, Lany

Subject:

RE: RFR Larry. ' Based on this e-mail. I have taken this CAR to Initiate for you. Enclosure, page 38 Thank) ****-orIgin., Message-_** Ftom: Crlsdol'le, Lany S. 11, Z0068:14AM

'10m,-=-ria Sent: ary 11, 2006 7:35 AM I To: CrIScIone, Larry S. SUbject: RE: RFR 200600140 i wiii dLtcuIMd writeup. --eng111l1 1'rDIIt: Lany s. ---volunteered to answer AFR 200600140 yestenfIIY atnce he WB.lnvolved with the odgInaI Mod Package.-rr=prwllmlnary answer soon. I Intend to work an INis with ReacIor Operata' and tralnfng department reaourcea In !he Ilmulator In February.

I do not want to w"fe lIlelr IIma.down

• wrong pllh. I did nollJlve a due dale on the AFA Iince I do not reed a formal wlftten answer unUlearly next year. However, I need someone to look at the Issue Ills month to lei me know definitively If Itie P-4/S64*F FWIS can be bypassed In Mode 1. Thanks. LaITY sat: lBnuatY 10, 2006 10:59 AM Enclosure, page 39

, UNITED NUCLEAR REGULATORY WAIHINGTON, D C SAFETY EVALUATION BY THE OFFICE Of NUCLEAR REACTOR RI!LAIEQ TO AMEE!JOMENT NO 128 IO FACILITY OPERATIE!JG LICENSE NQ Npf.3Q UNION ELECTRIC kQMpANY . CALLAWAY PlANT. UNIU OQ'I5ET NO 50=483 1 0 INTRODUCTION

• By letter dated August 8,1997. as supplemented by letter dated November 10, 1997. the Union EleClI'lC Company (UE) requested changes to the TeChnical Speaficallons (Appendllf A to FaCIlity Operating License No NPF-30) far the Callaway Planl, Unit 1 The proposed changes would reVIse the Technical Specifications (TS) to change feedwater Isolation engineered safety features aduatlon system (ESFAS) functions In IS Tables 3 3--3. 33-4. and 43*2 The November 10, 1997. supplementalleUer proVIded additional clanfyJng information and did not change the staffs anginal no slgtuficant hazards determlnatlOl'l that was published In the Federal Register on December 11. 1997 (62 FR 66144) 20 TECHNICAL Speclfl,ATIQN CHANGES AE!JD eYAL.UATIQE!J 2 1 Actuation LogiC Applicability The applrcable modes for Functional Ul'1Its 5 a 1); Automallc actuation LogiC and Actuation Relays (SSPS). and 5 a 2). automatic Aetwtlon logIC and Actuation Relays (MSFIS), In Tables 3 3-3 and 4 3*2 would be reVISed to add MODE 3 thiS change IS proposed because the automatic actuatIOn logIC for closure of the ITIaIn feedwater lSOIabon valves (MFIVs) mU$t be available an MODE 3 to establIsh a pressure boundarY, preventing dIVersion of auxdlary feeclwater (AFW) flow, thereby ensunng delIVery of AFW flow to at least two Intact steam generators under aCCIdent conditions As a result of thl, change In applicability, the end pamt of the action statements Will be changed to hot shutdown ThIS IS more reltncWe and IS consIStent WIth the appllcabtllty of other TS related to decay heat removal by the auxdl8ry feedwater (AFW) system This change IS acceptable

• Enclosure, page 40

" *. 22 New Steam Generator Level Low-Low Functional Umt A new FunctIOnal Unrt 5 d, Steam Generator (SG) Water Level Low*Low (for feedwaler IsolatIOn only), would be added to Tables 33-3. 3 3-4. and 4 3-2 This change IS more restncbYe The main feedwater Isolation valve (MFIV) IsolatIOn on SG water level low. low Isolation was added to the plant desJgn to address 8 concem that AFW flow could be fed back through the MFW system Instead of to the SGs under caMln break condilions ThiS ISolatIOn SIgnal IS credited In the analyses for the loss of non-emergency AC power, loss of normal feedwater, and feeawater system pIpe break events ThIS ISOlation Signal was not InCluded In the original TS, which were based on the Westmghouse Standard Technical SpeCllicatlons (STS). because neither the STS at the lime nor the current STS Include thiS Isolation Signal While thiS IsolatIOn signal had not prevIOusly been Included In the TS. the IlC8I1$1e stated that they have always performed surveillances on thiS Isolation Signal conSIstent With other automatic acluaben logiC and aduaben relays applicable In MODES 1-3 ThIS enaflge IS acceptable 2 3 Tnp Time Delay Applicability The applicable MODES In Table 3 3-3 for auxIliary feedwater (AFW) SG Water Level Low-Low Funcllonat Untls 6 d 1) c). Start Motor Dnven Pumps Vessel Delta T (Power-1, Power-2), and

• 6 d 2) C)I Start Turblne-Df/ven Pump Vessel Delta T (Power-1.

Power-2), would be reVIsed to delele MODE 3 FunctIOnal Unit 6 d 3) In Table 4 3*2 would also be reVISed to delete MODE 3 ThIs functIon IS used to change the tnp time delays depending on power level At reactor* thermal power less than or equal to 10 percent, the malOmum tnp tll11e delay IS enabled, and the maximum top tame delay should always be eno;bJed In MOOE 3 This change IS acceptable 24 Feedwater Isolabon on P-4ILow Tavg The Bases for Funchonal UM 11 b. Reactor Tnp P-4. tn Table 3 3*3 would be reVised to add iii note allOWIng the feedwater Isolabon functlon on P-4 (reactor top and bypass breakers open) COIMCldent With low Tavg (Tavg s 564 0 F) to be blocked The reason for the change IS to decrease unnecessal)'

cycling of the MFIVs and AFW system whIch adversely Impacts startup and shutdown evolutions Thts feedwaler IsolatIon functIon proVIdes baCkup protection for exceSSive cooldown events and IS not credited In any FSAR analyses The hcense8 has proposed to Install a bypass SWitch to block thiS Signal dunng srartup and shutdown evolullons WlthTavg s 664°F JUst pnor to opent"9 the reactOr trIP breakers The feedwater Isolallon functIon would be restered by manually defeabng the bypass pnor to entering MODE 2 ThiS change IS acceptable 2 5 Conclyslon The staff has reViewed the licensee's proposed T8 changes to revise the feedwa!er IsolatIOn ESFAS functIOns Based on the review. thestsff concludes that the proposed TS changes are acceptable

• Enclosure, page 41

, ,'. '. 30 $lATE CONSULTATION In accordance With 'he CommissIOn's regulatIOns.

the MIssourI Stale officIal was notified of the proposed Issuance of the amendment The State ofllaal had no comments " 0 ENVIRONMENTAl CQNS!DERATION The amendment changes a reqUirement With respect to the installation or use of a faCility component located Within the re5Crlcted area as defined.n 10 CFR Part 20 and changes surveillance requJlements The NRC staff has determined that the amendment Involves no SIgnificant Increase In the amounts, and no Significant change In the types. of any effluents that may be released offslte. and that there IS no SIgnificant .ncrease If1 endl\ltdual or cumulative occupational radIation exposure The CommlS$\on has previously Issued a proposed findll'lg that the amendment ,nvolves no SignIfICant hatel'ds conSlderalJon.

and there has been no publIC comment on such finding (62 FR 66144) Accorchngly.

the amendment meets the ellOlbl/lty entena for categoncal elCCluslon set forth In 10 CFR 51 22(c}(9) Pursuant to 10 CFR 51 22(b} no enVironmental Impact statement or envIronmental assessment need be prepared .n connection With the Issuance of the amendment 5 0

• The CommiSSion has conduded.

based on the conSIderatIOnS discussed above. that (1) there IS reasonable assurance that the and safety of the publiC wdl not be endangered by operetton In the proposed manner, (2) such activilles Will be conducted In compliance WIth the COmmiSSion'S regulations, and (3) the Issuam;. of the amendment Will not be mun.calto the common defense and security or to the health and safety of the publIC F>rmclpal Contributor A Cubbage Aprl1 23, 1998 Date

• Enclosure, page 42 50.59 SCREEN AdiYitylDocument Number. OTO*ZZ-OOOOS Addendllin 1 ReYislon

__ in MODE 2. Continued Applicability Olher applicable processes identitied during the applicability determination:

No other procCSIICS (If a CA2S10 was NOT tiled, Indicate that no otiter processes or document the buis ror not perfOrming II CA2510). List the documents (FSAR. Technical Specifications.

Technical Specification Bases. and other documenls) reviewed, including section numbers. whre relevanllnformalion WIIS found. (Relevant dol::umenrs Iisled in the !Went document (RFR. Procedure.

etc. need not be repealed).):

OL Amcndmenl 126 dated 42398; ULNRC*03681 d3tcd 11-10-97; ULNRC-03628 dated 8-8.97; COMN 43387; CAR 200700230; CAR 200700222; CAR 200607357; CAR 20060624 J: CAR 200506711 Continued 0 fdentify the relevant PSAR-described SSC(s) atJd lite aS10Ciatcd dcsign function(s) (Sec Sections 3.3 lind 5.1.1 of the lOCPRS0.59 Resource Manual (RM>>: Reactor Trip Sysrem (TS 3.3.1. FSAR Section 7.2); Enslnecrcd Safety Features Actuation SYltem (TS 3.3.2; PSAR Sections 7.3.7 and 7.3..8) -design functions Include tripping the reactor lind Isolating an 8utOmlltic feedwater isolation sisna) Continued 0 50.59 Screenlq Questions: (Che<:lt the correct responses below. Attach additional pagc(s) to provide jUllificalion for "YES" if desired. See Section 5.2.2 of RM for additional

&uidancc.)

L Docs the proposed activity involve cbange 10 an SSC such Ihat it adversely affeclS an FSAR-0 YES 181 NO described desiln function? (See Seclion 5.2.2.1 of the RM.) 2. Doc5tbe proposed aClivity involve I chanp to. procedure web lhat it adversely .frects how 0 YES 181 NO FSAR* described SSC de:lign function$

arc performed or conlrolled? (See Section 5:Z.2.2 of lhe HM.) 3. Does lhe proposed activity involve revising or replacing an FSAR-dcscribed e'IIIWllion 0 YES 181 NO melhodology that is used to establish the desisn bases or used in the safcty analyses? (See Section 5.2.'2.3 of tile RM.) Docs Ihe proposed activity involve IIlestorexpedmem not described in tbe FSAR. where anSSC 0 YES 181 NO is utilized or controlled in a manner tlull is outside .he reference bou.1ds of the desip ror that SSC or is inconsi$lent with analyscs or descriptions in lite FSAR? (Sec Section 5..2.2.4 of the HM.) 5. Does the proposed activity involve or requlrc a chanSC to 1l1c Technical Specifications?

.. 0 YES 181 NO (Sec S,;cllon 5.2.2.' of the RM.) VI. If all quc::stiol'\$

arc answered NO. lIten implcmcnt thc activIty per the applicable planl procedure ror the type of activity without obtaining a License If screen question S is answcn:d yes, then request and recoivc a License Amendment prior to Implementation of activity.

.. U screen question 1.2,3 or 4 is answered YES. then a 50.59 Evaluation shall be perrormed and approved prior Implementation or the

.. A lOCFRSO.59 eVllluatlon is NOT required for FSAR alld Technica.

Specification (TS) Bases change made strlClly conform the FSAR and TS Bascs to Technic::!1 Specifica.ion' chanses approved by tbe NRC. Such changes are typicllUy implemented in conjunction with Implementation of the NRC'approved TS changes wilbout further evulualion.

VII. If rile conclusion of the sCreening questions Is that a $0.59 EvalUalion Is not n:quircd.

pl'ovide an overall JUilific;!lion for Ihat determination.

' sceattached Continued 18 VllI SSgnolTs:

PrCl)llrer:

• • • • • Date: E:-I.::II' 1tJ::Z. osure, page CA2S1l APA-ZZ-00143 09112105 50.59 (Print name) Dale: ...J...J.-1J 01 Qualified Reviewer:

IX. Has II oopy or the c:ompleted screen been provided 10 Uc:ensing'l nc osure, page Page2of2 CA2S11 APA-ZZ.OOJ43 0911210.5 AttammeRt to lOCFR50.59 SerelR CA2511 Section VII OTG-zz..eooos AdcI. J Rey. 0 'ip2oU SCREENING QUESTIONS 10CFR50.59 screening question 1 is answered "No-since the proposed procedure changes do not involve any physical alterations to the plant (no new or different type of equipment will be installed).

There are no design changes involved to the reactor trip system (RTS), engineered safety features actuation system (ESFAS), or any other structures, systems, and components (SSCs). 10CFRSO.S9 screening question 21s answered VNo" since the proposed procedure changes do not adversely affect how any FSAR*described SSC desIgn functions are performed or . controlled.

There are changes to the procedural controls tor tripping the plant and isolating one of the FWIS functions, as discussed abOlie. These changes deal with the performance of a manual reactor trip and the isolation of the FWIS that is derived from the coinCidence of P-4 (satisfied by reactor. trip breaker position switches showing the breakers are open after a reactor trip) and low reactor coolant system (RCS) T*avg (enabled at 564 R F). The proper timing of when to manually trip the reactor has always been under licensee purview and requires no further evaluation under SO.S9. The procedure changes related to blocking FWIS in MODE 2 have already been reviewed and approved by the NRC in OL Amendment 126. 1OCFR50.59 screening question 3 is answered "No" since the proposed procedure changes do not revise or replace any FSAR-descrlbed accident evaluation methodology that is used to establish the design bases or used in the safety analyses.

No Changes to the Area S steam line break hazards analysis In FSAR Appendix 36 are required.

as discussed funher below in the response to screening question 5. 1 OCFA50.59 screening question 4 is answered *No" since the proposed procedure changes do not involve any tests or experiments not described in the FSAR. 10CFR50.S9 screening question 5 is answered NNo" since there are no changes required to the Technical Specifications (TS). TS Table 3.3.2-1 Function 8.a reqUires that the P-4 permissive be OPERABLE In MODES 1,2, and 3. However, the operability at the P-4 permissive is unaffected by these procedure changes since no.hin associated with the RTB osition switches or SSPS cabinet des' n is bel chan ed. The permissive itself will remain OPERABLE in MODES 1, 2, and 3; however. a non-critical enabled function -downstream-of P-4 will be blocked In MODE 2. This non-crilical enabled function, FWIS on P*4 coincident with low RCS T-avg, Is not a TS required SSC. If it were a TS required SSC, It would be required to be listed as a sub-function under T:$ Table 3.3.2-1 FUnction 5. It is not. FWIS on P-4 coincident with low RCS T-av does not meet an of the four criteria for TS inclusion In 10 CFR SO.36(c){2)(II).

This actuation signal is not requited to mitigate any accident This a ua on signa IS modeled in the analysis of an Area 5 steamline breaR. but only for the purposes of providing conservatively eariy'feedwater Isolation to minimize the time to SG tube uncovery for mass and energy {MlE) releases In j\rea 5 for break Sizes 5 1.2 square feet per Section 6.6.2.2.3 and Table 6.6.2*4 of ASG WCAP-16140 and WCAp*16265.

In other words, it Is only modeled where it makes the results worse -it is nOt a required design function.

Therefore, a 50.59 Evaluation (form CA2512) Is nOt required for the proposed procedure changes..

Enclosure t page 45 CA2511 Section VII Attachment tv IOCFR5D.511 Screen OTG-zz..ooots Add. 1 Re¥'.O'I.e Jon BACKGROUNp OTG-ZZ-OOOOS Addendum 1 Revision 0 was developed from steps in existing OTG*ZZ-ooooa Addendum 3 and simulator scenarios as one of two methods, in the case of this addendum by performing a manual reactor trip. for inserting the control and shutdown banks. Some differences exist between this new OTG*ZZ*OOO05 Addendum 1 Revision 0 and existing OTG* ZZ-oOOO6 Addendum 3 due to different entry conditions, Including different operatIonal MODES and 1I1e fact that the shutdown banks will now be Inserted by the manual reactor trip (MODE 2 In the case of new OTG..ZZ"()()OO5 Addendum 1 Revision 0 during a plant shutdown from 20% RTP to MODE 3 vs. a plant shutdown from MODE 3 to MODE 5 In. the case of existing OTG-ZZ-OOOO6 Addendum 3). However, these ancillary changes In the Purpose, Scope, Precautions and Limitations.

and Prerequisites are to be expected.

Changes to the Procedure Instructions are also required to reflect the different plant *conditlons that will exist at addendum entry (such as a heightened concem over excessive RCS temperature reductions and measures enacted for letdown control, different desired SG levels, different set of main control boards alarms to be expeded, checking that intermediate range NIS channels indicate a lowering neutron flux, expectation that E-O (Reactor Trip or Safety Injection).

be entered if unexpected conditions arise. and additional steps tied to restoring and stabilizing SG level, RCS pressure, and RCS temperature).

AU of these changes are also to be expected given the different set of plant conditions associated with addendum usage. Changes have also been made to replace a specific SG narrow range Jevelln some steps with a reference to the control or program band, but these are equivalent changes requiring no further screening.

Therefore, the timing of rod Insertion and blocking an automatic feedwater Isolation signal is at the crux of this 50.59 screenIng form. In the NRC's Safety Evaluation for OL Amendment 126 dated 4-23-98,lhe NRC staff speoillcally reviewed and found acceptable our feedwater Isolation signal (FWIS) bypass switch deSign and our using it to block the FWIS initiated by 1I1e coincidence of P-4 and low T-avg as long as its use was limIted to the followIng plant conditions: T*avg less than or equal to 564 I F (the plant can be In MODE 1 or 2, but T*avg must 564 1t F) just prior to opening the reactor trip breakers (RTBs) which satiSfies the P-4 portion of th's feedwater isolation signal's logic. These limitations will be met by step 5.2.11.b of OTG-ZZ-00005 Revision 25 which is being issued concurrently with this OTG*ZZ.QOOO5 Addendum 1 Revision O. NRC also wanted this particular FWIS function to be restored by defeating 1he bypass prior to entering MODE 2 ascendIng during startup from an outage. Thls"lImitatlon is met by step 4.16 of existing OTG-ZZ-oOOO2 Revision 36. As long as these limitations are observed, t!'le plant will operate within the bounds of an amendment preVfously reviewed and approved by NRC. Enclosure, page 46 This page and the following 4 pages are taken from NRC IR 05000462/2009004 which is found in ADAMS as ML093140a03.

The cause of the finding has a problem Identification and resolution crosscutting aspect in the area associated with the corrective action program because Wolf Creek failed to thoroughly evaluate the failure mechanism such that the resolutions address the causes and extent of conditions, as necessary.

Specifically Wolf Creek did not properly consider the possibility of common-cause pitting failures which could have Impacted the essential service water piping Train A structural integrity thereby aft'eding Its cooling loads. including the Emergency Diesel Generator A (P.1(c) (Sedion 1R15). Green. The inspectors identified a noncited violation of Technical Specification 3.8.1, Required Action B.4.2.2 on March 24, 2009 when the licensee performed eledive maintenance on safety bus relays and removed equipment from service that was required by the technical specifICation and the NRC Safety Evaluation Report (SER) while In an extended diesel generator outage. The maintenance had the potential to open the normal offslte feeder breaker. This issue has been entered Into the corrective adlon program as Condition Report 15727. The inspectors determined that the failure to Implement requirements of Technical Specification 3.8.1 and the associated NRC safety evaluation was a performance deficiency.

The finding was more than minor because it is associated with the equipment performance attribute for the Mitigating SYstems Comerstone and affected the comerstone objective to ensure the availability, reliability.

and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding was determined to be of very low safety significance because the issue did not result in the Train B offslte power being inoperable for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and did not involve extemal events such as flooding.

Additionally, the cause of the finding has a problem identification and resolution crosscutting aspect In the area associated with the Corrective aelion program. SpecifICally, Wolf Creek did an extent of condition review In response to a previous violation which Included Procedure STS IC*20BB, but still failed to prohibit performance of STS IC*20BB during the 7-day diesel outages (P.1(c)(5ection 1R19). ' Gre On ugust,

• e Inspe ors ent a none te violatIOn ofTechnical pecification 3.0.3 in which both trains ofTechnlcal Specification 3.3.2 engineered safety features actuation system interlock function B.a were bypassed with Jumper wires in accordance with a plant procedure.

Fundion B.a is the Interlock for reactor trip signal coincident with 10 Tave Signal. Wolf Creek blocked the signal from the feedwater valves with jumper wires during control rod drive motor-generator testing in Mode 3. The Inspectors and the NRR technical specification branch found this to be eontl'8ry to the Updated Safety Analysis Report. the technical specifications.

the technical specification bases and the NRC safe 'evaluatiOns su rtl the technical s deatlons.

he licensee entered this issue in their corrective action program as Condition Report 1'931B. The inspeclors faun that the fal ure to imp ment Technical Specification 3.3.2 interlock.

function B.a was a performance defICiency.

The inspectors determined that this finding was more than minor because it Is associated with the design control attribute of the Mitigating Systems Cornerstone and it affected the cornerstone objective to ensure the availability.

reliability, and capability of mitigating systems that respond to initiating ,events to prevent undesirable con!!9uences (I.e., core dams e. 8 ,nspec ors evaluated the significance of this finding using Inspection anual Chapter 0609.04, EnclOsure Enclosure, page 47

• Phase 1 -InlUal Screening and Characterization of Findings:

and screened the finding to Phase 2 because the finding represents a loss of a system's function.

The inspectors used Inspection Manual Chapter 0609. Appendix A and screened the finding to'the NRC senior reactor analyst for review because there was not an acceptable equipment deficiency in the pre-solved worksheet.

The senior reactor analyst determined tliat the finding Is Green bacause he solved Table 3.10 of the Risk-Informed Inspection Notebook for Wolf Creek Generating Station, Revision 2.1a and found that the loss of feedwater isolation sl nal for less than 3 de s resulted In a 1 E*7 Green outcome. The inspectors also determined that the cause of the finding has a crosscutting aspect in the human performance area associated with decision making because Wolf Creek failed to make a risk significant decision using a systematic process. This issue was evaluated more than once and those evaluations sought to Justify bypassing the interlock rather than seek the full r ulato basis for the interlock H.1.a 1 R15 * .G.ru.!1.

The Inspectors identified a nonclted vlolation of 10 CFR 50 Appendix B, Criterion III, "Design Control," for failing to translate the boric acid design basis into procedures that ensure time sensitive operator actions are completed to achieve lIie core shutdown margin specified In the core operating limits report. Performance Improvement Request 2005-34611dentifl8d that ifthe room coolers were started while lake temperature was low, the boric acid solution temperature decrease below the solubility limit. Corrective actions for heat tracing and room temperature logging took approximately 3 years to implement and stopped short of addressing boric acid system operation when nonsafety power is lost to the *heat tracing alid the plant must be taken to cold shutdown in accordance with technical specifications.

The licensee entered this issue in their corrective action program as Condition Report The failure to translate the design bases Into procedures that ensure the function of the safety-related boric acid system upon loss of nonsafety-related heat tracing is a performance deficiency.

The inspectors determined that this finding was more than minor because this issue aligned with Inspection Manual Chapter 0612. Appendix E, example 2.1, because the pipe temperature was required to stay above the boric acid solubility limit and the loss of the heat tracIng and or room temperature decrease will block the boric acid system. This Issue was associated with the equipment pe.rformance aUrlbute of the mitigating systems cornerstone and affected the cornerstone objedlve to ensure the availability, reliability.

and capability of systems that respond to Initiating events. The inspectors evaluated the significance of thIs finding using Phase 1 C?f. Inspection Manual Chapter 0609, Appendix A, -Significance Determination of Reactor Inspection Findings for At Power Situations," and determined that the finding screened to phase 2 because the issue was a design or qualifICation deficiency conflf1T1ed to result in loss of operability or functionality The Inspectors evaluated the signiflcan<:e of this finding using Phase 2 of Inspection Manual Chapter 0609, Risk Informed Inspection Notebook for Wolf Creek Generating Station. and determined that the finding was of very low safety significance because loss of the boric acid system in Table 3.9 for one year resulted In a 1E-7 CDF when giving recovery credit for the refueling water storage tank. The Inspectors determined that this finding haS a crosscutting aspect In the area of problem identification and resolution associated with the correctfve action program component becaiJse Wolf Creek did not take appropriate corrective actions to resolve known deficiencies in the design and operation of the boric acid system for ' .4* Enclosure Enclosure, page 48

• I t i.On August 22, 2009. the inspectors e . led a Violation of Technical SpecificatIOn 3.0.3 In which both trains of a Technical Specification 3.3.2 Interlock In the engineered safety features actuation system were bypassed with Jumper wiNS In accordance with plant procedure. . The inspectors reviewed the technicel speclflcatlon bases for the engineered safety features actuation system interlocks and function 8.a. The bases and USAR state that the functions of the interlock are to: 1) trip the main turbine, 2) isolate main feed water coincident with 10 Tavg, 3) allow manual block of the automatic re..actuation of safety injection after a manual reset of safety injection.

4) allow arming of the steam dump valves and transfer the steam dump from the load rejection Tavg controlier to the plant trip controller, 5) prevents opening of the main feed water isolation valves If they were closed on safety injection or steam generator hi*hi water level. The Inspectors found that this was consistent with the standard Improved technical specifications for Weslinghouse plants and the Wolf Creek USAR, Table 7.3-15. "NSSS Interlocks for Engineered Safety Feature Actuation System: Under License Amendment 123. Wolf Creek converted to improved standard technical specifications In December 1999. The P-4 Interlock description has not changed since 1999. The licensee submittals acknowledged that the functions of P-4 were not part of a design basIs analysis, but wera retained in the technical s !Hcatlons to limit reactor coolant stem cooldown 'ollowio a reactor tri . Technical peciflcatlon 3.3.2 states that .ESFAS [engineered sa ety eatures actuation signaij instrumentation for each Function in Table 3.3.2 shall be OPERABLE According to Table 3.3.2-1: Function 8 of Table 3.2.*1 covers interlocks and specifically interlock B.a. P-4, is required to be Operable In Modes 1, 2. and 3. The inspectors found. that function 8.a is required in Modes 1,2, and 3. The Inspectors consuhed with the . Office of Nuclear Reactor Regulation's lec:hnical specification branch and found that statements in the bases provide a summary of the technical specification and do not override requirements.

The sentence In the biiSUG that states: *Thls Function must be OPERABLE in MODES 1, 2, and 3 when the reactor may be critical or approaching criticality,*

clarifies why it Is required in Modes 1, 2. and 3 and does not permit P-4 to be inoperable If the reactor is not approaching criticality.

Operators are trained to anticipate criticality such as during control rod-drive motor-generatortesling during August 22-23, 2009. During interviews, Wolf Creek stated that it was necessary to bypass the P4/FWIS in order 10 perform rod-drive motor-generator set testing that cyded the reactor trip breakers.

Wolf Creek contended that the P-4IFWIS was not necessary to assure oompliance with the* plant safety analysiS.

lastly, Wolf Creek stated that during Mode 3 after refueling outages, it was necessary to install jumpers and bypass the P-4/FWIS for Enclosure Enclosure, page rod-drop testing because operation of the main feed water system in automatic level control was more desirable than having an operator manually control steam generator levels with auxiliary feedwater.

The inspectors agreed that this interlock is not assumed in Chapter 15 of the USAR. but the inspectors found that the Wolf Creek technical specification bases state that *eSFAS Instrumentation satisfies Criterion 3 of 10 CFR 5O.36(c)(2)(ii)" which is identical to the generic standard specifications approved by the NRC. The inspectors found that there are several technical specification systems such as steam generator atmospheric relief valves, the condensate storage tank, and pressurizer power operated relief valves that are not in Chapter 15 ofthe USAR but are required to be operable under technical specifications per 10 CFR 50.36. Thus, the inspectors found that the interiock's absence In Chapter 15 of the USAR does not mean it Is not required by the technical specific8tloi'l.

Wolf Creek previously evaluated this condition In Performance Improvement Request 2001-0041 which conc!uded this P-4/FWIS was not required to be operable In any Mode. because it Is not credited In Chapter 15 of the USAR. Wolf Creek also used other plants with NRC approved safety evaluations to justify the use of Procedure SYS68-122 rather than requesting a license amendment.

The Inspectors found that these conclusions are Incorrect.

after Inspector questioning, and afterward, operators incorrectly applied Technical Specification 3.3.2, Condition F, which allowed 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> to return one train of the interlock to service. Wllh both trains of P4 bypassed, Technical Specification 3.0.3 applied and Wolf Creek had 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> to be In Mode 4. The P-4 interlock was Inoperable for approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> from August 22-23, 2009. Wolf Creek missed the lransition to Mode 4. e nspectors oun a contro room opera ors I no og e nopera I! Y 0 Anal Is. The nspectora found that the failure to evaluate implement Technical Specification 3.3.2 Interlock, function 8.a was a performance deficiency.

The inspectors determined that this finding was more than minor because It Is associated with the .design control attribute of the Mitigating Systems Cornerstone and it affected the cornerstone objective to ensure the availability.

reliability, and capability of mitigating systems that res ond to InlUat! events to nt undesirable consequences i.e** core *de e. The inspectors eva usle the srgn' canee of this finding using Inspection Manual Chapter 0609.04. 'Phase 1 -Initial Screening and Characterization of Findings,n and screened the finding to Phase 2 because the finding represents a loss of a system's function.

The inspectors used Inspection Manual Chapter 0609, Appendix 'A and screened the finding to the NRC senior reactor analyst for review because there was not an acceptable equipment deficiency in the pre-solvfld worksheet.

The senior reactor analyst determined that the finding Is Green because he solved Table 3.10 of the ' Risk-Informed Inspection Notebook for Wolf Creek Generating Station, Revision 2.18 and found that the loss of feedwater isolation sl nal for less than 3 dB 8 resulted in a 1 E-7 (Green) outcome. The inspectors also determined that the cause of the finding has a cros u ng aspe n the human performance area associated with decision making because Wolf Creek failed to make a risk ,Ignlflcant decision using a systematic process. This Issue was evaluated more than once and those evaluations sought to 'ust' b ass!n the interlock rather than seek the full re Uleta basis for the interlock.

H.1.a] Endosure Enclosure, page 50 Enforcement.

Wolf Creek Technical Speclflcation, Table 3.3.2.1, function 8 Includes engineered safety features actuation system interlocks.

Function 8.a, the P-4lnterlock, requires two trains to be operable in Modes 1. 2, and 3. Function 8.a does not provide a required action for both trains of engineered safety features actuation system interlocks inoperable.

Wolf Creek Technical Specification 3.0.3 requires the plant to be In Mode 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> if there is no required action specified for a limiting condition of operation that cannot be met. Contrary to the above, from August 22 to August 23, 2009. Wolf Creek failed to change modes from Mode 3 to Mode 4 when both trains of engineered safety features actuation system Interlock function 8.a, P-4, were inoperable for greater than 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Specifically.

from August 22 to 23. 2009. Wolf Creek failed to change modes from Mode 3 to Mode 4 lNtIen both trains were removed from service for approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Because this violation was determined to be of very low safety significance and was placed in the corrective action program as Condition Report 19318 this violation is being treated as a nonciled violation in accordance with Section VI.A.1 of the Enforcement Polley: NCV 0500048212009004*04, "Failure to Implement J:ngineered Safety Features Actuation System Technical,Speclfication Results in Missed Mode Chanae." 1 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications (71111.17) a. Inspection Scope The Inspectors reviewed the effectiveness of the Iicensee'S implementation of evaluations performed In accordance wtth 10 CFR 50.59, "Changes, Tests, and Experiments," and changes, tests, experiments, or methodology changes that the' licensee determined did not require 10CFR 50.59 evaluations.

The Inspection procedure requires the review of 6 to 12 licensee evaluations required by 10 CFR 50.59. 12 to 25 changes, tests. or experiments that were screened out by the licensee and 5 to 15 permanent plant modifications.

The Inspectors reviewed 9 evaluations required by 10 CFR 50.59. These included:

2006.001, Radiological Consequences of a Fuel Handling Accident, Revision 0* 2008-0006.

Wolf Creek Generating Stlltion (WCGS) Simplified Head Assembly* (SHA) Drop Analysis, Revision 0 200a.0008, Use of Dedit:ated Operator for SI Pump B Room cooler* Replacement.

ReviSion 0 2005-004, WCGS Rod Withdrawal at Power Event Safely Analysis, 0* 2008-001.

Evaluations of Voids In the ECCS Suction Piping. Revision 0* 2008-002.

Evaluations of Voids in the ECCS Discharge Piping. Revision 0* 2006-002, Power Operation, Revision 54* Enclosure Enclosure, page 51 This page is taken from NRC Integrated Inspection Report 05000482/2009005 which is found in ADAMS as ML100430713.

The finding was more than minor because H was associated with the configuration control (reactivity control) attribute of the Barrier Integrity Cornerstone, and It affected the cornerstone objective to provide reasonable assurance that physical design barrie", (fuel cladding, reaelor coolant system. and containment) protect the public from radionucllde releases caused by accidents or events. ThE! .inspectors evaluated the significance of this finding using Phase 1 of Inspection Marrual Chapter 0609.04. and determined that the finding screened to Green because the P-6 interlock only affected the fuel barrier (Section 40A2). This findlF1g was not assigned a crosscutting aspect because the C8\Jse was not representative of current performance.

Cornerstone:

Occupational Radiation Safety The Inspector identlfled a nonclted violation of Technical Specification 5.1.2.a.1 for failure to maintain administrative control of door and gate keys to high radiation areas .with dose rates greater than 1 rem per hour but less than 500 rads per hour (referred to as locked high radiation areas). Specifically, as of October 21, 2009, the licensee did not have administrative controls over a single master key to locked high radiation areas. This issue was entered Into the licensee's corrective aellon program as Condition Report 20973. Failure to maintain administrative control mthe master key to locked high radiation areas was a performance deficiency.

This finding Is greater than minor because If left uncorrected the finding has the potential to lead to a more significant safety concern in that an Individual could receive unanticipated radiation dose by gaining access a locked high radiation area without the proper controls and briefing.

This finding was evaluated*

using the occupational radiation safely slgnificence determination process and determined to be of very low safety Significance because It did not involve: (1) as low as Is reasonably achievable planning or work control issue, (2) an overexposure, (3) 8 substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, the violation has a crosscutting aspect in the area of human performance associated with the work practices component because the lack of peer and self-checKing resulted In inadequate control of keys to high radiation areas H.4(a) (Section 2OS1). Cornerstone:

Miscellaneous Severity LevellY. The Inspectors identlfied a Severity Level IV noncited violation of 10 CFR 50.73 In whlctl the licensee failed to submit a licensee event report within 60 days following discovery of events or conditions meeting the reportablllty criteria.

On December 31,2009. the Inspectors identified a licensee event report that was no timely. Licensee Event Report 2009-009-00 was not Issued within 60 de s for a condition ohl lted b technical s eclflcatlons, and the event report did nOlldentlfy that the disabling of both trains of the P interloc on ugUSI 22. 2009 was also reportable per to CFR 5O.73(a)(2)(v).

The P*4 Interlock was required by Technical Speciftcatlon 3.3.2, function 8.a, and Is discussed in USAR. Section 7.3.8, "NSSS Engineered Safety Feature Actuation System.* Wolf Creek licensee event report 2009-009 was correct in that the Inter10ck Is not credited In accident analysis.

However, NUREG 1022, Section 3.2.6, specifies that inoperable systems required by.the technical specifications be reported, even if there are other diverse operable means of accom lishin the safet function.

-10.* Enclosure, page 5a Ellclosure2

,-,', NRC FORIII HI u.s. NUCLEAR REGULAtORY COMIIIIIIION APPROVED BV OM8: NO. )101).0104 EXPIRES: 0lIl1/1.010 NM7) [ This LER is In ADAMS as ML 100890421.

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.. t..aR HUMBER 7. REPORT DATil ** OTHER FACIUl1EIINVOLYI!O MONTH PAy YE:AR YEAR I SEQUENrlo\L I REV MONTH DAY YI'AR ' 'lICl.I1'Y IIi\IIE DOC:IIJT_R NUU8ER NO. 05000 FACIUT\' IWIII!. DOClIrTNUIoIII!'

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/WIIi Richard D. Flannigan.

Manager Regulatory Affairs (620) 364-4117 13. COMPLEll!

ONE UN!! FOR EACH gOllPONEN1' fAILURE DESCRIBEO

..THIS REPORT I ! t.MNU. REPORTAIILE CAUSE COMPONENT lWIU. REPORTABLE CAuse SYSTEII ! COMPONENT FACTURI!A i TOf-PIX SVliTlito' FAClURER TOIiPIX 14. SUPPLEMENTAl.

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'/'/lIs} On August 22. 2009. at 0540 hCNJra Central Daylight 11me (COT). wit" 1"9 plant in Mode 3, Control Room ataft' defeated the feedwater isolation on lowT811g coincidenlWith P-4 Function using procedure SVS S8-122. "Enablirlg/Disabllng P""/Lo Tavg FWlS." This p'ocedullJ was perfonned fnr restoring main feedwater now Ihrough the main feedWater isolation valves (MFIVs) to supply water 10 the steam generators.

On August 23,2009, al 0125 hours0.00145 days <br />0.0347 hours <br />2.066799e-4 weeks <br />4.75625e-5 months <br /> the jumpers Installed for defeating the feedWater isOlation on low Tavg coIncident with P-4 functIon were removed and procedure SYS S8-122 completed al 0140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br />. The Nuclear Reguletory Commission (NRC) Resident questioned the defeating of the leedwater i6Olation on k7N Tavg coincident wtth P-4 function while In Mode 3. Technical SpecI1icatton (TS) 3.3.2, Table 3.3.2-1 spec:ifies the appHcable Mode fOr Function 8.a. (Reactor Trip. P-4) as Modes 1.2. 3. Defeating the feedwaler Isolation on loW Tavg coincident with P-4 function using SVS S8*122 results In both channels being def.ated.

There Is no TS Condition for two inoperable trains. limiting Condition for Operation (LCO) 3.0.3 specifies thatwhert an aSSOCiated Action Is not provided, action shall be Initiated within 1 to place the plant In Mode 4 in 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Action was not take" 81 required by the TSs. NRC Enclosure, page This document is found in ADAMS as MLIOII00391.

This is the document whereby Wolf Creek' formally requested to change its Technical Specification to allow blocking the Lo-Tavg!P4 FWIS during MODE 3. U. S. Nuclear Regulatory commISSion April 13.2010 ET 1D-0014 AnN: Cocument Control Desk Washington.

CC 20SSS Subject Docket No. 50-482: Application To Revise Technical Specification 3.3.2. *Englneered Safety Feature Actuation System Instrumentation,'

Table 3.3.2-1 ' Gentlemen; Pursuant to 10 CFR 50.90. Wotf Creek Nuclear Operating Corporation (WCNOC) hereby requests an amendment to Renewed Facnity Operating License No. NPF-4I2 for the Wolf Creek Generating Station (WCGS). The proposed amenClment revises Table 3.3.2-1, Function 8.8., (ESFAS Intertocks, Reactor Trip, P-41) of Technical Specification (TS) 3.3.2. -Engineered Safety Feature Actuation System Instrumentaticn:

WCNOC Is proposing to add footnote (m) to Function 8.a. to identify the enabled functions and the apPlicable MODES for the Reactor Trip, P-4 Interlock Function.

Attachment , through IV provide the Evaluation, Markup of TSs, Retyped TS pages. and proposed TS Bases changes, respectively, in support of this amendment request. Attachment fIJ, proposed changes to the TS Bases, Is provided for Information.

only. Final TS Bases cI'Ianges will be Implemented PUl1luani to TS 5.5.14. "Technical Specffic:ation (TSl Bases Control Program: at the time the amendment is implemented.

Attachment V provides I List of Regulatory Commitments made by WCNOC in this Submittal.. . It has been determined that this amendment application does not Involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22{b).

no environmental Impact statement or environmental assessment needs to be prepared in connection with the Issuance of this amendment.

This amendment application was reviewed by the Plant Safety Review Committee.

In accordance with 10 CFR 50.91. a copy of this amendment application.

with at1achments, is being provided to the designated Kansas State official.

Bnclosure.

page ,54 P.o. Sex "11 I SUl1IlIIJIon.

KS flll83V; PhOne: (620) 364-8831 Art lqual Oppo!1unly IM'{HI;N!T March 30. 2011 Mr. Matthew W. Sunseri President and Chief E)(ecutive Officer Wolf Creele Nuclear Operating Corporation Post 0I'f\ce Bo)( 411 Burlington, KS 66839 UNrrED NUCLEAR REGULATORY WASHI"IGTON, D.C.

ADAMS as MLl10550846.

This is the document whereby the US NRC granted Wolf Creek permission to change their Technical Specifications such that the LO-Tavg/P-4 FWIS could be blocked in MODS 3.

WOLF CREEK GENERATING STATION -ISSUANCE OF AMENDMENT RE: REVISE TABLE 3.3.2-1 OF TeCHNICAL SPECIFICATION 3.3.2, -ENGINEERED SAFETY FEATURE ACTUATION SYSTEM (ESFAS) INSTRUMENTATION-(TAC NO. ME3762)

Dear Mr. Sunseri:

The U.S. Nuclear Regulatory Commission (the Commission) has I$$ued the enclosed Amendment No. 194 to Renewed Facility Operating License No. NPF-42 for the Wolf Creel< Generating Station. The amendment consiets of chenges to the Technical Speclfications (TSI) in response to your application dated April 13. 2010, as supplemented by letters dated October 13 and December 21, 2010, and January 1a, 201 1. The amendment revises TS Table 3.3.2-1, Function a.8 (Reactor Trip, P-4) by addil'19 footnote (m) to identify the enabled functions and the applicable modes for the Reactor Trip, P-4 interlock func1ion.

A copy of our related Safety Evaluation is enclosed.

The Notice of Issuance will be included in the Commission's neld biweekly Federal Register notice. Sincerely, J-BalWant K. Singal. Senior Prolect Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nudear Reactor Regulation Docket No. 50....a2 Enclosure.:

1. Amendment No. 194 to NPF-42 2. Safety Evaluation cc w/encls: Distribution via Listserv Enclosure, page S5 ESFAS Instrumentation 3.3.2 TltH 3.3.2*t (pege II 01 'I El\VltIftrld s.r*.,. tatllfe AcWtiol\ Splem 111&"""" ....1101'1 APPLlCA81E MODES OR OlllER SPEtlFIED REQUIRED SURWll.l.ANCE AlLOWABLe CONDITIONS tlWlNElS CONDITIONS REQUIREMENTS VAlUEII) 1. /l1IIomItUc; SwIGhoYe' to ConIIIInment Svmp I. AVIoIl1ltlh:ActuaIIon LQVltandAIlII.!IIIIOn Relay. D. Rtlutlinl!

Will' Slcnge T.nI; (RW8T) L8YtI1 *l_Low '.2.3,. 1.2,3,. 4 c 8R3.3.22 SR3.U4 BR3.l.U )( SR3.U.1 a'\3.u.1I 8R33.2.' SRU.2.10 F SRl.:U" IiA 1.2,3 :I L SR3.U.S SR 3.3.2.9 e. Readot Tlip. p.*I"") 1.2.3 je) Th, AlIowatit V'lIuO dtIInu .ho Llmllina SaI'eIr 61""'" Settingl.

See I_lot!he Tlip palnlS.

• Pr_nt, Cl*'ino of MFIV,' cIoHd _ SlIM' 80 W.., Lew!-Hig/I HIOI'I-MOO!'S 1. 2.111'K1 3 1m) _I81-MODES '. 2, end 3 ThiS is page 6 of MLl10550B46.

It shows how the Wolf Creek Technical Specifications were revised under Amendment 196 to allow wolf Creek to disable the P-4/564°F FWIS during MODE 3. Wolf Creek -Unit 1 3.3*35 Amendment No. 123. 126. 133. 183. Enclosure, page 56 194 This sheet can be found in ADAMS as page 63 of It shows how Table 3.3.2-1 ESFAS Instrumentation function 8.a, appears in Callaway Plant's 3.3.2 Technical Snecifications as of Julv 29 2011. TIbII 3.3.2*1 (PIliii' 10 of Engineered Safety AdulUoro Sy.l_

APPliCABLE MODES 00 OTHER SPECIFIEO ReQUIRED SURVEIL.Li'NCE ALLOWASI.E FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE"" Aulomallc SwIIchowIr 10 Conteitlmlllll Sump e. Automalie ACtuallon LogIC anr:lAelulllion RIllY' (SSPS) b. Refueln\ Will., .I1k TIL._I. LcwLow Cotnctdel'll with S.'aty Inl.r;lion ESFAS Inl"loo", R.Klor Trip, P"" P'.Slurlz" Pressure.

P*'1 Automatic:

,","&\.11'1111' Actuellen AulOlNlllc Aculian LoIiJIc Inc! Actuation ReleyflSSPSI

b. Pl'Hlur\zar Prnlura w HigPl 2lrainl C Sf( SH 3.J.2 1:1 12.3.4 4 SR3.:j.:n 35.2'1l SRl,3.25

!:iR 3.3.:/ Ii) R,r,r 10 FuACIlon 1 (Salelylr4ltC11C1n) for ,U InttJalion funcUonl and requlfem."...

1.2.3 1.2.3 1.2.3 U.3 (a) The AII_IIbIB Valla **nn!he . B.d.(1). Ind 8.(1.(2) (!tie NomJIIIII Trip S.... lor IN Nonlftal T,ip SelPQln\t.

2 r:t,.;n, lTaillS 3 21rain1 .. F SR 3.3.2.11 NA L SA 3.3.2,5 SR J 3.2.9 s 1981 psig H SF< 3.3.2.i! SR 3.3.2.'1 SF! 3.3.2.'" Nil. D SA 3.302.1 SJl3.3.Z.5 SR 33.2.9 s2l10 11&111 IIfII'ItIy

.".......

lor Func1Iana

' *** 4 .*.(1). S .... 5 *.(1). 5.*.(2). , cllIIInH ..NII\ *ng.a"'Y .y.l..., ..Wng for .... FunC!lOnI).

See !hi Note that, just like at Wolf Creek prior to Amendment No. 194, the Callaway Plant Technical Specification for function a.a is applicable in MODE 3. CALLAWAY PLANT 3.349 Amendment No. 202 , Enclosure, page S7