ML20037B182
| ML20037B182 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 09/20/1977 |
| From: | Desiree Davis Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUDOCS 8009080561 | |
| Download: ML20037B182 (2) | |
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UNI DSTATESNUCLEARREGULATORYCOMMfSSION DOCKET NO. 50 COMMONWEALTH EDISON COMPANY NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE The U. S. Nuclear Regulatory.Comission~ (the Commission 1 has issued Amendment No. 21 to Facility Operating License No. DPR-2, issued to the Comonwealth Edison Company (the' licensee), which revised Technical Specifications for operation of Unit 1 of Dresden Nuclear. Power Station-(the facility) located in Grundy County, Illinois. The license amendment is effective as of its date of issuance.-
The amendment (1) authorized operation of the facility with.
additional 6 x 6 fuel assemblies as replacement for some of the existing
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fuel assemblies, (2) incorport.i.ed revised Minimum Critical Heat Flux Ratio limits that assure conservative operation with respect to thermal hydraulics during Cycle 11, and (3) incorporated new Maximum Average Plarar Linear Heat Generation Rate limits to assure that the reactor is '
operated so as to continue to meet the emergency core cooling system performance criteria.
The application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations.
The Comission has made 1
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. appropriate findings as required by the Act and the Comission's rules and regulations in 10 CFR Chapter I, which are set.forth in the license amendment.
Notice ~of Proposed I*.,suance of Amendment to Facility Operating License in connection with this action was published in the Federal Recister on June 23,1977 (42 F.R. 31845). No request for a hearing or petition for leave to intervene was filed following notice of the proposed action.
The Commission has determined that the issuance of this amendment will not result in any significant environmental impact and that pursuant to 10 CFR 151.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
For further details with respect to this action, see (1) the application for amendment dated April 11, 1977, and supplements thereto dated May 23, August 3, and September 8 and 15,1977, (2) Amendment No. 21 to License No. DPR-2, and (3) the Comission's concurrently issued related Safety Evaluation. All of these items are available for public inspection
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at the Comission's Public Document Room,1717 H Street, N. W., Washington, D. C.
20555. A single copy of items (2) and (3) may be obtained upon request addressed to the U. S. Nuclear Regulatory Comission, Washington, D. C.
20555, Attention:
Director, Division of Operating Reactors.
Dated at Bethesda, Maryland, this 20th day of September,1977.
THE NUC R REGULATORY COMMISSION
.v Don K. Davis, Acting Chief Operating Reactors Branch f2 Division of Operating Reactors
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.sinilarity to the previous fuel, which was accepted and which-has demonstrated acceptable perfornance for several core cycles,'and based -
on acceptable _perfornance of Exxon Nuclear Company fuel of similar design ~ in other reactor.s, v.e find the mechanical design of Type XN 71 fuel acceptable, dE
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2.3 Thernal-Hydraulic Desinn
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ex The' licensee's application for reloadN stated.that "no new (transient) or accident) analyses are indicated," based on sinilarity of the Cycle 11 core to previous cores that have been used successfully in Dresden Unit No. 1.
However, it was the IRC staff judg1hent that the licensee should
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present new transient and ' accident calculaiions applicable to Cycle 11 and based on an acceptable critical-heat-flux (CHF) correlation, such as
-2.
The licensee responded with a commitaent to provide such analyses by ovember 1,1977.
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-lL SJ 4Lf-fic mEme existing anadydes were performed assuning
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/ previous cycle, using analysis techniques different from those currently i 3
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employed by most other licensees (for example, equating a steady-state' f
overpower reactor state to a transient reactor state) and using the
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\\ Janssen-Levy critical heat flux correlation which hds been shown to be
_A hon-conservative for non-uniforn axial power shapes. -
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A.f ter several discussions between the licensee and the HRC staff, ar-altern +e method was proposed to co,nservatively bound thornal hydraulic -
perfo..nce of the Cycle 11 core during postulated transients and.
accidentsf That alternate method and the bases for its acceptability aFe presented in the following Jser'o n3%._ _ -
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i & Mm= Jhe new transient and acci nt analyses will be performed'.and submitteo
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iand operating restrictions alic.cd tc ':: %h Tent h thejconservative
- 7.. g QA ' approach described below c n be relieved following -revicw and approval
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'of the new analyses. The analyses will also be used as a conservative /
Q fWI / j" reference cycle" analysis which can be utilized by the licensee for 14 d,if j future relo,ad en
~)Jpw2_. 3.~)Interin Safety Linit MCHFR 1
s The liccnm has wrfomo a nrT:e of semit' vite studies crw 1:
Minirmj,ri t.ici um ri x sc do i.Ci.R ) d.. e s ch ts em nej oy -iu y
Janssen-Levy correlation to Mininun Critical Potter Ratio WCP2) 'uluss de nr Nket ty :ne :D-2 corr el ati rn.
Inese sidies are :.4
, cc.:-
paring stedy-state reactor ccn:;tiens nith dif ferent flcus, xuer
, levels, axial power sM,ss, cad different inlet enthalpics (i.e.,
tenpsrcture and voic content).
For each such reactor conci; ion, bGtn the MCHFR and MCPR values were calculated.
Using the comparative resul.ts, the licensee has determined the acst conservative credible
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" conversion factor" and has used that " conversion factor" to'show that if the liCHFR for Dresden Unit fjo. I as determined by the Janssen-Levy correlation does not go belou a value of.l.9, then it conservatively assures that the liCPR for Dresden Unit !!o. I as determined by the.
XI!-2 correlation does'not go below a value of 1.32.
The value of -1.32 -
Er (determined by Xil-2)Lhas been found by the staff to-be an acceptable safety limit 11CPR for. other BUR Technical Specification safety limits.
Therefore, a Janssen-Levy-!1CHFR value of 1.9 is an acceptable limit for C
Dresden Unit No.1.
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2.3.3 Interin Operating Limit IEHFR
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To insure operation of the plant above the safety l'init !1CHFR, it' is necessary to determine the maxinua change in t1CHFR MCHFR) that could b^
occur during an anticipated transient.
Adding tha FR to the safety
- -[i linit MCHFR and requiring the plant to operate;ab he. resul ting sum (called the operating limit liCHFR) will sssure tha+> *"
fuel will not go below the safety limit tiCHFR during a transientN g/
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In order to determine the maximun credibic-MCHFR for Dresden Unit flo.1,
. =i the licensee and the staff reviewed transient analyses of several. other BWR's that had used the Janssen-Levy correlation to. determine the worst
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case llCHFR.. In the ~ Turbine TripLWithout Bypass transient, which is
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believed to be the limiting transient for Dresden Unit !!o.1, the largest _
AMCHFR in the analyses of other plants was about 0.50..
However, to e
assure conservatism, the licensee and the staff established the largest '
/>1CHFR for all transients in other BUR's, not only the one believed to be the.liniting transient for Dresden Unit !!o.1.
It was'found:that
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this largest value of AMCHFR is 0.59 for the overpower trip.
This
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aliCHFR was then added to the safety limit I CHFR for Cresden Unit Mo.1.
To account for any plant differences between the Dresden Unit and the other DUR's surveyed, or for inaccuracies in the analyses-used to cal-culate the safety limit MCHFR or the 0.59 atiCHFR, a nargin' of 0.31
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was added to the A!!CHFR.
The _ total 'lCHFR operating liait is therefore 1-2.80, which the IJRC staff agrees is an acceptable, conservative
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operating limit itCHFR for Dresden Un, t lio.1.
' en analyses to be.
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confira the conservatism in this a
approach and provide a technical basis for es ablishing a less coascrvative oparzting limit.
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2.
Accicant Analysis N
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- 2..i ECCS t.ccendix :. An21ysis
. On December P7,1974, the Atoaic Energy Comission isrund an Crder for
!adification of License inplcuenting the requircr.cnts of 10.CFR EU.46 m._ T
" Acceptance Criteria and Emergency Core Cooling Systeas for Light
. lfater fluclear Pcuer Reactors." One of the requirements of: the Order '
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5' was that, prior.'to any license amendment authorizing any core r'eloading, A
"...the licensee shall submit a re-evaluation of ECCS cooling performance calculated in accordance with an acceptable evaluation model which conforms to the provisions of 10 CFR Part 50.46."
The.0rder also required tnat the g
evaluation should be acconpanied by such proposed changes in Technical ~
Specifications or license-anendments as nay be necessary to implement the evaluation results.
By a !!emorandua an'd Order issued *by the 11RC on August 21,1975,N d
Commonwealth Edison Company was granted 'an exenstion, in force until December 31,1977, "from the requirements of and underlying 10 CFR 50.46 with respect to the design and diversity of energency systems.or the
.p diversity of energency power sources (for Dresden Unit !!o.1) 'but not,frca e
the specific performance requirements of the FAC. Credit was given fg p"
certain equipment and for offsite power availability.
In the Order, the Connission noted that " Commonwealth Edison submitted on !!ovenber 1, 1974, a preliminary evaluation of the reactor's abi'ity to comply uith the FAC, not necessarily including all detail and documentation called for by Appendix K, but nevertheless based on conservative assunptions anc provioing a conservative assessnent of ECCS performance."
P,ecent staff concerns regarding stean effects on core spray distribution were examined to assure that operation during Cycle 11 would.be.within ECCS perforaance requirements of 10 CFR Part 50.46.
The staff recueste j'
that Connonwealth Edison define a " spray distribution that conservatively
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accounts for the steas atnosphere that would exist in the Dre: den-1, reactor vessel following a postulated -LOCA" and further, '_' Based on this spray distribution and conservatively determined heat transfer coefficients.
k calculate the maximum allevable bundle-power ana linear heat generation rate F
that will assure that the performance requirenents of Paragraph 50.46 of p
10 CFR are not exceeded."
y i.i In response, CECO referenced spray flou distribution calculations ncde U
previously that were based cn single nozzle tests in~ air. -The air tests E
were used to determine an individual nozzle's spray distribution. The p
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conbined effect (flow) from all. nozzles'uas then determined for ecch bundle, Dased on geometric calculations.
For each of three concentric-core reciens, CECO selected the aininco flew predictec for any sinole j'
buncle a w c'ivided thn nir.inun fic.. Dy a f actor of b0 to Acccen '.r ef fecr.s c" = ne:a envirar ent thn
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ceter.,in:d a tM 0:
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This procecure was utilizeo to proouce a conservative precictior, cf tae mininum sray flou available to any buq<ile in each of me three ccre
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regions, considering effects of the stean environment., The 1!RC' staff H
agrees with CECO that a factor of two reductio'n to acccunt for steam effects is conservative for this purpose.
Dresdegnit No.1 utilizes 1/2H30 n0zzles which proauce a large droplet size Large dreplets have been shown gee affected less by a steam environment than' smaller drop-Z let sprays.
Based on for similar type nozzles,ggle nozzle-test results in air and-steam
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we believe the factor of tuo reduction J.E -
conservativcly accounts for steam effects on core spray distribution for
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U 0**'" A CECO then determined an assembly naxinun, power for each of the-three regions such that the min 5 mum spray flow' predicted as above for that re') ion (considering steam effects) would provide adequate cooling (i.e.,
so that the spray cooling coefficients / justified).
This was done by
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exanining numerous reports concerning Full Length Emergency Cooling Heat Transfer (FLECHT) experiments, and examining minimum spray flows
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conservctively predicted to be present_in other 8:!R's of various types with the corresponding spray cooling coefficients assumed for those reactors.
It was noted that a certain " vaporization flou" can ne defined for each fuel assembly in the reactor or in the FLECHT tests, ii.'
such that vaporization of_ that amount of water will exactly remove the
, JJ total anount of heat being produced in the bundle at the earliest calcu-lated time of spray initiation follcuing a postulated LOCA.
It was further noted fron the FLECHT tests and from conditions present in other q
operating Bt!R's that if the mininum spray flow available to a fuel.
bundle (considering steam effects) is a factor somewhere between 1.3 l
and 2.0 above the above defined " vaporization _ flow," then the sprey cooling coefficients assuned in the ECCS-LOCA calculations are con-i:
servatively justified.
CECO determined a maxinum bundle power for each region such that the minimua predicted spray for any bundle.in the region (considering steam offects) is a factor of two above the " vaporization
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fl ow. " The net effect is an approxinate 30*, derate in the naxinun power allouable from the highest power bundle in the cost affected region.
e note tnat the net effect-of the procedure descriud in this paregraph
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l and in the prece/cing paragraph resuits in a factor of four bet.cen the nininun flow to any bundle (predicted without consideration of stean effects) and the " vaporization flow" of the highest power bundle.
'his cens9rvatively acccunts for steam ef fects. is accept 6ble.
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TR - = w "a nilouabic pm.r feoa any o'. mum in Uch "
the $rea corc regions (determined as.cescrioca aDove) uill ce incor +.' rated.i;g0 the
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- w. chic tW h4)o 14,'k W' t Technical Specifications, and pe iodic surveillance wil.1 be required to insure that these regional bun e power limits are not exceeded.
On the above basis, se conc d tpat when ee Dresden Unit No.1 is.
tWe(il i mi t s7 i = 1 u 97" 9ME3hdjli AMise r H Ud operated in accordance wit
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-e-vake-s-contained -in th -revised--Techn4 col-Spec-ific+t-ions 8 peWodanceO '
requirenents of' the FAC 4 )L act, and therefore the original basis for granting thc excaption is,still valic.
Dresden Unit ho. I th-:refere reets all requirements of 10 CFR 50.46-(except for certain of those recuire -
nents from which they are specifically exempted until, Decenber 31, Wi7) and operation of Cycle 11 is therefore acceptable fron loss-of-coolant
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accident considerations within the limitations imposed by that exemption.
L 2.4.2 Steamline Break Accident I-t.!-
Steaaline _ break accidents which are postulated tn occur inside containnent P
are considered in the ECCS analysis discussed in Section 2.4.1.
The analysis of steamline breg)cccidents occurring outsice contain ent were e
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presented by the licensee to support operation of previous cycles.
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Since results of the stea'11ine-break-outside-of-coqtainment analyses.
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would not be changed by the reloaded core, the previcus analyses are acceptabl e. -
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Fuel loading errors are discussed in Refererce 2a for a fresh XN ;
6 x 6 fuel bundle placed in an improper location or rctated 90 cr 100
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degrees in a' location near the center of the core. The infomation
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presented indicates that the worst fuel loading error results in a ainiaun critical power ratio (MCPR, based on the acceptable XM-2 cor-relation) of greater than 1.32, which is an accentable value.
The licensee calculated effects for several other mislocdingshh Wre l_c.c : ::"c~ 'De#: :~:
N-thereby providing the bcsis f,r our concle-sion that the misicadir.g ekror describcd above is the nn t severe-M w bw.
2.4.4 Control Rod Dron Accident rusM y L J. On M % 4 g 4 o
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d' Anclyses of control rod drop eccic;ents were presenttd ny the licens3e' 2l O sL;;' Ort Q2rStinn cf pr9Viaus CyCicS.
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t r, ccnneQU2t.Ces of a cons.rol ros1 Urop acciuent.
We Qrec that W nr2vious analyses are ccceptable.
2.s.5 Frel Handling Accir:ent ts:1 hlyses of tne fuel handling accident were presentec by the licen:me
to support operation of previous cycles.. The resuits of those anelyses n
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