Text

Forwards Response to NRC 930129 Request for Addl Info to Complete Review of Util 920701 Response to GL 92-01,Rev 1, Reactor Vessel Structural Integrity
	ML20012G458
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	02/25/1993
From:	Beckman W; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	GL-92-01, GL-92-1, NUDOCS 9303020499
	Download: ML20012G458 (10)
	v • d • e

{{#Wiki_filter:e r* s ConsumEIS POWER 3 William L Beckman { l'lant Manager POWERING MICHIGAN 5 PROGRESS Big Rock Point Nuclear Plant,10269 US 31 North, Charlevoix, MI 49720 i i February 25, 1993 4 NUCLEAR REGULATORY COMMISSION t DOCUMENT CONTROL DESK WASHINGTON, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT - RESPONSE TO A REQUEST FOR ADDITIONAL INFORMATION IN REGARDS TO GENERIC LETTER 92-01, REVISION 1, " REACTOR VESSEL STRUCTURAL INTEGRITY; 10 CFR 50.54(f)". By letter dated January 29, 1993, you requested additional information to complete your review of Consumers Power's response dated July 1, 1992, to Generic Letter 92-01, Revision 1, " Reactor Vessel Structural Integrity". The following information is provided on Attachment I as requested. j u,n o y (? V "f William V Befkman (Signed) William L Beckman Plant Manager cc: Administrator, Region III, USNRC NRC Resident Inspector - Big Rock Point Attachments 010145 bg: 9303020499 930225 PDR ADOCK 05000155 j P PDR A csg gygggy(gymyy

r s f i ATTACNMENT CONSUMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET 50-155 RESPONSE TO RAI TO GENERIC LETTER 92-01, REVISION 1, " REACTOR VESSEL STRUCTURAL INTEGRITY - 10 CFR 50.54(f)" 9 i b e I I f

r i ^ RESPONSE TO RAI - GENERIC LETTER 92-01, REVISION 1 1 Reauest Please provide an analysis to demonstrate that the vessel's beltline plates and j l welds will have Charpy upper shelf energies greater than 50 ft-lbs at the end of its licensed life. This analysis should consider the effect of all surveillance t data and should follow the method documented in Regulatory Guide 1.99, Revision 2.

Response

i i The attached analysis based on BRP surveillance data indicates E0L U.S.E. to be approximately 60 ft-lbs for base plate material and 52.5 ft-lbs for the weld i metal. This analysis is in accordance with paragraph C.2.2 of Regulatory Guide 1.99, Rev. 2. i Reauest Provide the orientation of the axis and notch of the Charpy specimens relative to 3 the rolling direction and surface of the plate. ]

Response

Per GECR-4442, " Reactor Pressure Vessel Material Surveillance Program at the l Consumers Power Company Big Rock Point Nuclear Plant", Charpy and tensile ^ specimens were made from the base metal used to fabricate the vessel. These specimens were made from flat slabs cut parallel to and one-quarter plate thickness from both of the plate surfaces and with their longest dimension transverse to the plate rolling direction. i A test weld representing a vessel longitudinal joint was fabricated from vessel base material. Charpy and tensile specimens representing the weld heat-affected zone (HAZ) and the weld metal were made from the test weld material. The weld Charpy specimens were made transverse to the weld direction; thus, only 1 the central section of the specimen was necessarily composed of weld-deposited metal in the relatively narrow type of weld used for the vessel. Specimens were i taken throughout the weld section to a distance of 1 inch from the weld root. Their long axes were parallel to the plate surface and normal to the weld length. t The Charpy notch was made parallel to the plate surface. l j Reauest J l Provide the heat of materials used by Lukens Steel Company to fabricate the d beltline plates. 3 j

Response

All four (4) beltline plates were poured from a single heat of steel, Lukens heat

19246. The plates are identified as #19246 slab 1, #19246 slab 2, #19246 slab 3 and #19246 slab 4.

4 1,

s r a t I RESPONSE TO RAI - GENERIC LETTER 92-01, REVISION 1 2 t Reauest i i Provide the available weld information that demonstrates that the surveillance welds are representative of the actual beltline welds. L

Response

Combustion Engineering Inc., Dwg # E-201-794-8 (Vessel Forming and Welding) indicates combustion weld specification SAA-33A(3) was used in reactor vessel beltline welds. GECR-4442 for the BRP's Surveillance Program required welding of the surveillance material to be by the same Combustion Engineering Weld Specification SAA-33A(3). Welding material requirements were then the same for surveillance material as for the reactor vessel. t Reauest Provide whether the surveillance and beltline were fabricated using single or tandem electrode procedures.

Response

Combustion Engineering Weld Specification SAA-33A(3) required the vce of a single electrode. Reauest t Provide the unirradiated upper shelf energy for each surveillance plate and weld. { l

Response

Per NRL Memorandum Report 2027 dated August 1969, unirradiated base plate U.S.E. is 82 ft-lbs and unirradiated weld metal U.S.E. is 95 ft-lbs. l Reauest Provide the irradiated upper shelf energy and corresponding neutron fluence for i 3 each surveillance weld and plate that was tested after irradiation in the Big Rock Point reactor vessel. 1

Response

Irradiated U.S.E. data for capsules 119,127,122, and 124 is provided in Table 9.2 of NRL Memorandum Report 2027 and Table 5-1 of WCAP-9794 for Capsule 125. Both data tables are attached. See Table 1 of Consumers Power analysis for fluence / capsule identification. ) i

{ . cannew: BIG ROCK POINT NUCLEAR PLANT EA. GL-92-01-01 ,,r%M ENGINEERING ANALYSIS WORK SHEET Sheet l_of._fz Title Evaluation of the Upper Shelf Energy (USE) of the Big Rock Point Reactor Vessel (RV1 Base Metal & Weld Metals INITIATION AND REVIEW Initiated Rev Method Check (,/) Technically Rev'd Rev Alt Det Qual Description By Date Calc Rvw Test By Date O Griginal Issue (', M. J/j/tr / t BO-t' . 'hh1 _// / i e I i 1 l i z. n._. __w _,_ i . _. ] I. OBJECTIVE: i l \\ i s To determine the end of life (E0L) upper-shelf-energy; (USE)l. of..the } i reactor vessel base and weld metal. l y..., - !i- -.1 -- L.) _ II.

REFERENCES:

{ i j-- ~~ 1)~ 10CFR50 App.G. ~~ 3 I~ 1 ~ l

2) -- Reg. Guide 1.99 Rev.- 2, May 1988 j-4 L

=Li r 3) G.L. 92-01 Rev. 1, March 6, 1992i - 4) WCAP-9794, Analysis of Capsule 125 from the CPCo~BRP Nticle~a'r TI~ Plant RV Radiation Surveillance Program,-Sept.,-1980 L 3-:-H-l 5)~~~NRL Memorandum Report 2027, August 1969 i 6) GECR-4442, December l'963 - ~ L ~l~l i t ~ T *~ 7)- Letter to R.E. Kettner, CPCo from Fi AdHollenbach, Gen Electric +- I i i i ..Co., dated Feb 26, 1962 8) NRLMemorandumReport1638,Aug.l15,1965 ~ i 9) NRL Report 6349, Jan. 31,~1966 r 10) Nuclear Engineering and Design, Vol. II, No. 3, April 1970,- C.Z. Serpan and H.E. Watson, pp.'393-415 11) ' ASTM E185-82 ~~- - 4-1-. _)._, _._-.4. III ANALYSIS INPUTS: All ' inputs for this analysis are extracted from references 2, 4, 5, l__ l and 7 above. l i i 1 - -IV ASSUMPTIONS: -- A Hy y. 'i m a. I I J-i . 1._, _1__ N i ~~j 3-- 7 I i t l 1 i i BRP 090 02/01/88

W censumer wwer BIG ROCK POINT NUCLEAR PLANT EA GIc92-01-01 pewsmus ENGINEERING ANALYSIS CONTINUATION SHEET Sheet __2_of 6 an==ws russaEss new a 0 V ANALYSIS: ~ A. Method 1 The USE is derived from Figure 2 of-Reg.-Guide l.99 Rev 2 (Attachment 1) for both base metal ~and weld metal' as described in paragraph C.2.2'of thi~s~ame' document. This ~ -is done by establishing a line parallel to-the existing. upper bound lines based on_%.. decrease jn;USE as measured by surveillance specimens for each materiali(base & weld) of interest. To determine the USE at any fluence of-interest, simply identify _the % decreaselin USE on the ordinate which corresponds to the fluence value taken off the material specific parallel line as-drawn above..This - j percentage decrease in USE can.be compared to the unirradiated base line USE to identify the USE for that fluence;- j.~ '- La J l .._.R_ .m l i B. Surveillance Data Five (5) Surveillance Capsules.have be'enlevaluated.from the BRP_ RV to date. Two (2) capsules (capsule's 119 wall and 122 ~~ accelerated)' were evaluated in 1964Tand two',(2)'more (capsules ~ 127 wall and 124 accelerated) in:.1967 by the Naval Research 1 Laboratory (NRt.). Wall capsule 125 was evaluatediin'1980 (Ref. '~4) by Westinghouse Corp. in~ conjunction withTPRI1 Table 9.2 ~ ~of Ref. 5 (Attachment 2) identifies -values for ' full-shear - absorption energy (USE) of capsules 119, 122, 127:and 124. ~ Identification of these: values is in~ accordance ~with7the'~ ' definitions presented in paragraphs 4.17Jand-4.'18-of; ASTM E185 82 (Attachment 3). The USE for capsule 125 is indicated via ~ T ) i - capsules is presented below. -]--[USE f6Fallffive(5)[ ~ ~ '-- Table 5-5 'of Ref.~4-(Attachment ) ~. ~~ ~ - -+- J._ _._J__ ~ ^ surveillancecapsuleItSEData Capsule TYPE F X 10" ~ BASE

I6ecreasA~

WELD ** 'E Decrease ~~ USE (ft*1b) Base USE USE (ft'lb) Weld USE i I 80 ' 15.8 119 WALL .15 82 'O 127 WALL .71 73 11.0 ' 70 ! ' 26.3 \\ 125 WtLL 2.27 68 17.1' 5 I 74 i i ! 22.1 122I accel-2.3 '62 24 5 - I !57 I I i 40 : i 124 ACCEL 10'.7 70 14.5' ! 65 l I I I 31.'6 l i i

lhirradiated base metal USE = B2 f t*1bs

.-_...._...___.l i L ?.__

- thirradiated weld metal USE = 95 f t*lbs

..L w. i i BRP091 02/01/88 i

i o censuma power BIG ROCK POINT NUCLEAR PLANT EA. GIe9 2-01 ' ,wwrmans ENGINEERING ANALYSIS CONTINUATION SHEET Sheet 3 of 6 amaramaan consens Rev # 0 C. ~ Fluence Data ~~~ ~ The most recent fluence (f) data comes from capsule 125 '(Attachment ~4, Table 5-1) having a measured fluence of 2.27 X 109 nyt, E ) 1 MEV. The thermal history at the time capsule _ 125 was removed indicates 17,858,371.2 MwHrt ( E ^ ~~ ~ page 1,3,4 and 5). 1Therefore, j . + _. -. y. . - + Calc. I

2. 27 X 10 'nyt, D1MEV=1. 271X10 a nvt M'

2 t 17,858,371.2MwHrt MwHrC L__ _ _.;_a _1_ g. - Reference plant operating data for fe,b-1979,-ih Aktachment As of.12-31'-92, the'BRP RV has 35,494,707 MwHrt (Attachment 5, ; page 2){ The fluence as of this date is.~thenr ~ - --T .i L _J._. d _L . _, _.J _._ _. _ i I Calc.'2 I y.+.-+.-.-, r 1 ~12/31/92 H_H._q = i 1.271 X 1dialnVC, D 1M ~ ~ f = 3 5, 4 9 4,7 07 MwHr t .&._.{MwHrt_ _ _ _. . = 4. 51 X 10 ' nvt, D1MEV 0- - J.. ' - - - - 1 I i I i I i 4 j t

4. _1

.__ j i i_ b Note tha't Calc 2-is the fluence at the wall' capsule

position-

= and not at the Vessel _ID or the'l/4t po.sition. As of 12-31-92, BRP has only 89 months left on its license (to-the date of May 31, 2000) per License No. DPR-6 (Attachment 6). _ l Typically BRP operates on a twelve (12) month cycle with a planned refueling outage of sixty.(60)_ days.. Conservatively,. forty-five (45) day refueling outages; will be used to calculate E0L fluence at the wall capsule location with no outage-in the year 2000. Therefore, ~ ~ ~ Calc.~3 ~~ ~~ ~' t 4 Y OU O'. _. .1. w, _,. _ - _ 7 yrs

315 Days Outage Time Remaining

= - -- 1. Yr i l q _. -- P ~. -... _, _, _-4 4 BRP091 02/01/88 6

pe BIG ROCK POINT NUCLEAR PLANT EA Gr.,-92-01 powouns ENGINEERING ANALYSIS CONTINUATION SHEET Sheet 4 of 6 mammurs passetss Rev # 0 Calc. 4 ~ ~7 -~ 2708; Days, -7. Yrs.* 365 Days + 151 Days- + 2 Days. l _1 Yr (Year 2000). (Leap Yrs.) - Calc.5 .u--+u -a r I i . _ - ; 27 08 Days - 315 DayE - =i-2393 operating Days Remaining l-L..--- 1 i .- ~. -. - . _ u_ q_ .u 1 4 i Calc. 6 d ' -- -~ Hu 1 i 2393 Dayi *

N40Mwt = 13,7 83',6 8 0 MsHrt Remain ng ~T ^

1 Day _._. % ._ j _ i . ~ _. - - 4 i i i i i..Calt;' 7 -- ' - - - ~ ' ~~- HY HY-t .ii r ~ 34,494,7 07 MwHrt +- 13,7 83 ; 6 80 MwHrC ~=" 49,27 8,3 87 MwHrt@EOld ..1

g.. p.

.L.. (-_.. i ~ Calc.~ 8 T ~ ~ - ~ ~ " ~ ~ - - " " i

rT -d ' - ~ ~t --

i l i 22#U# f = 49,27 b[387 MwHrt:

1.271 X 10 :

- -F-MvHrC.-~l - +q 7 - - Sy 1-ygy. 6.263110" vt,gog g ygy3 capguy, _ n \\ = n -- i i i 4 -_.y Therefore, the maximum E0L; fluence ' t the' wall ! capsule locition forj a the BRP RV i s ~6.~263~ X~10" ~nvt T E31 MEV. 4 i;, ~~ ~1 '~ _ E.l_...! _ 1. 1 ~~J u l _ L, (Attachment 7 from General Electric Corp ~ calculated a lead factor for. the wall capsules to be 1.25:1 when compared to the Vessel Wall, (I.D.). Therefore, the EOL fluence at the Vessel ID is: 7-Calc.9 f "' 6.263 X 10nn yt, E> 1MEV E > 1 MEV, D = 5,011 x 10 nvt-, got f @ Vessel 1.D. ~ 1.25 i .-a Per paragraph C.I.1 Formula'(3) of Attachment 1, thef fluence thru the, vessel wall is given by the following function-J-

1. -

1 .2-- 7 (,.n,3.- .q. _.7 . ~ _. ~~~ F., = fluence! at; Vessel I.D. where: F, = fluence at 1odtion' X ~~ ~ ~ ~ 1 r = depth in-inches into the vessel - - * - ~ - wall from' the Vessel I.D. BRP091 02/01/88

mnrw BIG ROCK POINT NUCLEAR PLANT EA GL-92-01-01 * ,,wrass ENGINEERING ANALYSIS CONTINUATION SHEET sheet 5 og 6 marammavs p==sem Rev # 0 i e ~ Noting'that the BRP RV wall ~(with cladding) 'is 5.40625 inches' thick,~ ~ ~ -the distance x to the 1/4t location is - -H-l _ 2_ _. Calc.10 i 4 --- - + (.25)(5.40625 in.)~=,1.3515625 inches i-4-f-t p!- i 1 l i i i i I. = t I i 4._.,! _ _ _ t i i i l l 4 + i 2 TheI~E0i flueM e at'~h'e 1/4t locafich~iI T C ~_ ~ ~- ~~ ~~ t j _4_ 4 l i - _ Cal c.11 _.- L_ i j l l - _ ___ f. 7 ( 5. 011-X' 10'8) [ et a4:n ass 2si)_._J

y'_--

_, :.. nyt, - E '> -1 MEV T-i i i l 3 i 3.. nyt, E'> 1 MEV " f - 5.011i x 10, [e 22437s] .. f 15- 011: x -10-{.7229791)r -"4cM{ -bnvt,L Ea 1 MEV - i i i e i i f T 3.623" x 10 nyt,1 E'> 1 MEV ~ Y ~[ T } ~ ]E f l ~~ ~~ ~~" ~ - M = '(EOL; O 1/4 t location) m -M-l + i i l { 1 + .- ;... i --. i 1 .-._.,_.7__ f j i i i D. Paraaraoh C.2.2~ Analysis 7 4 1 1 --Since the measured USE from surveillance specimens must bei i l plotted parallel to the existing lines on figure-2 of ~ Attachment.1,_it.is determined _thatLcapsule 122_dataJs.'mosta _, limiting for both base metal andiweld metal. This is because; j ~ T ~the:% decrease in USE'is greatest-for these-~ capsule specimensi v i -~ Accordingly,ua.line is drawn. parallel _to..thelexisting upper I ]__ and! weld ! bound data for both base metal (24.4% decrease in.USE) n'vt,[E i ~(40% 'deUease in~ USE) ~at~a~ fliience"of 2.3T10' -. _ _.. _. ~ _ _ _ - _. _.. _.- 1 MEV. Refer to Figure 1 of-this-analysis. --i+A i 1 Per Figure 1 of this ' analysis (Figure 2 of Attachment'l),~ E0L - ~ ~ - USE of the base metal will-experience a 27.6% reduction and the-weld metal will experience. a 44.7% reduction. The final EOL i USE'S are give below l E0L USE @ f= 3.623_X.10 nyt, E>1MEV 0.1/4t BASE METAL: (1.0 .267) 82 - 60.1 ft*1bs i E0L USE.0.f= 3.623 X 10 .nyt, E>1MEV.'..1/4t 0 WELD METAL: (1.0 _.447) 97 =_52.5 ft*1bs ---...w.4 n- .._.es--. n. ..n.- .1-

. :a BRP091 02/01/88 l

i w n.

,- ::'i 1: : l' 1'

-: c: Il.l.i:: :. 1'.? .lj.: gt.:.:.:-t 11 I f i. Hi,!.!.i.t .,,i.i.t a t, i,n l 4 j . 2.; 1 .13 s. g.. ., i,, 1, 4,,,J 2 r.. i' L 2 1+ 50, ".' lili. l' 1 R L; di.:- '..: . u.. L;L_ IE. p'

2. n..

e": as '.r. i,, y. i. i..... i: i g e8 ..i 1. t.. i... mpu s.,. .m, .i, ti, p*7 ! [ u'l 'Ii l$! lIIl j

'stii, id' l

14.i l..! ~8 E. .i.dk -l' j ~ N,- M, _,,,.p*='C',,5!p, 3-g ea*] gHF,_%, D'-% ,.a. l m. -!.,N!h! S. i 4_ b a 40 4 7 '% - :h::15 h bMk 5 bN, 3 J._l._... .! I i J

', LA. ' oppE

". -c- [.. yrw p+ fT ; ,Q;M"* 'l i 9 ii1; 1 3 30 r. ]jJ8 - [ METAL WELDS g ,,, wr,9,cf,.,.chqp.p,pp ,,;,.J.- -,p,- r,. g,-p;; e j '_N l I. t 0.35 0.30 - ^_ ',

,p' '

l -

7.. g' 7

g,' 4 3 j l M. a 7

1.,,

---.-- a< 4 '"' .w -_ ,,u, 7 y I i' b{f $ M b f-~',I. I I [ if li!: :lli 0.30 0.25 - Q e ~ 'b' M d=b'T !# hb;;M b20 f 0.26 0.20 - h.l at

i

s::d.c$

hd.,,' 4 i '11rh.P N,. - U Af a; !b Il 'l l! m i :l 0.20 0.15 - # s i.t l s-I 1. i i o-i i t-_,- l m } 4; d~h l g l l s d* ,,p, g 4",,, I t-i

j..

'~ ir l 0.15 0.10 - F L a- [.. I

f.;I Tl-, i

. 4. m S,-( '} l!',: E' i M-i j l J. .lo L. E o 1 c g 'I i .Il I. a i I i I :n u 0.10 0.05 - #1"' I ..I 1 I i_4.1 ), 4 r i .lH.i l i? i 5 jo

j-b: :[

s,,; p p_t u + : i 'it i .E 1 if ji 4 i ,,d y i'- i i'i Lil i 5:. l,i41 C1 c

., N FH: u. k EL.

i2... I, !

r...,

~t it:

e u.

- 4..i 'n n .12 i ! :.. :p:: ;:--.w: i-6: i ..i: m: a c t ~ i: to:, i. :.:

di

Ii I

s i ;,.

g. :1; :i:.
c: ii*

I , :t: s - J s i+ st - i:

jj.: i:3;.j!!. ];:p:; iiii~"~ Ti~3::.

l. :: }I:f l; i i g g-f

ql;.-t :!

t.,i :..

i

: l ;.!

j Ig s.

jg.

l l:-

g.

g i

II.;.j

j.;.

;

l

!!:: it _ .r t

.:; g

l. t. :. jj!

S I' .::. ::9 ;c. ... F.tj: E, .;:1

:?j g !. ::'

!:i: ,i.; g5.. it

!.: -l. :

t m

.3 er, m

EA-GL-92-01-01 '.:n tt*. ,3 j ? t i- ,.. h-

r. : *: : *'. : -

i. :;::

Sheet 6 of 6

I1

.- 13't *;- I';;- .'l

9. !.

1 d' l

l. ;

.i t ".e,,Il u'- ,.~ .i.. .n.. Rev 0 ". 6 - ~1 1 . g. ., u. .,,1 1,,.. ,{_

l

.g.

n.. ;

l? i. . 4j,.g. l ,j g. 6ga: .. 3 .';j

. - M M.

.-a }'l'. :7.-

_;: ;.j.:l p:: p1: : :.;:

F q;.: 1 -~ ~ -- t tr . : : j: .q ::y tr +--- r p 3 {j; ;;

" t-.

_i_.l 2 t!.; - ;!...-.::3 j a; 3

st:

tit 3 I: t --1. :.4. u:1 2 __ _j 1 i i: : P.. tp:. ;;j! ._;.: 2 t n.

a: a:. ::g : ::

s ... _.u. a..n p,

u

t. ;. :.

T: I.:;l ::,.; p R... H. i,M.... d.?:.iLI

.!i

1 3y! ;d;- ii..L.

E...j:iI!i

2., l, d,M.

1::';

j

@,n!.b: !11 1*': l: n:! un 1J d,, E.

h.. n.

u._rD 3: Ii t n, t 'm . lj

!! jj?.

i

U ii 'If

11-

!1 It -l l ::'- l.+. t it: ::1. :.P ti'

t

lit i:n

..; t.;

44. t.:

i ,l t e. _;.1,.. a.[,jl.tti, I: l.. e-3 .- im 4t

13 r

f rl' t i an

t...

-.4a. _i.. ..,t .t n. .1 . i.. r i i

i..

a a ..n

p.....

.,i ;.,. y . a. h:. q-4 i l g 4 3 h-h. 18 i r-t i-t 4* t r i t- --* l7 4' 6 8 10 2 4 6 8 10 2 4 18 19 2 X 10 2.3 3,g,p_ FLUENCE. n/cm2 (E > 1MeV) rol Vyt FIGURE 2 Predicted Decrease in Shelf Energy as a Function of Copper Content and Fluence

EA GL 92 Ol-Attechnent,D1 ,4 P .m...

8dIIes 8 l

. ~ f u.se NUCLEAR REGULATORY COMMISSION neer use @..g/) REGULATORY GUIDE f i OFFICE OF NUCLEAR REGULATORY RESEARCH l REGULATORY OUiOE 1.90 i (Teak ME 30548 RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS A. INTRODUCTION tion toquarenats in 10 CFR Part 50 have been cleand urder OMB Clearance No. 31504011, I General Design Critenon 11, ** Fracture Preversion of Reactor

8. DISCUSSION Coolant Pressure Bourdary," of Apperdix A Gezeral Design 1

Critena for Nuclear Power P.aras " to 10 CFR Part 50,* Dmesoc Ocensing of Prtduction ard Utilization Factlities," requires, in Some NRC regairemetes d,at neceuitate calcdauon of radia-. tion embnttlemes are: past, that the reacwc coolars pressure bourdary be designed with sufficient margin to craure that, when strened urder operating, rruinterance, testing, ard postuisted accars cordions. (1) the

1. PargrghV.A of Appendix 0 requires the effects of neutroe I

bourdary beh ves in a nonbntde mar 2rer and (2) te probability rata:be to tse predited fnxn de resJts of patmers rd=nr= cfhan of rydfy propagsting fracture is minirnized. Gercral Design stadies. This guide provdes asch results in the fann of cakstative j e Criterion 31 aho requires that the design reflect the uncertainties procedures that are acceptable to de NRC, in determsrung the effects of irradisuon on rnateral proprties.

2. Parsgraph V.B of Ap;endia O desen'bes the basis for scaing Apperdia G. "Frs:ture Toughness Requirements," rd A;perdia the upper 1mut for pressure as a funcoon of teaperature during

[ H. "Reactre Vene! Matenal SurwiCarte Prtram Rearer ents " w h.ch unplerr.cnt, in part, Criterion 31, necewtate f.e cakulation hearup ard ced.dow o for a given service period in terms d the ? of chariges in frecure tmghress of reactor vessel ma:cruh caused predicted value of the nQusted referen:e temperanare a the cod ty neutroo rhw dauug%n the servre life This gue desenbes of the service period. i ger.cral prtcedures accepable to the NRC staff for calcu!a:ing ese

3. The definition of reactor vessel bcMhe giveri in Pargraph effects of neutroo rMia:ke embntdecent of the new-alloy steels cunendy used for light-.ater<ooled reacer sessels.

D.F of A;pendia G requires Meat facaticeof resbes of the ressor sessel that are predicted to experierre sufrecient neutron rm'.snam 4 ernbntdenent to be consucrtd in te selecten of de arms lanszng D.e calculative procedures given in Regulatory Position 1.1 of this gude are tot te same as those givtn in ec Pressunzed Therms! matenal. Pangra;6s ID.A and IV.A.1 aprcafy the ait:ianal test S'.ock rule d 50 61, "Tracture Toughness Requiretrou for Pro-reqdrerrmts for ScMine rascrir.is taas su;piemeza de rapitterzas ? for rea: tor vessel ma:cria!s generaity tect>oo Against Pressurued Thermal Sbxk Everns " of 10 CFR Part 50) for calculating RTpTS, de reference ternperstare that is i to be com;sred to the screening cnterian given in ce rule. De

4. Pangraph U.B of Appetd.s H im.mc. ASTM E 115 j

informatsoa co which eds Revtuon 2 is tused may also affect de by refercoce Paragrege 3.1 of ASTM E 11542,"Stardard Prac-basis for the FTS rule. The s:aff is presendy consdering wicther tice fue Cocductag SurmLnar Tests for IJgisNana Cooled i I Wdear Powie Fr.w Vessds" @af.1), mpires ihm 8e amamais to propose a change to i 50.61. to be placad in m.rvehre be shoec that may limh c9erstma of The Admory Commmee on Reactor Saftguards has been coo-de reactor durirq its lifdice,i e., tone expecsed to brmhe highea [ sulted coooerning this guae ard has cocrurred in te regdstory adjcaed reference tezzpers'ure or de lowest Charpy apper shelf energy a crd oflife. Both mesanes of redistice embnert,, nan tuna position. be consacred. In Pangng6 7.6 of ASTM E ltS42,1be require-l menu for the numter of capales and the siddrawal schedule are Any irtfurmarko cruecrim n:tivoes erreired in this regulatary tused on tbc calculated amaum of raianon embatdemem a end guMe ase contamed as rexprecats in 10 Cl R Part 30, stich pro-i of life. vides de reguistory tesis for this gude. The irJannsnon conce-e I U1,N RC D ECUt.ATORY C utDt.S The penoes are hswed en tte fo4o*fas ten broed did:W Re,wtatory Culoes are haves to eescrite and snake svarat.te to the h, a naportation evas o e it s, t 6e e tah est psactors f No C r.t"m"70' 'O aW'3"'.MT%f= "#'"- !: M'%"*am W"' I: 2=*."J' h.ew i i nc rs subetJtutes for reput tions, and seempitence wtt S. Materisas ene Mant Protection le Generas O w edes are ne Nt ocer D.rton ~.se et. 4.rmri - g,0,,gedggg..g,*go,,, or a e nt 9 oss wit ,ogg eyg,,, g fa,Y.b PN1 omca Sea This pAde was kswed after consideration of comrneats recarved froen 3 Fest, we# Wytor( DC toeb7 eat,Yenophone (202)27s 2060 or l the putdtc. Corromer ts and tutgest6cns (or lmswovements in these (2c2)27s 2171. yvsoes ar.e enco_ur t9ed _at arr t_hea_s, and.~puloes,.wf.ri be.w tt.e_d, as rei rr .e s to n ct n . r-to 4 _. m.b .e rc,_.,_ t~ s _,. _ to t~_Rutas and P_roce_duees m,w.r _aneilne o.rder b. asea D.et.e'.s_on thes entorm.stnc.n Ser on a st We.rtten comments enay be Dutwnttted.r s s ss r co g w,y,w. m. me f4. W4 7d 70 / 8 %,, ! i n-

EA CL 92 01 01

~

5 Paes 2 ef te, nov 4 g

w,, gemur : of radados embrit$ amens used la 61s rules Os6 ele's derived fweeles (Saf. 2) ese at T far seih and 11 T. are dimbi fmn the 8'88tte of he Charpy Y noes impece inst. ka base metal deepte onesehe eSers to tai e need diet nduced pn O to 10 CFR Pet 30 ml@es eut a M oes of abeeded 6e % mon ha ne see of scenense hea tem a 3 ea N g versus enttgers:we be obssaned through he eaGe4oMale reecest Os pleos of te cekulashe procedates ghes le 61s gukle) tre.nsidon temperature segba. The edM=* of 114 rsference requires =% ergbewin@lgmset to eveless de credb5-sernpersture. ART DT.k denred in Appendix 0 es te tempere-Iry of the da:a and assigs suhable margiss. When surwmance data N cure duft in die Charry curve for de irradiated materh! re!stbg from 6e reasor in questkui become avaDahle, the neight gives to that for the unitradised materk! measured at de Dfmctound m diem relatM to de informanon in this pide mi!! depend on de energy level aM the des than formed the basis for this guide oere credNiry of de survei!! axe data as fadged by de following arL= criscrit: 3o_goot.pnund shift values. The seco d measure of r crnbrinlerners is de decrease in de Carpy gTer-delf energy level, wbch is defined in ASTM E 185-82. This Revision 2 opdates the

1. Materk!s in de capsdes should be those jdged most likely calculadve procedures kn de adjustment of referexe tempersture; to be controlling on regard m rad;a6oo embrn:lemes according homever. calculative pocedures for the decname in gycr-shdf m the recommeMations of this pide.

caergy are unchanged tecsse d e preparatory wort had not bece canpleted in ume to indude them in this revhion.

2. Scacer in the plots of Charpy energy versus ternperature for.

de irradiated aid unirth conditions should be sma!! enough The basis for Equanan 2 for ARTNDT (in Regulatory Poahion to perrnit de deterrninatbn of the Dfoot-pound a:mperature ard . g of dus guide) is ccwnmed in publications by G. L Guire (Ref. de upper-shelf energy unambiguously.

2) arid G. R. Odette ci al. (Ref. 3). ELxh of these papers und surveillarce data fmm commercial pact reactors. The bases for

3. When there are two or mort sets of surveTara data frorn their regression corretabons were 6ffertat in that Odette ma$e ore reactor, the scacer of ARTNDT values aba a test-fit line grearer use of physical sedels of rda6on embric: cme:x. Yet, the drawn as describcd in Regulatory Position 2.1 normaDy ab:iuld be two pape rs cor tain sirriar recommerdadons: (1) separate correla-less can 237 for utkis and 177 for base rnetal. Even if de Sueue ex.n functions sbAd tused for weld ard base metal. (2) the fune.

range is large (two or rrore orders of magedtade), the scacer should two should be the prtduct of a d.emisry fawsor afd a flueme factor. rre eacted twKe tbse values. Even if de data fail this criterion (3) 0.e parwrters in the chemistry factor should be the elements for use in shaft calculations, ciey may te errdible for desermining cgper and n;cket, aM (4) ce fluerce factor shmid provde a trerd 6xittse in unct-steif errigy if ese upp:t shelf can be cleady deter. trJned, foDourg the definit >oc given in ASTM E 115-82 (Ref.1). curve s]cpe of ab.t 0 25 to 0 30 on k.g log paper as lod n'cm8 (E > I MeV), steeper e low fluexes ard Sacer at high fbences. Regulatory Posioon i 1 is a blerd of the correlataca fanctioos C The irraianon ternrcrature of the Charpy specirnens in ce prcuoted by these audors. Some test rea: tor data stre used as capsule should enatch vesse! wall temperature at the etamg/ base a gude in establishmg a eutoff for de cherr.istry faaor for new-cetal interface m,11n 1257. cwger rr. ate nals. De aca tee for Regu's:ory Poacco 1.2 a e.at ginn by Spetver H. ILnh (Ref. 4).

5. De suricillaxe data for the corre:auc.o morator rr.atenal in de capsde shodd fa!! mv.hin the scaer.r bard of the des base for The measure of fhrace used in this gude is de auraber of that c.stena!.

ricumns per 4are centincw hanrg energes greater tan I nuisan electroo volts G > 1 M*). The differexes in energy spectrs e To use the surveilluce data from a specs 5e plas instead of die surveillarce capsule azd the veasd inter surfwe k;ca.>ons do Regdatory Poutoo 1, one rnust develop a rdanooship of ARTNDT not a; pear to be great cretsh to marrant de use of a damaic fax-to fbest for La pla:W. Because such da:a are timbi in number tics such as displacements per stan (dpe)(Ref. 5) in the anahais ard subject to scaner, Regula:ory Positbo 2 desertes a prtzedure of the surveillaxe da:a tee (Ref. 6). in stich er forzo of Eg.ation 2 is to be exd and the fhence fac-tar certin is tetained, but te cherrus:ry factw h determmed by However, ce neutran energy spearum daes charge signifca:c!y de glas survam data. Of several possible ways ) fu such da:a, with icotmo in the vessel msll; bence for @W the a:teron-die me6ad 6e trmmm de mms of the squares of the errors ticio of radiatoo crieir 1"we through the vesac! wall, k is was che a:xnentat artusr0y.1:s use is justiflod in part by de necesaary to use a damage functiac so determite LRTMIT versus fact 6st **least squares" is a comrnon wM far curve fkting, radial distame imo de nil. The crus *Mely acxrped damage f.:no-Also, s'en dere are cdy to'o data poires, de least sparcs method tioci at this time is dpa, aM the amaxion formda (Equanoe 3) gists gnater

tight to de poi:n with de higher HTmyp; this given in Regulatory Prarna 1.1 is ts.ned on the anera.atke of dra seems reasucable fut fitting survei:laxe data, tecause genera:ty thrtiugh the vessd.at.

de higScr da:a poi:t will be te mort recerr a:d certfac mi: repre-sers core exi$crn procedares. nsitivity to neutron rafation ember:lerners rray be aficcted .a t y clernczas oder t'an crpper and n cid. The original verdi and C. REGULATORY POSITION Revision I of this pi5e had a posporus tertn in de chemistry factor, but de stadies os 3 hich this rnisbo 3 as tesed fau:d cdser

1. SURVLTilANCE DATA NOT AVAILO!2 eletrrots such as phpeus to be of Sean!ary impor*.ance,i e.,

s including them in de ana'ysis 6d not prr.da a !gmfcaat!y bct-When credible surveiIlaxo da:a fracn 6e reamt in gacetion ses fit of de da:a. are tot availatie, cskulari,o of ocutrte rdatbo cobricleznent d te teh'ine d reste team 4: ofligSt+ers reactors shxad be tesed Sc3ner in ce data tase used for this guide is relatively sig=5-on 6e procedures in Regulatory Ptdions 1.1 and 1.2 witin de cant, as eviler.ced by t's fact that e,e staadard deviations for limha*ims in Rege! story Posiin 1.3. IW2

4l EA OL 92 01 et Attestemng 1. I Page 3 et 4 g,,, 3.3 g,ssed Referesse Tm,.=.; Here, og is te snaded devission Gw 6e inhint RT spy. W e 3 asceswed voies af inkial RTMDT nc es enseerial in ponies in I svenamie,.i i mbe ssamen.d sroen en pr.ew f e.mm seed. The gj.ned referena imnper em (ART) for ach iraterial ia i she teldine is gives by 6e foDowing eapression: If not, and generic sness values foe Get class of manwist are used, i og is the sundard deviados obtained fkorn te set of Asa seed to ART - Initial RTNDT + ARTgor + Margin (1) esablish the emesa. NDT s de refercice temperature for de vsdrtMisted 1he stardard devtstion for ARTNDT eA. is 28'F for melds and i trutial RT traterial as defined in Paragraph NB-2331 of Section III of de 177 for base netal, eacept that e A need eat exceed 0.50 times I ASME Boiler and Pressure Veuel Code (Ref. 7). If treasured vduesthe sneaa value of ARTNDT-l of irutial RTNDT or de material in ques 6on are not available. f geriera incan values for that clau* of rraierial may be used if there 1.2 Charyy Upper Shelf Emergy an sufficient test resuhs to esablish a mean and sardard devia-tion for the class. O.arpy upper-shelf energy should be assumed to decrease as a function of thence and eqper contest as indicated in Figure 2. ARTNDT s $e mesa value of the adjustmern in reference Linear interpotaion is permmed. i ternperature caused by irradiation and should be calculated as 1.3 Ltrn!tations foCows: ARTNDT = (CF) f(0.28 - 0.10 log f) (2) Applicanon of the foregoing procedures should le subject to i de following firruta6ona: CF (*F) is de ctemistry factor, a furstion of ecuer and nickel corgers. CF is given in Table I for welds aid in Table 2 for bue I. The procedures apply so the grades of sA 302. 336. 533. metal (plates and forgingst Linear interpolation is permitted. In aid 508 steels having nunimum specir.ed yield strerghs of 50.000 i Tablea 1 and 2 "*cight-percera cop,ct" aM "*eight-percera psi ard urder and to their =cids ard hear-affected sanes. r6cici" ase the best. estimate va!ues for the tr.ateral, which will nonr. ally be d.c rnean of de sneasured values for a plate or forging

2. The prucetres sie vald for a narniralirintation acrnperature or for weld sampics made with the *cid wire heat number that of 550*F. Irrakauon below 525*F d ouM be consdered to pro-I matches die ersuca! vessel weld. If such values are rx* available.

duce greater embrm!ctrent, ard irrMistion above 590*F erey be ce u;pec lirrutt:g values given in the rneenal specifcauans to wtuch consdered to proixe less e di.n.a. The cornet on fator used e,e vessel s u built rr.ay be used. If ext as adable, conscrs a6ve should be psnTed by reference to xnal data. estimates (mcan plus one surdard devianon) based on genere data may be used if justreinon is provded. If there is m infonr.a6on

3. Application of eiese procedures to fluerme ic*els or to cop-avsilable 0.35% ecyper and 10% rackel sWid be assumed.

per or racLt.1 cr.rscs beycrd de ranges given in Fipue I ard TaNes The neutron fluence as any deph an te vessel wall, f(10* n/czn'. I and 2 or to materials havir.g chernical composinons beyord the E > I MeV). is deternured as followv range fcund in e e daa bases used for this gude shW be jusuf ed i = fsurf (c -Ur) (3)

2. SL'RVE. ILLA 5CE D ATA AV AILABLI i

where fsurf (10* n/em', E > I MeV) is the calculned value of ce rieutroe fluence at es inner wetied surface of ee vessel at the Men two or more credible surveinance data sets (as defined locadoo of die pos:ulated defed, and s (in inches) is the depth irso in the Discussion) become available frorn de reactce in g.aestaca. d.e vesse3 wsfl measured from de veue! inner (wened) surface. dary inay be used m> determtne de djusied referen: mrngerature Ahematively, if dpa calculanons art made as part of the fluctce and the Charpy upper. shelf energy g the beltiine matenals as l analysis. the ratio of dpa at the deph in question to dpa at de inter descnted in Regulatory Positions 2.1 and 2.2. respective!y. f surfxe may be subs 6:unod for the exponcatal anenuatbo factor in F+atsoo 3. 2.1 Adjusted Reference Tesaperstart The fluerre faaor,f018 - 0.10 log f. is &termined by "1, A-The adjusted tefeterre temperature should be ebunned as tion or frorn Figure 1. foDows. First,if dere is clear evdence that de ecger or sectef f caraers of die surveinance weM differs frurn d.ar of the vessel wid. I " Margin" is the gancity. T. d.at is to be Moed to obuin coo-i.e., daffers from the averige for de weld wire nest nurnber i servauve, u;$cs-Mad values of Mjusted refererxt ternpersture ance:ated with t!e veuct weld and de surveitlance weld, the i for die calcu!rions required by Apperdia O to 10 CFR Part 50. nessured values of ARTNDT should be adjusted 17 inultiplying them by the tudo of the chemistry factor far the vene! wid to that i for d.e surseinaxe weld. Secord, de survei!!arrt da:s should be f.fted using Equanon 2 to obtain the eda:boship of LRTNDTto Margin - 2 V og + oj W flucace. To do so, calculase the cherrustry facsor, CP, for the best, NDT y its oarresponding fa by enuhiplying exh m" justed ART b floence fxtor, sumaung tbc products, and divding by the sutn of i d" 7*17M Q"$N"E~ N JE d'c sqt. ares gthe fLeme factms. T1e res4 vaiur e(CF M.es su we mmt. 9 en asTw s<.msue s anrcara entered in Eqaanoo 2 wiIl give the relanonship of ARTNDTm> i e 1.99-3 4 i i

2 ~ .~ E 8L 92-01 01 Attac6 merit 1 l Page 4 of 18, Rev 4 TAttA t r* ss<me CHEMtSTRY FACTOR FOR WELDS. *F I i Nchst. W.5 wa-5 0 0.20 0.40 0.00 0 to 1 00 1.30

Coppet, i

O 30 20 20 30 30 30 30 0 01 20 20 20 30 30 20 20 0.02 21 26 27 27 27 27 27 i 0.03 22 35 di 41 dl di dl 0.04 24 43 54 54 54 54 54 i 0.05 26 49 67 68 68 68 48 32 32 O.06 29 52 77 82 82 0.07 32 55 15 95 95 95 95 0.0s 36 58 90 106 108 los 103 l 0.09 40 61 94 115 122 122 122 I 44 65 97 122 133 135 135 49 68 101 130 144 148 143 0.10 O.12 52 72 103 135 153 161 161 0.11 ~0.13 58 76 106 139 162 172 176 a 0.14 61 79 109 142 168 it2 lag t 0.15 66 84 112 14 175 191 200 0.16 70 18 115 149 175 199 211 O.17 75 92 119 151 184 207 221 l 0.13 79 95 122 154 187 214 230 0.39 33 100 126 157 191 220 238 l i 0 20 18 104 129 160 194 223 245 f 0.21 92 108 133 164 197 229 252 O 22 97 112 137 167 200 232 257 O 23 101 117 140 169 203 2% 263 0.24 105 121 144 173 206 239 -268 i 0 25 110 126 148 176 209 243 272 O 26 113 130 151 190 212 246 276 0 27 119 134 155 184 216 249 280 i O 2B 122 138 160 187 218 251 284 O 29 128 142 164 191 222 254 247 l 0 30 131 146 167 194 225 257 290 O 31 136 151 172 198 228 260 293 i 0 32 140 155 175 202 231 263 296 O 33 l'4 160 190 205 134 266 299 0.34 149 164 184 209 238 269 302-l 0.35 153 168 187 212 241 272 305 0 36 158 172 191 216 245 275 308 0.37 162 177 196 220 248 278 313 0.38 166 182 200 223 '250 281 314 O.39 171 185 203 227 254 285 317 0.40 175 139 207 231 257 288 330 t to obuin rrean values of shA. ARTNDT. In calcdaca6 e nnargma. d fluence dat fits de plant surveitance dets in such a may as so de value of ea may be reduced from the values gives in the last~ minimize te sum of the squarcs of fic errors. paragraph of Regulatory Pcamos 1.1 by an amount tobe decwted To calcv!ste the trargin in etis case, use Equation 4; de va}ues on a case-bys:ase basis. depending co where de incasund va!uss l fall relative to the mean calculated for ec survcGlarse materials. l given there for og traay be cut in t.aff. l If this procedure givea a highet value of afjusted referetcc l 2.2 Charpy 14per-Shelf Furry tem;erature than that given by using de procedures of Regulatory j Position 1.1 ec sarvemaxe da:a should be med. If this prtxxdure The decrease in upiedelf energy may te obtained by plot. gives a lower value, eider snay Ic wed. ting de reduced plant surveEmpce data on Figure 2 of this guide -[ t Tor p! arts having survemaroc data that are credNe in aD respectsan5 fitting the data with a line drawn parallel to the etisting linen cacept dat de material kes not represca the critkal material in as de upper tound of aD the data This lire sismid be caed in c.e vessel. de calcutanve procedures in this gdSe shmid te used preference to ee existing gra4 l 7 1.94 4 i 1 ~ i

e.. EA Of.-92-01 01 Attachment I Paes 5 ef Q tey e D - TAM 2 2 CHEMisTitY FACTOR FOR RASE METAL. T Ccipper* Nittel. Wl4 w4 0 0 20 0.40 0.60 0.00 1.00 8.20 0 M M M M M M M 0 01 20 20 20 20 20 20 20 0.02 20 20 20 20 20 20 20 0 03 20 20 20 20 20 20 20 0.04 22 26 26 26 26 26 26 0 05 25 31 31 31 31 31 31 0.06 28 37 37 37 37 37 37 0 07 31 43 44 44 44 44 44 0.0s 34 48 51 51 51 51 51 0 09 37 53 58 58 58 58 58 0.10 di 58 65 65 67 67 67 0 11 45 62 72 74 77 77 77 0 12 49 67 79 E3 86 86 86 0 13 53 71 15 98 96 96 0.14 57 75 91 100 105 106 106 0.15 61 80 99 110 115 117 117 0.16 65 54 104 lit 123 125 125 0.17 69 Es 110 127 132 135 135 0 13 13 92 115 134 141 144 144 0 19 71 97 120 142 150 154 154 0 20 82 102 125 149 159 164 165 0.21 86 107 129 155 167 172 174 0 22 91 112 134 161 176 181 154 0 23 95 117 118 167 184 190 194 0 24 100 121 143 172 191 199 204 0 25 104 126 148 176 199 208 214 0 26 109 !)0 151 ISO 205 216 221 0 21 114 134 155 154 211 225 230 0 23 119 138 160 187 216 233 239 0.29 124 142 164 191 221 241 248 0 30 129 146 167 194 225 249 257 0 31 134 151 172 198 228 255 266 0 32 139 155 175 202 231 260 274 0 33 144 160 180 205 234 264 282 0.34 149 164 164 2C9 238 268 290 0.35 153 168 187 212 241 2 72 258 0 36 158 173 191 216 245 275 303 0 37 162 177 196 220 248 278 308 0 38 166 182 200 223 250 181 313 0.39 171 115 203 227 254 285 317 0 40 175 129 207 231 257 288 320

3. REQLTREMENT FOR NEW P1JLS'TS D. IMPLEMENTATION For beldine materials in the reactor veuct for a rew plant. en 1he p2rpase of eds section is to provMe infacmaxm e applicants conters of resi&al elecents wch as copper, pbphorus. sulfur.

and Ikeraces regan!mg de NRC nafts plans br maing this and *aradium shauM be ernwrtdied to low tevels.' The cryper con-replatory pide. Eacep in those cues in m&h an a;plicam pro-icra dexM te wch e.w er cakvlated a$juped refererax temperture poses an acceptata ahernative rnethod for corrghirpth spectred at the 1/4T pwtion in die venel wsfl at crd of life is less than portkms of de Commisskm's regulations. the ereciods described 200*F. In selectmg the comurn annuns of niel to te used. ris in tha pdc will te used as follows: &leterkus effect oo teaton embrittlemerd diould te tehnced against as benefcial me:ahrgical effects and its terderry to lower I. The medods &acrited in Regulatory Ptsuris 1 and 2 of de irdtial RTNDT. this pdc usU be used ty ec NRC gaff in evaluting all predic- %r must r&wrous. ses en Amendu en ASDI $wderd Spachings A $33 G and H to 10 CF1t Pari 50. cm,g n 1.99 5

EA E 92 0101 - i .~ Attachesnt 1 Page 6 of C, tev s 1

2. Holders ef Ikenses and perses shouW eos abe aneeds TN Specifksimas is oder es esadams en 4 de described in can guide to predict the effes of neuvon ruheden en repiremesas of Secues V of Appendia 0. M CPR Pen 38.

i reactor veuel maienals as required by Paragrupk V.A of Appse-dis O so 10 CFR Part 30, veks dey can junify the use of dir-

3. The ww ** et itegulanery Position 3 are senee.

fererd rnethods. The su of de Rev'sion 2 methodobgy may resuk tinfly unchanged from those used to evaluane consuucasua permk in a modification of the preuvre4emperature lirruts contained in applications docketed on or ahu June I.1977. l i i I i ~ J k I t i 'e i i k i ii f r r i 4 i i 3.99 4 t

_.. ~ _ _ ~ EA OL 92-01 01 .~ '.

44 Att.cs at t Paes7etto,sw a t

ntrtaracts

1. Amencan sockey sw Tesias med Meurt.h. "staaktd Prw-

5. Amerken sadny k Tombs and Manednis. standmed pr.e.

tu h Coraxting savechne Tests for u k-hw cooled ske sw c% Mem hpo.ww h twdde %is k s Nacar Fmt Reeca Venek." ASW E IBM, My 1981.* Terms of DV-per Ame(DPA)." ASW E 69M9 Augua 1979.* l

2. O. L. Outhne. "Oarpy Trend Cwves Besed on 177 PWR Dets

6. W. N. McBroy. ** LWR Pressee Wesel Swveh Dommary l

Point.s." in " LWR Pressure Vese:I Survehwe Doumetry Im-W pro 5rma.1WR Pm Reacsor SurveDhnce provemes Prognra." NUREG/CR.3391. Vol. 2. prepared bF Mysics-M'W Dans Base P7' #~== " N11REV I HarJord Erigineering Developrnent Laboratory. HEDL-TME CR 333, prepared h Hanford Engim Developrnes 13-22. Apnl 19M.** taboratory. HEDI-TME 85 3, August 1915." 7' h 3*C N d N N b. W E

3. G. R. O&cte et al.,"PhysicaDy Baud Regrenion Correlatnas of Embrittlernent Data from Reactor Preuure Venel

" N # ' Po**' II"" C P " Of A3M E 800'r W I"83"" I'88'I C8d'. New York (updated frequemly).tt SurveUtarre Programs." Ektne Power Research Institute. NP-3319. Jawary 1954.t

8. Arnerwan Socicry for Testie6 and Matenals. " Standard Specif =+m for Prtssure Vessel Plates, Anoy Stect Q cac.%!

t

4. S. H. Bush. " Structural Matenals for Nuclear Power Plants."

and Ternpered. Maganese.Motytderern and Mar.gamese. in four=21 of Trsnag ad EmLank. Amencan Society for Molybacam-Nickel." ASTM A 333/A 333M 82. Sepember Testing and Materials. Novemter 1974.e 1982.* h s i [ i 1 E b 6 e f f 6 b 1 .p

Cque esy k otwned fnns em Anar as Soeury Le Teamg and M 8914 Rats Svest. % PA 194
- Cges owy be otened fram on $4=rsuredicar er Dacissemas. U S. Omrware Pristas OEss, Pusa Ccas Daa pus 1 Wadigva. DC20011-10E2.

<gus any be e%mel fnus de Derek Pv=er Research tasteht. Mt3 l'h A*uset. PhiO MID. (4 h3oi-tTorum ewy k etuewn! fnns 9e Amerwas Soce+y or Messwal !Jgnacts. MS E. C6 Street. New Yart. WY 100t1. i l.%7 i i i

LA*GL*93 01 01 l

. o s

.,4 E808 0 #f $$, ggy g i + t h e __-h i e 1 e. 2 =EE9 ' _ - = =. _ a. =# 1 --=- ~ -j -n 6 l. g 1 1 g n I B M I I g 1 I I E y I I E s u i e i n I l i 1-t E-g i i i i-i z 3 { l I-5,,- H i gg ____. = []p =gjj- @ -il.T 67[3== ""3== = E. I ? t -~ ~ ~ 1. g _=3_=a==r 2-qssE jg-2i== IL sat== --t

== = + .;= _ : 1 ;- __r:_=_5m-4 -+ g ^ .X ~"- m L,._Q

e p -

g _n y L W. ~ g 2 E5 =_eiEU=~== ===--s? i:-P==--iEithE2Ek?EW 25 = = = = ' .==Ar= --W3 i E.. -- u_=;=== n=z- - - =2, I e5 ? = g = _ _ - =.r.- r - r 1 =,...= m= " _ _.n.,_ A n 3 e --. S= g___ e,st .m_ t - s

ee

-u E h 5 1 I c I 5 4 g ',M M u \\; p 1 i gy 1 1 1 E i ,f E . 3 'f gi l 1 I 1 , i I E \\ i I A i \\ l _I, I -I 3 i

== .=m - 1 ?=2 m 4 MyL = =

== ^=e= = -- e -~ - -i =j s a- =e s-I _ ( C E Z ZZ ~ -4 _~: =

.:=

g ._ =.- g= c= g -

== =O D 1 'g E='_~255%== .~.L 5' __L =

== = '

2 =. g L _. =5 -- [ '"~ =O I = b m".'. IM i i u a i I I I I l a L n 1 E A R, i i 1 11 E a i B I I' \\ l l l 4 e.1f I i i T I, - i a e <t a e n *

R N

1 t ed I 3 *sossed saueMd J 3 Sol 000 K 4 1.99-8 1 4 .l; 4- ~

A. EA Gl. 92 01 01 ..] p. Attactment 1 F 1.o page 9 of 10, Rev 9 k,h g.- p. a f,. o.. .3 g p

T arrt:

p

.- p e.

g gf

!e

= a ..e;..mm- - -t-.,1+7r. ~ %gpy -O y agga: T {i, + : e t y .l =

n. m.

~ '~~~-- ~

.ph.: ~"r

. :.r i b r-th:"1 j hx j ..J - ka;;Vtr

t t

t,.

- ) i;;; gr:: q ny, j@" :.g ,:d -: c:- 9:.2:ea tr=l r h di n.. 4 . t.. h2 r.1:: .2' ......+ .6. . 1' t g

i

-.t.f2;T.t..,..... 1 > 'p.

- H f*Ti*4 h...

w 4, '3 - l i i-Y I Y i. bk f*** 'd It$td.1 IiH d..brif: dh M.-plth. -h n i - ? !i iji **t : W}!4 o vqsgs I'. 2bHI M c

f... u t'

n -t-- .u p3 qu c 1.: )h.- a .L n 4 g,1 =3

i

., n... #.,..g_. J. l.>8,.; ...L r r i +L. .. g 1 e 7 L 3 i Q. r ge m m o .! T I:i'"r I I PIT'..t n: L_ a l &.i.%. ?%p# pp., r.- a s w :" @ iN. h W' E i.%nW M4i W ' #% kW M. i&$.t-c w. . y.. %..-d.-nW~ .pa F1._J 2 9y: -JT g.. [i.k.:. N,! l,i,f _d. g.y r'$-... ig-M4'Edi w 4 r t ,.- '_ 2.t : y. d,rp:. y'.pp.g h.;..L +

E. d:=I.I1Q):-A'-:tcd;"r

~ . i: =_=_ . $ct :-.. _ _ _ _ _a-

=.

hn .x 3 .t =

- e

=

.r.p. t.t r.c ;a.a +.e r .a., 2 r._ .".[2.:._ag.n(':.__ r .J..

E r

r .,m () _1.. 4.. ,m. _ a .. 1.._- p,4. w g T h'. y ,t 4H. %n

1.-

n.: T : :-

. :.. = ...; i.i d Ai n.- N g =.w

:

= : 75.

.:.: c y_. =..T.:.4 :

m.4,: E g .u.= -m r- --E;si 34 .e sm

=

===.:=

t-+ r=n.

dI--M

== t.:--.- m.mn .....3 . _ =, _ 2r..t :. :: c q.. r-

g.,m.

s . - _...r.T 3..:::.:.:. :. ^ L

g:

__.:. _u,._t: r,,, u +c c w M (,) D .y

- .7....

._t... r 1 p;. aE .,.,Ly,.4 _, .f.- ..s 4. 3-N O -f w y .. CC e ..w ..l. .., y u. . L-y .._ a.. i c ...__7 1' 1 7 .u 1... m. .,.. o .l[ .,. r o l i q-a! Cir di to l ! ll l l i2,RMID liW il h @ 'ft' i r W W WI 3 8R R

o 8 WM #ii@ Mi @ % WMr W1 o d

d..O $3 Nil. id @) Y: tM 5.M NF EN gr { dd d d

.a c.

a -r

.L ::2 :.

i:.. :.::.- N t.. W .m n -: ::t ;.*

;- =.

.:ap--. g ... ;. r g .c: b b ri .J E ,.. A W )4 .,4...,,.,, - .a. cc lE o. oo o o 6 . ". '. ', "...*. ~ ~.

7... '. '. '. ' * * *. *. * "

o e o. ge. N. N. e, e=.'~..* ' ' ~ .~.. - y . nr m.r n ..i.. ; .w_.,..dh..:-==: m : n~ -r=rd,.. '= N -,.n _= =. -

=r :; s t n

,m .=~*~g_.., -7:t

p :n :::=3. :p- ::::=- $l 3.'i.'g

_333;:2.Q + ^-4 n u ;n

r.i-

=

2.. --:: :-

n 2: :

-m ::.x ;.-==

. :;nc-

=.x=a. : ::~._,. : : ... x--...:= .. c;

..m-: .:

+- r .r. .:/ - f:- u- = c-m=: 3 1,4--. ,_w e, . 4 _4 a7 X o o o o o o N 40 T M N nuassed '46'au3 ;lays us awasoao i 99 9 1 ~l ) i l J l

EA-OL 92 01 81 Attechnent l' i J rose is et t?, s. REGULATORY ANALYtt8 A an or de not tory andnia repwas fx e Reptanwy a sw = es ca minian's rubinew mean a 1717 N su m i OMe 1.99, hvision 2. k avdable fw inspecuon and copyW fw NW., WasMacon, DC, andw helawy One 1.99. Revidos 2. i i i ? t l i 4 1 r 6 i t 1 F l k i l t i l I e l 1.99-10 r

EA 4 4 -M -01 Attestuunt t..... p%st i '.. -. ~y . c l. +.. .,. ~. -.. a u ..:., a.s. a, -

.: . e..,.

r 4 .. g. ..,.,a,..,.,y .,3...V'.,,.* J' " y...,g j f i .u.... F -p., s. -: t -' 4 vri% :- 'r ~.

QiC:

c W ..;. h ~' W. 2-.'4- ,.,- iv v.X. G MYi.. /g ,ja;1U .2 r, Mio.J. R. HAwTug.;..s '.T. ~ ' ~. 'F? .g v-1 2 :.- g.- .r.

< -;. e jwhests brach -

f:.;j?RQ:-.' l i %~yi-n'.Ppu.@ .-q.- M s.a...

x-n: .s v.-

w

~

f,,.', .' * ~:i,,, ^., ..f.,.., (~ - t .~ .- m.. - l. m_e. ~~l,$..p g T $::, :p';' .,gg g, g,; p ;,,;. .r-ig . p-1.-~~

y Y'i.- S *

.w.n.
~~.-i.i.*-Q_,' 's- ;. ~ - l 4. .= - -m.. s :- -y..~..,e.' :.0 s:e .: m: r l August 1969 l f -w.. .., v,. -.. d I. .5, ~ ... '. ':b:.?@tl-{. ~ j ..,,.,.... m..s. r '.~~
2. .*. {.., p ' f - -
--T.,' or l .f--[9. - . t. ' 3.t fa. 7 ..'e n, ..'.,..s. .;.4. i p '::;.g. f~ n l a ..cW'so: .t-(% . a. % ',c +.. - ...'q.. 00 ) e... t yn l%$fl?, L..L..& ..s.- 2L 52.. iP ~ $.$$'b _, 4* g.
~~w.,~~
~~u .; -.. m .f'. ' lYy.. -.9 oy r. Wo - =4 -...h.;;;rtn.~~
~~.?,~~
+ 1 ...6,._.... .c 3' 5 49 'hy.;- ' n ".. n.g : - ; :. __.. o'- ..,; l u \\ ?l,. ..ma 4.. "./ y =.. -. - ! '. x. $$.._.h.g ,Q. s. o l ....,g. V~4 ~.,.[,,; j & Y 'ag4 [ m b ($r h ts k w e M sa k hs. We W'd : I.9 I --
EA eL N 01 01 t .:...*i.o- .-~ (: - .Attadment 2. '.d, .p e. a.e s, n=v .-a i l 't v (d j; yt .,.rf-1 ~? Mre"el hfi ' ree ..II"C.YN8ig fluences.'~ ,j g i ;,. /
.h
~~...+.. 0" ., e. e.y 7 5f his experiment are quite interesting fj , sainoe-@ tit'can' sow be 's;of t e conclusions uggosted that.a Ni-Cr-Mo steel (AS43).~~
~~'willgip"o.ad to neutron irradiation damage by ' thermal and f ast" neutrons at <240*P in a similar manner to low-alloy These A543 and carbon-silicon steels ( A302-D and A212-B).~~
data should be quite valuable for future inclusion into a damage function description of radiation damage 'foi this apd other types of steel. IN SUPPORT OF POWER REACTORS II. STUDIES Mechahical property and Neutron Spectrum Analyses i of the Big Rock point Reactor pressure Vessel A. and tienry E. Watson Charlos Z. Serpan, Jr. The neutron-induced nochanical property changes of thepres l Big Rock Point Reactor (BRPR) Interim tored by a pressure vessel surveillance program (2). results of evaluating irradiated BRpR surveillance specimensOne fina have been reported (3,4). now been received, and the results of analyzing these are presented. 1 The chemical composition of the DRpR surveillance steel, determined at NRL,' is combined in Table 9.1 with the heat i More details treatment and processing history from Ref. 2. l of the surveillance program are presented in Refs. 2, 3, and 111e results of evaluating the Charpy-V specimens from the veillance program are presented in Figs. 9.2 through 4. e met al specimen results are shown in Fig. 9.2, weld BRPR s ispecimens in Fig. 9.4. and specimen 9.5; Fig. 9.3, liAZ, f rom the reactor builder's standard are in Fig. 9.5. metal. he un- . figures, the data points used to deterrine t resul. te.d condition curve (3) are shown in the top portion; In th li irrad the unirradiated condition data thercEf ter in the figures, are n5t retainod so that new data points can be seen more The unirradiated condition data for Fig. 9.5 wereAll o l casily. taken from Ref. 5. BRPR surveillance program are summarized in Table 9.2. v{ ] 6 4
~~j.g,3~~
, 'I... g.5 .. g... I 4,,, ..g.'d:., ' + $hl' iyif',E@fk.'.l.7 Table 9.2 i,- <s '....... /. < 'Charpy-V Notch Ductility Data from 3 er. - ~ vyt ~ Program . ',; Rock Point Surveillance , ;p.: V} ' bt, r ( C, 30 f t-Ib Transition Full Shear Energy Ahorption, ft-lb ,,./ jf.
~~F T a pteratusw~~
~ ~n/cm >1 MeV 68 ab Unirrad. Irrad. br Unirrad. Irrad. -fs e,':fg, ?- c. 82 88 .n. -5 +10 15 4 Metal' Tpermal Control 82 ~ 82 ~ ~5 ~0 ^ 18 -5 73 id 'p,V 1.5 x 10 -5 55 60 H2 7.1 x 10 82 62 -5 65 70 2.3 x 10 82 70 0 -5 145 150 .g)i(.q,.., 1.07 x 10 {g -55 15 95 92 -70 Thermal Control Metal 0 -70 -15 55 95 80 I-1.5 x 10 -70 65 135' 95 70 '*d.- 7.1 x 10 95 57 19 -70 120 190 2.3 x 10 95 - 66 2n -70 160 230 gj4 9.. l.. 18 -55 ~-55 0 100 ~100 1.07 x 10 r;.p j kheatAffected 1.6 x 10 18 -55 -30 25 100 78 7.1 x 10 k. 19 -55 15 80 100 79 ? 20 -55 105 160 100 75 =gg 2.3 x 10 g. y 1.07 x 10 k f..,.,... 18 0 15 15 96 70 2. i 4 . 5tandard 1.5 x 10 96 82 ,a2 u g 18 0 40 40 s ,,'Aeference 7.1 x 10
:t t*
= ,g If4. +. ~~ Includes 15'F hr from therre.hI aging. 'n ~e i- .f
SA 4 92-01-91 s. utomaan n G &. M' x x. 3 .. h r T. J. <. p'.- a
. r

. r...
~~~.f... - -.Yf_jyui[*"'; ., v.d;D*lh..v.. Q \\ a.= c ? cf'. w,+ - r~~
~~73.. -~~
~~c:. ..>.. c .' a... w...- w -~~
~~.4.*.~~
~~+..& , ink %*K. dm;~~
M.
**g,
~~'t,' Op.J. R. HAwmokNs '4q', w.:.,:~~
~~s.. >...,.f.~~
l
~~..f 1~~
~~'i 1 .% erials branck -~~
~~gy J%,. e:~~
~~.r. p,. .s~~
~~Q' i.~~
~~1.*:2. _ * .. ;. m v e'.-...' - a.. n '.P D M don di k. c... g,.. g :.y..;.= e-3 9' ~~~
~~.y' -~~
~~' -' :..n ., < > Y: '. J~~
tN
~~.a . U. ~t;.>.l. b.K. . ~., * a i' e $ ? :~~
g.. f

. 'I
. s:s'g %.;- ,y Q i b'i ^. .?-p gy :l._ .,..,q. p . m .n :- I fl. y;. w.,,. n-n.... a,. 4= x.;.... ~- ~ ?, w.:C. '.%U..., *e,40 e.-;&- rt f August 1969 ' - ~ ..,s.,%*
~~i.- ?~~
~~l 1 i,., t, f, * * 7, "... [., 24's.y.'Ti~E-:l5 % ' ~ .M.s^E.5&mVlk'. .~ ::=. ._.. u..' z - e; . -i;. .w~~
% s' '.*. a.

p. l ~
~~- ) s' m'~~
~~N.[~~
~~.r., ~\\ '.3. c.'., ~ . g. ~.~~
** Y.

q.
.N 5 .. W. eens d e f. ' -*jt i * ** l ' .,,M .,.%,. *r,.$,':h *D '. _* ' '1 k',f;, * * 'a ' h. - 7*
~~Y.i.~~
~~p [ a*- 9 3."..,','~~
~~4 A. W' *h* s.~~
~~%.af g.", C'#e' e. s c-~~
4.
~~s. t s~~

~~f.~1 a~~
~~.~.s m. ;,.. w; ....t.E.* . Y,* I (~~
~~)#,.~~
~~.,4f. .a~~
4.,

S
~~.s ... e ,(~~~
Q.*
~~. r ,v.: ...n..~~
~~f. g.o :....~~
..r... - ; ;-... y 4, k. gs '\\; 71 4< _~_'*; m _?* l*. p.
EA-at.92 01 01 i-E "ddk' . Attademnt 2,,W .Pege 3 of 3, see N y.?, g .~. ? g{. ~ 1, c$fh ro $on
~~increatig fluences.~~
the conclusions' f this experiment are quite interestin; . n a..:,y ~. - .. saindeMt'can' sow be 's,0 uggested that.a Ni-Cr-Mo steel (A543)'. 1 Tjrill-(hdutrons at <240*F in a similar manner to low-alloy ^ f as t" nese'A543 and carbcn-silicon steels (A302-D and A212-B). data should be quite valuable for future inclusion.into a damage function description of radiation damage for this apd other types of steef.. STUDIES IN SUPp0RT OF POWER REACTORS - II. 11echahical property and Neutron Spectrum Analyses of the Big Rock point Reactor pressure Vessel A. Charles Z. Serpan, Jr. and !!cnry E. Watson ,r The neutron-induced mechanical property changes of the pressure vessel are being moni-Rock Point Reactor (BRpR) Interim tored by a pressure vessel surveillance program (2).re Big One final group of specimens have have been reported (3,4). now been received, and the results of analyzing these are presented. The chemical composition of the BRpR surveillance steel, determined at NRL,' is combined in Table 9.1 with the heat More details treatment and processing history from Ref. 2. of the surveillance program are presented in Refs. 2, 3, and l i The results of evaluating the Charpy-V specimens from the BRPR sgveillance program are presented in Figs. 9.2 4. Z specimens in Fig. 9.4, and specimen 9.5; ~ Fig. 9.3, HA. f rom the reactor builder's standard are in Fig. 9.5. metal. the un- .figuros, the data points used to deterrine resu te.d condition curve (3) are shown in the top portion; In Lii the unirradiated condition data f.rrad there3 Citer in the figures, are n5t retained so that new data points can be seen more no unirradiated condition data for Fig. 9.5 wereAll of easily. taken from Ref. 5. BRPR surveillance program are summarized in Table 9.2. 4 er ~ 6
~. _. "j.g,,,
~~d,... ;QQ -~~
~~?,')[yi',,,:p'$s:-%..,:>.? Table 9.2 g,. 2 n. ' 'c. ' y y.<... - . 3) . Chd,py-V Notch Duetility Data from ~ er. ,^ - e. ?.'(( ', Rock Point Survoillance Progran~~
~~s&T~~
~~9;. t ;~~
~~C ,s : C 30 f t-1b Transition Full Shear Energy~~
~~F Absorption, ft-lb j~~
'4 P' ]~ y I Temperature R/cm >l WoV d.'" - 68 mb Unirrad. Irrad. bT Unirred. Irrad. D,.h..- \\ .d S' \\,/, S' 43. - -5 +10 15 82 88 4 3 tal* T)ernal Control 82 ~ 82 ~ -5 ~0 18 -5 . fi 1.5 x 10 i 18 -5 55 60 82 73 I 19 -5 G5 70 82 62 7.1 x 10 i 2.3 x 10 82 70 .u a ..;;ik.,.. -5 145 150 1.07 x 10 . h -55 15 95 92 -70 . d' Metal' Thermal Control 10 -70 -15 55 95 80 I-1.5 x 10 95 70 65 l't S -70 N..- 7.1 x 10 95 57 I -70 120 190 x 10.,9 -70 160 230 US ~ 65 N 2.3 1.07 x 10'g v. l.. $*, f' 18 -55 ~-55 0 100 ~100 hiest Affected 1.6 x 10 -55 -30 25 100 78 7.1 x 10 19 -55 15 80 100 79 0 -55 105 160 100 75 =gg 2.3 x 10 1.07 x 10 "E4 18 0 15 15 96 70 1 St$mdard 1.5 x 10 96 82 S {" '(neser.ac. 1e 0 40 40 -.. g 7.1 x 10 E ,5? r i.- 4 Includes 15*F hr from theriaal aging. '*r
~ Attestmens 3 e i.e a, aw e 8 peelenoson:E 105-82 n-Standard Practice for Conducting Survelilance Tests for Ught-Water Cooled Nuclear q Power Reactor Vessels, E 706 (IF)' w 'I4% nis mad rd *= immed inda e nied desis=6 e t issane numter immediady m.us ne dese=uo. i.du.ie m,mc or orisinal adopuos or.in the caer of revnion. she year erinst avmaa. A number is parresheme ladwates the year erlast napproval A supencript esmaos 4)indkanes an editorial chaase since she last revumas er reappro=ut h a r*rrr-sw6ae s.2.3 was can ned edisoriant and the dc.isn.oon d.= -as sensed Jury i, int 2. '8 Nott-Tbc tsar aus dnaged editorissy in July 19ss. eters are made throughout the service life of the reactor
g. Scope vessel to ac ount for radiation effects. Because of the 1.1 This practice covers procedures for monitoring the f' ndiation-induced changes in the mechanical properties of variability in the behavior of reactor vessel steels, a surveil-lance program is warranted to monitor changes in the i
ptic materials in the beltline oflight-water cooled nuclear properties of actual vessel rnaterials caused by long-term 4 per reactor vessels. His practice includes guidelines for exposure to the neutron radiation and temperature environ-gsigning a minimum surveillance program, selecting mate. ment of the given reactor vessel. His practice describes the ris!s. and evaluating test results. criteria that should be considered in planning and imple-1.2 This practice was developed for alllight-water cooled menting surveillance test programs and points out precau- } pclear power reactor vessels for which the predicted max. tions that should be taken to ensure that: (1) capsule anum neutron fluence (E > 1 MeV) at the end of the design getime exceeds 1 x 10 ' n/m (I x 10" n/cm') at the inside expasures can be related to beltline exposures, (2) materials 2 2 stface of the reactor vessel. selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor
~~2. Referenced Documents vessel, and (3) the tests yield results useful for the evaluation of radiation effects on the reactor vessel.~~
2.1 ASTM Standards: 3.2 The design of a surveillance program for a given A 370 Test Methods and Dermitions for Mechanical reactor vessel must consider the existing body of data on Testing of Steel Products: similar materials in additiot. to the specific materials used for g E 8 Test Methods for Tension Testing of Metallic that reactor vesset he amount of such data and the-d Materialsa E 21 Practice for Elevated Temperature Tension Tests of similarity of exposure conditions and material charactenstics will determine their applicability for predicting the radiation h Metallic Materials) E 23 Test Methods for Notched Bar impact Testing of effects. As a large amount of pertinent data becomes Metallic Materials) available it may be possible to reduce the surveiBance effort E 208 Test Method for Conducting Drop. Weight Test to for selected reactors by integrating their surveillance pro. Determine Nil-Ductility Transition Temperature of grams. Territic Steels' E482 Guide for Application of Neutron Transport
~~4. DeneJtions I~~
~~Methods for Reactor Vessel Surveillance~~
~~J E 560 Recommended Practice for Extrapolating Reactor 4.1 cdjustedreferenutemperature-the reference temper.~~
3 Vessel Surveillance Dosimetry Results* ature adjusted for irradiation effects by adding to RT,er the 2.2 American Society ofMechanical Engineers Standard: transition temperature shin (see 4.85). Boiler and Pressure Vessel Code Sections 111 and Xi' 4.2 base metal (parent materia 0-as-fabricated plate ma-terial or forging material other than a weldment or its
~~1. Significance and Use corresponding heat.affected. zone (HAZ).~~
3.1 Predictions of neutron radiation effects on pressure 4.3 behline-the irradiated region of the' reactor vessel j iessel steels are considered in the design oflight-water cooled (shell material including weld regions and plates or forgings) y andear power reactors. Changes in system operating param-that directly surrounds the effective height of the active core, and adjacent regions that are predicted to experience suffi-j
~
cient neutron damage to warrant consideration in the 'T** 3== is =5a theiur=dk*on of AsW Commina E 10 on Nucsear selection of survedlance material, i 4.4 EOL-end-of-life; the design lifetime in terms of s d July I.19s2. Pubrated septemtur 1932. OrisinaDy years; cffective full power years; or neutron fluen(2.' 11 - Basshed as E lss - 61 T. tmt prevuxa edinae E is5 -79. 4.5 index femperature-that temperature corresponding W I l4*=aar Aao& (AS7v standants, veh 01 A14 AS and 03.01. to a predetermined level of absorbed energy, lateral expan-gnu hQ Sion. Or fracturt appearance obtamed from the average (best a - 4 ' A'unab6e troen the Amencan society of Aviomodwe Ensinan 345 E. (rth 4 L Nr. York. NY 10017.
~~61) Charpy transition curve.~~
349
.f." u.rA-n et-ot Attechnent 3 O E 145 P.e. z et g, aw e 4.6 faction mengrA-in a tensBe test, the lood at fractort materials shan include one heat of es base amend, ces
~~l divided by the initial crW_ arcs of the test speo-weld, and one weld best-effecnod.aoes (HA23. The I'~~
imea. metal, weld metal, and HAZ (Note 1) materials indaded N* 4.7 fracture stress--in a tensDe test, the bad at fracture the program shall be those id.ai to be most divided by the cross. sectional area of the test specimen at with regard to setting pressure 4cmperature limits, foe ,,,,'I time of fracture. tion of the reactor to compensate for radiation effects d 4.8 heat-cgerted-zone (HAZ)-plate material or forging its lifetime (Note 2). De beltline materials shall be eva! material extending outward from, but not including, the on the basis of initial reference temperature (RTm), weld fusion zone in which the microstructure of the base predicted changes in the initial pmperties as a function TatSgec j metal has been altered by the heat of the welding process. chemical composition (for example, copper (Co) and .il Type. 4.9 leadfactor-the ratio of the neutron Dux density at phorus (P))(Note 3), and the neutron fluence during the location of the specimens in a surveiBance capsule to the operation. eens correst liethods A 3 neutron flux density at the reactor pressure vessel inside Nort 1-The bue metal for tbe weld heatdcted-rone (HAZ)a 1 85 surface at the peak fluence location. monitored shad conespond to one of the base metals ad M for 4.10 neutron fluence-the time integrated neutmo flua surwmance prosmn size an i Nort 2-The data used for the selection of surwine test as ad shape de l density, expressed in neutrons per square metre or neutrons per square centimetre. M.sha:1 be that obtained in accordance with ASME Code SectionI needed. A W 4.11 neutron flur density-a rucasure of the intensity of
~~3"#y*'~~
residual /anoy etements such as Ni. Si. Ma, Ma,a V nc ! p neutron radiatica within a given range of neutron energi$ C, S. and V may contribute to oversn radiation behavior of Ar:s to. l the product of the neutron density and velocity, measured in materiak neutrons per square metre-second or neutrons per square 5.1.2 The base metal and the weld with the highes adjusted reference temperature at end-of-life shall be sel Wf 41 neut spectrum-the distribution of neutrons by fw &c meWance pmgram. K ec Charpy upper gg energy levels impinging on a surface, which can be calculated energy of any Ithe belthne materials,is predicted to drops plates : based on analysis of multiple neutron dosimeter rneasure-a marginallevel(currently considered to be 68 J (50 A Ib0s ating ments, on the assumption of a fission spectrurn, or from a the quarter thickness (% T) location) during the op ,,,, t calculation of the neutron energy distribution, etime dbe s mel, pmisions shall be made to also mcid 12.7 r ! i 4.13 nil-ductility transition temperature (Tm}-the that material in 6e sunemance pmgram, preferaW in 6 g,gg,, maximum temperature at which a standard drop weight fmm of fradun toughness speciment Rese aW sot! l specimen breaks when tested in accordance with Method sp cimens may substitged m part fu specimens # surface ai I E 208 material least Rely to be imutmg. Mes, or nor f 4.14 reference temperature (RTm)-See subarticle NB-5.1.3 The adjusted reference temperature of the materid g; 2300 of the ASME Boiler and Pressure Vessel Code, Section .m e nada vessel beltime shall be determmed by aM8 ggg > 111. " Nuclear Power Plant Components.. the appropriate values of transition temperature shift to
~~be orim -~~
~~4.15 transition temperature shyt (ARTm) or adjustment reference temperature of the unitradiated material T Ethe HAZ~~
of reference temperature-the difference in the 41-J (30-transition temperature shift and Charpy upper shelf eneT { perpedie

ft lbf) index temperatures from the average Charpy curses dmp can & dete,rmined from relationships of fluence d b,3,3gg measured before and after irradiation.
~~chenu, cal composition. kommeMo ) 4.16 transition region-the region on the transition tem-M #*'"I#I N~~
~~b."#~^ "I"Imum test pmgram ns is.~~
per.ature curve in which toughness increases rapidly with c nsist d 6e matenal selected.in 54 taken from # be N ' rising temperature. b,d change from a pnmarily cleavage following locations: (I) base metal from one plate or fagi8 tenns og fradure appearance, at is characterized by a rapi used in the beltline,(2) weld metal made with the same W j j (crystalline) fracture mode to pnmasily shear (fibrous) frac-of weld wire and lot of Dux and by the same welding prs:@ as that used for the selected beltline weld, and (J)
~~ture mode.~~
4.11 Charpy transition curve-a graphic presentatson of g g, g4y y .d Charpy data, meludmg absoibed energy, lateral expansion, ,3gy,' and fracture appearance, extendmg over a range including 5.5 Archive Materials-Representative test stock tom tbc lower shelf energy (<5 % shear), transition region, and least two additional capsules with test specimens of the W# the upper shelf energy (>95 % shear). metal, weld, and beat-affected-zone materials used i8 4.18 upper shc7energylevel-the average energy value for program shan be retained with full documentation M all Charpy spectmens (normaDy three) whose test tempem identification. It is recommended that this test stock be is
ture is above the upper cod of the trans: tion region. For form of full-thickness sections of the original ms38M specimens tested m sets of three at each test temperature,6e (plates, forgings, and welds).
set having the highest average may be regarded as definmg 5.6 fabrication History-ne fabrication history (s# 1 6e tapper sheEenergy. itizing, quench and tempering, and post-weld best ment) of the test materials shan be fu!!y representative d ,t. 5, Test Materials fabrication history of the materials in the beltline DI# i 3 5.1 Materials Selectiorc reactor vessel and sha!! be recorded. 5.1.1 Surveillance test materials shall be prepared from 5.7 Cherr.ical Analysis Requirements-The samples taken from the adual materials used in fabricating analysis required by the appropriate product specif8 3 l'* the beltline of the reactor vessel. These surveillance test for the surveillance test materials (base metal 350 1 )
EA-GL 92 01 01 Attachmord 4 Peps 1 cf 4, tev s l i L l t i i ANALYSIS OF CAPSULE 125 FROM THE -l2 CONSUMERS POWER COMPANY BIG ROCK POINT 'A NUCLEAR PLANT REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM (WCAP 9794) %[ l ,s EPRI RESEARCH PROJECT 10213 TOPICAL REPORT September 1980 7., i Prepared by WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P. O. Box 355 Pittsburgh, Pennsylvania 15230 ' I T. R. Mager, Principal Investigator N$ 2 Prepared for ELECTRIC POWER RESEARCH INSTITUTE l i 3412 Hillview Avenue Palo Alto, California 94303
~~a' T. U. Marston, Project Manager~~
.g. U, r 1 s-a I Y -a j. *> i
... ~ EA GL-92-01 01 - ?? p Pese a of 4, ser e .i u i t LIST OF TABLES i i Title Page-i Table f f 41 Chemical Composition and Heat Treatment of l.j Big Rock Point Reactor Vessel Surveillance 4-2 Material 51 Big Rock Point Charpy V-Notch Toughness After Irradiation to 2.27 x 1019n/cm4 (E > 1 Mev) 5-4 5-2 Instrumented Charpy impact Test Results for I' 55 Big Rock Point Capsule 125 Vessel Base Material f i 53 Instrumented Charpy impact Test Results for 56 Big Rock Point Capsule 125 Weld Material 5-4 Instrumented Charpy impact Test Results for Big Rock Point Capsule 125 Weld Heat Affected 57 Zone Material 5-5 Summary of Big Rock Point Reactor Vessel Surveillance Program Charpy impact Test Results 5 56 Tensite Test Results from Big Rock Point 59 Capsule 125 61 Nuclear Parameters for Neutron Flux Monitors 6-6 l 67 6-2 21 Group Energy Structure 6-3 Irradiation History of Big Rock Point Surveillance 64 Capsule 64 Measured and Saturated Activities of Fast Neutron 6-13 Flux Monitors Removed from Capsule 125 i 6-5 Spectrum Averaged Reaction Cross-Sections for ,l 6-14 Big Rock Point Capsule 125 I 64 Summary of Neutron Dosimetry Results for Big 6-15 l Rock Point Capsule 125 i 4 i [ l i
7 I* t.A GL N 01-01 i f, Pese 3 of 4, tw e TABLE 51 j BIG ROCK POINT CHARPY V-NOTCH TOUGHNESS 18 N/CM2 (E > 1 MEV) AFTEf1 IRRADIATION TO 2.27 x 10 [ Base Metal Lateral Specimen Test Temp. Energy Expansion Shear Number ('C) (*F) (J) (ft4b) (mm) (mils) (%) l 41 A -18 0 13.6 10 .14 5.5 0 414 10 50 31.2 23 .51 20 5 418 24 75 28.5 21 .39 15.5 10 41C 38 100 24.4 18 .52 20.5 12 41D 52 125 47.5 35 .69 27 19 415 66 150 59.0 43.5 .99 39 36 413 66 150 51.5 38 .93 36.5 30 412 93 200 86.8 64 1.37 54 100 411 121 250 94.9 70 1.61 63.5 100 i 416 149 300 90.8 67 1.52 60 100 l Weld Metal 4K6 -46 -50 23 17 .39 15.5 0 t 4K3 -18 0 23 17 .24 9.5 5 4K2 10 50 25.8 19 44 17.5 -10 4)Y 24 75 58.3 43 .95 37.5 52 4JB 38 100 66.4 49 1.09 43 49 4JC 52 125 61.7 45.5 1.07 42 62 4K4 66 150 73.2 54 1.31 51.5 72 4JT 93 200 74.6 55 1.35 53 85 4K1 107 224 94.9 70 1.63 64 100 4)U 121 250 104.4 77 1.83 72 100 4K5 149 300 99 73 1.79 70.5 100 HAZ Metal 552 -46 -50 28.5 21 .33 13 0 51C -18 0 31.9 23.5 .30 12 5 555 10 50 36.6 27 .63 25 26 554 24 75 69.1 51 .76 29.8 50 557 24 75 56.9 42 .97 38 48 556 38 100 74.6 55 1.18 46.5 41 54 U 52 125 70.5 52 1.02 40 45 551 66 150 115.2 85 1.77 69.5 100 i 51B 93 200 86.8 64 1.33 52.5 92 553 121 250 94.9 70 1.50 59 100 54T 149 300 134.2 99 1.94 76.5 100 5-4 l
_ _ ~... 7 F* EA a-92 0101 +g Pese 4 ef 4, aw e i TABLE 54
~~SUMMARY~~
OF BIG ROCK POINT REACTOR VESSEL SURVEILLANCE PROGRAM CHARPY IMPACT TEST RESULTS t 41 Joule 68 Joule 30 ft-Ib 50 ft-lb Upper Shelf t Trans. Temp Trans. Temp Energy Fluence increase increase Decrease Material (1019 n/cm2) (*C) (*F) (*C) ('F) Joule (ft-Ib) i Base metal / 09w. 5 6' O O O O O O s e r.w.7167 33 60 44 80 12.2 9 f > #25 a 2.2719 67 120 67 120 19 14 / irt A 2.3 G1 39 70 50 90 27.1 20 isfA40.7 47 83 150 75 135 16.3 12 Weld metal .15 31 55 56 100 20.3 15 .71 75 135 78 140 33.9 25 2.27 75 135 83 150 28.5 21 2.3 72 130 111 210 51.5 38. I*I 94 170 128 230 40.7 30 10.7 ~ HAZ .15 0 0 0 0 0 0 .71 14 25 33 60 29.8 22 2.27 61 110 72 130 27.1 20 2.3 44 80 72 130 28.5 21 10.7 89 160 106 190 33.9 25 Standard .15 8 15 11 20 35.3 26 reference .71 22 40 22 40 19 14 Tears. ten iernverawe.ncreases hie ty a*st>onaue due to tarse watter en osta. n a. 5-8 j
o v..v.qt, .s ) v DPERATING DATA FOR THE MONTH OF ) ,$4 ' FEBRUARY 1979 YEAR ^ ,'h .. Off MONTH TO DATE CUMULATIVE ) SERRS'MICAL. 29.7 773.7 98899.4 N TING 29.7 773.7 96719.7 6 ICAL GENERATION,MWHE 1444.e 42914.0 5457942.0 f1ET*FEANT ELECTRICAL DUTPUT, MWHE 1321.4 39335.1 5353998.4 k-N STATIW1 POWER,lRIME43N DAYK GENERATING 122.6 2474.9 393063.4 W.' -13N DAYK MOT GENERATING 529.5 529.5 28124.9 'eD REACTOR'lEAT PRODUCED,MWHT 4748.8 136829.7 17959371.2 r !4P'(StIT CAPACITY FACTOR, X (MDC) 3.1% 44.1% 56.3% 30.3% 30.7% 31.7% D GROSS UNIT EFFICIENCY, % gCYCLE HEAT RATE, BTU /KWH (GROKK) 11271.4 11115.7 19774.3 12317.2 11871.6 11384.1 I' e PLAIET M AT RATE, BTU /KWH (NET) 4.4 54.6 72.2 ItEACTM AVAILABILITY, X 4.4 54.6 83.5 [, NIM-GEMRATOR OPERATING AVAILABILITY, % 0 0 489 TIM M CRITICAL 1 1 119 TIas$ TRIPPED (CAUKING BLADE MOTION) 14931. NWD/KT CORE AVERAGE EXPOKURE ,eEXPWIME FOR LEAD F AKKEMBLY,(F57 ) 20034. MWD /KT m' 'EXPOSINtE FOR LEAD G-LEAD AKKEMBLY,(G92 ) 27860. MWD /KT G1-PU AKSEMBLY,(G21 ) 23384. MWD /KT F*FORLEAD FOR LEAD Gi-U AKKEMBLY,(G292) 15302. MWD /KT 7567. MWD /KT FOR LEAD G3-U AKKEMBLYs(G399) p ?. . i. e 4a v iY . t.' ' b. {. ", 2 ,.'.'t,.>- g 2 s*.. ij',.;.%',;* ?i . ')
~
~~LA r,L 92 01-01 Attadiment 5 % 2 d 5, an e PLANT OPERATING DATA FOR THE MONTH OF. DErIDGER 1992 - CYM F 26 i t i*f.*;l. .'=. YEAR~~
~~a l~~
~~} MONTH TO DATE GQtglLAT.I.E. ! / l 663.7 4790.5 188159.6~~
~~HOURS CRITICAL 654.1 4700.6 185145 7~~
_ HOURS GEN RE ATING 46716.0 287419.0 11272085.0 2; GROSS ELECTRICAL GENERATIONsMWHE 44361.7 271422.8 10663180.4 W NET PLANT ELECTRICAL OUTPUTS MWHE STATION POWERS MWHE-ON DAYS GENERATING C 51,3 15.93.A.2,_ 608t o4..._6 _. l -ON DAYS NOT GENERATING 92.6 3156.8 51626.4 196262.0 896869.0 35494707.0 1" REACTOR HEAT PRODUCEDsMWHT _ NIT CAPACITY FACTOR 1 (MDC) 89.0% 46.1% 60.7%_.( U 31.5% 32.0% 31.8% GROSS UNIT EFFICIENCYs %
~~' CYCLE HEAT RATES BTU /KWH (GROSS) 10831.8 10650.0 10747.2 W ANT HEAT RATES _jTU/KWH (NET) 115106_.6 11277.7 ii360 9,..~~
t 89.2 54.5 74.1 REACTOR AVAILABILITY, % i TURBINE-GENERATOR OPERATING AVAILABILITY, 1 87.9 53.5 81.6 j" i 13 569 ~l 5 TIMES CRITICAL O 3 156 1 TIMES TRIPPED (CAUSING BLADE MOTION) ):[ 12 13 I4 IS ISS I6 13007. MWD /ST CORE AVERAGE EXPOSURE 3
~~EXPOSURE FOR LEAD I2 ASSEMBLY s (1206) 24278. MWD /ST i~~
EXPOSURE FOR LEAD 13 ASSEMBLY,(1313) 23271. MWD /ST 4 - EXPOSURE FOR LEAD I4 ASSEMBLY s (I407) 19942. MWD /ST EXPOSURE FOR LEAD 15 ASSEMBLYs (I503) 12263. MWD /ST lEEP_QSUBE_lpR LEAD 155 ASQiRLY, (111_9) 102211rW.CL/_ST. l EXPOSURE FOR LEAD I6 ASSEMBLY s (I608) 4155. MWD /ST i 'i
~~s FW/(5TM - CRD) = 1.005 l~~
~~.4 3:s 1 a )n. y~~
~~f l~~
i )"5 l r J . cf J i a 1 evi m 1 .y l e, w ]
EA-GL.92 61.g1 Att.ciiment J.. e >.o. ANALYSIS OF CAPSULE 125 FROM THE CONSUMERS POWER COMPANY BIG ROCK POINT fk NUCLEAR PLANT REACTOR VESSEL i RADIATION SURVEILLANCE PROGRAM .g, (WCAP 9794) c $^ 'i EPRI RESEARCH PROJECT 10213 'f;h 4 TOPICAL REPORT September 1980 Prepared by WESTINGHOUSE ELECTRIC CORPORATION j Nuclear Technology Division 1 P. O. Box 355 f Pittsburgh, Pennsylvania 15230 T. R. Mager, Principal Investigator i,.9 S* :j Prepared for ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue Palo Alto, California 94303 [. e T. U. Marston, Project Manager s i
ke
+ .y ? i
o EA-OLo92-01 01 Page 4 of 5, sev 0 LIST OF TABLES Table Title Page Chemical Composition and Heat Treatment of 4-1 Big Rock Point Reactor Vessel Surveillance 42 Material 51 Big Rock Point Charpy V Notch Tou hness After Irradiation to 2.27 x 1019n/cm (E > 1 Mev) 5-4 1 52 Instrumented Charpy impact Test Results for Big Rock Point Capsule 125 Vessel Base Material 5-5 5-3 Instrumented Charpy impact Test Results for Big Rock Point Capsule 125 Weld Material 5-6 Instrumented Charpy Impact Test Results for 54 Big Rock Point Capsule 125 Weld Heat Affected Zone Material 57 5-5 Summary of Big Rock Point Reactor Vessel Surveillance Program Charpy impact Test Results 5-8 5-6 Tensite Test Results from Big Rock Point 5-9 Capsule 125 6-1 Nuclear Parameters for Neutron Flux Monitors 66 67 62 21 Group Energy Structure 63 Irradiation History of Big Rock Point Surveillance 64 Capsule 64 Measured and Saturated Activities of Fast Neutron Flux Monitors Removed from Capsule 125 6 13 6-5 Spectrum Averaged Reaction Cross Sections for 6-14 Big Rock Point Capsule 125 6-6 Summary of Neutron Dosimetry Results for Big 6-15 Rock Point Capsule 125 i V i
gg. a.92-01 01 I tage 5 of 5, tw 4 TABLE 63 (cont) IRRADIATION HISTORY OF BIG ROCK POINT SURVEILLANCE CAPSULE P P trradiation Time Decay Time 'I I j max Month (Mw) (Mw) P /P (days) (days) j max 7/77 109 240 .4 54 31 897 8/77 - 9n7 0 240 .000 61 836 10n7 20 240 .084 31 805 11n7 180 240 .752 30 775 l 12n7 201 240 .838 31 744 l 1/78 83 240 .346 31 713 2/78 187 240 .778 28 685 3/78 209 240 .872 31 654 4n8 187 240 .779 30 624 5/78 207 240 .864 31 593 i 6/78 193 240 .803 30 563 7n8 200 240 .834 31 532 8/78 199 240 .830 31 501 9D8 26 240 .108 30 471 10/78 0 240 .000 31 440 11/78 180 240 .750 30 410 12n8 192 240 .801 31 379 In9 2n9 175 240 .729 32 347 8 Total EFPS = 2.72 x 10 l l a. Decay t.rre <s referenced to Ja% ry 15. 1980. 08 00 l 6-12 I
EA CL 92-01-01 . UNITED STATES Attedmont 6 NUCLEAR REGULATORY COMISSION t et 2, w e WASHINGTON DC 20555 .;,G ^ pU CONSUMERS POWER COMPANY DOCKET NO 50-155 BIG ROCK POINT PLANT FACILITY OPERATING LICENSE License No OPR-6 A. This license applies to the Big Rock Point Plant, a boiling water reactor and associated equipment (the facility), owned by the Consumers Power Company (the licensee). The facility is located in Charlevoix County, Michigan, and is described in the licensee's application dated January 14, 1960, and the Final Hazards Summary Report; as supplemented and amendedby subsequent filings by the licensee. B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Consumers Power Company: (1) Pursuant to Section 104b of the Act and 10 CFR Part 50, " Licensing of Production and Utilization Facilities" to possess, use and operate the facility at the designated location in Charlevoix County, Michigan, in accordance with the procedures and limitations set forth in this license; (2) Pursuant to the Act and 10 CFR Part 70, "Special Nuclear Material," to receive, possess, and use at any one time up to (a) 2500 kilograms of contained uranium 235 as fuel, (b) 10.32 grams of uranium 235 as contained in fission counters, (c) 150* kilograms of plutonium contained in Pu0 -UO fuel rods, and (d) 5 curies of plutonium encapsulated as a p$uto nium-beryllium neutron source, all in connection with operation of the facility, subject to the following conditions: (a) A technical specification will be instituted to assure that the plant will not operate if the capacity of the makeup line to remove heat frcm the spent fuel pool is insufficient for the heat generating capacity of the pool inventory, con.'idering the power history of each assembly in the pool, the number of assemblies in the pool and the effect of lake temperature on the heat-removal capacity of the makeup water tystem.
~~Note:~~
(For internal CP Co clarification.) Amendment He 4 dated 12/6/72 established that CP Co could have 150 Kg of Pu in mixed-oxide fuel bundles. However, this was limited to 50 Kg until the GESHO question was finalized. Reference 10/13/75 letter and NRC order dated 8/11/75. Amendment 107 February 19,1992
. FA-GD-92-01-C ~Y,. W[4p _. ;...' 3h[k;_ ',.. Attaciummt, P = page"2?'ar'2, arv ; w 4 ~ ,,,,,,u,. nt is effeitiveyacf d,.,as olthe,$dgmr7 '~ ^ ":Q" d**hf... ~ date o D. This license ame shall expire at' midnight, May 31, 2000.W E k*s#! % i .,ge - ' j ~ s., 4..._., FOR THE NUCLEAR REGULATORY C0lWISSION i L 8 Harsh (Signed) I j L B Marsh, Director i Project Directorate 111-1 j Division of Reactor Pr.)jects - III/IV/V Office of Nuclear Reactor Regulation l i i
~~Attachment:~~
Appendix A Technical Specifications l Date of Issuance: October 19, 1992 / r 1 4 i I e t a I -1 I i Amendment M7,108 October 19, 1992 l
u.a.,2... 3 Paes s.e 2, en e M..,. NT-(. ess: ~ pek 1 . 52. ': Qlbs - h~ip f 5 ~
~~1' ~.' h"i..~~
~~A %.'.;y. -~~
~~3..1.~~
4 T'i.. Director.'dr.Weleak;<;,A tivitie... h.,./- .e. . )tr.' R.9 E...KettnerJ- . +. i d ~ s s / 'A ;' '; . f. - i 4 .:pi ) a e..' '1 %: 7 {d' Consumers' Power Cc' p'any212 Weat Michis nMvenue '. ',.%,j @jj $$ 1
l s

c
~~-W' r i- > '. Jackson,' Michigan\\'P..'\\ '.Y '.: i.%%.... 4'.f: '..:. ~ ' *:,- > %. kr 9.f. ~'6 6, *:e ' .;~% L ',SubjectrF BIO' ROCK' POINT M~~
~~PLA:c..;ngr;qc3~~
^ - '( , ;i,J: ~ VE3SEL MA72 RIALS
PRO 3RAl*
~~.' a -~~
~~Dear Sob:~~
. An apprcptiate reactor vessel su.ve111ance pro ~g:am for the Big Rock Point Icactor would consist of the i: radiation o.' samples in the Consumers'. reactor for a period of about 32 years. Tne proposed details o.' this progran a:e attached. follows: A sumr.ary is as ~ r: .l. Sampling period,s'; 1, 2, 4, 8,15, and 32 years. 2. Samples will be% harpy "V" notch and tensile specibens of'; ~ . Base ;Ietal,. Weld. Metal, Heat-affected zone, and Standard Heat of *%tei'ial; .. : p.f. ; g. /:3 ..? 7' .{ .?, R3 .)
~~t Each capsule w lb ontain 12 cir~~
~~' y and 3 tensile specimens-for a total o( samples a~~
~~34 apsules for the 32 year Program..,?~~
~~, i-g .J j,. ..p~~
~~4.. Sample capsules will be locateil inside the thermal shie94 to Sive a 10:l'rhtio when compared to the vessel; and insiith the~~
., vessel >:all toi 1 ratio yhen cocipared to the vessel. ,. Also capsules %give a 1.25: ill be placed. in.the reactor. steam line chamber i effects and iio 'ir' radiation. ' '. ~ to give. therma'l[.Q': '~ ^ ^.pp - l *, . r., 5 control specimeris.neither thermally exposed af ter machining. nor iriadiated will be analyzed during,the program..- ,~ A detailed'p}os._ 'hdiation,, examination'.will 6e conducted:. -i :
~~.:, W~~
~~. 6. ~ a tenaile,W.%*k;%.&,,h %:. :i[. f p?W. if,.- and metalingraphic' tests.~~
~~including im~~
~~!g_ ?, k =. c=ing. ' $ (to 'prepdratf9n;~~
~~14M at.~~
.s;.x les and ca7sulp: ?. 3 1 Time Md unde take isfa I'7,y L pro 4, a S% .,. y m 1:' t. $j y.:. c e? U ..} -~ .v:,: -1 n
I n.rA-92 0101 . > - ~Y::..i. &.. \\, ...n...... ..:... :y ,..W v. ., I h. s. 7, 7ebruary 26) 19M '. >.". ' 2.:- E. E. Kettner s. It is recocznended that youir.stitute this program in order to provide authoritative information en it radiation eNects for Ic
~~plant operation, state and rational re.;ulato:y organir.stions,~~
~ insu::ence purposes and to Jurther ',c.owled,e for the acomic industry. It is proposed that you bear the cost of the prepa:ation or sampics and encapsulation. APED ifould provide the scaples,. do the necessary work to engineere the details of'the ' program and w l examine the p:operties.of the contiol specinens not subject to
~~le would like to :eceive, as soon as possible, you.~~
1:cradiatio:.. thoughts on this surveillance prog.am ahd will be happy to quote, l through the appropri. ate channels, the cost of' suitable sample preparation and encapsulation lor the B16 Rock Point.eactor. i Very truly you;s, /b" (7. A. Hollenbach, canacer NUOLEAR PLANT SYSTENS PE0JECTS FAH:lfIH/el t 6 L \\ Enc 1 l. A_ ,..u. r ,..: 4 't s sj -... ..g a St . ~ ?- ^' . 1g. f, : ,,,.y ~ >. a., , >.a^. !,.= i4'. 't','5f? b M..b. .5 ~ ::^ ? b Y.. Yh:" J' .? ~ .+ f w. .a....t;v.uaav.~y:m:....n -w:.a. c.;. c.w;. 8 v- ,.-;.n 4yp ..q . n
~~c...,~~
~~.f. $ . as, .,,75.~~
~~.g.. 4,.~~
.e. c. m: F
' *~ * r 1 l I i m;
y..

Q ;
~~1 i . p.. t 7'~~
.er 74 6, a M >' ~
~~- n..:1.9Y-y.[. ' O-'C.') p, C. i 4 i 2.. d h,. '*',:;,, l~~
~~. 7,..,,,~~
~~M yPN. u.5, pd %*.. 2.c-4 .Ig;, y,., q, j ~, % m \\ ...x;.. .*et.-<--~~
~~.r..~~
~~,1{ v 4f. . z.. n?;.j .e-h,: u.*f;. ,s~~
s
~~.h., t.,. _,* ~. g a * *t D .w,T;, ' "tffafl5;rC. vg*-~~
~~m. '~~
~~t- ': ' ~ ~< '19 x,; 6 " - i r. . * *I 4.*, ' -c, f~~
~~~ ff L 'q,~~
~~,tq. ~ fM.k,...j.,/g,4 E. SHE., C. Z. Sriw, Jn.. 'R. A. Qppy, Ja.,N;P A. Sm - 2. * ;, ) k.u5"/INM.3 I r 1~~
~~p. bA~~
~~.wr r'... M, Ns. m,.. %:p 4."., 3: .. '#.. v tW2%N.~~
?-
^ Riactor Materials Branch .. m... y. ~, e Metallurgy DMslon .yz.. .. x../s v::.y:- -... :, a*rg~... v -.... w. - .c ~ g '. m..w e. ,,- y:~ r ?..w - *,'.h$5[&- ~ ;s ' '%U' f t h **.U ',g, f ., 7,. ~, - .y . ~~. .. ~. Ap ,1 1 i J i e 1 h i j August 1969 i l f i* I e j i I e ) -e ,f d \\ d i7 g' g gs Q zn ~ Y N l i... %g .s l
4-

4.,,
1 h.. . h*
)**
~~i ,9 .t.,. p ,.g ;. 7 - f !!.iM '. '9 N.h. VAL RESEAR.C..H..I.A...BORdOR,Y ' 4 ,. -h... ..~ a.^......--. .g.... ;5 0 '. ~ \\ -m i.-e v'.., r..., ua... ,...,.u ~ Q.,q...,f.' ~ *1 ..-.l.~~
*[ 'l.of
%Q RC,'* m Axanc e.t has been apprord for r.Mc n1c.de an.4 5 k. ns distr:5/., n is t e.'M-d I t l ij
'* ~
~~fQ-yyy,~~
~~,.g '0 Table 9.2 aC 4 stW A.2);.5". y~* g .? ,.;t ?,, t gt t.h;',,.nr : y Q v ~ ,odO - i.,Suasipurg of4Charpy-V Notch Ductility Data from -+;~~
~~y g'0 y~~
~~+.t (. e ;.st, r- ^ g a ,y j.:,,~~
~~gTc 57~~
~~+ U C 30 f t-Ib Transition Full Shear Energy~~
~~jft y e.#.?~~
fp y Y;, 4 2 Temprat m 7 Abs orption, ft-lb vem y1 ,,y I 5, }> f 'f* 'jki-) E " - 68 mb Unirrad. I r r a il. OT Unirrad. Irrad. t; ji' ,l:)} %* *~ s i L{,f Y. ff"?' ??. i r,l',i ' B&se Meta 1 ' Thermal Control -5 10 15 H2 88 r 6..
~~8. A 2-~~
-5 ~0 82 ~ 82 j :$f' k) jNS.s 1.5 x 10 -5 ~ 7.1 x 10 -5 55 60 H2 73 1 hI(,g ((#2;p 2.3 x IO -5 65 70 82 62 I '$(j.j:f.",, 3 -} 1.07 x 10 -5 145 150 82 70 f+Mf; .l, W I WIh:W. 'c dweld Metal Thermal Control -70 -55 15 95 92 1 % ew 1.5 x 10 ~70 -15 55 95 80 1 [. w.p'. ' h,k 5 'v.. 33 -70 (15 135 'la 70 - n - 3gi y gyg '., 7.1 x 10 ....y' 19 -70 120 190 95 57 -* ' i.! 2.3 x 10 1.07 x 10 ~7 0 160 230 95 - 65 20 h.... [:. ,s. :. y'. 3g Heat Affected 1.5 x 10 -55 ~-55 0 100 ~100 J 87 18 kne 7.1 x 10 ~55 -30 25 100 78 1 [ k. 2.3 x 10 -55 15 80 100 79 1.07 x 10 -55 105 160 100 75 0 !!..an9 ~ 18 5tandard 1.5 x 10 0 15 15 96 70 3 18 l "Aeference 7.1 x 10 0 40 40 96 82 jj g& [ev a x- . 5*. f hi; s Includes 15*F AT from thermal aging. sw M.3 l4.n '.*flj1 Q'. n..
ff
~~..,e ...,.D.. ,_-_m..m ... s.-... . ~...,. -. -,.... .,,,_...__.,,._-..._..m.~~
~~,i '~~
e .,w. W. e t ANALYSIS OF CAPSULE 125 FROM THE CONSUMERS POWER COMPANY BIG ROCK POINT NUCLEAR PLANT REACTOR VESSEL v RADIATION SURVEILLANCE PROGRAM (WCAP-9794) t EPRI RESEARCH PROJECT 1021-3 TOPICAL REPORT 't Septernber 1980 P Prepared by WESTINGHOUSE ELECTRIC CORPORATION I Nuclear Technology Division P. O. Box 355 Pittsburgh, Pennsylvania 15230 T. R. Mager, Principal Investigator Prepared for ELECTRIC POWER RESEARCH INSTITUTE 3412 Hillview Avenue Palo Alto, California 94303 T. U. Marston, Project Manager 1
~~r..,,.~~
u o TABLE 51 BIG ROCK POINT CHARPY V NOTCH TOUGHNESS AFTER 1RRADIATION TO 2.27 x 10 N/CM2 (E > 1 MEV) 19 Base Metal Lateral Specimen Test Temp. Energy Expansion Shear Number ( C) ('F) (J) (ft-lb) (mm) (mils) (%) 41A -18 0 13.6 10 .14 5.5 0 414 10 50 31.2 23 .51 20 5 41B 24 75 28.5 21 .39 15.5 10 41C 38 100 24.4 18 .52 20.5 12 41D 52 125 47.5 35 .69 27 19 415 66 150 59.0 43.5 .99 39 36 413 66 150 51.5 38 .93 36.5 30 412 93 200 86.8 64 1.37 54 100 411 121 250 94.9 70 1.61 63.5 100 416 149 300 90.8 67 1.52 60 100 Weld Metal 4K6 -46 -50 23 17 .39 15.5 0 4K3 -18 0 23 17 .24 9.5 5 4K2 10 50 25.8 19 .44 17.5 10 4JY 24 75 58.3 43 .95 37.5 52 4JB 38 100 66.4 49 1.09 43 49 4JC 52 125 61.7 45 5 1.07 42 62 4K4 66 150 73.2 54 1.31 51.5 72 4JT 93 200 74.6 55 1.35 53 85 4K1 107 224 94.9 70 1.63 64 100 4J U 121 250 104.4 77 1.83 72 100 4K5 149 300 99 73 1.79 70.5 100 HAZ Metal 552 -46 -50 28.5 21 .33 13 0 51 C -18 0 31.9 23.5 .30 12 5 555 10 50 36.6 27 .63 25 26 554 24 75 69.1 51 .76 29.8 50 557 24 75 56.9 42 .97 38 48 556 38 100 74.6 55 1.18 46.5 41 54 U 52 125 70.5 52 1.02 40 45 551 66 150 115.2 85 1.77 69.5 100 51 B 93 200 86.8 64 1.33 52.5 92 553 121 250 94.9 70 1.50 59 100 54T 149 300 134.2 99 1.94 76.5 100 5-4 I}}

ML20012G458

Contents

Text

Response

Response

Response

Response

Response

Response

Response

REFERENCES:

:

SUMMARY

Attachment:

Dear Sob:

Navigation menu

ML20012G458

Text

Response

Response

Response

Response

Response

Response

Response

REFERENCES:

:

SUMMARY

Attachment:

Dear Sob:

Navigation menu

Search