ML19274F085

Domestic Licensing of Production & Utilization Facilities: Fracture Toughness Requirements for Nuclear Power Reactors, 10CFR50

ML19274F085
Person / Time
Issue date:	05/09/1979
From:	Chilk S NRC COMMISSION (OCM)
To:
Shared Package
ML19274F083	List: ML19274F082 ML19274F085
References
RULE-PR-50 NUDOCS 7906130077
Download: ML19274F085 (20)

Text

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e Arrernts 0--FnacTtr.s Tovoiurs.a RBetFMEMSJr?.

L GA v

..wM.2r3 rCoPS (345 MPd. and to those with specified minimum nt ap:.a

.cc aununum fr.ciur.

ns.cha.

r.ca e.m.au ror r.mta m.w.

yleld strengths greater than 50,000 psi (345 MPa) tou 2uu or pra.or..,.t. :a co==.

u a: e.

r ctor ca a pr.uur, o uce rr.er ".*:.'

but not over 90,000 psi (621 MPa) if qualified by ooot.a.po..e r..c.cn to -c.

using methods equivalent to those described in o.

norms op.r.uon. mciu ar :cocm2u awca or..r.tr amn cr

.n e atice."* *-

paragraph G-2110 of the ASME Code (defined in

.r.itoo u occurr.ne...na.

.t. tie t. u. to.neh..th. rm.rit.m hrttro.

.pa. agraph II.A. below).1 ur. ** ***'.

, iu ur.= w".ce.7 r

m.,

n. suoi ci.4.

n. r.2uu e...u or int..p:.cau.ppir to th. foslo.tDf m.t.rt.JS

• A. Carbon.nd ;aw..' lor f.mtie ste.1 plat..

corging., ca.3t:rg. and pt;e with rpectC.4 mintrnw23 rt.ld.tr.ng*hs not o..r 30,000 pT1 B. W.ld.

3d v.l4 h..t..r!.ct.4 soc.a '

th. mater 1Cs r:9cL3e4 12 se; tion !A.

C. Ef.t.rt.ts fc,r bcitta g and ctaer tTpes of fast.S.r3 with.

=151 mum 11 14

.tr.nch. net..., Tect?..d (896 MPaj.

other r.ma,of th. o.:ao.ynuurie.r aut :. e.=me.tr.u4 i:

AC.qu.c7 f.c.ur. tought.

of to th. Cc:n:ntasten 03.A Inda.tdu.1 c.s b

ts.

1 The latest edition and addenda permitted by paragraph 50.55a(b) at the time the analysis is made shall be used.

79001soonp7,

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g.=

2

g

a. Asus cIeC. S'. Amertean Se=

[Ifnosectionisspecified,thereferenceisto

.'W e t*

e_"**sstoo a".mes ar

.tety of L!achanical rr.=ur. ve..e4 cod.D.ruct.u row., e.ua componenu. -

"Section XI" meansSection XI, Division 1, tse coutrucuca ot

~

osu2Toh

" Rules for Inservice Inspection of Nuclear i so.s s.. coa.i ntenu =.esma cueen and Power P1 ant CCmponents "

If no edition or a.

remue m ing su turae uio7. u ts. 4u ne '=*

addenda is specified, the applicable AS'1E low-alter ste.a. nainer aucy r,eia includ.

muscre and ereci,uuan r.uaenm.i cae=

Code edition and addendaA and Iim.itations msa a preatnumanur w rc.n=re coue and modifications thereof are

'"c.**P7.t.m tre.:au. te.sa means au I

s preoperattonal syn. m 'se.aan tai br*ro-statto pressure testa and sai r7. tam lesa=

age and hy3restatt: preraure testa portmed during the sert;ce 1Lfa of th e prttaure boundary La compliance with the.A3M2 Code. secuan XI. "Ru!es fer leurvee En= _

a spection of Nue:aar ~

Po'. der riant CCCponents A

wer-D. 'Spectf!ed minimum yteld stra n.gth*

mes.no the cunimum y;eid stronph (in the untrTadiated conction) of a :na*,rtal r?ecta fled in the construccon ecde 'anaer vntcn the compocent is Dutit pursuant L2150Ms.

= '

-. le

__.m 7-f",3 -neference to:ngerature

• meant the reference temperature, atTo ma de ned in a

p,

_ the A3MZ Coce.

^

diCoonas gnay be obta4aed froft t.he' Amer.

less Soc:ety of NoemanteAl 2::gaceers. United Engineering Canter. 344 East 47th Street.

New Tort. N.Y.10017. Cbptea are amushte for inspection at the Ccunmiselon's PubL:e Document Room.1717 H $t. N.W Washlag.

Lon. D C.

~$#

satC'..~.

3, Section 'l of this

1. • % ' Adj u eted reference tembersture

means the reference temcers Lre sga2s;C cor tiradiation seeet. e see, Appene s.W si addtnt to RTao, the tamoersture astti La the average the Charpy heesee cune rar the artsd:sted mater.u reau,. io unt rar e..,n-14cea

_3

--mater:st. measured,.at tee -

4. 30 ft lb. (41J) level-e s.m -senur.* ' -

~ ~ " ' ' "

+ho recion of the reactor vessel (shell Lemi asees. zone., mai cuerur orreu=J

$aterial including weld regions and plates or no ese:uve neig

,=rr.nne ans an,nt or tn. ruei.;emeni forgings) that directly surrounds the effec-uait:::a netcat or as.'

naterta ror waren e. pewie e.s tar.. mer tive height of the active core and adjacent er,iroter.ne. t.meeran e ai eos or,":c4 regions that are predicted to exper4ence suf-

--.orse.a r

=5 5 n--

mewresu mna ro, te. p.acement er r,a, tor ficient neutron radiation damage to warrant setune mat.r u specur..n. a see consideration in the selection of +he mas *w withdraw %s2 sqd teettst or such'apecimensi.==4 ene o m run c.rpeno4w r*=iar 1imitin9 material' with regard to radiation to monnor oee.. -nic. n:ror m.,e-i.

g chaeges ta the er ure tonganess propert.es damage.

or the heit1Lne as a It or e u osure t neutron tr?t41stion -

'e thermat environ nacs.

J.

"Integra surret..ce protrama neans the ce

.sst:on or '.:. *13ual.riate tal survet '-

s p NTama as a,

' M to or.i

'Ir me actor, ease.4 to rwd res..

e n tc.'

to mer tor taa casans It "m'

.f.7

-.s...

G 7-ge.'

s

u n

and, for the beltline materials, the test a nacme me====

4. n d.monatrat. compusace with th.

requirements of Appendix H.

For a reactor

-um imtu,w sougana.a requa. menu of.ecuona Iv and v of iba appendts far.

vessel, that was constructed to an ASME Code

",,"**'"""**"2",,"**'""***'"**

earlier than the Summer 1972 Addenda of the c

r 1971 Edition (pursuant to S50.55a), the program of fracture toughness tests and

_..m r_ 2..cc c_.r : c 7

}",8&"' g**,;; "3,i =*,'ag'Cf' analyses shall be supplemented in a manner 5

i.2 of the uMa Code. to adiuon. wm aoproved by the Commissicn to demonstrate air.d er the aMr Cod.. untrradta

. uc maansa shut s. a.imd er me

. compliance with the fracture toughness 7 *2tI'7tt ' E x*Cosr' O onasas requirements of this appendix.

= isade for supp:ementaa testa in er petaa atuabora euch u that dancribed in on 8.

Test methods for supplemental fracture

-r.C. g ' *~"

2gha 'J' te ah I*.*"In'fic"** 'n"' oore, toughness tests described in paragraph V.C.2.

di i

i mee 4th the foilowtns requinmanp:

shall be submitted to and approved by the

1. Lo4ation and orientauon of impgles te,

,eeune a shaan tempir with the quir Commission prior to testing.

nanta paragraph ;fB-2222 of M

Code.

2. uma riale used to prepare wet ep" C.

All fracture toughness test programs be representative of t$e act nana saa the naiaw n,on at u r l conducted in accordance with paragraphs n.

anais (

7u' ired nr n. apptleant. rui.e o the co.';

and 3 of this Section shall compi<v with ASME prueuen cede under wnich the ponen.

Code requirements for calibration of test is butit pursuant to $ $0 &Sm. vicopt th*

' emus matketna intended for *be reacto.

re ei wiuit. neon aui comfir with thd equipment, qualification of test personne.i, gduions r utamenta of cp m.C. o' and retention of records of these functions

2. Cuiersucw of umperatu tn,trum.nu and of the test data.

md Charpy Vknotch tmpact 4 machined Ased in tmpact testing shall mply with 12.4 wtusromenta a paragraph 2384 of the LSME Code.

4. Individunit rfortning fracture tough <

sens test.s shall 9p qualLS by tratnLng and papertence and sh411 have d monstrated com.

>etancy to perforn the to ta in accord wtt2 rritten procedures f the omponent manu<

acturer.

5. Practure toug es; fest resulta shall be

'ecorded and aball L$clude a certi$ cation b%

,Be Itcensee or per pert rming the testa

'or the licensee tBatt /

m. no tasta have fu performed in com<

>11ance with the quLrementa of thu bppend LI.

b. "fte test data at rrectly repor*.ed and identtSed with the tasterial intended for a pressure-retaintag cpp9nent.

c. The testa havet been conducted usica r.achines and Lastrumarntation with avm3<

knie records of per*pdic ca4Lbration. and

d. Records of We quajt$ cations of tBe ad171duale perforfnLag t.h testa are stad<

bble upon request nection mA. of y[o the test requirements adina appendtz. testa on maj C. In adottson jtertala of the reactor vessel bottline shall _og onducted in accordance with t.be foLlowing ir_tmum req mentat

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1. Charpy V notch (C.) !aadmet testa shall conduc at appropriate \\tamperatured over a temperature range auf:ctent to de**4 th. C. test cyrves (including t.h4 upper.ahe;(

levelsi La terms of both fracturt energy a.nd lateral espaAsnon of spectmana. Docation and d

nelantation of impact test specknana anaal compit with the requirementa or paragraph NB-3222 of the ASMZ Code.

2. Matartala uand to prepa.ro test specunens for the reactor vessel beltline region shal be taa en directly from escoes mate ttal and welds in the veneen shell courseten following motetion of the production :enstrudLnal eld ;o

t. and subjected to a heat trettmant hat p oduces metallurgtest efecta egutta.

est those produced in the vessel matarts.

throu hout tia f EDrication procesa. In actord.

knce tta paragraph 3rB-2211 of the.qM21 pde Where eencueen shelt forgings are t.tsedj or ere the same welding procese :a :1 sed) for ongttudinal and circumferential wald.sq pa :stes 13e test spect'nens may be taann from a separate weldment provided that such b eldment 14 prepared using escasa material the shall forging (s) or plates, ma appily reie, the same heat of 11:er material. and

'ne same production weldtag conditions sai

--v...

Q.

n. mcm. me

,,se. a.e-,r,e A, he pressure-rotalning componenta of th.,,seto, ee iant pr,s.ure b'"u"i -t me r fracture tcughness requirements of the ^' sic ma mad. of f.rnsi matenus ah

". dunng.r.sem hrdn.ut s ""an4anr Code supolemented as follows for

~

condition of normal operation. Including an.

ticipated operationaa occurrencesg"

_ _ - ~ _ _ _ _

dedards of paragraph N3-230 of the AASQ dode, and the requtrements of eegttonJJ V.A. 3 and 4 and !72. of this nypeijrdtz.

l j

(

2.

t vessele, esclustee of botting ct ot.hel yten rs:

/

l eC culated strees intanalty factdre shal 14 lome thaa the reference stresa /stensiti actors the margina spectaed in pse ASLO d

t' ode Ap ediz O. " Protection Agpnat Non Juctile Pd$ lure *. T12e callutation/ procedure.

hall comphy Flth the proceduresiepectf ed it he ASME e AppendLa O. byt editttonas and alte tre procedures map be used W the Comm on determines that they provtdli, equitaloit mpgtne of safety agatnat fracture, 7tmalag.ppropriate a13cwanc( for att uncerl ainctes in theidsta and anettfee.

h. Por nWes. flanges a84 abell rTgtoni 1 ear geometri disconticultJes, the data as

procedures tred in addttton to th ipectf!ed in th AShfE bde shall provid

nargins of safe comparabte to thm..

autred for stella god heads remota from dis

'ontinultles.

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c. Whenever the core /ts critical, the tne Lernperature of 12.s rvactor Teenel shall bq l

an adequate margicl high enough to prort of protection agaJ:1p fracture, taatng inta accou n t suca f acto as the potenttal fon overstrees and the I shoca duttog antici.

pated operattenal o $rrenas in the controt of reactivity. In go sang when the core ta erttical tother then foir the purpose of low-levet physice testd) shalt the temperature of

' L%e reactor vessel be las4 than the minimum I ermissible temperaturg (or the Lnservice sys-p -

ta'm hydrostaud pressur% test nor less thazu 44'P above tant terapetature required b) nection IVM.1

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d. If therefs no fuel La the reactor durtng the inttlal preoperational s9 tem tenha.ge and hydrostatte g:vasure teats, the minimum per-missible tot temperature uan de eeta r.

mined in Accordance wtth peragraph 0241d bf the ASMR Code except that the f actor cd ety swited to each term matlng up the cuistaa stress tatanasty factor may de r,-

uend t4 to. In no came aban tae i et sam-braturp be tema than RTaett l4*P.

f o

3. Materials for piping pumps, andl "thea, and materials for bolting and!

Oth r fasteners, shall meet t e rei jul ments of the ASME Code, tral gr.,hs NB-2332 anr4 Na ""

_ j,' " '

L.,i e

,y_

/..

s

o.

(' have minimum Charpy upper-shelf energy of

7. _4aP a**ctor venien seit::n. m.e.rt.:,,hau x- --

4 75 ft ib (102J) initially and shall main-f tain minimum upper-shelf energy througnout tre m en r n v.n m 3.,,.,,- :

y$'n**.ra-2EE **"'T.ggl,f the 1ife of the vessel of 50 ft 1b (68J),

i r unless it is demonstrated in a manner

! tt

• utan -

,.t.

,a to em iEE"' *. er.fl,','O $',*.'.,'.;,, g approved by the Commission that lower values T,, =ern aim er: vie.

4.w.2, rn.rre of upoer-shelf energy will provide margins

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of safety against fracture comparable with those required by para 3raph 2.

2.

When the core is not critical, pressure-temperature limits for the reactor vessel shall be at least as conservative as those obtained by following the methods of analysis and the required margins of safety of Appendix G of the ASME Code' supplemented by the Nguirements of Section V of this Appendix. The justification submitted for the pressure temperature limits shall describe the methods of analysis used and shall demonstrate that when the limits are controlled by the nczzles, flanges and shell regions near structural discon-tinuities, the margins of safety for those. recions are comparable with those required for' the belt-line when it is controlling.

3.

When the core is critical (other than for the purpose of low-level physics tests) the temperatgre of the reactor vessel shall not be lower than 40 F above (22 C above) the minimum permissible temper-ature of paragraph 2. of this section nor lower than (i) for pressurized water reactors, the minimum permissible temperature for the inservice system hydrostatic pressure test, or (ii) for boiling water reactors when water level is within the normal range 0

for power operation, 60 F above (33 C above) the reference temperature of the closure flange regions that are highly stressed by the bolt preload.

4 If there is no fuel in the reactor during system hydrostatic pressure tests or leak tests, the minimum permissible test temperature shall be RTNDT + 60 F (RTNOT ^

33 C).

1The latest Edition and Addenda cermitted by cara-graph 50.55a(b) at the time the analysis is made shall be used for the ourposes of caragrach IV.A.2.

-d-E: '.

X 5.

If there is fuel in the reactor during system pressure tests or leak tests, the require-ments of paragraphs 2 or 3 shall apply, depend-ing on whether the core is critical during the test.

B. 4 -

B.

Reactor vessels for which the pre-1,w. -- _ _ =

.g *Lw,lugaaja="A=0'**"." **;

dicted value of upper shelf energy at end of ca.rms -tne,sr-ve.nt sa -er =**-

life is below 50 ft lbs or the predicted no mue pro.rsi.. : r.rnias matario of the reester Teasel belt 3De.

9

- 7' -

6::....~

S.

,. n, effects of neutron radiation on the refererce t w_=_ _ru-= mnam, temoarature and upper-shelf energy.

=e=. mI. rim. &Uu"",'*','t "'gy=,

predicted from the results from pertinent radia-tion effects studies in addition to the results

.a a

_y fad of the surveillance program of Appendix H.

a

3. a.=

,==t.

ma

=t*4 onir for a.

r,o,.nunu. i. w e,.

r

c. p.nw

'ata tae requi,em.st. or.=oca rvu,,%.,.

~ 70$ U.'"Jae*. *.f"*=*i'**

"'"l, 7,,*;-

and the predicted valve of tha upper-4

" ha"=E" Yum uT! *d.".,

sheif energy Loh. h. [; -

ON These predictions shall be made for the radia-Q'.*... _MM tion conditions at the tip of the assumed flaw

c. tm a at its deepest part. The highest adjusted

"""** v2 annot w =u e.4. n= tor,=.

s reference temperature and the 1owest upper-Ew~iEOof-1Is'ng"tr.E"o7J.""dj f shelf energy level of all the bbitline mater-ials shall be used to verify that the frac-(ture toughness requirements are satisfied.

-f E" :: '. E

h*

( materials that do not satisfy the require-Voimme,.,,e e,.

[Lnents of Section '/.B.,

1._ _..,,.

, associated welds shall be made and any flaws a.unaian of th. d.iuin.,

__,acluding 100. percent..of anyA-i

, evaluated according to emi t

2. Addauonal evidence of the -

,and as speCifjed bY the Commission

= s-con M of the ASME Cod *&

fracture toughness of the be! Cine materigs expc.ur. to neutron tradae w

e3 after tan aus n. ontain.d from r uat. of sup.

piemental

---T--- _.. ONj fracture tou-}hness tests.

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3. A analysis shall be performed that conservatively demonstrates, making appropriate allowance. for all une.rtaintsee.

the estatence of adequate marginspot con.*

unued opermuon.

of 5afety D. If the prorwduree.,f section v.C. do not indteste the eusstence of an adequate safetr margin. the r. actor us.ei h.ittine regen aan d. eunsected i. a thermas ann. ann

. ~ S."dIgree il eus N the fracture toughness of the material.

r.conry anaa de meneur.d er tasting addt.

nonai ep.ctmens that han d.en withdrawn

' t ave been

,h from th..urvem a-program c.p.ute. and e ann.wed under tne.am. time.at-t.mpera.^

ture candluona na those gtna the tenuin'g -

- ( tor est331ishing

-materts. ne ruurt. shui proeide tne es..

ue adm.ted wrereoe.

temperatur.r ter ano.aan. n. reacier f and upper-shel f enercy c~

eenmet may conttnue to be operated only for 7

that service period within which the pre.

. dicted fracture toughness of the belutne region matertMs satiaSes the requirements of secuan ITA.& ustng the values of ad.

justed reference temperatur. that include th. scocts of annealtng and subsequerts irradlauen.

E. De proposed programa for satisfyteg the requiremente of sections V.C. and VD.

shall be reported to the Director of Nuclear Reactor Regulation. U.S. Nucint Reguistory '

Commasen. Washington, D.C. 205 5 5.t for re.

view and approema on an tndartdual cw besas at least 3 years prior to the date wnen the predteted fracture toughness tevels will no longer satisfy the requtrementa of nec.

tion YA g

e

/G Artsarsts 3-Raacna VaassL Marrazas Stravs21.3.aarca Ptossaad it.sectazusJrTo L saracevettom The purpose of the matarial survettlanca program requtrod by this appendiz is to soonttor changes Ls the fracture touganess propert.se et terr 1tta materts.a La the power reestors reeutting f um t.nett esposure-

/tS. Y W4Yfdr* /tVC fftVP*

re.

-toe e-t neimn. re. ion ore-to neutron term 41auen maa tae therme en.

,11 nment. ender tat. procr m. cr.eture touanne te. esta.ro omtat-a e rial SpetLmens m ~ *W

• • • r'"5a' 'k..sG!a wn pettor.14May from exposed in surveillance capsuIes, which are

" * * "S e _ _

p

..._. m,um-- -as.A, N(Accendix G.be used as described in Sections ho, -

(

.:ture II. scarlatAanca Paoomau carranza _

A. No matertaa survetitance pregram la re.

quired for reactor vessels for which tt can be conservatively demonstrated by analyt.

ical methods. appued to erperimental data and testa per*ormed on ecmparable vessais, snaming approprtste allowancas for all un.

certainttaa in the measurementa, t.h st the pean neutron Suenw(E>1MeV'as the end on the easign tire of the veneel wci nct esceed 1CP' n/cm*.

.. j.

f/*s ! ?~

M

m. a s.,

i. -

. hat

-- : oo nm -n t

. conome.

or,is ua..a.a a.

tn.tre.uan. -

N oaragrapn.

CirQ'Q*? * '""I'"** **""'

I '.do.4ht$ ror sur,.ia'anc. r, wtor _

materiais 1

. stb, e 135, exceot as nodified bv t$1s woe a ior v..ei.. AST.t CepgE ttana S

=

r.-i i

p. ss.

e.e.pt- = mooe

'r tat.

.p.

a tz.

appendix.

c eu: ne.

wr.=.nau m..t th.. n o. -

followtag r.a%

est.:

1. Surveu!.n etmen. an.Al B.

t.A.o from locattorpalocrs, t3e fr ture tougn.

n test c*. men. req. ed by meetton III of App als O. The.peem t types on.a corn with the req'atremen'

't seet ton i

. of Appendtz O f enes.>t that dic,.

Mt I

1.

That cart of the surveillance program 1 ' ^~ ~. ~

~ ~ ~ ~ ~ ~-

conducted prior to the first capsule with-drawal shall meet the requirements of the k edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased. For each capsule withdrawal, the test procedures and reporting requirements shall meet the requirements of the edition of E 185 in effect on the date of capsule withdrawal, to the extent practical for the configura-tion of the soecinens in the capsule. For any part of the surveillance program, later editions of E 185 including only those editions through 1979 may be used instead of the editions specified above.

2. surveinance specimen espsules shall be located near the inside vessel wall in the beltline restan. so that the specimen irradiation history dupil-cates to the extent practicable, within the physical constraints of the system, the neutron spectrum, temperature history, and maximum neutron fluence experienced by the reactor vessel inner surface. If the espsule holders are attached to the vessel wall or to the vessel cladding. construction and inservice inspection of the attach-ments and attachment welds shall be done according to the requirement.s for permanent structural attachments to reactor vessels gtven in the ASME Code 8 Sections III and XI. The design and location of the capsules shall NTNEPM.

~~

permit insertion of replacement cap-sules. Accelerated irradiation capsules may be used in addition to the re.

quired number of survetilance capsule

/ L.

specified infJ.

i 7 tandard Recommended Practice for j Surveillance Tests for Light Water j Cooled Nuclear Power Reactor 'lessels.

Copies may be obtained from the i

American Society for Testing and g

Materials, 1915 Race St., Phila-s.een rrom n. Am.ne.h. se....

o..

delphia, Pa.

~-

1o 103. Cool'es will

.n ro, rest.

taewa u.i.rw.. me a.c sc en.i.oei.

be available for inspection at pas r..

. eita.e upw.r

t. ~.., e-

• "$9 '*2 the Commission': Public Document "J'llEs" ".. *M'[:

EM3'" *'1'lM g

Room, 1717 H St., NW., Washington, D.C.

rat mr.wontoca.a c.

W-c.,.

j' P e

e e

F'rst capsu1%e.fou.%h service life Second capsule-Three. fourths nervice life T

capeute-6tandby

. the event that the surveillance t-me exhibit, at one-quarter of the to l's serv life, a shift of the reference tem;ers.

ture greater than or*.gtnally prod' for utn111 matertal as recorded in 134 appi to techn al spectScation the remaining tth=

arawn. schedule saan be modtSee a4 !. lows:

Ravtssa WrrManawaL Sensnots Secon capsule-Cne half servi life Third c$poule-Standby

b. Fcur hector veemets vulch a not eet the condiqons of section U.CJ A. but for which it be conservatively detsonstrsted by erperime tal data and testa performed on comparante esat steets that e adjusted reference to ersture wiu not scoed 200*F st the end of the service lifet e of the re.

actor teamel, a least four a sillance cap-eules than be rovided for e subseggest withdrawal na f ' lows:

WerM awaL SC VL3 Ftres capsule==At the its when the pro.

4 ted shf.*t of the adjusted to eran temperature is ap-pr -

ly WP or at one.

fo h rytc4 LLfe, whleh-eve is artier.

Second capsule-At ros!mately one. half of e time interval be=

t n l'.rst and third cap-e thdra wal.

Third capsule-Th f

ha servtce 1Lfe.

Fourth capsule-.4 db,

c. For reactor v is htch do not meet the conditions of tio U.C3.3 at Issas live surveClance $psules hall be provided for subsequent w thdraws as follows:

k Q

WrrH aawaL SC TLE V>

First capsule-t the time ben the pre.

gv dicted shttt the wijusted forence tem.

j s"'

' prettmately * *F or st one-i perature la s

fourth servi e info.Titlehe?'r eartter.

5econd and capsulee--At prettmately

,,\\

,i -

one. third d two-thirds of

'e time in.

,s terval be een first and to

.11 capsule Q

withdraw.

Pourth capfule-Three. fourths of rvice lif am

\\

Fifth cap 41e-Standby.

1. FroeAston shall also be med for ada dtttonal urvectance testa to morVter the reacts o annes1 Lng and subsequent' radia-tion.

e. W drawat schedules may be 12sd to cot ctde with those refueling oute es or plant / shutdowns most ciceely approethirt the withdrawal achedule.

f.

I acceterated trvadiation capsule are was ofed in addition to the minimum re-

%ul number of surveclar.e capsules. he

withdrawal schedule may be modised. tagg h/a account the test resulta obtained freys jtesting of the specimens in the accelerstedl yApsules. The proposed modtned vtthdrswa i

, schedule tn such cases anait be approved by

--_a

.s.

Proposed wtthdrawal schedules mase-ed-c

/ ' ana11 be submitted, witn a tech.

nical justtScation therefor, to the Cotsmts.

alon for approval. The proposed schedule shall not be implemented without prior Cocamlasma approval.

A m-J_

n

<mesulas and their wtthersemi achedules a f $tpws:

j n.

actor vennets for whics de scamervati demonstrated bye rtmental esta and tse formed our

.parshle ves-regr*

anowsaces foe 41 steel maalag a

43 uncertaintiaa in ensurements, that the adjusted refer %

• perature estabW Mabed in acccarnate with se%n !!!.B. wtB mot ne%:wr at the.nd e meret

'.Ifetime t#Nte reactor vesset, at le survessnee capsulee shan be provided%

rf -

• n ano o a*

/!

C.

An integrated surveillance program may be considered for a set of reactors that have similar design and operating features. The materials chosen for surveillance from each reactor in the set may be irradiated in one or more of the reactors. No reduction in the requirements for number of materials to be irradiated, specimen types, or number of specimens is permitted, but the amount of testing may be reduced if the initial results agree with predictions.

Inte-grated surveillance programs must be approved by the Commission on a case-by-case basis. Cri-teria for approval include the following con-siderations:

(1) The design and operating fea-tures of the reactors in the set must be suf-ficiently similar to permit accurate comparisons of the predicted amount of radiation damage as a function of total power autout, (2) There must be adeouate arrangements for data sharing between plants, and, (3) there must be substantial advan-tages to be gained in the credicted reductions in power outages and personnel exposure to radiation.

n Integrated survebance p~rhamM.

su rtzed by the CommLaaton osvan in tvidual basm. dependtag the de=

p' no of com eJ1ty an4 th redacted me.

verity of irrtelat.

tzr. rnac tras o

nsrs A. Practure

• ganess testin f the spect.

nana w1'

. awn from the capsui alJ be en

g__ 4 in scoordance with the tre.

:,.,, _: : x k e naha re[eren tamperatunsM

.3. % m.w. hesi.ne.c.,4.one. an esa n.iu mi de tats.4 from tae

.t re-mits by ache to tne reference - peratun b amount oK4ne temperat satti in the

harpy sees cur,bdeswee e unternatst,a nat,w an4 sae mat.rm mea..

ne tn. no foo i.eet or t3at

- at th.

m%t,ee expan, ton Levet, whchever perature4ntft la greater <

The highees ustaa referencQmperature pW the loeresa upper.ahelf one level of h.L1 the tutne materials aball b.

to t the fracture toughnese tre-

, of section 7.3. of Append 12 0 e

wpe p

en gg,

,y g

/ M

~

f HI.

w or rast amatrLys U

A m empends withcLrswal and th resulta or in. tricsun tousana. sau anan * **

Tw1 thin 90 days after comoletion ein3.co at a r==.

y tacanical nport to De

,,,,,4 4 to in. cir.cror or Nuclear Reactor of testing. Tn,e Commission shall

%au. v.s. Nwlear Regulatory Commta-

.on. w stoa. o.c. 20s55

-N j De notified at least 30 days in 3

advance of the capsule with-QQ -'" * * "*1,t:o drawal, giving the expected date u m -te t s n=u2ta and Fa*

E,4 ror ih. "n.ctor e 17 of completion of testing and sub-mittal of report for approval by o

wthe Cormission.

-._ _ne r. cort anan===* L=en>d* **r; -

s.

-= -

c=2-

% y,,1*""**b%.C #

anu.O

' data required by ASTM E 185, and the results of all fracture toughness yy,r'***or,,,g'""*.g",g

)testsconductedonthebeltiinemater-e

e. ortats.uy pnat ua p _

N~,, _. ~.. 4__ _ -.-

f als in the irradiated and unirrated conditions.

.m..

._.:. # ~ ar s~

3 -- =g g:-

- -.. -- -_ y y.

y__

C.

If a change in the ope ating pres-sure-temperature limits given in the Technical Specif' cations is required, the revised limits shall be submitted with the report, including any changes made in operating procedures required to meet such limits.

J y

w/

m.'