ML17194B410

From kanterella
Jump to navigation Jump to search
Application to Amend License DPR-25,revising Tech Specs to Allow Use of Fuel Assemblies Designed & Mfg by Exxon Nuclear Co for Ensuing Cycle 9 Reload & Future Reloads
ML17194B410
Person / Time
Site: Dresden Constellation icon.png
Issue date: 12/21/1982
From: Rausch T
COMMONWEALTH EDISON CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
Shared Package
ML17194B411 List:
References
5584N, NUDOCS 8212290190
Download: ML17194B410 (18)
v • d • e


Text

Commonwealt~ison One First National riJP.Chicago, Illinois Address Reply to:.Post1 Office Box 767 Chicago, Illinois 60690 December 21 ~ 1982 Mr: Harold R: Denton, Dit.ector Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Dresden Station Unit 2 References (a):

( b) :

Dear Mr:

Denton:

Proposed Amendment to Appendix A Technical Specifications to Support Operation with Fuel Supplied by Exxon Nuclear.'Company NRC Docket*No~-50~237* -

T: J: Rau~ch letter to H: R: Denton dated January 11, 1982.

J.* D. Hegner letter to L. DelGeorge dated April 29~ 19,82:

(

P8rsuan~ to 10 CFR ~o:59, *commonwealth Edison proposes to amend Appendix A, Tech~ical Sp~cifications, to Provisional Operating License DPR-25 for Dresden Unit 2:

This am~ndment is being submit-ted to allow the u~e of fuel assemblies designed and ~anufactured by It I

Exxon Nu clear Co mp any Inc: (ENC). for the ensuing' Cy c 1 e 9 reload and 00 future reloads at Dresden Unit 2.

s 1/'/0 N&tl*"-?

. Attachment 6 to this letter provides the changes proposed I.* ~..n to the Technical Specifications and Bases.

A detailed description q~ 1~ of these changes, along with a general discussion of the Dresden 2 Cycle 9 Reload is provided in. Attachment 1.

~~~

'z ~~;>

These proposed changes have received On-site and Off-site

~

1 ~p rev i e w a fl d a p prov a 1.

~' ' "'

,.,,,~ 1,.,?

Attachments 2, 3 and 4 to this letter provide the Dresden 2

~J plan.~ specific reload, transient and LOCA analysis report prepared t:~:, p by ENC..

Attachments 2 and 3 contain information proprietary to. the io:t.'t o'.Exxon Nuclear Company.

As such,,._they are accompanied by an affida-

?J.o? ;.,J, v i t ( At t a chm en t 5 ) s i g n e d b y E NC, the o w n e r o f t h e i n f o rm at ion.

Th e

  • affidavit sets forth the basis on which the information may be with-held from public disclosure by the Commission, and addresses with specificity the considerations listed in paragraph (b) (4) of Section

, 1

/2{.790 of the Commission's regulations:

jl.&co' w c fc.h: /Yooo.. oo r'

8212290190 821221

\\'

PDR ADOCK 05000237 p

PDR I~!

\\

(

,jl

~* \\.

2.:..

December 21; 198 2 Ac co rd ingly; it is respect fully requested that the information which is proprietary to Exxon Nuclear Company, Inc; be withheid from public disclosure in.accordance with 10 CFR Section 2;790 of the Commission's regulations; Correspondence with respect to the proprietary aspects of this application for withholding or the supporting ENC affidavit should be addressed to R: B: Stout; Manager Licensing and Safety Engineering; Exxon Nuclear Company~

2101 Horn Rapids Road, p;o: Box 130;,Richlqod; Washington 99352.

In Reference (a); Commonwealt~ Edison requested nearly identical changes to operate. Qre*sden:.Unit 3 with* fuel:~supplied by ENC; Amendment 63 to the Diesden 3 Li~~nie DPR-25 authorizing these changes was transmitted by Reference (b):

Because this request is similar* to that previously,re"viewe*d* and,authpriied in R.eference (b);

it does not involve a ~6m~lei.i~su~; t~e~efote th~s*ctinstitutes a 10 CFR 170 Class III request.

As such, a $4,000 fee remittance is enclosed;

  • ~I, ;..

~

Please address any questions you may have to this office; Three (3) signed o-~iginals and thirty-seven (37) copies of..

this letter with Attachments l; 4; 5, 6; 7~ 8 and 9 are provided for your use~

In addition; six (6) copies of this letter with proprie-tary Attachments 2 and 3 are also beihg provided at this time; lm Ve r y tr u 1 y y o u r s ;

'.,$~--"/ ~cwc*L

'Thom*as J: Rausch Nuclear Licensing Administrator cc:

Region III Inspector - Dresden Attachment (separate sheet) 5584N

, *I Attachments (1):

Dresden 2 and Cycle 9 Reload Discussion and Description of Technical Specification Changes 5584N (2):

Dresden 2~ Cycle 9 Reload Analysis Report XN-NF~82-77(P); Rev~ 1 dated November 1982 (3):

Dresden 2; Cycle 8 Plant Transient Analysis Report; XN-NF~82~ff4(P); Rev: 1 dated November 1982

( 4) :

( 5):

( 6) :

. Dresden Unit 2 -LOtA Anal~~ii Using the Exem "Eval.uation Model MAPL..HGR Results; XN-NF-82~88,

,*Rev>:l~.: dated November 1982.,*

I Affidavit of R: s: Stout Attesting to the proprietary nature of.XN-_NF-8l-75(P), dated Dec.~.f!l~.er* 1982..

Pro~osed Technical Specification Changes to DPR-19:

Dresden 2; Cycle 9 Reload Analysis Report; XN-NF-82-77 (Noh Ptriprietary); Rev~ *l, dated November 1982~

(8):' Dresden 2; Cycle 9 Plant Transient Report; XN-NF-82-84 (Non Proprietary); Rev~ 1, dated December 1982 (9):

Errata and Addenda Sheet No~ 7 to Ores.den and Quad Cities LOCA. Analyses; NEDD 24146

e ATTACHMENT 1

  • Dresden 2 Cycle 9 Reload Discussion and Description of Technical Specification Changes Dresden 2 Cycle 9 will represent the first reload of Exxon fuel in Unit 2, and the first use of 9x9 fuel in a.Jet Pump BWR~

The following safety evaluation addresses the fuel design; reload analyses and Technical Specification changes suppbrting operation of D2C9 Reload XN-1.

The evaluation is divided i~to four sections as follows:

I~.

I I~

I I I~

IV~

Reload Fuel and Core Design Transient and Accident Analyses Technical Specifications Summary Sections I and II are based on the Dresden Station Unit 2 Cycle 9 Reload A~al~sis; XN-NF-82-77 (Attachment 2)~ the Plant Transient Analysis for Dresden 2 Cycle 9; XN-NF-82-84 (Attachment 3); and the Dresden Unit 2 LOCA Analysis using the ENC EXEM/BWR Evaluation Model MAPLHGR Results~

XN~NF-82-88 (Attachment 4)~

Section III describes the proposed Tech.nical Specification changes required for Cycle 9~

Following the Section IV summary is a list of references primarily consisting of ENC Topical

. Reports on their generic Jet Pump BWR methodology~

I.

RELOAD FUEL* AND-CORE* DESIGN Dresden 2 Reload XN.:.1 will consist of 220 ENC 8x8 reload assembli.es designated as type XN8D3~02 and 4 ENC 9x9 lead assemblies (LA's) designated as type D2 9x9 2:91~

The core loading will consist of the following:

  • Number-of*. Bundles 92

. 24 384 220 4

Bundle average enrichments 5584N

    • Fuel* Type**

GE 8x8-2~50%

~E 8x8-2~ 62%

GE P8x8R-2~6.5%

XN-1 8x8-2~83%

xN..:.1* 9x9-:-2~78%

e*

.:.. 2 -

A:

Fuel Mechanical Design The mechanical design of the 8x8 reload fu~l is described generically in Reference i:

In_genera1; design. crit~ria are establi~hed to limit,the~sttess; strain and.overall duty on the*

fuel rod or bundle d~ring noimal and.transient operation:

In addition; the.fuel is ~esign~d to be.methanica~ly compatible with other reaceor:, ioterrials'.,.fuel,.handling equipment.and existing fuel:

The fo~r' 9x,9*; L:A'{~,h-ave*.been~9e~lgrie'd to Q,~ m.echan_ically and hydrauJ_Jcall'y *compatible.)iitb** the *co.:..re.si_qent BxB fuel and **

reactor iriternals> *The D2 9x9 2~97 fuel design i*s a 9x9 array with 80 fuel.rods (6 containing Gadolinia) and one spacer capture. water rod~'~"The; active fuel length is 145:24 inches which includes a 6.i~6h blanket of:nat~raJ U at both top and.

b~ttom: Enriched fuel pellets h~ve a 1% ~i~h volu~e whi*le ~he natural fuel pellets have a 0:5% di~h volume:

Fuel rod pitch is m ai ntai r:ied via seven Z ire aloy.:.. 4 spacers with In cone! springs:

  • Lower tie plates are drjlletj to improved reflood capability a~d employ* a* spring seal at the tie plate c*hannel' interface to limit coolant ieakage to the bypas'.s r*egion as a result of channel side w*a11 d e-for_ma t ion (bulge) with exp o*sure ~

Th~ 9x9 LA fuel-d~sign utilizes Zircaloy.:..2* cladding with t~e

  • exception of 6 Zr-4 cl~d fuel. iods in one -0f the.4 9x9 LA 1s:

Zr-4 differs from ~r-2 in that zr.:..4 contain nd nickel while the o~~er alloying metali ~iist in about the same 9oncentrations in both Zr-2 and Zr-4:

Zr-4 is already used in BWR channels and.

spacers as well as for fuel pin cladding in PWRs:. *

  • B:

Thermeil Hydraulic Design 5584N The primary thermal hydf~ul~c design ~riteria foi 8x8 XN-1 fuel are identified in Reference i: *As discussed in the D3C8 NRC SER

( A~ end men t 6 3 i s ~>LI e d A p r i 1 2 9,

  • 19 8 2) ; 8 x 8 X N-1.. f u e l i s t he rm a l

~ydraulically compatibl~ with GE fue1:

.. Because XN-19x9 fuel is also *thermal hydraulically compa"tible with. XN-1 8x8 and GE 8x8 "fuel~ the core thermal hydraulic *

  • response is expected to be ~imilar to previous reJ_oads:

Analyses.made during the cCllculation of the* Fuel Cladding*

Integrity Safety Limit demonstrated that a MCPR Safety Limit of 1.05 provides assurance that at least~99:9% of the 8x8 and 9x9 fuel rods in the. core would be expected to avoid. boiling transition during steady state operatidn *at the safety limit.

Refer to Reference 2 for*further* discussion of the methodology.

3 -

c~

Fuel Centerline Melting and Cladding Strain 5584N One of the ENC's primary thermal hydraulic design criterion for D2 reload XN-1 fuel is that fuel design and operation will be such that f~el centerline melting *is not expected f6r antici-pated operational occurences (transients) throughout the life of the fuel~

To demonstrate compliance with this criterion, ENC has performed transient overpower analyses for a fuel rod history (peak LHGR vs: exposure) which.represents a conservative upper bound on peak rod power over the life of the fuel bundle:

The results indicate that substantial margin to centerline melting is assured for both 8x8 xN.:..1 arid 9x9 XN-1 fuel:

The cladding strain at 120% overpower condition was calcuiated and determined to be less than the ENC design criteria of :2%

plastic strain for 8x8 XN.:..l fuel:

For previous reloads, GE provided an LHGR design limit~ which was incorporated in the Technical Specifications as an operating limit, to ensure margin~du.r'ing transients to the 1% plastic strain Safety Limit assumed iri the GE licensing basis:

In addition, a Jechnic~l,Specification provision for reducing the APRM scram and rod.btpck s.e:tt-fng _by FRP/MFL.PD was incorporated to ensure margin to 1% plastic strain during transients initiated from r~duced core flow with exc~ssive peaking, (i:e:

peaking whtch would resuEt.. in an LHGR i~- e~~~~s.of the operating limit if recird flow w~re increased.to rated}.

For D2C9. this approach and the corresponding ~echnical Specification provisions will be mai,ntaine_d for* GE.fueL For XN-1 reload fuel, the fuel desi~n *is.~~ch~that margin to centerline melt and the design ctiterlon of 0:2%. pla$tic st~ain is a~sured. for overpower conditions throughout the life of the fuel as demonstrated by.the Fuel Design Analyses:

Since this inherently protects against 1% plastic strain and assures ma*rgin to the expected threshold for strain induced claddiri~ failure; an LHGR Safety Limit is not necessary and has not been specified for ENC fuel:

As a result; no Technical Specification LHGR Operating

However; to ensure applicability of the Fuel Design Analysis; proposed Technical Specifications 3;1:s and 4;1:s~2 will require a daily surveillance on power distribution for ENC fuel~

In most cases, operation within the MCPR and MAPLHGR limits will ensure that the power distribution for ENC fuel remains* within the assumptions of the Fuel Design Analysis~ *However; this surveillance will ensure under all conditions that the peak LHGR for ENC fuel is procedurally controlled to provide margin to centerline melt for overpower conditions initiated from rated power or reduced flow:

}

4 -

D ~

Nuclear* Design The 8x8 XN-1 fuel design consists *of 63 fuel rods and one water rod~

The average assembly enrichment is 2:83% which includes a six inch natural U blanket'at both-top and bottom:

The average enrichment of the central region (excluding blanket) is 3~02%.

Five burnable p'oiscl'n rods containing.*a Gd203...:.uo2. mixture are utilized.to red~te initial bundl~ reacti~ity.

The specific neutronic design parameters and pin enrichment distribution are provided in Section 4 of Attachment ~~

The 9x9 XN-1 fuel design consists of 80 fuel rod_s* ancf one *wate*r.rod~

The average assembly enrichment is 2:18% whiQh also includes a six inch natural U blanket at both top and bottom:

The average center zone enrichment is 2:97%~. The 9x9 fLlel design contains six poison rods utilizing a Gd203~U02 mixture to reduce initial bundle reactivi~y.

Core Reactivity ~ The calculated BOC9 cold core keff values at all rods out and all rods in are 1.111 and 0.958 respectively.

The shutdown margin with the strongest control rod out was determined to be Ll%A k with the most reactive point in the cycle being Boe:

Therefore R is merely equal to.0:*04% Ak to account for inverted tubes in the control blades.

The Standby Liquid Control System (which is ~esigned to inject a quantity. of boron that. produces a concentration of no 1 es s tha*n 60 O ppm of boron in the reactor core in less than 100 minutes) was calculated to provide a shutdown margin of 4;4% Ak for cold conditions with all control rods in their full power positions~

The calculated shutdown margins for the strongest rod out and SBLC system are both well in excess of their respective Tech n i ca 1 Spec i f i cat i o n Re q u i rem en t s ~ O : 2 5 % + R an d 3% ~

Core Stability~ The decay ratio for the D2C9*core was determined to be 0~46 at the intersection of the natural*

circulation line and the 100% Flow Control Line:

Fuel*Storage*Vault/Pool Criticality - Technical Specification 5.5 requires that the keff of the spent fuel pool be.~.95 and that of the new fuel storage vault

-<~90 when, dry ( < :95 when flooded).

In NEDE-24011, GE states that these criteria will be met for GE fabricated racks if fuel bundle reactivities are limited to koo

.S1:31 fort.he rack dimensions utilized in the Dresden spent fuel pool and

~ 1~30 for the rack dimensions utilized in the new fuel vault, where k00 is calculated in an infinite array of similar fuel in the core configuration (as opposed to the storage configuration):

GE has calculated koo's for their fuel designs and demonstrated that the criterion is satisfied for all GE fuel~

ENC has calculated k00 for 8x8 and 9x9 XN-1 reload fuel and for a.comparable GE fuel design.

Based on the comparison in Appendix A of Attachment 2 and the criteria from NEDE-24011~ it is concluded that arjequate margin to the Technical Specification keff limits exists for storage of both 8x8 and 9x9 XN-1 reload fuel in the vault and pool (for GE designed racks).

,:_t I

  • t 5 -

For the.high density fuel storage racks designed by Nuclear Service~ Corporation; criticality anal_yses have been performed for ENC fabricated fuel with a* center zone'enrichment of 3~02 w/o and demonstrated that the 0~95 keff requirement is met~

Based on bundle reactivity co~parison provided by ENC, the high density fuel storage racks meet the 0:95 keff: requirement for

  • both ENC 8x8 and ENC 9x9 fuel when gadolinia is considered~

I I.~ TRANSIENTS AND ACCIDENTS.

A:

Anticipated Operational 6tc~~re~Ees (Transients)

In order to deterl)line. 0p~rat_i.ng. limits fo_r D2C9~ ENC has considered eight categories.of. co'i*e:-wide potential transients

(~s described in Reference 3) and provided analyses results for the following tl:1ree transient~ to -q_e_termine~*t.he thermal *margin for D2C9.;'

  • \\
  • 1

~

1 7-'

~?

  • . I

,.. ~,* ~. ~

/'

Gener~tor Load Rejection-wit~Out*Bypass (LRw/oB)

-

Loss of Fe ed:wa t er* 11e*~.t ing -<<Lo FWl~ ><*

. The. other core'.:::-wide_ trC\\nSi.ents are:inherently non-limiting or b.ounded by one of the above*:* In adition; two loca+ even'ts; Rod

  • Withdraw~! Eiror a~d.F~el L~ading.Error; were analyzed as

~escribed in Reference 4 and determined to be non-limiting~

The*

result's of the core.:.wide q.nd.local transient. analyses are _..

provided in Attachment 2 (XN-NF-82~77; D2C9 R~load Analyses) and. (XN-NF~s2~s4; D2C9 Pla~t Transient Analysis Report).

The.9x9 XN-1 *fuel desl.gn was incorporated.into the analysis of these events by choosing limits that. maintain the. same bundle

  • power for*9x9 fuel as. fdr 8x8 fuel and then verifying the limits by transient analyses~

The Generator Load Rejection without Bypass was determined to be the limjting event* for D2C9;.

resulting in a A CPR of o ;26 for.8x8 fuel and

~30* for 9x9. which, when combined with the*1;05 Safety Limit; requires a MCPR operating limit at 1:31 _for. all 8x8 fuel types an~ a MGPR operating limit of 1.:35 for"9x9 XN..:.1 fuel~

Core*Wide Transients - The. plant transi~nt model used to evaluate the LRw/oB and FWCF events was ENC's COTRANSA code (Reference 3) which incorp6ra~es a one-dimensio~al neutronics

_model to account for shifts in axial power.shape resulting from rapid pressurization and.void collapse:

The LRw/oB event was found to be the most limiting event and therefore analyzed statistically while the non-limiting FWCF was analyzed determi-nisticall~ (using bounding values as input parameters)~

The LOFWH event was analyzed deterministically with ENC'S PTSBWR code (Reference 3) which uses a point-kinetics neutronics model since rapid pressurization and void collapse do not occur for this event:

Both codes utilize a multi-node steam line model to accomodate pressure waves in the steam line:

6 -

One of the statistically varied inputs to the LRw/qB analysis w a s th e con t r o 1 r o d s p e e d du r i n g th e re a c t o r s cram.

  • Act u a 1 scram time data from previous cycles on Dresden 2 were used to generate the scram time distribution assumed in determining the A CPR distribution for this event:

In order to assure the applicability of the LRw/oB analysis to cycle 9* operation; compliance with the assumed scram tim~ distribution must be.

verified throughout cycle~ as required by proposed.T.S.

4~3~C:3~ If the* current cycle scram speeds deviate from the assumed distribution;. an adjustment to the MC PR oper~t i ng 1 imi t may be: required.

The method for checking compliance and adjusting the MCPR operating limit is provided in propo~ed Technical Specifications 4~~:c~3 and 3:5:K:

Local-Transients As shown in Attachment 2, the results of the FQel Loading Error and Rod Withdrawal error were bounded by the LRw/oB event and

  • are therefore non~limiting: Based on the RWE results~ the rod blo~k monitor setpoint will be increased from the currerit value of 107% to 110% to provide additional flexibility. in utilizing the allowable power/flow operating region above the 100% flow control line:

The A CPR for the RWE event with a 110% full flo.w RBM setpoint.was 0:13:

TheACPR for the fuel loading error e vent w as O

~ 14 :

A 11

  • o f t h e A c PR s a re 1 es s t ha _n the 1 i mi t i n g value of 0.26 calculated for the LRw/oB event~

Reduced Flow Operation

~NC has prqvided MCPR operating limits for manual and automatic flow control reduced flow operation in Attachment 2~. These values are based on the analyses provided in Reference i6 fdt 03C8 which ENC has indicated are applicable to D2C9.

ASME Overpressurization*Analysis In order to d~monstrate compliance with the ASME Code Overpressurlzation criteria of 110% of design vessel pressure, the MSIV closure event with failure of the MSIV position scram was analyzed with ENC's COTRANSA code:

The maxi~um pressure observed in the analys~s was about 1349 psig or 108% of reactor vessel design pressure., The corresponding steam dome* pressure wa~_about 1325 psig~ for a vessel differential* pressure of 24 psi.

Thi.s includes the... effects of th'e A TWS RPT which was a s sum e d t 0 *i n i tia t e a t. a h 0 m in a l' p*i'e s s u re s et p 0 i n t 0 f 12 4 0 psig.

The ASME limit for peak vessel pressure of 1375 psig (110% *a f desigr pres sure) is the re for~ e,qu i valent to a dome pressure limit of.1351 psig (1375-24) ~

... The Technical Specifi-cation S~fety~Limit of 1325 ~sig is b~sed on dome pressure and therefore conservatively assumes a 50 psi vessel dp (1375-1325).

The pr.oposed safety :l~mit* of 13.45 psig dome pressure is based on

l e

e*

7 -

a 30 psig vessel dp which removes excess conservatism while_

continuing to bound expected.differential pressure behavior, especially when the lack of forced flo~ imposed by RPT is considered:

The choice of 1345 psig th~s assures compliance with th~ ASME criterion of l375 psig peak vessel* pressure while

.alsb mai8taining consistency with the DJR7C8 pr~ssure safety limit.

s:

P6stulated-Accidents In support of D2C9 operatiqn~ ENC has*reanalyzed the. Loss of-.

Coolant Accident (LOCA) to determine MAPLHGR limits for XN~l fuel and the Rod Drop ~ccident (RDA) to_demonstrate compliance with the 280 cal/gm Technical Specificatibn limit:

The results of. t'hese *analyses are presented in Section 6 of Attachment 2.

The methodology f6r the RDA analysis is.described in Reference 4 and that for the LOCA analysis is*prbvided in References 6-thru 13 :*

Loss: of Coolant* Accident~ Reference 6 de~cribes ENC's generic jet pump BWR3 LOCA break spectrum analysis which defined the limiting br~ak for BWR 3's tq be a double-en~ed guillotine break in the recirculation piping on the suction side of the pump.

  • The.a*nalys1s of this event f'or Dresden* 2 is provided in - and summarized in Attachment 2: *Operation w.ithin the MAPLHGR ~imits of Section 6:1 -f~r ~NC 8~8 fuel and Table* A:~

fa r - th e 9 x 9 LA ' s ( At t a chm e ri t 2 ) w i 11' ens u r e t ha t th e p e a k cladding* t'emperature remains tielo,w

  • 22000F ~.lo.cal Zr-H20 reaction remains be1ow. 17% arid core~wide: hydrogen prod.uction remains qelow 1% for the *limiting LOCA event:

The LOCA analysis of Attachmept 4 was performed for an e*ntire core of ENC 8x8 *

  • reload *fuel and.therefore provide MAPLHGR 's for ENC fuel only:*

The MAPLHGR limits for D2 9x9 XN-1 'LA's were established to.

maintain n o.dal powers equivalent to the* ENC 8x8 f ue 1 -

assemblie,s:

ca*nfirmatory ECCS analyses were performed to verify that the 9x9 MAPLHGR limits maintain peak clad temperatures and.

local oxidation tractions within lOtFR5D Appendix K limits.

As discussed pr~~iously; ENC reload fuel is hydraulically ~nd.

" n e U tr on i ca* 11 y compatible wit h G

~ E : f u e 1

  • There f o re, the

-~xisting G:E:*LocA 'Analysis (Reference 14) and MAPLHGR li~its will remain applicable during D2C9 and*future cycles with GE/ENC mixed cor'es: '.

~ad Drop Accicient -

ENC'~ methodolcigy ~or snalyzing the Rod brop Accident. (RDA) is described in Reference 4 and utilizes a generic parametric analysis ~hich calculated the fuel enthalpy rise duririg postulated RDA's over a wide range of reactor operating variables:-. For D2C9; Section 6 of Attachment 2 shows a value of 111 cal/gm for the maximum d~posited fuel rod enthalpy during the worst c~se postulated RDA:. This value is well below *the Technical. $pecifjcation limit of 280 cal/gm.

To ensute compliance with t~e~RDA ~naiysis assumptio~s; control rod sequencjng below 20% core thermal po~er~must comply with G:E: 's Banked:Positidn*Witrldrawal* Sequen~i~g 6o~~traints (Reference 15):

.IIL*

.:. 8 -

TECHNI~AL-SPECIFICATIONS

.. provides *proposed Technical Specification ~hanges to support D2C9 operation with ENC 8x8 and 9x.9 fuel~ *The following sections highlight. the major areas requiring revision and identifies the associated sections* of the Technical Specifications~

A:

GENERAL.

Throughout the Technical *Spe*cifications and bases; sections have been r~viserl to reflect the appropriate.Exxon.M~thodologie~ and references ~nd delete Gener~l Electric methods and references where necessary~

Also; fo.r each* revised specl fic.ation as i~entified below~ the.corresponding section oft~~ bases has beeri revised as required:

8:

LHGR As described previous 1 y ;

  • no L H GR s a f et y Li.mi t o r operating Li mi t is specified for ENC fuel.

Operation within the MCPR and MAPLHGR limits; and th~-power distribution assumptions of the Fuel Design Analyses will prote~t against fuel centerline*

melting ~nd thereby protect> against strain-induced cl-adding.

failures.

All Technical' Specif,ication sections referring to LHGR or FLPD have been.revised to apply only to GE fuel:.. New speci fic.ations ha.ve* been. prop_qsed which require surveillances on ENC fuel to ehsur*e *a p'pllt:abf1 i ty ci f the Fu el Des~ gn An_alysi s<

In addition **~d *the;.above i*:a:J.1 refe;ren_ce"$ tp;'-:7_x7 f\\JeJ and the*

power spikl.ng peh'a'lty: have been deleted *sinc:e there will be* no 7x T fuel i'n D2C 9 arid SU f_f icien t margin to" LHGR 1 imi ts exist tn accommodate the expected power spiking due to fuel densific~tipn~.

r

~.

Ls~

  • Sectio.n i,.. * *o*escriptfo.n
  • LK 1 ~ l ~A ~ 1) 2: l : 8: l 3 ~ 1 ~ 8/ 4 ~ 1 ~ B Table 3:2:3 Note 2 3 *: 5 : J I 4 : 5 : J Definition 6f FLPD revised to a~ply to GE fuel only:

APRM Scram and Rod Block equations revised to provide MFLPD/FRP -adjustment f d r GE f u e 1 only ~

F o r.E NC f u e 1, a requirement to ensure compliance with the FuSl Design Analysis has been added~

Revised to iequire LHGR limit and

  • surveillance for GE fuel only; *Deletes reference to 7x7 fuel and power spiking.

C:

MCPR T ~ S*: *Section 1: LA 3:5:K Figure 3.5.2 9 -

Description MCPR Safety Limit changed to 1:05 MCPR LCD changed to 1.31 for all 8x8 fuel typ~s and 1:35 for the 9x9 LA's.

Revised to indicate new curves for determining MCPR limit during operation at reduced flow and to require adjustment of the limit if scram times fall outside the*

distribution assumed in the transient analysis:.

Replaced with new figures for determin-ing MCPR limits during operation at reduced flow:

D~ Reactor* Coolant* System PressQre-Safety* Limit (Section 1:2)

Changed from 1325 to 1345 ps~g~ Previous value assumed a vessel

  • pressure drop of 50 ps.C New valve is conservative compared to the actual pressur~ drop as determined by analysis~

E~

RBM Setting (Table 3:2:3)

  • Changed from ;65Wd+42 (107% at full flow) to :65Wd+45 (110%)

based on result5 of RWE analysis:

F:

Sectioh 3~5~0~3~a*

This section which allowed operation with o~ly 4 ADS valves-during 02 Cycle 7 has been deleted since it is no longer applicable:

Analytical support for such operation will have to be purchased from Exxon and licensed later if desired:

G:

MAPLHGR.

Section 3:s:r14:5:1 Figure 3*.s:~1 Description Revised to distinguish GE MAPLHGRs which are functions of nodal exposure from ENC MAPLHGRs.: which are dependent on bundle average exposure.

. *Add MAPL.HGR 'curve* for ENC fuel types XN so3:02.and 02 9x9 2:97 while deleting curves for ?x7:- fueL, Extend MAPLHGR curve s f o r,. ~ E f u e 1. tip es P 80 R 82 6 5 H,

Section Figure 3.5~~1 (Cont'd) Description 8DRB265L; P8D~B282 and P8DRB2~5L out to 40,000 MWD/ST.

In addition; since Errata and Addenda Sheet No. 7 to.NED0-24146 provided ex tr a ma rg i n for P 80 R 82 6 5 L o v e r 8DRB265L; a s*eparate curve is now

  • pr6vided:

E&A Sheets fat the changes to the GE fuel type MAPLHGR curves are provided as Attachment 9~

Table 4P of E&A Sheet No~ 7 shows MAPLHGR values for P 80 R B2 6 5 L as a pp 1 y in g on 1 y t o

  • Qua d Cities~ General Electric was contacted regarding this situation an~ indicated that Table 4P does apply to Dresden as we 11 a s Qua d Ci t i e s ~

H

~

S c r a m -Ti in e

  • Su r v e i 11 an c e* "

Specification 4~3~C~3 has been added to require verification after each set of scram' ti*min.g data t_h*at the cur:r-E~n*t scram speeds fall within th.e distribution assumed in *the tran'sient analyses~

L Li c e n s e cci n d it i o n

  • M
  • The previously approved.provisions to allow operation with one recirculation loop d~t of *~~~vice.~~e ~eing revised slightly to refer to correct Technic~l Specifitation~~refet~nce~ and to only specify

~03 CPR adders as oppose~ to actual MCPR values~

These changes are strictly for incre~sed clarity and to pieclude having to change values every cycle.

TJR/lm 5584N

REFERENGES L

XN-NF-81-21( A)~ "Generic Design Repo:i;t-Mechanical Design for Exxon Nu cl e*ar Je. t. Pu mp B WR Fuel Assemblies," dated Oc~obe r, 1981.

2~

  • XN-NF-524(A); "Exxon Nuclear.Critical Power Methodology for Boiling Water Reactors" dated Novem6er~ 1979~

3.*

XN-NF-79-71 Revision 2 "Exxon Nuclear Plant Transient M~thodology for Boiling Water Reactors" dated November; 1981*~

4~

XN-NF-80~19(A) ~

Volum.e *1 (Supplements l *and 2); "Exxon Nuclear Methodology for Boiling Wate~ Re~ctors Neutronics Methods~for besign and Analysis" dated M~y 1980.

5~

XN-NF-81-22(P)~ September.1981 Generic Statistical Uncertainty Analysis Mettiolo.logy 6~

XN-NF-81-7l(A)~ October 1981 Ge n e r i c J e t-Pu mp : B W R3 L 0 C A An a 1 y ~ i s Us i n g t h.e E NC E XE M Ev a 1 u at i on Mod~l.

7~

XN-NF.-:82~88~ "Dresden,. Unit 2 LOCA Analysis Using the ENC EXEM/BWR Evaluation Model-MAPLHGR Results" dated November*~ 19~1

,-, (At tach.ment 4) ~

  • 8~

XN-NF-80-19(A); Volume 2; Revision i~ Ju*ne 1981.*

Exxon Nuclear Metho~o1ogy fd~ Boiling Water Reac~ors

~v alua*t iori Model; Summary D~ scr iption*,.

9~

XN-NF-8o:...19(A) ~

Volume iA; Revision l~ June 1981 Exxon Nuqlear Methodology for 8oiliAg Water Re~~tors RELAX: A RELAP4.Based Co~puter Code* for Calculating Slowdown Pt:ienomena

10.

XN-NF-80-19(A), *volume 2B; Revision 1; June 1981 Exxon Nuclear Methodology for Boiling Water Reactors FLEX:~ Computer Cod~ fo~ J~t PLlmp~BWR.Refill and Refloo~ Analysis

11.

XN-NF'-80:..19(AL,.Volum.~.~2c,.-..;June_1~81,-" *-,

  • Exxon Nuclear Mett:io.c;Jofogy f.'ox '86,i-ling~-W.ater: Reactors*

Jerification and Qu~llfication of EXEM XN-CC-33( A)*, R*evisi.on~_.L~-*,.Nbvembe.x 19.i.5 ~ * :

HUXY: A GeneraXi z i:! d~ Multi rod Hea tu p 'code{ with lOCF R50 Append"i x K Heatup Option.

12.

13

  • X N - NF 5 8 ( P),

A~-g us t: 19 8.1. _. :*

RODEX2 Fuel Rod Tti*ermal-Mechanical Response :!::valuation Model.

5584N

e REF ERE NEES

.(Con' t) 14~

  • NED0.:...24146A Revison l; 0 Loss of Coolant Accident Analyses-Quad Cities 1/2~ Dresden 2/3" dated April 1979~

15; NED0-21231; "Banked Position Withdrawal Sequence" dated January 1977; 16; XN-NF-81-84(P) ~ "Dresden Unit 3 Analysis for Reduced Flow Operation"; November 1981; 5584N

ATTACHMENT 5 A F F I D A V I T STATE OF Washington COUNTY OF Benton SS.

I, Richard B. Stout, being duly* sworn, hereby say and depose:

i.*

I am Manager, Licensing and. Safety Engineering, for Exxon Nuclear.Company, Inc. ("ENC"), and as such I am authorized to execute this Affidavit.

2.

I am familiar with ENC's detailed document control system and policies which govern the protection and cbntrol ot information.

  • t_~

. "* 1

3.

'I.arnfqmiliar with the po~ument.s (1) XN-NF~82-77(P), entitled

~.

I,

~.

~

~

"Dresden Unit 2 Cycle 9 Reload 'Analysis," and (2) XN..:NF-82-84(P), entitled "Plant Transie~t Analys'.is for Dresde.n:unit:2i'Cyc,le.'9,";*reforr.ed to as "Docu-

'I i'*li ments".

Information contained i~ these Documents has been classified by ENC as proprietary in accordance with the control system and policies established by ENC for the control and protection of infor~~tion.

4.

The Documents contain information of a proprietary and con-fidential nature and is of the type customarily held in confidence by ENC arid not made available to the public.

Based 6n my experience, I am aware that other companies regard information of the.kind contained in the Document as being proprietary and confidential.

\\

5.

The Documents have been made available to the United States Nuclear Regulatory Commission in confidence, with the request that the information contained in the Document not be disclosed or divulged.

2

6.

The Documents contain information which is vital to a com-petitive advantage of ENC and would be helpful to competitors of ENC when competing with ENC.

7.

The information contained in the Documents is considered to be proprietary by ENC because it reveals certain distinguishing aspects of safety analysis 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 not know or use the information contained in the Documents.

8.

The disclosure of the proprietary information contained in the Documents 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 safety analysis methods, and would result in substantial.harm to the cbmpetitive position of ENC.

9.

The Documents contain proprietary information which is held 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 Documents has been made available, on a limited basis, to others outside ENC 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 secured file or area and distributed on a need-to-know basis.

(

3

12. These Documents pro vi de information which reveals safety analysis methods developed by ENC over the past several years.

ENC has invested millions of dollars and many man-years of effort in developing the analysis methods revealed in the Documents.

Assuming a competitor had available the same background data and incentives as ENC, the c~mpetitor might,. at a minimum, develbp the information for the same expenditure of 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 1 s competitors are sufficiently similar to the corresponding background data and incentives of ENC to reasonably expect such competitors woul~ be in a position to duplicate ENC 1 s proprietary infor~ation contained i~.the Documents.

THAT the.*: state~e~,ts' made~\\~'r/~.iri*aboy*e_,.ia'~e, *to *:_t~e best of my kno~ledge, information, and belief, truthful and compl~te.

~

1..' '. :

  • ~

! l

  • f.

FURTHkR AFFIANT SA~E,TH No{*

SWORN TO AND SUBSCRIBED before me this -:;

day of

~-'

19JZ--**

£~~

Retrieved from "https://kanterella.com/w/index.php?title=ML17194B410&oldid=5130114"