ML20049A340
| ML20049A340 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/03/1980 |
| From: | Ahearne J NRC COMMISSION (OCM) |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20049A339 | List: |
| References | |
| NUDOCS 8012020443 | |
| Download: ML20049A340 (85) | |
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UNITED STATES m
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WASHINGTON, D. C. 20566
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s September 3, 1980 7
CH A RMAN m
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Director, NRR.
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John Ahearne SUBJECTi F'ARLEYPOWERLOPERATINGLICENSE T
' As 'I mentianed at the conclusion of today's meeting on the low power proposal for Farley, I have a number of questions I would like-s s
answered prior to approving full power license for Farley. These are indicated on the attached pages from the August 25, 1980 compliance l'etter-of the Alabama Power Company and from Supplement 4 to the N
Farley 2 SER.
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'Any questions of interpretation should be referred to E
Dr. George Sauter of my office.
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Att'achmen't
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cc: Commis'sioner Gilinsky Commissioner Hendrie l
' Commissioner Bradford l
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Se;tember 7, 1920, c d.AlR M AN
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9 MEM3RANDUM FOR:
Director, NRR FROM:
John Ahearne
[f ji{ wjK hPERATINGLICENSE S'JEJ ECT:
FARLEY POWER /
As I mentioned at the conclusion of today's meeting on the low power proposal for Farley, I have a number of questions I would like ar.swered prior to approving fu'.1 power license for Farley. These are indicated on the attached pages from the August 25, 1980 ccmpliance letter of the ' Alabama Power Company and from Supplemen: 4 to the Farley 2 SER.
Any questions of interpretation should be referred to Dr. George Sauter of my office.
A::achment
- i C:mmissi:ner Gilinsky C:m-issi:r.er Hendrie C:--issi:ner Bradferd Secretary Acting E30 4
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s In 5.::lement N:. 3 :: tre SER,.e als: i:er.tified t.: ;eneri: issues. Our status f review
- f inese iss.es is i- ;Et se:-i:ns in:i: ate: beicw.
( *. ) Envir:- en ai :calificati:n cf safety-relate e ui; tent (ie:ti:n 7..2)
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3 r
s 1,. a. e... a. s,.,.. s. 4............ s..a.
e....
s.
':1 ne e areas a; na.e ar# san si :e issua :e :f Su:;1ere.
N:. 3 :: tne !!:., inclucir.;
TM* iss.es, a.e :een sa-is'a: :-il;.
es:'ve: f:r f.el ica:ir; an: 1: ::.er testin; exce:
...... s.... i.... e. s,.. s.,
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ts M, #.,..
r
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e
- 4.... gx.g..e.
............ s s.
ts'-
su:;ites wili te ins alle: :n auxiiiary fee: a er flew ::ntr:1. valves (II.E.1.2, I
Se:tien 22.2).
1(dS
- i. (2) Prie- : ::n:c: ting the ac;rentet 1:< :: er tests, a::licant ::st re:eive NR a::r:vai
\\sb21s/
Of its safety ar.aiysis re::rt :n :ne :en::: :f tr.e tests (I.G.1 Secti:n 22.2) 1.9 Un ese!vec Sa'etv !ssues Or. N:ve::er 23, 1977, the Ateci: Safety an: Licensin; A;;eal 5:ard issued a :ecisien (ALAI-aat) in c:nnection with its c:nsiceratien of the a:;1icatien for the River Eend Statien, Unit Nos. I and 2 (Oceket Nos. 50-a53 and 5C-459) wnica established specific recuirements for accressing unrescived safety generi issues in ::nne:tien with cur li:ensin; pr:ceed-ings. These recuirements are a;;11catie 1: tse J:seph M. Farley Nuclear ?lant, Unit 2 a;;11 cation.
A:;endix C := this su;:lement ; resents inf:r:t-i n for the J:se:h M. Farley Nucitar Piant, U..it I a;;1i:ati:n in ::nferman:e witn ne A:: tai I:ar ca:ision anuncia e: in ALAE-444 9
i i
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y
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a.='...s. <...c -.'~......... ce..=. i..a e.
. at er :.e :rigi.ai tes s a-c a.alysts.ere a:e: ste.
- r
- revi:us evie. t' estin;.:use g, ;,..y :. a...- -. s '. e - *. ".... *. *. ". a. -. s. ' -. '.. *..- -
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w.
aly 7-;0, 1550 ce:er.ine wnetner :ne cualifica-i:n :f :..e ecui::e.t. as installed in Farley 2, ;erfermed in a:::r:ance with ne re:ecures c' IEEE Stancard 3:4-15 1 c:ald c.eet i
cu rent li:ensing criteria as cescri:ec in 57.7 Se::ien 2.10.
During this reviev *e e<aivate: a re:resentative sa: pie of.hirty-f gr :ie:es :f Seismi ateg:ry me:na..ical, i..stru:entatien, an: ele: rical e:uipment. Our review u.::vered relatively few pieces of equi; ent for whi:n it was not clear that the seisti: cuali'itati:n was a::e: table in the lign of curre.: licensing criteria. F r ext::le, :ne :a.:ery charpr in :na service water
- ui'.cing was :untet flat :n the test table,.nila it is :antilevered cff the.all in the
'ield.
Alsc, the sciencid valve in *he river water builei.; is field ::unte: in su-h a way that it may be susce:tible to low fre:veney (telew 20 nert:) input, yet the test was a::arenti.v c:ncu:ted eni.v for a fre:uenev ran e te.venc 20 hert:. The :etails cf these s
snorte:.ings anc ethers in the ecui: men cualifica.icn are cescribed in the report of cur July 7-10, 1550 tri; :: the plant.
Fe-nese fe= items, ne a:plicant has ::.i".ec ::
sa::i aceitional inferma-icn, clarification, a3: resciutien for :ur review :rier := a:;reval c,. e..11..wer
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We ::n:iude that the a;;licant nas ne; ;tevice: ne: essa y and suffi:ier.: informati n ::
ce.9:nstrate 'ull ::;11ance with Paragra;ns !. A and IV. A.2.a.
The a:;iicant.as stated 1..at :ne infermatien necessary :: fully satisfy this re uirement wii; te previced t: us by Se:terter 30, 1980. The a::licar.: has previded sufficient infer:ati n to a11:w us t: cete ine that prict te ner:al full p:wer Opera.icn, the safety cargins re uired 'er I:w ;:wer ;eration will be achieved and za'ntair.ed. On this basis, we enneluce that n
.,:. O
!i:w ;;.ar 0;eratien is a :e::aele. We will ::::lete our review prier to full p:wer
~.'
l
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h::eratien t: ::nfi that ace vate saf ety car; ins will aise be maintained curing n - al p ;eration, in:1uding :;erati:nal transie.ts, in :: ;1iance with Paragra:hs I. A and ic p V.A..a e' A;;endix G t: 10 0.:Pt Part ~0.
#r gI J
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s.
se:ervis: y persennel. Ee:ause these tes.s are reiatively reu*ine in nature a.c are
.... e.
.. a '. l >. e i.n '-. e. '. a..c *. *. i r. *.9 e l a.- a. a "... ;", w e.e r.-1 "- a. "... -. i. I s ".n11"Ke i.y "...a*. *...e tests were cencu:.at impr::erly. 0:nsecuer..ly, we ::n iuce that an exe :tien 'er ne:
. e. e.... <.,..a *.es.s
'.a. a..--.. a a.c e v '. *....'. '. *.*. =..,... e.".. =. s i s.i u s '.'. '. i e d.
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f
(a). tic seam P. y is re;reser.te: :y t.e a:: n1:nal cata; (t)
.re a:citiona'. :ata can :e use: t: extra;:iate to the u;;er shelf 1:r weld seas 2e..:n.. an.a i.)
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i
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a.. a. s-. '. a '. s s ".. v.. l '1 a.. - *.... -..=.. *..*. a.. u s.. e e.
a
- a.
v, the re:uire en s of ASTF. Standar: E ~13-73, "Stan:ar Re::: enced Fr'actice 'er Surveillan:e Tests 'er Nu: lear Rea:::t Vessels." a.: A;;en:ix M, 10 FF. Far. 50. We have evaluate: the 4 a::iicant's inf:rmatica for cegree :' :::;11ance t these recuirecer.ts and have cen:1uded ij nat tne a::licant has met all re : ire en s of A:;en:ix H,10 0.:R Fart 50, ex:ept for
'. Paragra:n C.!, for which su'ficient i..f:rmati:n nas been su;;1ied t: justi'y an ext:: tion.
pa.g..a.g.t-
- ...,uj.es.6.e.e1 14..,....... e. c, u.a e.e a.... y e s s e ).. *. e. n j.. e d. by a w
surveillance program cc:;1ying witn A5 l' Stancar: E-185-73.
Ac:Ording :: this standard the base metal and weld cetal te be inciu:ed in the program sneuld represent the caterial that may limit t..e coeratiens of -he rea:::r curing its. lif etime. This selecticn is based on ini-ial transiti:n tec:erature, u:per shelf energy level, and estimated increase in tran-sition tem:erature ::nsidering enerical ce=; siti:n (c:;;u an: ;ncs;herus) and neuti:n fluence.
A:::r:ing :: cur evaluatien, plate 1212-; and weit sea: 11-923 are the ::st li iting base a....
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ii.i.s :ase: en :ne greater of ne 1:ll:.ing: (a) ine actual snift in reference tem:erature f.,.....e.:..-2.4.,
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2,
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cu 3ne ali n:enal :lan: e erati:ns, and thus, the exemptien t: caragra;n *.5, A;;endix H,
- r. :a.. ca j 5 j..s. j e. s.,..
e' I
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6
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5-
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4 i =,.
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e
. a.1..... s..u.. =. a an... "... C.
e.... e,..... n x s: and :a-=,.a o-L' t..:. '. A- *.-'.x ;..
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...s.s.n..ese.a.a.....s...:.s
.egut..
- 4. 7..*. a.. s. :. s.... o. s..,.
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... g ;...,. i.'.-. e a s e #. *. *.*. I e v e l '.. '. a '. #..v * *. s a '. e.Y. ~ ' s.,,.a...'..,a'.
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a, "..- l i. '..".. a e < *..
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, ".. s "..i.n..,,,
i
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... e -=. c c i - a.ca..r.s. ' "... ".. e a.=,. = y o's a r =. 3. t.e.d.
o m..e..:x., n
...e.
- 4....,. a i n s '..N... ^. ". -. '. l e r" a '.1 ". *., "
.( e.. i.a.. ". *. '. "... e n*.t.u..: : ile.
a...'
e r....
? essure Vessel C ce, will be usec.ith fra:ture ::v;nness test results re u.irec by e
'.C u*.:R Da...",
- .a
. a ' -". l a. a. '. 9 e r e s -..- - -.. l a ".... e ". ".. =. '... "...d a. 2" a.....
.,s. ao-' W
... g 3...... e. g.g. o..e s :.. i. g <.. : a.1 e.v U..#. '. N a.. 2.
T.5e fra::cre :ughness tests receired by the ASME C: e and A::endix G,10 C R Fart 50, vili
- r
- vice retscna:ie assuran:e that ate:vate sa'ety =argins against ini ;cssi:ility :f non-ductile benavier er ra: idly prepa;ating 'racture :an te established for all pressure retair.ing ::::enents of the rea:ter ::clant ressure Scundary. The use of A::endix G, Se:.icn III cf the ASME Cede, as a guide in establishin; safe ::eratin; ; ece:ures, and use f the results of the fracture tougr. ness tests perfer:ed in a::ercance with tne ASME Cece a.. i:..e.,.1a.i.os, vil'i *,.*.* vide a.e ".a.e sa'.e v
=a., ins.-"..4.*,
-.a. a.4.9, **s*.ir.
ai..-
d a
3
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- e..,, 4. a.. e.... s.e s e
- 4.. e,...v i s t o s a. a.
W rege:ati:ns ::nstitutes an a::e;;a:ie : asis f:r satis'. vin ine rervire ents :f General
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(4) In':rmati:n regarcing the c: era:iiity c'f the ; urge valves.
The resciuti:n :" these issues wili ;tevice in:rease: assurance that the valves will c;erate un:er a::i:en ::ncitiens anc radicactive releases will be minimi:ec. However, c;tratien f
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'7[1 systems shall te maintained as 1:* as pra::ical an: leak tests shai.1 te run perieti: ally.
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'.'e wiii re;;r cur evaluation Of leak testin; c'.hese systers in a future su::ie:en :: Our y m'-
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a :esign eview :: cete :-i*.e.hether or 3: u;cn rese: c' an engineered sa'ety feature a:::.atien signa.i, al.i asse:ia.e:
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Unit 2.
The res;:ense inci:ated that ne ceficiencies wa e identi'iec as a result of a::licant's review of Farley Uni 2 in a:::rda.:e wi.h tne a::icn items in the *uiletin.
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Uni t *.. :n :3 :. A en: ents 70 an: 72, the a::licant ena..;et ne design bases for the Unit 2 scent fuei ;;ci : a s acin; of
- .3 in:hes.
The :esign an: cesign :ases fer *he mcdificatices :: tne Uni 2 spent 'uel ;;ci are :ne same as these 'er tr.e Unit 1 spent fuei ; col. Ve enclu:e :na tne Unit 2 esign modifietti:ns are tne same as tnose a;:reved fer Unit 1, est tr.e re: irerents of General Design Criterien 52, " Prevention cf Criticality in.: vel St: rage an: Handling," and are therefere a.e..a.wie.
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"* n summa ry, the pare Fr: tecti:n Pr:;ra fer the Earley Nu: lear Siar. witn'the ie:reverents alrea:y ra:e, is ace:cate f:r :ne ; esent ti e and, -ith :ne scheggiec
- ifications, -i'.i e:. the ;;iceiines ::ntainte 'n A;;endix A : it? A51 9.5-1.
Tne Ti e Fr:te: icn Fr:;rt: as curren-ly :esigne: anc installed eetts Ga tral Oesign Criterien 2 and is a::e;;a:it."
Ey letter cated a ;ust
'.3,
- .530, the a:;iicant nas stated :nst all :::fficati ns will te u
c:::1ete: ;rier te fue'. *.:a:ir.; ex:e:t f:- tne f:11: win;-
1.
5 :se :ste:ti:n systers f-ne auxi'ia y :ui',:ir;.
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n:se s atices in - e Lt.it 2 :a:1e tunnels :e:-een :ne :iesei. ;ene at:- :uiicir.; an:
e 5 3 '# )
iI auxiliarv builcin;.
4is:s'tlp, 50th :f inese re:if t:attens will :e :::Oleted my 0:te:er 30,1950, :r ;rt:r :: initial
."g:riticality. -nienever is earlier.
'*e ::n:iuce :nat fuel 1: acing is a::e: acle :e:ause all fire ;r:te::icn e:vi::ent recuired by :ne a;; roved plan inside centaineen will te installed before fue.l lettin;. We will
- ncition -he licensee te re;uire all fire pre e:-icn e:vi; rent :: te int:alie: and Opera:1e prier :: Neve :er 1,1950 in a:::rd with the C:mmissien's May 23, 1950 Me :rancum and Orcer.
Ve will in: luce Fire Frcte:tien Syste: Techni:ai I:e:ificati ns in :.,e ::erating license fer Unit 2 that are the same as inese revie-e: and a;;r:ve: for Uni: 1.
Tne Office of -
v;).34L Ins;ectica and Enferee:ent will verify ::::letion Of all :::ificati:ns at the times stated ga::ve.
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Ire te:r...ical sce:if t:ati:es fer tne ::erati:n f Farley su: lear cia *.t.
U.it ',:. 2, rtcuire a :r:: ass ::ntr:1 ;re;rar fer the sci':'f t:a f:r. an :acca;ir." :f -ast'es Oy t e Psef =7:tivC'-
s
.s:!!: -aste systet. 7?e r::ess ::n;-:'. :r:; t. :urre.ly f r are 'er arley -it 1 is '
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. aste fs sni;:e:. A::*.i:an: nas agree: :: ;r:vi:e f:
-ev'a-s: at:-d a' :j ::::er *.,
a,, *..**p sQ.!!:, 1 e :ases and justift:stien f:r the pr::ess ::nte:1 ;r:grac ::. assure tr.at shi;:e:
t.: ~,
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sH,s :11: wastes will ::nf:r: : a::li:a:1e :urial gr:un re:uire entJ. We =ill ::ncitien :ne 4
lit:erse 1: re:uire the su::ittal :f :nis inf:--ation. 34 sed :n ins ct: rage :::4:ity 'er d(
self: ra:-aste at Farley, tne staff :elieves tnat a:;reval f the finai pr::ess ::ntrol i
s
~
involve a~ safe:'f
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.Su:.f e: :: 4;;r: val cf the ;r::ess ::ntroi pr:grr:, e ::n:1u:e inat, tN ;re:essing an:
st: rage facilities for solic ra:icactive waste cateriais are acecuate for Operatien of both units, in:1ucing anti:i; ate: ::erati:.ai :::;rrences, meet the a;;11:ible re:virements of General Design Criterien 60, "0:n:rci of Releases f Ra:ica:tive' Materials t: thdEnvirn-ment," and therefore are a::estat.le.
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ra:i:1:;ical sa::tage as re:wi e: :y ;0 #F:. Fart 72.55.
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As a result of su:se:uent revisiens, :ns a::r:.e: :ian ::nsists f a :::u ent entitle:
"::se;n.. Fariey Nu: lear Flant M: ifie: Aten:e: Se:urity ::an" cate: August 30, 1575.
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..e s.a.. has again reviewe the :nysical security :la.. agt'ns the re:vice ents 4 10 lFF Fart 72.55 and nas : ster.med that the plan is a::e;ta:ie ex:a;; as n:tec teie.
a Sy ieher : ate: August 13, '.950, tne a;:li:ar.: ::..itte: t: i::leme.ing :er ain :nanges in
}Q q ;nis ;hysi:a1 se:urity pregram.
Satisf act: y i :le:e.tati:n :f these ::mit. ents is re:uired
,. TO rier te fuel leading of Unit 2.
The Office :f Ins;e:tien an: En" r:erent will verify
'# ~ f.
d'i.-:letentaticaprior:=fuellee:ing.
L In a::itien, we re: wire tnat tee a::;i:a". '.tiy ::::1y -itn tae e:virement f 1 CFP. Part
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T.e i:erti:sti:n c' vital areas 4.:,*asurts use: :: ::ntr:i a::ess 1: tnese areas, as ces: rite: in ne ian e.ay :e su:'e:-.: t e.:rer.s in t$e future tase: en a ::n'i-at: y evai a.i:n Of ' Unit 2 t: :etermine in:se areas.nare a: s Of sa:: a;e signt cause a reitase of ra:i:nu 'iens in suffi:1e. :.a-.ities : rescit in ::se rates e:ual t: er exces:ing 10 CF?.
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U :e *..e issuar:e f Su::le ert.N:. 3 :: tre Safety i.a'.ati:n :t::-t, t*.e a::14:1-. has s.:-t te: a en: ents t: its cuaHty ass.: ar:e : :;ra. :ss:-i:ti:n f:- tre ::erati: s :r.ase cf t e J:se:n M. Farley Nu: lear Fiar..-
Our review cf t.e er.anges t: tne t.a',ity assuran:e pr:gra nas verified that the criterta cf A::encia i t:
- 0. R Part 50 have :een a:acuately '
j a::resset in Section 17.2 of tr.e 13 R as a ta.ded tar v;r. A. en::ent 72.
The staff ?.as recently devel:;ed a reviseg ;r::ecure f:r ::n:u: ting tr.e revie. of tne list of safety-relate: stru:tures, sy e:1, tr.: :: ::nents (C-list) 1: -ni:n tat ::ali:y assura.:e i
- tegram a
- ;11es. This review inv:1ves all :ran:nes that save res;:nsittlity f:r reviewin;
.the FSAR an: significantly ent.a :es the staff's ::nfi:er.:e in the a::e: 4:ility :f tre l
p Q-list. Staf f re-review of the Q-list using the revise: Or :edure is 1-::rtant f:r ;r:;er l
1 maintanar:e f til. safety-related e:vi::er.: Over the ;1 ant lifeti e (40 years); h:.ever, its l
3)
- 1etion is n t deemed to be necessary pri t t
- granting auth:-ity : 1:ad fusi tne
]perfer: 1:w ;:wer tests, be:ause the ne e:vi;:ent is n:t likely 1: re:uire :aintenan:e in
, g the sher; tire internal of ;erati:n at 1:w : wer. This re-review is presently un:er way P
l e.
%g'. Ianc :ne results will be re;:rted prf:r to full ::-er ::erati:n.
i I
I 4
1 1
i h
e
~
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t 4
l f
4 I
t
+
f I
l' I
1 i
e.
- ~ - ~,.. -,,.- -
T
- e 'i. 'f'ict f I*.s;e.tti:n an: fr.f:r:tre.t will review the :r::stures after they are a.::'f't, a..:.ill -assure tr.at tne a::r::riate :::i" Cati:ns as stated at:ve are ea:e :ri:-
a..e...a.
- . :.r evit*, we as:erta' red tra: ine 571.ill he in':rre: :' t e resu!:s =f evaluatier. :f i.s.
g.g4.., r,. e.. : g.... s 4..- -. e.-.'. e e a.*. i.- r.. t a s.. a s..'. s *. '..*. ' e s c '.. *6 *'a.'..-
'. n
.a...e...
..j.,..
,s e e..,e../. 4...,s
.....e.
,e.s...
- a. e)s.t.: :e.e.
a. e u..........r.e:
a y........,.
4
...,.,.4...
g g 3, g g.. e. g. 4., a g. a. a, i...... < 4.. e. g s.., :ve..
- e....g.
. e....e.~..es
..a.
l tre 3* P : : state ina: "Tr.e SF ir:.: sr.a'.; :r:s u e ;e.e at e ;ireerin; t.:::r: f:e One 5*A
....e. c:
o...
.a.s..e
..e.<.....,..e.a.4..a.
assess.e......s....e
..,a. 4......a.
i
... ex.s.s
.a. s..a.y.e.e3a.<s.e....e
. _....s.
a.
e
- ....n, s..
a..a. _ a.s.. w.n..
e a
issue :: the STAS."
s
.('N i
i f,a.:
4.........r.
- 4...a *.'. :.n 'v *.9 e * *. "... e '. P.. s e C *.'... i a...:.. '. -.. e. e r.*.
'..'.a *. *...e
- r..- F... e s.'. a v e -,
1
.,, )p c9
'l :een ::t'fiec :: s:ecify Shif t Su:ervis:r -Ins:e:-in; res::.si:ilities and assig. en.s as e
. s...s s e. a :v e a... a..,.... e,... e s. as. e.e e... s. a.. e.. s. e. s. a. y.......a l.d g <. e S i a u
N t: :ne ::ntr:1 r::: up n learnin; :" a significar.t abn:rmality er in:ipient emer;en:y. we 4
l
- n:1cca that the re:viremer. s ::n=ernin; STAS f:r fuel lea:in; and,iew p:wer testin; have f
been met. In a:::rdan:e with our Cate: Re uire:en (I. A.1.1 in Se:tien 22.5 ef this l. su::la ent) APC: is re:uired to have STAS wh: have.*:::lete: all training re:uire:ents :n
\\,
- uty by "a..ua y 1, 1521.
l I
1.A.1.2 Shift tu:erviser A: inist stive Dut'es Recu' se.er.
i.
I s e s :. t.
.e a.;.4.s..a.
ve
...<es
- o...e s.<s.. s...e../ss. a...e.
.a.e.,...i.s...a.
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g., 3.. ~.. 4.
g.,
.6.,...g,....., 3... s #..- ' ' * *.. '.-- a s s.. #..,- s a'. s. -.e' a*.'....'..*.
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.e.. 4... e..e.. a,. e _ c.,.
1 a.....e.s
, ;e..e..e. o a........e. :.: :
e..
t
- ., s.. a.
t T.e ni; es: level f ::rperate cana;e ent f ea:. iicansea shall issue an: ;tri::i:aily reiss e a :anage ent cire: ive
.t. e::nashes ne :ricacy canageme.-. res:enshilhy of
. e s.,,. s...e...,s
.,.. sa.e....a.4.
.,.. e.. a..... a. a.1.. n a.
- 4.. s. i 4. s s
- 4.,..
a :
a :learly es.2:lishes.is ::: an: :uties.
- . g...... e... e g 3.g..
7.<..
... a s s. - i....=.. - =.... ' e s.. a. s.... s '. - '. i '... a. s. a..-
.,..e s '.'. s.e.'s.-
a.....-.--.' -.- -. =..'..- - s a. a. -. -.=.. ',.-=. " e...-
3.......
t e....e el.a..a..e.
.a.
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e a. a.....<..,
so
...g
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s..,
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3.g.4..
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- g........
g...
ge, g.,
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24.......
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u
.ase.......
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-, - +
+, + +
r-.+.c+y-t-.
gg..,
.y-s w---c
-+
.p 9
- -:1usier.s ke r. ave revie.ec the infernati:r. n sat'; ra. nim; an :ve-t' e :r:,1:e: ty A::: in 1..e SAE as aren:ec, a..: tr.e suunitials cate: Aw;ust 7, * *,. an: 22, !!:, ar.: c:n;are: the infernati:n
........s:r 3(e,s,...::...
3..: r,),.(.<), (.a,. <3),(.),
......e a.. 3,a e...
- 4..s c,...,::
s,
s.
6 s
.a.. ::....,..., 4.e.....$.e.'a a.t..'.1, '.,..:.:..
r
..... e........,..,
- 4. e..s e e.-. -. r e s.-. e. a. e i. a.-.- -. a. e. '..'....' e. e,. l a. i..: a.-
... 4.. t. 4. a..... e v a. a... s.. s.. a., 4. s..... a.....
- 4. e s. a....... e.. s.. e. e.. s
..,....4..
a
.u....
a: is a::e:ta:1e.-
.:r the text fe-3:..t s,'n:.ever, tre :ersee nas ::
't. SEOs li:ense:
a...s.....e
.5...a..,.a.g, a..gg.....se. a.
.e..a.se..a.
....e..
....t.. g.. g. 3,y g....
. i.
.i..........
., sate :y usin; *hree SFOs en es:n snif t i sten: Of t3e t.: re:uire: :y the Interim riteria.
.F.
l* ief:re tSe license : ::erate Farley Und 2 is issue:,.e e:cire:
fi t? ( 1 !RO F:r the 1er; term, t c ;er sa'f are re: wire:, li:ense: :n ::th us.its.
In the i
D interir. ;cri:d, wnen n: en: ugh SEos license: en Unit 2 are availatie, three :er shift are 7 require:, at least ene licensed as SEO :n Ur.it 2, ;1 s a snift fere an - :;erating, license:
1 en init - an: traine: in the cifferer.:es :et.een tne units.
L i
(2?. 40 - Three per shift, :ne licea.sec en ee:n unit anc :ne license: en both units.
.l '
~
! The nue:er of candicates wn: t::k the E0 and SEO exa..icati:ns, and the nu=ter f trainees, jl):aren:
i ih sufficient :: su:::rt a finding at inis time that tne a::licant can in fact fulfill gr re:wirements. K: wever, ine a::lica.. is :evel::ing a::iti:nai centingen:y ;1ans whien,
- 'n ::rjun:tien with examinati:n results s::n :: :e avaiia:ie, seem te us likely :: :r::u:e
.i".1
- rt tra -esu'ts :' :gr evaluati:r. :f t,nis a::iti:nai tre : asis fer sa:n a fin:ing, kt
- -f:- ati:n in a su:;iesent t: tr.is infety I.alua.'.an.
e.....,
- .a.,,
g...
4.,.'a '.
.'.t s'-. :.=1 '
t.'-
s
- :.' t e-;
A'l ea:::r :: erat:r license a:;ii:ar.ts snail taxe a.r' ta. ext:inati:n witt, a ne. Cate;:ry
- ealing.itn :ne :rinci:les Of r.ea; transfer an: flui re:na..i:s, a ti e lici; :f nine -
n:u s, an: : :assin; gra:e f 50 :er:en :verali anc 70 :er:er. in each cate;: y.
.. se <...........a..-. t.i.e.se a. ..a..s s.t'.1.ake
..-a.
. t a.- *..-. c a, e =..- a x._ '....=.'. a. ~.,
an ::e stin; test, and a senter reat :- :: erat:r.rittan ext 'nati:n with a
- e. : ate;;ry
...:..,.....e.......,,<.4.,
.g.4 3.
g.4,.
<.<.,' sa.va.- -.-. s, t. -.a
.... ~...
., s,...,... s. t., :
.t..e..
. g.g.-
g..
4.
,1..
.............s
..,.t.
6.
..ese....,.....,.,,..e.,..,,.., e g 4
.,...s.
..s...
..t 4.....
m-
- 'estin;n:cse su::ittec '. analyses f:r s:all-:reak a::i:er. s in ic;ical P.e;:rt W AP-95:0, "En::-; cr. 5 all Eresi A::1:ents ':r Westir;n:cse N555 Systa:"; June 1570. E er;er:y ;r:-
i
- e:gre ;ci:e*iras were then : eve *::e: fr:: t'est aralyses :y :ne Vest n;r.:cse :lar.: 0.ners d
Or:.:. ~ntst guidelines were re.de-e: an 4::r:ve: :y tr.e staff in 4:ver:er 1975. it.e
[
stafd evie. :f these analyses a : guitelices as :erf:-re: :y Be Eulletin ar.: C :sts Tast
- r:e as is :::grente: in their t::r :r. '.estir.;n:use it:::rs, "*eneri: Ivaivat':r. cf et:.1:n-T-ansients an: 5:411 ! et< L:ss-:f ::: ar.: ::':en;s in kestiegn: st *esi; ed
- t t:' ; lar s," ?C:.!;-Cil".,.'t.cary 153: (s::en:ix IX, se:-':n 2.2).
ke have revie.e:
- . t :est;n features :f the Far'.e,. L-!: 2 ;'t : ar:.e ::r.:1.:t tra: : e revit. at: a::r:vai
- f
?.t stail-:reak '.0:A ar.a'ysts an: ;u':al'res a::1y
- a*. :: ne Earley, ai: ; ;1 a..
3-
).
- e.. e. g g. e..i. n e a* 3, 9..e
- C. *..*.e l i.* e.'.s e t s " ~* i '.*. e ". * * *
- e -".* e s
- . 0 r l e s s * * #. * *..-. I a..*. a.* * '..* e.*.*.
.s o
(in:iu:f r.; small treaks), ina: : att ::re ::: ling, anti:i;t;e: transients witn:ut trip, s et: generat:r tute ru;;ure, an: icss f main fee:.ater. it.ase prece gres are require: ::
- e tvie e: ty ine staff and ::rre::e: :y ne a;:licant :ri:r to full ;;.ar c:erati:n. (See re:wi trer.t I.C.! cf Par-2 of YJE!; Cis:.)
j East: c::n our review to cate cf the pre:etures su::itte: ty the lic2nsee, we fin: they are
- tnerally c
- nsistent with ne gui: alines f:r Vestingneuse :lants. There are a nu::er c' ein:r in::nsistancies with specif': catails of the guf:tlines and s:ce instru:stens :: :ne c: erat:r are vague. These matters are tein; dis:vsses with the li:ensee. Our detailed i
c:: er.ts en the ;rocetures were trans:itte: :: the licensee, and we met with the licenses t
- is: css ;rece:ure revisions require: for te:nnical and sequer.tial acecuacy. Sele:ted revised e erger.:y prececures will be walked througn a simula :r an: ne : tant and further :nanges I
ma:e, i' ne:essary. The resulting eevise: t ergen:y ;rt:::gres iil :e in::r;; rated it.::
..e..a....a. :n,,.r...t.
a a..
.a.a.t...
a..
....e...es.
f.
...e s... t. a...e
.. e s e e e.. t..... e s.e s <.... g....... e.. s. a.e *. t
,=..'e..='i.-' f..- - < '. s. t...
s
.4
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'* **1.1 a*.*..**.*.t*.*.
8.-
..t
,. 4. g j e.
a s
- ..~ *. t s *, ' *,* * *.. a* 4 *. ' a - *. s*.
t!8 ****t.*ti
.~.e.E**'.~.
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- e *..* t * * *. e.. - e. s '. s. n ' *. * ' a * *. %
It
- 1. s ' '.s '. e '. - - a.v t.. e., e.. *g-l
...ss
. t.
.g...
..... :g a...g',
s
..t
...... aa s.eg. y.e a. !:..g
..g g. s.:
.gs... '..
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'- /
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s.t.t...a 1
.t.
1
.t s......
..t.a.z.......... : v, a.
c
.a
<e..
- -t ning, st will rt
- r :u evaluat': Of the ::::lete:
i i
.::.tr f:r 1:. ;:.er testing ar.:
p
$.'d:.,.[~
d.:r::::u as in Su:;1ement $ :: :gr Safe y Ivai' a*ien ES:i*, :ri:- : full ;;wer ::erati:n.
~
. 0. *
- -ift Relief a-: Turn:ver : ::ecures
- e:.'-t e ;
g..............g..
4..,, 4.. g... g. i. 4..,......,..
- t.
3,.
3.......g..,3
<. g....g :ge.
.4 g. g. t...,
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4..-
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....,..t
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4
......g..x.e...........,...g
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,,4..,
I
. 2 3 t :..
- 1... a. t s.,..e.,.e
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. a.... 1. t. t. 3. a.. !. s..s.t s....s, a.,4.,.4....,
2 a
l
..y
,.:...:..:..:g.e...
4...
.t..... 3.,-
.i<..g <.-
2..
e e
r -
ew y
-u-rr-*
m e
-4
+---s-
---+--------------r------3------r
-mc v-
l..*..
i;t.e.c *.e....
n*,s.s.e.=
.s.
p... e..... s
- e... s. 4. e...
J
..e.. a. s. e a. s... *....y s.ys.t.
r iz.e.e.t ) d e.... e V. e.,...... a e... e s. j.,..., e... e 5 4
...a4..
s s.
- o...,. t.. t. t. oy..t..
a t.,..a.y.
.~
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.t,.
.e <.,.i 3.ac4...
l.
4
.:s. ss..
s..
... 1..sicts s.
!. I ine a:;* f: ant has summitted the 1c.-::-er ;nysi:s test ;re:e:gres t: '.estingn:vse for revie.-
(;* #
, at: Vestin;r.:ese ::t ents have teen Pt:e'vad at ne *arley ?'ar.:.
.e re:uire tha :: ren's
/
I!als: :s :r: vide: ty Vestingneuse ':r :ne au;. rented 1 - ::.ar tests (! e-I..1 c' this i
0 'i. se: ':n).
The Office =f Ins;ection an: infer:e:6nt wili Serify f: 'illeent of this re:uire-4
..t ) "'.
i I e.: :r'er := fusi lea:ing.
i
.i I.C.1 C:.atrel Re:S Cesien i
l
- cu' e-ent e
Il, Serfer: a ;reliminary assessment of the ::nti:1 rec to i:enti'y signi'icant human f a:::rs ce'f:Te.:Tes and instrumentation ;retless and esta:iish a s:he:ule a;;r:ved by the NRC f:r
....e..s..,.e f i
- a. e n.4. e s.
a This recuirerea.t shall be met bef:re f-el 1:adin;.
....s...
... 1..s...s
.s s. ss;..
t..
s i...
..I s., s. a a..<.s.... a.,...e.u..&
a
.s.
e..,..a s.a<.<.e..z..as
..a.
a
.g.<..
4.
..,<.g..
...I..t.3 g...
...<.g..,,........<..,
4.
,.3,3 a.,
i
.e..t..
a4 !X.e...ese..yje.s...e 4
- 4.. '. a *. s* ".#.*'.~*
.. - *.. e x *. s e '. 8.. a I *...*.. s a *.*. 8 t
. ~.
- t: :y tr.e en: ef 1582. As a.. interim measure, *'a:ssa ?:.er *:::any ( A?C ) was l
e l
l e:.3 e: :: ;erf:re a preliminary casi;r. assessment of tr.e Unit 2 ::nts:1 rec: :: ice-ti'y i
si;r.'": ant r.usar. "a: crs ce'iciencies an: instrumentatic.. ;r::lers. Results c' A:*:'s a s s e s se er.: are ;r:vice: in a June.0, *.550 ie ter :: :na NR*.
The NRC staf' anc i s
- Esultzn felle.ed up the ASC: assessment vita a 5-cay ersite ::ntr:1 ree: aud't.
The i
~
a.. t.. : a....e, s. 3...
.es.e.....
.ew.., assess.,.
.o.
.....; g
.a. s.,3 a,y ge... a f...,
z
... n..
. s g a.. a.. y a.,..,...., e. e. e s s
- a. s e,. e... s...e..,e i
.s~.,.
.....,,.,, z.a...
.a.t.s..,.
I,
....t....es.
.t
,~..a.... a s... s... e...
.,a.s.
e.a..e...s.,.....c...e.......a.,........ -..s.
a.
s l
.......,.e.............e..,..a.
-,. -. =.. t..- s.= s... =..v.a'.4.-
g..
.5
... 3......
.,...g, i
t
,3.
I l
o,...,
,ss..,..a. s,
- s..,..,.c..,.a............,
I
..,..a<>,.
g.,
t
....'..-=
t-.'.-
a.
.at.s.s,-=..-.- -.-.-.
2..-.
.e :.-. 4.'.' - *.-.
- c........
3
...t-
..i.........,
..s..
,4 l
.4
.,.....,.....g e.
y....
.4 l
.g l
- 1 l - -,
, alrea:f *n=e.ay.
n:.ec,er, n:ne cf :nese :sficiencies ef'er any significant safety risk ;c 1
j fuei *.:acin; an: Ic. ::.e testir; :e:ause stere are larger tr.er:a1 =ar; ins :: ne ers e; :.'
e4:te:1 ; '.el cesi;n 'irits an: safety ;ir.its curin; i:. ::.er c:erati:n : tan :uring full
,.-t...e.a...n.
j' s];,,
I *n :r:er :: ::rre:t : ese :e'icien:ies, AFC: an: the staff nave a;ree: that ex:e;; as r.::e:
. u.. ?.'.a.,.
in *:e-1~,
- e "
- 1 *:.'.; ::1utiens.i',; :e 4::le-ente: :-i:t : es:aiati:n :ey:.: five o..
,..; s. %,. e.. e.... e..
e
..........,;.a.yg.
. g
.sg.,.,...~....... a. 3 g. <. e.
g.,. a......, g a......i..<....,....g
.....e, s
..e
..... 1....,ti*.i.,.
.-a 4+
ar.
=.11
.ei.a.e. 4
- ..e
.... 3,i e...s t.s a u..
.... e an: re:alance ne "i: rates ste:v;n: : :ne system. This reic:stien and re: alan:in; sn:u;: e:::e ice ta:k;r:unc n:ise :: an a::e;;a:1e level (less :han 5~ :5(A)).
1 8
2.
4-.u :iat:r 7-ie-it':stien.
AP : 111 :evel p a list cf annuncia :rs :..a snculd e:tive n:re :: erat:r attention. Teese an un:ias:rs wiii be ;ricritt:e: ty ::ler, t
3.
aeru :dat:r Aleres. A?Co will increase ne main ::ntr:1 beard, balan:e :f plant and e er;en:y ;.er b:are annun:iater alar: levels :: 5-ic!(A) a cve the a::ien noise level. Fecu:tien f ne ta:k;r:un ncise vill result in ::re aucible alarm levels.
l 4
Acci: ental A::uatien. APCc will extend the h:ri: ental ;ortion of :ne main ::ntrol
- ar: :: ;revent inadvertent ;eration :f ::ntreis.
Y i
5.
Celer C::i c.
ace will review tre ::ler ili: : for :emar:ati:n. A ::ler will :e 1
uset. a Or: vices signi'i:an:e and St ta:e wili :e :t-sa ently installed. Celers use: ':- system :is:risina-i:n.11' :e evie.e: i.. :r:er :: re: :e :Se nu :er cf ::1:rs s..
a... ase s,..s.e
.a.s.
.a.,.a......
..e.,...
<.t.
..e..... a:...
..,..e
.,.e e..t.e.
1
.e..e.e.
.,2..
.c.
....a.e.
.,s.e..z...........
..i..
1
.a..
t..
g g g,.
4.,.. < e. 4.. g. 4....... a *.. - '. - *. -..a. s d. s *. a.. v.
I i
- -::ers : : uter. A? c will ::rre:: :ne ;a:tr 'ee: :r::le: an: ensure nat b:tn aln
- ..te:1 r::= :ata::e ray tu:es are ::eratie. A h::: will be a::e: :: :ne cathese ray tu:e 1 :ste: en :ne main ::ntr:1 ::ar reacter anel :: recu:e the glare. A :r:ss I
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e s.e <fi..,s.s.....se a.e g g.... g, s..... ) a. 4....es..
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5.
Natural cir:ulatien at re:c:ec : essure; l
i E.
C:01:own :a:atility of the : tar;in; an: let::.n system; 7.
Simuhte: icss of all onst e a.: c'fsite a: ;.er S.
Establishment cf natural cir:viation fr::: stagnant :en:'tiens; and M
i 9.
For:ec circulatica cccicevn (Far-A) anc ter:n eixing a.: :::1c:wn (Part !).
!a:n ap;11 cant f:r a full :::wer ::eratin; li:er.se mus :t ': : tests similar :: the a::ve
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its s:retale: star cf :ne tests :- :::::er 1, 1910.
II.!.
Edrin ':- "itica:In: : e Ca a;e
- e:.' e e :
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a..s..... a..
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t. s.. e.
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o 4
sysit-s. 4.:*.ti ; systats that are ?:: e*;i.et t: sa't:y features, ar: 'r.stra entat':- ::
i
...e.-.t.t,.e se.e t'.*.*...a,a...
...... g.....
.r..... a-.d..e...s d....-d-.
t i
T-is re:.'rerent sna11 te cet Saf:re fuel 1:a:in;.
I I
revd-de-Tne s t" re:uires tr.at Me a::l': ant :evel:: a :r:;rt: :: ensure that all ::eratin; :ersonnel are trained in the use of instal'et :lant systars :: ::n r:1 :r mitigate en a::iden; in 1
(
.i:n ne : re is severely cacage:. Tne trair.in; :r:;ra: shall inclece the fellowin; :::i:s.
A.
I ::re Instrumentatien q
1.
Use of fixec er ::vatie in::re cete:::-s te determine ex:ert Of cere damage an:
ge::etry :Banges.
2.
Use of there : vples in :etermining :tak ten:eratures; reth::s 1:r exten:ec range restings; re:hecs f:r : ire:: rea:in;s a: te minal.' unctions.
1 In:: e V;:'ea- : stru enta-':- ". 51
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' a * : strase-tatien 1.
- nstrurer.;a-icn res: rse in an a::i:er.: env'r:r.:er. ; f ailure se;uer.ce (tine :: -
fatiure, etne: Of f ailureh 'ncicati:n :f reita:it' y (a:: cal vs in:ica e: '. e v e i ).
'I: trna.ive cett::s fer retsurir.; ' :.s, :ressures, *.evels, zet tan: epa;; es.
1..
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- arn:t :e c.et feasi,tly :y Ja.uary 1. "I-50, a jus-iff:a ':n steule to :r:vi:e: f:r less tnan seis: 4 cualif.1:stien an: a s:Be:St sr:v M te se:r hte: h r v:;ra:e :: se re: iret seismic ;ualificati:n.
.. e. 3.. t.
4.
4. g. <. 3...u... e. 3
- 4.,.,. g..,.. j. s t. -.. - ' a. e e.. d.
.. s' a.. e
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transie : cr accicent.hien -:wM :ause t.e relief cr sa'ety va*.se :: *i't).
If the e... $.... e.. a 1... a g j e. s.,. a. d. a n..,. a.
'. -... ' s. s d. *. ". -. '. ".. a. '.. a.*.....'.e..-...'ia..t.'
. a.. a.y,., s.e..e m.. a....:s e. s..... i t s....... e. u... o.. e e. s..... e.. c.,. a e. o.. a. t...
. r., - a s..- ". I d.
e *f V'.
ed.
. s s,.s..i...... t. u s 4.
. s T.: ::.er-::erated relief valves (FCEV) an:.hree safety valves (SVT cen.e:te: :: the ::p o'
- ne cressuri:er are er:leye
- revke ever:ressure :r:te:. bn f:r the react:r 00 lant sys e: a -arley Unit 2.
7:sitive 7 8.V ;:shi:n in:kai'en '.s
- .aine: :y stat :rcua.te:
limit s.itenes -ni:3 ::nte:1 indi:ating lignis :unte: :n tae main ::n r:1 ::ar:. Tne li-it s i*: es are :unted t sense the fully ::en an: fully :1: sed valve sta. : si f=n.
Lic h s'. h:nes are ;ost-accident envirence..t cualifie an: seismic excitat'en cualified sw h enes, an alar: nas teen ad:e" in tne main : ente:1 r::: : b:h tte wnen any FORV is n:t fully c':se:. Inis aiare is har:sirec, i.e.,
41ar: :;eratilhy is n:t ca;eecer.: en the :lant
- :ute. The indica::rs,' one set c' red and green lign** ;er PCRV,.are ::were: "re the Class II ce distribution system. The PCRVs are air c:eratec em:1:ying a 5:len:ic :: c:ntrcl 1
l instru:ent air. The FORV solenoi is :: ered 'r:: ne same *! :us as.ne c:rres;cncing a
valve ; sition inci:ation. Centrol and inckatien Of a PORV ill be 1:st in the event that the vs is i st.
The PORV is cesigned :: f ail :leset :n 1:ss f ::we.- :: the ::ntrol sciencie. This ::n'igurati n is ::nsicere: a::e:;2:le.
te:-.:gr.e: lieh swhenes als: are :ur. e: :. ea:n sa'e:'.a've ster :: :r:e.3a ::e. a :
r i
t
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,E 4-ala P. s '. tr.e :ti n ::n.r:1 r::-.: in:i:sts
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a.y safer, veive is n::
'.'.iy c!:st:.
- /.' as: SV ::sition beicators are single-:nannel sys.e?.s.
As :a:ku: in:Ha:1:.., there exist tem: era.ure :e e:: s :n s'i slie' an: sa'ei.v valve ail :ie:es..ict..cin a ::- :n i
Sea:er anc :f:ing *un :: the press;.ri:er relie' ang.
Te::erature, pressure, an: levei a.e
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- iscussi:n cf Exis*ia:
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- Ce*.*."...**.*.
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s' :... s ' s'.'. -. a o.-.' - - i c 1 e s,. e..-. a - e ' '. '..-. - e *. ' e -. - -... '. s. s s'.. =.. w..s-t.\\, a.-
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- s.v s *.e an. "...e -- = s s. - 4.. a..
a'. e. "... a... s "- -. '. '. "..,.e. c>
'... '. a.v r.
- nsis.s f t.c analog an: eigital etters ::unte en the main : ente:i ::ar:. Tne.:arley Uni; 2 will use :ne cecicate: eigital :alculat:r :: :aiculate margi- :: sa ura:ica usin; 4.....
..... e 3 :, e s.... s s.' '.. e... e s s ". =. a..- "... e.". '..,. e s. '.... '. t, : '.. t_-. a.. a. ".. e measure-ent :r : re exit trer==::v:les. The :arrent main ::ntr:1 ::ard rea::ct is pressure s a.".-a.i c.. ::er en y.,cea. u r e s e s -.. e "... e u'.'. l i a *.'. -. '. *...a. *"..-..,'i~.,.
- n. i... a..d a:;en:e: :crtions of the staa: ta:1es :: : ster:ine su::::iin; ::n:itiens in cegrees Eahrenheit.
L)[
I A14:ama Fewer Cc= any is ;ursain; with *.'estin;n:use Electri: C:r:: ration a ciner change ::
.,?
- . previce main centrol board reaccu in degrees Tahrenheit. A ces:ri;;ien cf the : dificatien
,w.w.
' re: wired t it;1ecent this :nange -ill te ;resente: ::.ne !G pri:t :: its :::;1etion.
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...... ".... 4. s i.'.s *.a '. l e d
, ' ' i
[anc::eratienalpricttofuelloacing.
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.... 4 e.
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- ar:. The u:cer limit c' :ne rea :ut is in ex:ess :" 2200'~.
!..e se::n: means availa:Te f r :: nit:rir.; tre ::::v:1e tes:erature is the in-::re there:-
.... e. =. a.- - -...> n e 'i.. a *. e. 4. s' a e.....-. e s a '. s.,.' a. s s e ~.. '.. n.. '. "...a.
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=..m....a..... ss.
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... is
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et-.-.
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~.e ycw8.,
m'.s a.....s a..... -.. a. t -..;" s. e.= m c"....- (=* s-.t. t..*. ')
an.
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's e.s e r.*. c,.
's ' ^. - a. - t... '.
)
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O g.?. q y... t.. s
(,.*.a.. : *
- t,.te
- 4. 3 s
r 7.2} have sn:.n that : rima y syte- :ressure in:reases Of less snan 100 ; un:s ;er s: care 33 3 a.e -.+. '..-.ed
'. -..s e-.e.. s.a. 1.ad. - e.' s -. '.... s '. - --. a. e. '. 'a 'i s ~.. e. a... *..-. i.1
- ~
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...aa. '. ".. --".. s -*.. s.-"..*. e i n..5 w u 1 -..-..- e....e o
s.
FORV. Fr:: :rese results, it can be reasent:1y esticated that a 50 percent lead refection frc c:erati:n at 50 percent =cwer w:uld :r:: :e similar pressure transients; n: wever, analyses :f su:n an incident was r.ct included in the stu:ies.
4Ine a::licant has incicated in a rettir.; with :ne staff ina: the 1:ad refe:tien transient fr:: 50 :erce.. ;cwer will be analy:e: an:
..1 test :sta exists *nien scuid serve ::
ff )-
s c........e a ai.v.4. a3...,4.
- 4..s.
... j g..., <
., <-. 4.3
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3.
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a.. t t.....
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.'et:rstrate na; the FORV ins a'le: in :ne ::ar. nas a failu e rate ecuivalent : Or less i
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a-ergen:y res:ce.sibilities. The e-er;e.:y ;ia. Ces: Hies the ::c-c'nati:n :f the a--an;e-ran s and a; ee.ents :at.een :ne lice.see an:.ese a;e.:'.es.
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for Oeveic::ent an: Ivaivation e' State at: '.::a1 hverr. en: Eaci:1:;icai 2:er;en:y Res;cese Plans of Fi[et Nuclear Facilities" (NURE; 75/111). ' :le:in; Sv::leren N:. 1 to ina.
- u
- lica:icn :ated F.a.r:n 15, *.577, wni:h 1:entifies n:se ite:s esse..tial fer NRC's ::r.-
- green:e in a 5:ste :ian. As a result f this revie. a : in a:::r:an:e iu :ne previsier.s
- f the Feteral Register Neti:e (Vole e 20, N:. 245, Oe::: ar 22, 1975), ne NR-'::n:urrac fcrea,,1y in
.,.e e.....n:r.e :n re:rua y.,
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Revisiens :.he State c' Alataca and the State c' Oecr;ia Racicio;ical E..ergency ;erations Flans are being su::itted to FEr.A for review. These cra't lans were written to meet the esser.tial re:wirements of NUREG-0554 By letter :: e: Au;ust 23, 1980, Fi.".A fines our ree:::endatien 'cr issuin; a 'uel 1: acing and 1:w ::.er testing license to be reas:nacle (See A::endix to this se:;1ement).
As a rescit cf the Oct.ission's a:.icn :la. f:r Sr::::iy '.':;ra:in; Emer;e.:y are:arteness at 7:.er Rea:.:rs (55 Y 75-453), na Irergan y Elan '.; Ee t't.'
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ra:icis:::;i and chemical analysis :f rea:::t :: clan. and :en. air.:ent at::s;nere sar ;1es un:er ce;racec-: re ::ncitions with::: ex:essive excesure.
This e:uirement shall be ce* by,.'anuary 1,
- 951. (See NUR~0-0573, Se.ica 2.1.34 an:
letters Of Ce;*e r,er 27 anc Nove :er C, 1973.)
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a CONTROL NO'O 9 5 6 ACTION CONTROL QATES FROM; coua' oE AouNE 9/p/WL Chairman Ahearne ACvNoWLEoGYENT t,/
OATE OF OOCUMENT 9/3/80 INTERIM REPLY PREPARE FOR SIGNATURE TO; plNAL REPLY j
CHAIRMAN Dir., NRR FILE LOCATION O Executive eiREcTOR Denton
- oTsER, D ESCRIPTION O LETTER OMEuo O REPORT C QTHER SPECI AL INSTRUCTIONS OR REMARKS Requests response to questions concerning ji the Farley Power operating license prior g
to approving full power license f PRIORITY MAIL Y
-Q, C,,1 ::_ ca.c
%% 's g x y pc.fe l. N A
$y)L
',*J f
CLASSIFIED DATA ctAssiricATioNj S b"C ~
EM h occuuENTecopy No.
cATEcoRv NuveER or Pacts O NSI C Ro O fro aoSTAL REGISTRY No.
ASSIGNED TO:
OATE INFORMATION ROUTING LEGA L REVIEW C elNAL 0 CoPV NO LEGAL Os;ECTiONs Denton 9/5/ SO UlrcKS rianauer AssicNEo To:
oATE NOTIFY-DEisenhut
! 09 0A Cornell Ross g
Rehm Schroeder g x y, COMM ENTS. NOTIF Y:
f_
/ h U-w Case Snyder ext. ____
I Denten Vollmer C YES C NO l
l Grimes PPAS
- CAE NOTIFICATION RECOMMENDEO
NRC FORM 232 EXECUTIVE DIRECTOR FOR OPERATtONS DO Nor REMOVE TN/S COPY
. PRINCIPAL CORRESPONDENCE CONTROL 7:M***-
P r
i e
ENCLOSl:RE COMPLIANCE OF FARLEY NUCLEAR ptAT.T UNIT 2 WITH TiiE NRC REGULATIONS OF 10 CFR PARTS 20, 50, AND 100 Regul a tion
, 10 CFR)
Comoliance
(
20.1(a)
. This regulation states the general purposes for which
-Je part 20 regulations are established and does not impose any independent cbligations on licensees.
20.1(b)
This regulation describes the overall purpose of the part 20 regulations to control the possession, use and transfer of licensed material by any licensee, such that the total dose to an individual will not exceed the standards pre-scribed therein.
It does not impose any independent obligations on licensees.
M A'.
20.1(c)
Confor=ance to the ALARA principle stated 7n this recula-tien is ensured by the imolementation n# APC colicies
'Tkg 9 and appropriate Technical Specification and health
/-S T physics procedures. Chapters 11 and 12 of the FSAR WM.w,,J R describe the specific ecuipment and design features utili:sd in this effort.
20.2 This regulation mer:1y establishes the applicability of the par: 20 regulations and imposes no independent obligations on those licensees to which they apply.
20.3 The definitions con ained in this regulation are adhered to in all appropriate Technical Specifications and procedures, and in applicable sections of the FSAR.
20.4 The Units of Radiation Dose specified in this regulation are acceptable and conformed to in all applicable FNp procedures.
g wu u a
20.5 The Units of Radioactivity specified in this regulation are accepted and conformed to in all applicable FNp procedures.
20.5 This regulation gov 5rns the interpretation of reguia-tions by de NRC are does not impose independent obligations on licensees.
20.7 This regulation gives the address of the NRC and does not incase independent obligations on licensees.
O g
a n p.-.w Lta,sy
~ut w aa us..e4 y&
q -a$k -@dQ m a a u a u m At.ru a w h w'- A c.aAbd a d wM ~?
J 20.101 The radiation dose limits specified in this regulation are ecmolied with throuch the implementa' tion of and adherence to administrative colicies and controls anc appropriate health physics procedures developed foF this purpose.
Conformance is documented by the use of appropriate personnel monitoring devices and the maintenance of all required records. (See FSAR Chapter 12) 20.102
. When required by this regulation, the accumulated dose for any individual permitted to exceed the exposure limits specified in 20.101(a) is determined by the use of Fonn NRC-4. Appropriate health physics procedures and admini-strative policies control this process. (See FSAR Chapter 12) 20.103(a)
Compliance with this regulation is ensured through the implementation of appropriate health physics procedures relating to air sampling for radioactive materials, and bicassay of individuals for internal contamination..
Ad=inistrative policies and controls provide adequate margins of safety for the protection of individuals against intake of radioactive materials. The systems and equipment described in Chapters 11 and 12 of the FSAR provide the capability to minicize these ha:ards.
20.103(b)
A;;ropriate process and engineering controis and equipment, as described in Chapters 11 and 12 of the FSAR, are installed and operated to maintain levels of airborne radioactivity as icw as practicable. When necessary, as deter =ined by FNP administrative guidelines, additional precautionary procedures will be utilized to limit the
~
poten-ial for intake of radioactive materials.
20.iC3(c)
The FN? Respiratory Protection Procedure implements the requirements of this regulation by ensuring the proper use
~
of a;; roved respiratory protection equipment.
The FNP Respiratory Protection Procedure incorporates fully the l
stipulatiors of Regulatory Guide 8.15 " Acceptable Pro-gra=s for Respiratory Protection." (See FSAR Chapter 12) 20.103(d)
This regulation describes further restrictions which the Commission may impose on licensees.
It does not impose any independent obligations on licensees.
20.103(e)
The notification specified by this recura ion was made as required on February 23, 1977.
l 20.103(f)
The Respiratory Protection Program is in fuli conformance with the requirements of 20.103(c).
20.10:
Conformance with this regulation is assured by accrorr4 2te APC :olicies regarding employment of individuals under the a;e of 1S and the FNP Heal-h Physics Manual restric:-
in; :nese individuals' access :o restricted areas.
a E
(vii) not applicable, since the operating license application was filed before February 5,1579 (7) technical qualifications - Chapter 13
.(8) operator requalification program - Chapter 13 and letters of 6/20/E0, 7/15/80, 7/2;/50, 8/1/50, and S/5/80.
The latter letters des: ribe the training program including a requalification program meeting the new requirements of NUR5G-0594 p,,,,) 1 g7M. C.34(c)
A ;bysical security clan has been prepared and will be n
y 7
impl =....a.n.s...
50.34(d)
A si'e;rn-Hs centincenev clan has been prepared and will be i ;lemented.
50.34a These regulations are relevant :: the construction permit 50.35 stage rather than the operating stage.
50.35 Technical Specifications have been prepared and will be imple-tented, in:luding items in each of the categories specified, in:iuding:
(1) safe y limits and limiting safety settings, (2) lini-ing conditiens for operation, (3) surveillan:e requirements, (4) design features, and (5) administrative O ntrel S.
50.35a The Enviren: ental Technical Specifications, Part I, inilude s;e:ifications which recuire c =pliance with 10 CFR 50.34a.
(releases as icw as is reasonably achievable), and that ensure that concentrations of radioactive effluents re-leased :: unrestricted arees are within the limits specified in 10 CFR 30.lC5.
The reporting requirements of 10 CFR 50.35a(a)(2) are aisc included in these specifications.
50.37 This regulation requires the applicant to agree to limit access to Restricted Data. APC's agreement to do s: is in the opera-ing license application for the Farley Nuclear Plan: Units 1 and 2 in Section 11.
50.35 This regula-ion prchicits the NRC frc issuing a license to foreign-controlled entities. A?C's statement that it is no cwned, controlled, or dcminated by an alien, foreign corpora-ion, or fcreign government is in the operating license application for the Farley Nuclear Plan:.
50.39 This regulation provides that ap;iications ar.d related documents may be made available for public inspecticn.
This ic;dses ne direct cbligations en applicants and licensees.
.ic-
O 8
50.40 This regulation provides considerations to ". guide" the Commission in granting licenses as foll:ws:
50.40(a)
The design and operation of the facility is to provide reasenable assurance that the applicant will comply with NRC regulations, including those in 10 CFR Part 20, and that ne health and safety of the public will not be endangered. The basis for APC's assurance that the regulations will be met and the public protected is contained in this enclosure and in the license applica-tien and the related correspondence over the years.
More-over, the lengthy process by which the plant is designed, constructed, and reviewed, including reviews by APC's own staff, the NRC staff, the ACRS, and NRC licensing b:ards, -
provides a great deal of assurance that the public health and safety will not be endangered.
In ; articular, :ne At::ic Safety and Licensing Board, after an extensive review, concluded that APC had the cc:n.it:=nt and technical cualifications necessary to cperate Farley Nuclear Plant safely and in c =pitance with all applicable radiclogical health and safety requirements.
50.40(b)
An ther c:nsideration is that the applicant be technically and financially qualified. Ecth APC's technical qualifica-tiens and its financial qualifications were reviewed in hearings before the At: ic Safety and Licensing 50ard.
50.20(c)
Another consideration is that the issuance of the license is no: :: be inimical to the cc:ren defense and security cr to the health and safety of the public.
The individual shewings of compliance with particular regulations contained in :nis enclosure, as well as the contents of the entire FSAR and related correspondence over the years, plus the lengthy process of design, construction, and review by APC, its NSSS vendor, and the government, provide APC with considerable assurance that the license will not be inimical to the health and safety of the public.
As for the cecon defense and the security, there is considerable assurance that the license will not be inimical in that APC has an approved security plan for Farley Nuclear ?lant tha: A?C is not c:ntrollec by agents of foreign ceuntries, and that APC has agreed to limi: access to Restricted Data.
1 50.20(d)
The final 50.40 " consideration" is that the applicable requirements of Part 51 have been satisfied.
Part 51 concerns compliance with the National Environmental Policy -
Act of 195g. A?C has submi :ed a Final Envii once-a'. Re; ort and the NRC staff has reviewed the rep rt and *?C ;ublished a final Environ ental I :act Statement ;ursuant :: 10 CFR 50, A u k @?
A;;e. dix 0.
u Envir:r ent'i Tecnnicai 5 e:ifica:ic s are a
__;ending. '
6 50.01 This regula: ion applies to Class 104 licensees, such as those for devices used in =edical therapy.
Farley Nuclea Plan: has not applied for a class 104 license, and so 50.41 is act applicable.
50.42 See:1on 50.42 provides addi:1enal "considera: ions" :o
" guide" :he Cc==ission in issuing Class 103 licenses. The two considera:icns are:
(a) tha: the p:cposed activi:ies will serve a useful purpose propor:icnate :o the quantities of special nuclear :aterial or source =aterial to be utiliced and (b) that due account will be :aken of the 1
j an:1 :us; advice picvided by the A::orney General under subsee:1:n 105c of :he Ato:ic Energy Ac:. The "useful purpose" to be served is the p cdue:1on of electric power.
The need for the power was deter =ined by the licensing board at the construction per=i: stage. A1: hough conditions afft:iing the need for power are cens:an:1y changing, A?C peri:dically =akes load ;;ojec;1:ns, and in A?C's judge =en:
he need for Tarley Nuclea: Plant is s:ill substantial.
As fe the a:cun: cf special nuclea: =a:erial or source
=aterial used, there is no reas:n :o believe that their prep ::icn in relation :o :he pcwer produced is substancially grea:e: than :ha: Of other c:::ercial pcuer reac: ors in this ccen::y. As fc: the an:itrus: review, the C ::ission referred the ta::e: to an A:::ic Saf e:y and Licensing 3:ard (Decke:
5: s. 5 0-31
- A an'd 50-3 61A) f e hearing and decer:1:a:icn.
Such 3:ard on April S, 1977, issued its Opinien, Findings
.and 0:dar in which 1: was de:ertined :ha applican:'s
~~
activi:1es under :he Farley ?lan: license would rain:ain a si::a:1on incensisten: with certain an:i;rus: laws and :ha: i: w:uld be necessary :: at:ach additions :c the li:ense which will prevent such a resul:.
Subsequen:1y en June 21 such Board a: :he foc: of :he fur:her proceeding rela:ing :o :he =a::er of wha: cendi:icas should be a:: ached :o the license, issued a further order prescribing definitive license conditions which were appended to :he l
cpera:ing license for Farley Uni: 1.
Exceptiens to the
(
decernina:1cns made by such 3:ard were filed by applican:,
Depar::en: ef Jus:1ce, the S:aff cf :he Nuclear Regula:ory C ::issien, and two in:erveners.
Se:S exec;:icns are new pendint before an Atomic Safe:y and Licensing Appeal Board of :he Cc :issicn.
l 3C. 3 This regulation i=peses cer:ain de:ies on the SEC and addresses :he a: licabili:y Of :he Federal Tower Act agencies :o eb:ain NEC l
and :he right of govern =en:
licenses.
I: i= poses no direc: ebliga:icns on licensees.
l 50.44 The Farley Nucles: Plant cerbus:ible gas cen::01 sys:c:
is described in TSAR Se::icn 6.2.5.
The sys:c-is d(signed l
- o main:ain :he hyd: gen c:::en::a:icn in cen:ain=en: a:
i l
a safe icvel fell: wing a LCCA, vi:h::: purgin; :he con:cin-nen a::: sphere, as specified in :: 07; 30.ia(c).
The 3,
t s-I
J 50.46 In FSAR Amend:ent 72 dated P. arch 7,1950, ApC provided the (Cont'd) results of a LOCA-ECCS analysis for Farley Nuclear plant using an NRC-approved evaluation m: del which is in ccm-pliance with Appendix K to 10 CFR 50. The analysis, based on an overall peaking factor (Fq) of 2.32, provided resul ts in cc=pliance with the criteria of 10 CFR 50.46(b).
The Fq limit is reflected in the Technical Specificati:ns.
,The Staff has proposed new standards for fuel cladding swelling and rupture medels in draft NUREG-0530.
If these 42 }{ <
standards were to be adopted for the currently approved evaluation codels, a decrease in the Fq Technical Specifica-3
33 4 tien limit would be required. However, by letter dated August 6,1980 ApC provided an assessment of this reduction as well as the benefits of modeling it;rovements described to the NRC by k'estinghouse Electric Corporation.
It was
, concluded that application of these improve;ents would more than effset any Fq reduction resulting frcm the revised staff cladding model; therefore, the existing Technical Specification limit is acceptable.
50.50 This regula-ion provides that the NRC will issue a license upcn determining that the application teets the standards and requirements of the Atomic Energy Ac and the regula-tiens and that the necessary notifica:icns to other egencies er bodies have been duly made.
It impcses no direct obliga-tiens en licensees.
50.51 This regulation specifies the maximum duration of licenses.
Ccapliance will be affected simply by tne Commissien's writing the license so as to comply.
50.52 This regulation provides for tne combining in a single license of a nu=ber of activities.
It imposes no indc-pendent obligation on the licensee.
l 50.53 This regulation provides that licenses are net t: be l
issued for activities that are not under Or within the i
jurisdiction of the United States.
The operation cf l
Farley Nuclear plant will be within the United 5:1 es and subject to the jurisdiction of the Units: States, as is evident from the description of the f acility in the operat-ing license application.
50.54 This regulation specifies certain condi:icns that are i
inc:rporated in every license issued. C0mpliance is l
effected simply by including -hese conditions in the license when it is issued.
Indeed, cuch of 50.5
- merely provides : hat other provisions of the law apply, wnich w:uid be the case even witncu-50.5 *.
13-l l
l l
50.55 This regulation addresses conditions of construction per-mits, not operating licenses, and so it is not relevant at this point.
50.55a(a)(1) Various chapters of the FSAR dis:uss design, fabrication, erection, construction, testing, and inspection of
. safety-related equipment.
For example, Chapter 14 pro-vides information on testing of safety-related systems.
Chapter 17 provides information concerning the Quality Assurance program that was utili:ed. As a further example of a specific system, Chapter 5, Section 5.2, " Integrity of the Reactor Coolant System Soundary," discusses the design of the reactor coolant system.
50.55a(a)(2)
This paragraph is a general paragraph leading into paragra;hs (c) through (i) of the regulation.
50.55a(b)(1) These paragraphs provide guidance concerning the ap; roved 50.55a(b)(2)
E:ition and Addenda of Section III and XI of the ASME B&pV Code.
50.55a(c)
Design and fabrication of the reactor vessel was carried ou-in acc:rdance with ASME Section III (1958) Class A.
Inf rmation can be found in Chacters 5 and 5 Of the FSAR.
5".55a(d)
Rea:::r coolant system pipin; mes:s the rsquirements of
~
ASME Section III (1971) Class 1.
Information can be found in Chapters 3 and 5 of the FSAR.
50.55a(s)
Rea: tor coolant pumps meet the requirements of ASME Se::icn III (1971).
Information can be found in Chapters 3 and 5 of the FSAR.
50.55a(f)
Tne valves within the reactor ecolant system pressure
~
boundary were designed and fabricated in accordance with the re;uirements of ASME Section III,1971 edition. (See FSAR Chapters 3 and 5) 50.55a(g)
Inservice Inspection (ISI) recuirements are delineated in this part and are specified in the Techr.ical Spe:ifica-tions, paragra;h t.0.5.
As permi ted by*this part ar.d the Technical Specifications, certain exer::icns hare
?
been recuested for the inservi. e ins ection of vcy.ious.
'"N g -
systems and the inservica testinc cf va-icus Ouros and va l ve s._ Sy letters dated Mar:h 12, 1950 and July 25, 1980 the Farley t;uclear plant-Uni: 2 Ir. service Testing pr: gram for pur.ps and '/alves ar.d Inservice Instsc:ics ?r:grar were _
- cketec. The preservice Inspection pr;;rar and reiisf recuests for this pr:grar are ine:r;;ra:sd in FSAR Se:-i n 5.2.5.
-la.
50.100 These regulations govern the revocation, suspension, and 50.101 modification of licenses by the Ccr. mission under unusual 50.102 circumstan:es. No such circumstan:es are present in 50.103 the Fi;p proceeding, and these regulations are not a;plicable.
50.109 This regulation specifies the conditions under which the NRC may require the backfitting of a facility. This regulation imposes no independent obligations on a license unless the NRC proposes a backfitting requirement, and so this regulation is not applicable.
50.110 This regulation governs enforcement of the Atomic Energy Act, the Energy Recrganization Act of 1974, and the NRC's regulations and orders. No enforcement action is at issue in the FNP proceeding, and so this regulation is not applicable.
Appendix A GCC 1 Se: tion 3.1.1 of the FSAR describes the design' provisions cade :: ensure that these requirements are met.
Cedes and s.andards utilized for the unit are specified throughout the FSAR.
Chapter 17 describes the cuality assurance pr:gra: and the provisiens for maintenance of records.
u.~
2 FSA.:. Section 3.1.2 addresses the design considerati:ns for f
natural phenc:ena, which are described in detail in Cha;;ers 2 and 3.
Appropriate considerations have been cade in de design basis for hist:rical data, ccmbir.ed effe::s of normal and accident conditions wi-h the effects
~
of natural phenomena, and the imp-ortance of the safety functier.s to be performed.
GDC 3 FSAR Section 3.1.3 cescribes in general the ceasures which have been taken to minimize the probability and effects of fires and explosiens. Section 9.5.1 describes f
the fire detection and protection systems.
In addition, improvements to the fire protection systems have been and are being made in a::ordance with i;RC requirements based on Appendix A to BTP ApCSB 9.5-1.
These difica-
?
9-ewillbe e-l eted bv Neve-be-1. 1950.
This s:nedule Y
is in accordan:e with tne C.missioner's Order dated May 23,1930 i
N jLn wyu.4 % my.
aw p.
~
17 I
1 i
GCC 4 Structures, systems, and cc:p$nents important to safety are designed to acco=modate the effects of any environmental conditier.s at the Farley site that are associated with n:rmal operation, maintenance, testing, and postulated accidents, including icss-of-coolant accidents, and to be cc ;atible
. with these c:nditions. Design criteria and implementation
. are ; resented in Sections 3.5 and 3.6, and environmental facters are described in Sections 3.11, 6.3, and 6.4.
These structures, systems, and ec=penents are appropriate'.
protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result fr:: any postulated failure of equipment or structural design or frc environmental conditions and accidents cutside.the nuclear power unit.
Chapter 7.0 lists the ::t:rs, instrumentation, and asse-ciated cables of protection and safety features systems located inside the containment structure. It als5 gives the desien
~
requirements in terms of the time that each must survive the extreme environmental conditions following a LOCA.
Details of the design, environmental testing, and construc-ti:n of thase systems, structures, and cc:ponents are included in :ther sectier.s of Chapters 3.0, 5.0, 6.0, 7.0, and g.0.
Evaluation of the performance of safety features and analyses of p:ssi:le accidents are contained in Chapter 15.0.
~-
In addition, a review cf sco:ed class 1E electrical equipment inside and cu: side the cor.tainment against PJREG-0535 is near completion. This review has confirred that ::st electrical ecuic:ent has been adecuately demonstrated to be cualified. The results of this review are scheduled to be submitted to the staff my Sectember 15, 1950.
4 g
g --M g% q
?
m
GDC 54 piping systems penetrating tne centaircent, in so far as (Cont'd) practical, have been provided with tests vents and test connections er have other provisions t: allcw periodic leak testing as required. Section 5.2.1.4 has further details en testing. See Section 5.2.4 f:r general contain-ment isolation details.
G;C 55 As discussed in FSAR See:icn 3.1.48, each line that is part of the reactor coolant pressure boundary and that penetrates the reactor contairment is pr vided with containment isolation valves, in accordance with this criterion, as de-scribed in FSAR Section 6.2.4.
Isolatien valves outside the conuinment are located as close to the centainment as practical, and autora fe isolation-valves are designed to take the position that provides greater safety upon icss of actuating power.
C her appropriate requirements to minimize the probability cr'.cen-sequences ef an accidental rupture of these lines er of lines-connected to them are provided as necessary to assure adequate safety.
Determination of the appropria eness of these require-ments, such as higher quality in design, fabrication, and testing, additional provisiens fer inservice inspection, protection against more severe na ural phenc ena, and additional isolation valves and contai=en, includes censidera-ti:n of the ;cpulation density, use charac eristics and physical characteristics of the site environs.
GDC 55 As discussed in FSAR Sectien 3.1.49, each line that connects directly to the containment a=: sphere and penetrates the '
pri:.ary rea::Or contairment is provided with containment isolation valves in accordance with this criterion, as discussed in Se tion 5.2.4 Isolation valves cu: side the containment are lect ed as close to the centai=en as practical. Upon 1 css of a:: ating power, aut::atic isolation valves are designed to take the position that provides creater safety.
t l
GDC 57 As described in FSAR Section 3.1.50, each line that penetrates containment and is neither par: cf the reactor c:clant pressure
~
boundary r. r c0nnected directly to the containment a=: sphere is provided wi-h an appropriate ::ntai=ent isolatier. valve arrangement. Refer to Section 5.2.4 for a cc= pie:e descripci:n of the design and operatica f the ccr.:ai=ent isolation systa.
GCC 50 As described in FSAR Section 3.1.51, centrei cf waste gas effluents is accc plished by hoidup cf.aste gases in decay tanks until the activity cf tank contents and existing environmentai conditions perni: dischar;es within 10 CFR 20 and 10 CFR 50 recuiremer.:s. Waste gas efficents are :: nit:re:
at the point Of discharge fcr radica:tivity and rate cf fic...
Su ffici er.: waste gas h:ic.;; ca;a:ity is ;r:vided as discussed in Section 11.3, :: c:pe ui:n all anticipated operaticnai occurrences and site er.vir:rcental c:ndi:icns. A de:ay unk
/M M # 'yd burst would ro: result in an a::ivi:v release creater~;han r
~
M[ A N.
10 CFR 100 li9its, based en cre :eriert failed fuel.
i M. J N A M g *IA. M d pb
? f Appendix E This revised plan, which see:s the criteria in NURIC-0654 (Cont'd) has been sub=1::ed to :he NRC s:aff.
Appendix ?
This Appendix applies to fuel reprocessing plants and rela:ed was:e managemen: facilities, no: :o pewer reactors and is
- herefere no: applicable to :his proceeding.
Appendix G Fracture :cughness requirements of this Appendix and progra=
require:ents given in Appendix H f ort :he basis for Technical Specifica:Lon surveillance require en:s dealing with the use of surveillance specimens. Addi:1onal infor a: ion to demon-s::a:e cenpliance can be fcund.in FSAR Chapter 5 and A?C le::er to :he NRC dated July 30', 1980.
Hea:up and coolious
~
li=1:s censis:en: wi:h the requiremen:s of this Appendix are es:ablished in :he Technical Specifica: ions.
Appendix H React:r vessel ca:erial surveillance program requiremen:s are delinea:ed in this part. Technical Specifica:ica 4.4.10.
1.2 and Operating procedures have been established :o inple:en:
- hese require ents with a fur:her receire:gn: c update :he "hes:cp" and "cooldown" curves provided in the. Technical Specifi-
~
cations.
Further infor:t: ion is provided in FSAR Chap:er 5 and AIC le::er :o :he NRC da:ed July 30, 1930.
Appendix !
This A;;endix provides nu:erical guides for design objectives and liri ing conditions for c; era: ion to ree: the criteria "as 1:w as is reas=nably achievable" for radicac:ive za:erial in e
ligh:-wa:er-cooled nuclear p:ver react effluents. A?C has M'W.*M filed with the Cc :issien the necessary informa: ion to d
2 '.
per=i: an eva ua: ion of :..e FS? vith res:ect to the requirements
^ ^ ^
ed se:: ice II.A.,
II.3, and II.C of Arrendix I-In :his subnit:al A?C provided :he necessary informati:n to shew cen-f or:ance vi h :he Cc =issien's September 4, 1975 amend en: to Appendix ! ra:her :han perform a de: ailed cost-benefi: analysis required by See:icn II.3 cf Appendix I.
Appendix J Rea ::: c:n:ainmen: leakage testing fer wa:er eccled power l
reae: ors is delinea:ed in :his Appendix.
These require:en:s are given in Technical Specifica:ica 3/1.6.1.
Additienal infer:ation
[
c:ncerning ecepliance can be fcund in ?SAR Chapters 3 and 6.
l Appendix K This Appendix specifies fea:ures of accep:chle ECCS evaluatien
- dels. As noted above fer 30.L6, :he analysis for 757 has been conducted usin a codel which has been accep:cd bv :'. e s
C:::issien s:sff as mee:ing :he require =en:s of :his Appendix.
Appendix 1 The an:1:rus: review rela:ing to the license applica:icas is new pending before an A:ccie Safc:y and Licensing A;;eal 3 card.
l f
_31-
4 In Supplement No. 3 to the SER, we also identified two generic issues. Our status of review of these issues is in the sections indicated below.
(1) Environmental qualification of safety-related equipment (Section 7.7.2)
(2) Anticipated transients without scram (Section 5.4.1)
All the new areas that have arisen since issuance of Supplement No. 3 to the SER, including TMI issues, have been satisfactorily resolved for fuel loading and low power testing except the following items, which will be license conditions.
)* '
o# ~
Priortoexceedingzeropower(i.e.,thatrequiredforphysicstests),redundantpowef (1) supplies will be installed on auxiliary feedwater flow control valves (II.E.1.2,
'j Section 22.2).
(2) Prior to conducting the augmented low power tests, applicant must receive NRC approval of its safety analysis report on the conduct of the tests (I.G.1 Section 22.2)
- 1. 9 Unresolved Safety Issues On November 23, 1977, the Atomic Safety and Licensing Appeal Board issued a decision (ALAB-444) in connection with its consideration of the application for the River Bend Station, Unit Nos. 1 and 2 (Oceket Nos. 50-458 and 50-459) which established specific requirements for addressing unresolved safety generic issues in connection with our licensing proceed-ings. Those requirements are applicable to the Joseph M. Farley Nuclear Plant, Unit 2 application.
Appendix C to this supplement presents informction for the Joseph M. Farley Nuclear Plant, Unit 2 application in conformance with the Appeal Board decision enunciated in ALAB-444 l
l s
s I
3 y
Equipment for Nuclear Power Plants," and IEEE Standard 344-1975, "IEEE Recomended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Dower Generating Stations." The principal change in our criteria is to require consideration of equipment multi-mode response and biaxial coupling effects. In view of these changes we considered it prudent to further review the Farley 2 equipment qualification program against SRP Section 3.10, to determine whether the original tests and analyses were adequate. Our previous review of Westinghouse equipment for the Farley Plant considered the effects of multi-mode response and biaxial coupling, and found this equipment adequately qualified. This evaluation addresses the qualification of BOP electrical and mechanical equipment.
Our Seismic Qualification Review Team (SQRT) performed a review at the plant site on July 7-10, 1980 to determine whether the qualification of the equipment / as installed in Farley 2, performed in accordance with the procecures of IEEE Standard 344-1971 could meet current licensing criteria as described in SRP Section 3.10.
During this review we evaluated a representative sample of thirty-four pieces of Seismic Category I mechanical, instrumentation, and electrical equipment. Cur review uncovered relatively few pieces of equipment for which it was not clear that the seismic qualification was acceotable in the light of current licensing criteria. For example, the battery charger in the service water cuiloing was mounted flat on the test table, while it is cantilevered off the wall in the field. Also, the solenoid valve in the river water building is field mounted in such a way that it may be susceptible to low frequency (below 20 hert:) input, yet the test was apparently conducted only for a frequency range beyond 20 hert:. The details of these shortcomings and others in the equipment qualification are described in the report of our July 7-10, 1980 trip to the plant. For these few items, the applicant has committed to submit additional information, clarification, and resolution for our review prior to approval of full power operatien. In addition, the SQRT has requested, and the applicant has provided, pertinent documents as well as test and analysis reports for five (5) pieces of equipment in orcer that we can conduct a followup in-cepth confirmatory review.
Based on the results of the review of installed equipment by the Seismic Qualification Review Team, we conclude that there is no severe discrepancy in the equipment qualification program with respect to SRP 3.10 criteria, and there is reasonable assurance that low power operation of the plant can be permitted at this time without endangering public health and
. y, safety. We will complete wr confirmatory in-depth review and require the applicant to D
clarify the seismic qualification of the ecuipment identified in our trip recort prior to 3
jaiJ
- y full power oceration of Farley 2.
e e
4
has provided data to demonstrate compliance with the Charpy impact energy requirements of Appendix G, 10 CFR Part 50. However, the applicant has not sufficiently addressed the generic requirements for high strength materials.
According to Appendix G,Section III of the ASME Code, the applicant must supply fracture mechanics data from at least three heats of the material and from a sufficient number of specimens to cover the temperature range of interest for any ferritic steel having a specified minimum yield strength greater than 50 ksi. This data must include static, dynamic, and crack arrest critical K values, all of which must be equal to or g
above the Kgg (reference) curve of Figure G-2210-1 of Appendix G,Section III, ASME Code. The applicant also mu*' 'emonstrate that the calculated stress intensity factors are icwer than the refe' 7 stress intensity factors (KIR)
Y. the margins specified by Appendix G of the ASME t. ode and as required by Paragraph IV.A.2.a. Appendix G, 10 CFR Part 50, to provide adequate safety for normal operation of the ferritic pressure boundary of the pressurizer.
We conclude that the applicant has not provided necessary and sufficient information to demonstrate full compliance with Paragraphs I.A and IV.A.2.a.
The applicant has stated that the information necessary to fully satisfy this requirement will be provided to us by September 30, 1980. The applicant has provided sufficient information to allow us to determine that prior to normal full power operation, the safety margins required for low power operation will be achieved and maintained. On this basis, we conclude that
' low power operation is acceptable. We will complete our review prior to full power y
- Q..
operation to confirm that adequate safety margins will also be maintained during normal 4
operation, including operational transients, in compliance with Paragraphs I.A and IV.A.2.a of Appendix G to 10 CFR Part 50, 2.
Paragraph III.B.4 requires that the testing personnel shall be qualified by training and experience and should be competent to perform the tests in accordance with written procedures. For Farley Unit No. 2 component testing, no written procedures were in existence as required by the later regulation; however, the applicant has supplied sufficient information to demonstrate that the intent of Paragraph III.B.4 has been met. The applicant has stated that individuals who conducted the testing were quali-fled by schooling, training, and years of experience and were certified by qualified supervisory personnel. Because these tests are relatively routine in nature and are continually being performed in the laboratory, we conclude that it is unlikely that the tests were conducted improperly. Consequently, we conclude that an exemption for not performing the tests in accordance with written procedures is justified.
3.
Paragraon IV.B requires that the reactor vessel beltline materials have a minimum j
unirradiated upper shelf energy of 75 ft-lbs in order to provide adequate margin for deterioration from irradiation. In weld seam 10-923, two of nine specimens tested had impact energies below 75 ft-lbs at a test temperature of 10 degrees Fahrenheit; no additional testing was conducted at higher temperatures to define upper shelf energy.
The applicant has proposed a correlation between Charpy impact energy and temperature for welds fabricated with the same type of wire and lot of flux as used in weld seam 10-923 of Farley Unit No. 2.
We have evaluated the applicant's additional data, which incluces a broad temocrature range over the lower shelf, transition and upper shelf temperature regions and have found that:
e
r (a) told seam 10-923 is represented by the additional da2a; (b) the additional data can be used to extrapolate to the upper shelf for weld seam 10-923; and (c) the minimum upper shelf energy is at least 100 ft-lbs.
Based on this additional information and our evaluation, we conclude that an exemption to Paragraph IV.B, Appendix G,10 CFR Part 50, is justified.
Comoliance with Accendix H, 10 CFR Part 50 The toughness properties of the reactor vessel beltline materials will be monitored througn-out the service life of Farley Unit No. 2 by a materials surveillance program that must meet the requirements of ASTM Standard E-185-73, " Standard Reccamended Practice for Surveillance Tests for Nuclear Reactor Vessels," and Appendix H,10 CFR Part 50. We have evaluated the
{ applicant'sinformationfordegreeofcompliancetotheserequirementsandhaveconcluded i that the applicant has met all requirements of Appendix H,10 CFR Part 50, except for bParagraph II.B, for which sufficient information has been supplied to justify an exemption.
Paragrapn II.B requires the beltline region of the reactor vessel to*be monitored by a surveillance program complying with ASTM Standard E-185-73. According to this standard the base metal and weld metal to be included in the program should represent the material that may limit the operations of the reactor during its lifetime. This selection is based on initial transition temperature, uoper shelf energy level, and estimated increase in tran-sition temperature considering chemical composition (copoer and pnosphorus) and neutron fluence.
According to our evaluation, plate 87212-1 and weld seam 11-923 are the most limiting base and. eld materials, respectively; the base plate B7212-1 is predicted to be the more limiting of tne two. The Farley Unit No. 2 surveillance program contains material from base plate l
B7212-1 and weld seam 19-9238. Because weld seam 19-923B is not the most limiting weld in l
the reactor vessel beltline region, the applicant's material surveillance program is not in l
full compliance with Appendix H.10 CFR Part 50. To have an acceptable surveillance program I
for Farley Unit No. 2, the apolicant must use the following analysis for every capsule removed and tested.
During the plant's life the applicant must recalculate the pressure-temoerature operating limits based on the greater of the following: (a) the actual shift in reference temperature l
for plate 37212-1 as determined oy impact testing, or (b) the predicted shift in reference temperature for weld seam 11-923 as determined by Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation hmage to Reactor Vessel Materials."
l Although material frca the most limiting weld seam,11-923, is not contained in the Farley l
Unit No. 2 materials surveillance program, we have found that an exemption to Paragracn II.2 of Appendix H, 10 CFR Part 50, is justified for the following reasons: (1) the applicant has included in the surveitlance program the beltline material predicted to be most limiting; and (2) we have conservative methods of analysis, contained in Regulatory Guide 1.99, to determine the radiation characteristics of the limiting beltline weld. For these reasons, c onclude that the integrity of the reactor coolant pressure boundary will be ensured 1
10
durino all nsemal plant coerations, and thus, the exemption to Paragraph fl.B. Appendix H.
10 CFR Part 50, is justified.
f" W
k (b '
c Based on our review, we conclude that it is impractical for the applicant to meet Paragraohs III.B.4 and IV.B of Appendix G and Paragraph II.B of Appendix H.
Imposition of requirements in these paragraphs would result in hardships or unusual difficulties without a compensating increase in the level of quality or safety. The granting of exemption from these paragraphs is authorized by law and will not endanger life or property or the common defense or security and is otherwise in the public interest. Therefore, pursuant to 10 CFR 50.12(a), exemptions from the requirements of these paragraphs are granted.
Appendix G, " Protection Against Non-Ductile Failure,"Section III of the ASME Boiler and Pressure Vessel Code, will be used with fracture toughness test results required by Appendices G and H, 10 CFR Part 50, to calculate the reactor coolant pressure boundary pressure-temperature limits for Farley Unit No. 2.
The fracture toughness tests required by the ASME Code and Appendix G, 10 CFR Part 50, will provide reasonable assurance that adequate safety margins against the possibility of non-ductile behavior or rapidly propagating fracture can be c?tablished for all pressure retaining components of the reactor coolant pressure boundary. The use of Accendix G,Section III of the ASME Code, as a guide in establishing safe operating procedures, and use of the results of the fracture toughness tests performed in accordance with the ASME Code and NRC regulations, will provide adeouste safety margins during operating, testing, main-tenance, and anticipated transient conditions. Compliance with these Code provisions and NRC regulations constitutes an acceptable basis for satisfying the reouirements of General Design Criterion 31, " Fracture Prevention of Reactor Coolant Pressure Boundary."
The material surveillance program, required by Appendix H,10 CFR Part 50, will provide information on material properties and the effects of irradiation on the material properties so that changes in the fracture tougnness of material in Farley Unit No. 2's reactor vessel beltline region caused by neutron radiation can be properly assessed, and adequate safety margins against the possibility of vessel failure can be proviced. Compliance with ASTM E-185-73 and Apoendix H, 10 CFR Part 50, satisfies the requirements of General Design Criterion 31 and General Design Criterion 32, " Inspection of Reactor Coolant Pressure Boundary."
5.2.2 Pressure-Temeerature Limits Appendix G, " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel Material Surveillance Program Requirements," 10 CFR Part 50, describe the conditions that require pressure-temoerature limits for the reactor coolant pressure boundary and provide the general bases for these limits. These appendices specifically require that pressure-temperature limits must provide safety margins for the reactor coolant pressure boundary at least as great as the safety margins recommended in the ASME Boiler and Pressure Vessel Coce,Section III, Acpendix G, " Protection Against Non-Ductile Failure." Appendix G, 10 CFR Part 50, requires additional safety margins whenever the reactor core is critical, except for low-level physics tests.
11
A description of the 18-inch purge valve'cperability program for Farley Unit 1 is given in applicant's letter dated December 10, 1979. A comparison of the Unit 1 purge system operation with Branch Technical Position CSB 6-4 is given in applicant's letter dated February 5, 1979. By letter dated June 30, 1980, applicant has stated that this information is also valid for Unit 2.
We are currently reviewing the Farley Plant purge system. In order to complete our review of the purge system, we require the following information:
(1) A description of the containment purge system design that assures blockage of the purge valves by debris will not occur. The description should include quality and seismic classification of the blockage prevention measures.
(2) A cescription of the means for detecting high radioactivity conditions prior to opening the purge valves.
(3) An estimation of the time period that the purge valves will be open during the year with justification for the duration.
(4) Information regarding the operability of the purge valves.
The resolution of these issues will provide increased assurance that the valves will operate under accident conditions and radioactive releases will be minimized. However, operation of the Farley iuclear Plant is not contingent upon the resolution of these issues because there 7 is reasonable assurance that the valves will perform their accident function from the 50-
[f cegree open position. The resolution of this matter will be reported in a future supplement ll to the Safety Evaluation Report, prior to full power operation.
6.2.5 Containment Leakage Testino Prooram l
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We have reviewed the applicant's containment leak testing program as presented in Section 6.2 of the Final Safety Analysis Report, as amended through Amendment 72, for com-pliance with the containment leakage testing requirements specified in Appendix J to 10 CFR
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Part 50, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."
Compliance with Accendix J provides adequate assurance that containment integrity can be verified througneut the service lifetime and that leakage rates will be periodically checked during service on a timely basis to maintain leakage within the scecified limits. Main-l taining containment leakage within specified limits provices reasonable assurance that, in l
the event of a radioactivity release within the containment, the loss of containment atmos-phere througn leak paths will not be in excess of the limits specified for the site.
The aDolicant has provided a detailed discussion of the containment integrated leak rate (Type A) test procedure and acceptance criteria. All systems penetrating containment will be vented to the containment atmosphere so that the differential pressure expected during an accident will exist across the containment isolation valves for tne Type A test.
The applicant has listed all the containment penetrations and has itemized all the local leak testing that will be performed. Senematic drawings of each piping system penetrating 25
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The applicant will retain the 6-month leek test of the airlocks at a test pressure equal to the calculated peak accident pressure, in accordance with Appendix J.
Additional staff effort on containment leak testing that will lead to a revision of Appendix J is being done in conjunction with the Office of Standards Development. The revised Aopendix J will be applicable to all plants depending on their licensing status and design.
Closed systems outside containment (e.g., the emergency core cooling system and the con-tainment spray system) will become extensions of the containment boundary following a loss of coolant accident. One of the requirements for full power operation (III.O.1.1) in NUREG-0694, "TMI-Related Requirements for New Operating Licenses," is that leakage from such j systems shall be maintained as low as practical and leak tests shall be run periodically.
pt We will report our evaluation of leak testing of these systems in a future supplement to our Safety Evaluation Report prior to full power operation.
6.3 Emergency Core Coolino System 6.3.3 Tests and Insmections In Farley Unit 2 there are four intakes that take water from the containment floor following a loss-of-coolant accident and recirculate it to the safety injection system and the containment spray system. In Supplement 3 of our Safety Evaluation Report, it was concluded that tests of a full scale model of intake number one of the four intakes had led to intake design improvements. These improvements have subsequently been applied to models of intakes 2, 3, and 4 with resultant demonstration of vortex suppression with up to 50 percent intake screen blockage. Pressure loss coefficients were developed for each of the four intakes. Results are reported in Appendix 6C to the Final Safety Analysis Report, as amended through Amendment 72.
Preoperational tests were performed on the plant safety injection system and containment spray system while crawing water from the refueling water storage tank. Loss coefficients for major sections of pump inlet and pumo discharge lines were developed from these tests.
When considered in combination with the intake loss coefficients, it was determined that pump runout flow will be higher than the available net positive suction head (NPSH) would allow without cavitation. Flow restriction orifices were sized and installed to limit the runout flow and thus the NPSH required for each pump. This provides a margin in excess of 2 feet for each low pressure safety injection pumo. The tests and results are discussed in the FSAR as amended through Amendment No. 72.
The model test program demonstrated that without vorttu (uppression equipment in place, severe vortex conditions occurred. The Farley-2 Techrdcal Specifications will include a condition to assure that the plant will not be operated without the intake trash racks, screens, and inner cages properly and completely installed.
The staff is conducting a ' generic program (A-43, " Containment Sumo Performance") that addresses emergency core cooling system hydraulic performance during recirculation as affected by potential break locations and debris from insulation or other sources.
Accitional studies are needed on the use of insulation inside containment and the response 27
i 1
of insulation and other materials to loss-of-coolant accident conditions. Until such time as resolution is achieved, four near-term actions are being required: (1) reevaluate the NPSH available to each safety system pump and verify a margin of 1 foot or more over the required NPSH at limiting runout conditions; (2) establish a housekeeping program to assure that the plant is always restored to "as-licensed" cleanliness prior to power operations; (3) reevaluate the insulation used inside containment to assure that insulation debris would not be expected to block approach paths, trash racks, or screer-in such a manner as to jeopardize intake performance and that debris penetrating the intake screens would not be expected to compromise safety system life or performance or degrade core cooling; and (4) describe the available instruments and controls, and provide procedures permitting the operator to detect problem conditions and to take corrective actions to maintain adequate core cooling even if air entrainment, cavitation, or debris entrainment were to occur.
As Farley Unit 2 has successfully completed tests demonstrating operability with up to 50 percent blockage of the intakes, Action 1 has been acceptably completed. Action 2 will be accomplished by including a technical specification requiring it.
Action 4 is required i
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to be accomplished before full power operation. Our_ evaluation of Action 4 will be reported 4,jt;>(du07 3
in Supplement No. 5 to the SER. Action 3 is required to be accomplighed before startup after the first refueling.
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G 7.0 INSTRUMENTATION AND CONTROLS 7.3 Engineered Safety Features Actuation and Control On Novemoer 7, 1979, Virginia Electric and Power Company reported that following initiation of safety injection at North Anna Unit 1, the use of reset pushbuttons alone resulted in certain ventilation dampers changing position from their emergency mode (closed) to their normal mode (open). On March 13, 1980 the Commission issued IE Sulletin No. 80-06 " Engineered Safety Features (ESF) Reset Controls" to all licensees with operating PWR and BWR facilities (including Alabama Power Company) requiring:
(1) a design review to determine whether or not upon reset of an engineered safety feature actuation signal, all associated safety-related equipment remains in its emergency moce; (2) a schedule for the performance of tests to verify that equipment remains in its emergency mode when actuation signal is removed or manually reset; and (3) a description of corrective actions if equipment is found to not remain in the emergency mode following reset of its actuation signal.
AlaDama Power Company has responded to IE Bulletin 80-06 by its letter dated June 13, 1980,
[ for both Farley Units 1 and 2.
We are currently reviewing a;;plicant's information and will
,gg i report our evaluation in an SER supplement prior to full power coeration. Plant operation 1*
! during completion of this confirmatory review is acceptacle cecause the applicant has se reviewed the potential problem and stated that he will perform confirmatory testing prior to fuel loading. The Office of Inspection and Enforcement will verify completion of the tests l
prior to fuel loading.
7.7 Environmental and Seismic Oualifications l
7.7.1 Environmental Oualification of Pressure Transmitters 1
The corresponding section in Supplement Ne 3 to the Farley Safety Evaluation Report identified four groups of process instrumentation transmitters which were required to be replaced with environmentally requalified transmitters. These four groups were (a) pres-surizer level transmitters, (b) reactor coolant system wice range pressure transmitters, (c) steam generator level narrow range transmitters, and (d) steam generator level wide l
range transmitters. These transmitters were Westinghouse supplied Barton Lot 1 instrument transmitters which were identified by the staff as not fully meeting the post-accident long-term-monitoring envir'nmental qualification recuirements. Requalified transmitters for o
the above aoplications were required to be installed in Unit No. 2 of the Farley Station prior to its initial fuel loading.
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e in documentation will be corrected by providing confirmatory test data or justification on a schedule compatible with completion of the staff's review by February 1, 1980, as required by the Commission's order.
We conclude that operation at low power is acceptable while the confirmatory review is conducted. We will recort our preliminary evaluation of applicant's review in a future suoplement prior to full power operation.
7.9 Loss of Non-Class IE Instrumentation and Control Power System Bus During coeration On November 30, 1979, the Office of Inspection and Enforcement issued IE Bulletin 79-27,
" Loss of Non-Class IE Instrumentation and Control Power System Sus During Operation," to all power reactor facilities with an operating license and to those nearing licensing. This bulletin outlined actions *a be taken to address control system malfunctions and significant loss of information to the control room operator as a potential consequence of the loss of 120 volt alternating current control power to these plant systems. Further, IE Information Notice 80-10, issued on March 6,1980, proviced information relating to the Crystal River Unit 3 event of February 26, 1980, in which a significant loss of information to the coerator resulted from a loss of power to a portion of the plant instrumentation system.
By letter dated July 17, 1980, applicant has provided its response to IE Sulletin 79-27 for Unit 2.
The response indicated that no deficiencies were identified as a result of applicant's review of Farley Unit 2 in accordance with the action items in the Bulletin.
The staff will complete its review and resolution of this matter before authorizing operation g
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- acove five percent power. Plant oceration is acceptable pending comoletion of our confirma-(tory review because the applicant has found no deficiencies.
7.10 Temeerature Effects on Level Measurement On August 13, 1979, the Office of Inspection and Enforcement issued IE Sulletin 79-21, l
" Temperature Effects on Level Peasurements," to all utilities operating pressurized water reactors including Alabama Power Company. This bulletin required licensees to consider the effect of containment temperature under accident conditions on the reference leg water column of steam generator level instruments and the resultant error in indicated water level.
l Acolicant has reviewed its steam generator level instruments in accordance with IE Bulletin 79-21 requirements and recorted the results of its review for Units 1 and 2 in a letter cated November 1, 1979. The applicant has stated that the steam generator narrow range reference legs have been insulated to minimize the error due to snort-term reference l
leg heatup during a high energer line break inside the containment. In addition, the applicant has included an allowance for residual temperature effects in the steam generators l
level trip setpoints and has agreed to modify emergency operating procedures to assure that
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the operator will account for temperature effects on reference legs for both steam generator j
and pressuri:er level instrumentation curing the post-accident monitoring period.
l As a part of our review of the Farley level instrumentation, we assessed the method used for establisning the low-low steam generator level trip setooint. This setpoint is adjusted 31
o General Design Criterson 17. The proposed modi 9fcations will protect the class 1E equipment and systems from a sustained degraded voltage of the offsite power source. The proposed changes to the technical specifications meet the criteria for testing of protection systems and equipment. Staff positions 1, 2 and 3 have been met by the applicant. Staff position 4, the applicant's proposed method for correlating the measured values with the analysis results is acceptable. The implementation of this commitment and the adequacy of the results y
obtained will be verified by the Office of Inspection and Enforcement prior to fuel loading.
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8.3 Onsite Power System.
8.3.1 A-C Power System By letter of June 20, 1978, applicant reported that design conditions existed which could render swing Diesel Generators 1C and 1-2A inoperable when both Farley units are in operation and loss-of-offsite power (LOSP) on both units or LOSP on both units and less-of-coolant accident (LOCA) on one unit occur. By letter of August 15, 1978, the licensee reported that after the two units go into operation, only one emergency source would be dedicated to the river water system pumps. The failure of this emergency source could leave both the units with no river water pumps available (required for safe shutdown). Subsequently, by letter of November 17, 1978 the licensee proposed design changes to eliminale these design deficiencies. Our evaluation of these design changes is provided below.
Loss of Offsite Power and Loss-of-Coolant Accident One unacceptable cesign condition existed which could render swing Diesel Generators 1C and 1-2A inoperable when both Farley units are in operation and LOSP on both units or LOSP on both units and LCCA on one unit occur. This condition is the result of the manual alignment of the power supply to the motor control centers (MCC) which feed the auxiliaries of the suoject diesel generators. Under this postulated condition, Diesel Generator 1C is a source of power for Unit 2 Train A hot shutdown loads and Diesel Generator 1-2A is a source of l
power for Unit 1 LCCA loads. If, at the tire of the accident occurrence, the MCC that feeds
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Diesel Generator 1C auxiliaries is aligned to Unit 1, it will be deenergized and Diesel Generator 1C will lose its auxiliaries. At this time the MCC has to be connected to Unit 2 manually. If this manual action is not accomplished within the specified time Diesel Generator 1C becomes disabled. If Diesel Generator 1C is disabled, no source of power for Unit 2 Train A hot shutdown loads will be available. Thus a single failure on Unit 2 l
Train B can render Unit 2 incapable of being shutdown. Similarly if the MCC that feeds Diesel Generator 1-2 auxifiaries is aligned to Unit 2, at the time of the accident, it will be deenergized and Diesel Generator 1-2A loses its auxiliaries. If Diesel Generator 1-2A is disabled, no source of power for Unit 1 Train A LOCA loads will be available. In this case if we apply the single f ailure criterion to Unit 1 Train B, Unit 2 will meet the shutdown requirements, while Unit 1 cannot meet the LOCA requirements.
The licensee has taken the following corrective actions to correct this deficiency:
1.
Replacement of the manual alignment of the MCC's that feed Diesel Generators 1C and 1-2A auxiliaries by automatic realignment.
We have reviewed the information provided by the applicant and' conclude that the design modification of the onsite power systems for farley Units 1 and 2 will ensure that the plant has adequate riv,er water supply with both units operating in the event of the loss of the cooling water pond dam, loss of offsite power and a single failure. 'de further conclude that this design meets the appitcable requirements of General Design Criterion 17, " Electric Power Systems," and is therefore acceptable. The Office of Inspection and Enforcement will gg ) $d.!
verify that these design changes have been implemented prior to fuel loading.
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o 9.0 AUXILIARY SYSTEMS 9.2 Fuel Storace and Handlino 9.2.2 Scent Fuel Storace The Safety Evaluation Report, dated May 2, 1975, evaluated the storage of fuel assemblies in the Unit 1 spent fuel pool and the Unit 2 spent fuel pool based on a spacing of 21 inches between the center lines of adjacent assemolies. In FSAR Amendment 55, the applicant changed the design bases for the Unit 1 spent fuel pool to a spacing of 13 inches. In Amendment No. 8 to License No. NPF-2, the staff approved the increased storage capacity for Unit 1.
In FSAR Amendments 70 anc 72, the applicant changed the design bases for the Unit 2 spent fuel pool to a spacing of 13 inches.
The design and design bases for th2 modifications to the Unit 2 spent fuel pool are the same as those for the Unit 1 spent fuel pool. We enclude that the Unit 2 design modifications are the same as those aproved for Unit 1, meet the requirements of General Design Criterion 62, " Prevention of Criticality in Fuel Storage and Handling," and are therefore acceptable.
- 9. 3 Coolino Water Systems 9.3.1 Auxiliary Feedwater System l
By letter dated January 10, 1978, the applicant reported a design deficiency in accordance with 10 CFR 50.55(e). The design deficiency was that a single failure of a Class 1E direct current emergency power bus would result in loss of one of the two motor-driven auxiliary feedwater pumos and also would result in loss of speed control for the turbine-driven auxil-lary feedwater pump. Farley plant design bases (FSAR Section 6.5) require that two of the three auxiliary feedweter pumps must start and deliver feedwater to steam generators in the event of a steam or feedwater line rupture.
By letter dated December 20, 1978, the applicant stated that a separate 3-kilovolt ampere uninterruptible power system would be added to the auxiliary feedwater system to sucply power for controls of the steam-driven auxiliary feedwater pump.
We have reviewed the design modifications and conclude that with the corrective action, a single failure would not prevent two of three pumps from starting and operating in the event l,
of a steam or feedwater line break. We further conclude that the modified design of the power supply for the auxiliary feedwater system meets the recuirements of General Design gJ Criterion 44, " Cooling Water," and is therefore acceptacle. The Office of Inspection and Enforcement will verify that these modifications have been made prior to fuel loading.
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"In summary, the Fire Protection Program for the Farley Nuclear Plant with the improvements already made, is adequate for the present time and, with the scheduled modifications, will meet the guidelines contained in Apcendix A to BTP ASB 9.5-1.
The Fire Protection Program as currently designed and installed meets General Design Criterion 3 and is acceptable."
Sy letter dated August 18, 1980, the applicant has stated that all modifications will be l
completed prior to fuel loading except for the following:
1.
Smoke detection systems in the auxiliary building.
2.
Hose stations in the Unit 2 cable tunnels between the diesel generator building and the auxiliary building.
Both of these modifications will be completed by October 30, 1980, or prior to initial criticality whichever is earlier.
We conclude that fuel loading is acceptable because all fire protection equipment required by the acproved plan inside containment will be installed before fuel loading. We will condition the licensee to require all fire protaction equipment to be installed and operable prior to November 1, 1980 in accord with the Commission's May 23, 1980 Memorandum and Order.
We will include Fire Protection System Technical Specifications in the operating license
'for Unit 2 that are the same as those reviewed and approved for Unit 1.
The Office of gM 3 Inspection and Enforcement will verify completion of all modifications at the times stated b
above.
a facilities to accommodate approximately 570 drums will be provided within the auxiliary buildi.ng for eacn unit.
The technical specifications for the operation of Farley Nuclear Plant, Unit No. 2, require a process control program for the solidification and packaging of wastes by the radioactive
, solid waste system. The process control program currently in use for Farley Unit 1 is
! acceptable for interim operation for Unit 2 as well, pending final approval based on the information to be submitted by the applicant in an October 1,1980 submittal. During this l period of interim authorization, the applicant commits to shipping no waste that does not Q
neet applicable conditions incorporated into the license of the burial ground to which the maste is shipped. Applicant has agreed to provide for review and aoproval by October 1, rMf d'
1980, the bases and justificat' ion for the process control program to assure that shipped solid wastes will conform to applicaDie burial ground requirements. We will condition the 1
license to require the submittal of this information. Based on the storage capacity for solid radwaste at Farley, the staff believes that approval of the final process control program can be delayed without affecting low power testing and does not involve a safety question.
I Subject to approval of the process control program, we conclude that the processing and storage facilities for solid radioactive waste materials are aceauate for operation of both units, including anticipated operational occurrences, meet the applicable requirements of General Design Criterion 60, " Control of Releases of Radioactive Materials to the Environ-ment," and therefore are acceptable.
a 13.0 CCNDUCT OF OPERATIONS 13.5 Physical Security Plan The applicant's physical security plan was originally approved by the NRC staff on February 23, 1979. The approved security plan addresses the protection of both Units 1 and 2 against radiological sabotage as required by 10 CFR Part 73.55.
As a result of subsequent revisions, the approved plan consists of a document entitled
" Joseph M. Farley Nuclear Plant Modified Amended Security Plan" dated August 30, 1979.
These security documents are withheld from public disclosure in accordance with Section 2.790(d)(1) of 10 CFR Part 2.
In conjunction with the Unit 2 application, the staff has again reviewed the pnysical security plan against the requirements og 10 CFR Part 73.55 and has determined that the plan is acceptable except as noted below.
By letter dated August 18, 1980, the applicant committed to implementing certain changes in his physical security program. Satisfactory imolementation of those commitments is required
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prior to fuel loading of Unit 2.
The Office of Inspection and Enforcement will verify b
implementation prior to fuel loading.
In addition, we recuire that the applicant fully comoly with the requirement of 10 CFR Part 73.55 wnich states that: All keys, locks, combinations, and related equi,pment used to control access to protected or vital areas shall be controlled to reduce the probability of compromise. Whenever there is evidence that any such key, lock, combination, or related equipment may have been compromised, it shall be changed. Upon termination of employment of any employee, such keys, locks, combinations, and related equipment to which that employee had access shall be changed. This requirement will te made a condition in the license.
The identification of vital areas and measures used to control access to these areas, as described in the plan, may be subject to amendments in the future based on a confirmatory evaluation of Unit 2 to determine those areas where acts of sabotage might cause a release l
of radionuclides in sufficient quantities to result in dose rates equal to or exceeding 10 CFR Part 100 limits.
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17.0 QUALITY ASSURANCE 17.3 Ouality Assurance Program Since the issuance of Supplement No. 3 to the Safety Evaluation Report, the applicant has submitted amendments to its quality assurance program description for the operations phase of the Joseph M. Farley Nuclear Plant. Our review of the changes to the quality assurance program has verified that the criteria of Appendix 3 to 10 CFR Part 50 have been adequately addressed in Section 17.2 of the FSAR as amended threaqh Amendment 72.
The staff has recently developed a revised procedure for conducting the review of the list of safety-related structures, sytems, and components (Q-list) to which the quality assurance program applies. 7his review involves all branches that have responsibility for reviewing the FSAR and significantly enhances the staff's confidence in the acceptability of the p Q-list. Staff re-review of the Q-list using the revised procedure is important for preper j
maintenance of all safety-related equipment over the plant lifetime (40 years); however, its completion is not deemed to be necessary prior to granting authority to load fuel and perform low power tests, because the new equipment is not likely to require maintenance in m*Y the short time internal of operation at low power. 7his re-review is presently under way g
- I and the results will be reported prior to full power operation.
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The NRC Office of Inspection and Enforcement will review the procedures after they are modified and will assure that the apprcpriate modifications as stated above are made prior to fuel load.
In our review, we ascertained that the STA will be informed of the results of evaluation of Licensing Event Reports and corrective action measures that mignt be useful to him in carrying out his emergency advisory role. A newly formed Systems Performance Group (SP Group) will assess operational data, including Licensing Event Reports. The procedures for the SP Group state that "The SP Group shall provide general engineering support for the STA function. When SP Group personnel performing operational assessment conclude that informa-tion exists which may be relative to the function of the STA, such information will be issued to the STAS."
Subject to confirmation by the Office of Inspection and Enforcement that the procedures have y
been modified to specify Shift Supervisor -Inspecting responsibilities and assignments as
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(6 discussed above and that procedures have been modified to specify prompt calling of the STA to the control room upon learning of a significant abnormality or incipient emergency, we I
conclude that the requirements concerning STAS for fuel loading and. low power testing have been met. In accordance with our Dated Requirement (I.A.1.1 in Section 22.5 of this supplement) APCo is required to have STAS who have C0mpleted all training requirements on
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cuty by January 1, 1981.
b I. A.1.2 Shift Suoervisor Administrative Outies Reouirement Review the administrative duties of the shift supervisor and delegate functions that detract fecm or are subordinate to the management responsibility for assuring safe operation of the plant to other personnel not on duty in the control room.
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This requirement shall be met before fuel loading. (See NUREG-0578, Section 2.2.la, Item (4),
and letters of September 27 and November 9,1979.)
I Position 1.
The hignest level of corporate management of each licensee shall issue and periodically
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reissue a management directive that emchasizes the primary management responsibility of the shift supervisor for safe operation of the plant under all conditions on his shift and that clearly estaolishes his command duties.
2.
Plant procecures shall be reviewed to assure that the duties, responsibilities, and authority of the shift supervisor and control roem coerators are properly defined to effect the establishment of a definite line of command and clear delineation of the command decision authority of the shift supervisor in the control room relative to other plant management personnel. Particular emphasis shall be placed on the following:
a.
The responsibility and authority of the shift sucervisor shall be to maintain the broacett ;:erscective of operational conditions affecting the safety of the plant i
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Conclusions We have reviewed the information on shift manning and overtime provided by APCo in the SAR as amended, and the submittals dated August 7, 14, and 22, 1980, and compared the information with the applicable portions of 10 CFR 50.34(b)(7), 10 CFR 50.54(i), (j), (k), (1), (m),
10 CFR Part 55, and the Interim Criteria dated July 31, 1980.
For the long term, the licensee proposes to operate in accordance with the regulations and the Interim Criteria; our evaluation shows that his program complies with the requirements and is acceptable. For the next few months, however, the licensee has too few SR0s licensed on Unit 2 to comply with the Interim Criteria, and has proposed an acceptable plan to compen-sate by using three SR0s on each shift instead of the two required by the Interim' Criteria.
8efore the license to operate Farley Unit 2 is issued, we require:
(1) SRO - For the long term, two per shift are required, licensed on both units. In the interim period, when not enough SR0s licensed on Unit 2 are available, three per shift are required, at least one licensed as SRO on Unit 2, plus a shift foreman - operating, licensed on Unit 1 and trained in the differences between the units.
(2) RO - Three per shift, one licensed on each unit and one licensed on bcth units.
The number of candidates who took the R0 and SRO examinations, and the numcer of trainees, are not sufficient to support a finding at this time that the applicant can in fact fulfill our recuirements. However, the applicant is developing additional contingency plans which, in conjunction with examination results soon to be available, seem to us likely to produce the basis for such a finding. We will report the results of our evaluation of this acditional i
7 information in a supplement to this Safety Evaluation.
t I.A.3.1 Revise Scooe and Criteria for Licensinc Examinations Recuirement All reactor operator license applicants shall take a written examination with a new category dealing with the principles of heat transfer and fluid mechanics, a time limit of nine hours, and a passing grace of 80 percent overall and 70 percent in each category.
All senior reactor operator license applicants shall take the reactor operator examination, an operating test, and a senior reactor operator written examination with a new category dealing with the theory of fluids and thermodynamics, a time limit of seven hours, and a passing grade of 80 percent overall and 70 percent in each category.
These requirements shall be met before fuel loading. (See letter of March 28,1980.)
Discussion and Conclusion We informed the Applicant that the scope and criteria for licensing examinations would be cnanged as stated in the above requirement.
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submitted analyses for small-break accidents in Topical Report WCAP-9600, Westinghou: 4
" Report on Small Break Accidents for Westinghouse NSSS System"; June 1979. Emergency pro-cedure guidelines were then developed from these analyses by the Westinghouse Plant Owners Group. These guidelines were reviewed and approved by the staff in November 1979. The staff review of these analyses and guidelines was performed by the Bulletin and Orders Task Force as is documented in their report on Westinghouse reactors, " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants," NUREG-0611, January 1980 (Appendix IX, section 2.2).
We have reviewed the design features of the Farley Unit 2 plant and we conclude that the review and approval of the small-break LOCA analyses and guidelines apply in total to the Farley, Unit 2 plant.
By letter dated June 30, 1980, the licensee submitted procedures for loss-of-coolant accident (including small breaks), inadequate core cooling, anticipated transients without trip, steam generator tuce rupture, and loss of main feedwater. These procedures are required to be reviewed by the staff and corrected by the applicant prior to full power operation. (See requirement I.C.8 of Part 2 of NUREG-0694. )
Based upon our review to date of the procedures submitted by the licensee, we find they are generally consistent with the guidelines for Westinghouse plants. There are a numoer of minor inconsistencies with specific details of the guidelines and scme instructions to the operator are vague. These matters are being discussed with the licensee. Our detailed comments on the procedures were transmitted to the licensee, and we met with the licensee to discuss procedure revisions required for technical and sequential adequacy. Selected revised emergency procedures will be walked through a simulator and the plant and further changes made, if necessary. The resulting revised emergency procedures will be incorporated into the plant training program and operating procedures.
I j As we stated acove, the selected procedures in their current state are generally consistent
!with the guidelines for Westinghouse plants. These procedures are in place at the plant and
,i are availaole for any emergency. Since the procedures deal primarily with the c'coldown of
{ the reactor and steam cycle and since the decay heat lead at 5% of rated power is minimal, t
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we find the procedures in their current state to be acceptable to succort oceration up to 5%
power for low power testing and training. We will report our evaluation of the completed
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,, rocedures in Supplement 5 to our Safety Evaluation Report, prior to full power operation.
p I.C.2 Shift Relief and Turnover Procedures Recuirement Revise plant procedures for shift relief and turnover to require signed checklists and logs i
to assure that the cperating staff (including auxiliary operators and maintenance personnel) possess adequata knowledge of critical plant parameter status, system status, availability, and alignment.
This reouirement shall be met before fuel loading. (See NUREG-0578, Section 2.2.lc, and letters of September 27 and Novemcer 9,1979.)
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I.C.7 N555 Vendor Review of Procedures Reouirement Obtain nuclear steam supply system (NSSS) vendor review of low-power testing procedures to further verify their acequacy.
This requirement must be met before fuel loading.
Discussion and Conclusions The applicant has submitted the low power physics test procedures to Westinghouse for review and Westinghouse comments have been received at the Farley Plant. We require that comments i
also be provided by Westinghouse for the augmented low power tests (Item I.G.1 of this a
section). The Office of Inspection and Enforcement will verify fulfillment of this requi"e-M ment prior to fuel loading.
I. D.1 Control Room Design Recuirement Perform a preliminary assessment of the control room to identify significant human factors deficiencies and instrumentation problems and establish a schedule approved by the NRC for correcting deficiencies.
This requirement shall be met befora fuel loading.
Discussion and Conclusions As part of the staff actions following the THI-2 accident, the staff requires that all licensees and applicants for operating licenser conduct a detailed control room design review. We expect these reviews to be initiated within the next several months and be completed by the end of 1982. As an interim measure, Alabama Power Company (APCo) was required to perform a preliminary design assessment of the Unit 2 control room to identify significant human factors deficiencies and instrumentation problems. Results of APCo's assessment are provided in a June 10, 1980 letter to the NRC. The NRC staff and its consultant followed up the APCo assessment with a 5-cay onsite control room audit. The review included the assessment of control and display panel layout, annunciator design, labeling of panel components, and the usability and comoleteness of selected emergency procedures. The audit was performed by means of detailed inspection of the control panels, interviews with operators, and observation and videotaping of operators as they walked through selected emergency procedures.
Although our review identified some human factors deficiencies, in general we found that the control room was designed to promote effective and efficient operator actions. The controls and displays are functionally grouped and generally well integrated. The audio alarm system is designed to provide a directional as well as tonal differentiation. The first-out annun-ciators provide information to assist the operators in rapid diagnosis of system conditions.
g'already underway. However, none of these deficiencies offer any significant safety risk to
' fuel loading and low power testing because there are larger thermal margins to the onset of
' exceeding fuel design limits and safety limits during low power operation than during full
~* power operation.
In order to correct these deficiencies, APCo and the staff have agreed that except as noted (j
in Item 10, the following solutions will be implemented prior to escalation beyond five
. udpercent power:
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1.
Control Room Noise. The background noise originated from the air conditioning ducts located in the control room ceiling. APCo will relocate the volume control diffusers and recalance the flow rates throughout the system. This' relocation and rebalancing should reduce the background noise to an acceptable level (less than 65 08(A)).
4 2.
Annunciator Prioritization. APCo will develop a list of annunciators that should receive more operator attention. These annunciators will be prioritized by color.
3.
Annunciator Alarms. APCo will increase the main control board, balance of plant and emergency power board annunciator alarm levels to 6-Sc8(A) acove the ambient noise level. Reduction of the background noise will result in more audible alarm levels.
4 Accidental Actuation. APCo will extend the horizontal portion of the main control board to prevent inadvertent operation of controls.
5.
Color Coding. APCo will review the color utilized for demarcation. A color will be used that provides significance and the tace will be permanently installed. Colors used for system discrimination will be reviewed in order to reduce the number of colors and to increase system discrimination.
6.
Ooerator Aids. All operator aids will be mace permanent.
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7.
Labelino. APCo will review the main control board and will relabel as required for easy identification and for consistency.
8.
Process Computer. APCo will corract the paper feed problem and ensure that both main control room cathode ray tubes are operable. A hood will be added to the cathode ray tube located on the main control board reactor panel to reduce the glare. A cross t'
index of data point accresses will be provided for the operator.
9.
Controllers. All reverse acting controllers will be consistently labeled (open/close).
- 10. Tolerance Rances. APCo will provide for normal, alert and alarm ranges for the sig-nificant main control room meters. As a first priority, meters identified in emergency procedures will be completed before exceeding 5 percent power. Other signficant meters will be completed as information is available, but will be finished before comoletion of the initial refueling outage.
t
.1
6 The low power test program conducted at Sequoyah Unit 1 consisted of nine tests, eight of which involve natural circulation in the reactor coolant system at low power conditions, but at normal, or nearly normal operating pressures and temperatures.
The specific tests proposed are:
1.
Natural circulation test; 2.
Natural circulation with simulated loss of offsite ac power; 3.
Natural circulation with loss of pressurizer heaters; 4
Effect of secondary side isolation on natural circulation; 5..
Natural circulation at reduced pressure; 6.
Cooldown capability of the charging and letdown system; 7.
Simulated loss of all onsite and offsite ac power; 9.
Estaclishment of natural circulatioa from stagnant conditions; and 9.
Forced circulation cooldown (Part A) and boron mixing and cooldown (Part 3).
Each applicant for a full power operating license must perform tests similar to the above tests concucted at Sequoyan except for Test 8 and Test 9b.
Test 8 may be deleted if training for each operator is provided on a simulator that has been updated as necessary using Westinghouse and TVA test data collected during performance of Test 8 at Secuoyah. Test 9b must be performed but may be modified and deferred until completion of the power-ascension program and manufacturer's acceptance test, provided that it is performed immediately following the manufacturer's acceptance test. Other exceptions to the test pregram will be considered if unique, plant-specific differences could cause one or more tests conducted at North Anna and Sequoyah to be unsafe.
Discussion and Conclusions l
Sy letter dated July 17, 1980, the applicant committed to performing a special low power test program which will consist of Tests 1 througn 7 and 9a prior to exceeding five percent of rated power. In addition, the applicant committed to perform Test 9b after completion of the power ascension program and the Westinghouse NSSS acceptance tests. Chapter 14 of Farley Unit 2 FSAR will be modified to describe this test. In lieu of performing Test 8, the aoplicant has committed to providing Farley Nuclear Plant operators training on a simulator that has been modified using test data collected by Westinghouse and TVA at Sequoyah.
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}Itisconcludedthatthelowpowertestprogramdescribedintheapplicant'sletterdated
%{
July 17, 1980, will satisfy Requirement I.G.I.
Prior to conducting these tests, the acplicant 7
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must sucmit a test, description, procedures and safety analysis for review and acoroval by 7a
a
[thestaff. Applicant has agreed to provide this information by Septemcer 1, 1980, based on its scheduled start of the tests on October 1, 1980.
II.B.4 Training for Miticatino Core Damace Recuirement Develop a training program to instruct all operating personnel in the use of installed systems, including systems that are not engineered safety features, and instrumentation to monitor and control accidents in which the core may be severely damaged.
This reouirement shall be met before fuel loading.
Position The staff requires that the applicant cevelop a program to ensure that all operating personnel are trained in the use of installed plant systems to control or mitigate an accident in which the core is severely damagcd. The training program shall inclece the following topics.
A.
Incore Instrumentation 1.
Use of fixed or movable incere detectors to determine extent of core damage and geometry changes.
2.
Use of thermocouples in determining peak temperatures; methods for extended range readings; methods for direct reajings at terminal junctions.
B.
Excere Nuclear Instrumentation (NIS) 1.
Use of NIS for determination of void formation; void location basis for NIS rescense as a function of core temperatures and density changes.
C.
Vital Instrumentation 1.
Instrumentation response in an accident environment; failure sequence (time to failure, method of failure); indication of reifacility (actual vs indicated level).
2.
Alternative methods for measuring flows, pressures, levels, and temperatures.
a.
Determination of pressuri:er level if all ievel transmitters fail.
D.
Determination of letdewn flow with a clogged filter (low flow).
c.
Determination of other Reactor Coolant System parameters if the primary method of measurement has failed.
s cannot be met feasibly by January 1, 1980, a justification should be provided for less than seismic qualification and a schedule should be submitted for upgrade to the required seismic qualification.
5.
The position indication should be qualified for its appropriate environment (any transient or accident which would cause the relief or safety valve to lift). If the environmental qualification program for this position indication will not be completed by January 1,1980, a proposed schedule for completion of the environment qualification program snould be provided.
Discussion and Conclusions Two power-operated relief valves (PCRV) and three safety valves (SV) connected to the top of the pressurizer are employed to provide overpressure protection for the reactor coolant system at Farley Unit 2.
Positive PORV position indication is obtained by stem-mounted limit switches wnich control indicating lights mounted on tne main control board. The limit switches are mounted to sense the fully open and fully closed valve stem position. Limit switches are post-accident environment qualified and seismic excitation qualified switches.
An alarm has been added in the main control room to indicate when any PORV is not fully closed. This alarm is hardwired, i.e., alarm operability is not dependent on the plant computer. The indicators, one set of red and green lights per PORV, are powered from the Class IE de distribution system. The PORVs are air operated employing a solenoid to control instrument air. The PORV solenoid is powered from the same IE bus as the corresponding valve position indication. Control and incication of a PORV will be lost in the event that the bus is lost. The PORV is designed to fail closed on loss of power to the control solenoid. B is configuration is considered acceptable.
Stem-mounted limit switches also are mounted on each safety valve stem to provide open and closed indication. These limit switches will control indicating lignts mounted on the main control board (one red and one green light per SV as provided for eaca PCRV). Indicator power is taken from a Class IE de bus. The switches and associated electrical hardware are post-accident environment qualified and seismic excitation qualified. As for the PORV, an alarm is in the main control room to indicate when any safety valve is not fully closed.
PORV and SV position indicators are single-channel systems. As backun indication, there exist temperature detectors on all relief and safety valve tail pieces which join a common header and piping run to the pressurizer relief tank. Temperature, pressure, and level indication for the pressurizer relief tant are provided on the main control board and alarmed utilizing the plant computer.
Based on the applicant's suomittals cescribing the system and ciscussion with the applicant's l
staff representatives, the position incication system cescribed above is considered acceptacle.
n
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of9 4
- g The Office of Inspection and Enforcement will inspect for comoliance prior to fuel loading.
y 1
m
s p ussion of Existing Instrumentation O
Descriotion of Subcooling Monitor The subcooling meter provides continuous main control board indication of margin-to-saturation conditions. The applicant will install a primary coolant saturation meter prior to fuel load. A summary of information required for the subcooling monitor was provided in Table II.F.2-1.
This system has temperature inputs from resistance temperature detectors (RTDs) (2 hot and 2 cold legs per enannel), in-core thermocouples (8 per channel), and temperature reference for the in-core thermocouples. Pressure inputs are taken from both the reactor coolant system and the pressurizer. A redundant subcooling meter display consists of two analog and digital meters mounted on the main control board. The Farley Unit 2 will use the dedicated digital calculator to calculate margin to saturation using input from the lowest pressurizer pressure and the highest of hot leg RTD temperature measurement or core exit thermocouples. The current main control board readout is pressure saturation. Emergency procedures describe the utilization of the subcooling monitor and appended portions of the steam tables to determine subcooling contiitions in degrees Fahrenheit.
f Alabama Power Company is pursuing with Westinghouse Electric Corporation a minor change to p :; provide main control board readout in degrees Fahrenheit. A description of the modification f required to implement this change will be presented to the NRC prior to its completion.
I I The Of fice of Inspection and Enforcement will verify that the subcooling monitor is installed
'and operational prior to fuel loading.
Descriotion of In-Core Thermoccuole Monitoring A description of the in-core thermocouple measurement system was provided by the applicant in transmittals dated July 17 and July 24, 1980. The primary means of monitoring in-core thermocouple temperature is the core subcooling monitor system. Each channel of the sub-l cooling monitor receives inputs from 8 thermocouples (2 per core quadrant per channel, for a I
total of 16 thermoccupies). A digital readout of any of the 16 single thermocouple l
temperatures may be obtained at the subcooling monitor panel located behind the control boarc. The upper limit of the readout is in excess of 2300*F.
t The second means available for monitoring thermocouple temperature is the in-core thermo-couple readout panel located adjacent to the safeguards section of the main control board (MCS). Any of the 51 in-core thermocouples may be selected by toggle switch positioning and
[
read on an analog readout. The readout range is 100-700 degrees Fahrenheit. If the readout should go off-scale high, thermoccuole temperatures may be measured directly by connecting a "Digimite" or millivolt potentiometer to the thermoccuple inputs at the readout panel.
I A third means available for monitoring thermocouple temperature is the plant process computer. The computer constantly monitors all 51 in-core thermocouple temoerature values.
'when any value exceeds preset alarm limits (700 degrees Fahrenheit hi, 1200 degrees w
D C.3.10 Requirement For Westinghouse-designed reactors, if the anticipatory reactor trip upon turbine trip is modified so that it will be bypassed at power levels less than 50 percent, rather than below 10 percent as in current designs, demonstrate that the probability of a small-break LOCA resulting from a stuck-open PORV is not significantly changed by this modification.
Discussion and Conclusion The licensing basis for Farley Nuclear. Plant includes an anticipatory reactor trip upon turbine trip which has been modified to be bypassed at power levels of 50 percent or less.
The Westinghouse design criterion is that load rejections up to 50 percent should not require a reactor trip if all other functions operate properly. The power mismatch is
~
accommodated by steam dump (40 percent) and automatic control red insertion (10 percent).
Analytical studies performed by Westinghouse for the two Farley units (WCAP-8318, Section 7.3) have shown that primary sytem pressure increases of less than 100 pounds per square inch are predicted for 50 percent step load rejections form rated full pwoer and from 75 percent power. Pressure increases less than 100 pounds per square inch would not open the PORV. From these results, it can be reasonably estimated that a 50 percent load rejection from operation at 50 percent power would produce similar pressure transients; however, analyses of such an incident was not included in the studies.
IfTheapplicanthasindicatedinameetingwiththestaffthattheloadrejectiontransient
- from 50 percent power will be analyzed and that test data exists which would serve to support the analytical predictions. This data will be furnished by September 1, 1980.
Review of this information and data will be required to reach a conclusion on the accept-iability of the anticipatory trip bypass below 50 percent power. We conclude that operation I, up to 5 percent of rated power is acceptable because the anticipatory trip is always
- oypassed for power levels below 10 percent of rated power. We will comolete our review g j prior to operation above 5 percent of rated power.
C.3.11 R-cuirement Demonstrata that the PORV installed in the plant has a failure rate equivalent to or less than the valses for which there is an operating history.
Discussion and Conclusion The applicant has indicated that Farley Unit 2 has PORVs furnished by Westinghouse which are of the same type used in a majority of Westinghouse-designed plants, including Farley Unit 1.
Information furnished on operating Westinghouse-designed plants has shown that for 60 known cases of challenges to PORVs of the type used for the Farley units, no failure to reseat following tne challenges was experienced. Section 3.2, Appendix VIII of NUREG-0611 further indicates that the summary prepared for the B&O Task Force was incomplete and further documentation was recommended and will be furnished in January 1981.
l@7
emergency responsibilities. The emergency plan describes the coordination of the arrange-ments and agreements between the licensee and these agencies. Provisions have been made for an annual review of the emergency plan and for periodic testing, updating, and improving procedures based on training, drills, and exercises. The scope and content of the applicant's emergency plan is substantially equivalent to that recommended in Annex A, " Organization and Content of Emergency Plans for Nuclear Power Plants," to Regulatory Guide 1.101.
Based on review of the applicant's emergency plan, we conclude that it meets the regulatory position statements of Regulatory Guide 1.101.
The Alabama Radiation Emergency Response Plan (ARERP) updated February 16, 1978, was reviewed against the guideline standards of the Nuclear Regulatory Commission's " Guide and Checklist I
for Development and Evaluation of State and Local Government Radiological Emergency Response Plans of Fixed Nuclear Facilities" (NUREG-75/lll), including Supplement No.1 to that publication dated March 15, 1977, which identifies those items essential for NRC's con-currence in a State plan. As a result of this review and in accordance with the provisions of the Federal Register Notice (Volume 40, No. 248, Decemoer 24, 1975), the NRC concurred formally in the ARERP on February 9, 1979.
Revisions to the State of Alabama and the State of Georgia Radiological Emergency Operations Plans are being submitted to FEMA for review. These draft plans were written to meet the essential requirements of NUREG-0654 By letter dated August 28, 1980, FEM.A finds our recommendation for issuing a fuel loading and low power testing license to be reasonable (See Appendix 0 to this supplement).
As a result of the Commission's action plan for Promptly Upgrading Emergency Preparedness at 1
Power Reactors (SECY 79-450), the Emergency Planning Review Team conducted a site visit and technical meeting with the applicant, State, and local officials. In response to our visit, the applicant summitted on December 29, 1979, a prooosed revision (Rev. 2 and 3) to the Farley Nuclear Plant Emergency Plan. This proposed revised plan is currently under staff review and the results of this effort will be reported upon prior to granting a full power licensa; however, preliminary review reflects that tne licensee has designated an interim Emergency Operations Facility, established an interim Technical Support Center, and designated an onsite Operations Support Center (Joseph M. Farley Nuclear Plant Unit 2 l
" Response to the TMI-2 Action Plan," transmitted by applicant's letter dated June 20, 1980) l which we find meets those additional items in the interim upgraded criteria necessary for the issuance of this fuel load license.
l h In summary, based on our review of the comoined applicant, State and local emergency plans, we conclude that the current plan provides an acceptable state of emergency preparedness for l
a fuel load and low power license.
l Deficiencies to ce Correcteo for a Full Power License l
h.uN
- 4 FEMA /
Current efforts by the staff, the Commission and FEWA to upgrade rules and guidance in the dwi 5 7 area of emergency planning should result in definitive and uniform acceptance criteria in l
, the near future. The proposed -evision to Appendix E to 10 CFR Part 50 will include required p.g l#
imclementation schedules for acplicants a..d licensees. In the meantime, the NRR staff has y
informed LWR acplicants and licensees of its new requirements in the emergency planning area 1/@
i s
1 e
4 I. C.1 Short-Ters Accident Analysis and ;'rocedure Revision
=
- Requirement l
Analyze the design basis transients and accidents including single sctive failures and considering accitional equipment failuret and operator errors to identify appropriate and inappropriate operator actions. Based o these analyses, revise, as necessary, emergency procedures and training.
This requirment was intended to be compl eted in early 1980; however, some difficulty in completing this requirement has been experienced. Clarification of the scope and revision of the schedule are being developed and vill be issued by July.1980. It is expected that.
this requirement will be coupled with Task I.C.9., Long-term Upgrading of Procedures. (See
~
NUREG-0578, Sections 2.1.3b and 2.1.9, and letters of September 27 and November 9, 1979.)
Discussion and Conclusion s
?
e
!!. B.1 Reactor Coolant System Vents l
4 Recuirement Install reactor coolant system and reactor vessel head high point vents that are remotely ocerable from the control room.
This requirement shall be met before January 1, 1981. See letters of September 27 'ar.c November 9, 1979.
Discussion and Conclusions BylettersdatedJune20andAugust1,1980,theapplicanthAsdescribedasystemfor venting the reactor vessel head. The system is operable froarthe control room. Applicant has committed to install the system by January 1,1981, or prior to full power operation, whichever is later. The design was scheduled for completion August 8,1980, and all material is scheduled for shipment by September 1,1980.
,p f'
We will review the results of the design orier to full cower coerttien and report the 1
conclusions of our review in a future supplement to the SER. We conclude that app 11 cant has u
taken adequate steps to cate toward meeting this requirement. We will pursue an acceptable
, schedule for installation af ter January 1,1981, on a casts similar to that taken for the il North Anna 2 and Sequoyan facilities.
I II.B.2 Plant St':1dina Recuirement
\\
x Complete modifications to assure adequate access to vital areas and protection of safety equipment following an accident resulting in a degraded core.
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a a
This requirement shall be met cy January 1,1981. (See NUREG-0578, Section 2.1.6b, and letters of Sectammer 27 and November 9,1979.)
Discussion and Conclusions By letters cated October 24, Novemoer 21, Cecemoer 31, 1979, and January 21 and June 20, 1980, the applicant has provided a description of the shielding design review to be done, fApplicanthascormittedtocompleteitsreviewandmakenecessaryshieldingchangesinthe
.Af/
plant orier to full co-er oceration or January 1,1981, whichever is later.
t fWewillreview9eresultsofthedesignandreportourconclusionsinafuturesupplement htotheSER. We conclude that applicant has taken adequate steps to date toward meeting this
.Irequirement.
II.B. 3 Post-Accident Samoling Recuirement Complete cerrective actions needed to provide the capability to promptly ebtain and perform radioisotcoic and cremical analysis of reactor coolant and containment atmosphere samples uncer degraced core c nditions without excessive exposure.
This requirement shall be met ey January 1,1981. (See NUREG-0578, Section 2.1.Sa and letters of Septemov-27 and November 9, 1979.)
Discusjior. and Conci nions By letters dated June 20 and August 1,1980, the applicant has provided a descripticn of eculpment and procedures to be used to sample reactor coolant and containment atmosphere toilewing an accident in which there is severe core damage.
We will review the acolicant's procedures for sampling and report our conclusions in a l
future sucplement to the SER. We conclude that applicant has taken adequate steps to date
' toward meeting this receirement. We further conclude that installation after January 1, 1981 is acceptable because reactor safety is not significantly affected by several months extension at low pcwer operatien.
II.D.1 Relief and Safsty Valve Test Recuirements Recuiremeit j
Ccmtlete tasts to salify the reactor ecolant system relief and safety valves uncer expected I
operating conditions for cesign casis transients and accidents.
This requirement shall Se net by July 1, 1981. (See NUREG-0578, Section 2.1.2 and letters o,f Seotemese 27 and Novemoer 9, 1979.)
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