ML17341A863

Forwards 150-day Response to NRC 810821 & 1218 Questions Re Pressurized Thermal Shock to Reactor Pressure Vessels. Emergency Operating Procedure 20002 (E-2), Loss of Secondary Coolant, Encl.Reactor Integrity Demonstrated

ML17341A863
Person / Time
Site:	Turkey Point
Issue date:	01/21/1982
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML17341A864	List: ML17341A863 ML17341A865
References
L-82-26, NUDOCS 8202080248
Download: ML17341A863 (73)

Text

~ ).S REGULATORY I RMATION DISTRIBUTION SYST (RIDS)

ACCESSION NBR'8202080248 DOC ~ DATEs 82/01/21 NOTARIZED: YES DOCKET FACIL:50 251 Turkey Point PlantE Unit 0E Florida IPowereand Light C

05000251 AUTH,NAME AUTHOR AFFILIATION UHRIGs R ~ E ~

.Flor ida Power 8Light, Co, RElCI'P ~ NAME RECIPIENT AFFILIATION EISENHUTED.GR Division of Licensing

SUBJECT:

Forwards,150-day,response:to NRC 810821 K 1218 questions re pressurized thermal -shock to reactor pressure,vessels.

Emergency Operating Procedure 20002 (Eee2)<

"Loss. of Secondary

'Co slant rmencl,Reactor tntegrtty demonstrated.

DISTRIBUTION 'CODE:

A0090 iCOPIES RECEIVED:LTR J.

ENCL j SIZE:.

TITLE: Ther mal Shock Tto Reactor Vessel NOTES:

RECIPIENT ID CODE/NAVE ACTION:

ORB ¹1 BC 01 INTERNAL: ACRS ABBOTTpE AEOD COt'I LIAWEB IE 07, NURLEYgT NRR DIR NRR HAZELTON NRR KLECKER NRR ORE ILLYp P NRR THRONzE NRR/DE D IR NRR/DHFS DIR NRR/DL D IR NRR/DL/ORAB 11 NRR/DS I/RA8 NRR/0 IR 05 RE VAGINS pM RES/DRA

<<COP IES LTTR ENCL 7

7 1

1 1

1 1

1 2

2 1

1 1

1 1,1 1

1 1

1 1

1 1

1 1

1 1

-1 1

0 1

1 1

1 1

1 1

1 1-1

'RECIPIENT ID CODE/NAKE 12 ACRS IGNEEE CON AUSTIN ELD IE WOODSIER NRR CLIFFORD

'RR GOODWIN E E NRR JOHNSON NRR LOISr L NRR RANDALL NRR VISSINGpG00 NRR/DHFS DEPY09 NRR/DHFS/lPTRB NRR/DL/ADSA NRR/DS I D IR NRR/DS I/RSB NRR/DST/GIB RES BASDEKAS RES/DET COPIES LTTR ENCL 1

1 1

1 1

0 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 EXTERNALe ACRS NRC,PDR NTIS 10 02 16 16 1

1 1

LPDR NSIC 03 06 1

1

.1 1

TOTAL NUMBER OF COPIES REQUIRED:

LITTR 65 ENCL 63

t I

n

Office of Nuclear Reactor Regulation Attention:

Mr. Darrell G. Eisenhut, Director Division of Licensing U.S.

Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Eisenhut:

.O. BOX529100 MIAMI,F L 33152

/Avzih FLORIDAPOWER & LIGHTCOMPANY January 21, 1982 L-82-26

/

f

<j~",p Oj

'+Q 4P~+i4p

~4"~

"4 Re:

Turkey Point Unit 4 I'ocket Nos.

50-251 Pressurized Thermal Shock to Reactor Pressure Vessels Please find attached our "150 day" response to the questions in the NRC letter dated August 21, 1981.

The attachments also address the additional NRC questions and concerns which were provided in Mr. Novak's letter to FP8L dated December 18, 1981.

We have reviewed the report prepared by our NSSS vendor to respond to the NRC

concerns, and we conclude that these analyses demonstrate that Turkey Point Unit 4 may continue safe operation through the design lite of the reactor pressure vessel without modifications to address pressurized thermal shock.

Our responses demonstrate the integrity of the reactor vessel.

Based on this we concluded that no further development of an action plan is warranted.

however, work is being carried out to evaluate the current low leakage core loading pattern and to assess its effect on reducing'he reactor vessel fluence for Turkey Point Unit 4.

We have also included a separate discussion of operator training and management involvement as requested in Mr. Novak's letter.

It should be noted that credit for operator action was only assumed for the postulated main steam line break accident analysis.

The actions required by our operators to mitigate pressurized thermal shock during a steam line break are included in our emergency procedures.

(See attached procedure)

Very t uly yours, obert E. Uhrig Vice President Advanced Systems and Technology REU/PLP/mbd Attachment cc:

Mr. James P. O'Reilly

. Region II Mr. Harold F. Reis, Esquire

",,18202080248 820121

',. Pj)R. ADQCK 05000251 p)"'pw,',,

PDR, PEOPLE...SERVING PEOPLE

Re:

Turkey Point Unit 4 Docket No. 50-251 Pressurized,Thermal Shock To Reactor. Pressure Vessel,s I.

INTRODUCTION The primary purpose of this submittal is to address the information requested of Florida Power 8 Light Company in U.S. Nuclear Regulatory Commission letter titled, "Pressurized Thermal Shock to Reactor Pressure Vessels",

dated August 21, 1981 with r egards to the Turkey Point Unit No.

4 reactor vessel.

Florida Power II Light Company responded to guestions (1), (2) and (5) of this letter via FPEL Co. letter titled, "Pressurized Thermal Shock to Reactor Pressure Vessels",

dated October 21, 1981.

As a result of reviewing this

response, the U.S.

NRC identified additional information needed.for Turkey Point 4 as outlined in a December 18, 1981 U.S.

NRC letter to the Florida Power 5 Light Company.

This submittal also responds to this requested additional ir.formation.

WCAP-10019, "S.nmary Report on Reactor Vessel Integrity far Westinghouse Operating Plan".s"~..,

prov.'." s the fundamental methods and assumptions used in the Turkey Point Unit 4 thermal shock evaluation and, therefore, is a primary reference in assessing the information contained herein.

Plant specific details are provided herein to amplify'eference [lj and to address the requested information such that a more complete evaluation is available for the Turkey Point Unit 4 reactor pressure vessel.

II.

IRRADIAT'ION INFORMATION k

The analytical methodology and the"design basis used to predict time averaged. fast neutron flux and fluence levels within the pressure vessel/surveillance capsule geometry have been discussed in some de-tail in WCAP-10019, "Summary Report on Reactor Vessel Integrity for Westinghouse Operating Plants"Llj.

The geometric, material, and power distribution information included in this submittal are fully consistent with the methbdology outlined in WCAP-10019 and provide a sound basis for the prediction of the long term fast neutron environment to which the pressure vessel will be exposed.

Also included'ith this submittal is a

summary of the results of the "latest design basis neutron transport calculation performed for this vessel as well as an updated evaluation of neutron dosimetry from each

. of the reactor vessel surveillance capsules which'have been withdr'awn to date.

This dosimetry re-evaluation not only reflects advances in dosimetry analysis methodology and nuclear data, but in addition es-tablishes dosimetry results for all capsules on a consistent basis suitable for direct comparison with analytical predictions.

Geometric information for use in neutron transport calculations is provi~ed in Figures 1 through 3.

In Figure 1, a plan view of the re-actor at the core midplane is depicted., This figure shows the reactor core, lower internals, pressure

vessel, and the inner diameter of the primary biological shield.

Pertinent dimensional information is also included on 'Figure l.

In Figure 2, a detailed description. of the surveillance capsule geometry and associated structure is provided.

This information is sufficient t'o allow accurate determinations of capsule lead factors as well as spectr'um averaged reaction cross-sections for dosimetry applications.

In Figure 3, the azimuthal locations of each of the capsules included in the reactor vessel surveillance program are illustrated.

~

~ j ~

~

NN~g~Oy>> +4'+~ +

++ ~i

+

+'y <r 1~ p r~

~

= 0, 10, 20, 30 and 40 degrees

1. 380 2.959

'.849 g61 6 ~ 591 3.4 4.826 thermal shiel d Figure 2

- Surveillance capsu'ie dimensions (cm)

270o I

y 1/

/Po 30 300 00 lo io' g00 THERMAL SHIELD REACTOR VESSEL Fi'gure 3

Arrangement of Surveillance Capsules

The material descriptions for each of the major zones shown in Figure 1 are listed in Table 1..

The data is presented in terms of volume fractions of solid material in the defined zone of interhst..-

Since neutron transport computations for fluence determinations are of the fixed source variety, fuel enrichment is of no consequence.

However, for consistency the U02 material listed in Table 1 is taken to contain a nominal 3.2 weight percent U-235.

The core power distributions for use in the computation of time averaged. neutron flux and long term neutron fluence levels are given in Figures 4 and 5 and in Table 2.

In Figure 4 the relative fuel assembly power levels are given for one core octant.

The information is presented relative to a core average of 1.0.

Also presented in Figure 4 are a series of fuel assembly numbers which are used to re-late spatial power distribution gradients listed in Table 2'ith core location.

All fuel assemblies labeled Number 1 are assumed to have a.flat power distribution; that is, no spatial gradients exist within these assemblies.'patial gradients for assembly. types 2 through 9

are tabulated in Table 2.

The data in Table 2 is oriented such that the power value in the upper left hand'corner of the table represents the portion of the fuel assembly that is closest to the center of the reactor core.

Values for these spatial gradients are uniformly spaced within each fuel assembly.

The time averaged axial power distri-bution for use in neutron transport calcula'tions is shown graphically in Figure 5.

As discussed in WCAP-10019, these design basis power distribu'tions are statistically based and have proven to yield satis-factory results for long term fluence predictions.

Results of neutron transport calculations for the geometry shown in Figure 1 are presented in Figures 6 through 8.

In Figure 6, the calculated maximum neutron flux levels at the surveillance capsule centerline, pressure vessel inner radius, 1/4 thickness

location, and 3/4 thickness location are presented as a function of azimuthal angle.

In Figure 7, the radial distribution of maximum fast neutron flux

Table 1

MATERIAL DESCRIPTION FOR USE IN NEUTRON'RANSPORT CALCULATIONS Zone Material Volume Fraction Reactor Core plater U02 Zirc - 4 Inconel - 718 Stainless Steel - 304 O.S8864 0.29967 0.10035 0.00281 0.00062 Core Saffle Core Harrel Stainless Steel - 304 Stainless Steel - 304 1.0 1.0 Thermal Shield Stainless Steel - 304 1.0 Surveillance Capsule Structure Stainless Steel - 304 1.0 Surveillance Specimens Low Alloy Steel 1.0 Pressure Vessel Low Alloy Steel 1.0

~ ~

I-figure 0 Lona Te~ Radial Power Ois-.,r"'.bunion C

~

3

'. 0. 73 0.83 0.97

'10 1.05 0.95 Qi 1.07 Qi

0. 97.

1.16 Qi 1.00 1.11 1

1.11 1.13 1

1.03 0.92 1.11 0.97 1.02 1.04 0.85 Oi Qi Qi 14 0.95 1.07 1.12 0.80

0. 93 0.77

.Assembly

~

Power Oistribution Assembly

cO CL EO (f1 I

((I 1.2 1.0 1

~ loKl

~ ff

~ a ~

a(al

~ Iilil "fl (i) el

(::j

~~

'.:I 0.8

~

~

~

~ I

il al I!:;I s

~ ~

fa)l

I.i

~ ~

1')ll I:-:

0>>6

~

~

0>>4 I>>"...I::

111 II(

Q Q

~

~ ls II

~ ~

~

'.i'.I

I

~ a

il'

~ ~ I

,I:I 1'. I

.1<<

~

I>>

.':.::.!l

".::I (Iia

..11

~ ~

~ ol

'1;1 Iof I!

1'>>I II'.1

>>I

.:1(

I.::

il'i i(if ljII o ~I:li

~

~

~

~

~ ~

-.II

~ ~

~ ~ ~

i:2)

<>i

>> jl

~

~

(l( >>

i!Ij fll I>> I I-'ll

~!Ljj ill I ~ '.

sir

~

~ ~

I'.I:

I>>"

It" I-',i j(':i

~I':

ifl'

~

, -I lf:(

~

~

~

~

~ I

~ ~

L':

I;I I

~

!ji Ij>>l Ts 1 Ii" I:)i IIII li)l

~

~

~

Ill

~<<

ill'II II(

liil l)II Vv!

I.

II I

I fa

':}

Ill

~ fII

~ ~ ~ a

~ (

.tf(

l ill

~ ~ ~ ~

>>ll

Ii

!!Ii II

}

II

!i

~ ~

fLs II:I jP'I::I

'lj ilf)

)I);

Il

'j I

II I

~ ~

~

.Il

':l I

~

I

I:)

'i'I

ll)

I::I j:,'Ii 1"

i)i'Il

>>>>I

>> il I

~ ~ ~

I:I!

s

)',(I ill

~ I

~

~

mill lI il jl

~:ll

)I.i I 1 (I

~

j!

I jll II lll (i:I

(((

)li I:

,"I

-:I j

I II IIl

}.

~ o

':I il

.I (I

il ll I

I 1

I

~ ~

a ~

a I(

~ ~

'I I!

I I)

I(!

1ilI:I i))l I (!

I I

I li Ir I

1I.

!>>lj

,lii II:

I(I I:';::I I,I;

'I II>>(l il

I:i I

lji i!'::I

~l!

.:jl}

I::i flI!

il:

ll>>f

~

~

a.asI.':ll I:a) m'I

.)I) "(ff I

ll'. ll!/

(:.I Il) li fll (Il

(!I':.

I ii!i'll

);il fff:

(I) i!i'.

.!I:

I!I i,'(

I>>.

I

.'ll

I;

~

~

(

I i.sl a

/I,I Iifi I>>

~

~

)

s

Ial

III

~ ~ I

I!I

~ I

~

LI

I;

'lli ji)

>>[,,

I 1(

I Il II'I I>>

ll(i II'a lli i;I!

!!l~

II:

I!ii rl:

'Ifi llli Il:i IrI I'

~ili 1

~!iii la)l iIli (l(l (a:,:

'.ii; 1 (I)

I

~ a>>

I I',)(

~ sf

}

Iii!

!ssI II,~

li.ij

".Ii

~

~

I:I>>

~

~

~ >> ~

~ ~

!j,:i "i'

I "i'"

I'

)!I:

I':al a>>

I i

I:I:

>>>>I' o

~

~

1:I;

~ ~

I~,:I

~ ~

Ill

~

~

BI

il

',I:

sl>>>>

':I

il:

'i:I

~ ~

~ ~ ~

1

~ ~

I

~ ~ \\

~ ~

~ o o<<

if" I:ii

'1>>

',(i:

!il!

I IiI

~

1.':ii

~ ~ ~

~

~

~ ~ ~

I: s

Il, "I

~

~

(I!

(

~

II f>>)a

~ s JI a:I

~ ~ ~

1 I

~

~I('i'i jill

~ (

~

~

'.:I:

I'>> I

I.I Ii::

~ I

~

I 'I

': ~ I

'I

~ ~

~

~

~ ~ I

osl

I

~ ~ ~ ~

I:(

~

~

I>> ~ I

)II:

s ~I:

I:ili i.':,i (ii

1:

::I:

I

.:. I.':

~

il s.r II ) iil 0.0 0>>2 0.4 0.6 0 ~ 8 1.0 Fraction of Core lleight'

ASSEMBLY TABLE 2.

eee ev 'Ree'>ouee er+reurretr

n. Oo 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 Oooo 0 ~ 00

0. 00 0 ~ 30 0 ~ 00 0 ~ 00 oo 00 0 ~ 00 0 F 00 0 ~ OO D.OO 0 ~ 00 0 F 00 0 F 00 0 ~ GO 0 ~ CQ 0 ~ 00 0.00 O.an 0 F 00 0 ~ GO 0 ~ Ca 0 F 00 0 F 00 0 ~ 0')

0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 F 00 0 F 00 0 F 00 0 ~ 00 0 ~ 00 0 ~ 00 0 F 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 O.oo o.ao n.oo 0 ~ 00 0 ~ 00 0 ~ 00 0 F 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ OO 0 F 00 0 ~ 00 0 F 00 0 F 00 0 ~ 00 OoOD 0 ~ 00 0 ~ 00 0 ~ 00 Oooa 0 ~ 00 noOO 0 ~ 00 0 ~ 00 0 ~ CO Oo 00 Conn 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 Ooaa 0 ~

DO'oaa 0 oioa Ooao Oooo 0 F 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 60 0 ~ iaa 0 ~ iao 0 ~ iO0 0 ~ 100 0 ~ iQO 0 ~ 00 0 F 00 Q ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 0 ~ 00 O.nO 0 ~ 00 0 ~ 00 0 F 00 0 F 00 0 F 00 0 F 00 0 F 00 Doao 0 F 00 Oooo O.en

n. eo 0 ~ 00 0 ~ 00

a. no 0 ~ 00 0 ~ 00 0 ~ 100 0 ~ 00 0 ~ 00 0 ~ 00 0.:oa.

0 ~ -00 0 ~

00'.

eo 0 ~ 00 lo 21.

1 ~ 23 Oooa 1'3 1 oZi O.aa lo }7 1 ~ 13 0 ~ 00 1 ~ 04

~ 99 oo 00 F 89

~ 83 0 F 00 o67

~ 55 lo 21 l~ '21 lo25 1 o'21 lo 19 1 ~ 21 lor14 1 ~ ll lola 1 ~ OZ

~ 96

~ 87

~ 82

~ 77 o65

~ 55 1 ~ '21 1 ~ 12}

}o 25 1 o'2}

l ~ 20 Lo'20 i}4 L ~ lL lola L ~ 02

~ '97

~ i96

~ 87

~ 82

~ 75

~ 65

~ 55 1 o 21 1 ~ 24 0 F 00 1

25 l ~ 24 0 ~ Go 1 ~ 10 lol4 0 F 00 1 ~ 05 looa 0 ~ GO

~ 90

~ fi5 0 F 00

~ 67

~ 55 1 ~ 20 lo 20 1 ~ 25 1 ~ 25 le 22 l ~ 21 L ~ L4 L ~ 10 1 ~ 09 1 ~ 02

~ 97

~ 89

~ 85

~ 77

~ 65

~ 54 1 ~ 19 1 ~ 18 1 ~ ?3 0 ~ 00 l ~ 24 1 ~ 22 lo 14 1 ~ 10 1 ~ 09 1 ~ Ol

~ 96

'o 96

~ gp 0 F 00

.o 74 o63

~ 53

)4}8 lo )6 Lo 18 1 ~ 22 1 ~ Z2 0 ~ 00 1 o l6 1 ~ L2 0 ~ 00 1 ~ 03

~ 98 0 ~ 00 89

~ 02

~ 70

~ 62

~ 52 lo 17 1 ~ 14-1 ~ '15 1 ~ 15 1 oil 5 1 ~ 16 L ~ ll lo07 l ~ 05 1 ~ )F.

1 o 1'.

. 1 ~ 1 *.

lo l'll:

loll 1 ~ oc l oo'.

leo]

~ 03 o77

~ 68

~ 61

~ 52

~ 01

~ 7.0

~ 61 obC

~ 5 2

~ 98

~ 9t.

~ 'r9 3

~ 91 92 BC iS SEEABLY 1 ~ 40 1 ~ 37 1 ~ 35 1 ~ 34 1 ~ 34 1 ~ 3?

Lo?8 l ~ 23 1

L9 Lo13 l ~ 08 Lo03

~ 95

~ 68

.80

~ 70

~ 60 3

I ~ 39 1 ~ 36 1 ~ 35 L ~ 3'5 1

~ 35 1 ~ 36 L

29 1 ~ 24 Lo??

Lol4 L ~ 08 1 ~ Ob

~ 96 F 88

~ 79

~ 69

~ 50 1 ~ 39 1 ~ 37 l ~ 38 1 ~ 4L 1'o 42 0 ~ 00 L ~ 34 1 ~ 20 0 ~ 00 1 ~ 18 lo}2 0 ~ 00 1 ~ 00

~ 92

~ AO

~ 69

~ 59 1 ~ 40 LE 38 1 ~ 44 0 ~ 10 1 ~ 43 L ~ 38 l ~ 30 1o24 L ~ 22 1 ~ 13 1 o 00 L ~ 07 1 ~ 00 0 ~ 00 o83

~ 69 o58 1 ~ 40 1 ~ 40 1 ~ 45 1 ~ 44 1 o 39 1 ~ 37 1 ~ 28 1 ~ 23 lo2}

1 ~ 12 1 ~ 06 1 ~ 05

~ 96 9?

~ 82

~ 69

~ 58 1 ~ 14) lo 143 Oo60 1 ~ 142 1

139 Ooaa L.i30 1 ~ l25 0 ~ 00 1 ~ f}4 1 ~ 07 0.60

i96

~ 90 D ~ iao

~ 170

~ 157 1 ~ 39 1 ~ 38 lo4}

1 ~ 3h 1 ~ 32 1 ~ 32 1 ~ 24 lo19 lol7 1 oa0 1 ~ 02 1 ~ 00

~ 91

~ 04

~ 78

.66

~ 55

l. 38 1 ~ 36 1 ~ 39

'l ~ 3R 1 ~ 30 lor29 1 ~ 22 1 ~ 16

}~ 114 1 ~ 06 lo 00

~ 97 F 88

~ 102

~ '75

~ 165

~ 54 1, ~ 36 l~ 37 0 F 00 Jo33 1 ~ 31 0 F 00 I ~ Z2 1 ~ 17 0 ~ 00 1 ~ 06

1. ao 0 ~ OO

~ 80

. ~ 81 0 ~ 00

.64

~ 52 loi34 lo 32 lo 34 lo R7 1 ~ 24 1 o'23 lo 16 loll 1 ~ 08 1 ~ 00

~ 94

~ 92

~ 83

~ 77

~ 70

~ 60 1 ~ >31 1 ~ '29

}oi31 g ~ i)4

}~ 2}

1 ~ I?Q 1 ~ LZ 1 ~ 07

}o65 96

-. 9n

~ '08

~ fi0

~ 73

~ 67

~ 58

~ '47 l ~ 28 1 ~ 29 0 ~ 00 los%

l ~ 22 0 ~ 00

'1 o 12 1 ~ Ob 0 ~ 00

~ 95

~ 89 0 F 00

~ 79

~ 72 aoGO

~ 57 o45 1 ~ 24 1 ~ 21 1 ~ 24 1 ~ gg 1 ~ 15 1 ~ 12 1 ~ 04

~ 96 F 08

~ 83

.81

~ 74

~ 69

~ 62

~ 52 o42 lo 18 1 ~ L5 1 ~ 18 Oo 00 1 ~ 13 1'o 07

~ 99

~ 93

~ 91 F 83

~ 78

~ 76

~ 7}

0 F 00

. ~ 57

~ 48 1 ~ 13 1 ~ 08 1 ~ 07 a.O7 1 ~ 05 0 ~ 00

~ 95

~ 90

.0 F 00

~ 79

~ 74 0 ~ 00

~ 65

~ 59

~ 5}.

~ 43

~ 36 lo 06 o99

~ 96 o94

~ 92

~ 90

~ 04

' 00

~ 77

~ 70

~ C5

~ 63

~ 57

~ 51

~ 45

~ 39

~ 33

~ 96

~ 8 fI

~ 8'.

~ 8$

~ 81

~ 7(

~ 7>>

~ 71

~ 6)

~ br

~ 5i

~ 5-

~ 4r

~ 4' 41

~ 3(

~ 3

S SEHAl.f TABLE 2.

RDD SY RDD PDVER DISSUTIDN (Con't) 98

~

~ 97

~ 98 1 ~ 01 lo01 1 ~ 02 1 ~ 01 1 ~ 01

~ 99

~ 99 1 ~ 00

~ 98

~ 95

~ 94

~ 97

~,96 F 00 lo02 le05 le03 1 ~ G3 1 '4 1

~ Cl 1 ~ Ol I ~ 03

~ o8

~ 95

~ 93

~ 98

~ oA le Ol 1 ~ 05 l ~ 08 0 ~ 00 I ~ 07 1.07 0 ~ 00 1 ~ 05 1 ~ 05 0 ~ 00 1 o 05 1 ~ 03

~ or

~ 93 1 ~ 00

1. Oh 0 ~ 00

? ~ 10 1 ~ 09 1

05 1 ~ 05 l.o07 1 ~ 03 1 e 03 1 ~ 07 1 ~ 06 0 ~ 00 1 ~ 02

~ 96

~ o3 1 ~ 00 1 ~ Ol 1 ~ 08 1 ~ 09 1 ~ OA 1 ~ 09 1 ~ 06 1 ~ 05 1 ~ 0>

1 ~ 04 1 ~ 03 1 ~ 07.

1 ~ 04 1 ~ 06 1 ~ 03

~ 97

~ o4 le61 le64 Do60 1 ~ 109 lo69 0 ~ 00 1 ~ I09 1 ~ F08 0 ~ 00 1 ~ 67 1 ~ 66 0 ~ 60 1 ~ 65 1 ~ 65 0 F 00 1 ~ 60

~ I9%

'1 ~ 01 1 ~ 02 1o07 1 ~ 05 1 ~ 06 F 08 1 ~ 06 1 ~ 05 1 ~ 07 1 ~ 04 le03 1 ~ Oh 1 ~ Ol 1 ~ 01 F 01

~ 97

~ 93 1 e'00 1 ~ F02 1 ~ 07 lel04 1 ~ 05 1 ~ 08 1 ~ 06 1 ~ 05 1 ~ !07 1 ~ 03 1 ~.02 1 ~ 05 1 ~ 01 1 ~ IO0 1 ~ 00

~.96

~ 92 1 ~ 01 1 ~ 04 ODOO F 07 F 08 0 F 00 F 00 BOA 0 'O 1 ~ 06 1 ~ 05 0 ~ 00 1 ~ 03 1 ~ 02 0 ~ 00

~ 97

~ 92 le 00 le 01

.1 ~ 06 le 04

~

le 05 le 08 lo 05 1 ~ 04 1 ~ 07 1 ~ OZ 1 ~ 01.

lo 04

~ 98

~ 99

~ 94

~ 9D 1 ~ 00 leIO1 1 ~ 05

>1. ~ 04 1 e 05 1 ~ 07 1e65 1, ~ 04 1 ~ 06 1 ~ 02 1 ~ 01 1 ~ 03 99 98

~ 98

~ l93

~ 89 1 ~ CO le 03 0 ~ ('0 1 ~ CB le08 0 ~ 00 1 ~ C7 le 06 0 ~ CO 1 ~ 04 1 ~ C3 0 ~ CO 1 ~ 01 1 ~ 00 0 F 00

~ 93

~ 87

~ 98 1 ~ 00 1 ~ 06 F 08 le06 1 ~ 07 1 ~ 04 1 ~ 03 1 ~ 04 1 ~ 00 1 ~ 02

~ 98

~ 99

~ 95 89

~ 84

~.96

~ 99

~ 97 98 1 ~ 04 0 F 00 1 o 03 1 ~ 06 F 08 F 07 F 03 0 ~ 00 F 04 1 ~ 03 1 ~ 02 1 ~ Ol

~ 99 0 F 00

~ 92

~ 85

. ~ Al 0 ~ OD

~ 96 93

~ 86

.81

~ 77

'1'3 0 F 00 00

~ 98

~ 99

~ 95

~ 94

~ 96

~ 97 1~OD 1 ~ 02

~ 99

~ 98

~ 95

~ 94

~

~ 95 F 96

~ 97-

~ 97

~ 97

~ 96 F 07

~ 82

~ ~77

~ 74

~ R4

~ 79

~ 75

~ 70

~ 95

~ O3

~ 94

~ 91

~ 95

~ 90

~ 90

.S SENSLY 1 ~ 13 1 ~ 12 lel3 lo 14 1 ~ 15 1 ~ 15

~ le 15 1 ~ 13 le 12 1e09 to07 1 ~ 05 1 ~ 01

~ A5

~ 76 5

1 ~ 12 lelL lol3 l ~ 14 1 ~ 16 io19 1 ~ lh F 14 F 15 1 ~ 10 1 ~ 0>

lo07 1 '1 o96

.89

~ Bl

~ 7 7 1 ~ 12 le ll 1 ~ 15 le20 1 ~ Z2 0 ~ 00 l.e 19 1 ~ 17 0 ~ 00 1 ~ 13 1 ~ 10 0 ~ 00 1 ~ 04

~ 99

~ 89

~ BP

~ 70 1 ~ 12 1 ~ )3 1 ~ 20 0 ~ 00 le 23 1e21 lo lh Eel I 1 ~ 15 1 ~ 09 1 e06 1 ~ 07 1 ~ O'S 0 ~ 00

~ ol e79-

~ 69 1 ~ 13 1 ~ 13 1 ~ 21

] ~ 22 1 ~ 20 1 ~ Zh 1 ~ 15 1 ~ 13 1 ~ 14 l ~ 08 1 ~ 05 1 ~ 06 1 ~ 00 98

~ 91

~ 79

~ 69 1 ~ I1,3 1 ell6 0 ~ 00 1 e'21 1 ~ IZ0 Oo60 1 ~ I17 1 ~ 116 0 ~ OD 1 ~ l09 1 ~ 06 0 F 00 le00

~ l95 OSLO

~ IBO

<<lh9 1 ~ 13

}e13 1 ~ 19 1 ~ 15 1 ~ 15 1 ~ 17 1 ~ 13 1 ~ 11 1 ~ll 1'5 F 02 1 ~ 02 90

~ 86

~ 76

~ 67 1 ~ 12 le 0. 3

,1 ~ 18 1 ~ 1%

1 ~ I14 1 ~ 16 1 ~ 12 1 ~ 09, 1 ~ 10 lo 04 1 ~ (00 1 ~ ~00 e93 F 88 F 84

~ 75

~ 65.

1 ~ 12 le 14 0 F 00 lel7 1 ~ 17 0 ~ 00 1 ~ 14 1 ~ 1Z 0 F 00.

lo06 1 ~ 02 0 ~ 00

~ 94

~ 89 0'0

~ 75

~ 64 1 ~ 11 le 12 le 17 1e 13 la 13 1 ~ 14 lo 10 lo07 lo 08 1 ~ 01

~ 97

~ 97 90

~ 85 F 81

~ 72 '

63 1 ~ 01 1 ~il1 1 os16 1 ~ '12 1 ~ 12 1 ~ 'l4 1 ~ 09 1 ~ 06 1 ~ Oh F 00

~ 96

~ 96

~ (88

~ 84

~ ISO

~ 70-

~ 6l.

1 ~ 11 1 ~ 13 0 ~ 00 lolb lol5 0 ~ 00 1 ~ 11 1 ~ 08 no 00 1 ~ 02

~ 98 0'O

~ 90

~ Pb OeCO

~ 71

~ (0 1 ~ 10 1 ~ 10 1 ~ 16 1 ~ 16 1 ~ 13 1 ~ 13 1 ~ 07 1 ~ 04 F 04

~ 98

~ 94

~ 94

~ 88

~ 84

~ T8

~ 6T e58 1 ~ 08 1 ~ 08

~ 1'3 ODOO 1 ~ 14 1 ~ 12 1 ~ 06 F 03 1 ~ 03

~ 96

~ 92

~ 93

~ 88 0 F 00

~ 74

~ 65

~ 57 1 ~ 06 1 ~ 05 1 e'07 le 11.

1 ~ 12 0 F 00 F 07 1 ~ 04 0 F 00

~ 97

~ %3 0 F 00

~ 86

~ 80

~ 70

.62

~ 55 1 ~ 05 F 03 1 ~ 04 lo 04 1 ~ 05 1 ~ 06 1 ~ 02

~ 99

~ '98

~ 92

~ 88 87 F 80

~ 73

~ 66

~ 60

~ 53 1 ~ 06 1 ~ 03 1 o03, 1 ~ 02'

~ 02 F 01

~ 94

~ 09.

~ 05

~ A3

~ 77

~ 71

~ 65

~ 59

~ 52

TABLE 2.

ROO BV ROO ROBER OIQRIBUTIOH (CRRt) aSSEHBL 1 ~ 4l 1 ~ 36 lo 36 1 ~ 35 1 ~ 35 1 ~ 33 1 ~ 29 l o? 5 1 ~ 23 le 18

.le 12 1 ~ 07 lo00

~ 92 F 84

~ 75

~ 66 Y

6 1 ~ 34

-1 ~ 37 1 ~ 36 lo37 1 ~ 36 1 ~ 37 lo2>

1 o?7

l. o26 1 ~ 17 1 ~ 1Z 1e10 to00

~ 92

~ 82

~ 73

~ 64 1 ~ 39 1 ~ 37 1 ~ 39 F 43 1 ~ 4?

0 F 00 1 ~ 35 1 ~ 30 0 ~ 00 le?0 le 16 0 ~ 00 1 ~ 04 95

~ 83

~ 73

~ 63 le 39 1 ~ 39 1 ~ 4I 0 ~ 00 lo 4Z 1 o 40 le 30 1 ~ 25 lo 24 1 ~ 16

l. 10 1 ~ 09 1 ~ 03 0 ~ 00

~ 85

~ 72

~ 62

]o40 1 ~ 40 1 ~ 45 1 ~ 44 1 ~ 39 1 ~ 38 1 ~ 28 1 ~ 24 1 ~ 23 1 ~ 14 1 ~ 04 1 ~ 07

~ 99

~ 93

~ 84

~ 7?

~ 61 1 ~ t40 1 ~ l%3'.

Ooe0'oA3 1 ~ l39 OROO 1 ~ l30 1 ~ I26 0 ~ 00 1 ~ I15 1 ~ I10 OOOO

.t98i

~ l91 0.60

~ IT2

~ V)5 1 ~ 38 1 ~ 37 1 ~ 42 1 ~ 35 1 ~ 32 1 '2 1 ~ ?4 1 ~ 19 1 ~ 18 1 ~ OE) 1 ~ 04 1 ~ 01

~ 92

~ 84

~ 79

~ 58 1 ~ 36 1 ~ 37

~ 1 ~ 39 1 ~ 32

~29

'29 1 ~ 21 1 ~ 16 1 ~ 15

1. 06 1 ~ 01

~ 97

~,88

~ 81

~ 77

~l66

~ 56 1 ~ 36 1 ~ 38 0 F 00 1 ~ 33 1 ~ 30 0 F 00 1 ~ 2?

1 ~ 17 0 F 00 1 ~ 06 1 ~ 01 O.no-

~ 88

~ 82 0 F 00

~ 65

~ 54 1 ~ 33 1 ~:31 1 ~ 33 lo 27 le 23 lo 23 1 ~ 14 1 ~ 10 1 ~ 08

~ 99

~ 94

~ 9L

~ 83

~ 76

~ 71

~ 61

~ 52 le~33 1 ~ '28 1 ~ 30 1 ~ 24 1 ~ '20 1 o'1 4 1 ~ 10 1 ~ 06 1 ~ 04

~ 95

~ AO

~ 88

~ 80

~ 73

.6S-

~ 158

~ 49 le 27 1 ~ 28 0 'O 1'4 1 AD ?0 0 ~ 00 le09 1 ~ 04 0 o'00

~ 94 89 0 ~ 00

~ 79

~ 72 Oo GO

~ 57

~ 47 1 ~ 23 1 ~ 20 1 ~ 23 1 ~ 20 1 ~ 13 1 ~ 10 1 ~ Ol

~ 96

~ 94 F 87

~ 82

~ 80

~ 73

~ 69

~ 62

~ 52 F 44 1 ~ 17 1 ~ 13 lo15 0 ~ 00 le 09 1 ~ 04

~ 96

~ 89

~ 82

~ 77

~ 75

~ 70.

0 ~ 00

~ 57

~ 48 F 41 1 ~ 10 1 ~ 05 1 ~ 04 1 ~ 04 1 ~ 01 0 ~ 009l

~ 87 0 ~ 00

~ 70

~ 73 0 F 00

~ 64

~ 58

~ 50

~ 43

~ 37

.1 ~ 02

~ l96

~ 9S

~ 91 F 88

~ 87

~ Bl

~ 77

~ ~ 75

~ 69

~ 64

~ 61

~ 55

~ 58

~ 44

~ 34 o7'7i

~ 7 o7

~ 6,

~ 6

~ 5 o5

~ 4

~ 4

~ 3

~ 3

. WSSENBt.Y 1 ~ Ol lo00 1 ~ 00 1 ~ Ol l ~ 02 1 ~ 03 1 ~ 02 1 ~ I32 l ~ 0?

1 ~ Ol lo00 1 ~ 00

~ 9 9

~ 97

~ qq

~ 94

~ Q4 7

F 00 1 ~ 00 l ~ OP.

F 04 lo07 1 ~ 04 1 ~ 03 lo05 F 02 F 02 1,04 lo01

~ 98 o95

~ 94

~ 93

) F 00 1 ~ 00 1 ~ 0'5 1 ~ OA F 09 0 ~ 00 1 ~ OB F 07 0 ~ 00 1 ~ 06 I ~ 07 0 ~ 00 1 ~ 06 1 ~ 03

~ 90

~ 94

~ 93 1 ~ 01 1 ~ 03 ie Oe

.0 ~ 00 1 ~ 11 1 ~ 10 1 ~ 06 le 05 1 ~ 07 1 ~ 05 1 ~ 05 1 ~ 08 1 ~ 07 0 ~ 00 1 ~ 02

~ P5

~ 03 1 ~ 03 1 ~ 04 is 10 1 ~ 11

.1 ~ 09 1 ~ 10 1 ~ 06 lo C5 le 08 1

0<

1 ~ 05 1 ~ 07 1 ~ 05 1 ~ 06 1 ~ 03

~ 96

~ 93 le04 le67 OoOO l ~tll 1 ~ I10 0 ~ 00 1 ~ 09 1 ~ '09 0 ~ 60 l ~ ~08 lool 0 ~ IOO 1 ~ 06 le05 0.00

~ I90

~ 194 1 ~ 03 F 04 F 09 1 ~ 06 1 ~ 06 1 ~ 10 1'6 1 ~ C'6.

1 ~ 08 F 04 1 ~ 04'

~ 06 1 ~ 02 1 ~ 00 1 ~ 01

~ 96

~ 93 lo63 le 04 1 ~ 08 1 ~ 105 1 ~ 06 1 ~'09 1 ~ 06 le 05 1 ~ 108 1 ~ 04 1 ~ l03 1 ~ 05 lo 00 1 ~.00

~ '95

~ 92

~

~

1 ~ 03 1 ~ 06 0 F 00

~ 1 ~ 08 1 ~ 09 0 ~ 00 1' 09 F 08 0 F 00 1 ~ 06 lo05 0 F 00 le02 F 01 0 F 00

~ 96

~ 91-1 ~ 02 le 03 le 07 lo 06 1 ~ 06 le'08 1 ~ 05 1 ~ 04 1 ~ 06 lo 02 le 01 le 03

~ 99

~ 97

~ 98

~ 9Z

~ 90 leal 1 ~ 03 1 ~ IOB 1 ~ 01 1 ~ 06 0 ~ CO

) ~ 04 1 ~.03 lo'05 0<

1 ~ 00 1 ~ 03

~ 99

~ 96

~ iq6 92

~ 80 1 ~ 06 leC5 0'O 1 ~ 03 1 ~ 03'

~ 00 lo 00

~ 98 0 ~ 00

~ 92

~ P6 1 ~:0'J 1 ~ 09 1 ~ 05 1 ~ 08 F 68

. 0 F 00 le01 1 ~ 02 1 ~ 07 le 08 1 ~ 06 F 07

) ~ 02 1 ~ 01 F 03 F 00

~ 98 1 ~ 00

~ 97

~ 98

~ 94 o87

~ 83

~ 99

1. ~ 00 1 ~ 05 0 ~ 00 1 ~ 07 1 ~ 06 le 01 F 00 lo02

~ QB

~ 97

~ 99

~ 98 0 F 00

~ 91

~ 83 o00

~ 97

~ 97

~ 99 1.03 F 04 0 ~ 00 lo02 1 ~ 02 0 F 00

~ 99 e97 0 F 00

~ 95 F 84

~ 79 o76

~ i96

~ 46

~ 97

~ 98 lo00

~ 98

~ i96

~ 97

~ 93

~ 92

~ 93

~ i86

~ 84

~ 79 e75

~ 71

~ 9

~ 9

~ 9

~ 0

~ q

~ 9

~ q

~ 9

~ g

~ 0

~ 8

~ 0

~ 8

~

1

~ t

0

TABLE 2.

ROD BY ROD ROVER DI+tTOUT TON (coe)

~S SF!18L 1 ~ 36 1 ~ 34 le 34 1 ~ 34 1 ~ 35 lo35 1 ~ 33 Le31 1 ~ 31 1 ~ 27

l. ~ 24 L ~ 71 le17 1 ~ L}

1 ~ 05

~ 97

~ 87 1 ~ 3'>>

1'7 LE 37 LE 35 LE 36 1 ~ 3'I 1 ~ 33 L ~ 32 l ~ 33 l e? 6 1 ~ 23 Lo23 l ~ 15 04 F 01

~ 01 1 ~ 32 lo 32 1 ~ 35 le 40 1 ~ Cl

0. 00 1 ~ 30 L ~ 35 0 ~ 00 1

~ ?4 le 26 0 ~ 00 L ~ La 1 ~ 10

~ 99

~ 77 1 ~ 32 1 ~ 33 l ~ 39 0 ~ 00 l ~ 40 1 ~ 40 l ~ 32 1 e29

~ 24 1 ~ 73 l ~ 20 L ~ 20 le 15 0 ~ 00 L ~ 00

~ 87

~ 75 1 ~ 33 1 ~ 33 1 ~ 34 1 ~ 40 1 ~ 37 1 ~ 37 1 ~ 29 1

~ 26 1 ~ 27 1 ~ 20 le 16 lol6 1 ~ 09 1 ~ 06

~ 98

~ 05

~ 73 lo82 1o135 0 F 00 1 ~ I41 1 ~ 136 0 ~ 00

l. ~ 13 0 L ~ 127 O.OO 1 ~ 120 1 ~ Il6 0 ~ OO 1'7 lo61 0 ~ OO

~ 6'>>

~ I73 1'9 1 ~ 24 1 ~ 35 1 ~ 30 lo20 1 ~ 29 lo23 1 ~ 20 lo20 1 ~ 13 1 oba loC7 F 43

~ 89

~ 79

~ 69 1 ~ 26 1 ~ 28 1 ~ 31 1 ~ 26 1 ~ 24 1 ~ 26 1

19 le 16 lo 15 1 ~ F08 1 ~ 04 1 ~ 03

~ 95

~ 89

~ 85

~ I75

~ 66 1 ~ 26 1 ~ 28 0 F 00 1 ~ 26 1 ~ 25 0 ~ 00 F 19 1 '5 0 F 00 F 07 1'3 0 F 00

~ 93

~ 88 0 F 00

~ 73

~ 63 le 22 1 ~ 22 1 ~ 24 lo 19 1o 17 1 ~ 17 loll lo 07 lo 07 1 ~ 00

~ 95

~ 9'>>

~ 87

~ 81

~ 77

~ 67

~ 59 1 ~ 10 le l.a 1 ~ 21 "lo15 1 ~ I13 LE 13 LE 05 1 ~ 02 1 ~ 02

~ 95

~ l90

~ 90

~ 8 2

~ 77

~ I73

~ 53

~ 55 lo 15 le 17 0 ~ 00 le 15 1'2 0 F 00 le 04 1 ~ 00 0 F 00

~ 92

~ 09 OeGO

~ 73

~ 75 0 ~ GO

~ 61 e50 1 o LO 09 1 ~ 12 1 ~ 10 1 ~ 05 1 ~ 03

~ 96

~ 9.2

~ 91

~ 85 F 01 F 00

~ 74

~ 70

~ 64

~ 55

~ 48 1 ~ 04 1 ~ 02 1 ~ 04 0 ~ 00 F 00

~ 96

~ 09

~ 85

~ 05

~ 79

~ 75

~ 74

.e9 0 F 00

.5a

~ 50

~ 43

-.. 9a

~ 44

~ 93

~ 9'>>

~ 42 0 ~ 00

~ 04

~ 01 0 ~-00

~ 74

~ 70 0 F 00

.e3

~ 50

~ 51

~ 44

~.90 F 85

~ 02

' 80

~ 70

~ 70

~ 73

~ 70

~ 60

~ 63

~ 60

~ 58

~ 53

~ 48

~ 44

~ 39

~ 35

~ al

~ 7t,

~ 7l

~ 61 oct R

~ t)i

~ 5

~ 5t

~ 5(

~ 4 I

~ 4(

~ 41

~ 3l

~ 3(

~ 31

.~ S SE>>lBLY 1 ~ 06 1 ~ 04 1 ~ 04 l 04 l, ~ 06 l ~ 07 1 ~ 06 l ~ 05 l ~ 06

]e05 l.n5 I e04 1 ~ 04 l ~ 02 l ~ 3L l ~ 00 1

e 00 9

1 ~ 04 1'3 1 ~ 03 1 '5 F 07 L ~ 09 leoe le06 1 ~ 0>>l 1 ~ C5 1

~ 05 le07 le04 1.02

~ 99

~ 90

~ 97 1 ~ 04 1 ~ 03 1 ~ 06 lo LO 1~lL

o. on le 11 1 ~ 09 O. On 1 ~ 00 1 ~ 09 0 F 00 1 ~ 00 1 ~ 06 1 ~ 00

~ 97

~ 96 1 ~ 04 le 05 Lo 10 0 ~ 00 1 ~ 12 l o 12 L ~ 00 1

~ 06 l ~ 09 L ~ Oe 1 ~ 06 1 ~ 09 1 ~ 00 0 ~ 00 1 ~ 03 90 1 ~ 06 1 ~ 07 1 ~ 11 1 ~ 12 1 ~ll le ll 1 ~ 07 1 ~ 06 1 ~ 09 1 ~ 05 l ~ 05 1 ~ 07 1 o 05 1 ~ 06 1 ~ 04

~ 98

~ 95 loOT 09 0 ~ 00 1 ~ 112 l ~ 111 0 ~ IOO 1 09 l ~ 109 0 ~ 60 1o07 1 ~ Ob 0 ~ 00 1 ~ ~05 1 ~ 04 0 F 00 199

~ 145 l.oe 1 ~ 06 1.~ 11 1 ~ 00 lo07 1 ~ 09 1 ~ 06 1 ~ 05 1 ~ 07 1 ~ 03 1 ~ 02 1 ~ 04 1 ~ 00

~ 99 l ~ 00

~ 96

~ 93 1 ~ ~05 1 ~ 06 1 ~ 09 1 ~'06 1 ~ 09 1 ~ 05 lo 04

1. Oe 1 ~ 02 1 ~ 01 1 ~ 02

~ 99

~ ~97

~ 94

~ 91 lo06 1.08 0 F 00 1'9 F 09 0 F 00 F 07 1 ~ 06 0 F 00 F 04 1 ~ 02 0 F 00 lo 00

~ 99 0 ~ 00

~ 94

~ 90 1o 05 lo 05 1 ~ 08 1 ~ 06 1 ~ 05 le 07 1 ~ 03 1 ~ 02 1 ~ 04

~ 99

~ 40

~ 99

~ 96

~ 94

~ 90

.aa

) ~ 66 le 05 1 ~ 06 1 ~ 02 1 ~ '01 Lo62

~ 98

~ 97

~ 199

~ 95

~ 93

~ 93

~ 89

~ 06 1 ~ 09 1 ~ 07 0 ~ GO 1 ~ 04 1 ~ G2 0 ~ GO

~ 99

~ 99 0 ~ 00

~ 96

~ 94 0 ~ 00

~ 00

~ P.4 le 65 1 ~ C4 1 ~ 05

. 1 ~ 07 l ~ 09 0 ~ 00 1 ~ 04 F 04 1 ~ 08

-1 o 08 1 ~ 05 1 ~ 05 1 ~ 00

~ 99 F 00

~ 96

~ 95

~ 96

~ 92

~ 93

~ 40

~ 84

~ 80 lo 02 1 ~ 02 1 ~ 06 0 F 00 1'6 F 04

. ~ 99

~ 97

~ 94

~ 93

~ 94 o93 0000

~ 86

~ 79

~ 76

) ~ 01

~ 99 l ~ 00

') ~ 03 1 ~ 04 0 ~ 00 1 ~ 00 o'99 0 ~ 00

~ 45 e93 0 F 00

~ 90

~ 06

~ ao

~ 75

~ 72

~ 98

~ 97

~ 98

.9a

~ i99

~ 96

~ 94

~ 9'>>

~ 90

~ 84 F 88 F 04

~ 79

~ 75

~ 71

~ 68

~ 9;

~ 9i

~ 9]

~ 9i

~ 9' 9:.

~ 41

~ 91

~ 8I

~ at

~ ac

~ 0(

~ 7t

~ 7i

~ 6(

~ tT'

~ 00 1 ~ Ol

I (E

> 1.0 Nev) through the thickness of the pressure vessel is shown.

The relative axial variation of neutron flux within the vessel is

'iven in Figure 8.

I The data given in Figures 6 through 8 can be used directly 'to develop I'ead factors relating each surveillance capsule to any point in the pressure vessel; or, in conjunction with appropriate full power operating times, to derive fast neutron fluence distributions within C

the vessel.

Since initial startup, two surveillance capsules have been withdrawn from the Turkey Point Unit 4 reactor.

In 1976, Capsule T was removed frcm the 0'zimuthal position while in 1979 Capsule S was withdrawn

~ frcm the 10'ocation.

Neutron dosimetry from both Capsules T and S

was evaluated by Southwest Research Institute and the results were documented in SWRI-02-4221 and SWRI-02-5380.

For the purposes'f this submittal, the SWRI radiometric counting data have been extracted from the appropriate reports and the fluence determinations have been updated to reflect the following changes in surveillance dosimetry methodology.

l.

Application of the best available nuclear data 2.

Use of spectrum averaged cross-sections which include capsule perturbation effects 3.

Spatial gradient corrections to measured count rates to permit neutron flux evaluating at the geometric center of the capsule.

Suruei ttance ca jsules at radi i90.900 csi Surveillance capsules at radius

= 192.488 cm Reactor vessel waLL inner surface at radius

= 197.635 cm 4

Reactor vessel walL, L/4 thickness, at radious

= 202.485 cm 5

Reactor vessel wall, 3/4 thickness, at radius

= 212.487 cm 1.GGE~i2 7.GCE+ii

.GCEtii 3.00E+ii 2.00E+11 AZIHUTHAL DISTAUBUTIOH OF NEUTRON FLUX tH/CH -SEC)

~ MITHIN PV

).OOE+i'.00E+'.9 c

5.00E~i0 CA s

I 3.POE-!0

~ '2.00E-l3

~'.OOEtiO

~~ 7.00E~04 5.OOE+09

~ 3.00E 04 L'-OOE+OS f.OOE&9 CD CD CD CD CD CD CD CD CD CD CD CD CD CD C3 CD CD

~ ttj CD CD CD CU CD CD CD CD CD CD LA Pl CD CD CD e

CD AZ IHUTHAL AHCLE (OEG)

Figure 6

I

~

I

~

>)0 9

8 7-'.-

e I

J I

5 Z

Q a

2 44 II v5 I'

U 5

~ g V

~0 Xv C

I I

IA I41 C

QC Og AL

~ ~

XB 4IJ Id 444 m, yolO 9

~

~

~

~

3.00

Y 3

- GUre 8

).0 9

d.

7 5

3 2

I i '

~

I I

I I

I

~

9 d

7 6.

I

~ I-'

I I.

I I.

~

~

~

\\

I I

~

~

~

I

~

~

~

5 2

0,0) 8 L

"oRO 0

I)00,

/DO 300

t

Updated results based on the FE (n,p)

Hn reaction are summarized 54 54 in Tables 3 and 4 for Capsules T and S, respectively.

The Capsule T

data is in excellent agreement with prediction with the average measured value deviating from the calculation by only 7X.

However, the Capsule S data indicates a larger mismatch with prediction with the measurement exceeding calculation by some 37/..

A disagreement of this magnitude is not typical for Westinghouse Plants.

Furthermore, the Capsule S data appears to be inc'onsistent with Capsule T data 'in that the measured fluxes agree within 10Ã, yet flux levels at the 10'osition should be 35 - 40/ below the 0'lux level.

A recheck of the dosimetry or supplementary measurements will verify that the Capsule S data are unrealistically high.

Over the last several fuel cycles'eginning with fuel cycle No.

5 (August 1978) the Turkey Point Unit 4 reactor pressure vessel has used forms of a low leakage core'onfiguration.

A qualitative assessment of'he effect of these core patterns on vessel fluence indicates that the current rate of flux may be as much as 30K below design values.

Further work is being carried out to evhluate the current low leakage pattern and to assess more exact core configurations for reducing the rate of flux as much as possible.

MONITOR SATURATED ACTIVITIES FOR CAPSULE T (FLA)

Adjusted Saturated Saturated Activity Activitv (DPS/mg)

(DPS/mg)

Fast Neutron Flux (n/cm2-sec)

Radial Reaction and Axial Location Location (cm)

.Capsule T

Ca 1 cul a ted, TABLE 3 COMPARISON OF MEASURED AND CALCULATED FAST NEUTRON FLUX Fe (n

)Mn W-17, W-21 S-67 S-69 S-73 Average 191.46 191.46 191.46 191.46 191.46 7.52 (3) 8.36 (3) 8.80 (3) 8.61 (3) 8.83 (3) 7,14 (3) 7.94 (3) 8.36 (3) 8,18 (3) 8.39 (3).

. 1.64 (11)

. 1.82 (11) 1.91 (ll) 1.87 (11)

~

1.92 (ll)

. 1.83 (11) 1.67 x 10

0,

TABLE 4 COHPARISON OF HEASURED AND CALCULATED FAST NEUTRON FLUX HONITOR SATURATED ACTIVITIES FOR CAPSULE S (FLA)

Radial Reaction and Axial Location Location (cm) 54

)

54 Saturated Activity (DPS/mq)

Adjusted Saturated Activity (DPS/mg)

Capsule S

Calculated Fast Neutron Flux (n/cm2-sec)

Top Hiddle Bottom Average 191.46 191.46 191.46 7,37 (3}

7,15 (3) 7.31 (3) 7.03 (3) 6.82 (3) 6.97 (3) 1.61 (11) 1.56 (11) 1.60 (11) 1.59 (11) 1,19 (11)

Top Middle Bottom Average 192.46 192.46 192.46 6,39 (3)'.58 (3) 5.85 (3) i 6.94 (3}

6.25 (3) 7.41 (3) 1.74 (11) 1.59 (11) 1.70 (11) 1.68 (11) 1.19 (ll)

III. VESSEL WELD MATERIAL INFORMATION A.

Weld Locations and Chemistry P

Figure 9 provides the axial locations of the vessel weld seams with respect to the core.

identified with each weld location using the Figures 6 through 8.

These indicated values leakage core configuration patterns employed since August 1978.

Turkey Point Unit 84 reactor End-of-life fluences are flux distributions given in do not account for the low at Turkey Point Unit 4 The weld chemistry along with the weld wire heat number is given'-in Table 5

for the critical weld seam, the intermediate shell to lower shell weld.

The estimated m'ean copper content, including the range and standard deviation, are also reported in Table 5 based upon all reported measurements for the critical weld heat.

B 'TNDT Values Based Upon Turkey Point Unit 3 Survei 1 1 ance Test Resul ts The surveillance weld for Turkey Point Unit 3 had the same weld wire heat number and the same flux lot number as the beltline weld in Unit 4.

Although it has been suggested that the Unit 4 Capsule T surveillance result should be applied to predict BRTNDT the Unit 3 surveillance information is still more representative of both Units 3 and

• 4 beltline welds for the reasons provided below.

Figure 10 provides a plot of thirteen (13) surveillance results for several BSW reactor

vessels, including those for Turkey Point Units 3 5 4, against the NRC Reg.

Guide 1.99 trend curves.

The copper content of these data range from 0.25 to 0.35% wt, which. are representative of the Turkey Point

'eltline welds.

All of the points are of high nickel content and have experienced fluence levels equal to or greater thar, the subject weld.

As can be seen from Figure 10, 12 of the 13 surveillance results show RTNDT values less than those predicted by the Reg.

Guide 1.99 trend curves.

Coat Uct

(

I ~~q,z Cur.ca

~ o Dure.HM+>>

(~p g j zs H/4ligiBCE)

Figure 9.

Turkey Point Unit 5'4 Reactor, Vessel Weld Locations and Fluences

Critical Weld Seam I

TABLE 5 TURKEY POINT UNIT 4 WELD CHEHISTRY Chemical Composition (Wt.X)

C Hn P

S Si CR.

tli Ho Cu Inter Shell to Lower (b)

Shell Girth Weld(

)

(c)

.07 1

28

.021

.014

.52

.17

.57

.36

.21

.076 1.26

.011

.018

.66

.14

.~7

.42

,31 NOTES:

(a)

B&W Weld Code No.

SA1101 (Wire Heat No..71249 - Linde 80 Flux Lot No. 8445)

(b)

B&W Weld Hetal gualification Analyses (c)

W Unit 3 Surveillance Program Analyses (Same as B&W Meld Code No.

SA 1101) 0 Copper analysis on as deposited weld metal for Weld. Wire Heat No. 71249 Cu

~x

.23

.21

.19 B&W Wild Hetal gualification Tests

,31

.30

.35

.34

.32 W Surveillance Test Hean

= 0.28 Range

= 0.19 - 0.35 Standard Deviation

= 0.06

0

Ul CD UJ OC Uj UJ UO I

Lal I

CA Ch 500 IGLOO 300 200 IOO L~eend 0

~0.35K Cu

~0.30K Cu

~0.25K 'Cu Unit 3 Unit 4 Fracture Analysis Curve lgq I/2 IIRTRRT = 340+

I 000( ( Cu-0.08)+ 5000(4 P-0.008)]

lf/IO ]

UPPER LIHIT Ch LLI I

Ch, Uj CC O

I CL I-OC<<I 50' tpl 7 i

0.3S~gu 0.30<Cu 0.25,~CU 0.20)0Cu 0. IS<Cu 8 lple 2

0.10gcu IlP = O.OI2 6

ego'OWER LIHIT Cu = 0.08 P

~ 0.008 6

e 1020 FLUENCE, n/cm2 (F.> IHeV)

Figure 10 Plot of Surveillance Weld Data for B&W Reactor Vessels Against Regulatory Guide 1.99 Trend Curves

A

The Unit 4 surveillance result is the only point showing a positive increase in RT>0T when compared against the trend curve 'prediction and, thus, appears to be an outlier.

A review of the irradiated Chhrpy V notch data for the Unit 4 weld metal

~ indicates that a sufficient number of tests were not performed in the 'transition region to adequately define the transition temperature.

This lack of sufficient data probably accounts for the higher transition temper'ature shift reported for this weld metal.

The Turkey Point Unit 3 surveillance result appears to be near the mean of the 12 points showing a lower shift.

Therefore, the Unit 3 surveillance result is considered to be.representative of the beltline weld in Unit. 4 and has been used in recent fracture analyses for pressurized thermal shock.

To further support the use of the Unit 3 result, Figure ll shows the results of two dynamic fracture toughness tests on IX-MOL fracture mechanics 18 2

specimens from the Unit 3 weld metal, which were irradiated to 5.7 to 10 n/cm in the Unit 3 surveillance capsule T

The comparison of these results L33 with-the adjusted KIR curve show that use of an adjusted KIR curve based on Chai py transition temperature shifts is very conservative for this material.

The fracture analysis results presented in'the following section for the small loss-of-coolant-accident, small steam break, and Rancho-Seco transient are based upon the Unit 3 surveillance result and the NRC Reg.

Guide 1.99 trend curve.

It was assumed that the trend curve for Unit 4 passes through the Unit 3 surveillance data point and follows the slope of the Reg.

Guide trend curve until it intersects the upper limit line following this lower slope until end-of-life as shown in Figure 10.

The initial RTNpT value is 3'F.

In conjunction with this information, the predicted shift is 211'F for a fluence of 1.1 x 10 n/cm (current total f')uence on the inner wal])

19 2

using Reg.

Guide 1'.99.

,f

250 225 200 175 150

~

~\\

125 100 75 50 0

25 W.O.L. Results from Ref. [3j Irradiated to y=5.7xl018 n/cm2 KIR ADJUSTED FOR QRTNDT 1 55 F

CALCULATED FROM IRRADIATED CHARPY SPECIMENS FROM CAPSULE T 53j

-200

-100 100 200 300 T, 'F TEMPERATURE TURKEY POINT UNIT 3 SURVEILLANCE DATA (WELDMENT)

FIGURE

11.

C.

Rate of Increase of RTNDT J

Future rates of RTNDT increase have been previously generated for the Turkey Point Unit 4 reactor vessel using the slope of prediction curves presented in proposed ASTM Standard, "Predicting Neutron Radiation Damage to Reactor Vessel Material".

However, based upon the information given above in combination with the Reg.

Guide 1.99 trend curve, RTNDT is increasing at the rate of ~ I3.5'F/EFPY for the next 10 effective full power years and ~3.5'F frcm then to the end of design life.

These values, which are averages since the rate of increase in RTND does not vary linearly with time, apply to the. inner wall of the reactor vessel beltline weld.

The.low leakage core patterns, which have been employed since August 1978, are currently under evaluation for Turkey Point 84.

The rates of increase of'TNDT per EFPY taking these configurations into consideration are unavailable at this time.

However, the values are anticipated to be lower than those presented above.

D RTNDT mimit and Criter ia for Continued Oper'ation RT should not be utilized as a sole parameter to determine the acceptability of the integrity of the reactor vessel for any specific plant.

RTNDT also should not be used as the sole parameter to compare the relative acceptability of different vessels.

However, if a limiting RTNDT is to be defined for a specific vessel, it should be qualified to the specific methodology utilized to calculate an acceptable lifetime for the specific vessel.

The basic methodology that has been utilized in calculating acceptable lifetimes for specific vessels is given in Reference

[1j.

This report describes the

'arious operational and non-opPj'ational transients considered, the thermal and hydraulic methods, the fracture mechanics

methods, and the acceptance criteria.

In addition to the Owners'roup work, plant specific analyses have been performed to evaluate the impact of key plant specific parameters such as fluence, transient characteristics, material properties, vessel geometry and weld locations on the acceptable vessel lifetimes for Turkey Point Unit 4.

In summary, the most appropriate criterion for continued operation should be based upon fracture analysis results.

These results incorporate all of the variables which can affect vessel integrity.

~

IV.

~

N ANSIENT FRACTURE ANALYSES SHOWING BASIS FOR CONTINUED OPERATION Detailed integrity assessments have been carried out for the Turkey Point Unit 4 reactor vessel postulating the occurrence of the four most limiting thermal 'shock events applicable to the subject plant:

o Large loss-of-coolant-accident (LOCA) o Small LOCA o

Large steamline break o

Small steam break The Rancho-Seco transient was also arbitrarily evaluated although it is an event which i.s not directly representative of the Turkey Point Unit 4 plant.design.

A.

Transient Development A complete discussion of the applied transients relative to the analytical methods and assumptions used in the transient development and to the description of the associated temperature,

pressure, and flow rate histories are provided in Reference [1].

Additional similar transient information for the large LOCA-and large steamline break events, which were analyzed on a plant specific basis for the Turkey Point Unit 4 reactor vessel in late 1976 and early 1977, are given in Reference [4].

Since this transient information is well documented in Reference [lj and [4], it will not be repeated here.

However, some conservatisms

. inherent in the applied tragsients for the Turkey Point Unit 4 design are discussed in Section V.

1. Probabilit of Transient Occurrence The transient events which have been postulated for evaluation in this report range from small breaks to the instantaneous complete severance of a primary or steam pipe.

The large breaks are very unlikely to occur, and have been assigned probabilities in the well-krown Rasmussen report [5].

The smaller breaks which have been analyzed herein will be discussed separately.

The small loss of coolant accident transient analyzed results from a sequence

-2 of events which can be assigned a probability of 2 x 10 per vessel per year of operation.

The small steam break transient chosen for analysis is actually one of three scenarios which could occur, and represents the most severe transient of the three, that of a stuck-open safety valve..The probability of this occurring is about 10 per vessel per year.

The other two scenarios, either a stuck open steam dump valve or stuck open atmospheric dump valve, are somewhat more pro-

bable, but result in less severe transients.

The Rancho-Seco transient was caused by a control system fai lure that resul'ted in an excessive feedwater addition transient in the primary system.

This transient produced a significant reactor vessel thermal shock;

however, a simi-lar excessive feedwater addi tion event in a Westinghouse PWR would not pro-duce a primary system transient of significance to reactor vessel integrity.

This is due to the fact that the Westinghouse PWR has a large secondary system thermal interia which would minimize the primary 'system cooldown rate.

The primary system pressure and temperature transient that occurred in Rancho-Seco would be similar to the primary system transient that could occur in a Westing-house PWR due to a low probability small steam break, Also, the repressurization in the Rancho-Seco transient

(~2,100 psi) is not possible for the Turkey Point Unit 4 plant since the capability of the high pressure safety injection (HPSI) pumps is much less (1,400 psi) than the applied values.

B.

Fracture Mechanics Analysis 1.

General Methods and Acce tance Criteria A detailed discussion of the basis for the thermal, stress, and fracture analyses completed for the Turkey Point 4 reactor vessel has been previously provided in Reference [Ij, so it will not be repeated here.

Although the acceptance criteria were also provided in Reference [1],

it is deemed worthy to provide this,information below.

The results of the fracture mechanics analyses of postulated longitudinal and circumferential flaws are presented in Section IV.C below in terms of the maximum number of calendar years the reactor vessel will conform to the following criteria:

t

.. 1; Minimum 'critical fla>I depth for crack initiation is greater

~

than 1;0 inch, or 2.

Crack arrest occurs within 75 percent of the'esse1 wa11 thickness.

I C

~

I t

I The initiation criterion is based on the ultrasonic inspection limitations, and the arrest criterion is'set to be consistent with Appendix A of Section XI, ASME Code.

. It should be noted that the acceptable vessel lifetimes given in the report are bpsed on these acceptance criteria, and therefore, the def'ined acceptable 3ffetime does not indicate catastrophic failure of the vesse1.-

~

~

2.

Warm PrestressilA The results obtained for the Turkey Point Unit 4 reactor vessel made use of the principle of warm prestressing to demonstrate integrity for the remaining design lifetime for some of the limiting transients provided in the following section.

The technical basis for the use ot warm prestressing in demonstrating vessel integrity has been given in detail in References

[1j and I 6J, so it will not be repeat d here.

The application oi warm prestresssng to some of the transi ents results in an excellent behavior with regard to vessel integrity because warm Pr estressing occurs very early in each transient.

Specific details are provided in the following section.

C.

Results of Analyses The minimum number oi'dditional years of vessel operation without violation of acceptance criteria for the above transients relative to v ssel thermal shock consideration were provided in Reference I 1j for the Turkey Point Unit 4 plant.

however, further analyses have been recently completed using plant specific property information and the benefits of warm prestressing for the large LOCA, large steamline

break, and Rancho Seco transients.

These new results are provided in Tab'le 6 along with the

.previously submitted small LOCA transient result.

Additional information with respect to fundamental inputs to each analysis are also provided in Table 6.

Corresponding plots giving specific results of warm prestressing are shown in Figures 12, 13, and 14 for the applicable limiting.transients where this benefit was used.

The updated results in Table 6 are con~~idered to be more appropriate to the Turkey Point 4 reactor vessel, and therefore should replace the results previously submitted in Reference L1j.

The results obtained on all four of the limiting transients applicable to the Turkey Point 4 design demonstrated that vessel integrity would be maintained throughout the design lifetime of the plant.

An end-of-life value would also be expected for the Rancho Seco transient, which is not applicable to the Turkey Point 4 plant, if plant specific information would be taken into account.

D.

Summary Detailed analyses have been carried out for a number of postulated transients which result in" thermal shock to the reactor pressure vessel.

The results of these ana'iyses are summarized in Table 6 and show that reactor vessel integrity will be maintained for the Turkey Point Unit 4 vessel throughout its design iitetime.

TABLE:

~TURKEY POINT UNIT 84 MINIMUM NUMBER OF ADDITIONALYEARS OF VESSEL OPERATION WITHOUT VIOLATION OF ACCEPTANCE CRITERIA FOR THERMA SHOCK TRANSIENTS Transient Large LOCA Small LOCA Minimum Number of Add'1 Years*

33 Vessel N]d

~Geometr Location Plant Plant Specific

-Specific Remarks

'aterial Pro erties Transient Characteristic Plant Specific Limiting Generic 3

Loop 2" Break - No Mixing Case Plant Speci-fic as given in Table 5, Item (c)

Plant Speci-fic as given in Section III.B Fluence Vr cad Benefit of Profile Curve Ilare-Prestressii Plant R.G.

1.99 Specific (See Figure 1')S (See I-figure 13 Large Steam Break Small Steam Break 33 33 Plant Speci-fic as given in Table 5, Item (c)

Plant. Specific

.as given in Section III.B Plant Specific Generic (See Figure 14

-YES NO Rancho-Seco Transient+"

25 0

Generic.

Longitudinal Generic Category 3-Loop Weld in Peak in combinationl Fluence Lo-with properties cation given in Section III.B.

Generic 3-Loop YES

• Accumulated EFPY as pf 10/31/81 is 5.66 years.

The values shown reflect the number of years before donservative acceptance criteria L

are exceeded (does not indicate actual vessel failure) with the use of a 0.8 plant usage factor (capaci ty factor).

• "This transient is not applicable to the Turkey Point 4 design.

However, it was arbitrarily applied.

~.'40 HuM~ C~zzlz

~.0 ~

~.00

0. ~ 0 I

I I/

I Ps,4wrmairl

/~i7i~c'M Cif/JVd

~.0 ~

I I

I I

I I

I I

I I

cf 4 ~i~r~W~

7>ltl~Pa 4ieVd

~. ~ 4

~.00 IOO ~. ~ 0 000 ~.00

)0 ~0.00 F 000.00 4 F 00.0 ~

'f tRE CAtT ICOSI.

CAACK SIZES Figure 12 Warm Prestressing Plot for Unit 4 Lar ge LOCA Transient

Figure 13 Warm Prestressing Plot for Unit 4 Small LOCA Transient

4~..i (000.00 0000.00

%000.00

%000.00 0000.00 T 1IIE CRLTLCRL CRRCK SIZES Figure 14 Warm Prestress]ng Plot for Unit 4 Large Steamlfne Break Transient

\\

V-OTHER CONSIDERATIONS A.

Uncertainties and Hargins Although a quantified assessment of the sensitivity of fracture analyses recently completed for the Turkey Point Unit 4 beltline weld has not been evaluated at this time relative to uncertainties in input values (e.g.

initial crack size, copper content, fluence, and initial RTNDT), a detailed discussion of the conservatisms inherent in these analyses is provided in Reference [1].

Conservatisms due to the generic analytical approach and margins associated with transient development, fluence calculations, and stress and fracture mechanics analyses are outlined in the December 1981 report on reactor vessel integrity [1].

To further quantify the margins inherent in the analyses for Turkey Point Unit 4, the following plant specific conservatisms are noted:

o The large and small LOCA transients development assumed a refueling water storage tank temperature of approximately 40'F.

Considering the warm climate ot'he plant location, this water temperature value is considered to be quite conservatives o

The small steam break transient (and Rancho Seco transient) have included repressurization values

(~2100 psi) well beyond the capability of the Turkey Point Unit 4 high pressure safety injection (HPSI) pumps (1,400 psi).

o Forms 'of a low leakage core pattern have been in place since fuel cycle No.

5 (Aug.

1978) and this information has not been utilized in the evaluation of vessel integrity.

True dosage to the reactor vessel is expected to be somewhat lower than assumed in the analysis and the results obtained are conservative.

O B.

Remedial Actions

~

e The Westinghouse Owner's Group report Ll] provides a qualitative assessment of the feasibility and/or usefulness of the following remedial actions that could be used to resolve vessel integrity concerns, including a reduction in rate of further vessel embrittlement:

(I)

Increasing the ECC water temperature via heating of the refueling water storage tank (2)

Limiting auxiliary feedwater flow (3f Design of control systems to mitigate challenges to reactor

'vessel integrity.

(4)

Core modifications to reduce further neutrdn radiation damage at the beltline

~ (5)

Recovery of material toughness by in-place annealing of the reactor vessel Work is being carried out to evaluate the current low leakage core pattern and to assess more exact core configurations for reducing the rate of flux as much as possible for Turkey Point Unit 4.

C.

Action Plan Since integrity has been demonstrated for the design lifetime of Turkey Point Unit 4, there is no need for an action plan.

37

0 0

VI.

REFERENCES

[1].

Meyer, T. A, "Summary Report on"Reactor Vessel Integrity for Westinghouse Operating Plants",

WCAP-10019, December 1981.

[2].

Norris, E. B., "Reactor Vessel Material Surveillance Program for Turkey Point Unit No. 4, Analysis of Capsule T," Final Report, Southwest Research Institute Project 02-4221, June 14, 1976.

[3].

Yanico, S.E., et. al, "Analysis of Capsule T from The Florida Power and Light Company Turkey Point Unit No.

3 Reactor Vessel Radiation Surveillance Program",

December 1975.

[4].

Meeuwis, 0. et al, "Fracture Mechanics Evaluation of the Florida Power'nd Light Company Turkey Point Unit No.

4 Reactor Pressure Vessel Subjected to Postulated Accident Transients",

WCAP-8945, May 1977.

[5].

"An Assessment of Accident Risks in U.S.

Commercial Nuclear Power Plants",

WASH-1400, August 1974.

[6].

McGowan, J. J., "Application of Warm Prestressing Effects to Fracture Mechanics Analyses of Nuclear Reactor Vessels During Severe Thermal Shock," Journal of Nuclear Engineering and Design, V51, No. 3, Feb.

1979.

STATE OF FLORIDA

)

)

COUNTY OF DADE

)

ss

~

Robert E. Uhri, being first duly sworn, deposes and says:

That he is

'Vice President Light Company, the here>.n; of Florida Power 6

That he has executed the foregoing document; that the state-ments made in this said

document, are true and correct, to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Robert E. Uhrig Subscribed and sworn to before me this day of NOTARY PUBLIC, i and for the County of Dade, State of Florida Notary Public, State of Florida at f.argo h1y Commission Expires October 30, 1983 COmmiSSiOn eXpireS:

Bonded thru Maynard Bonding Agency

p t

l.

A