· f Ad PR tFCF Non Proprietary I 9. INTERNAL PRESSURE 9.1. Phenomenon The gas pressure within a fuel rod changes continuously during irradiation. Since fuel rod internal pressure

f Ad PR tFCF Non Proprietary

I 9. INTERNAL PRESSURE

9.1. Phenomenon

The gas pressure within a fuel rod changes continuously during irradiation. Since fuel rod internalpressure is a design criterion, it is necessary to ensure that the COPERNIC internal gas pressurepredictions are accurate.

9.2. Model

The gases within a fuel rod are considered to be ideal and the BOYLE-MARIOTIE law can thereforebe applied:

P.Vj - n= *R T Eq. 9-1

where:

P : fuel rod internal gas pressure,

Vi : jth internal void volume,

n : number of moles in the Vi void volume,I R : universal ideal gas constant, and

iT : mean gas temperature in the V; void volume.

2 The pressure P for n inter-connected volumes can be expressed as:

8 xlni

| P= R V Eq.9-2V.

The Xini term represents the total free gas inventory. The various internal volumes are:

[D-]

_

~. ~.FCF Non Proprietary

i' 1 Chapter 9 PAGE 9-2

[b.1

Eq. 9-3

[b., d.]

9.3. Experimental Validation and Uncertainties

The void volume and internal pressure measured to predicted comparisons for U0 2 are presented inFigures 9-1 and 9-2 respectively, for both the Zircaloy-4 and Alloy 5 claddings. This data is listed inTable 9-1. The application of the code internal pressure predictions to fuel rod design analyses,including fuel-clad lift-off, is described in Chapter 12.

9.4. Adantation to Mixed Fuels

&O

2

8LU

U.

The calculations that are performed to predict the internal gas pressure of U0 2 fuel rods are identicalto MOX and gadolinia fuel rods.

The void volume and internal pressure measured to predicted comparisons are presented in Figures9-3 and 9-4 (Fable 9-2) for MOX fuel and Figures 9-5 and 9-6 (Table 9-3) for gadolinia fuel,respectively.

9.5. Validity Ranee

The validity range of this model corresponds to the code benchmarking ranges described in Chapter 1.

II

i0w

C9I-

IL

FCF Non Proprietary

s ' -' Chapter 9 PAGE 9-4

FIGURE 9-1 MEASURED AND PREDICTED U0 2 FREE VOLUME

COMPARISON

-I

'u

2

80

I

[b.]

rm COP RNXC FCF Non Proprietary

I FIGURE 9-2 MEASURED AND PREDICTED. U0 2 INTERNAL PRESSURE

COMPARISON

IE2 [b.]

0

ql!'".p i.-~muFCF Non Proprietary

' k Chapter 9 PAGE 9-6

FIGURE 9-3 MEASURED AND PREDICTED MOX FREE VOLUME

COMPARISON

w

0C.au

0I-

[b.-

rm a aOs .W - _alI FIGURE 9-4 MEASURED AND PREDICTED MOX INTERNAL

PRESSURE COMPARISON

(0

,DS

[b.]

00UUS

0I-

K , ; FCF Non Proprietary

f Chapter 9 PAGE 9-8

FIGURE9-5 MEASURED AND PREDICTED U02 -.G0 3 FREE VOLUME

COMPARISON

0

w

U.-

[b.]

.FCF Non Proprietary

f 4¢ 2S iCChapter 9 PAGE 9-9

P FIGURE 9-6 MEASURED AND PREDICTED UO2-Gd2O3 INTERNAL

PRESSURE COMPARISON

0:

I

I.-

FCF Non Proprietary

Chapter 9 PAGE 9-10

w

0

'Li

0

This page intentionally left blank.

FCF Non Proprietary

Chapter 9 PAGE 9-11

TABLES

iIL.

0C,'U

0

I-

FRAMATOME COGEMA FUELS

TABLE 9-1 GLOBAL MEASURED/PREDICTED COMPARISON (UOi2)

.red. M Prd, Mea& PYA.. MeaL. Pred.MFulRdaa- Fs Dmi ze es Void Vold. Jilnt-... Ct F.Rd Fod dl E .Col I F.CoName Prtog. dIg ethience(MWdI G G o I res rs.tLLALI LLIA/¶~~~~~j~~~~~eA (11m) t) [AIA% a 3) (m ) (a) (bt % % AM (~

- - -i

k(b.

0

I(1)

co

-n0*In

0

-u

CD


%A

Wi

C N W,,.

0

Co

mcoI

Xo

,n

0-nC)

0-oaOPZ

I .


.b -; Ii 3-AI .C10 b (G)DS( OL| °id|1 do TRR;-d kt| oL~j 0

-T R -. . -. : - . : . '-p

0

co

-LI

-n

z0

0-ci

-0

FRAMATOME COGEMA FUELSI

0 l

0

0w

CDED

co

m

EDc

-nC)0nz

0

-u0

0la

I - I


FnlRdCad at luru -Pried Mens id s. .Wi -Pred.; Mes. Prd. Me.Na e rg, dngstec (Mu dIi .FMR FGR Voi Void. Dt In F.f t .Ro F.cot. F. C.oL ,(%) VoL Vi.]Pes P s ALL A : A I/

(ib.]to

0

-IE.

(0

0)

-n

z0

20=L


Db.]


I: I Prd e . Po. Mes. Col.Mss Pe. Mes..Fuel Rod aa F fliii Id",d Void--: i. .RdI. o .ol .ClNme d :Pr | dting | Fuiene |Mwd/ DGR | FGR V. L VoL 1 Prss | Pe) | A | AX

% .14

<l'

,}{:.. +

: or:

e}e i;pin;.,:.. .:::- ;ri ;. >

P .eRe: idm ! ash iS,;z, .j:s

F.,+ ';.''a s >f4!r

o. at -S' ... ,.,,;

ts An,;

'i'':. pi'

0

<0

co

Xrn

CD

0

z0-u

M

a*0

03


[b-]


I I Prtd. Met rd 4 Mes.kri.Me Pied. Mei. .Fuel Rod Cld at Bmp Pd. es.Void 'Void Ut -:Jt.: F. Rod E Ro~d t. Cot.' F.C

' Db.]

Ri4

0-I

(0

-UMT1(0

-n

-oX,

010


-- - - -

0

vitr?'1bw

!t

l

Cry

-oCDco

C)mto

ax3

m0

nD2:.9a

* - .b -


- --- Pred. Mes. Pred.' Meas. Pe.mews. rd Me.:fte-RoI..Idt- -oi --- L, ut .6 -~E1d I -

Name Prog. ding Eene (MWd -FGR- FGlt: Vl o. Pes.Pes LL LL W sJ'Iype (nig 2 tUW)~ (%90) (% (a ) (c(br () () () ()

[bMY

-u00-R

co

m

'1

z0

-u-alaa0so

FRAMATOME OOGEMA FUELS

I

hc%..4-wv

ilaj.!lv-t 4

0

m

m

ZoCD,

'1C)

z0

0QM4.


- CTi- -id -liru -rd. e rdass.Po.Mes r.Ms.Frd es

Name Prog. ding Fluence.. (M~I G FGR -Vl rs.Pes J L! LL AI| : S eI Ro Clad-| > 0 ~Adst Bunu Prd Mes Vo |FR ild Viioid mt. mt F.:6 PRod 4 1o F. Co Cl

Typo. (f

ly

i 7..

co

G)mIto

2Zn0B'1

z-0

-0-oCD

FRAMATOME COGEMA FUES

.0,G)IV

to1wVI-

Cl0a

co

-omenco

ali

-n-nz0

-0

-apom

D'

Q

L _


ai - - --1 - T - - " -I ast Buzp Pb&Mas rd. Medas. P -. - M. as. Pre MRats PrE. CoTRding J I ~~~~Vol':.. Vol. Pre"S§W` es IA/ LL LL A/

11 - I, - I -

_ _..

0-f-V -

I z

-o

0

10


W.

IV.wM:VI

0m

co

G)mco

'1mz00

cDa

a. - I -


-.. - . -.- Y

I Frd. Mess. ;Pred Mag Fred. Mes read . MLS.,''FtelRod.Ca- Fs Bru rd Mess. Void: : Vdld n. t. FRo F.od .CAL F KCo

Ne~ ~ ~ j~~'IFI G oVolol. pii. pres ALJ ALt ALIL MJL

Crn

.~ I>4 r

tLI

0--

..

S) ,

_ _

FRAMATOME COGEMA FUELS-y

0

D)

mco

'Co)

z

10

(D

I - I -


- | -b I - t - > 1 * - - - --I 1Pt~d. Mas IPred Men. Pred. 'MeMg Pr. mein;-Fuel Rod Ca- Fs iru rd es odjVi u. Tt F n .Rd ECL F.Cl

Nae Frg dn Idence- (M dWG G o o rs.Pes LL A/ LL AI-Nme.-W - -Pen- .

[b.]

'I..

-I

co

~9Az

0

i13

P)

-0O

ci,


[b-]


Nmn Prg | .iz FJiec (M1d aG Em Void VoMd mt. it. I . U ~dFCLFCI

D*.]

.3

0

C)

_X

mCo

C)0mZo

10

CD-4.LtQ

I


TABLE 9-2 GLOBAL MEASURED/PREDICTED COMPARISON (MOX)

f. ̀i'

He ~9nIVf

Mw

Cl0(D

4

X

m

lD

8o

'1

nz0-U-o0-vSt

-


Fast BuRnuVo'd Void mt.e lita. FVIid F. .rId F.d Cc. M. As.

;ue, :", .d, .. , D..'. ),.::''(....... '' si'G_ _ M . P. M ._ _ d. . ;.

. M. C,

. .E,,4,,

(b.]

0'0

(0

m

::lp i..

w


[b.]


FuelkodNae j Prog. Fluecne -(M4d| FGR o Vs.RAd. OLR A AL/

W-]

M od" ,

ii

�1

'V;l.

0 3

(0

-0

Im

'1

z0

laM3.-4.

a3


[b.]

FRAMATOME COGEMA FUELST

:. ::: -; d.. -. .;P.e...Mes--....; : : : : -M .:..Mes..

FadRod NS t m Prog.~ The Md/! TOR- FR Vo VL Pes rss A A/ L A/

jb.J4.

�,

�"3-

*�A1�P

pC,

CD

CD

-n

0

I4.D

mco

Ca,


TABLE 9-3 GLOBAL MEASURED/PREDICTED COMPARISON (UO2-Gd2 O3 )

0

0

~10.

0- z

XoZ

DIV0

0 i,

m nCa m

Cd,CC)


. - -. . R ' - - i- .e

.g lmence ( / FG F|GR i .- ; od .CoL .. ''. ;) I o.j ::,i_. Prss. Press AL "L AL Mb.I .% % e,(bairy- j% Ž ()

'1

Cm,0=1

mD

co

m

Ia

'10z0

-0a

vQ2

_ _

rM ttOP T-sR )-nic FCF Non Proprietaq

10. MATERIAL PROPERTIES

10.1, DensiVy

10.1.1. Notations and Units

I

p.:

Mi

N.

ai

enrPu:

pgado:

x

y;

density (g.cm73 - material 100% dense)

atomic masses (g.mol-1 )

Avogadro number= 6.022 1023 (atomes.moli")

lattice parameter (cm)

plutonium weight fraction

Gd2O3 weight fraction

deviation from stoichiometry = (oxygen/uranium) - 2.000

mole fractions (U235, Pu, Gd)

10.1.2. Zircaloy

p = 6.54 Eq. 10-1

CO

00C,

0

10.13. Advanced Alloy 5 Cladding

p = 650 Eq. 10-2

10.1A. U0 2

4(( - YU235) MU238 + YU235 MU235) + B .( + 2) MOPU02 = 3

2

Eq. 10-3

where

3U02 = (5.4704+ 0.251x). 10 8 Eq. 10-4

10.1.5. MOX

4 ((1-YU235 -Y ) MU23+YU235 MU2 3s+ YP Mp)+8 (+ )-MOP(U.PU)0 2 = 3 Eq. 10-5

'(U, Pu)0 2 N

; ; = l FCF Non ProprGetary

E R , C Chapter 10 PAGE 10-2

where

b., c., d.l Eq. 10-6

10.1.6. U0 2-Gd203

4 YU235 Ygadod MU2 38s YU235 MU2 3 5+ Ygado Mgado)+8 2 (i +2).

PU02 - Cd2 03

IIUO 2- Gd 203 N

Eq. 10-7

where

Ib., c., d.] Eq. 10-8

a,L1

08

0

I-U.

10.2. Meltine Point

10M1. Notations and Units

Tic : melting temperature in °C

Tf3 MOX melting temperature in °C at zero burnup

Bu burnup in MWd/tU

BUM burnup in MWd/tM

y plutonium content in weight fraction

10.2.2. Zircaloy

Tf = 1850 Eq. 10-9

10.2.3. Advanced Alloy S Cladding

T = 1850fc Eq. 10-10

. G FCF Non Propuetary

Chapter 10 PAGE 10-3

10.2.4. U02 , U0 2 -Gd2O3

The melting point is given by:

[b., c., d.] Eq. 10-11

10.2.5. MOX

The melting point is given by the following equations:

[b., c., d.] Eq. 10-12

[b., c., d.] Eq. 10-13

10.3. Thermal ExPansion


Ax/x : thermal expansion (fractional)

Do diameter at [e.]

C b: length at [e.]

TC : temperature in °C

Tk : temperature in K

LU e0 : thickness at [e.]

10.3.2. Zircaloy

The thermal expansion values (from 20 to 800 °C) for stress-relieved Zircaloy-4 cladding are:

-6(radial) = 9.62* 106* (T -20) Eq. 10-14co C

AD (circumferential) = 7.40.10 *(T -20) Eq. 10-15D 0 C

FL(axial) = 5.49.6 10 (T, - 20) Eq. 10-16I LO.. ,. I

*jfk~, '> FCF Non Propietary

n e eChapter 10 PAGE 10-4

IU.i.3. Advanced Alloy 5 Claading

The thermal expansion expressions (in %o) for the advanced Alloy S cladding are:

Eq. 10-17

Eq. 10-18

Eq. 10-19

Eq. 10-20

Eq. 10-21

Eq. 10-22

to-j

03-

IL.

[b., c., d.J

Eq. 10-23

Eq. 10-24

Eq. 10-25

a

FOF Non Proprietary

. o N Chapter 10 PAGE 10-5

10.3.4. U0 2, MOX, U02 -Gd2O3

The thermal expansion (from 273 K to 3120 K) of uranium dioxide is:

AL=_0.00328+I.179 105 *Tk- 2 .4 2 9*10 9 Tk +1.219*10 7 TL e ik

The thermal expansion is the same for U0 2, MOX and U02-Gd 2O3 fuels.

Eq. 10-26

10.4. Thermal Conductivity


2 : thermal conductivity (W.m~l K-1)

Bu : burnup (MWd/t)

Tc temperature (°C)

Tk temperature (K)

kB Boltzmann constant = 8.6144 - eV.KI

S Zirconia thickness (m)

p : porosity (fraction)

x : deviation from stoichiometry

y : plutonium content (weight fraction)

z : gadolinia content (weight %)

0)-J

2

'U

U-

10.4.2. Zircaloy

The thermal conductivity for stress-relieved Zircaloy-4 cladding is:

0 T¢S455 )(Tc) = 12.291 +0.0I08-TC

455 S TC S 1049 X(TC) = 9.925 + 0.0160 -TC

1049S TC S 1500 X(TC) = - 7.067 + 0.0322. TC

Eq. 10-27

Eq. 10-28

Eq. 10-29

< -'-.... '' .. FCF Non Proprietary

E R ! Chapter 10 PAGE 10-6

1OA43. Advanced Alloy 5 Cladding

The thermal conductivity for the advanced Alloy 5 cladding is:

lb., c., d.] Eq. 10-30

10.4.4. Zlrconia

Under PWR operating conditions, the thermal conductivity of zirconia (ZrO2) is:

[b., c., d.]

[b., c., d.]

[b., c., d.)

lOA.5. U02

The thermal conductivity for 100% dense U0 2 between 273 K and 3120 K is:

[b., c., d.]

Eq. 10-31

Eq. 10-32

Eq. 10-33

0

0 Eq. 10-34

The following correction for porosity is used (Ref. 10-1):

A = 1ioo (I'- p) where a = 2.58-S.* 0 1 .4*T Eq. 10-35

I

-FCF Non ProprietaryChapter 10 PAGE 10-7

10.4.6. MOX

Eq. 10-36

[b., c., d.]

10.4.7. UOGd2O3

(b., c.,d.I Eq. 10-37

0-J

8w

a:UL.

tb., c.,d.] Eq. 10-38

10.5. S~eC~flc Heat


: constant = 535.85 K

ED constant= 157.7707 103 j.mol 1

K: constant = 80.1314 J.mo1.K 1

K2 : constant = 32.845 104 J.mO11.K 2

K3 constant= 23.62183 106 J.mnol-l

R ideal gas constant = 8.314 J.mol'.KI'

TC : temperature (C)

.-;- ,- . Tt 1.., I FCF Non Propretary/ r Chapter 10 PAGE 10-8~§.. .j?.

Tk temperature (K)

Cm : specific heat (Jmol-1.K71)

CpC : specific heat (J.g1 .'C1 )

Cpk : specific heat (J.g-I.K 1 )

10.5.2. Zircaloy

The specific heat values for Zircaloy-4 cladding are:

05TC5570 c = 0.28796+8.67486 107 Tc Eq. 10-39

570oTc!825 c = 0.3632 Eq. 10-40

8255T¢ S957 c = 0.6970 Eq. 10-41

957:STc: 1500 c = 0.3597 Eq. 10-42

10.5.3. Advanced Alloy 5 Cladding

The specific heat values for the advanced alloy 5 cladding are:

Eq.. 10-43

Eq.10-44

W [b., c., d.]0

Eq. 10-45

Eq. 10-46

10.5A. U0 2, MOX, U0 2-Gd2O3

The specific beat of uranium dioxide is:

I' kf dxP( K3 E D -E D Eq10429 1STk:S3 120 cm =-2 1+)2 K +2K 2Tk _-. R Eq 10-47

T ske.(Cxp(ihet-)- R .Tkf , MO a

The specific beat is the same for U0 2, MOX and U0 2-Gd203.

f 02N .FCF Non Proprietary

w 10.6. Emissivitv


: emissivity

Tk temperature (K)

TC : temperature (°C)

10.6.2. Zircaloy

The emissivity for stress-relieved or recrystallized Zircaloy-4 cladding is:

E = 0.808 Eq. 10-48


The emissivity for advanced Alloy 5 cladding is:

oi [b., c., d.] Eq. 10-49

u.

8 10.6.4. U0 2 , MOX, U02 -Gd 2O3

0 The uranium dioxide emissivity (Ref 10-2) is:

Tk5 1O0 E 0.8707 Eq. 10-50

IL-4IOODSTk 2050 £= 1.311-4.404-10 *Tk Eq. 10-51

Tk22 050 £ = 0.4083 Eq. 10-52

The emissivity is the same for U02, MOX and U0 2-GdO 3 fuels.

FCF Non Proprietary

C I _

10.7. Young's Modulus


E Young's modulus (GPa)

Ep : Youngs' modulus (GPa) at a given porosity p

Tk temperature (K)

TC temperature (TC)

p porosity (fraction)

10.7.2. Zircaloy

Young's modulus for Zircaloy-4 cladding ([b., c., d.]) is:

[b., c., d.] Eq. 10-53

10.7.3. Advanced Alloy S Cladding

Young's modulus for the advanced Alloy 5 cladding ([b., c., d.]) is:

I[b., c., d.] Eq. 10-54

Di

10.7.4. U02, MOX, U02 -Gd2O3

The adiabatic Young's modulus of stoichiometric uranium dioxide at 100% theoretical density(Elco in Pa), is given by the relationships:

STc 2337 EIlOO= 222.46-1.8749.10 *TC -9.74610 *T Eq. 10-55

2337!5 TC S 2656 E100 =-1146.24 + I15.25S8*10 2 *TC - 260.352*106F6 *TC2 Eq. 10-56

Porosity effects are taken into account with:

EP = EIO-(I-25-S-P) Eq. 10-57

Young's modulus is the same for U0 2, MOX and t30 2 7d 2 03 fuels.

I


10.8. Poisson's Ratio

1Q8.1. Notations and Units

v Poisson's ratio

E . Young's modulus (GPa)

G : Coulomb's modulus (GPa)

Gp : Coulomb's modulus (GPa) at a given porosity p

T : temperature (°C)

p porosity (fraction)

10.8.2. Zircaloy

Poisson's ratio for stress-relieved or recrystallized Zircaloy4 cladding between [b., c., d.] is:

lb., c., d.] Eq. 10-58


Poisson's ratio for the advanced Alloy 5 cladding between [b., c., d.] is:

[b., c., d.] Eq. 10-59

10.8A. U0 2, MOX, U02-Gd2G 3

Poisson's ratio for uranium dioxide is:

,2* I Eq. 10-60

Coulomb's modulus of stoichiometric uranium dioxide at 100% of theoretical density (GjOg inGPa), is:

OS5 TSG2337 0 = 84.14-0.7203-10 *T-3.747-0 . -IT2 Eq. 10-61

2337S T!5 2656 100 = -446.6+447076*102 *T - 100.939 *10 * T Eq. 10-62

Porosity effects are taken into account with:

Gp = G 1O -(I - 2 .2 5 p) Eq. 10-63

Poisson's ratio is the same for U02 , MOX and U0 2-Gd2 O3 fuels.

2

0

I

FCF Non ProprietaryCf = P

:; 41k.s I.:r.f< Chapter 10 PAGE 10-12

10.9.0.2% Yield Strength


: fluence (1021 n/cm2)

RPO.2 : 0.2% yield strength (MPa)

T temperature (°C)

10.9.2. Cold Worked Stress Relieved Zircaloy

[b., c., d.]

[b., c., d.]

Eq. 10-64

Eq. 10-65

[b., c.,d.)

[b., c., d.] Eq. -10-66

a

w

8

10.93. Advanced Alloy 5 Cladding

[b., c., d.]

[b., c., d.]

Eq. 10-67

Eq. 10-68

[b., c., d.]

[b., c., d.] Eq. 10-69

10.10. Inltinate Tensile Streneth


CR

T

fluence (1021 n/cm2 )

ultimate tensile strength (MPa)

: temperature (°C)

fM C FCF Non Proprietary

Chapter 10 PAGE 10-13I _ __--


[b., c., d.] Eq. 10-70

[b., c., d.] Eq. 10-71

[b., c., d.]

lb., c., d.] Eq. 10-72


[b., c.,d.]

[b., c., d.]

Eq. 10-73

Eq. 10-74

[b., c., d.]

[b., c., d.] Eq. 10-75~0~11

w

ILI

10.11. Uniform Elongation


4) fluence (1021 n/cm2 )

8 uniform elongation (fraction)

T temperature (0C)


[b., c., d.] Eq. 10-76

l , ' ' ,FCF Non Proprietary

J. .'. Chapter 10 PAGE 10-14


[b., c., d.] Eq. 10-77

[b., c., d.] Eq. 10-78

[b., c., d.]

(b., c., d.] Eq. 10-79

0)

w0

E

REFERENCES

Ref. 10-1 J.C. VAN CRAEYNEST, J. P. STORA, Effet de la porosite sur la variation de iaconductibilit6 thermique du bioxyde duranium en fonction de Ia temperature, JNM37 (1970), pp 153-158.

Ref. 10-2 R.E. MASON, Matpro-Version 11 (Rev.2), A Handbook of Material PropertiesforUse in the Analysis of Light Water Reactor Fuel Rod Behavior, NUREG/CRO497,TREE-1280, Rev.2, US NRC (Aug. 1981).

80.

I-

FCF Non Proprietary

J , Chapter 10 PAGE 10-16

-

0Uw

I


,m c a rP'L~ -~r-c FCF Non ProprietaryChapter 11 PAGE 11-1

11. USER'S GUIDE

I .I. General

lb., d.]

LIq

w

0

LU

1., ;FCF Non Proprietary

f Chapter 11 PAGE 11-2

1 1.2. Input Description

[b., d.]

w

8LQ0

I

r oFCF Non Propietary

1 Chapter 1 1 PAGE 1 1-3 ]11.3. Output Descriptlon

11.3.1. Files

[b., d.]

2

I0I-

113.2. Code Error Messages

[b., d.]

11.3.3. Code Performance

[b., d.) Eq. 11-1

J. *::o; ; ? - , , .FCF Non Proprietary

_ Chapter 11 PAGE 11-4

-J

0

I-


FCF Non Proprietaryf .COERNIm Chapterl_ PAGE11-5

FIGURES

w0

-.T '^,j'i , FCF Non Proprietary


FIGURE 11-1 PROCEDURE TO RUN COPERNIC

'U

2

m00

U.I2

[b.)

I

rm COP nlRHI;FCF Non Proprietary

0 FIGURE 11-2 SAMPLE INPUT DECK

0

I-O.2.

Chapterli PAGE 11-8

0

C.,

ui

[bJ

_"P R I FCF Non Proprietaryrm I$60 Pw X Chapter 1 1 PAGE 1 1-9

22

08w0

U-.

[b.]

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TABLES

0


TABLE 11-1 INPUT VARIABLE DESCRIPTION

COPARNIC List of input variables

Name Group |Din. VDesciption DOult Unit

[b.], Vb

01~-AC

CD 0

G)

M ,

-U

C>.:t, 'm.s''eD-] fii~-L6s!

_ _. * _._ hi

1 )

_ FRAMATOME COGEMA FUELS

COPERNIC List of input variables.N p

. a]es


COPERNIC List of input variables

Same .Gt'o | , Din . : Deslpi

[b.J



N.]



Nanie.: GtiuPI: Dhii l: H iPI

[D]

PRAMATOME COGEMA FUELS


[b.

. _ . . .



[b.J

_ _



[b.]



NieiGk~u-: Dfiii. Di

[b.]



[b]



.Nam *eGmtp | flj 7 . DesIpl

[b.J


COPERNIC Ust of input variables

Mb.]



Ninie " I P1 ID m.. -|

[b.]



[b.]



.N & i | G r u p . | P i . : * . : : N e r p t o : :: A u t : it

1 -

0-o

X

0b _i

|| {Adi?

L~-

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[b.]



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12. APPLICATION METHODOLOGY (UNITED STATES)

Fuel performance codes are used in a wide variety of disciplines: thermal, mechanical, nuclear andmaterial. The COPERNIC code is no exception. It will be used primarily as a design tool for lightwater reactor fuel rods. The following design criteria will be used with the COPERNIC code toverify fuel rod designs.

Fuel Rod Internal Gas Pressure: the internal gas pressure of the peak fuel rod in the reactorwill be limited to a value below that which would cause (1) the fuel-cladding gap to increasedue to outward cladding creep during steady-state operation and (2) extensive DNBpropagation to occur.

LOCA Initialization: LOCA initialization predictions will be input into LOCA evaluationmodels that are used to verify two principal LOCA criteria: (1) fuel rod fragmentation mustnot occur as a direct result of the blowdown loads, and (2) the 10 CFR Part 50 temperatureand oxidation limits must not be exceeded.

Fuel Melting: fuel melting during normal operation and anticipated operational occurrencesis precluded.

Cladding Strain: the maximum uniform hoop strain (elastic plus plastic) shall not exceed I%;steady-state creep-down and irradiation growth are excluded.

Creep Collapse Initialization: cladding collapse is precluded during the fuel rod design life.

Cladding Peak Oxide Thickness: the cladding peak oxide thickness shall not exceed a best-estimate predicted value of 100 microns.

wThese design criteria satisfy the fuel cycle review recommendations defined in Reg. Guide 1.70(4A.41) and the licensing requirements defined in 10 CFR 50.46 and SRP 4.2.

0 The COPERNIC code will also be used to provide data for analyses that have no explicit basis in theE regulations. These include best-estimate fuel temperatures fornuclearanalysis codes such as NEMO

(Ref. 12-1) and initialization data for core thermal-hydraulic codes such as LYNXT (Ref. 12-2). TheB COPERNIC code will also provide best-estimate fuel performance predictions for other similar

analyses as appropriate.

The manner in which the COPERNIC code will be applied to the fuel rod design criteria is discussedbelow. These analyses are applicable to uranium dioxide and urania-gadolinia fuel with low tinZircaloy-4 and advanced alloy M5 cladding.

12.1. Fuel Rod Internal Gas Pressure

12.1.1. Fuel Rod Internal Gas Pressure Methodology

The COPERNIC code will be used to predict fuel rod internal gas pressures that are used to verifythe fuel rod internal gas pressure design criteria are met. The following analysis method will be used.

Bounding steady-state internal gas pressures will be determined from COPERNIC internalgas pressure predictions. These bounding pressures will be used with an approved fuel rodgas pressure criterion to determine the limiting internal gas pressure that will result in theonset of fuel-clad lift-off. The Fuel Rod Gas Pressure Criterion (Ref. 12-3) was approvedfor Zircaloy4 cladding July 1995 and extended to advanced alloy MS cladding November1999 (Ref. 124). The bounding pressure used in this analysis is composed, at any giventime-in-life, of a COPERNIC best-estimate predicted pressure plus a pressure uncertaintyallowance. The pressure uncertainty allowance is composed of a COPERNIC codeuncertainty allowance and allowances due to fuel rod manufacturing variations.

[b.]

Steady state and transientpower history effects will be evaluated with COPERNIC. The treatment of codeuncertainties, manufacturing variations, power histories, and transients is described indetail below.

Code Uncertainties: The COPERNIC code contains

[b.]

Nominal fuel rod designcharacteristics and thermal-hydraulic conditions will be used in these evaluations that aresimilar to those listed in Tables 12-1 through 12-6 and Tables 12-7 through 12-10,

= respectively.

Manufacturinst Variations: The effects of fuel rod manufacturing variations will beincluded in the pressure uncertainty allowance. COPERNIC best-estimate cases will be

us run with orninal fuel rod design characteristics0

[b.]

The COPERNIC cases used to generate the pressure allowances described above willcontain cladding oxide formation and the additional following characteristics.

Power History: The rod average power history selected for the COPERNIC internal gaspressure analyses will vary according to the stage of the fuel design process the analysis issupporting.

fr C4-,ME&RNTCI FCF Non Proprietary


Fuel rod design analyses, for example, will be performed with

lb.]

Applications analyses performed to support fuel reload operations will

[b.]

The power histories used in the COPERNIC internal gas pressure analyses will containtransient effects which are defined below.

Transients: The power histories discussed above will include both Condition-I andCondition-II transients. A transient is defined as a temporary change in the local powerlevel of a fuel rod.

0i-

A Condition-I transient may occur during normal operation or the maneuvering of a plant.Condition-I design transients will be used in the fuel rod internal gas pressure analyses. ACondition-I design transient bounds all transients that are expected to occur during normaloperation. lb.] Condition-I design transients will be included in the COPERNIC powerhistories. [b.]

In addition, if the power history contains regions of low poweroperation (such as reactor coast-down), the Condition-I design transients will be placed atthose times in life that are at or near full power operation.

[b.]

This method of defining the Condition-I transients for the fuel rod internalgas pressure analyses will be applied to both uranium-dioxide and urania-gadolinia fuelrods.

Lb]

A Condition-Il transient or Anticipated Operational Occurrence (AOO) is an event ofmoderate frequency. These events may result in a reactor trip but the plant will be capableof returning to operation. These events, by definition, will not propagate to cause a moreserious event such as a Condition-II or IV event and are not expected to compromise thefuel rod integrity or cause an over-pressurization of the reactor coolant or secondarysystems. The limiting power distributions that could occur during those Condition-IItransients that would result in a fuel rod internal gas pressure increase [b.]

~~~~F i P.ACF Non Pr~oprietaryChapter 12 PAGE 12b4

This method of defining the Condition-II transients for the fuel rod internalgas pressure analyses will be applied to both uranium-dioxide and urania-gadolinia rods.

12.1.2. Fuel Rod Internal Gas Pressure Examples

Six typical fuel rod designs will be used for all examples presented in this section. These fuel roddesigns were selected from two uranium-dioxide and two urania-gadolinia typical fuel cycle designs.

Mark-B, Uranium-Dioxide [b.],

Mark-B, Urania-Gadolinia, [b.],

Mark-B, Urania-Gadolinia, [b.],

Mark-BW17, Uranium-Dioxide [b.],

2 EMark-B WI7, Urania-Gadolinia, lb.], andwCD8 Mark-BW17, Urania-Gadolinia, [b.].

0

| Mark-B and Mark-BW17 are fuel rods designed for the Babcock & Wilcox (15 x 15) andWestinghouse (17 x 17) plants, respectively. The fuel rod characteristics and thermal hydraulicconditions for typical examples of these fuel rods are listed in Tables 12-1 through 12-6 and Tables12-7 through 12-10, respectively. Note that these tables contain [b.]

[b.] shown in Figures 12-1 through 12-6 were selected for the fuel rodinternal gas pressure methodology examples. These figures also contain (b.]

that are predicted to have the greatest end-of-life internal gas pressures. Thetransients Ab.]

that are applied to the power histories are presented in Tables12-11 through 12-16. These tables contain the following information for each transient: the fb.]

were determined to be necessary for these

rf COPWX\RUfic I FCF Non Proprietary


examples based upon the manufacturing variations criteria presented above: [b.]

The bounding fuel rod internal gas pressures that were predicted using themethodology defined above are shown in Figures 12-7 through 12-12. The best-estimatc internal gaspressure predictions which were generated [b.] (Figures 12-1 through 12-6) are also shown inFigures 12-7 through 12-12. The best-estimate cases were run with nominal fuel rod characteristics(Tables 12-1 through 12-6) and thermal-hydraulic conditions (Tables 12-7 through 12-10) andwithout the additional transients (Tables 12-11 through 12-16) and manufacturing variations (Tables12-1 through 12-6) that are necessary for the bounding analysis. Figures 12-7 through 12-12illustrate the internal gas pressure differences that may be expected to exist between best-estimateand bounding COPERNIC cases.

12.2. LOCA Initialization

12.2.1. LOCA Initialization Methodology

The COPERNIC code will be used to generate LOCA initialization predictions. These predictionswill be used for LOCA evaluation models such as BWNT LOCA EM (Ref. 12-5) for B&W plantsand RSG LOCA EM (Ref. 12-6) for Westinghouse plants.

[b.]

Eq. 12-1

[b.]

The following method will be used to generate the LOCA initialization predictions.

Nominal fuel rod characteristics and thermal-hydraulic conditions will be used that aresimilar to those listed in Tables 12-1 through 12-6 and Tables 12-7 through 12-10,respectively. The COPERNIC predictions will include cladding oxide formation. The rodswill be analyzed with

at

uwU-

0

0

lb.]

FC onPoritr

12.2.2. LOCA Initialization Examples

Examples of the [b.] , obtained usingthe LOCA initialization methodology described above are presented for six typical fuel rod designsin Figures 12-21 through 12-26. Note that the power levels of the urania-gadolinia rods are initiallyrestrained by the presence of the unburned gadolinia. [b.]

were used for these examples.

123. Fuel Melt

123.1. Fuel Melt Methodology

The COPERNIC code will be used to predict the linear heat rates where the onset of fuel centerlinemelting occurs. Fuel melting is not permitted during normal operation or anticipated operationaloccurrences.

Centerline fuel melt analyses will be performed with COPERNIC best-estimate predictions andnominal fuel rod design parameters. [b.]

The best-estimate fuel melt temperature relationship from Chapter4 is:

Wat

w0

A:us

i)

[b.]

where:Tm = best-estimate centerline fuel melt temperature, 0C, and

Eq. 12-2

Bu pellet burnup, GWd/tU.

(b.]

fr COPRjICv I FCF Non Proprietary


(b] Eq. 12-3

The following fuel melt analysis method will be used.

Nominal fuel rod characteristics and thermal-hydraulic conditions will be used that are similar tothose listed in Tables 12-1 through 12-6 and Tables 12-7 through 12-10, respectively. TheCOPERNIC predictions will include cladding oxide formation. The COPERNIC cases will be run[b.]

12i3.2. Fuel Melt Examples

Examples of the local linear heat rate predictions obtained [b.]are presented for the six typical fuel rod designs in Figures 12-28 through 12-33. [b.]

were used for these examples.

12.4. Cladding Strain

12A.1. Cladding Strain Methodology

The COPERNIC code will be used to predict the local linear heat rates at which the cladding uniformhoop strains equal 1%. Cladding uniform hoop strain is limited to 1% during normal operation oranticipated operational occurrences. Cladding strain analyses will be performed with COPERNICbest-estimate predictions and nominal fuel rod characteristics across the range of operationalburnups.

The cladding uniform hoop strains [b.]The induced strain, therefore, is defined as:

'Iq

0

[b.] Eq. 12-4

w ,':FC No Ppeary


where:

4,oop = cladding uniform hoop strain,

[b.]

Cladding strain analyses will be performed in the following manner.Nominal fuel rod characteristicsand thermal-hydraulic conditions will be used that are similar to those listed in Tables 12-1 through12-6 and Tables 12-7 through 12-10, respectively. The COPERNIC predictions will includecladding oxide formation. The COPERNIC cases will be run

(b.]

12.4.2. Cladding Strain Examples

Examples of the linear heat rates obtained where the cladding uniform hoop strain is 1% arepresented for the six typical fuel rod designs in Figures 12-28 through 12-33.

[bD

CO

I.-

12.5. Creep Collanse Initialization

12.5.1. Creep Collapse Initialization Methodology

The COPERNIC code will be used to generate cladding creep collapse initialization predictions.These predictions will be input into cladding creep collapse codes such as CROV (Ref. 12-7).Cladding collapse is not permitted during the fuel rod design life.

Cladding creep collapse predictions will be generated. Conservatism will be incorporated in theseanalyses [b.]

COPER\,TICFCF Non Proprietary


The following method will be used for generating the creep collapse predictions.

Nominal fuel rod characteristics and thermal-hydraulic conditions will be used that aresimilar to those listed in Tables 12-1 through 12-6 and Tables 12-7 through 12-10,respectively. The COPERNIC predictions will include cladding oxide formation andthese cases will be run [b.].

Creep collapse evaluations will be performed during the fuel rod design process. Thepower history envelopes selected for these evaluations would be expected to bound theoperating power levels of all future fuel cycle designs without introducing excessiveconservatism into the design process. The validity of these envelopes will be verified aspart of the reload design analysis.

12.5.2. Creep Collapse Initialization Examples

Examples of the fuel rod internal gas pressures obtained with the creep collapse initializationmethodology defined above are presented in Figures 12-34 and 12-35. The pressure differencesbetween the uranium-dioxide and urania-gadolinia fuel rod designs shown in these figures illustratethe effect that the power histories have on the predicted creep collapse initialization internal gaspressures. [b.] that were used for these examples are shown in Figures 12-36 to 12-41.

12.6. Cladding Peak Oxide Thickness

12.6.1. Cladding Peak Oxide Thickness MethodologyU.

Ma The COPERNIC code will be used to generate cladding peak oxide thickness predictions. The peak

C. cladding oxide thickness will not be allowed to exceed a best-estimate predicted value of 100microns.

The following method will be used to generate the cladding peak oxide thickness predictions.

Best-estimate values will be used for all predictions. Nominal fuel rod characteristics andthermal-hydraulic conditions will be used, similar to those listed in Tables 12-1 through12-6 and Tables 12-7 through 12-10, respectively. (b.]

A sub-batch isdefined as fuel assemblies within a given fuel batch that have the same make-up (fuel roddesigns, number and placement of urania-gadolinia rods, etc.) and that are inserted anddischarged from the core at the same time so that the fuel assembly residence times areidentical.

[b.]

f . ;g N |FCF Non Proprietary

(b.]

The COPERNIC cladding oxide model was developed from European PWR data. The fuelrod designs for these reactors are generally similar to those used by FCF. The fuel cycledesigns and cycle lengths of European reactors, however, often differ significantly fromUnited States reactors.

[b.]

to ensure that the model provides best-estimate predictions at the 100 micronlevel.

12.6.2. Cladding Peak Oxide Thickness Examples

Examples of the COPERNIC cladding peak oxide predictions obtained are presented in Figures 12-42 through 1247. These examples contain the predictions for both low-tin Zircaloy-4 and M5advanced alloy claddings, and illustrate the cladding oxide thickness margin gains obtained with theMS advanced alloy cladding. (b.] used for these examples areshown in Figures 12-36 through 1241.

(0

tLi

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FCF Non Proprietary

_ Chapter 12 PAGE 12-11

REFERENCES

Ref. 12-1 BAW-10180-A,Rev. 1, NEMO- NodalExpansion MethodOptimzued, March 1993.

Ref. 12-2 BAW-10156-A, Rev. 1, LYNXIT - Core Transient Thermal-Hydraulic Program,August 1993.

Ref 12-3 BAW-I01 83-A, Fuel Rod Gas Pressure Criterion (FRGPC), July 1995.

Ref. 12-4 BAW-10227P-A, Evaluation of Advanced Cladding and Structural Material (M5) inPWR Reactor Fuel, November 1999.

Re. 12-5 BAW-10192P, BWNT Loss-Of-Coolant-Accident Evaluation Model for Once-Through Steam Generator Plants, Februaly 1994.

Ref 12-6 BAW-10168P-A, Rev. 3, B&W Loss-Of-Coolant-Accident Evaluation Model forRecirculating Steam Generator Plants, December 1996.

Ref. 12-7 BAW-10084P-A, Rev. 3, Program to Determine In-Reactor Performance of BWFCFuel Cladding Creep Collapse, July 1995.

w

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This page int l lG b l

This page intentionally left blank.-m

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FIGURES

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FIGURE 12-1 Typical Mark-B Uranium-Dioxide Cycle

[b.I

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FIGURE 12-2 Typical Mark-B Urania-Gadolinia Cycle

Ib.]

9

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.Chapter 12 . PAGE 12-16


[b.]

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FIGURE 12-4 Typical Mark-BW17 Uranium-Dioxide Cycle

[b.]

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U.'

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FIGURE 12-5 Typical Mark-BW17 Urania-Gadolinia Cycle

[b.]

[b.]

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FIGURE 12-6 Typical Mark-BW17 Urania-Gadolinla Cycle

[b.]

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[b.]

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[b.]


FIGURE 12-9 Typical Mark-B Urania-Gadolinfa Cycle

[b.

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[b.]

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[b-]

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FIGURE 12-13 Typical Mark-B Uraniulm-Dioxide Cycle_

b.]

aN-

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| Chapter 12 PAGE 12-27.


[b.D

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Chapter 12 PAGE 12-28 l

FIGURE 12-15 Typical Mark-B Uranla-Gadolznia Cycle

[b.]

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[b.]

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.. Chapter 12 PAGE 12-30


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lb.)

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FIGURE 12-19 Typical Mark-B Fuel Cycles

:2 [b.J

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FIGURE 12-20 Typical Mark-BW17 Fuel Cycles

[b.]

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I Chapter 12, PAGE 12-35

FIGURE 12-22 Typical Mark-B Urania-Gadolnia Cycle

[b.]

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FIGURE 12-27 Typical Mark-B and Mark-BW17 Fuel Cycles

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FIGURE 12-30 Typical Mark-B Urania-Gadoilnia Cycle

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[Chapter 12 PAGE 12-47

FIGURE 12-34 Typical Mark-B Fuel Cycles

[1.-

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LU0I-

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Chapter 12 PAGE 1248

FIGURE 12-35 Typical Mark-BW17 Fuel Cycles

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[ Chapter 12 PAGE 12-49


[b.]

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.Chapter 12 PAGE 12-50


[b.)

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LChapter 12 PAGE 12-53


lb.]

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[b-1

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(b.]

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FIGURE 1243 Typical Mark-B Urania-Gadolinia Cycle

[b.

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lb.]

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FIGURE 12-45 Typical Mark-BW17 Uranium-Dioide Cycle

[b.1

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FIGURE 1246 Typical Mark-BW17 Urania-Gadolinia Cycle

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. ................ Chapter 12 PAGE 12-60

FIGURE 1247 Typical Mark-BW17 Urania-Gadolinia Cycle

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-Chapter 12 PAGE 12-61

TABLES

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IL

*Chapter 12 PA'GE 12-62

TABLE 12-1 Typical Mark-B Uranium-Dioxide Fuel Cycle

Fuel Rod Characteristics

E ,

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,m e'i , . y f?-



TABLE 12-2 Typical Mark-B Urania-Gadollnia Fuel Cycle


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TABLE 12-3 Typical Mark-B Urania-Gadolinia Fuel Cycle


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1IChapter 12 PAGE 12-65

TABLE 12-4 Typical Mark-BW17 Uranium-Dioxide Fuel Cycle


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Thermal-Hydraulic Conditions

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TABLE 12-9 Typical Mark-BW17 Uranium-Dioxide Fuel Cycle


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TABLE 12-12 Typical Mark-B Urania-Gadolinia Cycle

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/ Chapter 12 PAGE 12-78

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_KChapter 13 PAGE 13-1

v Pi 13. MOX APPLICATION METHODOLOGY (UNITED STATES)

The U.S. Department of Energy (DOE) has recommended that a significant portion of the nation'ssurplus weapons-grade (WG) plutonium be disposed of by reconstituting the plutonium intomixed-oxide (UO2-PuO2, MOX) fuel rods and irradiating them in commercial light waterreactors. The COPERNIC code, developed utilizing the extensive European experience withreactor-grade (RG) MOX fuels, will be used to perform fuel performance analyses for MOX fuelwith weapons-grade plutonium in support of this Material Disposition Program.

The MOX fuel produced from weapons-grade material will be virtually identical to the fuelproduced from reactor-grade material in terms of physical characteristics and performance. Themanufacturing processes, plutonium isotopics, impurities, and pellet microstucture will becontrolled to ensure this equivalence. The fabrication process will use the COGEMA/BELGONUCLEAIRE-developed MIcronized MASter blend (MIMAS) process currentlysupplying MOX fuel to 32 reactors in Europe. The use of WG plutonium will significantly reducethe PuO2 content of MOX fuel relative to RG material. The WG material is about 95% fissile,whereas the RG material contains significant amounts of absorber isotopes (Pu-240, Pu-242).Thus, MOX fuel from WG material will require Pu contents of 4% to 5% instead of the 8% to 9%for RG MOX. Gallium and other impurities will be effectively eliminated through the use of anaqueous polishing process step added to the manufacturing process being used to produce the WGMOX fuel. Due to the different isotopics, the WG material will have a fissile plutonium contentabout 50% greater than that of the RG plutonium. However the master mix of U0 2 and PuO 2 Willbe adjusted from the 70/30 ratio typical of RG material, to 80/20 for the WG material to ensurethat the fissile content of the plutonium-rich particles remains the same as the reactor gradematerial. Since the fission density and thus, the fission product concentration and distribution,

w will be comparable to the RG fuel, the WG fuel behavior will be consistent with that of theL_ European experience base.

wa The COPERNIC fuel performance code will be used primarily as a design tool for light water08 reactor fuel rods - both low enriched uranium (LEU) as discussed in Chapter 12 and mixed oxideW as discussed in this chapter. This chapter prescribes the application methodology and presents0

example calculational results of the COPERNIC code applied to MOX fuel. The same designcriteria applied to LEU fuel in Chapter 12 will be used with the COPERNIC code to verify the

It acceptable performance of MOX fuel rod designs, namely:

Fuel Rod Internal Gas Pressure: the internal gas pressure of the maximum pressure fuelrod in the reactor will be limited to a value below that which would cause (1) the fuel-cladding gap to increase due to outward cladding creep during steady-state operation and(2) extensive DNB propagation to occur.

LOCA Initialization: LOCA initialization predictions will be input into LOCA evaluationmodels that are used to verify two principal LOCA criteria: (1) fuel rod fragmentationmust not occur as a direct result of the blowdown loads, and (2) the 10 CFR Part 50temperature and oxidation limits must not be exceeded.

Fuel Melting: fuel melting during normal operation and anticipated operational

FCF Non Proprietary

v ,o 44 4 A Chapter 13 PAGE 13-2

occurrences is precluded.

Cladding Strain: the maximum uniform hoop strain (elastic plus plastic) shall not exceed1%; the impact of steady-state creep-down and irradiation growth is excluded.

Creep Collapse Initialization: cladding collapse is precluded during the fuel rod design life.

Cladding Peak Oxide Thickness: the cladding peak oxide thickness shall not exceed a best-estimate predicted value of 100 microns.

These design criteria satisfy the fuel cycle review recommendations defined in Reg. Guide 1.70(4.4.1) and the licensing requirements defined in 10 CFR 50.46 and SRP 4.2.

The COPERNIC code will also be used to provide data for analyses that have no explicit basis inthe regulations. These include best-estimate fuel temperatures for nuclear analysis codes such asNEMO (Ref. 13-1) and initialization data for core thermal-bydraulic codes such as LYNXT (Ref13-2). The COPERNIC code will also provide best-estimate fuel performance predictions for othersimilar analyses.

The manner in which the COPERNIC code will be applied to the fuel rod design criteria isdiscussed below.

13.1. Fuel Rod Internal Gas Pressure

13.1.1. Fuel Rod Internal Gas Pressure Methodologyco-jw2 The COPERNIC code will be used to predict fuel rod internal gas pressures that are used to verify

that the fuel rod internal gas pressure design criteria are met. The following analysis method whichLU is consistent with that described in Chapter 12 for U0 2 fuel will be used.pu Bounding steady-state internal gas pressures will be determined from COPERNICo internal gas pressure predictions. These bounding pressures will be used with an

approved fuel rod gas pressure criterion to determine the limiting internal gas pressurethat will result in the onset of fuel-clad lift-off. The Fuel Rod Gas Pressure Criterion(Ref 13-3) was approved for Zircaloy4 cladding in July 1995 and extended to advancedalloy M5 cladding in November 1999 (Ref. 134). The bounding pressure used in thisanalysis is composed, at any given time-in-life, of a COPERNIC best-estimate predictedpressure plus a pressure uncertainty allowance. The pressure uncertainty allowance iscomposed of a COPERNIC code uncertainty allowance and allowances for fuel rodmanufacturing variations.

.b.]

Steady state and transient power history effects will be evaluated withCOPERNIC. The treatment of code uncertainties, manufacturing variations, powerhistories, and transients is described in detail below.

f~r c�i¶Ic IFCF Non Proprietary

Chapter 13 PAGE 13-3mmI4P Code Uncertainties: The COPERNIC code contains [b.)

Nominal fuel rod design characteristics and thermal-hydraulic conditions that are similarto those listed in Tables 13-1 and 13-2, respectively, will be used in these evaluations.

Manufacturing Variations: The effects of fuel rod manufacturing variations will beincluded in the pressure uncertainty allowance. COPERNIC best-estimate cases will berun with nominal fuel rod design characteristics

[b.]

IThe COPERNIC cases used to generate the pressure allowances described above willcontain cladding oxide formation and the additional following characteristics.

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Power History: The rod average power history selected for the COPERNIC internal gaspressure analyses will vary according to the stage of the fuel cycle design process theanalysis is supporting.

Fuel rod design analyses, for example, will be performed with

(b.]

Applications analyses performed to support fuel reload operations will

[b.]

The power histories used in the COPERNIC internal gas pressure analyses will containtransient effects which are defined below.

II

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Transients: The power histories discussed above will include both Condition-I andCondition-II transients. A transient is defined as a temporary change in the local powerlevel of a fuel rod.

A Condition-I transient may occur during normal operation or the maneuvering of a plant.Condition-I design transients will be used in the fuel rod internal gas pressure analyses.A Condition-I design transient bounds all transients that are expected to occur duringnormal operation. [b.] Condition-I design transients will be included in the COPERNICpower histories. [b.]

In addition, if the power historycontains regions of low power operation (such as reactor coast-down), the Condition-Idesign transients will be placed at those times in life that are at or near full poweroperation. [b.]

This method of defining theCondition-I transients for the fuel rod internal gas pressure analyses will be applied toboth U0 2 and MOX fuel rods.

(b.]

A Condition-U transient or Anticipated Operational Occurrence (AOO) is an event ofmoderate frequency. These events may result in a reactor trip but the plant will becapable of returning to operation. These events, by definition, will not propagate to causea more serious event such as a Condition-E or IV event and are not expected to

Uicompromise the fuel rod integrity or cause an over-pressurization of the reactor coolant&a or secondary systems. The limiting power distributions that could occur during those

Condition-II transients that would result in a fuel rod internal gas pressure increasew

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This method of defining the Condition-II transients for the fuel rod internal gas pressureanalyses will be applied to both U02 and mixed oxide fuel rods.

13.1.2. Fuel Rod Internal Gas Pressure Example

One typical fuel rod design is used for the example presented in this section. This fuel rod design,the Mark-BW/MOXl (non-axial blanket) design, is representative of the planned design to be

~r~r ~FCF Non Proprietary

^ =~ Chapter 13 PAGE 13-5

employed in the partial-MOX core fuel cycles and is an adaptation of the Advanced Mk-BW fuelrod design. Partial-MOX fuel cycles utilize both LEU and MOX fuel assemblies in the reactor core.

The Mark-BW/MOXI is a fuel rod designed for Westinghouse (17 x 17)-type plants. The fuel rodcharacteristics and thermal hydraulic conditions for this example are listed in Tables 13-1 and 13-2, respectively. Note that these tables contain lb.]

(b.] shown in Figure 13-1 was selected for the fuel rod internal gaspressure methodology example. [b.]

The transients [b.]that are applied to the power history envelope are presented in Table 13-3. These

tables contain the following information for each transient: the [b.]

were determined for this example based upon the manufacturingvariations criteria presented above: [b.]

The bounding fuelrod internal gas pressure that was predicted using the methodology defined above is shown inFigure 13-2.

13.2. LOCA Initialization

13.2.1. LOCA Initialization Methodology

The COPERNIC code will be used to generate LOCA initialization predictions. These predictionsL. will be used for LOCA evaluation models such as the RSG LOCA EM (Ref. 13-5) for

Westinghouse-type plants and the LOCA EM Topical Addendum for MOX fuel (to be submitted).[b.]

0

(b.] Eq. 13-1

Ine following meumod will be used to generate the LOCA initialization predictions.

- ~ v . Be ,, < | FCF Non ProprietaryChapter 13 PAGE 13-6

Nominal fuel rod characteristics and thermal-hydraulic conditions will be used that aresimilar to those listed in Tables 13-1 and 13-2, respectively. The COPERNIC predictionswill include cladding oxide formation. The rods will be analyzed with

(b.]

V

13.2.2. LOCA Initialization Example

An example of the [b.] obtainedusing the LOCA initialization methodology described above is presented for a typical MOX fuelrod design in Figure 13-4. [b.]

were used for this example.

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13.3. Fuel Melt

133.1. Fuel Melt Methodology

The COPERNIC code will be used to predict the linear heat rates where the onset of fuel centerlinemelting occurs. Fuel melting is not permitted during normal operation or anticipated operationaloccurrences.

Centerline fuel melt analyses will be performed with COPERNIC best-estimate predictions andnominal fuel rod design parameters. [b.]

The best-estimate fuel melt temperature relationship forMOX fuel from Chapter 10 is:

[b.] Eq. 13-2

4I0where:

Tm = best-estimate centerline fiuel melt temperature, 0C,

is- cOImnEic I FCF Non Proprietary

I Chapter 13 PAGE 13-7- . _

P y = plutonium content weight fraction, and

Bu = pellet bumnup, GWd/tM.

(b.]

[b.] Eq. 13-3

The following fuel melt analysis method will be used.

Nominal fuel rod characteristics and thermal-hydraulic conditions will be used that are similar tothose listed in Tables 13-1 and 13-2, respectively. The COPERNIC predictions will includecladding oxide formation. The COPERNIC cases will be run

[b.]

13.3.2. Fuel Melt Example

An example of the local linear heat rate predictions obtained [b.]is presented for the typical MOX fuel rod design in Figure 13-6. [b.]

were used for this example.

13.4. Cladding Strain

13.4.1. Cladding Strain Methodology

The COPERNIC code will be used to predict the local linear heat rates at which the claddinguniform hoop strains equal 1%. Cladding uniform hoop strain is limited to 1% during normaloperation or anticipated operational occurrences. Cladding strain analyses will be performed withCOPERNIC best-estimate predictions and nominal fuel rod characteristics across the range ofoperational burnups.4IThe cladding uniform hoop strains [b.]

FCF Non Proprietary

Chapter 13 PAGE 13-8A

wThe induced strain, therefore, is defined as:

Db.] Eq. 13-4

where:

£hoop = cladding uniform hoop strain,

(b.]

Cladding strain analyses will be performed in the following manner.

Nominal fuel rod characteristics and thermal-hydraulic conditions will be used that aresimilarto those listed in Tables 13-1 and 13-2, respectively. The COPERNIC predictionswill include cladding oxide formation. The COPERNIC cases will be run [b.]

0

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13.4.2. Cladding Strain Example

An example of the linear heat rates obtained where the cladding uniform hoop strain is 1% ispresented for the typical MOX fuel rod design in Figure 13-6. [b.]

13.5. Creep Collapse Initialization

13.5.1. Creep Collapse Initialization Methodology

The COPERNIC code will be used to generate cladding creep collapse initialization predictions.These predictions will be input into cladding creep collapse analysis codes such as CROV (Ref.13-6). Cladding collapse is not permitted during the fuel rod design life.

l

f (C O :ESS N I$ |FCF Non Proprietary_ Chapter 13 PAGE 13-9

v t Cladding creep collapse predictions will incorporate conservatism [b.]

The following method will be used for generating the creep collapse predictions.

Nominal fuel rod characteristics and thermal-hydraulic conditions will be used that aresimilar to those listed in Tables 13-1 and 13-2, respectively. The COPERNIC predictionswill include cladding oxide formation and these cases will be run [b.]

Creep collapse evaluations will be performed during the fuel rod design process. Thepower history envelopes selected for these evaluations would be expected to bound theoperating power levels of all future fuel cycle designs without introducing excessiveconservatism into the design process. The validity of these envelopes will be verified aspart of the reload design analysis.

13.5.2. Creep Collapse Initialization Example

An example of the fuel rod internal gas pressures obtained with the creep collapse initializationmethodology defined above is presented in Figure 13-7. [b.] that was usedfor this example is shown in Figure 13-1.

13.6. Cladding Peak Oxide Thickness

w 13.6.1. Cladding Peak Oxide Thickness Methodology

The COPERNIC code will be used to generate cladding peak oxide thickness predictions. TheuJ8 peak cladding oxide thickness will not be allowed to exceed a best-estimate predicted value of 100

microns.

o The following method will be used to generate the cladding peak oxide thickness predictions.

Best-estimate values will be used for all predictions. Nominal fuel rod characteristics andthermal-hydraulic conditions will be used, similar to those listed in Tables 13-1 and 13-2, respectively. [b.]

A sub-batch is defined as fuel assemblieswithin a given fuel batch that have the same make-up (fuel rod designs, plutoniumcontent, etc.) and that are inserted and discharged from the core at the same time so thatthe fuel assembly residence times are identical. [b.]

COPERNICFCF Non Proprietary

Chapter 1 3 PAGE 13-1 0

The COPERNIC cladding oxide model was developed from European PWR data. Thefuel rod designs for these reactors are generally similar to those used by FCF. The fuelcycle designs and cycle lengths of European reactors, however, often differ significantlyfrom United States reactors. [b.]

to ensurethat the model provides best-estimate predictions at the 100 micron level.

13.6.2. Cladding Peak Oxide Thickness Example

An example of the COPERNIC cladding peak oxide predictions obtained is presented in Figure 13-8. This example contains the predictions for both low-tin Zircaloy-4 and MS advanced alloycladdings, and illustrates the cladding oxide thickness margin gains obtained with the M5 advancedalloy cladding. [b.] used for this example is shown in Figure 13-1.

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REFERENCES

Ref. 13-1 BAW-10180-A, Rev. 1, NEMO - Nodal Expansion Method Optimized, March l993.

Ref. 13-2 BAW-10156-A, Rev. 1, LYNXT - Core Transient Thermal-Hydraulic Program,August 1993.

Ref 13-3 BAW-I0183-A, Fuel Rod Gas Pressure Criterion (FRGPC), July 1995.

Ref. 13-4 BAW-10227P-A, Evaluation of Advanced Cladding and Structural Material (MS) inPWR Reactor Fuel, November 1999.

Ref 13-5 BAW-10168P-A, Rev. 3, B&W Loss-Of-Coolant-Accident Evaluation Model forRecirculating Steam Generator Plants, December 1996.

Ref. 13-6 BAW-10084P-A, Rev. 3, Program to Determine In-Reactor Performance of BWFCFuel Cladding Creep Collapse, July 1995.

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I Chapter 13 PAGE 13-15-_ _

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FIGURE 13-3 Typical Mark-BW/MOX1 Partial-MOX Fuel Cycle

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FIGURE 13-5 Typical Mark-BW/MOX1 Partial-MOX Fuel Cycle

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FIGURE 13-7 Typical Mark-BW/MOXI

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Chapter 13 PAGE 13-23la-q4

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TABLE 13-1 Typical Mark-BW/MOXI Partial-MOX Fuel Cycle


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Ir i m-;BSr BAW 10231(NPYA

APPENDIX A

Correspondence with dhe NRC, includingRequests for Additional Information and Responses

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I

' FRAMATOME COGEMA FUELSSeptember 16, 1999GR99-191.doc

U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555

Subject: Topical R port BAW-10231 , "COPERNIC Fuel Rod DesignComputer code."

References: 1. T. A. C1feman to NRC Document Control Desk,GR074.doc, January 22, 1998.

2. T. A. Coleman to NRC Document Control Desk,GR506.doc, July 14, 1998.

Gentlemen:

Enclosed are fifteen copies of topical report BAW-10231P,COPERNIC Fuel Rod Design Computer Code dated September, 1999.Also enclosed are 12 copies of the revised BAW-10231 which is thenon-proprietary version of BAW-10231P. This report describes theCOPERNIC fuel rod design computer code and its application tofuel designed and licensed by Framatome Cogema Fuels (FCF).

COPERNIC is applicable to natural, slightly enriched,reprocessed, and re-blended highly enriched uranium dioxide fuelsas well as urania-gadolinia, and mixed oxide fuels. FCFrequests that approval of BAW-10231P be limited to application touranium dioxide and urania-gadolinia fuels. An addendum to BAW-10231P to support application to mixed oxide fuel will besubmitted in August 2000.

COPERNIC will be used in the fuel rod design and analysis of thefuel that uses FCF's advanced cladding material. NRC review ofthe advanced cladding topical report is near completion and nowis an appropriate time to update BAW-10231P to reflect themethodology that will be used for that material. FCF also foundthat a revision of the COPERNIC thermal model is required withthis submittal. Incorporating these changes has impactthroughout BAW-10231P. Therefore the report is being resubmittedin its entirety rather than being issued as change pages.

Reference 1 was the original transmittal of the COPERNIC computercode topical report to the NRC. The same report was resubmittedto the NRC in reference 2 to clarify the proprietaryclassification of some of the material. FCF hereby requests that

f Fromatorne Cogema FumslFRAMATO ME 3315 Old Forest Poed. P.O. Box I0935, Lynchburg, VA 24506-0935T I C H N O L O G I 1 S' Telephone: B04-132-3000 Fax: 604-832-3663

these two submittals of BAW-10231P dated January 1998 and July1998 respectively, be destroyed and replaced with the updatedversion of the report.

The applications portion of BAW-10231P is still in preparation.The applications will be completed and documented to the NRC aschapter 12 of BAW-10231P. Chapter 12 will be submitted inDecember 1999.

In accordance with lOCFR2.790, it is requested that this reportbe considered proprietary. An affidavit supporting this request,is attached. In order to complete the engineering activitiesassociated with implementation of advanced cladding and other newproduct designs, NRC approval of COPERNIC for uranium dioxide andurania-gadolinia fuels is needed by December 31, 2000.

Very truly yours,

T. A. Coleman, Vice PresidentGovernment Relations

cc: J. S. Wermiel, NRCS. L. Wu, NRCR. Caruso, NRCS. N. Bailey, NRCC. E. Beyer, PNLM. A. SchoppmanR. N. Edwards

FRAMATOME COG EMA FUELSDecember 2, 1999GR99-234.doc

U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555

Subject: Topical Report BAW-1023AXComputer Code.n C"

RDiRod Design

Reference: T. A. Coleman to NRC Document Control Desk,GR99-191.doc, September 16, 1999.

Gentlemen:

The reference letter transmitted the updated version of theCOPERNIC topical report, BAW-10231P. As was noted in the letter,the applications portion of BAW-10231P was still in preparationand was not included with the submittal. The letter also statedthat- the applications portion would be submitted in December1999. The Applications Methodology (chapter 12) is enclosed withthis letter. Fifteen copies of the proprietary version andtwelve copies of the non-proprietary version are attached.

Please replace pages xvii and xviii and add pages xix and xx tothe Table of Contents in BAW-10231P. Place the retaining pages(chapter 12) at the very end of the report

In accordance with 10CFR2.790, it is requested that this reportbe considered proprietary. An affidavit supporting this request,is attached. In order to complete the engineering activitiesassociated with implementation of advanced cladding and other newproduct designs, NRC approval of COPERNIC for uranium dioxide andurania-gadolinia fuels is needed by December 31, 2000.

Very truly yours,


cc: J.S.R.S.C.N.R.

S. Wermiel, NRCL. Wu, NRCCaruso, NRCN. Bailey, NRCE. Beyer, PNLA. SchoppmanN. Edwards

' FRAMATOMETLC H N O LOG I L S

Framotore Cogems Fuels3318 Old Forest Road. P.O. Box 10936, Lynchburg, VA 245060035Telephone: 804.132-3000 Fax: 804432-3683

. AFFIDAVIT OF THOMAS A. COLEMAN

A. My name is Thomas A. Coleman. I am Vice President of Government Relations for

Framatome Cogema Fuels (FCF). Therefore, I am authorized to execute this Affidavit.

B. I am familiar with the criteria applied by FCF to determine whether certain information

of FCF is proprietary and I am familiar with the procedures established within FCF to

ensure the proper application of these criteria.

C. In determining whether an FCF document is to be classified as proprietary inforrnation,

an initial determination is made by the Unit Manager, who is responsible for originating

the document, as to whether it falls within the criteria set forth in Paragraph D hereof.

If the information falls within any one of these criteria, it is classified as proprietary by

the originating Unit Manager. This initial determination is reviewed by the cognizant

Section Manager. If the document is designated as proprietary, it is reviewed again by

personnel and other management within FCF as designated by the Vice President of

Government Relations to assure that the regulatory requirements of 10 CFR Section

2.790 are met.

D. The following information is provided to demonstrate that the provisions of 10 CFR

Section 2.790 of the Commission's regulations have been considered:

(i) The information has been held in confidence by FCF. Copies of the

document are clearly identified as proprietary. In addition, whenever FCF

transmits the information to a customer, customer's agent, potential customer

or regulatory agency, the transmittal requests the recipient to hold the

information as proprietary. Also, in order to strictly limit any potential or

actual customer's use of proprietary Information, the substance of the

following provision is included in all agreements entered into by FCF, and an

equivalent version of the proprietary provision is included in all of FCF's

proposals:

AFFIDAVIT OF THOMAS A. COLEMAN (Cont'd.)

"Any proprietary information concerning Company's or its Supplier's

products or manufacturing processes which is so designated by

Com~pany or its Suppliers and disclosed to Purchaser incident to the

performance of such contract shall remain the property of Company

or its Suppliers and is disclosed in confidence, and Purchaser shall not

publish or otherwise disclose it to others without the written approval

of Company, and no rights, implied or otherwise, are granted to

produce or have produced any products or to practice or cause to be

practiced iny manufacturing processes covered thereby.

Notwithstanding the above, Purchaser may provide the NRC or any

other regulatory agency with any such proprietary information as the

NRC or such other agency may require; provided, however, that

Purchaser shall first give Company written notice of such proposed

disclosure 'and Company shall have the right to amend such

proprietary information so as to make it non-proprietary. In the event

that Company cannot amend such proprietary information, Purchaser

shall, prior to disclosing such information, use its best efforts to

obtain a commitment from NRC or such other agency to have such

information withheld from public inspection.

Company shall be given the right to participate in pursuit of such

confidential treatment."

2


(ii) The following criteria are customarily applied by FCF in a rational decision

process to determine whether the informnation should be classified as

proprietary. Information may be classified as proprietary if one or more of

the following criteria are met:

a Information reveals cost or price information, commercial strategies,

production capabilities, or budget levels of FCF, its customers or

suppliers.

b. The information reveals data or material concerning FCF research or

development plans or programs of present or potential competitive

advantage to FCF.

c. The use of the information by a competitor would decrease his

expenditures, in time or resources, in designing, producing or

marketing a similar product.

d. The information consists of test data or other similar data concerning

a process, method or component, the application of which results in a

competitive advantage to FCF.

e. The information reveals special aspects of a process, method,

component or the like, the exclusive use of which results in a

competitive advantage to FCF.

f. The information contains ideas for which patent protection may be

sought.

3


The document(s) listed on Exhibit "A', which is attached hereto and made a

part hereof, has been evaluated in accordance with normal FCF procedures

with respect to classification and has been found to contain information which

falls within one or more of the criteria enumerated above. Exhibit "B",

which is attached hereto and made a part hereof, specifically identifies the

criteria applicable to the document(s) listed in Exhibit "A".

(iii) The document(s) listed in Exhibit 'A', which has been made available to the

United States Nuclear Regulatory Commission was made available in

confidence with a request that the document(s) and the information contained

therein be withheld from public disclosure.

(iv) The information is not available in the open literature and to the best of our

knowledge is not known by Combustion Engineering, Siemens, General

Electric, Westinghouse or other current or potential domestic or foreign

competitors of Framatome Cogenma Fuels.

(v) Specific information with regard to Whether public disclosure of the

information is likely to cause harm to the competitive position of FCF, taking

into account the value of the information to FCF; the amount of effort or

money expended by FCF developing the information; and the ease or

difficulty with which the information could be properly duplicated by others

is given in Exhibit "B".

I have personally reviewed the document(s) listed on Exhibit "A" and have found that it

is considered proprietary by FCF because it contains information which falls within one

or more of the criteria enumerated in Paragraph D, and it is information which is

customarily held in confidence and protected as proprietary information by FCF. This

report comprises information utilized by FCF in its business which afford FCF an

4

AFFIDAVIT OF THOMAS A. COLEMAN (Cont'd.l)

opportunity to obtain a competitive advantage over those who may wish to know or use

the information contained in the document(s).

THOMAS A. COLEMAN

State of Virginia)

City of Lynchburg)) SS. Lynchburg

Thomas A. Coleman, being duly sworn, on his oath deposes and says that he is theperson who subscribed his name to the foregoing statement, and that the matters and facts setforth in the statement are true.

7TAO* THOMAS A. COLEMAN

Subscribed and sworn before methis 44 day of _ g& 1999.

A etd,Notary Public in and for the Cityof Lynchburg, State of Virginia.

My Comnmission Expires

5

EXITlS A & B

EXHIBIT A

Applications Methodology - Chapter 12 ofTopical Report BAW-10231P, "COPERNIC Fuel Rod Design Computer Code"

EXHIBiT B

The above listed document contains information which is considered Proprietary inAccordance with Criteria b,c,d, and e of the attached affidavit.

6

-- -.-- . - . . . ..Jr.

UNITED STATES_ * NUCLEAR REGULATORY COMMISSION

WASHINGTOK D.C. 2055501August 11, 2000

49Mr. T. A. Coleman, Vice PresidentGovernment RelationsFramatome Cogema Fuels3315 Old Forest RoadP. 0. Box 10935 B HeLynchburg, Virginia 24506-0935

SUBJECT: REQUEST FOR ADDITIONAL INFORMATION - FRAMATOME TOPICALREPORT BAW-10231P (TAC NO. MA6792)

Dear Mr. Coleman:

By letter dated September 16, 1999, Framatome Cogema Fuels requested a review of TopicalReport BAW-10231P, "COPERNIC Fuel Rod besign Code." The staff has determined thatadditional information Is needed in order to complete its review.

The enclosed questions have been discussed with Mr. F. McPhatter of your staff. Pleaseprovide a response to these questions within 30 days of receipt of this letter. If you have anyquestions concerning our review, please contact me at (301) 415-1321.

Sincerely,

Stewart Bailey, Proj Ma Section 2Project Directorate loDivision of Licensing Project ManagementOffice of Nuclear Reactor Regulation

Project No. 693

Enclosure: Request for Additional Information

cc wlenci: See next page

Mr. T. A. Coleman Project No. 693

cc:Mr. F. McPhatter, ManagerFramatome Cogema Fuels3315 Old Forest RoadP.O. Box 10935Lynchburg, VA 24506-0935

Mr. Michael SchoppmanLicensing ManagerFramatome Technologies, Inc.1700 Rockville Pike, Suite 525Rockville, MD 20852-1631

_ -.. .-I .

REQUEST FOR ADDITIONAL INFORMATION

TOPICAL REPORT BAW-1 0231

"COPERNIC FUEL ROD DESIGN CODE"

1. Section 2.4 mentions iterations/convergence on gap conductance or contact pressureand also on axial interaction forces but does not mention an Iteration on fissions gasreleased (FGR and # of moles), however, Figure 2-4 indicates that the code may Iterateon number of moles released. Please discuss which is correct. If the code does notiterate on number of moles please discuss why this is satisfactory for code applicationsincluding transients.

2. Please compare the COPERNIC fuel thermal conductivity predictions to out-of-reactorUO2 thermal diffusivity data from References 1 and 2 and any other high bumupdiffusivity data that are applicable. 'The UO2 diffusivity data can be converted to thermalconductivity for these comparisons using the COPERNIC equations for specific heat. Inorder to fully understand the rim model thermal conductivity as applied to Haldentemperature predictions, please provide a one axial node calculation of temperatureprofile for IFA 562 at bumups of 60, 80 and 90 GWd/iMTU with and without the rimmodel. Please provide the radial bumup profiles used for this calculation.

3. The temperature uncertainties for LOCA and fuel melt analyses should ideally be based_ on data at linear heat generator rates (LHGRs) : 30 kWVm because these analyses are

performed at high LHGRs. The problem with determining temperature uncertainties forbumups greater than 30 GWd/MTU is that there is very little measured centerlinetemperature data with LHGRs > 30 kWlm. Please provide the COPERNIC comparisonsto data by plotting predicted minus measured temperatures versus bumup for LHGRs >30 kWMm to determine whether there is a change in thermal uncertainty with increasingbumup, and provide the uncertainties from this data comparison. Also, provide theCOPERNIC predicted minus measured centerline temperature data versus bumup forLHGRs > 15 kWlm, and the uncertainties from this data comparison. Thesecomparisons will help to verify that the uncertainties for the data that includes the lowerLHGRs are applicable to the higher LHGRs where LOCA and fuel melting analyses areperformed.

4. Please provide LHGRs and design Information for the EXTRAFORT test rod.

6. The comparison to IFA 432-1 inlet thermocouple only extends to a bumup of 9GWdIMTU, but the NUREGICR-4717 report provides data up to a bumup of 27GWd/MTU at the inlet thermocouple. Please provide the COPERNIC comparison up tothe limit of the data or a justification why this comparison Is not valid.

6. Is Framatome a member of Halden? If so, Halden has refabricated two high (- 59GWdIMTU) burnup rods (one with a functional thermocouple) and placed them first InIFA-597.2 (iHIWR-442) and subsequently in IFA- 597.3 (HWR.543) with measuredcenterine temperatures. Please compare COPERNIC code predictions to this data and

_ include this data in the response to Question 3 above.

,, ........................................ -. % ............ ,- ..t

- 2 -

7. The athermal fission gas release model (Section 5.2.2) is dependent on open porositybut no values are provided for what is used for Framatome fuel, What values are usedfor open porosity? If more than one value Is used, please provide the value for eachfabrication process.

8. The Section 5.2.3.5 explanation is not very clear about how the fission gas releasemodel applies to varying conditions of power and temperature. It would help to haveseveral examples for conditions of both increasing temperature and decreasingtemperature. Also, examples of fast and slow rate of change in fuel temperature arewarranted. It appears that the resolution thickness Is not used in the final equations InCOPERNIC for computing fission gas release. Is this Interpretation correct?

9. A comparison of the COPERNIC upperbound fission gas release predictions tomeasured data (with > 7 percent measured release) from U02 - Gd2O3 fuel rods withsteady-state power operation (from Table 5-3) demonstrates that the code underpredicts2 out of 6 rods (it is noted that one of the rods is only slightly underpredicted). Acomparison of COPERNIC upperbound predictions to the transient measured data with>5 percent release from U02 -Gd203 rods (from Table 5-4) demonstrates the codeunderpredicts 5 out of 25 rods. This indicates that the code's upperbound fission gasrelease model for U02 - Gd2O3 bounds much less than 95 percent of the data that arewithin the range of application for the rod pressure analysis. Also, the code does notappear to have been compared to the B&W segmented rodlets steady-state irradiated inANO-1 and power ramped in the Studesvik R2 reactor. If not, why was this comparisonnot made and presented because these rods are representative of U.S. designs?

10. The standard deviation of the gaseous swelling model is on the order of the inferred gasporosity from the measured porosity distributions. In fact there are only 3 data pointsout of 14 that have inferred gaseous swelring greater than 0.6, i.e., significantly greaterthan the standard deviation. Of these 3 data points only one of these is predicted wellby the gaseous swelling model while the other two data points are significantlyunderpredicted by factors of 1.6 and 2.9. Therefore, the validity and the accuracy of thegaseous swelling model appears questionable. What is the impact of the gaseousswelling model on rod pressure, melting and strain predictions? Does the gaseousswelling model use local burnup or pellet average bumup. The COPERNIC steady-stategaseous swelling model (Equations 6-13 to 6-15) has been programmed intoFRAPCON-3 with calculational results of 0.007 inches of displacement at a pelletaverage bumup of 62 GWd1MTU with a centerline temperature of 1200"C. Is thispredicted displacement with this model reasonable for these conditions? If not, furtherdiscussions are necessary to understand the gaseous swelling model.

11. Figure 6-8 (from September 1999 version) predicted versus measured densification datais significantly different from Figure 6-5 of the July 1998 version of COPERNIC;however, the densification and swelling models appear to be the same. Please explainwhy the data In the two figures are not the same.

12. The fuel column growth data in Figure 6-9 (September 1999 version) appears to containsignificantly less growth data than the same figure (Figure 6-6) In the July 1998 versionof COPERNIC. Please discuss why there is less data in the current version of

MMMMM O ~ ~ -__ ---- . - ... - -. . ..r .. ... ' - .4. . : '

-3 -

COPERNIC. Does the column growth data in Figure 6-11 include both ADU and AUCprocessed fuel or is It just AUC fuel?

13. Equation 7-1 for creep is a function of the shear stress component (a.- a,). Pleaseprovide a derivation of how this shear stress is determined to be the only activedeterminant of creep from Hills or Von Mises equations because these are not the onlyshear stress or stress components in these equations.

14. Will the FRAGEMA AFA 2G cladding that is fabricated in Europe be used in U. S.plants? Sections 7.1.2.2.1 and 7.1.2.3.1 refer to a number of cladding tube (AFA 2G)irradiation tests in the SILOE test reactor. Please provide further information on themanufacturing differences between the cladding from these tests and thosemanufactured commercially for U. S. plants, e.g., FCF Base Zr-4 and AFA 2G. Also,were the hoop stresses quoted In Table 7-1 posltve or negative? (The creep modelneeds to be validated against the current U.S. fabricated FCF Zr-4 cladding. Seequestion 15.)

15. Section 7.1.2.3.2 and Figures 7-20, 21 and 22 all refer to creep data from fuel rodsirradiated in the CAP test reactor. Pacific Northwest National Laboratory (PNNL) norNRC is familiar with this test reactor. Please provide the test reactor or loop conditionsthat are pertinent to in-reactor creep such as coolant inlet-outlet temperatures, fast flux,system pressure, etc. Also, provide predicted versus measured creep for the FCF Zr-4cladding used in the U.S. and the background information on this data.

16. Section 7.1.2.3.3 notes that creep data from one rod was excluded from the uncertaintydetermination because It was next to gadolinia rods. Does this mean that the creepmodel uncertainty does not apply to fuel rods near gadolinia rods? Also, Figure 7-20shows a considerable amount of measured-to-predicted data that are outside of theuncertainty bounds proposed. Please discuss why it Is ok to discard this data from theuncertainty determination for creep and those data in Figure 7-20 that are not within theproposed bounding creep uncertainty. Please identify those analyses whereoverprediction of creep is conservative and those analyses where underprediction isconservative.

17. Section 7.1.3.2.1 discusses the development of the M5 creep model from tubeIrradiations but no comparison to this data is provided, and the stress and temperatureparameters of this data are also not provided. Please provide this data andcomparisons to the M5 creep model. It Is also stated that the secondary thermal creeprate is independent of alloy type, but no data Is presented to corroborate this statement.Please provide this data.

18. Does COPERNIC consider the effects of cladding growth (Section 7.3.2) in thediametral direction or is this Implicit In the creep data? Also, the upperbound modelunderpredicts a significant amount of growth data in Figure 7-50. Please explain whythis is acceptable. The alloy 5 growth model, Equation 7-27, is not linearly dependentbut has a decreasing slope with fluence while the majority of Zircaloy growth data showa linear dependence with fluence. In addition, an initial examination of Figure 7-54appears to show that a linear dependent model would provide as good or better

-4 -

prediction of the alloy 5 growth data compared to the model proposed. Please providefurther Information on why Equation 7-27 Is more appropriate for predicting alloy 5growth even though a linear model would be more conservative and provide as good afit to the growth data.

19. Section 8.1 discusses the COPERNIC corrosion model and comparisons to data. Thecoolant inlet temperatures are provided for some of the reactors from which corrosiondata was taken but coolant outlet temperatures and LHGR are also importantparameters. What were the outlet temperatures and average LHGRlcycle for both theZr-4 and alloy 5 data (including the adjustment rod data) and Identify high duty, mediumduty and low duty plants (see 20 below)? Section 8.1.3.2 states that the alloy 5 data Isbased on the maximum average azimuthal oxide thickness over the span height.Please discuss how this is determined from actual measurements, e.g., is It an averageover a given length and how many azimuthal orientations are measured?

20. Also, oxide predictions and comparisons to data from ZrA4 are provided for all axial rodlocations; however, NRC Is most concerned with rod locations that experience maximumoxide thicknesses and corrosion from high duty plants. The axial locations wilhmaximum oxide thicknesses are typically in the next to last span or the next to last twospans from the top of the assembly depending on the number of spacer grids perassembly. Please provide predicted minus measured oxide thickness versus bothbumup and measured oxide thickness only for those axial spans with maximummeasured oxide thickness for each rod, and Identify high duty, medium duty and lowduty plants along with a discussion of the differences between the operating parametersof these different plants.

21. From examination of Figure 8-11, the COPERNIC code appears to significantlyunderpredict a large amount of measured oxide data from U. S. plants with Zr-4cladding. Please provide predicted minus maximum measured oxide thickness versusbumup and maximum measured oxide thickness only from rods from U. S. plants usingboth the COPERNIC and COROS02 corrosion models. Please provide predictions ofthis same U. S. data using the COPERNIC upper bound corrosion model. PNNL'scomparison of COROS02 and COPERNIC corrosion models at various temperaturesfor both Zr-4 and alloy 5 has demonstrated that COROS02 predicts the greater oxidethicknesses. Please discuss why this Is acceptable.

22. What Is the basis for the oxide layer thermal conductivity functions provided at thebottom of page 8-3? It appears that the oxide conductivity Is determined based on theoxide surface temperature. Is this Interpretation correct?

23. Please provide the average LHGR/cycle for the hydrogen pickup data provided inFigure E-22. The applicability of using only 5 cycle data to estimate the hydrogen pickupfraction Is questionable because there may be other factors (such as heat flux) in the 3and 4 cycle data that results in the 5 cycle data giving the lowest hydrogen pickupfractions.

24. Please provide the background data for the fuel melting temperature relationship usedby COPERNIC (Equations 10-11 and 12-2).

. .. e e .n .. , . I- 1. I"'-..

5 -

25. Section 12.0 notes that COPERNIC is used for initialization of core thermal-hydrauliccodes. Please list those calculated COPERNIC parameters used for initialization andthe specific applications of the thermal-hydraulic codes.

26. In Section 12.1.1 (page 12-2) under discussion on code uncertainties, It is noted that thecode has an option that conservatively bounds the fissions gas release data and thatthis option is used to bound the fission gas release for the rod pressure predictions.However, there is a concern that this option will not bound the U02 - Gd2O3 data withinthe fission gas release range that Is Important to the rod pressure analysis for UO2 -Gd203 rods (see Question 9 above) at the stated level of conservatism. Please discussthis issue further, particularly in relation to Question 9 above.

27. In Section 12.1.1 (page 12-3) under the discussion on transients, it is noted that plantspecific operating data may be used to establish simulated transients. Please explainfurther by what is meant by sufficient plant operating data and provide an example.

28. There is a concern that the uncertainty factor provided in Equation 12-1 may be toosmall it the predicted operating temperatures (stored energy) calculated for LOCAinitialization. Please discuss this issue further, particularly in relation to Question 3above.

29. Are any of the example calculations provided in Section 12 for fuel cores with two 24month cycles? It appears that there are no 24-month cycle results presented for theMark BW-17 design. If so, please explain because it Is anticipated that a large numberof plants will be switching to 24-month cycles In the next few years.

30. The axial power distributions for the transients were found for the example licensinganalyses, but the power distribution for the steady-state power operation were not foundin the topical report. Please provide these axial power distributions. If there are morethan 20 axial power profiles it would be helpful to condense the number down to 20 orless. Also, the steady-state power histories are only provided as plots versus bumup.Please provide these In tabular form to support the NRC audit calculation of thesecalculational examples?

31. Section 12.4.1 states that the cladding strain analyses will be run with ...... and thegaseous swelling option turned off. Performing these analyses without gaseousswelling ... produces more accurate predictions .at the very high local powerlevels that accompany these analyses. This appears to be contradictory to thecomparisons to data in Figures 6-17, 6-18, and 6-19 that demonstrate that COPERNICwith gaseous swelling option turned on provides an adequate prediction of diametralstrains. Please provide data and Information that supports the conclusion that theexclusion of gaseous swelling in COPERNIC produces more accurate strain predictions.

32. Section 12.5 states that COPERNIC will be used to generate cladding creep collapseinitial conditions and example rod pressure results are provided in Figures 12-34 and12-35. Are there any other initial conditions provided by COPERNIC for the creepcollapse analysis, e.g., cladding temperatures? If so, please provide predictions ofthese initial conditions.

. .... .1 - ,

-6-

References

1. Kinoshita, M. et al. 2000, "High Burnup Rim Project (II): Irradiation And Examination toInvestigate Rim-Structured Fuel," Proceedings of the ANS International Topical Meetingon LWR Fuel Performance, Park City, UT, April 10-13, 2000.

2. J. Nakamura, et al. 1997, 'Thennal Diffusivity Measurements of a High-Burnup UO2Pellet," In Proceedings of the ANS International Topical Meeting on LWR FuelPeformnance, Portland OR, March 2-6 1997.

IFRAMATOME COGEMA FUELS

September 29, 2000GROO-142.doc

U. S. Nuclear Reguiatory CommissionATTN: Document Control DeskWashington, D. C. 20555

Reference: Stewart Bailey, NRC, to. T. A. Coleman, Framatome Cogema Fuels,Request for Additional Information - Framatome Topical Report BAW-10231P (TAC NO. MA6792), August 11, 2000.

Gentlemen:

The reference letter transmitted thirty-two questions on FCF Topical Report BAW -10231 P, "COPERNIC Fuel Rod Design Code.' These questions are verycomprehensive in nature and will require a significant effort by FCF. Many of thequestions will require input from Framatorne in France. Some of the questions requitecomparisons of code predictions to data and some are direct requests for data. It willnot be possible to provide adequate responses-to the RAI in the thirty day time framerequested by the NRC. The current schedule is for submittal of responses to allquestions by January 31, 2001. If there are any questions or problems with thisschedule please contact C. F. McPhatter at (804) 832-2401.

Very truly yours,

(A. Coleman, Vice PresidentGovernment Relations

cc: J. S. Wermeil, NRCS. N. Bailey, NRCR. Caruso, NRCS. L. Wu, NRCC. E.:Beyer, PNNLM. S. SchoppmanR. N. Edwards20A13 File/Records Management

AraATatom MoEoem RlsF.RAMATOME .O. ox 11648. Lynchburg VPA24506-1646

T C N N 0 L 0 a I a S Telephone: 604-832-SOO Fax: 604-832-5167

; --

gFRAMATOME .,.W?

February 5, 2001GROI-021.doc

U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, D. C. 20555

Reference: 1. Stewart Bailey, NRC, to T. A. Coleman, Framatome CogemaFuels, Request for Additional Information - Framatome TopicalReport BAW-1 0231 P (TAC NO. MA6792), August 11, 2000.

2. T. A. Coleman, Framatome Cogema Fuels, to U. S. NuclearRegulatory Commission, GROO-142.doc, September 29, 2000.

Gentlemen:

Reference 1 provided a request for additional information (RAI) on Framatome CogemaFuels (FCF) topical report BAW-1 0231 Pt 'COPERNIC Fuel Rod Design Code.' Inreference 2, FCF agreed to provide responses to the RAI by January 31, 2001. Theresponses are enclosed. The responses are being submitted with the prefix 14 on eachpage. When the final NRC-approved version of BAW-1 0231 P is issued, the responseswill comprise chapter 14.

In accordance with the provisions of 10 CFR 2.790, Framatome ANP requests thatthese responses be considered proprietary and withheld from public disclosure.Attachment 1 is an affidavit supporting this request. Attachment 2 is the proprietaryversion of the responses and Attachment 3 is the non-proprietary version.

The approval of the COPERNIC code at this time Is requested for the advanced alloyM5Tm cladding only. Framatome ANP will continue to use the NRC approved codeTACO for applications with Zircaloy cladding. Responses for the Zircaloy portions ofthe RAI (Questions 14,15, 16,19, 20, 21, and 23) will be provided at such time thatapproval for COPERNIC application to Zircaloy cladding is requested by our customers.

Very truly yours,


3315 Od Foret Road P.O. Sa 10935 Lyncwhbug. VA 24506-0935 TM? 8043832-3000 www.fmatech.eom

I

cc: J. S. Wermell, NRCS. N. Bailey, NRCR. Caruso, NRCS. L. Wu, NRCC. E. Beyer, PNNLM. S. Schoppman20A13 File/Records Management

AFFIDAVIT OF THOMAS A. COLEMAN

A. My name is Thomas A. Coleman. I am Vice President of Government Relations for

Framatome ANP. Therefore, I am authorized to execute this Affidavit.

B. I am familiar with the criteria applied by Framatome ANP to determine whether certain

information of Framatome ANP is proprietary and I am familiar with the procedures

established within Framatome ANP to ensure the proper application of these criteria.

C. In determining whether an Framatome ANP document is to be classified as proprietary

information, an initial determination is made by the cognizant manager, who is responsible

for originating the document, as to whether it falls within the criteria set forth in Paragraph D

hereof. If the information falls within any one of these criteria, it is classified as proprietary

by the originating cognizant manager. This initial determination is reviewed by the cognizant

Section Manager. If the document is designated as proprietary, it is reviewed again by

personnel and other management within Framatome ANP as designated by the Vice President

of Government Relations to assure that the-regulatory requirements of 10 CFR Section 2.790

are met.

D. The following information is provided to demonstrate that the provisions of 10 CFR Section

2.790 of the Commission's regulations have been considered:

(i) The information has been held in confidence by Framatome ANP. Copies of the

document are clearly identified as proprietary. In addition, whenever Framatome-

ANP transmits the information to a customer, customer's agent, potential customer

or regulatory agency, the transmittal requests the recipient to hold the information

as proprietary. Also, in order to strictly limit any potential or actual customer's

use of proprietary information, the substance of the following provision is included

in all agreements entered into by Framatome ANP, and an equivalent version of the

proprietary provision is included in all of Framatome ANP's proposals:


'Any proprietary information concerning Company's or its Supplier's

products or manu1acturing processes which is so designated by Company or

its Suppliers and disclosed to Purchaser incident to the performance of such

contract shall remain the property of Company or its Suppliers and is

disclosed in confidence, and Purchaser shall not publish or otherwise

disclose it to others without the written approval of Company, and no

rights, implied or otherwise, are granted to produce or have produced any

products or to practice or cause to be practiced any manufacturing processes

covered thereby.

Notwithstanding the above, Purchaser may provide the NRC or any other

regulatory agency with any such proprietary information as the NRC or

Auch other agency may require; provided, however, that Purchaser shall

first give Company written notice of such proposed disclosure and

Company shall have the right to amend such proprietary information so as

to make it non-proprietary. In the event that Company cannot amend such

proprietary information, Purchaser shall, prior to disclosing such

information, use its best efforts to obtain a commitnent from NRC or such

other agency to have such information withheld from public inspection.

Company shall be given the right to participate in pursuit of such

confidential treatment. "

2

AFFAVIT OF THOMAS A. COLEMAN (Cont'd.)

(ii) The following criteria are customarily applied by Framatome ANP in a rational

decision process to determine whether the information should be classified as

proprietary. Information may be classified as proprietary if one or more of the

following criteria are met:

a. Information reveals cost or price information, commercial strategies,

production capabilities, or budget levels of Franatome ANP, its customers

or suppliers.

b. The information reveals data or material concerning Framatome ANP

research or development plans or programs of present or potential

competitive advantage to Framatome ANP. -

c. The use of the information by a competitor would decrease his

expenditures, in time or resources, in designing, producing or marketing a

similar product.

d. The information consists of test data or other similar data concerning a

process, method or component, the application of which results in a

competitive advantage to Framatome ANP.

e. The information reveals special aspects of a process, method, component or

the like, the exclusive use of which results in a competitive advantage to

Framatome ANP.

f. The information contains ideas for which patent protection may be sought.

3


The document(s) listed on Exhibit "A", which is attached hereto and made a part

hereof, has been evaluated in accordance with normal Framatome ANP procedures

with respect to classification and has been found to contain information which falls

within one or more of the criteria enumerated above. Exhibit 'B", which is

attached hereto and made a part hereof, specifically identifies the criteria applicable

to the document(s) listed in Exhibit "A".

(iii) The document(s) listed in Exhibit 'A", which has been made available to the

United States Nuclear Regulatory Commission was made available in confidence

with a request that the documnent(s) and the information contained therein be

withheld from public disclosure.

(iv) The information is not available in the open literature and to the best of our

knowledge is not known by Combustion Engineering, Siemens, General Electric,

Westinghouse or other current or potential domestic or foreign competitors of

Framatome ANP.

(v) Specific information with regard to whether public disclosure of the information is

likely to cause harm to the competitive position of Framatome ANP, taking into

account the value of the information to Framatome ANP; the amount of effort or

money expended by Framatome ANP developing the information; and the ease or

difficulty with which the information could be properly duplicated by others is

given in Exhibit mB',

E. I have personally reviewed the document(s) listed on Exhibit "A' and have found that it is

considered proprietary by Framatome ANP because it contains information which falls within

one or more of the criteria enumerated in Paragraph D, and it is information which is

customarily held in confidence and protected as proprietary information by Framatome ANP.

This report comprises information utilized by Framatome ANP in its business which afford

4


Framatome ANP an opportunity to obtain a competitive advantage over those who may wish

to know or use the information contained in the document(s).

THOMAS A. COLEMAN

State of Virginia)) SS. Lynchburg

City of Lynchburg)

Thomas A. Coleman, being duly sworn, on his oath deposes and says that he is the personwho subscribed his name to the foregoing statement, and that the matters and facts set forth in thestatement are true.

THOMAS A. COLEMAN

Subsc ibed and sworn before methiCfday od " 2001.

Notary Public in and for the Cityof Lynchburg, State of Virginia.

My Commission Expires

5

EXM1BITS A & B

EXHIBff A

Documented Responses to NRC Request for AdditionalInformation On BAW-10231P Dated August 11, 2000

EXEBIT B

The above listed document contains Information which is considered Proprietary Inaccordance with Criteria b, c, d, and e of the attached affidavit.

6

. s . .. I

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231

Table of Contents

Question I .. 14-11Question 2 ... 14-4Question 3 ... 14-9Question 4 ... 14-12QuestionS5... 14^20QuestionS6... 14-22Question 7 ... 14-24Question S .1. 4-25

Q uestion I .................................................................................................................................................... .... 14-32

Question 10 ... 14-34Question 11 ... 14-39Question 12 ... 14-40Question 13 ... 14-42Question 14 ... 14X44Question 15 ... 14-45Question 16 ... 14-46Question 17 ... 14-47Question 18 ... 14-53Question 19 ... 14-54Question 20 ... 14-56Question 21 ... 14-57Question 22 ... 14-58Question 23 ... 14-80Question 24 ... 1441lQuestion 25 ... 14-83Question 26 ... 14-44Question 27 ... 14-65Question 26 ... 14-67Question 29 ... 14688Question 30 ....... , .14-89Question 31 ... 14-222Question 32 ... 14-223

- l

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERND FUE RO DEINCMUE CD A-0

List of Figures

Figure 14-1: LHGR vs. Macro- and Micro-Time Steps [d; ..................................................................... 14-2Figure 14-2: Bounding Fission Gas Release Predictions vs. Macro- and Micro-Time-Steps Id] ................... 14-3Figure 14-3: COPERNIC and NFIR-11l Thermal Conductivity Comparison - 100% Theoretically Dense Fuel -

60 GWdItU Bumup .................................................................. 14-5Figure 14.4: COPERNIC and JAERI Thermal Conductivity Comparison - Sample No.2 - 100% Theoretically

Dense Fuel - 63 GWd/tU Burnup, 63-89% Initial Density Range ............................................................ 14-5Figure 14-5: COPERNIC and JAERI Thermal Conductivity Comparison -Sample No.3 - 10t 0% Theoretically

Dense Fuel - 63 GWdItU Bumup, 92-96% Initial Density Range ............................................................ 14-6Figure 14-8: Rim Effect at 60 GWd/tU - IFA 562 ..................................................................... 146Figure 14-7: Rim Effect at 60 GWdIIU - IFA 562 ..................................................................... 14-7Figure 1448: Rim Effect at 90 GWd/tU - IFA 562 ..................................................................... 14-7Figure 14-9: Radial Bumup Profiles - IFA 562 ..................................................................... 14-8Figure 14-10: Predicted Minus Measured Centerline Temperature Differences vs. Bumup ........................ 14-10Figure 14.11: Predicted Minus Measured Centerline Temperature Differences vs. Bumup ........................ 14-11Figure 14-12: Measured and Predicted Fuel Temperatures vs. Bumup ...................................................... 14-21Figure 14-13: Fuel Centerline Temperature Measurements and Predictions vs. Bumup, IFA-597.2 .......... 14-23Figure 14-14: Fuel Centerline Temperature Measurements and Predictions vs. Bumup, IFA-597.3 .......... 14-23Figure 14-15: FGR Transition Algorithm: Small Power Change ................................................................... 14-26Figure 14-16: FGR Transition Algorithm: Significant Power Increase .......................................................... 14-27Figure 14-17: FGR Transition Algorithm: Significant Power Decrease ........................................................ 14-28Figure 14-18: Rapid Change in Fuel Temperature [dl ...................................................................... 14-29Figure 14-19: Slower Change in Fuel Temperature [d] ..................................................................... 14-30Figure 14-20: Local Gas Concentration with Decreasing Temperatures (ld)..............................................14-31Figure 14-21: Upper-bound Predicted vs. Measured Steady-state Fission Gas Release for UOrGd2O3 Fuel

................................................................................................................................... 14-33Figure 14-22: Upper-bound Predicted vs. Measured Transient Fission Gas Release for U02, MOX and U02-

Gd2O3 Fuels...........................................................................................................................................14-33Figure 14-23: Bounding Internal Gas Pressure Wth and Without COPERNIC Gaseous Swelling Model Effects

- Typical Mark-BW17 Urania-Gadolinia Cycle, U02 Rods....................................................................14-35Figure 14-24: Fuel Melt With and Without COPERNIC Gaseous Swelling Model Effects -Typical Mark-BW17

Urania-Gadoinia Cycle, U02 Rods.......................................................................................................14-35Figure 14-25: 1 % Cladding Strain With and Without COPERNIC Gaseous Swelling Model Effects - Typical

Mark-8W17 Urania-Gadolinia Cycle, U% Rods .. ..................................................................... 14-Figure 14-26: Measured and Predicted Cladding Diameter Variations ........................................................ 14-37Figure 14-27: Measured and Predicted Cladding Diameter Variations ................................................ 14-38Figure 14-28: Measured and Predicted Fuel Column Growth (UO2) [d] ................................................ 1 4-41Figure 14-29: Advanced Alloy MS Creep Tests of Unirradiated Tubes at [bi...............................................14-49Figure 14-30: Advanced Alloy M5 Creep Tests of Unirradiated Tubes at lb) ............................................... 14-50Figure 14-31: Advanced Alloy M5 Secondary Thermal Creep Rate vs. Fluence [b] .................................... 14-51Figure 14-32: Combined Data Set of Creep Strain vs. Fluence lb] ......................................................... 14-52Figure 14-33: Oxide Thermal Conductivity vs. Temperature and Oxide Layer Thickness ........................... 14-59Figure 14-34: Typical Mark-B Urania-Gadolinia Cycle Predictions lb] ......................................................... 14-66Figure 14-35: Typical Mark-B Fuel Cycles - Creep Collapse Analyses Id)................................................14-224Figure 14-36: Typical Mark-BW17 Fuel Cycles - Creep Collapse Analyses [dl ........................................14-224Figure 14-37: Typical Mark-B Fuel Cycles - Creep Collapse Analyses [di ................................................ 1 4-225Figure 14-38: Typical Mark-6W17 Fuel Cycles - Creep Collapse Analyses [d] ........................................14-225Figure 14-39: Typical Mark-B Fuel Cycles - Creep Collapse Analyses cdi ................................................ 14-226Figure 14-40: Typical Mark-BW17 Fuel Cycles - Creep Collapse Analyses d ........................................ 14-226Figure 14-41: Typical Mark-B Fuel Cycles - Creep Collapse Analyses [d] ................................................ 14-227Figure 14-42: Typical Mark-SW17 Fuel Cycles - Creep Collapse Analyses [d] ....................................... 14-227

1i

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW 10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

List of Tables

Table 14-1: Design Information for the EXTRAFORT Mother Rod ............................................................ 14-12Table 14-2: Thermal-Hydraulic Conditions for the EXTRAFORT Mother Rod ............................................ 14-13Table 14-3: EXTRAFORT Mother Rod Conditions History..........................................................................14-14Table 14-4: EXTRAFORT Mother Rod Power Shapes ............................................................ 14-16Table 14-5: Design Information for the EXTRAFORT Re-fabricated Test Rodlets ..................................... 14-18Table 14-8: Thermal-Hydraulic Conditions for the EXTRAFORT Re-fabricated Test Rodlets .................... 14-18Table 14-7: Conditions History for the EXTRAFORT Re-fabricated Rod .................................................... 14-19Table 14-8: Calibration Database for Creep Hardening Effects ............................................................ 14-48Table 14-9: Plant and Fuel Rod Data ............................................................ 14-55Table 14-10: Unirradlated U02 Melt Temperature ............................................................ 14-61Table 14-11: Unirradiated (U,Gd)0 2 Melt Temperature ............................................................ 14-62Table 14-12: Axial Power Distribution Data Sets ............................................................ 14-70

Note: Tables 14-13 through 14-59 are listed in Table 14-12.

Iii

COPERNIC: FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL Roc DESIGN COMPUTER CoDs BAW-1 0231

Question i

Section 2A mentions Iterationslconvergence on gap conductance or contact pressure and also on axialinteraction forces but does not mention an Iteration on fissions gas released (FGR and # of moles), however,Figure 2-4 indicates that the code may Iterate on number of moles released. Please discuss which Is correct.If the code does not iterate on number of moles please discuss why this Is satisfactory for code applicationsincluding transients.

Response

Figure 2-4 Id]. There is no concern, however, that [d]. Micro-ftime-steps are generated at subdivisionsbetween the user-selected macro-time-steps. A micro-time-step Is [d]. lb, c]. This is Illustrated in Figures 14-1 and 14-2 where the macro- and micro-time-steps are defined with the larger diamond and smaller cylindricalshaped symbols, respectively. These illustrations were developed from the lel example of (e]. A total of [d)and [d] micro-time-steps were generated for the [d, eJ macro-time-step transient and [dl macro-time-stepentire Internal gas pressure case, respectively. These figures Illustrate the [d] additional time steps that aregenerated due to the micro-time-step feature. The [d] micro-ftime-steps [dl.

PAGE 14-1

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

FIgure 14-1: LHGR vs. Macro- and Micro-Time Steps

Figure 1441: LHGR vs. Macro- and Micro-Time Steps[el

Id]

PAGE 14-2


Figure 14-2: Bounding Fission Gas Release Predictions vs. Macro- and Micro-Time-Steps [dJ

[d]

PAGE 14-3

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNI C FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 2

Please compare the COPERNIC fuel thermal conductivity predictions to out-of-reactor UO2 thermal diffusivItydata from References I and 2 and any other high bumup diffusivity data that are applicable. The U02diffusivity data can be converted to thermal conductivity for these comparisons using the COPERNICequations for specific heat. In order to fully understand the rim model thermal conductivity as applied toHalden temperature predictions, please provide a one axial node calculation of temperature profile for IFA562 at bumups of 60, 80, and 90 GWdWMTU with and without the rim model. Please provide the radial bumupprofiles used for this calculation.

Response

COPERNIC fuel thermal conductivity predictions are compared In Figures 14-3 through 14-5 with fuel thermalconductivities that were obtained from three sets ( 4) of fuel thermal diffusivilty measurements. These figuresshow that the COPERNIC thermal conductivitles lb, d] with the Nuclear Fuels Research Program (NFIR) data'4 ) and somewhat (bd] the Japan Atomic Energy Research Institute (JAERI) data A. There are no end-of-lifedensity or porosity measurements presented with the Kinoshita "1), et el, data that are needed to calculate therim thermal conductivitles. This fact is illustrated In the following statements"':

it must be noted that the data In Figure 6"') are just measured TD values and the effect of porosity ofIndividual specimens were not considered. As the effect of coarsened pores, typical for the rim structure,on thermal resistance Is not clear, the comparison°" was made without corrections. Therefore, thispresentation must be considered still preliminary and the evaluation of thermal conductivity should bediscussed only after detailed analyses."

I4.

The COPERNIC predicted.radial fuel temperature predictions with and without the COPERNIC rim model areshown in Figures 14-6 through 14-8, at bumups of 60, 80, and 90 GWd/tU, respectively. lb, d]. Althoughthese differences are relatively small, KInoshital", et al, implies that these differences are very small or shouldnot exist at all. Note, however that the 1K(inoshfta°1 , et al, data Is preliminary and was obtained without stressInducing fuel restraints. UneW5, et el, suggests that fuel restraint may play an Important role In suppressingbubble growth within the rim and, therefore, In reducing thermal conductivity. If future work Indicates that theKlnoshita"), et al, preliminary conclusions are correct, jb, d]. [b, d).

The radial power distributions at bumups of 60, 80, and 90 GWdMtU, that were used in the COPERNICtemperature predictions of the IFA 562 rods, are shown In Figure 14-9. These radial power distributions [b,d].

PAGE 14-4

- -. COPERNIC FUEL ROD DESIGN COMPUTER CODE IBAW-1 0231

(-A I,

Figure 14-3: COPERNIC and NFIRIIl1 Thermal Conductivity Comparison60 GWdttU Burnup - [c]

bc,cid]

Figure 144: COPERNIC and JAERI Thermal Conductivity Comparison - Sample No.2.63 GWd/tU Burnup,783-89%b Density Range, [cl

lb. C. dJ

I.

(OI-.. PAGE 14-

i .. - ;

I* COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Figure 14-5: COPERNIC and JAERI Thermal Conductivity Comparison - Sample No.363 GWdItU Bumup, 92-96% Density Range, [c] I

lb,C. ,

Figure 14-6: Rim Effect at 60 GWd/tUIFA 662

lb, c. .

PAGE 1T46

(,

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1023i

Figure 14.7: Rim Effect at 80 GWdltUIFA 562

[b, c, d]

Figure 14-8: Rim Effect at 90 GWdltUIFA 562

[b, c, d]

PAGE 14-7


Figure 14-9: Radial Bumup ProfilesIFA 562

fb, c, d)

PAGE 14-8


Question 3

The temperature uncertainties for LOCA and fuel melt analyses should ideally be based on data at linear heatgenerator rates (LHGRs) 2 30 kW/m because these analyses are performed at high LHGRs. The problemwith determining temperature uncertainties for bumups greater than 30 GWdlMTU Is that there is very littlemeasured centerline temperature data with LHGRs 2 30 rWV/m. Please provide the COPERNIC comparisonsto data by plotting predicted minus measured temperatures versus bumup for LHGRs 2 30 kWMm todetermine whether there Is a change in thermal uncertainty with increasing bumup, and provide theuncertainties from this data comparison. Also, provide the COPERNIC predicted minus measured centerlinetemperature data versus bumup for LHGRs 2 15 MkWm, and the uncertainties from this data comparison.These comparisons will help to verify that the uncertainties for the data that includes the lower LHGRs areapplicable to the higher LHGRs where LOCA and fuel melting analyses are performed.

ResDonse

FRA-ANP (Framatome Advanced Nuclear Power) has been performing rod average burnup based LOCAanalyses for well over a decade. Cb, d, el

.The predicted minus measured centerline temperaturedifferences at LHGRs 2 [el and [e] kW/m are shown plotted versus bumup in Figures 14-10 and 1411,respectively. The predicted minus measured centerline temperature differences that bound 95% of the datawith a 95% confidence level are [d, e] and (d, e] for LHGRs 2 [el and [e] kWr/m, respectively. Theseuncertainties demonstrate that [by (b]

PAGE 14-9

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW410231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Figure 14-10: Predicted Minus Measured Centerline Temperature Differences vs.Bumup

[el

[b, c, d]

PAGE 14-10

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1023iCOPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231

Figure 14-11: Predicted Minus Measured Centerline Temperature Differences vs.Burnup

[el

[b. c, de

PAGE 14-11

COPERNf C FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNI C FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 4

Please provide LHGRs and design Information for the EXTRAFORT test rod.

Resoonse

The EXTRAFORT test rodlet was refabricated from a mother rod that was irradiated In a commercIal 900 MWPWR reactor for five cycles to a rod average bumup of 67.2 GWd/tU. [b, c, dJ.

Table 14-1: Design Information for the EXTRAFORT Mother Rod

[b, c, dl

PAGE 14-12


Table 14-2: Thermal-Hydraulic Conditions for the EXTRAFORT Mother Rod

[b, c, dj

PAGE 14-13


Table 14-3: EXTRAFORT Mother Rod Conditions History

lb, c, d)

PAGE 14-14


Table 14-3: EXTRAFORT Mother Rod Conditions History (Continued)

[b.cMd]

PAGE 14-15


Table 14-4: EXTRAFORT Mother Rod Power Shapes

lb.c,d]

PAGE 14-16


Table 14.4: EXTRAFORT Mother Rod Power Shapes (Continued)

[b, C, d]

PAGE 14-17


Table 14-5: Design Information for the EXTRAFORT Re-fabricated Test Rodlets

[b, c, d]

Table 14.6: Thermal-Hydraulic Conditions for the EXTRAFORT Re-fabricated TestRodlets

[b. c, di

PAGE 14-18

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Table 14-7: Conditions History for the EXTRAFORT Re-fabricated Rod

lb, c, dJ

PAGE 14-19

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1023iCOPERNIC FuEL. ROD DESIGN COMPUTER CODE BAW-1 0231

Question 6

The comparison to IFA 432-1 inlet thermocouple only extends to a bumup of 9 GWdllMTU but theNUREG/CR-4717 report provides data up to a bumup of 27 GWdIMTU at the inlet thermocouple. Pleaseprovide the COPERNIC comparison up to the limit of the data or a justification why this comparison Is notvalid.

Response

The IFA 432-1 inlet thermocouple data presented in NUREGICR-4717 extends up to a local bumup of 24.072GWdItU. A comparison of the COPERNIC centerline temperature predictions with this data is presented inFigure 14-12.

PAGE 14-20

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1023f

* Figure 14.12: Measured and Predicted Fuel Temperatures vs. Burnup

[dl

0._

-or,

PAGE 14-21

II . COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231

Question 6

Is Framatome a member of Halden? If so, Halden has refabricated two high (- 59 GWd/MTU) bumup rods(one with a functional thermocouple) and plsced them first In IFA-597.2 (HWRA442) and subsequently In IFA-597.3 (HWR-543) with measured centerline temperatures. Please compare COPERNIC code predictions tothis data and Include this data In the response to Question 3.1 above.

Response

the COPERNIC centerlIne fuel temperature predictions are compared with the IFA-597.2 (HWR-442) andIFA-597.3 (HWR-543) fuel temperature measurements In Figure 14-13. This rodlet attained a bumup of 61.5GWdAU0 2 or 69.8 GWdItU.

I.,

... .AGE 1422


Figure 1-13: Fuel Centerline Temperature Measurements and Predictions vs.Burnup, IFA-597.2 and IFA-597.3

Id]

Figure 14-14: Not used

M ]

(i

(O1PAG 14-2-3

I

COPERhilC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 7

The athermal fission gas release model (Section 5.2.2) is dependent on open porosity but no values are mprovided for what is used for Framatome fuel. What values are used for open porosity? If more than onevalue is used, please provide the value for each fabrication process.

Response

The open porosity Input to the COPERNIC code Is the percentage of open porosity to the total pefletgeometric volume. The open porosity percentage of the fuel supplied by FRA-ANP's present vendor Istypically [b, dJ. The [b, dA will be used until open porosity data obtained from the fuel vendor suggests a needto Increase this value to lb, d) pellet fabrication open porosity measurements.

PAGE 14-24

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DESiGN COMPUTER CODE BAW-10231

Question 8

The Section 5.2.3.5 explanation Is not very clear about how the fission gas release model applies to varyingconditions of power and temperature. It would help to have several examples for conditions of bothincreasing temperature and decreasing temperature. Also, examples of fast and slow rate of change In fueltemperature are warranted. It appears that the resolution thickness Is not used in the final equations inCOPERNIC for computing fission gas release. Is this interpretation correct?

Response

lb. c, d]

PAGE 14-25


Figure 14-15: FGR Transition Algorithm: Small Power Change

lb, c, d]

PAGE 14-26

COPERNI: FUEL ROD DESIGN COMPUTER CODE BAW-10231

Figure 14-16: FGR Transition Algorithm: Significant Power Increase

lb, c, d]

PAGE 14-27

COPERNIC FuEL ROD DESIGN COMPUTER CODE BAW-10231

Figure 14-17: FGR Transition Algorithm: Significant Power Decrease

(b, c, d]

PAGE 14-28


Figure 14-18: Rapid Change In Fuel Temperature [d]

[b, c, d]

PAGE 14-29

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW40231

Figure 14-19: Slower Change In Fuel Temperature [d]

(b.c.d]

PAGE 14-30

COPERNIC FUEL ROo DESIGN COMPUTER CODE -- -- - AW-10231

Figure 14.20: Local Gas Concentration with Decreasing Temperatures (id])

[b, c, d]

PAGE 14-31

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231Question 9

A comparison of the COPERNIC upperbound fission gas release predictions to measured data (with > 7%measured release) from U02 - Gd2O3 fuel rods with steady-state power operation (from Table 5-3)demonstrates that the code underpredicts 2 out of 6 rods (it Is noted that one of the rods Is only slightlyunderpredicted). A comparison of COPERNIC upperbound predictions to the transient measured data with >5% release from UO2-Gd 2O3 rods (from Table 5-4) demonstrates the code underpredicts 5 out of 25 rods.This Indicates that the code's upperbound fission gas release model for U0 2 - Gd2W 3 bounds much less than95% of the data that are within the range of application for the rod pressure analysis. Also, the code does notappear to have been compared to the B&W segmented rodlets steady-state irradiated in ANO-1 and powerramped In the Studesvik R2 reactor. If not, why was this comparison not made and presented because theserods are representative of U. S. designs?

Response

lb]It can be seen from Figures 5-15 and 5-16, however, that one steady-state and two

transient U0 2-Gd203 fission gas release data points are under-predicted by the UO2-Gd2%3 bounding fissiongas release model (another steady-state data point Is only very slightiy under-predicted). [b, di.

[b, di

The Mark-BEB rodlets have been run with the COPERNIC code and the best-estimate fission gas releasepredictions are tabulated below:

RI Rodlet R3 Rodlet

FGR Measurements 9.4 11.3COPERNIC Predictions lb, d] lb. d]

[b, d].

PAGE 14-32


Figure 14-21: Upper-bound Predicted vs. Measured Steady-state Fission Gas Releasefor U0 2-Gd2O3 Fuel

lb, d]

Figure 14.22: Upper-bound Predicted vs. Measured Transient Fission Gas Releasefor UO2, MOX and U02-Gd2O3 Fuels

[b. di

PAGE 14-33

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231Question 10

The standard deviation of the gaseous swelling model is on the order of the Inferred gas porosity from themeasured porosity distributions. In fact there are only 3 data points out of 14 that have Inferred gaseousswelling greater than 0.6, i.e., significantly greater than the standard deviation. Of these 3 data points onlyone of these Is predicted well by the gaseous swelling model while the other two data points areunderpredicted by factors of 1.6 and 2.9. Therefore, the validity and the accuracy of the swelling modelappears questionable. VVhat Is the Impact of the gaseous swelling model on rod pressure, melting and strainpredictions? Does the gaseous swelling model use local bumup or pellet average bumup? The COPERNICsteady-state gaseous swelling model (Equations 6-13 to 6-15) has been programmed Into FRAPCON-3 withcalculational results of 0.007 inches of displacement at a pellet average bumup of 62 GWd/MTU with acenterline temperature of 1200C. Is this predicted displacement with this model reasonable for theseconditions? If not, further discussions are necessary to understand the gaseous swelling model.

Response

Internal gas pressure, fuel melt, and dadding diametral strain predictions with the COPERNIC gaseousswelling model turned on and off are presented in Figures 14-23 through 1425, respectively. Theserepresentative examples were generated with the typical Mark-BW17 Urania-Gadoinla cycle U02 rod casesdescribed In Chapter 12. Note that the COPERNIC gaseous swelling model lb, d]. [b. di

Predicted diametral cladding strains that contain COPERNIC lb, d].

lb, d]

Although lb, d]

PAGE 14-34


Figure 14-23: Bounding Internal Gas Pressure With and Without COPERNIC GaseousSwelling Model Effects

Typical Mark-BW17 Urania-Gadolinia Cycle, U02 Rods

[b. d]

Figure 14-24: Fuel Melt With and Without COPERNIC Gaseous Swelling ModelEffects

Typical Mark-BW17 Uranla-Gadolinla Cycle, U02 Rods

lb, d]

PAGE 14-35


Figure 14-25: 1% Cladding Strain With and Without COPERNIC Gaseous SwellingModel Effects

Typical Mark-BWi7 Uranla-Gadoflinia Cycle, U0 2 Rods

[b, dJ

PAGE 1446


Figure 14-26: Measured and Predicted Cladding Diameter Variations

ab. d]

PAGE 14-37


Figure 14-27: Measured and Predicted Cladding Diameter Variations

(b, d]

PAGE 14-38

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DEsIGN COMPUTER CODE BAW-1 0231Question 11

Figure 6-8 (from September 1999 version) predicted versus measured densification data Is significantlydifferent from Figure 6-5 of the July 1998 version of COPERNIC; however, the densification and swellingmodels appear to be the same. Please explain why the data In the two Figures are not the same.

Response

The densification and so2fd swelling models in the September 1999 and July 1998 versions are lb]. Thegaseous swelling models In the two versions [b]. This [b) contributed to the predicted density [b] shown InFigures 6-8 (September 1999) and 6-5 (July 1998). Note that the measured and predicted axes In Figures 6-8 (September 1999) and 6-5 (July 1998) are interchanged.

PAGE 14.39

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW410231COPERNiC FUEL. ROD DESIGN COMPUTER CODE BAW-10231Question 12

The fuel column growth data In Figure 6-9 (September 1999 version) appears to contain significanty lessgrowth data than the same figure (Figure 6-6) in the July 1998 version of COPERNIC. Please discuss whythere Is less data In the current version of COPERNIC. Does the column growth data in Figure 6-11 includeboth ADU and AUC processed fuel or is it just AUC fuel?

Response

The fuel column growth data shown in Figure 6-6 (July 1998) was obtained from [b] Irradiated In commercialPWRs, [d]. The fuel column growth data obtained for commercial PWRs only, which consists of the data fromthe Initial [bJ fuel rods plus an additional [bJ fuel rods, [d]. The measured and predicted fuel column growthdata from the commercial PWRs as well as [d] are all listed in Table 9-1 (September 1999). The measuredand predicted fuel column growth data obtained from Idi are shown plotted in Figure 1428. The fuel columngrowth data shown in Figure 6-11 Includes both ADU and AUC processed fuel.

PAGE 14-40

COPERN IC FUEL ROD DESIGN COMPUTER CODE BAW-10231

Figure 14-28: Measured and Predicted Fuel Column Growth (UO2) [d]

[b, d

PAGE 14-41

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231

Question 13

Equation 7-1 for creep is a function of the shear stress component (as - Cr). Please provide a derivation ofhow this shear stress Is determined to be the only active determinant of creep from Hills or Von Misesequations because these are not the only shear stress or stress components In these equations.

Resoonse

The shear stress along the slip plane (Tgn) can be related to the principal stresses by the transformation ofstresses which, expressed in tensors) notation, is

Ins = aafr

where an, and aj are the direction cosines.[c]

IC]

Von MWses'10 ) was, perhaps, the first to recognize that triaxial yielding of an Isotropic material could bedescribed by Introducing a generalized stress defined as:

CF = a1/(02% . Leo )2 + (G* _ Or r )2 + (CG _ a' )2

PAGE 14-42

Oman

COPERNI{ FUEL ROD DESIGN COMPUTER CODE BAW-10231

H1ill"" extended the Von Mises formulation to anisotropic materials such as Zircaloy with the followinggeneralized stress equation:

=jR( a _)2 +RP(cTe _0=)2 +P(Oz _a,)20 V P(R + 1)

where R and P are the anisotropy constants which have been determined by testing(1° to be as follows for thecurrent FRA-ANP Zircaloy-4 cladding:

[c] IC]

Consider the following triaxial principal stress distibution(13 ) that may be considered typical for nuclear fuelrod cladding:

ChF Q 1 - 02

Ge9 =Q1 + 02

Cu= Q1where

Q .2 P. 2 PD-r -rb

b2 _ r,2

0 2 rb (P, -Pb)rm (r2 _ r,)

andr. = cladding Inside radiusrb = cladding outside radius

rm = + 2 = cladding mean radius2

Pe = intemal pressurePb = external pressure

[ci

PAGE 14-43


Question 14

Will the FRAGEMA AFA 2G cladding that Is fabricated in Europe be used In U. S. plants? Sections 7.1.2.2.1and 7.1.2.3.1 refer to a number of cladding tube (AFA 2G) irradiation tests In the SILOE test reactor. Pleaseprovide further information on the manufacturing differences between the cladding from these tests and thosemanufactured commercially for U. S. plants, e.g., FCF Base Zr-4 and AFA 2G. Also, were the hoop stressesquoted In Table 7-1 positive or negative? (The creep model needs to be validated against the current U.S.fabricated FCF Zr4 cladding, see next question).

Response

Approval of the COPERNIC code at this time Is requested first for the advanced alloy M5 cladding only. Aresponse for this Zr-4-based question will be provided at such time that approval for the COPERNIC codeapplications to Zr-4 cladding Is requested.

PAGE 1444

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1023iCOPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231Question 15

Section 7.1.2.3.2 and Figures 7-20, 21 and 22 all refer to creep data from fuel rods Irradiated in the CAP testreactor. Pacific Northwest National Laboratory (PNNL) nor NRC is familiar with this test reactor. Pleaseprovide the test reactor or loop conditions that are pertinent to in-reactor creep such as coolant Inlet-outlettemperatures, fast flux, system pressure, etc. Also, provide predicted versus measured creep for the FCF Zr-4cladding used In the U.S. and the background information on this data.

Response

Approval of the COPERNIC code at this time is requested first for the advanced alloy MS cladding only. Aresponse for this Zr-4-based question will be provided at such time that approval for the COPERNIC codeapplications to Zr-4 cladding Is requested.

PAGE 14.45

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW 10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 16

Section 7.1.2.3.3 notes that creep data from one rod was excluded from the uncertainty determinationbecause it was next to gadolinla rods. Does this mean that the creep model uncertainty does not apply to fuelrods near gadoflnia rods? Also, Figure 7-20 shows a considerable amount of measured-to-predicted data thatare outside of the uncertainty bounds proposed. Please discuss why it Is ok to discard this data from theuncertainty determination for creep and those data In Figure 7-20 that are not within the proposed boundingcreep uncertainty. Please Identify those analyses where over prediction of creep Is conservative and thoseanalyses where under prediction is conservative.

Response

Approval of the COPERNIC code at this time Is requested first for the advanced alloy M5 cladding only. Aresponse for this Zr4-based question will be provided at such time that approval for the COPERNIC codeapplications to Zr-4 cladding is requested.

PAGE 14-46

M�


Question 17

Section 7.1.3.2.1 discusses the development of the M5 creep model from tube Irradiations but no comparisonto this data Is provided, and the stress and temperature parameters of this data are also not provided. Pleaseprovide this data and comparisons to the M5 creep model. It is also stated that the secondary thermal creeprate is independent of alloy type, but no data is presented to corroborate this statement Please provide thisdata.

Resoonse

The advanced alloy M5 creep rate is modeled as lb, d].

PAGE 14-47

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Table 14-8: Calibration Database for Creep Hardening Effects[b]

PAGE 14-48


Figure 14-29: Advanced Alloy M5 Creep Tests of Unirradiated Tubes at [bi

[b, d]

PAGE 14 49


Figure 14-30: Advanced Alloy M5 Creep Tests of Unirradiated Tubes at [b]

[b, d]

PAGE 14-50


Figure 14-31: Advanced Alloy MS Secondary Thermal Creep Rate vs. Fluence

[b, d]

PAGE 14-51


Figure 14-32: [bI Creep Strain vs. Fluence[b]

[b, d]

PAGE 14-52

COPEURNIC FUEL ROt) DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 18

Does COPERNIC consider the effects of cladding growth (Section 7.3.2) in the diametral direction or is thisImplicit in the creep data? Also, the upperbound model underpredicts a significant amount of growth data InFigure 7-50. Please explain why this Is acceptable. The alloy 5 growth model, Equation 7-27, Is not linearlydependent but has a decreasing slope with fluence while the majority of ircaloy growth data show a lineardependence with fluence. In addition, an Initial examination of Figure 7-54 appears to show that a lineardependent model would provide as good or better prediction of the alloy 5 growth data compared to the modelproposed. Please provide further Information on why Equation 7-27 is more appropriate for predicting alloy 5growth even through a linear model would be more conservative and provide as good a fit to the growth data.

Response

The effects of cladding growth (Section 7.3.2) in the diametral direction lb, d]. FRA-ANP [b, d]. [b, d]. Recentadditional measured data(14) with rod average fluences up to I x 1 n/cm2, E >1.0 MeV, demonstrates lb, d].

PAGE 14-53

COPERNIC Fur-a ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 19

Section 8.1 discusses the COPERNIC corrosion model and comparisons to data. The coolant inlettemperatures are provided for some of the reactors from which corrosion data was taken but coolant outlettemperatures and LHGR are also important parameters. What were the outlet temperatures and averageLHGR/cycde for both the Zr-4 and ally 5 data (including the adjustment rod data) and Identify high duty,medium duty and low duty plants (see 20 below)? Section 8.1.3.2 states that the alloy 5 data is based on themaximum average azimuthal oxide thickness over the span height Please discuss how this is determinedfrom actual measurements, e.g., Is It an average over a given length and how many azimuthal orientationsare measured?

ResDonse

Approval of the COPERNIC code at this time Is requested first for the advanced alloy MS cladding only. Aresponse for the Zr-4-based portion of this question will be provided at such time that approval for theCOPERNIC code applications to Zr-4 cladding is requested.

The Inlet and outlet temperatures and the average linear heat generation rate (LHGR) for each cycle arepresented In Table 14-9 for the MS data. FRA-ANP fb,e]. However, [b, el. Average azimuthal oxidethicknesses are evaluated [b, d, e]. The maximum oxide thickness [b, el.

PAGE 14-54

COPERNIC: FUEL ROD DESIGN COMPUTER CO-DE BAW-10231

Table 14-9: Plant and Fuel Rod Data

CORE CORE Average LHGR I cycle PlantPlant INLET OUTLET (kWIm) Classification

TEMP. ( C) TEMP. ('C) lb Il

[b. d. e]

Plant Classification:lb. e]

PAGE 14-55


Also, oxide predictions and comparisons to data from Zr-4 are provided for all axial rod locations; however,NRC Is most concerned with rod locations that experience maximum oxide thicknesses and corrosion fromhigh duty plants. The axial locations with maximum oxide thicknesses are typically In the next to last span orthe next to last two spans from the top of the assembly depending on the number of spacer grids perassembly. Please provide predicted minus measured versus both bumup and measured oxide thickness onlyfor those axial spans with maximum measured oxide thickness for each rod and Identify high duty, mediumduty and low duty plants as well as defining the differences between the operating parameters of thesedifferent plants.

Response

Approval of the COPERNIC code at this time Is requested first for the advanced alloy M5 cladding only. Aresponse for this Zr4-based question will be provided at such time that approval for the COPERNIC codeapplications to Zr-4 cladding Is requested.

PAGE 14-56


From examination of Figure 8-11, the COPERNIC code appears to significantly underpredict a large amountof the measured oxide data from U. S. plants with Zr-4 cladding. Please provide predicted minus maximummeasured oxide thickness only from rods from U. S. plants using both the COPERNIC and COROS02corrosion models. Please provide predictions of this same U. S. data using the COPERNIC upper boundcorrosion model. PNNL's comparison of the COROS02 and COPERNIC corrosion models at varioustemperatures for both Zr-4 and alloy 5 has demonstrated that COROS02 predicts the greater oxidethicknesses. Please discuss why this Is acceptable.

Response

Approval of the COPERNIC code at this time is requested first for the advanced alloy M5 cladding only. Aresponse for the Zr-4-based portion of the question will be provided at such time that approval for theCOPERNIC code applications to Zr-4 cladding is requested.

The COROS02 and COPERNIC advanced alloy M5 models differ. The pre-trmnsition phase of theCOPERNIC M5 model uses a [el function rather than [e] used in COROSO2. This change provided [b]between oxide thickness measurements and predictions. Also, the Initial COROS02 model was developedwith id] (see the response to Question 22), while the COPERNIC corrosion model was developed with theoxide thermal conductivity relationships described In Section 8.1.2.1.

PAGE 14-57

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPENICFUELRODDESGN CMPUER ODE AW-023

Question 22

What Is the basis for the oxide layer thermal conductivity functions provided at the bottom of page 8-3? Itappears that the oxide thermal conductivity Is determined based on the oxide surface temperature. Is thisInterpretation correct?

Response

Experimental data from a CEA (Commissariat A I'Energle Atomique - Atomic Energy Commission) programprovided the basis for the COPERNIC oxide thermal conductivity relationships presented on page 8-3. TheCOPERNIC oxide thermal conductivities and the NFIR4II (Nuclear Fuel Industry Research Program) thermalconductiV.ty.16 are plotted together for comparison In Figure 14-33. [b, dj.

PAGE 14-58


Figure 1443: Oxide Thermal Conductivity vs. Temperature and Oxide LayerThickness

[b, d]

PAGE 14.59

COPERNIC FUEL ROD DEIGN COMPUTER CODE BAW410231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 23

Please provide the average LHGR/cycle for the hydrogen pickup data provided In Figure 8-22. Theapplicability of using only 5 cycle data to estimate the hydrogen pickup fraction Is questionable because theremaybe other factors (such as heat flux) In the 3 and 4 cycle data that results in the 5 cycle data giving thelowest hydrogen pickup fractions.

Response

Approval of the COPERNIC code at this time Is requested first for the advanced alloy MS cladding only. Aresponse for this Zr-4-based question will be provided at such time that approval for the COPERNIC codeapplications to Zr-4 cladding is requested.

PAGE 14-60


Please provide the background data for the fuel melting temperature relationship used by COPERNIC(Equations 10-11 and 12-2).

Response

[c] . Open and closed systems forheating the test samples have been used in fuel melt experiments. Closed systems*4,28) are generallypreferred because the test sample Is enclosed in a hermetically sealed crucible with a controlled atmospherethat restrains stoichiometry changes. Open systerns(a'm2 ) without controlled atmospheres, on the otherhand, are notorious for causing stoichiometry changes that introduce errors in the measured melttemperatures. Melt temperature measurements have traditionally been performed either by post-coolingobservations of microstructural changes or by the thermal arrest method where a marked change in the slopeof the measured fuel temperature Is observed. The most recent measurements have typically beenperformed with a closed system and the thermal arrest method because this approach Is generally consideredto produce more accurate measurements. The melt temperatures of unin-adiated U02 determined by variousInvestigators are listed below.

Table 14-10: Unirradiated U02 Melt Temperature

Reference Year Melting Point (C)Lambertson 0 1953 2878Wisniyl (19J1957 2760Ehlert Z 1958 2860Christensen 2" 1962 2790Christensen 22 1963 2800Pashos A'_ _ _1965 2800Hausner 1965 2805Bannister ) 1967 2860Benz A26) 1970 2810Rubin _ 27J 1970 2840Tachibana 1985 2845Chotard "" 1987 2852

lb, d].

Christensen's work("21J) has traditionally been used for fuel melt because It Is generally considered to beconservative. However, Christensen's measurements were performed In an open system and thestoichiometry of the test samples was not recorded. Furthermore, his initial datas'8 did not report a decreaseIn melt temperature with bumup.

[b, c, d]

[b, c. d]

PAGE 14-61

COPERNIC FUEL ROD DESIGN COMPUTER CODE * BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

[b, d . Yamamouch(31, et at data.

A review of the available UOGd2O3 fuel melt data'2 Indicates that there is no significant differencebetween the U02 and U02-Gd2O3 fuel melt temperatures for gadolinia concentrations less than approximately12wt.%.

Table 14-11: UnIrradlated (U,Gd)02 Melt Temperature

Gd2%3 Content (wt.%) Meltin Poin Referenceo 28574 28818 2861 Chotard>'12 286716 2865o 28444 2851a 2658 Buschm12 2836 Stoichiometric40 2791 oxides65 2585100 24444 28558 2856 BuschP2)12 2845 substoichiometic40 2786 oxides65 25970 28420 28816 2828

6.6 28688 2842 Watarumi43 '10 2828 (homogeneous13 2775 oxides)19 274525 275331 270737 26876 2836 Wataruml`&8 2836 (heterogeneous10 2829 oxides)

ld].

PAGE 14-62


Question 25

Section 12.0 notes that COPERNIC Is used for Initialization of core thermal-hydrautic codes. Please fist thosecalculated COPERNIC parameters used for initialization and the specific applications of the thermal-hydrauliccodes.

Resoonse

The COPERNIC parameters used for initialization of thermal-hydraulic codes (COBRA-IV, COBRA3C,LYNXT, etc.) are Id). The specific thermal-hydraulic applications where fuel performance code initializationpredictions are used Include those related to the evaluation of locked rotor, ejected rod, etc. events.

PAGE 14-63

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW410231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 26

Section 12.1.1 (page 12-2) under discussion on Code Uncertainties It Is noted that the code has an optionthat conservatively bounds the fissions gas release data and that this option Is used to bound the fission gasrelease for the rod pressure predictions. However, there is a concern that this option will not bound the U02 -Gd2O3 data within the fission gas release range that Is Important to the rod pressure analysis for U02 -Gd2O3 rods (see Question 9 above) at the stated level of conservatism. Please discuss this Issue further,particularly in relation to Question 9 above.

Response

Id); see the response to Question 9.

PAGE 14-64

�EE I.


Question 27

Section 12.1.1 (page 12-3) under the discussion on Transients it Is noted that plant specific operating datamay be used to establish simulated transients. Please explain further by what Is meant by sufficient plantoperating data and provide an example.

Response

The COPERNIC end-of-ife (EOL) internal gas pressure analyses employ [dJ. In addition, [d;. The Condition-Idesign transients discussed In section 12.1.1 account for lb, d]. All plants have the equipment andprocedures necessary to gather the operational data required for core follow activities. The plant datagathered includes the time-dependent behavior of the reactor thermal power level, regulating rod and (whereapplicable) axial power shaping rod positions, RCS boron concentration, average moderator temperature, andaxial power Imbalance. In general, these data are collected approximately on an hourly basis. Most plantshave the equipment necessary to electronically archive this data.

(el. It would never be possible to eliminate the [e] lb, d] gas pressure analysis because leJ. Generally, it mayonly be possible, due lb, dl, le]. The [el Condition I transient in the Internal gas pressure analysis examplespresented In Chapter 12 produced pressure lb, d] less than approximately Id].

The Condition-l des9ritransients used in the COPERNIC Internal gas pressure analyses are [b]. They areproduced with lb, d]m' (xenon distribution and control rod insertion) that would be lb, dJ. lb, d] the typicalUrania-Gadolinia cycle documented in Chapter 12 are shown In Figure 14-34. This example Includes lb, d]with [b, d] andi[e]. The [b] gas pressure difference Id]. This example demonstrates [d, el.

PAGE 14-65


Figure 14-34: Typical Mark-B Urania-Gadolinla Cycle Predictions [bi

lb]

PAGE 14X66


Question 28

(* There Is a concern that the uncertainty factor provided in Equation 12-1 may be too small at the predictedoperating temperatures (stored energy) calculated for LOCA initialzation. Please discuss this Issue further,partlcularly in relation to Question 3 above.

Response

[b, d, el.

I..

PAGE 14-7


Question 29

Are any of the example calculabons provided In Section 12 for fuel cores with two 24-month cycles? Itappears that there are no 24-month cycle results presented for the Mark BW-17 design. If so, please explainbecause It is anticipated that a large number of plants wil be switching to 24-month cycles In the next fewyears.

.O

Response

[b. di].

(O

PAGE 14-68

-' 'VIA ---- ... -.... -............... .. . ...


Question 30

The axial power distributions for the transients were found for the example licensing analyses, but the powerdistribution for the steady-state power operation were not found for the topical report Please provide theseaxial power distributions. If there are more than 20 axial power profiles Kt would be helpful to condense thenumber down to 20 or less. Also, the steady-state power histories are only provided as plots versus bumup.Please provide these In tabular form to support the NRC audit calculation of these calculational examples?

Resnonse

Forty-seven data sets of the normalized axial power distributions used for the Chapter 12 examples are listedin Table 14-12. The axial power distributions of each data set are provided in Table 14-13 through Table 14-59. The rod average bumups and linear heat generation rates are listed above each distribution presented.

Several different sets of axial power distributions are provided:- [bd,e]- lb,d,eJ- Ib.d,el

[e] are not presented as separate sets because [el.

All of the steady-state axial power distributions used for the Chapter 12 examples are provided. If thereviewers need 20 or less distributions, it Is left up to their discretion to select the appropriate distributionsfrom those provided.

PAGE 14-69

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAWl10231

Table 14-12: Axial Power Distribution Data Sets

Typical Mark-B U0 2 Cycle, U02 Power HistoriesTable 14-13Table 14-14Table 14-15Table 14-16Table 14-17Table 14-18Table 14-19Table 14-2DTable_14-21 Typical Mark-B U0 2-Gd20% Cycle, U0 2 Power HistoriesTable 14-21Table 14-22Table 14-23Table 14-24 id]Table 14-25Table 14-28

Table 14-27

Table 14-408d

TyIcal Mark-B UO24d203 Cycle UO2-GdO 3 Power HistoriesTable 14-29Table 14-30Table 14-31Table 14-32 [d]Table 14-33Table 14-34Table 14-35

Typical Mark-13W Ut~h Cycle, U02 Power HistoriesTable 14-36Table 14-37Table 14-38Table 14-39 dTable 14-40 dTable 14-41Table 14-42Table 14-43

Table14^44 Typical Mlark-SW U~h-Gdg~h Cycle, U02 Power HistoriesTable 14-44Table 14-46Table 14-46Table 14-487k

Table 14-49Table 14-50Table 14-51Table 14-52

Tabl 146 yloaf M~arkc4BW U02-Gd203 Cycle, UO2-Gd2O3 Power HistoriesTable 14-53

Table 14-55Table 14-566dTable 1457Table 14658Table 14-59

Apeffeations:

Id)

PAGE 14-70


Table 1413: Typical Mark-B Uranium-Dioxide Cycle

(d]

[dl

PAGE 14-71.


Table 14-13: Typical Mark-B Uranium-Dioxide Cycle(dJ

Id)

PAGE 14-72

COE I FUE RH EINCMUERlOD BAW-10231

Table 14-13: Typical Mark-B Uranium-Dioxide Cycle[d]

[di

PAGE 14-73


Table 14-13: Typical Mark-B Uranium-Dioxide CycleIdi

[d]

PAGE 14-74



[d]

PAGE 14-76


Table 14-15: Typical Mark-B Uranium-Dioxide Cycle

[d]

[dJ

PAGE 14-76


Table 14-15: Typical Mark-B Uranium-Dioxide CycleId]

Id]

.

PAGE 14-77



[do

[d]

PAGE 14-78

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW 10231


[d]

PAGE 14-79


Table 14-16: Typical Mark-B UranIum-DIoxide CycleId)

[dl

PAGE 14-80


Table 14.17: Typical Mark-B Uranium-Dioxide Cycle

[d]

PAGE 14-81


Table 14.17: Typical Mark-B Uranium-Dloxide CycleId]

[d]

PAGE 14-82


Table 14-17: Typical Mark-B Uranium-Dioxide Cycle{d]

]

PAGE 14-83

COPERNIC FUEL ROD DESIGN COMPUTER CODE 1BAW-10231

Table 14-18: Typical Mark-B Uranium-Dioxide CycleId)

[d]

PAGE 14-84


Table 14-18: Typical Mark-B Uranium-Dioxide Cycle[di

[dJ

PAGE 14-85

__ _

COPERNIC FUEL RoD DESIGN COMPUTER CODE BAW-10231

Table 14-18: Typical Mark-B Uranium-Dioxide Cycle(dl

PAGE 14.86



[d]

PAGE 14-87


Table 14-19: Typical Mark-B Uranium-Dioxide Cycle[di

[dJ

PAGE 14-88



[d]

PAGE 14-89


Table 14-19: Typical Mark.B Uranium-Dioxide Cycle[d]

[d]

PAGE 14-90



[dJ

[d]

PAGE 14-91



1~

PAGE 14-92

COPERNIC FUEL ROo DESIGN COMPUTER CODE BAW-10231

Table 14.20: TypIcal Mark-B Uranlum-Dioxide Cycle[d]

[d]

PAGE 14-93

COPERNIC FUEL ROo DESIGN COMPUTER CODE BAW-10231


[d]

PAGE 14-94


Table 14-20: Typical Mark-B Uranium-Dioxide Cycle[dJ

[dl

PAGE 14-95

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW.10231


[dj

PAGE 14-96


Table 14-20: Typical Mark-B Uranium-Dioxide Cycle(di

[d]

PAGE 14-97


Table 14-21: Typical Mark-B Urania-Gadolinla Cycle[d]

Id]

PAGE 14-98

COPERNIC FUEL ROD DESIGN CompurTER CODE BAW-10231

Table 14-21: TypIcal Mark-B Urania-Gadolinia Cycle[d]

[d]

PAGE 14-99


Table 14-21: Typical Mark-B Uranla-Gadolinla Cycle[d]

mN

PAGE 14-100



[dc

PAGE 14-101


Table 14-22: Typical Mark-B Urania-Gadolinla CycleId]

[d]

PAGE 14-102


Table 14-23: Typical Mark-B Uranla-Gadolinia Cycle[d]

(d]

PAGE 14-103


Table 14-23: Typical Mark-B Urania-Gadolinia Cycle[d]

id]

PAGE 14-104


Table 14-24: Typical Mark-B Uranla-Gadolinia Cycle[d)

(d]

PAGE 144105



[d]

PAGE 14.106


Table 14-25: Typical Mark-B Uranla-Gadolinla Cycle

(Id

PAGE 14.107

COPERN IC FUEL ROD D)ESIGN COMPUTER CODE BAW410231

Table 14-25: Typical Mark-B Uranla-Gadolinla Cycle[dl

{d]

PAGE 14-108

COPERNIC FUEL ROD DESIGN COMPUTER CODE * BAW-10231


[dN

PAGE 144109


Table 14-26: Typical Mark-B Uranla-Gadolinia Cycle[dl

[d]

PAGE 14-110


Table 14-26: Typical Mark-B Uranla-Gadolinla Cycle[dJ

[d]

PAGE 14-11 1


Table 14.26: Typical Mark-B Uranla-Gadolinia CycleIdl

[di

PAGE 14-112


Table 14-26: Typical Mark-B Uranla-Gadolinla Cycle[d)

Id)

PAGE 14-113


Table 14.27: Typical Mark-B Uranla-Gadolinla Cycle[d]

[d]J

PAGE 14-1 14

COPERNIC FUEL Ron DESIGN CompurrER CODE BAW-10231


[di

PAGE 14-115



[d]

PAGE 14-116


Table 14-27: Typical Mark-B Uranla-Gadolinia Cycle[d]

[d)

PAGE 14-117


Table 14-28: Typical Mark-B Uranla-Gadolinla CycleIdl

[d]

PAGE 14-18


Table 14-28: Typical Mark-B Urania-Gadolinia Cycle[dJ

[d]

PAGE 14-119


Table 14-28: Typical Mark-B Uranla-Gadolinia CycleId]

[Co

PAGE 14.120


Table 14.28: Typical Mark-B Uranla-Gadolinia CycleIdi

Id]

PAGE 14.121

COPERNIC IFUEL ROD DESIGN COMPUTER CODE BAW-10231

Table 14-29: Typical Mark-B Uranla-Gadolinia Cycle

[dl

PAGE 14-122



[d]

PAGE 14-123


Table 14-29: Typical Mark-B Uranla-Gadolinia CycleIdl

[d]

PAGE 14-124


Table 14-29: Typical Mark-B Urania-Gadolinia Cycle

[d]

[dJ

PAGE 14-125



[d]

PAGE 14-126



[d]

PAGE 14-127


Table 14-31: Typical Mark-B Urania-Gadolinia CycleId]

[d]

PAGE 14-128


Table 14-32: Typical Mark-B Urania-Gadolinia Cycle

[d]

PAGE 14-129

COPERNIC FUEL Roo DESIGN COMPUTER CODE BAW-10231

Table 14.32: Typical Mark-B Uranla-Gadoflnla Cycle[d]

[dl

PAGE 14.130


Table 14-32: Typical Mark-B Urania-GadolInia Cycle[d]

(dl

PAGE 14-131


Table 14.33: Typical Mark-B Urania-Gadolinia CycleId]

Id)

PAGE 14-132


Table 14-33: Typical Mark-B Uranla-Gadolinia Cycle[di

Id]

PAGE 14-133



tm

PAGE 14.134



[d]

PAGE 14-135


Table 14-34: Typical Mark-B Uranla-Gadolinla Cycle

id]

[dc

PAGE 14-136

COPERNIC FUEL ROD DESIGN COMPUTER C0ODE BAW 10231

Table 14.34: Typical Mark-B Urania-Gadolinla Cycle(d]

[dl

PAGE 14-137

COPERNIC FUEL ROD DESIGN COM1PUTER CODE BAW-10231

Table 14-34: TypIcal Mark-B Urania-Gadolinia Cycle[d]

Id]

PAGE 14-138


Table 14-35: Typical Mark-B Uranla-Gadolinia Cycle[d)

Id]

PAGE 14-139



[d]

PAGE 14-140

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231-


[dJ

PAGE 14-141


Table 14.35: Typical Mark-B Urania-Gadolinia Cycleid]

id]

PAGE 14-142


Table 14.36: Typical Mark-BW17 Urania-Dioxide Cycle[d]

[d]

PAGE 14-143


Table 14-36: Typical Mark-BW17 Urania-Dioxide Cycle(dJ

id]

PAGE 14-144


Table 14436: Typical Mark-BW17 Urania-Dioxide Cycle1

[dl

PAGE 14-145

COPERNIC FUE=L ROD D3ESIGN COMPUTER CODE BAW-10231

Table 14-36: Typical Mark-BW17 Urania-Dioxide Cycle(d]

1d]

PAGE 14-146


Table 14-36: Typical Mark-BW17 Urania-Dioxide Cycle[dJ

Id]

PAGE 14-147

I


Table 14-37: Typical Mark-BW17 Urania-Dioxide Cycle[di

[dJ

PAGE 14-148


Table 14-38: Typical Mark-8W17 Urania-Dioxide CycleIdi

[d]

PAGE 14-149


Table 14438: Typical Mark-BW17 Uranla-Dioxide Cycle[d]

[d]

PAGE 14-150


Table 14-39: Typical Mark-BW17 Urania-Dioxide Cycle[d]

[NQ

PAGE 14-151

COPERN1C FUEL ROD DESIGN COMPUTER CODE BAW-10231

Table 14-39: Typical Mark-BWI7 Urania-Dioxide Cycle

[dJ

PAGE 14-162


Table 14-39: Typical Mark-6W17 Urania-Dioxide Cycle[dJ

[dJ

PAGE 14-153


Table 14-40: Typical Mark.BWi7 Urania-Dioxide Cycle[Id

[d]

PAGE 14-154

COPERNIC FUEL ROD DESIGN COMPUTER CODE SAW-110231

Table 1440: Typical Mark-BW17 Uranla-Dioxide Cycle[d]

[d]

PAGE 14-1 55


Table 14-40: Typical Mark-BWI7 Uranla-Dioxide CycleEd]

[di

PAGE 14-156


Table 1441: Typical Mark-BW17 Urania-Dioxide Cycle[d]

[d]

PAGE 14-157



[d]

PAGE 14-158


Table 14-41: Typical Mark-BW17 Urania-Dioxide Cycle[dl

Id]

PAGE 144159


Table 1441: Typical Mark-BW17 Urania-Dioxide Cycle[d]

[dl

PAGE 14-160


Table 14.42: Typical Mark-BW17 Urania-Dioxide Cycle[d]

[d]

PAGE 14-161


Table 14 42: Typical Mark-BW17 Urania-Dioxide Cycle

[d]

[d)

PAGE 14-162


Table 14 42: Typical Mark-BW17 Urania-Dioxide Cycle[d]

[d]

PAGE 14-163


Table 14-42: Typical Mark-BWi7 Urania-Dioxide Cycle[dJ

Id]

PAGE 14-164


Table 14.43: Typical Mark-BW17 Urania-Dioxide Cycle

[dJ

[d]

PAGE 14-165


Table 14 43: Typical Mark-BWi7 Urania-Dioxide Cycle[d)

[d]

PAGE 14-166



[di

PAGE 14-167


Table 14-43: Typical Mark-BW17 Urania-Dioxide Cycle

[d]

[dl

PAGE 14-168

COPE-RNIC FUEL ROD DESIGN COMPUTER CODE BAW 10231

Table 14.44: Typical Mark.BW17 Urania-Gadolinla Cycle[d]

[dl

PAGE 144169


Table 14-44: Typical Mark-BW17 Uranla-Gadolinla Cycle[d]

[d]

PAGE 14-170


Table 14-44: Typical Mark-BWi7 Uranla-Gadolinla Cycle

[Id

Id]

PAGE 14-171


Table 1444: Typical Mark-BW17 Urania-Gadolinla Cycle[d]

Cd]

PAGE 14-172

C:OPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231

Table 14-44: TypIcal Mark-BW17 Urania-Gadolinla Cycle[dJ

Id]

PAGE 14-173


Table 14-45: Typical Mark-BW17 Urania-Gadolinia Cycle

(di

[d]

PAGE 14-1 74


Table 14-46: Typical Mark-BW17 Uranla-Gadolinla Cycle

[dJ

[CU

PAGE 14-175

COENCFULRDDSINCMUERCD BAW 10231

Table 14-46: Typical Mark-BWIT7 Urania-Gadolinla Cycle[d]

[am

PAGE 14-176

COPERNI FUEL RO DEsGN COPUE CODEs BAW-10231

Table 1447: Typical Mark-BW17 Uranla-Gadolinia Cycle[d]

[d]

PAGE 14-177


Table 14-47: Typical Mark-BWI7 Uranla-Gadolinla Cycle[dj

[dJ

PAGE 14-178

COPERN IC FUEL ROD DESIGN COMPUTER CODE : BAW-10231

Table 14-48: Typical Mark.BW17 Urania-Gadollnia Cycle[d]

(dj

PAGE 14-179


Table 14-48: Typical Mark-BW17 Uranla-Gadolinia Cycle[d}

[d]

PAGE 14-180

COPERNIC FUEL ROD DE-SIGN COMPUTER CODE BAW410231

Table 14-48: Typical Mark-BW17 Urania-Gadolinia CycleId]

Id]

PAGE 14-181


Table 14.49: Typical Mark-BW17 Urania-Gadolinla Cycle

[dJ

PAGE 14-182


Table 14.49: Typical Mark-BW17 Uranla-Gadolinla Cycle[d]

[dJ

PAGE 14-183


Table 14.49: Typical Mark-BW17 Urania-Gadolinla Cyclefdj

Id]

PAGE 14-184


Table 1449: Typical Mark-BWi7 Urania-Gadolinla Cycle[di

[d]

PAGE 14-185

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1D231

Table 14-50: Typical Mark-BW17 Uranla-Gadolinla Cycle[di

Id]

PAGE 14-186


Table 14.50: Typical Mark-BW17 Uranla-Gadolinla Cycle

(d]

PAGE 14-187

COPERNIC FUEL Rol) DESIGN COMPUTER CODE BAW 10231

Table 14-50: Typical Mark-BW17 Uranla-Gadolinia Cycle[d]

td]

PAGE 14-188


Table 14-50: Typical Mark-BW17 Uranla-Gadolinla Cycle(d]

[d)

PAGE 141 89


Table 14-50: Typical Mark-BW17 Uranla-Gadolinia CycleId]

Id]

PAGE 14-190

_OE I FUE RO DEINCMPTRCD BAW-10231

Table 14-51: Typical Mark-BW17 Uranla-Gadolinla Cycle[dj

id]

PAGE 14-191

COPERNIC FUEL ROD DESIGN COMPUTER CODE ~~BAW-10231

Table 14-51: Typical Mark-BW17 Urania-Gadolinla CycleIdJ

Id)

PAGE 14-192

COPERNIC FUEL Roo DESIGN COMPUTER CODE BAW-10231


[d]

PAGE 14-193


Table 14.51: Typical Mark-BW17 UraniaGadolinla Cycle(dl

Id1

PAGE 14-194

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1023iCOPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Table 14-51: Typical Mark-BW17 Urania-Gadolinia Cycle63 GWd/tU, U02 Single Uimiting Rod Analyses (Continued)

PAGE 14-195

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW 1023i

Table 14-52: Typical Mark-BW17 Uranla-Gadolinla CycleU02 Max Burnup Rod, Cladding Oxide

PAGE 14-196


Table 14-52: Typical Mark-BW17 Uranla-Gadolinla Cycle(di

[dt

PAGE 14-197

COPERNIC FUEL ROD DESIGN COMPUTER CODE * BAW-10231


Id]

PAGE 14-198



md]

PAGE 14-1 99


Table 14-52: TypIcal Mark-BW17 Urania.Gadolinta CycleIdi

Td]

PAGE 14-200



[d]

PAGE 14-201


Table 14-53: Typical Mark-BWI7 Urania-Gadolinla Cycle

(d)

[d]

PAGE 14-202

-COPERN IC FUEL ROD DESIGN COMPUTER CODE BAW 10231

Table 14-53: Typical Mark-BW17 Urania-Gadolinia Cycle

[d]

PAGE 14.203

COPERNIC FUEL ROD DEsIGN COMPUTER CODE BAW-10231

Table 14.53: Typical Mark-BWi7 Urania-Gadolinia Cycle[d]

[d]

PAGE 14-204


Table 14-54: Typical Mark-BW17 Urania-Gadolinla Cycle[d)

(dj

PAGE 14-205


Table 14-55: TypIcal Mark-BW17 Urania-Gadolinia Cycle

[d]

[d)

PAGE 14-206


Table 14-55: Typical Mark-BW17 Uranla-Gadolinla Cycle[d)

[dl

PAGE 14-207


Table 14-56: Typical Mark-BW17 Urania-Gadolinla Cycle[d]

[d]

PAGE 14-208


Table 1446: TypIcal Mark-BW17 Uranla-Gadolinia Cycle[d]

[dJ

PAGE 14-209


Table 14-56: Typical Mark.BW17 Urania-Gadolinla Cycleid]

Idl

PAGE 14-210

COPERNIC FuEL ROD DESIGN COMPUTER CODE SAW-10231

Table 14-67: Typical Mark-BW17 Urania-Gadolinla Cycle[d]

[d]

PAGE 14-211



(dj

PAGE 14-212


Table 14.57: Typical Mark-BW17 Uranla-Gadolinla Cycle[d)

[dJ

PAGE 14-213


Table 14-58: Typical Mark-BW17 Uranla-Gadolinla CycleId]

[d]

PAGE 14-214

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231 I

Table 14-58: Typical Mark-BWV7 Uranla-Gadolinla Cycle[d] I

Idi

PAGE 14-215


Table 1458: Typical Mark-BW17 Uranla-adolinla CycleId]

[d)

PAGE 14-216


Table 14-58: TypIcal Mark-BW7 Urania-Gadolinia Cycle[di

[d]

PAGE 14-217

COPERNIC FUEL ROD DEsIGN COMPUTER CODE BAW-10231

Table 14-59: Typical Mark-BWi7 Uranla-Gadolinla Cycle[d)

Itd1

PAGE 14-218



[IC

PAGE 14-219


Table 14-59: Typical Mark-BW17 Urania-Gadolinia Cycle(dl

[d]

PAGE 14-220

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAWl10231

Table 14.59: Typical Mark-BW17 Urania-Gadolinia Cycle[dl

id]

PAGE 14-221

COENCFULRDDEINCMPTRCD IBAW-1023iCOPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231Question 31

Section 12.4.1 states that the cladding strain analysis win be run with ...... and the gaseous swelling optionturned off. Performing these analyses without gaseous swelling ...... produces more accurate predictions .at the very high local power levels that accompany these analyses. This appears to be contradictory to thecomparisons to data In Figures 6-17,6-18, and 6-19 that demonstrate that COPERNIC with gaseous swellingoption turned on provides an adequate prediction of diametral strains. Please provide data that supports theconclusion that the exclusion of gaseous swelling In COPERNIC produces more accurate strain predictions.

Response

Ce)

Ramp tests were recently performed to study the effects of pellet cladding mechanical Interaction (PCMI) oncladding delormnationr 1 . Three rodilets, which were part of a segmented rod, were ramped to terminal powerlevels ranging from 39.5 to 41.5 kwlm. The ramp time to the terminal power levels was approximately 2-min.and the rods were held at the terminal power levels for 0 (zero hold time), 16-min. and 12-hrs. Although theterminal power level of these rods was well below fuel melt, It was sufficient to cause partial dish filling for therodlets with hold times of 16-min. and 12-hrs. These power ramps were simulated with 2 and 3 dimensionalfinie element codes that don't contain gaseous swelling models. The overall computational results of the 2Dfinite element code agreed well with the data from the rodlets with terminal power level hold times of zero and16-min. but the cladding diameter changes for the 12 hour hold time were underestimated because gaseousswelling was not included. The 3D finie element code, which is currently under development, tended tomoderately overestimate the measured cladding diameter changes for these cases. (el.

PAGE 14-222

COPERNIC FUEL ROD DESIGN COMPUTER CODE * ~BAW 10231.COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 32

Section 12.5 states that COPERNIC will be used to generate cladding creep collapse Initial conditions andexample rod pressure results are provided in Figures 12-34 and 12-35. Are there any other Initial conditionsprovided by COPERNIC for the creep collapse analysis, e.g., cladding temperatures? If so, please providepredictions of these initial conditions.

Besponse

The other initial conditions that are provided by COPERNIC for the creep collapse analysis Include le].

PAGE 14-223

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1023iCOPERNIC FuEL ROD DESIGN COMPUTER CODE BAW-1 0231

Figure 14.35: Typical Mark-B Fuel CyclesCreep Collapse Analyses [d]

[d]

Figure 14-36: Typical Mark-BWI7 Fuel CyclesCreep Collapse Analyses [d]

[Co

PAGE 14-224


Figure 14-37: Typical Mark-B Fuel CyclesCreep Collapse Analyses [d]

[d]

Figure 14-38: Typical Mark-BW17 Fuel CyclesCreep Collapse Analyses [d]

[d]

PAGE 14-225

COPERNIC FUEL ROD DESIGN COMPUTrER CODE BAW-10231


[d]

Figure 14.40: Typical Mark-BW17 Fuel CyclesCreep Collapse Analyses [dJ

md

PAGE 14-226



[d]

Figure 14.42: Typical Mark-BW17 Fuel CyclesCreep Collapse Analyses Idl

[d

PAGE 14-227


References:

1. M. Kinoshita, et al., OHigh Bumup Rim Project (II) Irradiation and Examination to Investigate RimStructured Fuel," Proceedings of the 2000 International Topical Meeting on LWR Fuel Performance, ParkCity Utah, April 10-13, 2000.

2. J. Nakamura, et al., 'Thermal Diffusivity Measurements of a High-Bumup U02 Pellet," Proceedings of theANS International Topical Meeting on LWR Fuel Performance, Portland Oregon, March 2-6,1997.

3. S. Yagnik, 'Thermal Conductivity Recovery Phenomenon in Irradiated U02 and (U,Gd) 02,' Proceedingsof the 2000 International Topical Meeting on LWR Fuel Performance, Park City Utah, April 10-13, 2000.

4. M. Uppens and L. Mertens, 'High Bumup U02 and (U,Gd)0 2 Specimens: Thermal DiffusivityMeasurements and Post-Irradiation Characterization,' Final Report, EPRI TR-106501, December 1996.

5. K Une, et al., 'Effects of Grain Size and PCI Restraint on the Rim Structure of U02 Fuels,' Proceedingsof the 2000 International Topical Meeting on LWR Fuel Performance, Park City Utah, April 10-13, 2000.

6. [cJ

7. Not used.

8. DA. Wesley, et al., *Mark-BEB Ramp Testing Program,' Proceedings of the 1994 International TopicalMeeting on LWR Fuel Performance, West Palm Beach, Fl., April 17-21, 1994.

9. l.H. Shames, OMechanics of Deformable Solids,' Prentice-Hall, Inc., 1965, p 21.

10. J.G. Merkle, 'Engineering Approach to Multiaxial Plasticity," ORNL-4138, Oak Ridge NationalLaboratories.

11. R. Hill, 'Yielding and Plastic Flow of Anisotropic Metals,' Proceedings Ray Society, A193, 1948, pp 281-295.

12. A.F.J. Eckert, H.W. Wilson, and CE. Yoon, 'Program to Determine In-Reactor Performance of B&WFuels -Cladding Creep Collapse', BAW-10084P-A, Rev. 3, October 1980.

13. A.P. Boresl, et. al., 'Advanced Mechanics of Materials,' John Wiley and Sons, 1978, pp 471-483.

14. J.T. Wilse and G.L. Gamer, 'Recent Results from the Fuel Performance Improvement Program atFramatome Cogema Fuels,' Proceedings of the 2000 International Topical Meeting on LWR FuelPerformance, Park City Utah, April 10-13,2000.

15. [ci

16. R. Van Nieuwenhove and E. Kolstad, 'In-Pile Determination of Thermal Conductivity of Oxide Layer onLWR Cladding, EPRI TR-107718-P2, Final Report October 1998.

17. I[c

18. W.A. Lambertson and M.H. Mueller, "Uranium Phase Equilibrium Systems: 1, U02 - A6037, Journal of theAmerican Ceramic Society, 36, p.329-331, 1953.

19. L.G. Wisnlyi and S.W. Pijanowski, 'The Thermal Stability of Uranium Dioxide", KAPL - 1702, UC-25,Metallurgy and Ceramics, November 1957.

20. T.C. Ehlert and J.L. Margrave, 'Melting Point and Spectral Emissivity of Uranium Dioxidew, Journal of theAmerican Ceramic Society, 41, p. 330, 1958.

PAGE 14-228

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-10231COPERNIC FUEL Roo DESIGN COMPUTER CODE BAW-1 0231

21. J.A. Christensen, 'Irradiation Effects on Uranium Dioxide Melting', HW-89234, UC-25, Metals, Ceramicsand Materials, Hanford Laboratories, March 1962.

22. JA. Christensen, 'Thermal Expansion and Change In Volume of Uranium Dioxide on Melting', Journal ofthe American Ceramic Society, 46, p. 607-608, December 1963.

23. T.J. Pashos, D.R. Dehalas, D.L Keller and LA. Neimark, *Irradiatlon Behavior of Ceramic Fuel",Proceedings of the Third United Nations Intemational Conference on the Peaceful Uses of the AtomicEnergy, Geneva 1964, Vol.11, p. 472-484, United Nations, Geneva, 1965.

24. H. Hausner, eDetermination of the Melting Point of Uranium Dioxideo, Journal of Nuclear Materials, 15,179-183, 1965.

25. M.J. Bannister, 'Melting Temperatures in the System Uranium-Uranium Dioxide', Journal of NuclearMaterials, 24, p. 221-251, 1970.

26. R. Benz, G. Balog, and B.H. Baca, 'U - U02 - UN2 Phase Diagram', High Temperature Science, 2, p.221-251, 1970.

27. B.F. Rubin, 'Summary of (U,Pu)02 Properties and Fabrication Methods', GEAP 13582, AEC Researchand Development Report, November 1970.

2. T. Tachibana, T. Ohmorl, S. Yamamouchi and T. Itaki, 'Determination of Melting Point of Mixed-OxideFuel Irradiated in Fast Breeder Reactor', Journal of Nuclear Science and Technology, 22, p. 155-157,February 1985.

29. A. Chotard, P. Melin, M. Bruet and B. Frangois, 'Out of Pile Physical Properties and in Pile ThermalConductivity of (U,Gd)0?, Properties of Materials for Water Reactor Fuel Elements and Method ofMeasurement, IAEA, Vienna, 13-16 October 1986, IWGFPT/26, p. 7747, 1987.

30. J.A. Christensen, R.J. Ailio and A. Bianchera, 'Melting Point of Irradiated Uranium Dioxide, Trans. Am.Nucl. Soc., 7, p. 390-391, 1964.

31. S. Yamamouchi, T. Tachibana, K Taukul and M. Oguma, 'Melting Temperature of Irradiated U02-2%Gd203 Pellets up to Burnup of about 30 GWd/tU', Joumal of Nuclear Science and Technology, 25, p. 528,1988.

32. RA Bush, 'Properties of the Urania-Gadolinia System (Part 1)', EPRI NP-7561-D, project X 101-3, Finalreport, January 1992.

33. K Watarumi, 'Determination of Physical Properties of (U,Gd)02 Compounds', Nuclear Fuel IndustriesLtd., NF-DK-1286, July 5, 1991.

34. BAW-1 01 22A, Rev. 1, 'Normal Operating Controls," G. E. Hanson, May 1984.

35. BAW-10163P-A, 'Core Operating Limit Methodology for Westinghouse Designed PWRs,' B. J. Delano,et al., June 1989.

36. S. Bourreau, et al, 'Ramp testing of PWR fuel and multi-dimensional finite element modeling of PCMI,'paper presented at the ANS 2000 International Topical Meeting on LWR Fuel Performance," Park City,Utah, April 10, 2000.

PAGE 14-229

UNITED STATESNUCLEAR REGULATORY COMMISSION

C WASKINGTON, O.C. 205S5.0001

N.^ l~ay 14, 2001

Mr. T. A. Coleman, Vice PresidentGovernment RelationsFramatome ANP3315 Old Forest RoadP. 0. Box 10935Lynchburg, Virginia 24506-0935

SUBJECT: REQUEST FOR ADDIONAL INFORMATION - CHAPTER 13 OFFRAMATOME TOPICAL REPORT BAW-10231P (TAC NO. MA9783)

Dear Mr. Coleman:

By letter dated July 31, 2000, Framatome requested a review of Topical Report BAW-10231P,'COPERNIC Fuel Rod Design Code.* The staff has determined that additional information forChapter 13, MOX Applications, is required in order to complete our review.

The enclosed questions have been discussed with your staff. As discussed with your staff, byJune 30, 2001, please respond to the uranla-related questions and provide a schedule forresponding to the MOX-related questions. If you have any questions concerning our review,please contact me at (301) 415-1321.

Sincerely,

Stewart Bailey, Per, Section 2Project Directorate IIlDivision of Licensing Project ManagementOffice of Nuclear Reactor Regulation

Project No. 693

Enclosure: Request for Additional Information

cc wlenc: See next page

Mr. T. A. Coleman Project No. 693

cc:

Mr. James MallayDirector, Regulatory AffairsFramatome ANP2101 Horn Rapids RoadRichtand, WA 99352

Mr. F. McPhatter. ManagerFramatome ANP3315 Old Forest RoadP.O. Box 10935Lynchburg. VA 24506.0935

Mr. R. SchomakerFramatome ANP3315 Old Forest RoadP.O. Box 10935Lynchburg. VA 24506-0935

Mr. Michael SchopprnanFramatorne ANP1911 N. Ft Myer DiveRosslyn, VA 22209

REQUEST FOR ADDITIONAL INFORMATION

TOPICAL REPORT BAW-10231P. CHAPTER 13

"COPERNIC FUEL ROD DESIGN CODE"

MOX APPLICATIONS

The questions provided below address COPERNIC evaluations related to normal operation. Asecond round of questions related to mixed oxide (MOX) fuel application to transient andaccident analyses wMll be issued separately.

1. It is recoginized that weapons grade plutonium will be used for MOX for commercialapplication in the U.S. However, the isotopic plutonium ratios are significantly differentbetween reactor grade (reprocessed LWR fuel) plutonium and weapons gradeplutonium. Please provide the plutonium ratios for reactor grade and weapons gradeplutonium and; also, the tabular values of pellet radial power profiles to be used forweapons grade plutonium and how these values were determined, If the reactor gradeand weapons grade MOX radial profiles are proposed to be similar, provide thecalculational results for both MOX types that demonstrate this conclusion.

2. Please provide the specifications (including nominal values) of oxygento-metal (O)M)ratio, PuO2 particle size, and grain size specified for the U.S. commercial application.

3. For the experimental thermal MOX data, what were the ONM ratios used for codeverification?

4. For the MOX fission gas release data, please provide the nominal and, range of PuO2particle size for the different experimental rods used for code verification?

5. The conductivity equation for unirradiated MOX (Eq. 4-44) defines te term, y, as Pucontent In weight-percent. but It appears that this may be weight fraction. Please verifywhich unit is intended. If the Pu content is in weight fraction, the correction for Puconductivity is small for 100 wI% PuO2 which appears to be too low (see questions 6and 8 below).

6. The Halden Reactor Project correction for unirradiated MDX Is an 8 percent reduction inthe Urania thermal conductivity (at all temperatures) when the Pu concentration Is equalto or below 12 WIS. This Is significantly higher than the value used In COPERNIC.Also, the COPERNIC model for urania and MOX pellet thermal conductivity at highbumups and nominal stoichiometry (x = 0.02) Is significantly higher in the range from600 to 1500 K than similar burnup-dependent models, such as those proposed byORNLiKurchatov (Popov, 2000), Halden (Wiesenack, 2000, HPR-589) and Baron(Baron 1 998). Please justify the higher thermal conductivity values used by COPERNICfor unirradiated and high bunup MOX (see question 8 below).

-2 -

7. Recent high-temperature data on unirradiated urania fuel pellet thermal conductivity(Ronchl et al., 1999) has indicated that the conductivity in the range from 2000 to3000 K Is significantly lower than the COPERNIC equations for urania and, byimplication, for LWR MOX also. Most of the current conductivity models (includingCOPERNIC) are based on very old data at high temperature from which there was Iconsiderable scatter. The more recent data appears to have less scatter and betterexperimental techniques to minimize the scatter due to heat loss and other effects.Please justify the higher estimates of COPERNIC conductivity In this high temperaturerange because the discrepancy affects the LHGR margin to center fuel melting.

8. The Integral MOX experiments provided, where centerline temperatures are measured,to verify the COPERNIC Integral thermal predictions of MOX fuel rods are limited to verylow burnup levels, I.e., less than 5 GWd/MTU. Please provide COPERNIC predictionsof at least three of the following Halden MOX instrumented assemblies, IFA-597.41.5/.6,IFA-606, IFA-610, and IFA-648.1, that achieved bumrups of approximately 24 GWdIMTMto 57 GWdWMTM, or suggest other Halden MOX instrumented assemblies. Pleasejustify the reasons for eliminating some of the data andlor assemblies for COPERNICcomparisons and the reasons for selecting others (this should be discussed with theNRC reviewer prior to Issuing a response to the request for additional information).Also, rod pressures due to fission gas release were measured for two experimentalHalden MOX fuel rods In IFA-597.41.5/.6. COPERNIC predictions of rod pressure arealso needed, where appropriate.

9. What are the gas production values (xenon, krypton and helium) used in COPERNIC forMOX. Justify their application to weapons grade Pu. Also, how are the release 4fractions for helium determined In the rod pressure analysis, LOCA analyses, and otheranalyses where it Is important?

10. Has Framatome (or other parties) examined the interface between MOX fuel and thecladding at high burnups to determine if there are any chemical reactions (such asZr-oxide formation or other reactions) between the fuel and cladding?

REFERENCES

Baron, D., 1998. "About the Modeling of the Fuel Thermal Conductivity Degradation at HighBumup, Accounting for Recovery Processes with Temperature" paper 2.5 In Proceedings of theNEA Seminar on Thermal Performance of High-Bumup LWR Fuel, Cadarache, France,3-6 March 1998.

Popov, S.G., et al., 2000. Therrnophysical Properties of MOX and U0, Fuels Including theEffects of Irradiation, ORNL/TM-2000/351, Oak Ridge National Laboratory, Oak Ridge, TN.

Ronchi, C., M. Shiendlin, M. Musella, and G.J. Hyland, I 999. 'Thermal Conductivity ofUranium Dioxide up to 2900 K from Simultaneous Measurement of the Heat Capacity andThermal Diffusivity," Journal of Applied Physics, Vol. 85 no. 2, pages 776 to 789.

-3-

W. Wiesenack and T. Tverberg, 2000. "Thermal Performance of High Bumup Fuel - In-PileTemperature Data And Analysis," In Proceedings of the ANS Intemational Topical Meeting onLight Water Reactor Fuel Performance, Park City, Utah, April 2000, pages 730 to 737.(As-modified for MOX by IHWR-589)

J 1FRAIMATOME ANP

July 27, 2001NRC:01 :033

Document Control DeskATTN: Chief, Planning, Program and Management Support BranchU.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001

Partial Response to RAI

Ref.: 1. Letter, Stewart Bailey (NRC) to T. A. Coleman, (Framatome Cogema Fuels),"Request forAdditional Information - Framatome Topical Report BAW-10231 P(TAC NO. MA6792)," August 11, 2000.

Ref.: 2. Letter, T. A. Coleman (Framatome ANP) to U.S. Nuclear Regulatory Commission,GRO-021.doc, February 5,2001.

Ref.: 3. Letter, Stewart Baliey (NRC) to T. A. Coleman (Framatome ANP), Request forAdditional Information - Chapter 13 of Framatome Topical Report SAW-1 0231 P(TAC NO. MA9783)," May 14, 2001.

Reference 1 provided a request for additional Information (RAI) on Framatome CogemaFuels (FCF) topical report SAW-10231P, "COPERNIC Fuel Rod Design Code." That RAIaddressed the U0 2 applications of the code. Reference 2 contained the Framatome ANPresponse to the RAI.

Reference 3 Is the RAI associated with the M4OX applications for COPERNIC. However,two of the questions In the RAI (numbers 6 and 7) refer to U0 2 applications. Theresponses to these two questions are enclosed. Framatome plans to apply the U0 2 portionof the COPERNIC topical report within the next few months. Therefore, receipt of the SERby August 31, 2001 constitutes a critical need in our schedule.

In a telephone conference with the NRC held on May 23, 2001, Framatome ANP agreed torevise some of the figures that had been included In the response to question 2 of the InitialRAI. The revised figures are enclosed. Since the topical report was printed double-sided,the enclosed change pages typically have revisions on one side only.

Power level hold time for LOCA initiarmation is discussed in Chapter 12, "ApplicationMethodology," of BAW-10231P. In orderto clarify the manner in which the hold time isdetermined, Framatome ANP has developed a method for linking the hold time to therequirements in the plant-specific technical specifications. A description of this method is

Framatome ANP Richland, Inc.

2101 Hom Rapids Road TeL (509) 375-D100Rlbamd, WA 59352 Fax: (509) 3754402

I

Document Control Desk NRC:01:033July 27,2001 Page 2

enclosed as a clarification to facilitate the NRC's review and should be included in yourevaluation of SAW- 0231 P.

In accordance wbth the provisions of 10 CFR 2.790(b), Framatome ANP requests that theseresponses be considered proprietary and withheld from public disclosure. Attachment I Isan affidavit supporting this request. Attachment 2 is the proprietary version of the RAIresponses, revised figures, and unsolicited response. Attachment 3 is the non-proprietaryversion. After the SER Is received, Framatome ANP will Incorporate all the enclosedmaterial Into ether the body or an appendix of the approved version of BAW-10231 P.

Very truly yours,

James F. MlaDrectorRegulatory Affairs

cal

Enclosures

cc: S. N. Bailey, NRCR. Caruso, NRCJ. S. Wermeil, NRCS. L. Wu, NRCC. E. Beyer, PNNLM. S. Schoppman20A13 FilelRecords Management

4

AFFIDAVIT

STATE OF WASHINGTON )) Ss.

COUNTY OF BENTON )

1. My name Is James F. Mallay. I am Director, Regulatory Affairs, for

Framatome ANP CFRA-ANP"), and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by FRA-ANP to determine whether.

certain FRA-ANP Information Is proprietary. I am familiar with the policies established by

FRA-ANP to ensure the proper application of these criteria.

3. 1 am familiar with the FRA-ANP Information included In two of the attachments

(responses to RAI and change pages) to letter NRC:01:033, dated July 27, 2001 from James F.

Mallay to (NRC). These two attachments are referred to herein as "Documents." Information

contained in these Documents has been classified by FRA-ANP as proprietary In accordance

with the policies established by FRA-ANP for the. control and protection of proprietary and

confidential information.

4. These Documents contain Information of a proprietary and confidential nature

and Is of the type customarily held In confidence by FRA-ANP and not made available to the

public. Based on my experience, I am aware that other companies regard Information of the

kind contained In these Documents as proprietary and confidential.

5. These Documents have been made available to the U.S. Nuclear Regulatory

Commission in confidence with the request that the information contained In these Documents

be withheld from public disclosure.

6. The following criteria are customarily applied by FRA-ANP to determine

whether Inforrmation should be classified as proprietary:

(a) The information reveals details of FRA-ANP's research and development

plans and programs or their results.

(b) Use of the Information by a competitor would permit the competitor to

significantly reduce its expenditures, in time or resources, to design, produce,

or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a

process, methodology, or component, the application of which results in a

competitive advantage for FRA-ANP.

(d) The Information reveals certain distinguishing aspects of a process,

methodology, or component, the exclusive use of which provides a

competitive advantage for FRA-ANP in product optimization or marketability.

(e) The information Is Vital to a competitive advantage held by FRA-ANP, would

be helpful to competitors to FRA-ANP, and would likely cause substantial

harm to the competitive position of FRA-ANP.

7. In accordance with FRA-ANP's policies governing the protection and control

of information, proprietary Information contained In these Documents has been made available,

on a limited basis, to others outside FRA-ANP only as required and under suitable agreement

providing for nondisclosure and limited use of the Information.

8. FRA-ANP policy requires that proprietary information be kept in a secured file

or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge,

information, and belief.

SUBSCRIBED before me this QI

day of .2001.

6''' - - ' . ' --

'�-l 1.0" &kkLValerie W. SmithNOTARY PUBLIC, STATE OF WASHINGTONMY COMMISSION EXPIRES: 10110/04

6. The Halden Reactor Project correction for unirradiated MOX is an 8 percentreduction in the urania thermal conductivity (at all temperatures) when the Puconcentration is equal to or below 12 wt%. This is significantly higher than thevalue used in COPERNIC. Also, the COPERNIC model for urania and MOX pelletthermal conductivity at high burnups and nominal stoichiometry (x = 0.02) issignificantly higher in the range from 500 to 1500 K than similar burnup-dependentmodels, such as those proposed by ORNIJKurchatov (Popov, 2000), Halden(Wiesenack, 2000, HPR-589) and Baron (Baron 1998). Please justify the higherthermal conductivity values used by COPERNIC for unirradiated and high burnupMOX (see question 8 below).

The COPERNIC U0 2 fuel thermal conductivity relationship [b, c]

The COPERNIC thermal conductivity relationship [b, cl

the Baron(6) relationship which has been shown to conservatively over-predict measured fuel temperaturesO.

The integrals of the above thermal conductivities are shown in Figures 1, 2 and 3 forMOX and U0 2. [b, c]

APPENDIX A

lb, c]

REFERENCES

[I] P.G. Lucuta, Hj. Matzke, I.J. Hastings, "A Pragmatic Approach to ModelingThermal Conductivity of Irradiated U02 Fuel: Review and Recommendations,' J.Nucl. Mater., Vol. 232, pp. 166-180(1996).

[2] W. Wiesenack, "Assessment of U0 2 Conductivity Degradation Based on In-pileTemperature Data," Proceedings of the International Topical Meeting on LightWater Reactor Fuel Performance, Portland, U.S.A., pp. 507-511 (1997).

[3] S. Yagnik, 'Thermal Conductivity Recovery Phenomena in Irradiated U02 and(U, Gd)02, Proceedings of the International Topical Meeting on Light WaterReactor Fuel Performance, Park City, Utah U.S.A., pp. 634-645 (2000).

[4] C. Duriez, J.P. Alessandri, T. Gervais, Y. Philipponneau, "Thermal Conductivityof Hypostoichiometric Low Pu Content (U,Pu)O2., Mixed Oxide" J.Nucl.Mater., Vol. 277 pp. 143-158 (2000)

[5] S.G. Popov, J.J Carbajo, V.K. Ivanov, G.L. Yoder, uThermophysical Propertiesof MOX and U02 Fuels Including the Effects of Irradiation" Report ORNL/TM-20001351

[6] D. Baron, "About the Modeling of the Fuel Thermal Conductivity Degradation atHigh Burnup, Accounting for Recovery Processes with Temperature," Paper 2.5in proceedings of the NEA Seminar on Thermal Performance of High-BurnupLWR Fuel, Cadarache, France, 3-6 March 1998

[7] D. D. Lanning, C. E. Beyer, "Assessment of Recent Data and Correlations forFuel Pellet Thermal Conductivity," Presented at the 2001 Enlarged HaldenProgram Group Meeting, Lillehammer, Norway 1-16 March 2001.Published in Halden Report HRP-356

[8] D. D. Lanning, C. E. Beyer, M. E. Cunningham, " FRAPCON-3 Fuel rodtemperature Predictions with fuel conductivity degradation caused by fissionproducts and gadolinia additions," Proceedings of the International TopicalMeeting on Light Water Reactor Fuel Performance, Park City, Utah U.S.A., pp.244-258 (2000).

Table 1: FRAMATOME-ANP database for the validation of the MOX fuelthermal conductivity relationship of COPERNIC

[b, c, dl

Figure 1 - MOX Thermal Conductivity Integrals

lb, c]

Figure 2 - U02 Thermal Conductivity Integrals

lb, c]

Figure 3 - U02 Thermal Conductivity Integrals

[b, cl

7. Recent high-temperature data on unirradiated urania fuel pellet thermal conductivity(Ronchi et al., 1999) has indicated that the conductivity In the range from 2000 to 3000K is significantly lower than the COPERNIC equations for uraniA and, by implication,for LWR MOX also. Most of the current conductivity models (including COPERNIC)are based on very old data at high temperature from which there was considerablescatter. The more recent data appears to have less scatter and better experimentaltechniques to minimize the scatter due to heat loss and other effects. Please justify thehigher estimates of COPERNIC conductivity in this high temperature range because thediscrepancy affects the LHGR margin to center fuel melting.

lb, c, e]


Figure 14-3: COPERNIC and NFIR-II Thermal Conductivity Comparison60 GWdItU Bumup - [cJ

(b, c, dl

Figure 14-4: COPERNIC and JAERI Thermal Conductivity Comparison - Sample No.263 GWdItU Burnup, 83489% Density Range, [c]

[b C d]

PAGE 14T

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW40231-

Figure 14-5: COPERNIC and JAERI Thermal Conductivity Comparison -Sample No.363 GWdttU Burnup, 92-96% Density Range, [q]

b c, d]

Figure 14-6: Rim Effect at 60 GWdtUIFA 662

(b, c, d]

PAGE 14-6

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAVW-10231

Question 6

Is Framatome a member of Halden? If so, Halden has refabricated two high (5 59 GWdIMTU) bumup rods(one with a functional thermocouple) and placed them first In IFA-597.2 (HWR-442) and subsequently in IFA-597.3 (HWR-543) with measured centerline temperatures. Please compare COPERNIC code predictions tothis data and Include this data In the response to Question 3.1 above.

Response

The COPERNIC centerline fuel temperature predictions are compared with the IFA-597.2 (HWR-442) andIFA-597.3 (HWR-543) fuel temperature measurements in FIgure 14-13. This rodlet attained a bumup of 61.5GWd/l:O 2 or 69.8 GWd/tU.

PAGE 14-22

COPERNIlC FUEL ROD DESIGN COMPUTER CODE BAW-10231

Figure 14-13: Fuel Centerline Temperature Measurements and Predictions vs.Burnup, IFA-597.2 and IFA-597.3

[d]

Figure 14-14: Not used

(d]

PAGE 14-23

- - -v

* . . IUUehKNIU FUEL ROD DESIGN COMPUTER CODE BAW-10231

Question 28

There is a concern that the uncertainty factor provided in Equation 12-1 may be too small at the predictedoperating temperatures (stored energy) calculated for LOCA Initialization. Please discuss this issue further,particularly hI relation to Question 3 above.

Response

[b, d, el.

PAGE 14-67

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAWDY10231

Question 29

Are any of the example calculations provided In Section 12 for fuel cores with two 24-month cycles? Itappears that there are no 24-month cycle results presented for the Mark BW-17 design. If so, please explainbecause it Is anticipated that a large number of plants will be switching to 24-month cycles In the next fewyears.

Response

[bd].

PAGE 14-68

[, .t;_j:- ^:. , - ; . ' ::. '-,t \. k-. -IZ ;"I . ;; 7 " ' .O {.:-. t..'W- t. .. .

II4.

Clarification of Power Level Hold Time for'LOCA InitialIzation

- Section 12 (Application Methodology) of BAW-10231P describes the methodology that will beused to predict the inifialtzation conditions of fuel rods for reload safety evaluations. Section12.2.1 specifies the methodology that will be used to generate the LOCA Initialization predictions.The fuel rod Is simulated to operate with a specified limiting rod power history and Is then rampedto the LOCA Fa limit at the time In life when the LOCA transient originates. The topical states[b, c

Plant technical specifications place limits on key controlled measurable parameters'such ascontrol rod Insertion, axial power imbalance, and thermal power level to ensure that the Initialconditions for accidents are maintained during operation. The limits define boundaries of coreoperation where power peaking factors could equal the LOCA Fa limit (or the maximum allowablepeakfng limit f'or another accldent If Wtis more limffing than the LOCA). Should one of the controlpara;nieter feach or exceed E fls nitM required actibns and competion times are specified by thetechnical specificatiors. The required actidns and completion times ensure that power peakingfacdo's are restored within their Umlfs promptly. These technical specification limits, together withtieir corresponding actions and completion times, limit the amount of time that the fuel couldoperate with power peaking factors ih excess of the specified acceptable fuel design limits.

Allowable completion times for these tech spec required actions typically fall In the range of 15minutes to 4 hours. Plants that operate with fixed Incore detector systems have the additionaloption of generating an Incore flux map at regular intervals (typically 2 hours) to provide a directcheck on the power peaking factors; this provides assurance that both the FQ and F44 peakingfactors are verified to remain within their technical specification limits.

Based upon the protection provided by the technical specifications, the amount of time that thefuel could operate at LOCA transient initialization conditions Is typically limited to a range of 15mInutes to 4 hours, depending upon the Individual plant tech spec requirements. [b. cl

The primary protection for theLOCA Fa peaking factor is afforded by the axial power imbalance (or axial flux difference) limits,[b, c

..

l

YIFRAMATOMEAN P

February 5, 2002NRC:02:010

Document Control DeskATTN: Chief, Planning, Program and Management Support BranchU.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001

Response to Informal Request on BAW-10231P, uCOPERNIC Fuel Rod Design Code"

Ref.: 1. Letter, Stewart Bailey (NRC) to T. A. Coleman (Framatome ANP), ORequest forAdditional Information - Framatome Topical Report BAW-10231P (TAC NO.MA6792), August 1, 2000.

Ref.: 2. Letter, Letter, T. A. Coleman (Framatome ANP) to U. S. Nuclear RegulatoryCommission, GRO-021.doc, February 5, 2001.

Ref.: 3. Letter, Stewart Bailey (NRC) to T. A. Coleman (Framatome ANP), "Request forAdditional Information - Chapter 13 of Framatome Topical Report SAW-1 0231P(TAC NO. MA9783), May 14, 2001.

Ref.: 4. Letter, James F. Mallay (Framatome ANP) to Document Control Desk (NRC),Partial Response to RAI,' NRC:01:033, July 27, 2001.

On several occasions the NRC has asked Framatome ANP to provide additional informationto assist in the NRC's review of BAW-10231P, *COPERNIC Fuel Rod Design Code.'Reference I transmitted a request for additional Information (RAI) on the U02 applications ofthis report Reference 2 contained Framatorne's response to that RAI. Reference 3provided an RAI that addressed primarily the MOX applications for COPERNIC. However,two of the questions in that RAI referred to U02 apprications. Reference 4 contained ourresponse to those two questions.

In addition to Information provided In References 2 and 4, Framatome ANP provided aninformal response to an NRC question concerning time In life for LOCA initialization.Attached to this letter Is a formal, referenceable response to that question.

Framatome ANP, Inc.

210i Hom Rapis RoadRlddand, WA tt352

TelW: (509) 375100Fax: (609) 375-40

Document Control Desk NRC:02:010February 5, 2002 Page 2

Based on discussions with the NRC, Framatome understands that this letter and attachinentwill provide an adequate basis for completing the SER for the application of BAW-10231P toUO2 .

Ve tr~ul yours,

James F. Mallay, DirectorRegulatory Affairs

Amk

Attachment

cc: J. S. CushingD. G. HollandProject 693

I

ATTACHMENT

Supplement lo RAI #25 on BAW-10231 P

Question:

Figures 12-21, 22, -23. -24, -25, and, -26 In BAW-10231P show a trend of increasingvolumetric average fuel temperatures with bumups for LOCA Initial conditions. This raisesa concern that the LOCA PCT may not be limiting in BOL FCF needs to address thetrend of Increasing fuel temperatures with bumups to allay this concern. Please evaluatethe LOCA PCT results for these figures as compared to the PCT limit of 2200 degree F.

Response:

The NRC-approved BWNT LOCA EM (BAW-10192P-A) and RSG LOCA EM (BAW-10168P-A) are the calculatlonal frameworks used to demonstrate compliance to the fivecriteria of iOCFRSO.46 for B&W-deslgned plants and Westnghouse- and CE-designedplants, respectively. The detailed methods used to show compliance are prescribed In 10CFR 50 Appendix K. Relative to fuel temperature trends, Appendix K Section IA.1 givesthe requirements for the initial stored energy in the fuel. It states:

'The steady-state temperature distribution and stored energy In the fuel before thehypothetical accident shal be calculated for the bun-up that yields the highestcalculated cladding temperature (or, optionally, the highest calculated storedenergy). To accomplish this, the thermal conductivity of the U02 shall beevaluated as a function of burn-up and temperature, taking into considerationdifferences in Initial density, and the thermal conductance of the gap between tleU0 2 and the cladding shall be evaluated as a function of the bum-up taking intoconsideration fuel densffication and expansion, the composition and pressure ofthe gases within the fuel rod, the initial cold gap dimension with its tolerances, andcladding creep."

An NRC-approved steady-state fuel code rike TACO3 (BAW-10162) or COPERNIC simplyprovides a set of inputs (namely fuel temperatures, hot and cold fuel pin dimensions, pingas pressure, and pin gas composition) that are derived-from the parameters listed In theAppendix K requirements. These inputs are then used In the analyses that showcompliance to 10 CFR 50.46 for the limits of fuel operation covering currently approvedburnups for Mark-B and Mark-SW fuel types.

LBLOCA analyses performed for the B&W-designed plants with BAW-10192P-A typicallycomplete five analyses at BOL, five analyses at a limiting MOL burnup (typically at 40, butranging from 20 to 55 GWd/mtU) and at least one analysis at the EOL condition based onInputs from the TAC03 code. The time-n-4lfe conditions at which the potential limitingPCT analyses are performed are strongly Influenced by the fuel stored energy and pinpressure predicted by the steady-state fuel code. After the COPERNIC code is approved,It can be used to generate steady-state LOCA initiarization inputs at several times In life to

provide fuel pin parameters that will determine how the PCT varies with bumup. Thismethod Is used to define an allowed linear heat rate limit versus burnup curve that willmaintain the calculated PCTs below the 2200 F limit for the entire fuel pin burnup range.

The LBLOCA analyses performed with BAW-10168P-A for the Westinghouse or CE plantsdefine a Up curve that Is used to restrict the total peaking as a function of time In life toensure that the PCT predicted by analyses performed near the beginning of life remainslimiting. The Kp curve Is set based on LBLOCA analyses performed at limiting fuel pinbumups that cover the licensed fuel pin bumup range. This bumup curve may need to beredefined when the specific NRC-approved steady-state fuel code used for the fuel reloadlicensing contract Is changed (i.e. TACO3 to COPERNIC).

Change Pages

Summary of Paees Replaced

Document Section Page(s)| ReasonTopical Report Table of v & vi Correct misspelling of Neodymium

ContentsTopical Report Table of xix Replaced to reflect addition of MOX section

ContentsTopical Report 4.3.3.1 4-16 Changed "weight %" to "weight fraction"Topical Report 4.5.4 4-20 Correct misspelling of NeodymiumTopical Report Figures 4-14 4-49 thru Correct misspelling of Neodymium in

thru 4-27 4-62 Figure titlesTopical Report 14 14-5, 14- Incorporate revisions specified in

6, 14-22, NRC:01 :033 (Partial Response to RAI)14-23, 14- dated July 27,2001

67NRC:03:027, "Final Attachment 2 1 Changed wording to "in Figures I throughResponses to RAIs 4B"on Chapter 13 of

BAW-1023 IP", dated4/18/03

NRC:03:027 Attachment 2 6 Changed Figure 4 to Figures 4A and 4B toI_ reflect separate data for each rod

* Note: Pages summarized in the above Table that were contained in the Topical Reportare the old pages from BAW-1 0231, Revision 0. The new replacement pages arecontained in BAW- 10231, Revision 1.

09

f C ̂ a",;) FCF Non Propretaryi 5 At rPAGEV

4.6.2. Validation of Mixed Oxide Fuels .4............................... -274.6.2.1. MOX ............................... 4-274.6.2.2. Uo2-Gd2o3 ............................... 4-27

4.6.3. Uncertainties ............................... 4-284.6.3. 1. FRAMATOME Database ............................... 4-284.6.3.2. US/NRC Halden Database ............................... 4-284.6.3.3. HALDEN PROJECT Database ............................... 4-284.6.3.4. Summary ............................... 4-29

REFERENCES ............................... 4-31

FIGURES ............................... 4-35FIGURE 4-1 ROUGHNESS FACTOR VERSUS DIMENSIONLESS GAP FOR DIFFERENT

-I

'U

0

I-

FIGURE 4-2FIGURE 4-3

FIGURE 4-4

FIGURE 4-5

FIGURE 4-6

FIGURE 4-7

FIGURE 4-8

FIGURE 4-9

FIGURE 4-1

FIGURE4-11

FIGURE 4-12


FIGURE 4-15

FIGURE 4-16

FIGURE 4-17

FIGURE 4-18

FIGURE 4-19

FIGURE 4-20

FIGURE 4-21

FIGURE 4-22

DIMENSIONLESS TEMPERATURE JUMP LENGTHS ..................................... 4-36COMPARISON OF MEASURED AND PREDICTED GASEOUS CONDUCTANCES . 4-37COMPARISON OF MEASURED AND PREDICTED GASEOUS CONDUCTANCES[b.] . 4-38COMPARISON OF MEASURED AND PREDICTED GASEOUS CONDUCTANCES[b.) 4-39COMPARISON OF MEASURED AND PREDICTED GASEOUS CONDUCTANCES[b.) 4-40COMPARISON OF MEASURED AND PREDICTED GASEOUS CONDUCTANCES[l.b .....-......-.......-... 4-41COMPARISON OF MEASURED AND PREDICTED GASEOUS CONDUCTANCES[b.] .... 4-42COMPARISON OF MEASURED AND PREDICTED GASEOUS CONDUCTANCESb.......... 4-43

COMPARISON OF MEASURED AND PREDICTED GAP CONDUCTANCES[b. .................................................... ,,.. 4-44COMPARISON OF MEASURED AND PREDICTED GAP CONDUCTANCES[b...4-45MEASURED AND PREDICTED GAP CONDUCTANCES vs PRESSURE[b.] .. .. 446MEASURED AND PREDICTED GAP CONDUCTANCES vs PRESSURE[b.] .. 4-47HEAT TRANSFER GAP MULTIPLIER X VS RADIAL MECHANICAL HOT GAP . 4-48COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES [b.] . .... ... 4-49COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES [b.) ............. 4-S0COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES fb.] ...... 4-51COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES b.) . . .4-52COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES [b.] ........... ., 4-53COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES [b.] ........... 4-54COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES fb]........................................................................................................................4-55COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES [b.) . .. ........... . . .4-56COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES lb.] .......... 4-57V

.':}:.. .z. -. .. A FCF Non Proprietary

vrm i4;, A J, . - 1 b eer PAGEVI

FIGURE 4-23

FIGURE 4-24

FIGURE 4-25

FIGURE 4-26

FIGURE 4-27

FIGURE 4-28


FIGURE 4-31

FIGURE 4-32

FIGURE 4-33FIGURE 4-34FIGURE 4-35

FIGURE 4-36

FIGURE 4-37

FIGURE 4-38

FIGURE 4-39

FIGURE 4-40

FIGURE 4-41

FIGURE 4-42

FIGURE 4-43

FIGURE 4-44

FIGURE 4-45

FIGURE 4-46

FIGURE 4-47

FIGURE 4-48

FIGURE 4-49

FIGURE 4-50

FIGURE 4-51

FIGURE 4-52

COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES [b.] ......... 58COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES .] .......... 4-59COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES b.]. ... ... ... 4-60COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES [b.] ........ 4-61COMPARISON OF MEASURED (NEODYNIUM) AND PREDICTED RADIAL POWERPROFILES - MOX (b. ................................................... 42MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP[b....4-63MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP [bj ................. 4-64MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP[b.] ................ 5MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUPlb.] 4-66MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUPb. ... . ...... 4-67

MEASURED AND PREDICTED FUEL TEMPERATURES VS TIME [b.] ........................ 4-68lei .................... .................................. . .4-69MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(FA 431-1, INT)............... .... 4-70MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 431-1. OUTLET) .. .......... . 4-71MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 431-2. INLET). ............ 4-72MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 431-2, OUTLET) ............ 4-73MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 431-3. INLET) ............. 4-74MEASURED AND PREDICD FUEL TEMPERATURES VS BURNUP(IFA 431-3, OUTLET) ............. 4-75MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 431-5, INLET) ............ _ 4-76MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 431-5, OUTLET) ............. 4-77MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 431-6, INI) .............. . . 4-78MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 431-6, OUTLET) ............. . ... .4-79

MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 432-1, INLET) . ... ........ _ . .4-80MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 432-1,OUTLET) ................. .4-81MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 432-2, NLET)......... .... 4-82MEASURED AND PREDICIED FUEL TEMPERATURES VS BURNUP(EFA 432-3, INLET) .. 4-83MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IPA 432-3, OUTLEI) ............. ... .4-84

MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 432-5, INLET) .............. _,,., 4-85MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP(IFA 432-6, INLET) . 4-86MEASURED AND PREDICTED FUEL TEMPERATURES VS BURNUP

(0-I

Lu

E

C ~APPENDIX A - Correspondence with the NRC, Including Re quests for Additional Informationand Responses

1. Letter from FCF (T.A. Coleman) to U.S. NRC, September 16, 1999.2. Letter from FCF (T.A. Coleman) to U.S. NRC, December 2,1999.3. Letter from U.S. NRC (Stewart Bailey) to FCF (T.A. Coleman), August 11, 2000.4. Letter from FCF (T.A. Coleman) to U.S. NRC, September 29, 2000.5. Letter from FRAMATOME ANP (T.A. Coleman) to U.S. NRC, February 5, 2001.6. Letter from U.S. NRC (Stewart Bailey) to FRAMATOME ANP (T.A. Coleman), May 14,2001.7. Letter from FRAMATOME ANP (J.F. Mallay) to U.S. NRC, February 5, 2002.

p(IL

us

0

0

rm COPEsNI FtCF Non Proprietarys C5Uehapter 4 PAGE 4-15

where

[b., e.],

[b., e.],[b., e-],

[b., e.], and

[b., e.I.

4.3.2.2. Porosity and RIM Model

The correction for porosity is a function of temperature as follows (Ref. 4-12):

Bu BlU;'POR = 1 (-a PR a = 2.58-5.8 104 -T Eq. 440

where

Tc : temperature (0C), and

POR: porosity fraction.

Fuel rims have been observed on fuel pellet peripheries when the local rim burnup exceeds a valueof approximately [e.] or when the average pellet burnup exceeds a value of approximately (e.]. Therim porosity has also been observed [e.]. The following rim model was selected based upon theseobservations:

Uiw0I-

[b., .j Eq. 4-41

where

Bs : [b., e-j,

BA : average pellet bumup,

C3 : [b., e.], and

C4 : lb., e].

and

[b., d.] Eq. 442

- . , r rFCF Non Proprietary

Chapter 4 PAGE 4-16

where

POR:

r

rs

C5:

porosity fraction,

[b.,e.]

[b., e.], and

[b., C.J.

Eq. 4-43[b., e.]

4.3.3. Adaptation to Mfixed Fuels

The COPERNIC U0 2 thermal conductivity relationship has been adapted to mixed oxide andgadolinia fuels with the following modifications.

4.3.3. 1. MOX

[b.,d.] Eq. 4-44

e

8

w

where

xF(y)=

F(Y) =

O/M :y

2.00 - O/M,

lb., d.],

[b., d.],

oxygen to metal ratio, and

Pu content (weight%).

4.3.3.2. U02 -Gd2O3

The ratio of the thermal conductivity of gadolinia bearing fuel to U0 2 fuel [r (zT) = 7(UGd)U0 2

J L(UO2)] is:

[b., d.] Eq 44 5

| m C0f E.,-RNid C FCF Non Proprietary

F 45. Fuel Pellet Radial Power Profiles

4.5.1. Phenomenon

The thermal power produced within the fuel corresponds to the volumetric distribution of fissions.The latter depends on the initial isotopic distribution within the fuel and the irradiation conditions(temperature, power, environment, ...). These conditions establish the radial distribution for boththe neutron flux and the resulting isotopic composition. The primary phenomena involved are:

- attenuation of the neutron flux towards the pellet center due to self-shielding and capture,

- fissile atom depletion within the pellet,

- the progressive enrichment of a fine Pu peripheral layer through captures in U238,

- burnout of a burnable absorber like gadolinium.

4.5.2. Model

The radial power distributions within the pellet were determined separately with the APOLL02neutronic transport code (Ref. 4-15). These predictions produced tabulated data which wereI incorporated in COPERDC.

Sensitivity studies were run to define the calculational basis for U0 2 fuel. These are:

- an infinite medium cell layout,

E - a uniform temperature for all isotopes (except U238),

o - a radial temperature profile for U238, andwo - a self-shielding calculation at BOL.0

The radial mesh must be sufficiently fine towards the pellet outer edge to correctly represent the

Ef peripheral rim effect. Better accuracy at high burnups is obtained with 20 rings.

4.5.3. Available Options

4.5.3.1. Generic Tables for U0 2

The generic tables for U0 2 were established for:

[d., e-]

- ~ :..1-~.FCF Non Proprietary

. R Chapter 4 PAGE 4-20

4.5.3.2. Tables for MOX

MOX generic tables were established for:

[d., e.]

4.5.3.3. Tables for U02-Gd2 O3

[d., e.]

4.S.4. Experimental Validation

Wo-I

4:

1u

w

The radial bumnup distributions obtained with the APOLL02 tables were validated with radialNeodynium profiles which were obtained with Electron Probe Micro Analyses (EPMA). The rodsused are given in Table 4-3.

The validation of the generic U0 2 tables was performed with a number of samples from PWR rodsthat had two initial enrichments and a variety of burnups (Figures 4-14 to 4-26). Good agreementbetween measurements and predictions is observed in these figures.

The measured to predicted comparison for the MOX generic tables was performed [e.]. Noallowance is made for the scatter in the radial power due to the presence of Pu-rich spots and thefuel is assumed to be homogeneous in the calculations. Therefore, the measured to predictedcomparison has a wide scatter, however, the temperature benchmark results for MOX fuel aresatisfactory as shown later in the last section describing the global experimental validation. Notethat the periphery effect for MOX is much smaller than U0 2 at a comparable burnup (Figure 4-27).

4.5.5. Range of Application

- U02 generic tables: U235 enrichment [d., e.j

Bumup [d., e.j

Pu content [d., e.]- MOX generic tables:

Burnup [d., c.]

- U 2O-d 2 03 specific tables: W2A3 content [d., e.

User specified input

- COPEjR>N-1 Ai,.FCF Non Propretary~'Chapter 4 PAGE 4-49

FIGURE 4-14 COMPARISON OF MEASURED (NEODYNIUM) AND

PREDICTED RADIAL POWER PROFILES (b.j

[b.]U,

0

0

ft -. ; ;|FCF Non Proprietary

; Chapter 4 PAGE 4-50


PREDICTED RADIAL POWER PROFILES [b.]

0

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[b.]

Jr COPE N)' Em, c FCF Non Proprietary^ -. Chapter 4 PAGE 4-51

I FIGURE 4-16 COMPARISON OF MEASURED (NEODYNIUM) AND


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PREDICTED RADIAL POWER PROFILES [b.A

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Chapter 4 PAGE 4-54

FIGURE 4.19 COMPARISON OF MEASURED (NEODYNIUM) AND


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PREDICTED RADIAL POWER PROFILES [A.]

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FIGURE 4-21 COMPARISON OF MEASURED (NEODYNIUM AND


a,

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I FIGURE 4-22 COMPARISON OF MEASURED (NEODYNIUM) AND


IColJ

z [b.]

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i:.Lir i 3f Chapter 4 PAGE 4-58


PREDICTED RADIAL POWER PROFILES Db.]

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PREDICTED RADAL POWER PROFILES Ab.]

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PREDICTED RADIAL POWER PROFILES - MOX [b.]

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COPERNIC FUEL Roo DESIGN COMPUTER CODE BAW 10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 D23 I

Figure 14-3: COPERNIC and NFIR-IIl Thermal Conductivity Comparison100% Theoretically Dense Fuel

60 GWdftU Bumup

lb, c, dc

Figure 14.4: COPERNIC and JAERI Thermal Conductivity ComparisonSample No.2

100% Theoretically Dense Fuel63 GWdNtU Burnup, 63-89% Initial Density Range

[be C, d]

PAGE 14-5


Figure 14-5: COPERNIC and JAERI Thermal Conductivity ComparisonSample No.3

100% Theoretically Dense Fuel63 GWd/tU Burnup, 92-96% Initial Density Range

[b,c, d

Figure 14-6: Rim Effect at 60 GWd/tUIFA 562

(b. c, d

PAGE 14-6


Figure 14-12: Measured and Predicted Fuel Temperatures vs. Bumup

PAGE 14-21

COPERNIC FUEL Ron DESIGN COMPUTER COOE BAW-1023iCOPERNIC FULRDDSG OPUE OEBW103

Question 6

Is Framatome a member of Halden? If so, Halden has refabricated two high (- 59 GWd/MTU) bumup rods(one with a functional thermocouple) and placed them first in IFA-597.2 (HWR-442) and subsequently In IFA-597.3 (HWR-543) with measured centerline temperatures. Please compare COPERNIC code predictions tothis data and Include this data in the response to Question 3.1 above.

Response

The COPERNIC centerine fuel temperature predictions are compared with the IFA-597.2 (HWR-442) andIFA-597.3 (HWR-543) fuel temperature measurements In Figures 14-13 and 14-14, respectively. This rodletattained a bumup of 61.5 GWd.tUO 2 or 69.8 GWdtU.

PAGE 14-22

COPERN IC FUEL ROD Dow COMPUTER CODE BAW-10231

Figure 14-13: Fuel Centerline Temperature Measurements and Predictions vs.Burnup, IFA-597.2

[d]

Figure 14414: Fuel Centerline Temperature Measurements and Predictions vs.Burnup, IFA-597.3

[d1

PAGE 14-23

COPERNIC FUEL ROD DESIGN COMPUTER CODE --- -BAW-10231COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231

Question 7

The athermal fission gas release model (Section 5.2.2) is dependent on open porosity but no values areprovided for what Is used for Framatome fuel. What values are used for open porosity? If more than onevalue is used, please provide the value for each fabrication process.

Response

The open porosity input to the COPERNIC code Is the percentage of open porosity to the total pelletgeometric volume. The open porosity percentage of the fuel supplied by FRA-ANP's present vendor istypically gb, di. The lb. d] will be used until open porosity data obtained from the fuel vendor suggests a needto Increase this value to [b, d] pellet fabrication open porosity measurements.

PAGE 14-24


There is a concern that the uncertainty factor provided In Equation 12-1 may be too small at the predictedoperating temperatures (stored energy) calculated for LOCA Initialization. Please discuss this Issue further,particularly in relation to Question 3 above.

Response

Based upon the analysis performed for Question 3 above, It Is recommended to lb, dl. This lb, d] wasdetermined to be lel in the Question 3 analysis. The recommended replacement equations that were derivedbased upon [b, d], therefore, are:

Equation 12-1 replacemenit(d, el

Equation 12-3 replacement[d, el

whereT = COPERNIC best-estimate temperature (CC),Tsw = temperature (CC) that bounds 95% of the data with a 95% confidence,TL limiting melt temperature CC), andBu pellet bumup (GWdItU).

PAGE 14-67

COPERNIC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231COPERNJC FUEL ROD DESIGN COMPUTER CODE BAW-1 0231Question 29

Are any of the example calculations provided In Section 12 for fuel cores with two 24-month cycles? Itappears that there are no 24-month cycle results presented for the Mark BW-17 design. If so, please explainbecause it Is anticipated that a large number of plants will be switching to 24-month cycles In the next fewyears.

Resnonse

[b, d].

PAGE 1468

Non-Proprietary

Document Control Desk Attachment 2 NRC:03:027April 18. 2003 Page 1

RESPONSE TO OUTSTANDING REQUESTS FOR AODMONAL INFORMATION

TOPICAL REPORT BAW-10231P. CHAPTER 13

"COPERNIC MOX APPLICATIONS"

Below are responses to the outstanding 1"k- and 2e-Round questions received on theCOPERNIC MOX Addendum.

Round 1, Question 8:

The Integral MOX experiments provIded, where centerline temperatures are measured, toverify the COPERNIC Integral thermal predictions of MOX fuel rods are limited to very lowburnup levels, Le., less than 6 GWd/MTU. Please provide COPERNIC predictions of atleast three of the following Halden MOX Instrumented assemblies, IFA-697A1.5I.6, IFA-606, IFA-610, and IFA-648.1, that achieved burnups of approximately 24 GWdlMTM to 67GWdIMTM, or suggest other Halden MOX Instrumented assemblies. Please justify thereasons for eliminating some of the data andtor assemblies for COPERNIC comparisonsand the reasons for selecting others (this should be discussed with the NRC reviewerprior to Issuing a response to the request for additional Information); Also, rodpressures due to fission gas release were measured for two experimental Halden MOXfuel rods In IFA-697A41.6.6. COPERNIC predictions of rod pressure are also needed,where appropriate.

Response:

Framatome ANP considers the lower-bumup experiment IFA-597.4/.51.6 to be atypical ofMIMAS fuel performance. The follow-on experiment IFA-597.7 showed very high fission gasrelease at the beginning of the Irradiation, as Indicated In Halden Status Report HRI 110. Thislevel Is unusual and not consistent with other experiments.

Therefore, Framatome ANP selected the experiments IFA-606, IFA-610.2, IFA-610.4, and IFA-648.1, which are more representative of MIMAS fuel, to demonstrate the adequacy of theCOPERNIC thermal predictions. Measured versus predicted central temperatures for these fourexperiments are provided In Figures 1 through 4.

The fission gas release for IFA-606, rodlet 3, which yielded the highest release fraction, waspredicted' to be 15.9% compared to the measured value of 12.2%.

It is concluded that COPERNIC provides very good agreement with the measured data.

0Non-PAhetw y

Attachment 2

ODocument Control DeskApril18, 2003

NRC:03:027Page 6.1

Figure 4

IFA648.1 Measured and Predicted Peak Teamperatures

I

[dl