ML20137P791

Withdraws Rev 2 to License Change Application 99 for License NPF-1.Changes Identified in Encl Capsule X Rept Will Be Incorporated in New License Change application.WCAP-10861, Analysis of Capsule X from Portland... Encl

ML20137P791
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	01/23/1986
From:	Withers B PORTLAND GENERAL ELECTRIC CO.
To:	Varga S Office of Nuclear Reactor Regulation
Shared Package
ML20137P794	List: ML20137P791 ML20137P807
References
TAC-59991, TAC-60677, TAC-63527, NUDOCS 8602050345
Download: ML20137P791 (13)

Text

-

y I a

s

- ammme

~amm Pbrden1d W M W Bad D Wehers Vre Presxet January 23, 1986 Trojan Nuclear Plant Docket 50-344 License NPF-1 Director of Nuclear Reactor Regulation ATTN: Mr. Steven A. Varga Director, PWR-A Project Directorate No. 3 U.S. Nuclear Regulatory Commission Washington DC 20555

Dear Sir:

Troian Reactor Vessel Radiation Surveillance Program In accordance with Section III.A of Appendix H to 10 CFR 50, attached is the summary technical report for radiation specimen Capsule I. Capsule I was withdrawn from the Trojan reactor vessel at 4.28 EPPY and was tested in accordance with 10 CFR 50, Appendices G and H and ASTM Specification E185-82.

The test results demonstrated the reactor vessel materials are less sensitive to irradiation than predicted by Regulatory Guide 1.99, Revision 1.

Please note Section 7 and Appendix A of the report identify changes to the surveillance capsule removal schedule and the heatup and cooldown limit curves based on identified changes in RT NDT. The removal schedule and heatup and cooldown curves are contained in Trojan Technical Specifications and are currently the subject of License Change Application (LCA) 99, Revision 2 which you are reviewing. Therefore, we would like to withdraw LCA 99, Revision 2 so that we may incorporate the changes identified in the Capsule I report. A new LCA will then be submitted.

Sincerely,

.=_ /

Bart D. Withers Vice President Nuclear Attachment c: Mr. Lynn Frank, Director State of Oregon Department of Energy, w/o attachment Mr. John B. Martin ,g[ /JA #l Regional Administrator, Region V U.S. Nucler.r Regulatory Commission, w/o attach  ;

Ol 8602050345 060123 ,,fI PDR ADOCK 05000344 p PDR a sw S1rren Dw n LM Ovr.n 9GA J

~

l M

APPENDIX A HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION A-1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTnor (reference nil-ductility temperature). The most limiting RTnor of the materialin the core region of the reactor vesselis determined by using the preservice reactor vessel material properties and estimating the radiation-induced ARTng7. RTyg7 is designated as the higher of either the drop weight nil-ductility transition temperature (NOTT) or the temperature at which the material exhibits at least 50 ft Ib of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.

RT yg7 increases as the material is exposed to fast-neutron raoiation. Therefore, to find

, the most limiting RTuo7 at any time period in the reactor's life, ARTNor due to the radia-tion exposure associated with that time period must be added to the original unirradiated RTuo7. The extent of the shift in ATyg7 is enhanced by certain chemical elements (such as copper and phosphorus) present in reactor vessel steels. Design curves which show the effect of fluence and copper and phosphorus contents on ARTyg7 for reactor vessel steels are shown in Figure A-1.

Given the copper and phosphorus contents of the most limiting material, the radiation-induced ART Nor can be estimated from Figure A-1. Fast neutron fluence (E > 1 MeV) at the vessel inner surface, the % T (wall thickness), and % T (wall thickness) vessel locations are given as a function of full-power service life in Figure A-2. The data for all ferritic materials in the reactor coolant pressure boundary are examined to ensure that no other component will be limiting with respect to RTyg7 I

. A-2. FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture toughness properties of the Trojan reactor vessel materials are presented in Table A 1. The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC A-1

. g Regulatory Standard Review Plan.111 The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the Trojan Vessel Material Surveillance Program. .

A-3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K i , for the combined ther-mal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Kig, for the metal temperature at that time. Kig is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code.l21 The K ig curve is given by the equation:

Kig = 26.78 + 1.223 exp [0.0145 (T- RTno7 + 160)] (A-1) where:

K ig = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTuor. ,

Therefore, the governing equation for the heatup-cooldown analysis is defined in Ap-pendix G of the ASME Codel21 as follows:

CKiu + K n5 Kig (A2) where:

K,y = stress intensity factor caused by membrane (pressure) stress l Kn = stress intensity factor caused by the thermal gradients J

Kig = function of temperature relative to the RTno7 of the material ,

C = 2.0 for Level A and Level B service limits ,

C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical ,

i A2

f

.~ At any time during the heatup or cooldown transient, K ig is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTgo7, and the

- . reference fracture toughness curve. The thermal stresses resulting from temperature i

gradients through the vessel wall are calculated and then the corresponding (thermal)

! stress intensity factors, Kn, for the reference flaw are computed. From Equation A-2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

i For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wal! During cooldown, the controUing location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, ,

which increase with increasing cooldown rates. Allowable pressure-temperature relations

are generated for both steady-state and finite cooldown rate situations. From these rela-tions, composite limit curves are constructed for each cooloown rate of interest.

I.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature,

~

whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

During cooldown, the % T vessel location is at a higher temperature than the fluid adja-

} cent to the vessel ID. This condition, of course, is not true for the steady state situation.

It follows that, at any given reactor coolant temperature, the AT developed during cooldown i results in a higher value of Kig at the % T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K,g ex-

ceeds Kn, the calculated allowable pressure during cooldown will be greater than the steady-state value.

1 The above procedures are needed because there is no direct control on temperature at the % T location and, therefore, allowable pressures may unknowingly be violated J

if the rate of cooling is decreased at various intervals along a cooldown ramp. The use i . of the composite curve eliminates this problem and insures conservative operation of l , the system for the entire cooldown period.

i i

, Three separate calculations are required to determine the limit curves for finite heatup j rates. As is done in the cooldown analysis, allowable pressure temperature relationships are developed for r,teady state conditions as well as finite heatup rate conditions assuming 1

! A-3 9

e-. -- ,- , . - -

---

,r - , - ~ ow----, ---,,,,.n- .,-,.,.,.,n,, , , , - - - , , , - , , , , ,,c----,,,-.v.,,-,--

the presence of a % T defect at the inside of the vessel wall. The thermal gradients ~,

during heatup produce compressive stresses at the inside of the wall that alleviate the ,

tensile stresses produced by internal pressure. The metal temperature at the crack tip ..

lags the coolant temperature; therefore, the K,g for the % T crack during heatup is lower

, than the K ig for the % T crack during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower Kig 's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the % T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady state end finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a % T deep outside surface flaw is assum-ed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are depen- '

dent on both the rate of heatup and the time (or coolant temperature) along the heatup ~

ramp. Since the thermal stresses at the outside are tensile and increase with increasing '

heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a com-posite curve based on a point-by-point comparison of the steady state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under ccnsideration. The use of the composite

, curve is necessary to set conservative heatup limitations because it is possible for con-ditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Then, composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on Figures A 3 and A-4. .

Finally, the new 10CFR50l31 rule which addresses the metal temperature of the closure ,

head flange and vessel flange regicns is considered. This 10CFR50 rule states that the ,

metal temperature of the closure flange regions must exceed the material RT Norby at least 120*F for normal operation when the pressure exceeds 20 percent of the preser-vice hydrostatic test pressure (621 psig for Trojan). Table A-1 indicates that the limiting A-4 1

I

_ .= ___ _ - _ _ _ - - - . . _ _ _ - . _ .-

." RT nor of 20*F occurs in the closure head flange of Trojan, and the minimum allowable

• temperature of this region is 140*F at pressures greater than 621 psig.

A-4. HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System j

have been calculated using methods discussed in Se tion A-3. The denvation of the limit j curves is presented in the NRC Regulatory Standard Review Plan.Pl i

Transition temperature shifts occurring in the pressure vessel matenals due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program. Charpy test specimens from Capsule X indicate that the surveillance weld metal and core region lower shell plate heat No. C5583-1 exhibited shifts in RT nor of 50*F l 2 and 95'F, respectively. These shifts at a fluence of 1.77 x 10'9n /cm are well within the appropriate design curve (Figure A-1) prediction. As a result, the heatup and cooldown I

curves are based on the ARTno7 given in Figure A 1 for the most limiting beltline material which is the lower shell plate heat no. B9883-1. The resultant heatup and cooldown limit curves for normal operation of the reactor vessel are presented in Figures A 3 and A-4 i

and represent an operational time period of 10 EFPY. These limit curves are impacted by the new 10CFR50 rule.

i' Allowable combinations of temperature and pressure for specific temperature change i rates are below and to tt's right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure temperature combinations

! are to the right of the criticality limit line shown in Figure A 3. This is in addition to other criteria which must be met before the reactor is made critical.

The leak test limit curve shown in Figure A 3 represents minimum temperature require-ments at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of References 2 and 4.

Figures A 3 and A-4 define limits for ensunng prevention of nonductile failure.

O 2

I i

A5 t

TABLE A-1 TROJAN REACTOR VESSEL TOUGHNESS TABLE Minimum 50 ft Ib/ Average Upper 35 mil Temp Shelf Energy Cu P NOT MWD NMWD RT wor MWD NMWD Component Heat No. Grade (%) (%) (* F) (* F) (* F) (* F) (ft Ib) (ft Ib)

Closure Head Dome B0048-2 A5338.cl 1 - -

- 20 8 28[al - 20 148 96[a]

Closure Head Torus C6096-1 A5338.cl 1 - - 0 10 30[a] O 149 5 97[a]

Closure Head Torus B0042-2 A5338.cl 1 - -

- 20 13 33[a] - 20 161.5 105[al Closure Head Flange 4436-V1 A508.cl 2 - -

20 - 20 O[al 20 151 98(a)

Vessel Flange 4437 V2 A508.cl 2 - - 10 17 37[a] 10 132 5 86[a]

inlet Nozzle 9-7987-1 A508.cl 2 - -

- 80 10 30[al - 30 105 68(a) inlet Nozzle 9-7260-1 A508.cl 2 - - 0 48 68[a] 8 111 72[a]

- 120 45 65I 'I 5 95 62[aj D Inlet Nozzle 9-7235-1 At08.cl 2 - -

O Inlet Nozzle 9-7208-1 A508.cl 2 - -

- 20 - 25 - 5[a] - 20 115 5 75[a]

- 30 0 2O[al - 30 126 82[aj A508.cl 2 Outlet Nozzle 9 7315-1 - -

Outlet Nozzle 9 7251-1 A508.cl 2 - -

- 75 - 25 - 5[a] - 65 120 78[a]

Outlet Nozzle 9-7301-1 A508.cl 2 - -

- 100 20 40[al - 20 126 82IdI Outlet Nozzle 9-7241-1 A508.cl 2 - - - 130 45 65[a] 5 101 66Idl Nozzle Shen C5570-2 A5338.cl 1 0 11 0 014 0 20 40[a] O 127 5 83IdI Nozzle Shen C5571-1 A5338.cl 1 0 15 0 013 0 -6 14'I l O 118 5 77Idl Nozzle Shell C6529-2 A5338.cl 1 0 14 0 010 - 20 10 30[a) - 20 131 85ldI intermediate Shell C5582-1 A533B.cl 1 0 12 0 009 - 10 - 10 60 0 128 5 101 Intermediate Shell C5587-1 A5338.cl 1 0 15 0 014 to 38 40 10 113 117 a) Estimated MWD - Lorgtudinal ants of Charpy specwnen onented in the map working direction NMWD - Longitudinal amis el Charpy specimen onented normal to the map working direction

• e ,

9

• 9 g g ,

• s . . , ,

e - .

. , , 9 TABLE A-1(cont)

TROJAN REACTOR VESSEL TOUGHNESS TABLE Minimum 50 ft Ib/ Average Upper 35 mil Temp Shelf Energy .

Cu P NDT MWD NMWD RT uoy MWD NMWD Component Heat No. Grade (%) (%) (* F) (*F) (*F) (* F) (ft Ib) (ft Ib) _

l A5338 cl 1 0 16 0 012 - 10 30 70 10 121 99

! Lower Sheu B9883-1 C5583-1 A533d.cl 1 0 15 0 011 0 18 60 0 113 84 Lower Shell l

- 20 15 35lal - 20 159 10'[al

> Bottom Head Torus C6123 3 A5338.cl 1 - -

4 16 IdI - 20 143 5 93 'I N Bottom Head Torus C5823-1 A5338.cl 1 - - - 20 Bottom Head Dome B0018-1 A5338.cl 1 - - - 50 - 25 - 5(a] - 50 177 115[a]

I 0 06 0 019 - 20 - 35 - 20 -

1025 weld Metal

- 60 -

- 10 - 60 - 110 Weld HAZ - -

a) Estimated uwD - Longitudinal amis et Charpy specwnen onented m the maior working carecteon N'.tWD - Long.tuomal amis et Cnarpy specimen oriented normat to the ma,or working direct on i

i i

\

.1R TNDT

• l40 + 1000 (% Cu 0 08) + 5000 (% P 0 008)l lI/1019l I/2 300 -

UPPEN L s

C 200 -

O*

sa /

b3 "J $ 100 -

e o5

> G-6 *c 0 35 0 30 0 25 0 20 % Cu 015% Cu 010% Cu A 4 40 -

%P=0012  % Cu - 0 08 O  % P - 0 008 e e PLATE METAL C5583-1 33 _

, 9 A WELD METAL 8 ,

E 20 I I IIIll l I I l I Ili l I I .

2m 1017 4 6 8 jo18 2 4 6 8 1019 2 4 6 j

! NEUTHON FLUENCE iE >1 Malin/cm 2; FIGURE A-1. PREDICTED ADJUSTMENT OF REFERENCE TEMPERATURE AS A FUNCTION OF FLUNCE, COPPER, AND PHOSPHORUS CONTENTS.

E ,

C0 0' .

0' 1

1 A1 x

. F x x R9 5 5 U8 T 6 T3 S2 %1 %3

- ' ~

2 " 2

, 3 9

- _ 0 o, _ 3 ii ;I

- - 8 2

L L

U

- - 6 2

F F

O I  ;, '

-' ' , jlI Iii' 4

2 N

O I

T

- C J

i I 1

1-l p ,

ii

! i  !

2 2

0 2

)

s r

N U

F A

S a

e A Y )

. ,l i

- ;1i ;i  !

8 1

s e

w V

e

' M o

P r

l 1

6 lu

' -  ! i 1 F e

y v )

t i

EY

. 4 c

e (P EF

'y#II

. t f 1 i L

(

CE E N(

EE eiM 2 iL f

.a l ' 1 i I  ; I ,

1 UF I LL E F C E 3 NC I

- 'i' ,I '-  !

V

' OIV 1 t

/

i

, t S R R T E

,I !1 s '

' I I!'

8 U S E

1. / NR

/

/ / 6 T E SW

/ AO FP

_. /

V

/

! i!II i;' i  :  ! 4 2-

. / E A

• Ii I 1~ 1j* I .I , ' -  !

2 R

- _ U G

0 g

8

_ r_ 7 O

I F

. 2 i 1 1

- 0 0 0 0 1 1 1 1 r ,

o s

pEuNv d5a[ zo$3z l

j i .!

., GRATERIAL .40PERTY BASIS: _.

COer/ ROLLING MATERIAL : R.V. LOWER SHELL COPPER CONTENT  : 0.16 WT%

PHOSPHORUS CONTENT : 0.012 WT% .,

INITIAL RTer  : 10*F

• RT , AFTER 10 EFPY  : %T.111*F 6

%T. 55'F ..

j iiiiiiiliiiiiiliiiiiiiiiiiIiiiillllilliiiiiil j jT CURVE APPLICABLE FOR HEATUP RATES UP TO 60*F/HR FOR  ;

THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS 7

- OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. _L

!Illi  !!Il Li TT T T ' i i i ltil I II 'l ll 1 I l  !!!!!!!

j ii, i ji li t fi t

-/ lljjj j l l LEAK TEST LIMITgj /j [ jjjj l ll 'QI fl '

'l 'llil ll l f 1 -

J ll1 l

_# 0 ll14 ll' I fl /I  !

! lil!

!? lll l I iiiuii!

l HEATUP RATES UP I u l j jjl 1 ll l l

$ -l l l ,l qj j j j j w TO 60*F/HR. 2 E Il I ui 't Y / / I I I! l -

E ll 'I L I Q" V , l'l / I IL i ll E

ill +

l I

/! /! ll i

I lIllll '

lll fj f:

n '

liliii!jj c UNACCEPTABLE r ACCEPTABLE

$ !III OPERATING AREA I /r !J OPERATING AREA N y I!ll ,rit r  !  ! ' l/! ((! Ii! I!!! !'iI!!

5 Ill 11 l I ll /!

! I Il  ! Il  ! ll!!!

E

' ill II 1 li/ l! !! Il li 'l li lil 111ll ll /

I I l i Illi 11!!!!!!!

l i fl c  ! Ililliti t '

g g a N CRITICALITY LIMIT BASED ~

4 .4 .i..;

l:

-l '

l ON INSERVl0E HYDROSTATIO -

] II!

k _

. I  !  !!  !' TEST TEMPERATURE (256'F) _

l l l

~

l FOR THE SERVICE PERIOD ! ,

} l l j UP TO 10 EFPY { ,

ll lli: 11 i I 'l  !' 'llll Ill I!lllI'T .-

il I!!! i i <

!  ! IlII !I!!!!!!!

il Ill :I I I  ! ll 1 l.Ill!:, i 00 100 0 200 0 300 0 400 0 500 0 INDICATED TEMPERATURE (*F)

FIGURE A 3. TROJAN REACTOR COOLAP4T SYSTEM '

HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 10 EFPY

O CATERIAL PROPERTY BASIS:

CONTROLLING MATERIAL : R.V. LOWER SHELL COPPER CONTENT 0.16 WT%

PHOSPHORUS CONTENT 0 012 WT%

INITIAL RT sar 10*F RT yor AFTER 10 EFPY '. . T. 1 1 1

• F

=

4 T. 55'F O.

3000 0 , ,

.i CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100*F/HR FOR ~+-~

THE SERVICE PERIOD UP TO 10 EFPY AND CONTAINS MARGINS OF _

I*l I!I 10*F AND 60 PSGI FOR POSSIBLE INSTRUMENT ERRORS II!If !I iI ! ili !It iii!

i i't 4 :ii, e .

Illil!lll  !!'!ile !i  ! ! lilil l,l'ilii ,;i:

'!i .ill! !

llllillij

!llii!!

1 i i>l,;;  ;

'!.lii ill i llill I!!!  !!..l  !!  !!!! ._t 1 it'Ili; i,  ! !!

illlii li lil I li lli'!i i ll!!iill lii! 'ill

!I!!!!!I lill!llll I

lillllll l! !!! I/!' !I!!!!II I

I ; ! ' ' ! ! LL lil!!!/!!t lllli't i I: : lili k  ! N'!!'lll lililll!l l!;!i l ii li;il it ll  !!!

E i.

un i!ijl!il! llllIlil! llllIlIll i!iiiiii> l t CC m

3

,ilil..I.lll

..llli.lllll . ..

Il .

if .li {liIlilli.  ! ll,!.ii

!!l!!il lIIIlIl{I l ' l LLili

$ Ililiilll. l If I!. !!

'I!

I UNACCEPTABLE l! I!IIII  ! II!II' III!!i! I

$ !II OPERATING AREA l /lII!i lll[!!!II IfIf!d 1 g 4

ll!ll!!;l lif!!!il j_/!lll!i; l/!!!!t i !!!!! !I!! l!!(( Ilk o  !!lll l

. =

l

=4  : =. ~.

lil lilli .llli!ill 5 'IiIii!II i!! l ll! i iliI Ii!  !! liill1i ilI(jil't 3 I I'!!Ii!I III l llj lill ACCEPTABLE 1000 0 lll;lll j

!!!IIIIII ll! , lIf t

OPERATING AREA Ill 1 llllllIll [I!I I 1

I l II lll I! I  !  ![i![

!llllll l fl llf .

ll  ;

l 1 .

Il,f! _!!t .[ _' .' ! ..

) f COOLDOWN RATES ~ :  ;  : -

{

=

l.,

' F, H R

!! II!ff!{I!

~

0 I '

l { f .fII- I' . iIl l 20 m -

l l lii;. i;jiti;  ;

40 M~ j: i'

j[j

, 60 '

!,.d_ fd_1_d

. 100.- W i~ l l  : -l l l ll l ll llll l s

. . _ gg

,lll

,q l lll' l l l f l lll ff I l! f fl.  ! lIll 00 100 0 200 0 300 0 400 0 500 0 INDICATED TEMPERATURE (*F)

VIGURE A 4. TROJAN REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE FOR THE FIRST 10 EFPY A.11

I .

1 i .

APPENDIX A -

! REFERENCES l

l l 1. " Fracture Toughness Requirements," Branch Technical Position MTEB 5 2, Chapter

{

5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG 0800,1981.

I

2. ASME Bciler and Pressure Vessel Code, Section Ill, Division 1 - Appendices, " Rules l for Construction of Nuclear Vessels," Appendix G," Protection Against Nonductile Failure," pp. 559-564,1983 Edition, American Society of Mechanical Engineers, New York,1983.

]

l 3. Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Require-

• i ments," U. S. Nuclear Regulatory Commission, Washington, D. C., Amended May 17, 1983 (48 Federal Register 24010). ,

4." Pressure-Temperature Limits," Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG 0800,1981.

l i

e i

9 e

F A 12 i