ML17252A552
| ML17252A552 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/11/1982 |
| From: | Rausch T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17194A407 | List: |
| References | |
| NUDOCS 8201190337 | |
| Download: ML17252A552 (19) | |
Text
~.
Commonwealth9ison One First National Plaza, Chicago, Illinois Aqdress Reply to: Post Office Box 767 Chicago. Illinois 60690 January 11; 1982 Mr. Harold R; Denton; Director.
Office of Nuclear Reactor Regulation flECEHfED u;s: Nuclear Regulatory commission Washington; DC 20555 JAN 18 1982B>
us aueim RffiUlATIJRV CllllYISSffil!
il~!:!:tiRff 1.1t.'l.l.5filli1T sa
Subject:
Dresden Station Unit 3 TIDC Proposed Amendment to Appendix Technical Specifications to Support Operation with Fuel Supplied by Exxon Nuclear Company NRC*Oocket* No: *50_;_249* .. - .....
References (a): J~ s; Abel .letter to D; G~
Eisenhut dated February 20~ 1981:
(b): J: S. Abel letter to D. G:
Eisenhut dated March 5, 1981.
Dear Mr; Denton:
Pursuant to 10 CFR 50.59, Commonwealth Edison ~r6poses to amend Appendix A, Technical Specifications~ to Facility Operating License DPR~25 for Dresden Unit 3~ This ameridment is being submitted to allow the use of fuel assemblies designed and manufactured by E~~on Nuclear Company Inc: (~nc) for the ensueing Cycle 8 reload arid future reloads at Dresden Unit 3; Attachment 6 to this letter provides th~ changes proposed to the Technical Specifications* and Bases; A detailed description of ~hese ~hanges; along with a gener~l discus~ion of the Dresden 3 Cycle 8 Reload is provided in Attachment 1:
- These proposed changes have received on:site and Off~site review and approval. tl..,,J': ,
..Si-~
At t a c h men t s 2 , 3 a n d 4 t o t hi s 1 et t e r p r o v i de t he Dr es e n 3 ?~ I N?
1 plant specific reload, transient and LOCA analysis reports- preparedc.i>bl ":
by ENC. Attachment 4 contains information proprietary to the Exxon~ns 1
~
Nuclear Company. As such, it is accompanied by an affidavit /t/S.ic 1 ,.,p (Attachment 5) signed by ENC, the owner of the information; The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission; and addresses with specificity the considerations listed in Paragraph (b) (4) of Section 2.790 nf the Commission's regulations.
Accordingly, it is respectfully requested that the information which is proprietary to Exxon Nuclear Company~ Inc: be withheld from public disclosure in accordance with 10 CFR Section 2.790 of the Commission's regu+ations: Correspondence with respect to t.hr* ~-~0-~r-~-~a-ry~ aspect*s-*s-f--~~-~;:l:s--apFJ/lication for withholding or
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H. R: Denton
- Ja nu a r y 11 ~ 19 8 2 the supporting ENC affidavit should be addressed to c: F: Owsley; Manager of Reload Fuel Licensing; Exxon Nuclear Company; 2101 Horn Rapids Road; p:o: Box 130; Richland, Washington 99352:
In Reference (a)~ Commonwealth Edison provided notification of our intent to operate Dresden Unit 3 with fuel supplied by ENC:
The 10CFR 170 Class III amendment fee of $4;000 for this proposal was provided previously in Reference (b):
Please address any questions you may have to this office.
Three (3) signed copies of this letter with Attachments l; 2, 3 5 and 6 are provided for your use: In addition, six (6) copies of this letter with proprietary Attachment 4 and the affidavit of Attachment 5 are also being provided at* this time: The non-proprietary version of ,Attachment 4 will be transmitted under separate cover.
Very tr~ly yours,
.~c;R' t(c.,c.e~
Thomas J. Rausch Nuclear Licensing Administrator Boiling Water Reactors cc: RIII Inspector ~ Dresden Attachments (1): Dresden 3 Cycle 8 Reload Discussion and Description of Technical Specification Changes:
(2): Dresen .3 Cycle 8 Reload Analysis Report, XN-NF-81~76; Rev. l dated December 1981.
(3): Dresden 3 Cycle 8 Plant Transient Analysis Report, XN-NF-81-78, Rev. l dated December 1981.
(4): Dresden Unit 3 LOCA Analysis Using the Exem Evaluation Model MAPLHGR Results, XN-NFG-8l-75(P) dated November 1981:
(5): Affidavit of James N. Morgan Attesting to the proprietary nat~re of XN-NF-81-75(P)*, dated November 12, 1981:
(6): Proposed Technical Specification Changes to DPR-25.
1 su~s1'¢ffIBED A_Np SWORN to bef.pre _rn~,):hi~ day 0 t:.** . *~ ;, (-. *"' 19 8 2 .
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DRESDEN 3 Cycl~ 8 Arm Description Of Technical Specification Changes Received wth ltr dtd 01/11/82
-*NOTICE -
. THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE DIVISION OF DOCUMENT CONTROL. THEY HAVE BEEN CHARGED TO YOU FOR A LIMITED .TIME PERIOD AND MUST BE RETURNED TO THE RECORDS FACILITY BRANCH 016. PLEASE DO NOT SEND DOCUMENTS CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY PAGE(S) FROM DOCUMENT FOR REPRODUCTION MUST BE REFERRED TO FILE PERSONNEL. . ,,DI
~'"""~.a 0*1~1irH!' _,,.. SO- "2..-r*r Cimtfol #8Z-Olf'l o~37 DEADLINE RETURN DATE liat~ f ~z f!,f f!ncmmrn fifG1.:Ii.J1Wffi' lJGCKH flLE RECORDS FACILITY BRANCH
- 1 ATTA CHM
- : NT l Dresden 3-Cycle.8-Discussion and Description.of. *'tec.nn.ica 1 *S,p~c if i ct-ion *Ch-ange*s
- Dresden 3 Cycle 8 will represent the first reload of a Jet Pump BWR utilizing fuel fabricated oy Exxon Nuclec:r company r ::NC).
Tne following discussion addresses tne fuel de~ign, reloao analyses and Technical Specification cnanges supporting operation of tne D3:8 Reloao utilizing XN-1 Enc fuel. The discussion is Oividej into *four sections as follows:
I. Reloao Fuel ano Core uesign JJ. Transient and Accident Analyses IJl. Tec~ni:al Specifications IV. List of References Se:tions I and II are based on the Dresoen Sta:ion Unit 3 Cycle 8 Reloao Analysis, XN-NF-81-76 !Attachment 2), the Dresden 3 Cycl! 8 Pla~: Transien: Analysis R~port, X~-~F-81-78 'Att~cnment 3) ano the Dresden unit 3 LDCA Analysis MAPLHGR Results, XN~NF-81-75 (Attacnment t). Section !JI oescrioes tne m5~or p:oposec Te:nni:al
-- Specifica:ion changes reouireo for Cycle 8. Sec~ion IV is a lis: of references primarily consisting of EN: Topical Reports on t~eir generic Jet Pump B~R metnooology.
Dres:en 3 Reload XN-1 ~ill consist of 22~ :eloa:
asse~:lies faoricatec oy EN: ano oesi~nated as type X~3)2.f9-5. T~~
co::: loc.:ii*1; wi.J.l consist of tne follo11.:ng:
Fuel Type GE SxS-2.50%
G:: 8x8-2.6'.:':'.;
200 GE PSxBR-2.65~
224 XN l Bx 8- 2. 6 77,;
As shown in Figure 4.2 of Attachment 2, the reloa~ fuei w!~~ De loaced in a 2 out of 4 sca::er with the ex:e~tion of tne c::e axes enc peripheral regions. This loading pattern is similar to that utilize: i~
previous re1oaos.
A. Fu e i ~echanical Design The mechanical design of the reload fuel is -0"2
- ri:-e:
.,r .
generically in Reference l. In general, oes~~ c:ite:ic. c.:~
established to limit tne stress, strain and ov all c~:y o~
tne fuel rod or oundle curing normal and transient operation. In addition, the fuel is designed to De me:nanically compatiole with otner reactor intern~ls. i~e!
handling equipment and existing fuel.
REGULAIDRY DOCKET fllE COPY
A. Fuel Mechanical Design (Cont'd)
The XN-1 fuel design is an BxB array ~ith 63 fuel rJcs (5 containing Gadolinia) ana one water roa. Active fuel lengtn is 145.2~ inch 0 s which incluoes a 6 i~cn blanket of natural u at b:~n top and oottom. Enri:ne:
fuei pellets are aisheo, natural pellets are not. Fuel roa pitch is maintained via seven Zircaloy-4 spa:ers with JncJnel springs. Lo~er tie plates are orillec tJ improve reflooa capability and employe a spring sea~ at the interface with tne cnannel tJ limit co~lant lea~a;e to tne bypass region as a result of channel sice wail ceformatiJn (bulge) ~itn exposute.
- 8. Thermal Hydraulic Design
/
Se:tion 3 of Attacnment 2 icentifies the primary thermal hydraulic desi;n criteria for XN-1 fuel. EN:
has performec testing on full scale assemJl!es to determine tne hydraulic resistances ano press~re crops far XN-1 8xS fuel anc G.~. BxSR fuel. The tes~ :es~:~s verify that XN-1 fuel is thermal hydrauli:ally
- Jm~atiole with G.E. fuel and core tnermal hyorauli:
response is expected to be similar to previous relJa:s.
The Fuel Claoding Integrity Safety ~imit was calc~la:e:
using a Monte Carlo tec~nique to convJlute uncertainties in t~e calculation of =~:e po~er distribution anc critical power ratio. Tne analysis demonstrateo tn:.t a MC:PR S:;fety ~irr:i.: c.f J.C5 ;:.r::-,*i::=s assurance tnat at least 99.9% of the fuel roes in tn~
co r := wJ:.; l c! u e e x p e ct e o t o_ a v o i c :; o i l i no t rans it .'. .J
during steaay state ~peration at thi S~fety ~imit.
Refer to Section 3.7 of ~tta:nment 2, ~ttach~ent 3 an:
Reference 2 for further discussion of t~e ffie:~:::lJ;,
C. Fuel Centerlin~ Melting during Overpo~er Con~:tiJns One of the ENC:'s thermal hydraulic des:gn criteria fJr 03 reloao XN-1 fuel is tnat fuel cesign ano opera:io~
will be sucn that fuel centerline melting is not expected for anticipated operational oc:uren:e5 (transients) tnroughout the life of tne fuel. To
.* demonstrate com;.iliance witn this criteria, ENC_t1as perfo-rmeo transient overpowe:- analyses for a *f~~l roe history rpeak LHGR vs. exposure) whicn repre-s~Jf::s 6 conservative upoer bound on peak rod power ov~W-tne life of tne fuel ou~ole fcatch average discharge ~~:nu:
of 30 GW~/MT). ENC has determine a that the con~itions
e C. Fuel Centerline Melting During Overpower Conditions (Cont'a) for minimum margin to centerline melt over the aoove mentioned fuel rod history occur at a fuel roe expos~r~
of 2l.2GwD/MT, whicn corresponas to a oeak L..HGR (stec::y state, 100% power) of 13.93 kw/ft. At tnese co~Gitio~s.
pea~ fuel centerline temperature was calculate: tc j~
3909°F. To demonstrate margin to ce~terline melt curing transients, power was escalated to l2~~ core power fLHGR=l5.7 Kw/ft) ass~ming neatflux in e~~i~i brium with neutron flux. At the 120% overpo.er conoition, centerline temperature was calculateo to ~e 4607°F, aemonstrating adequate margin to tne !'02 melting point at this exposure of a~out 49G0°F.
Since this calculation represents the most limiting point with respect to centerline melt tnroughout t~e life of t:-,e fuel, it is co:-icluoed t:-ic;t rPc:r,;;:n t:i centerline melt is assured for overpo~er con:itions tnrou~nout tne fuel lifetime.
e Since the fuel rod history assume: i' t:1e. a::i::,ve analysis is not !nfor:eo oy tne Tec~ni:al Specification, an upper bouno on the allowa~!e pea~
LHGR (at full power) over the life of the fuel ~as determined by multiplying the propJse: ~~c_HG? li~its for X~-1 fuel oy a conservative value for tne 1J:al peaking factor. Assuming tnis peak LHSR as a~ initia:
concition, po~er was again escalated tJ 120~ tJ yielc the maximum transient LHGR ac:-iievajle from full p:~~r operation witnin tne Technical Spe:ifi:atiJ~ ~~P~~~~
l i mi t s . These max i mum LHGK ' s were then com oar e d t :i* : :: :>
LHGR cotresponoin~ to centerline melting !teter~ine: c!*
EN:'s RODEX2 code) for various exposures tnrJug~Jut :~e life of the fuel. Tnis compa!ison snJwed tnat ~ar;i~
to centerline melting during overpower :onoiti:ins initiateu from full p:iwer is assureb tnrou;~out t~e life of the fuel oy Technical Specification M;~~HG~
limits. Therefore, an LHGR Operating Limit nas nJ:
been specified for ENC fuel.
ror transients initiated from reduced power and flJ~.
the peak transient LHGR could potentially exce~~ tnat of transients initiateo from full power, if-tMP*Jwer distribution is excessively peaked. Althougri1ffc:.sti:-.;
restrictions on power distri:>ution sucn as thW=reo ....:e:.:
fl o w MC PR l i mi t s o f Tech n i c a l 5 p e ci f i c at i o n 3 . .9'. k a;-, ::
tne RBM R0d block will limit the peaK LHGR and, e theref0re, the total core peaking at reouced flo~s.
- c. Fuel Centerline Melting During Ove.rpower Conditions (Cont'd) suoplemental protection will be PrJvided by a aaily surviellance on power distrioution. This surveillan:e, performed in accordance ~ith approved station procedures, will insure that the peak LHGR curing reduced flow operation is limited so tnat fuel centerline melting woulo not occur curing a transient term!nating at the 120% flux s:ram. Perfor~an.:e Jf this surveillance as required Dy pro~oseo Te:n!lical Sp e .: i f i c a t i on 4 . l. B . 2 , i p c on j u c ti o !1 v. le. :-. t n e ;:n ::. ;. J s e :. :
MAPLHGR limits for ENC 1.'fuel, will ensure thc.t rr.ar;i:, :J centerline melt is maintained unoer all operating an.:
transient conditions.
D. Nuclear Design Tne X~--1 fuel oesign consists of 63 fuel roes an: J!le water rod. The average a~se~oly enrichme~t is 2,6?~
wt-1i:h includes a six inch natural U i:..ilar1-<et a: b.:;::-, to;::,
and bottom. The average enricnment of the cen:rc.l regio~ (ex:luain; blanket) is 2.87~. Five ourna::.~e pJison rods containin~ a Gd2G3-UD2 rr.ixture are utilized to reduce initial ounole reactivity. Tne .
s~eci.fic neutronic oesign carame:ers anu oin enric~~e~:
o i st r i Du t i o n a r e prov i: e Ci i n Se: ti o n 4 o f t._t t :: : n::1 e ;-; : 2 .
Attac~rnent 2 ~*lso provides tne *r~sults of tne*variJws routine cycle-* specific analyses such as sht..:tCown m:;rgin. core s:a::iility ae:t:y ratio ac,d Stei:-;c:i 1 :....iq...:.:.:
Control System effective~ess.
Fuel St~rage Vault/Pool Criticality - Tec~nical S~e:ification 5.5 req~ires tnt:: t~e keff of t~e s~e~:
fuel pool be -~ .95 and t~at of tne nev. fwe~ st:~a;~
vault < .90 when dry ( < .95 when f~oooec). In NEJE-24011, GE states that tnese c:iteria will 8e re:
for GE facricateo racks if fuel buncle reac:ivi:ie~ t::e lirr.ited to kOO < 1.31 for th~ rack dime:-!sio;1s utilize .:.
in the Dresden spent fuel p~::il and < l.30 for t~i:> re::"
a i me n s i o n s u t i l i z e. d i n t h e n e w f u e l v a u l t , w ;-; er e ... OQ .i s calculated in an infinite array of similar fL.iel in ... _
core configuration (as opposea to the st~~'lQe
- configuration). GE has calculated koo's _fW' tneir f'..Je~
designs ano oe~onstrateo that the crite~i~fps sa::s:ie:
for all GE fuel. ENC has calculated KOO ~r X\-l reloao fuel and for a com~araole GE fuel aJsign one shown that XN-1 fuel is slightly less reactive. Sasec on this comparison and the criteria fro~ NEJ~-24J:l, ~:
is concluded that adequate margin to tne Te:h~ica:
Specification keff limits exists for stora~e cf Y\-:
reload fuel in the vault ano pool (for GE desi;nec racks).
- 5 -
Fuel Storage Vault/Pool Criticaiity (Cont'd)
For the high density fuel storage racks designed o; Nuclear Services Corporation, criticality analyses ~ave Deen performeo for ENC faoricated fuel which oemonstrateo that tne 0.95 Keff require~ent is me:.
Thi~ assures that the XN-1 reload fuel will meet tne .~~
keff Technical Specifi:ation criteria wnen stJre: ~n t~e high censity fuel racKs. Refer to Kin w. WJn~ r~::)*
testimJny dated January 2.i, .i981 for tne P.SL.B ne:r.:.:-;;s on the Dresoen Fuel Pool Mo8ification.
II. TRANSIENTS AND AC:IDENTS A. Anticipateo Operational Occurrences (Transients)
In order to determine oper~ting li~its for D3Ce, EN~ nas consicerea eignt categories of core-wide potential tra~sie,ts tas described in Reference 3) and provi~ea analyses res~lts f:r t n e f o l ..i. o ~ i n g t n ::: e e : r a :-, s i e ~, t s t o o e t e r rr. i n "= t n e t h e r rr. :: : r;, ;, r ; : r1 for D3C8.
G=:: n e r a ~ o r L~ 2 *= R~ .5e r: t i J n wi t r, :J :_j t =* y ;: a s s ' ~ =: v..* ... o =).
- Fee:water :ontrJlie::: Failure '.Fn:~)
- Loss of FeeJwater Heating ':..OF,.,.i..;).
The other core-wioe transients are inhere~t:y non-l~~it~~c :r bJ.Jnceo by one of the above. In aociti-on, t1<>:* i:ical ?ve--::s ... _._
withwrawal Error ano Fuel Loaoing Error, were a~alyzec as oescrioec in Referen:e L ano ceter~ineo to oe nJn-lirri:i,:. ~~?
resulti cf the core-~ide anc local Transients are provioe; !n Attacnments 2 anC 3. lne Generator Loao Reje:tiJn ~i:nou:
Bypass was determined to be the limiting event for D3:E, r e ; u ~ : i ng i n c; A C: P?. o f O. 2 5 w !-; .i cn , w n e 11 : o ~ '::i .i 0 e J \\ :'.. : ~ 1 : -. ? _ . _:
Safety Limit, requires a MCPR operating lirrit of 1.3~ for __ _
fuel types.
Core-Wice Transients Tne plant transient madel used to evaluate t:-ie '-.!=\..,./J3 a0c c":;-
events was ENC's COTRANSA cooe (Reference 3) wnicn incorpJ:::a:es a one-dimensional neutronics mojel to &ccount for shifts i~
axial power snape resulting from rapid pressurization dnd vJi:
c o 11 a p s e . T h e ~ OF wH e v e n t w a s a n a l y z e a w i t h E NC ' -s- TS 6 .~ ~ c o : e (Reference 3) which* uses a point-kinetics neutr6r{JLf,
- m:i*:el si:-::e J
rap:a pressurizc.tion ano voio collapse oo not 'JCCu . fJr t:1is event. Both codes utilize a multi-node steam line ~Joel fJr improved characterization of steam line oynamic oehaviJr.
Core-Wide Transients (Cont'd) uncertainties in input parameters for the LOFWH and FWCr events were assumeo to oe at boundinc values. FJr tne limiting event, LRw/oB, uncertainties in tne inout variaoles were hancleo statistically as oescriced in and Reference 5. Tnis results in a s t ':I t i s t i c a l d i s t r i b u t i Dn o f ~ CPR s , a r r i v e d c t c y convoluting the uncertainty distrioutions of tne inpu:
varia:iles utilizing a Monte CarlD proce:::L..Jre. l 1sin; tr,e mean value anc standard deviations of the~ cc~
o i s t r i b u t i o n , o 6 CPr\ o f O. 2 5 w a s o e t e r ;r. i n e o t a J o u n o 7 ; ;: : f tne possiole outcomes of the event.
Actual scram time data from previcius cycles Dn Dresden 3 was useo to generate the scram time distribution assu~ed in determinino tne rpR distribution for t~e ~~w/o~ event. :n order to assure the applicability of the LRw/oB analysis ta cy:le B operation, compliarice witn tne assu~ec scrai ti~e oistrio~tion must be verified tnrougnout cycl~ e as re~uirej Dy .PrJpc:;seo T.S. 4.3.C.3. Followin~ ea:~ set of full or nalf-core (not) scram testing, com~liance witn tne assume: oistrioution must o~ oemonstratec in 6CCor:3nc~
wi:n the station procedures based upon infJrmatiJn su~:liec by C:r\C:. Jf t:-ie current cycle sc:am speeos cevic.te f:::irr. v,;:-
assumed distribution, an adjustment to the MC?R operc.tin2 limit mc.y be rec;uireo. Tr1e Et*~C su~plied met'locs f:,:
cne:king com~li~n:e and adjusting tne MC?R o~erat!ng li~i:
fif necessary) will oe 1ncor~ordteo in Stati~n ProceJ0:es t*o ensure th? proper MCP~ opercting limit is use:
tnrciug~o~t tne :ycie.
Local Transients As shswn in ~tta:~ment 2, the results of tne Fuel L~c.:in; C:rror ~nd Rod Withcrawal error were oounoed Dy tne LRw/o~
evant ano are therefore non-li~iting. Basec on the ~WE results, the proposeo Tecnnical Specificatio~s roo blo:k monitor setpJint is increased from t~e cu:rent valwe of 107% to 110% to provioe adoitional flexibility in utilizing the allowaole power/flow operatin~ region abJve tne 180~
flow control line. The A CPR for the RWE event with a l~'.:::C:
full flow R6M setpoint is 0.15. Tne CPR for tne misl~~:eo bundle event was 0.16 whicn was larger than the .lli..JIPR calc,ulated for the misoriented bundle case (180':l -'i r o t a t i on ) . Al l
- o f t n es e 6 CPR ' s a ! e l e s s t n a n t1 ~ l i "* i t .:. n ;
value of 0.25 calculated for tne LRw/oB event. "
e
- 7 .
Reduced Flow Operation ENC has reanalyzed the necessary adjustment in the MCPR operating limit for transients at reduced flow. Proposed Tecnni:al S:ecification 3.5.K incorporates E~C genera~e:
MCPR limits for reouced flow operation wnich prote:t tne full flow MCPR Operating limit during Automa~ic Flow Control operation anc the M~PR Safety Limit 'during all flJ~
control mooes. ENC's tecn~ical report describing :ne analyses for reduced flow operation will be suomit:eo by (CECo) for your review in late January, 1982.
ASME Overpressurization Analysis In order to demonstrate compliance with the ASME Code Overoressurization criteria of 110% of oesig~ vessel pressure, tne MSJV closure eve~t with failure of t~e MSJV position scram was analyze~ witn E~:*s COT~Q~Sg coce. Tne maximum pressure ooservec in the an:lysis was 13~6 psiQ or 108~ of reactor vessel cesign pressure. Tne correspondin; steam dome press0re was 1324 psig, for a vessel differential pressure of 22 psi. Tnis in:lJues the effe::s e o f t r1 e AT wS RP T whi c :-l w. a s a s sum e ci t o in i t i a: e a t a no mi n c; l Dre s s c; re set p ,Jin t of l L 4':' ~ s i;. Tr-, e uS 11,~ l :. r..:.: f ::i: p e 2-<
vessel pressure of 1375 psig (110% of design press0reJ is therefore equivalent to a come pressure limit of 1353 ps~;
(1375-22). Tne Technical Specification Sa7ety Limit of 1325 psig is oaseo on dome pressure &n~ tnerefore conservatively assumes a 50 psi vessel dp (1375-1325). T~e propJseo safety limit of l3a5 psig co~e press~re is oasec on a 30 psig vessel dp whicn removes excess conserv~tis~
w~ile contin~ing to 6ouno expected cifferential ~resswre beh~vior, especia!ly when tne lack of forcec flJw im~2se:
by RPT is consi:ered. The choice of 1345 psi :n~s ass~res compliance with the ASME criteria of 1375 psi oeaK vesse_
pressure wnile also maint~ining consisten:y w t~ :ne recently proposec Quad Cities Unit 2 pressure safety li~~:.
E. oost~la~eo Accide~ts In support of D3:8 operation, ENC has reana~yzed the Loss of Coolant Accident (LOCA) to determine MA 0 lHG~ limits for XN-1 fuei ano tne Rod Drop Accident (RD~) to oemons:ra:e compliance ~ith the 280 cal/gm Tecnnical Specificcition limit. Tne results of tnese cinalyses are presenteorin Section 6 of Attachment 2. The methoaolo9y for L.hetRDA a:-ialysis is oescrioed in Reference 1.;. 8no tnat fo r~lt1e '-.'.)'.::L.
1 c:ir1 al y s i s i s pro vi de d i n Reference s 6 th r u l 3. ff e
Loss of Coolant Accident - Reference 6 descrioes ENC's generic jet pump B~R3 LOCA break spectrum analysis whicn defined tne limiting oreak for BW.R 3's to be a dou::ile--ende:
guillotine break in the recirculation piping on the sucti:~
sioe of tne pump. The analysis of tnis event for Dresoe~ 3 is provided in Attachment 4 and summarizea in.Section 6 of Attachment 2. Operation within the MAPLHG? limits of 7s~l~
6.1 (Attachment 2) will ensure that the peak claddin; temoerature remains below 22000F, local Zr-H70 rea:tiJn remains oelow 17% ano core-wide hydJogen pro3ucti:r1 rerr.a.:.:-.::
below 1% fJr tne limiting LOCA event. Tne LDCA a~alysis of Attachment 4 ~as performed for ENC Bx8 relost fuel anc tnerefJre provioe MA~LHG~ limits for EN: f~el onl;. ~s discussed'previously, ENC reload fuel is hycrauli~ally a~:
neutronicaliy compatible w.ith G.E. fuel. Tnerefo:e*, t11e.
existing G.E. LOCA Analysis (Reference lh) anj M~ 0 ~HG~
limits will remain a~~licable during D3:o anc fu:w:e :y:~es witn. GE/ENC mixed cores.
It sn~ulo oe noted tnat E~: MC 0 ~~G? limits are prJvide: as a function of bun~le avera;e exposure as 6p~Jsed to Goe=~
exp~sure for G.E. MAPL~G~ li~its. Tnis is cue to the different methooJlo~y employed by ENC for LOCA analyses.
Axi::.1 ex;:.iosure prJfiles which wou*1G trnse ex~e:te:: curir,;
nor~al operation are input into the calculation of fuel r~:
sto:ed energy, fissiJn gas releas~ and fuel rod hea:u~ frJ~
whi:~ MAPLHG~ limits are derivec. Since conserva:.:.ve axi~:
exoJs~re profiles are inherent in the me:nocolJ: .. ~~=L~:~
limits as a function of assembly ournup will ac~6Ja:elj prJtect the peaK ~ower plane! G.E. ~APLH~Rs will re~~~n a~
a function of nod2l exposure.
Rod Dre~ Accident - tNC's ~ethod)logy for analyzi~g tne ~J:
Jrop Accident (RDA) is described in Reference ~ ano utilizes a generic ~ar~me:ric analysis wnic~ cal:~la:eJ * -
fuel entnalpy rise during 8Dst~lateo RJD's over a wic~
r~n~e of reactor operating varia~les. Cycle specific parameters such as. maximum control rod wJrth, Do~~ler cnefficient, etc. are then applieo to tne p::.rametric results to determine the fuel entnalpy rise. For J;:E, Section 6 of Attachment 2 shows a value of 151 cal/~~ fJ:
the maximum aepositeo fuel rod enthalpy during t~e wors:
c2se postulated RDD. Tnis value is well celJw tne Technical Specification limit of 280 cal/gm. - c
--:i 11fJ I II. TECHNIC~L SP~CI~ICDTICN~
Attachment 6 provides proposed Technical Specificati~n tt changes to support D3C8 ope~ation with ~NC fuel. Tne following sections highlight the majot areas reouirin; revision and iaentifies the associated sections of tne Technical Specifications.
A. GENERAL TnrJugnout the Tecnni:al Specifica:ions and Bases, sections have oeen revised to reflect the appro;:>ria:e ::xxor, M~thciG01ogies ano references ano celete General Ele::ric methoos and references where necessary. Alsci, for eacn revised s~ecification as iaentified below, tne corresponding section of tne bases has ~een reviseo as required.
- 8. LHGR As oescribP.d previously, no ~H~M Operating Li~it is specified for ENC fuel. Operation within the MAPLHGM.
limits ano the power Distribution assumptions Jf the Fuel Design Analyses will protect against fuel centerline melting ourin9 transients initiateo from rateo or less tn~~
rated conoitions.
All Te:hnical SpecificatiJn sectiJ~s referring to L~~~
or FLPJ have oeen revis~O to apply only to c:: fuel.
New specificatiJns nave Deen prJpJsec w~~:h require surveillan:es on E~C fuel :J ensure t~at ~argin tJ centerline meltin:- is mair.:ai:-iec ot.:rin'.J transients initiated frJm an; allciw~Jle reactJr c6ntitions: For adcitiondl informatiJn re;arcing margin tJ ceGterli~e melt, refer to the previous tis:~ssio~ of Se::ion 2.:.
In acoition to tne a~ove, all references tJ 7x7 fuel and tne power soiking penalty nave oeen oeletej since tnere will oe no 7x7 fuel in D3CB.
T.S. Section De s c r i p t i o n
- l. K Definit!on of ~~P: revise:
to a~ply to G~ fuel on~y.
- l.l.A.l/2*.l.8.1 AP~M Sera~ ano Roa Bl:Jc~
3.1.B/i...l.t:l ~quations r~vised to pr~vise T~ole 3.2.3 Note 2 MFLPD/FRP ad~us:mentfor GE fuel cinly. For EN: fuel, a requi:r:::-;e::
to ensure complian~e wit~ tne F~~l Design Analysis has been aoces.
- -[
3.5.J/4.5.J Revised to require LHGR1 -l~it. an:
surveillan:e for GE fue l 1 nly.
Deletes reference to 7x7
- el anc power spiking. *
- Also revised to allow aajustment of APRM gains in lieu of reouciny trip settings.
C. MCPR T.S. Section De scr i pti on 1.1. A MCPR Safety Limit changeo .to l.G5 3.5.K MCP~ LCG changed to 1.30 fJr sll fuel types. Revisec to indicate new curves for determining MC~~
limit during operation at recu:e:
flJw and to require adjustment of the limit if scra~ ti~es fall outside tne ciistrioutiJn assuoe~
in the transient analysis.
Figure 3.5.2 Replaced with new figur~s for oeterminin; MC 0 R li~its duri~;
operation at reducec flJw.
D.. Reactor CoJlant System Pressure Safety ~i~it (Sectio~ 1.2)
As s:ussej earlier, this wi:l be chE~ged .frJ~ 1325 to 134. psig. Previous value assu~ed a vessel pressure d!J~
of 50 psi. The new value is conservative co~pared tJ t~~
a:tual ;ressJre Crop as Geterminec oy analysis.
- t. R~M Setti~c (Table 3.2.3)
Cha:ige::; frorn .:1vd+L:2 (107% at full flow; tci .c.51'>.:*L;.
(l~J~) baseo on results of RW~ analysis.
In acio~tion, Specification 3.2.C.2 is bein: revise~ to clarify RSM operability requirements and to De co~sistent witn Tac.le 3.2.3 r.ote 1 ..
F. Secti:n 3.5.J.3.a T~is section, wnich alloweo operatic~ ~ith o~ly ~ AJS valves during Cycle 6 operation, has been celetec s:.n:.::- i-c is no lJn;e; ap~licaole.
Se:-tion Jes:ri:ti*:Jn 3.5.I/Li.5.I Revised to distin~~isn_tt-ri"t G::_.
- ~'*P1_r.GP.s ::.:-e functions 1tT~*~i,,0:a.:.
- pJsu
- -e :ne!eas EN~ MA~ -=s 6:-e ceoenoent on bunole avero e exposure.
- 11 .
G. MAPLHGR (Cont'd)
Section Descriotion Figure 3.5-1 Added MAD~HGR curve for ENC t1pe XN8~2.69-5 fuel ano deleted c~rves for 7x7 fuel. Tne laoelling on tne curve for fuel ty~e 8J?5265-L nas oeen revised to incicate tnat tne curve apolies to P8~~~:65-~
(Prepressurizec) fuel alsJ.
µ. RPS Delay Time Specification 3.1.A is oeing revised to change the allowaole RPS oelay time from 100 msec. to SO msec to be consistent ~ith tne oelay time usec in ENC's oe:erminis:i:
t:-a:isient analyses (refer to I.E. CircLJlar 80-08). J;-, tne an3lysis of tne Lo2c ~e:e:tion witnJut ~y~ass ~vent, a conservative statistical distribution of tne RPS celay ti~e was 1..:se:, r:.t:-ier tnan a sin;le value c::s in ::1e oeterministic a~proach.
I .
Specification t.3.C.3 has been aao~d to req~ re verific~tion after edch set of s:r:.m timing a:2 tna:
tne ::::urre:-it s:rcr:. s:ieecs fall ... :tr.in tne cis :~:;..;::Jr, assumed in tne transient analyses.
Ol68T
-f 1,- 'ff-
{f
IV. REFERENCES
- 1. XN-NF-81-2l(P), "Generic Design Report-Mechanical Oesi*;in for Exxon Nuclear Jet Pump aw~ Fuel Assern:Jlies."
dated October, 1518i.
- 2. XN-".;F-524(P), "Exx:in Nuclea::: Critical Dowe:: Met;,JoolJ;y for Boiling water Reactors" dated NJverr::Jer, 1:,7:,..
- 3. X'\-N::--79-71 qevisicin l (Supplements l ano 2), "E:xx-:i:;
Nuclear Plant Transien~ Metncis:ilJgy for 8:iil!n; We:t~:
Reacc.ors" oated N::ivemoer, 1981.
- 4. X!\-~~F-80-19(?), Volu~e l (Supc~ements l e;n::; 2), "EoJ-Nuclear Metn:io:ilogy for bJilin; ~ater Reactbrs t'leutronics Metnods for Design and Analysis" bates t-t,ay 1980.
- 5. Xt\-NC"-51-22(P), Septem:ier 1981 Generic Statfs:ical uncertain~y Analysis Methol:Jlo;y
- 6. Xr-.; - r--:=- - El - 7 l ( P) , Oct J ::i e::: i 9 51 Generic Jet-Pump 8W~3 LOCA Analysis ~sin~ tne E~: E:XE:M Eval~atiJn ~ooel.
- 7. X~-r,::--:l-7::, "Jresce~ 1_,:-iit 3 :...J:.: r.-.e:.lysis ::s:>.;
ENC EXEM Evaluatio:-i ~Jcel-MAD~~:~ Fiesc.:lts" ce:te:
Oct::ioer, 1981 r ;..tta:r,;:-,ent L.:).
E. Xt~-~~=- .s*c:-l9:F), V=1lurne 2. Revisi~n l, J~ne ~;:~~
Exxc:-i Nuclear MethJCJlo~y for 8oiling ~ater Rea:tJrs t:x::r.J,: ::*ccs ~v2luc~i:ir1 M:;ce.i, S;_;mr:-1::/ Jes::r:.;;~i:i;a
- 9. XN-N::--SO-l9(P), Volume 2A, Revisicin 1, June l98l ExxJn Nuclear MetnocJlciov far 2Jilina water R?act::-;
FC:LCi.X: A t:..~:...APLi Baseo 6ii-,;:iuter CoJe fo:r [c; .. :u~a-.:.i:-1; BlowJJw:-i P~enornena iO. X~---l\~-80-l9(P), V::ilume 23, Revision l, Ju:-ie l5Sl ExxJn Nuclea: Metnooo:ogy fo::: 6Jiling wa~er Rec;:to:s FLEX: A Computer Cooe for Jet Fump 8~~ Refill ant Ref~ood Analysis
- 11. Xf\-!,F-80-19(P), valume 2C, J-.;ne l?o~
- Exxon Ni..;clear Metnooology for Boiling water Re_¥to:-s verification and Qualification of EXE~ -I - fot
- 12. XN-CC-33(A), Revision 1, ~8vernjer 1975 I ft HUXY: A Generalized Multirod Heatup Code with lfo:F~5:J Appendix K Heatup O;:ition
REFERENCES (Cont'd)
- 13. X~-NF-8l-58(P) t August 1981 ROJEX2 Fuel Rod Thermal-Mecnanical Response Evaluat!J~
M;:idel
- 14. Neoo--2Lil46A Revison 1, "Loss of Coolant Accident Analyses-Quad Cities 1/2, Dresden 2/3" oated A;::iril 1?7~.
_r
- *t I ~H t
A~S AF F I DAV I T STATE OF Washington ss.
COUNTY OF Benton I, James N. Morgan, being duly sworn, hereby say and depose:
- 1. am Manager,, Licensing and Safety Engineering, for Exxon Nuclear Company, Inc. ("ENC"), and as such I am authorized to execute this Affidavit.
- 2. am familiar with ENC's detailed document control system arid.
policies which gove~n the protection and control of.information.
- 3. I am familiar with the document XN-tff-2*1-75(P), entitled "Dresden Unit 3 LOCA Analysis Using the ENC. EXEM Evaluation Model - MAPLHGR Results," referred to as "Document". Ir.forrr,ation contained ir. this Document has been classified by ENC as proprietary in accordance with the control system and policies established by ENC for the control and pl"otection of
.information.
- 4. The Document cont~ins information of a proprietary and con-fidential nature and is of the type customarily held in c6nfidence by ENC and not made available to the public. Based on my experience, I am aware that other companies regard informa,tion of the kind contained in the Document as b'eing proprietary and confidential. r
-~:i
- 5. The Document has been made availatle to the UnHdf States Nuclear Regulatory Cormiission in confidence, with the request tliat the tf information contained in the Document not be disclosed or divulged.
2
- 6. The Document contains information which is vital to a com-petitive advantage of ENC and would be helpful to competitors of ENC when
~ompeting with ENC.
- 7. The information contained in the Document is considered to be proprietary by ENC because it reveals certain distinguishing aspects of ECCS analytical methods which secure competitive economic advantage to ENC for fuel design optimization and improved marketability, and includes information utilized by ENC in its business which affords ENC an opportunity to obtain a competitive advantage over its competitors who do not or may no.t know or use the information contained in the Document.
- 8. The disclosure of the proprietary information contained in the Document to a competitor would permit the competitor to reduce its expenditure of money and manpower and to improve its competitive position by giving it extremely valuable insights .into reactor core operating characteristics, and would result in substantial harm to the competitive position of ENC.
- 9. The Document contains proprietary information which is hel~ in confidence by ENC and is not available in public sources.
- 10. In accordance with ENC's policies governing the protection and control of information, proprietary information contained in the Document has been made available, on a limited basis, to others outside me only as required and under suitable agreement providing for non-disclosure and limited use of the information.
- 11. ENC policy requires that proprietary information be kept in a
- . r
~ecured file or area and distribu:ed on a need-to-know basis. --:t i,H tt
- 3
- 12. This Document provides information which reveals ECcs' analytical methods developed by ENC over the past several years. ENC has invested hundreds of thousands of dollars and many man-years of effort in developing the analysis methods and calculating the results revealed in the Document. Assuming a competitor had available the same background data and incentives as ENC, the competitor might, at a minimum, develop the information for the same expenditure 0f manpower and money as ENC.
- 13. Based on my experience in the industry, I do not believe that the background data and incentives of ENC's c;ompetitors are sufficiently similar to the corresponding background data and incentives of ENC to reasonably expect such competitors would be in a position to duplicate ENC's proprietary information contained in the Document.
TH.!\T the statements made hereinabove are, to the best of my knowledge, information, and belief, truthful and complete.
FURTHER AFFIANT SAYETH NOT.
SWORN TO AND SuBSCRIBED before me this ;u;::t day of Yr,,'('.~J. 19 ~ 1.