ML19317F709
| ML19317F709 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/16/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317F697 | List: |
| References | |
| NUDOCS 8001230601 | |
| Download: ML19317F709 (10) | |
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i SAFLTY EVALUATIQ3 BY DIE OPPICE CF 20 CLEAR R%T.R RIG LATIG SUPM.fCH C.Y.nEI!SC ?O.11 it LICCISC l0. !;PF-3 TOLEDO EDI Qi CQlPA:3Y MD CLEVEI).ND ELCCTRIC ILLUllDIKfLG CQtPMY 7
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i d J dq A NU I:CNCCCTIQi Dy letter dated Acril 10, 1973, tne Toledo Edisen Ccacany requestad a enange in the Tecnnical Specifications for tne "GiB liargin" reactor ecolant flow rates to acemlate the Departure From tiucleate Soiling Batio (aGR) penalty, as specified in license condition 2.C.(3)(i) of Pacility Ocerating Licensa IIPF-3.
Se proposed change involved balancing the Fuel rod Dowing Penalty of 11.2 percent (descrined in Section 4.4 of Suppleoent rio. I to our Safety Evaluatica report) cy taking credit for: (1) a 1 percent aua credit for tne Flow Area Feduction Facter; (2) a 1.1 percent crecit for the CuBR i-Ower Spike Factor, and (3) a 9.8 percent UCR credit for increasing tne requir-d reactor coolant flow by 5 percent.
In addition, by letter dated May 18, 1973, the Toledo Edison Cor:cany recuestw enanges in the Tecnnical Specificatiens because of recoval of tne aurnania Poison Rod Asserolles (2PaA) following evidence of wear of the hold dosn devices for tne EPPA's. On !!ay 26, 1978, the Toledo Edison Cccoany roviseu their May 18, 1978 request to include enanges in tne rechnical Specificctions Ncause of the trecnanical wear also oDserved on the Orifico Ecd Asserolics (CRAs). We Toledo Edison CcEcany stated that it was prucent to remove all SPBAa and all cut two of tne OFAs from the core internals of T,avic Jecs?, i.; nit 1 cefore the ccepletion of the first cycle of operation to avoid tne possicle darage to the plant from a potential failure of the holc down devices.
The reccval of the BPEAs and ORAs result in changes in various nuclear paramters, as well as resulting in an increase in core bypass flow. Changes to the Tecnnical Specifications are required as a result of c.ungas in ene nuclear parameters, as well as an increase in core cypass ficw.
Se Toledo Edison Coccany provided, as an attacntrent to tneir latter of May 18,1978, the Babcock and Wilcox document, PAi-1489, "A;olication to Amend Operating License for Recoval of Burnable Poison ibd Ascecolies -
Cavia-Besse Nuclear Generating Station, Unit 1," and by their letter of M
May 26, 1978 provided IWi-1489, Revision 1, " Application to Amend Operating oPFtCE D
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License for Removal of Burnable Poison Pod and Orifice red Assemolies -
Davis-desse Nuclear Generating Station, Unit 1." CM-1439 and DM-1489, Revision 1 provided aralyses supr)orting the proposed changes to the Tecnnical S;ecifications.
Jota requests fcr apclication to revise the wchnical Specificaticns (i.e. the letters of April 10, 1978 and day 18, 1978, as surplemnted) required staff evaluation of the thermal hydraulic design and tne accuracy of the ccserved reactor coolant flow rate in excess of design flow rate. Tnerefore, our evaluation of both requests for changes to tne Technical Specifications interface closely and are provided in tne discussion and evaluation provided below.
DISCUSSIGI SPPA. are used in the tirst cycle of MW reactors to control art of tne initial excess reactivity and to flatten the radial pcwer distributica.
Ce reactivity controlled by curnabic poison reduces the amunt wnica cust ce controlled by coluule coron and prevents tne cccurrence of a yci-tive aaderatcr cociticient above 95 percent of full pcwcr. The Davis-Eense, Unit I reactor has acnieved a first cycle burnup of 37 Cffective Pull Paa.cr Jays (2PCs) and some of tne burnable poincn nas oecn ourned out. ;Iowever, sufficient burnable poison remaina to recuire ccre cnangas in crder to offsot tho effect of its rec:cval. Tnese ccra changes were:
1.
Intercnange of fo"r internediate (2.63 w/c) enrien.nent bundles near tne center of the v ce with 4 Icw (1.9d w/o) enrien:acnt bunalas near tne core perinnery, 2.
Rearrangemnt of tne control roc groupings and deccupling of group 7 from tne withdrawal acquence. In tce regrouping, control roc group 7 nas been shif toc toward tne perinnery and rer.ains in ene core until a burnup of 145 FEPCs nas c4en reached.. This arrange ent sarves to turtner flatten the radial ;;over distrisutien ar.3 to replace some of tne fixed poison in tne core and tnus prevent tne scderator coefficient from cecoming positive.
Se Toledo Ediaon Company has perferoed an analysis of the a3dified core, assuming that the modification occurred at 80 FJPD and that the cycle length is increased frca 433 to 485 TPD. We analysis was performed using the same calculational.ethods and techniques that have been egloyed in tne oesign of otner sW reacters-including Davis-aesse, Unit 1.
The core physics parameters have teen calculated for the modified cycle--E to 145 FEPD with groups 5 and 6 partially inserted into the coro anc group 7 comcletely inserted followed oy 145 to 485 EFPD with groups 5 througn 7 nearly out of tne core. me recalculated parameters included snutdown ornese sunwa us >
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LL margins, rod bank worths, ejected and dropped rod worths,'5 tuck rod worth, cocolor coefficient, mderator coefficient, xenon worth, boron wortn, and critical boron concentration.
During recoval of the BPFAs it was discovered that sufficient wear was present on the holddown devices for the orifice rod assemblies (0:<c) to warrant tneir recoval. By letter dated May 26, 1978, the Toledo Edison Company suomitted Revision 1 to SMi-1489 to encompass the renoval of the cms frem the Davis-Besse, Unit 1 core.
All of the OFAs will be renoved with the exception of two modified orifice rod assecolies wnich are used with a primary neutron source. The rerrval of tne CFAs increases the flow througn tne guide tuces but r1ces not signi-ficantly alter the physics paraceters. Tnus, tne analyses presented in
%!-1439 remain in effect.
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- re nave reviewed the information presented in 2Al-1489 for tne values cf tne pnysics parareters and core flow and their effect on tne safety analyses for Davis-2csse, Unit 1.
For the roc witndrawal transients at full am ero powrs, the control rod misoperation transient, tne rod ejection acci-cent, tne nederator dilution transient, cold water accident, steam line failure accident, loss-ct-coolant accident, and loss-of-ncroal-fcccwater transient, tne significant parameters are shown to ce be'rded cy those used in tne Final Safety Analysis Report analysis. Thus, the consecuences of tnese transients and accidents will not be greater than tnoce coscribed in the Final Safety.'inalysis Eeport.
':ta loss of electric oower transient and tne steam generator tuce failure are independent of the significant paraceter changes anc the Final Safety Analysis Report analyses are, therefore, applicable for enese transients.
By letter dated June 6,1978 the Toleco Eoiscn Ccmcany su:raittec Fevision 2 to h-1439 providing a revised B&W analysis for the loss of ficw transient and the feedwater system calfunction transient. The minicu aina transient is the one-psp loss-of-flow transient.nica results in a minima niaa of 1.45.
It should be noted that the Davis-3 esse, Unit 1 Final Safety Analysis Report and BAW-1489 indicated that the most limiting loss-of-flow transient was a four-pump loss of flow transient.
The one-grap loss of transient became the most limiting transient wnen the power incalance/ flow reactor trip was adjusted to decrease inadvertent power imoalance/ flow reactor tripe. This trip adjustment was made prior to operation of Davis-Geose, Unit 1.
It should also be cointed out that incorporating margin to compensate for fuel rod bow results in a minimum required uma of 1.445, and thus the limiting loss-of-flow transient ra=1ano in a minimum DtiaR of 1.45 is acceptable.
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__J Ieraoval of all tne BPRAs and all but two of the oms from tne core results in a calculatcc increase of 4.7 percent in tce maxir.ua core bypass flow (trcn 6.04 percent to 10.75 percent). By letter dated Acril 10, 1973, the Toledo Edison Cor.pany requested that the minima allowaola reactor coolant ficw to increasec by 5 percent over the Final Safety Analysis F.coort design riow to cccoensate for the pctential e:fects of tuel rod ccwing.
Therefore, modified operating conditiens have been prococed to coaccnsate for cotn the increased cypass flow and the octential effects of rod cow on tne core thermal safety raargin. An analysis has ceen cerformed, casad on a minirun allowaolo flow rate of 110 percent of design flow anc a sligntly adjusted trip limit curve (Technical Specification Figure 2.1-1) for raact0r coolant core outlet pressure and cutlet tec erature. Ine analysis results indicate that operation at the procosc-J limits witn SPMs and OfiAs rcr.cVec culd not result in violation of acceptacle fuel design limit:. Reactor coolant system flow seasurements have indicated an actual e area ficw rata or at least 113 percent of tne pre!!ous limit (measure.renc crrors not included).
In a S&',v-cesignec :iucicar Steam Supply System (NGSS), Ocntile flowaters are ucca to ceasure Loop 1 anc I.cco 2 reactor coolant tios races (&i 133J have 2 loops with 2 punpo eacn). Tnese primary loop floweters are not callarated prior to installation. Loop 1 rd 2 feedsater ficw rates are reasured witn calibrated f1cuccters and a plant heat calance is used to calibrate the Centile ficmtera.
Ihe tctal reactor coolant flow rate for Davis-Eesse, :Jnit 1, as cetermined froa a plant neat calance, is 113.2 acreent of tne casi;n " aw rate. Based on the accuracies of primary and seccreary side ceasurcents recorted in l'able 1, tne licensee calculated tne reactor coolant flow rata accuracy to
- .e 1 2.2.cercent.
Heasurecent accuracies fer primary and secoraary side measurements used for calculation of reactor coolant flow rate are shown in Tacia 1.
E.xecpt for tne pressure uncertainty and flow AP uncertainty, these values are reasen-aole and consistent with industry practice. The ucat significant ter is in calculating accurate values of reactor low rate are reactor coolant to::rer-atures and feedwater flowaeter oifferential prescuras.
'Ibe seasurement accuracy reported for reactor coolant oressure isIO.77 percent; however, staff experience indicates atl percent is :: ore reasonacle.
Tne change totl percent pressure ceasurecent accuracy coes not affect tne final reactor coolant ficw accuracy as given to 3 significant digits.
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A_ a TABLE 1 ACCUP/J.'.Y OF PRIMR? MD GECCiiCARY SIDE ;EUUPOCCS USED POR CMULTTICt! CF TOIAL FC FLCliPATE lL\\SUICEC ACCUMCY PA.%' m ACCURACY - PEFCD E SPM CUITS a
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t 0.79 520 to 620F i- 0.79 F FC cold leg teco, r 0.79 520 to 620F T u.79 ?
Stea:a temp.
t 0.60 0 to 7002 2 4.2 F Feedwater teap I 1.13 0 to 600f I 6.o F o
Feed. tater pressure i1.0 0 to 15JJ : Gig I13 ;.si Stec:a pressure
- - 1.89 0 to 1200 psig I 23 psi EC pressure t 0.77 0 to 25J0 psig Ilspci
?cedwater Flow I 1.25 0 to 'Jod incnes 212. in=c3 (Std. 3 0) 2 4
N Flcurate 1.046 0 to 910 incnes
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The measure 0ent accuracy reported for reactor coolant flow rate AP
(!1.046 perc0nt) is for the 4P transmitter only. It is our position that a orift allowance for the flow element (Gentile tubo) is also needed.
Therefore, the staff has reevaluated the reacter coolant flow measure:aent accuracy, using a value of 12 percent for the reactor coolant flow rate AP reasurement. We effect of tnis change is to increase the total flow rate measurecent accuracy from.12.2 percent to 2.5 percent.
An important element in the error analysis is tne assumd independence of the uncertainties in measurement of feedwater flow fo( the two loops. The major potential source of dependency for the feedwater flow =easurement uncertainties is crud buildup in tne flow elements. Altnougn crua cuildup has been observed in the feecwater venturi's for at least one reactor vendor, tne once-through stea:2 generator feedwater chemistry control minimizes the increase of contaminants into the system ard tne cuildup of crud on tne flow elecents for Davis-Sesse, Unit 1.
Therefore, it is reasonable to ascume that tne feedwater flow measurement amuracies are indepere.:ent.
Flow requirements given in Taolo 3.2-1 of tne proxsed Tecnnical Specification revision, as provided in Fcvision 1 to SAW-1469, includea a ocaaurement un-certainty of'C2.2 percent factored into the 110 percent design flow required for potential rod cov effects and increasec cypass flow. Based on our deter-
- ination tnat the measuretent accuracy is22.5 percent, the Tecnnical Specification, Taole 3.2.1, nas been revised to reGect the increase in total ficw rate :easurerent accuracy frcn 12.2 percent to 3.5 percent.
Based on our calculations of bypass flow througn the guide tutes witn tne BPPAs and CFA3 recoveo, we have determined that an increase in tne reacter vessel flow of 5 percent is sufficient to corxensate for the increascu by:sas flow.
Also, we nave reviewed and evaluated the Toledo Edison Corrany's request for balancing tno Fuel Bod Sowing Penalty of 11.2 percent by taking creait for:
(1) a 1 pcreent DNBR credit for tne Flow Area aduction Sacter (as descrimi in Section 4.4 of the Davis-Sesse, Unit 1 Final Safety Analysis F.eacrt); (2) a 1.1 percent credit for the TJ2SR Power Spike factor (as descri>eci in agcreved Topical Report DAa-1401) r and a 9.8 percent DU3R credit for the effects of a 5 percent increase in reactor coolant flow.
We have reviewed each of these credit-m find them acceptacle. Also, we have determined that the revised Tecnnical Specification on CSSR Margin with tne allowed measurement uncertainty of 2.3 percent provides assurance that the treatment of the INGR penalty required to address the effact of
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fuel rod cowing, as specified in License Condition 2.0.(3)(1) is accept-acle. Therefore, we find that License Conditien 2.C.(3)(1) is no longer necessary and can te removed from racility License NPP-3.
.;e have also reviewed tne rodifieo crifice rod asseculy (eDM) for accept-aaility. A ICRA is a standard CFA codified for use witn a primary neutron source. During tne initial core operation of Davis-Pesse, Unit 1, two primary neutron sources are located in individual guide tubes of two fuel asserolics. Each source is held in a snroud tube unich rests on tne bottom of a guide tuco. A solid stainless steel rod is placed on top of tne source to hold it down against hydraulic lif t.
To provide further assurance tnat the source will not com out of the guide tuce during costulated accicents, a:xl CEA is latened to tne top of the fuel asseccly. The rods of tne GM plug tne too of eaca guide tune, inclucing tne guide tume containing the Ocurce.
To prevent the LCHA frcra causing wear of tne fuel assedly end fitting ara coming loose, the Toledo coison Cocrony prooosed to redify the primary source capturing arrangement. tiret,12 of tne rods in eacn of tne two CFA: remaining in the core are being removed, leaving caly the red aaove tne source and the 3 syraetrically-locatcd rods. 5ccendly, a retainer is to ce placed over the huo of the codified CSA and held down ny the reactor internals.
One design and testirr3 of this retainer cevice are described in tne Beacccx and Wilcox apport, DAW-149ti, "SPRA Retainer Design Report," ilay,1978.
From a rechanical design stand:cint, the casic concern is whether tne retainer provices enougn aciddown force to preclude loosening of tne HC3As.
From analyses of the static ard cynanic strasocs on the ratainer spring load arm and housing, recults of prototype testing in a flow test facility, and in-air mechanical tests, criteria for use of the ePRA retainer device witn rodified CFAs have been estaolished. The crimary criterion is enat the surgin to cor@cnent litt witn the retainer, taxing intJ account the
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hydraulic forces acting cn the ICBA, the !GA weight, and the retainer holdccwn tcree, should ce greater enan 30 pounds. This criterien is met with acceptanlo cargin by tne fact tnat ten the retainer device is used with tne acdificd CFA, the holddown force is greater than 35 pounds with all 4 reactor coolant gaps operating. A sccend critericn, which is related to fuel assembly growth, is based on a fuel assecoly burnuo design value that is used as a basis for the retainer design. Since the caxinm curnup used in one cycle of ooeration will be Icss than the burnup used as a design casis, tne fuel asseccly growth critorion is.et (note tnat the retainer will be used tor only one cycle of operation).
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__3 Altnough failure is considered unlikely, the potential consequences of a retainer failure have also been addresseo in'a letter froa J. Taylor (B&W) to S. Varga (NBC), dated June 7,1978. Se neutronic and tnermal-hydraulic concequences are censidered insignificant. Interference wita control rod rction, for exa= ole, would not, acecrding to analyses of stuck-out centrol red transients for B&W 177-FA plants, prevent sats snutdown of tne plant.
The major concern asscciated with retainer failure is plant dacago aru potential outages for repair. This darage caoulu ce precluded by the Loose Parts Sonitoring System (LRG). me Las is cesignea to uetect a f aileo retainer in eitner the reactor vescel or stca;a generator. Cven tncugn tne SPM retainer is designed fer only one cycle of operation, imi nas stated tnat it will rccc: rand that curveillance insoections ce code t'o11 cuing retair.er use. 3is should crovide additional confirration of acceptaale operation. c4W has also stated tnat cefinite plans regarding curvelliance will ce provided to N:C as they are formulated.
In s'.: ration, we conclude that, based on (1) analyses and test results on the BPM retainer device, (2) estaclisnment and meeting of criteria for use of tne device with oms ccaified for use with pricacy neutroa sourcea in Davis-Lesce, Unit 1, (3) analyses wnica indicate that failur2 of tne retainers, hcwver unlikely, uculd not prevent clant safe snutccun ana ( di failure detecticn capacility of tne Icose Parts ::enitoring System, enere is reasonacle acsurance that tne preposed use of the BPRA retainer witn two MORAs in Cavis-desse, Unit i vill pose no significant safety cercern.
Cecause of the modification of core loading, scme enar.ges :nvc ceen raue in ;o.er districations in eno core. Tnece enanges necessitato enanges in-the tecnnical cpecifications. Furtner enanges are necessitatec cy the re-pr:gra:riing of the rod grours, s
n.e new technical acccifications nave ceen estaclisncu, using procecurec
- hich nave been previously e @ loyed. New safety limits (Spec. 2.1.2) and Trip Setpoints (Fig. 2.2-1) and Allowable Values (hg. 2.2-2) :'. ave seen cpecified..iew red insertion limits (Spec. 3.1.2.d) nave coen specified along witn new axial irbalance limits (3;ec. 3.2.1) to ansure that peaking factor limits used as input to the LOCA-CCC3 analysis are not exceeded.
Ce rod program deceription has been changed (Scec. 3.1.7) to reflect tne codification in group assignments. We maximum l::cration cacacility re-quirements (page L3/41-2) has been cht.nged to reflect tne reactivity enanges resulting from the removal of tne EPRAs and the relocaticn of the fuel asserolics.
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a me procedurea used to estaclish the technical specifications on power distribution limits have been previously reviewal and approved. 3r. sed on this review and approval, we find the technical specification changes oescribed above to De acceptacle.
A furtner technical suecification change, unrelated to the core ccdifica-tion, was requested in Toledo Edison comoany's letter of.9ay 18,1978.
Inis request concerns tne modification of alarm set;oints on quadrant tilt to M =----hte a recently-discovered increase in the wasurement error associated with this quantity. Tne original uncertainty evaluation was performed in 1974, based on data obtained from prototype detectors.
Cbeervations of ancmalies in operating reactors lea to the reevaluation of this error. B&W has subsitted (letter, Taylor to Peid, dated May 11, 197d) a dccument describing the metneds used to perfor:s the statistical analysis of tne uncertainties and giving revised cuadrant tilt alara :et-points for Davis-Gesse, Unit 1.
We nave reviewed the dccument and cenclude that tne analysis method is acceptable. We have not reviewed tne cata base used to cotain numerical results but we know of no data that sculd make the application of the atnod to Davis-Besse, Unit I noncenserv-ative.
%, therefcre, fina the revised alarm setpoints on quadrant tilt to oe acceptable.
2e Toledo Edison Cor@any, in their subaittals of May 18 and May 26, 1976, stated tnat, atter coepletion of tne core modifications, startuo tests will be performed to assure that the various payaics parameters are bounded oy tacce in the Final Safety Analysis Report for Davis-2 case, Unit 1.
Iests vill to periorsed on roc drop times, critical borcn concentration, tert:er-ature coefficients, control rod worths, power districuticns, anc power coefficients. Succassful eccpletion of tests at each cower icvel will ce required before pecceecing to tne next higner pcscr level.
We reviewtd tne icw power pnysics tests and startup tests propossd oy tre Toleco Edison Ccccany and requested that caditional tests be cc:cleted.
Sy letters dated June 8 and June 13, 1970, tne Toledo Edison Ccccany cotuitted to the additional low pcwcr pnysics tests and startup tests wnich we requested.
Bassa on the use of approved calculational methocis, and on the sugmented low power physics tents and startup tests, whien we have reviewed and found acceptable, we find tne analysis of the physics parameters of tne core modification to be acceptable. We have also determined that tne actual reactor coolant system flow exceeds the design flow by an a: cunt sufficient to not only cocpensate for the increased bypass f1cw due to the removal of all the aPRAs and all but two of the &As, but also, that the excess flow is sufficient to C+ - h e the thermal margin t
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, required to address the rod bow effects on the decarture from nucleate
'ie have also determined that the two modified ORAs to boiliry ratio.
be used at Davis-cesse, Unit I will pose no significant safety concern.
In addition, our review and evaluation nad detarcined that revised limits necessary for safe operation have oeen incerocrated in tne revised Tecnnical Specification. Tnerefore, we find that Davis-Bessa, Unit I can be operated safely for tne duration of Cycle No. I without BPPAs and CHAs at the rated core power level of 2772 Megawatts-thermal.
CNIICELOTIAL, COtiSIDEPATIGI We have determined tnat the a;,ddr.st does not autnor1::e a change in effluent types or total aocunts nor an increase in power level and will nct result in any significant environmental ircact. iiaving made taia Getermination, we have further concluced tnat tne a enctwne involves an action whien is insignificant from tne standpoint of environmental iacact, and pursuant to 10 CFR Sl.5(d)(4) tnat an environraental ir: pact statement, negative declaraticn and environ.nental fenact appraisal need not be cre-pared in connection with the issuance of this a.mnc ent.
CG CIUSIC43 64e nave concluded, based on the consicerations discussed accve, that (1) cecause the amencment does not involve a significant increase in tne pr0cacility or concequences of accidents previcumly considered or a signi-ticant decrease in any safety margin, it does not involve a significant na::ards conaiceratien, (2) there is reasonacle assurance tnat ene nealth and satety of the puolic will not ce encangered by operation in tne pro-poseo manner, and (3) sucn activities will te concucted in compliance with the cc:nission's regulations and tne issuance of tnis anendcunt will not oc inimical to the cc:anca uefense and security or to the ncaltri and safety of tne a.:olic. Also, we reaffirm our conclusions as otherwise stated in our Jafety Evaluation Eepcet.
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