(i / i, cý% · 3.9 continued 4.9 continued m [f, a., 6. once whwnonehour andatkegoncehperelgthours...
TRANSCRIPT
(i
3.9 coant'd)
S - 3. om a-Wiftier the time that one of the Emergency Diesel 1Jc-r°el 0y�e-- enerator Systems is made or found to be inoperable! .
continued reactor operation is permissible for a period not
EIRA FS, to exceed 14 days ovided that the two incoming power 'Lt51 t.:e J) , ources are avaiable and that the remaining Diesel
Generator System Is operable.At the end of the 14 day
S• h/oode period the reactor shalf be placedina cold condition
CAcT1O4 F"-_ wit hours, unless the affec W diesel generator systemIs mzeOp e soonr m. MO
Ehc , tOmJ 4. When both Emergency Diesel Generator Systems are made or found to be Inoperable restore at least [Sl R
system to able status within two hours place th
reactoriWh col conition wit nt lowing 4
1iK~
Amendment No. 30,--, , Q!.,1 3, 190, 1.2, 232, 253
4.9 (cont'd) L• A2
3. The emergency diesel generator system instrumentation sshaill be ccheckkedd during the monthly generator test./,-
Ve-,',4:e L -la/
"simulated to demonstrate that the pair of diesel generators will start, accelerate, force parallel, and accept
' ' " .. •the emergency loads in the prescribed sequence, .::24-• 5. While the reactor is being operated in accordance with
Specification 3.9.B.3, the availability of the operable .
3 , - Emergency Diesel Generator System shall be " "" • demonstrated once within(II)•)rs by manual starting and ce-wal *he operability of the remaining
Emergency Diesel Generator System need not be
demonstrated if the affected Emergency Diesel Generator System became inoperable due to:
a. Preplanned preventive maintenance or restng.
b. An inoperable support system with no potential common mode failure for the remaining diesel generators, or
c. An independently testable component with no potential common mode failure for the remaining el generators.
217 -P. -44. I I
Revision G
/ I,
4L Cý%
Cti, B.3-13
(
JAFNPP
3.9 Continued 4.9 Continued m
[F, A., 6. Once whWnonehour andatkegoncehperelgthours B .thereafter, while the reactor Is being operated in
3.9.B.t .avao - ieJpower.,, 0 usb by , , I ver correct breaker aignmnerd and by verifying that
the associated off-site electrIcalikne Is en ATed
C c. Disl e a Fuel.I ov . . Thr wllme be a n*n*mmof64ADO gel. of diesel fuel on she for Once a rnonth the quany uaaae -an
Seach operable pdrof del generators. storage tank shall be manually measured and connpaed to the reading of the local level indicators to ensure the proper
S:•t heocl.t
1. From and Mer the time thaO Wuel oil storage tank level 1. Once a month a sample of the diesel fuel In each storage lnsnanie-taion is made or found to be Inoperable for any tank sheal be chocked for qualty as per the folowing: reason continued reactor operation Is permissible a tIn. indeinitely. provided that the level In the affected storage o "- " mx. tank Is manualy mneasured at least once/day. ~&Sdm~00%mx
SAsh 0.01% max. "IDLIMt~a Mo% Point 5540 rin. Viscosity (sW at F Q4o max. ,Sulfur w1% mx. . - -Copper'StripCorrosion No.3 max. / S#35mi.
Arnendiuiert No. A1ý64 218 6 4 '
Revision C
f
on Da.
JAFNPP
3.9 Coatinued
thereafter, wt&~ the reactor is ber opeatedI n accordance with Sp~ecficatilon 3,9,19,1 3,2.1.3 and 3.2.8A., On availatbW of of-site power @hai be aesmed by verifying corre beer alrned and by verfi that
tie Macdted -om Jtis We L
C. Diesel Fuel 7-C. Diesel Fuel
V7jLrC, - Ther wil t
Cft?? 1c. ab tVt/ __
for Once a month the quantity of diesel fuW va•e In each
Areading loe niators to menutfe proper~
1. Once a nmMth a ipale ofthdesel uel in each i e.4tnshallbe docked #ra
Flash TF 12WF mim
Wat 0.05% MW
Ansdmert No. ;)*,. ) 164
? a.-j 14
Revision C
218
-S -? e- C- I i- " C a t:
Spec 4c~o~fr
JAFNPP
4.9 Continued
2. Once per month demonstrat4 rS),ArfL7 operates to transfer fuel from
I
Str for ptxn ps per OI e 0 • IýG olte rail towr OS yst l 1 are < t _'• \ op~e •bloo continued reec •r opertion is pern ssible \ fo a •iod not to exceed 0 days total per •ir of \ diea 91 o1 s rovi led that the r aTlining fueal oil ttrr hsfer \ umps a) demonstrated to operable and daiy
pe.thereafter.•
Am endm ent No. $, o p 4 .205 P a e . 8 o f- v
JAFNPP
3.9 lcont'd) 4.9 lcont'd)
C. Fro and after the a that only one el oil tr sfer pump inDiesel Generator ystem is ound to be o rable, that Diesel nerator k3
System shall conpaidered inop able and continued rlctor operation sha be in "accord with Specificatio .9.B.3 above.
3. Whenever the diesel fuel on site for each operable pair of Diesel Generators decreases to less than 64,000 gallons as a result of operation of the Diesel Generators "to meet Technical Specification requirements." Specification 3.0.C does not apply. 48 hours we allowed to restore fuel oil storage tank uantity to a minimum of 64,000 gallons.
D. AC Power Operability During Cold Shutdown or Refuelina I
Whenever the reactor is in the cold shutdown or refueling mode, a minimum of one off site power source and one Emergency Diesel Generator System, capable of supporting required emergency equipment, shall be operable whenever any work is being done which has the ----. *;* e,• *....n.n, this uimmail *~l•nndrv aenntsinmant is
required, or a core or containment cooling system is required. When this condition is not met, initiate actions to suspend all work that could cause draining of the
vessel, suspend core alterations and handling of irradiated fuel assemblies in the secondary containment, declare required core or containment cooling systems inoperable and immediately initiate actions to restore required AC power sources.
). Not AKolicable
3 8T, D,
Amendment No. 83, 253 220 wCL-I *, 7 ýI I
Revision C
(
JAFNPP -" .&• " 7•,CO).7 3 , 2
3.9 Icontdl 4.9 (contd) A-i q. From a nd ifter te m im that 64 one o.,e oil t.'transfer pump in a Diesel Generator System is
Sfound to be operable, that Diesel Generator .--.- ?.
i contrue reactor operation shall be in accordance with Specification 3.9.8.3 above.• "•
Spair of Diesel Genw Itos deoeses to less thenIL 64,000 gallons as" a result of operation of the Diesel
S WGenerators "to meet Technical Specification s
vesl48 hours cre allowed to restore fuel oil storage tank4 " A1 z]~ 9..qant ity toI in m of or00 linaDllons.icbl 7
tt.)P htvwp- c4 D Nt Aoiiabl
powe wwr sousce ar c
Amenwdmn N 3ent. shall be2
pol f dr V': Int Saco mylin an mt i W h e th sis not .m et, initiate actions - - ' ' " " -Sto suspend all work that could cause draining of the •• / .'-O•
power sources. ft
Amendment No. ",. 253 220 Lo _. JA
Revision C
JAFNPP
4.9 Icont'd)
C. From and after the time that only one fuel oil transfer pump in a Diesel Generator System is
Sfound to be operable, that Diesel Generator / S 3, , I System shall be cosldeed inoperable and
continued reactor operation shall be in accordance with Specification 3.9.8.3 above.
[&C.:-,? VO A3 3. Whenever the diesel fuel on site for each operable
3. o'• 'FOGementa. Soeciation IWC done nqapply.ý -- ,'-.L - 48 hours we allowed to restore fuel oil storage tank
qunttytoam mk~nx gdwuPaon.,- .....
•4• wr Omaliitv D;Min old Sl~tdown oRfuino D. Not_ MJJodll • Whenever the reactor is In the cold shutdown or refnueling•
mode, a mi~nktx of one offsite power source and one Emorgefly Diesel Gen~ertor System, capal" of
operable whenever any work is being done which has the potential for dralfini the vessel, secondary containment is " _
requied, or a core or containment cooling system is required. When this condition is not met, initiate actions to suspend all work that could cause draining of the vessel, suspend core alterations and handling of irradiated fuel assemblius in the secondary containment, declare required core or containment cooling systems inoperable and immediately initiate actions to restore required AC power sources.
Amendment No. 86, 2S3220
Revision C
¼
ITS 3.9 Icont'd)
S :Fe_,ý.ce:o "C1 t;0, ? •
?0__1 -e 10"• o JA
I
nIt P. 'd
'e ac• C" JL)
'.A~)t aWspC sr~i flcd. bclpj
I K,cj1tir cqveratlonIf ilýIiif orel i.z14417)f L eomius V'
Ars shall he 1w1ui3 IRRNi a Yp * Or nw tiaRn r %iera onCo 4
Jb~~t~ibc or a Io retno 7 as tal Cal
.y ..... ~ f 13 o ng irnasuuratflts shall1 be 0 a" n:
a) voltat- of eadi (1-11 tO oens .01 v, b) sp1ecific
cjraviy of eadl cell, C) t. ra tlure of ever fit 111'cll.
S~ ri -s
13 Ofl
e 0""_
u' gravýý, atil overall tage ture is n - surecle w-xi dall there
r ri lot 11 vo tage,
tLor or tjK-- operabl bdtLO. A
r 'a
)e he afyfe ed c. 7he vailability of he ullaffe ted Ler 01 raljl
I'r cc ce [me gency Diesel Ge erator SY em 7b a I I ty of ul t
w t Sp 0 B . 5 t sh I be demonstra ed inn aacc ddaance
tv'atrI4'9 'oth Specificatio 4.9.B.
221Amendment 41
IfRe4 56~ 6-
3. *?. I
r-r
ACo3.03.Lý3-I R /%%,& E~CI -r~*S
k 3.e ý'.--
4.9 (cantd)
rET J.a trle~s
(5Vg ry w tk~w specilI gravty, ltage ancAj
,,ý1i Te-ble 3.94- ZuIMM of the pilot
ULIttery blorj%9un zavailable, __
s,kL'pairs shall he wnda Ainediately i-;wl &uitminwtua4 wctor kioratken
'Al~~ r' "'Onv~ K
its battery dharijer, aaul distribu~tionl system Iis
1,. Pilot 0011 voltage# specifia gravity. aid overall voltage mad toolerature In weasuted 1wa~liatcly wad daily tiorealter for tie cqirable battery.
c. The avai) ability of the unaffected Emergency Diesel Generator System shall be demonstrated In accordance with Specification 4.9.8.5.
V th1 M O M M U M11 tU M flU lo w in g , A t
iurtsiren&-s-hall be ~iai~:j evadt. lyk4 ,Volth(t! of each cull da Ie-kl( Voa-C4.
itof "eadch ml At~Ie"a
c ei(
/\ --jet (tCPOxo U 6/
te.cý4 1 .p~ C~r>)/J4~ i3-
Amenidment 4122
- ~ ~ ~ TT AfJAp~E AC 3Xg,3
0 MG. 11, 9 P 3.
221
k
5~ P e ik L) I 3.,g-7
1 . I Every ~ekthe specific gravity, voltage and WV'p eratule of the pilot cell and
overall battery voltage shal
be ieasured.
Dur3wjjpower qoerationbattery bec"I s unavailable,
repairs shall be nnde i1nidiLately,
and cnntifntyl!d ieacLor qperalion
is |jA!,,P3itblc for a period not
Lu exceed 7 days total/calendar
nonwh provided thai::.
o. %lie othor battery including its bxattery charger, anYd
diStrilAltiOn syste"3 iS operable.
Snr•asurcirents shall be inade: Sa) voltage of each cell to/
neiarest 0.01 v, b) specific
akAC7T6A C
Ilk:
1. Pilot ooll voltage, specific gravity, and overall voltage
ald Wt1eratUre is neasured jrnmiately and daily Lhere
a~ter for the operable battery.
c. The availability of the unaffected
Emergency Diesel Generator System
shall be demonstrated in accordance
Swith Specification 4.9.8.5.
221Al
(DI --- 9- (,o:.n t' d)
JAFNPP
3.9 (cont'd)
[Sit
"p0-rF S"P 3. C 14 3]~ 4.9 (cont'd)
3. Once every 24 months. • each station
btery sh"I be subjected to a service (duty Cycle) .test -1 O F 4. SIR . 3f? 3, ,a
4. Once every 60 months, each battery " be subjected to a fomIUM test for modified
fl , peomenc t his test shad verify thet the battery capacity is at least 80% of the anuMfactuMf's
rating.
IS . 3R 4,7.. "Acceleraomt ne test) abe besting Jon modified pewfognunce teatl shll be conducted an any battery:
c Annually if it has reached 85% of its service ile with cpacity < 100% of manufcture's fating.
dl Once every 24 months if it has reached 85% of its
~ ~1 ~service life with cepwAcy a 1001% Of the , manufacturer's rting. ..
" E~ch hi my ch,,,e, .heU "I-'-'96 ins l•ed w fkiV 6. Each battery chaerger shaM be ? , . ~ W8, and a performance test coMdAIE-- ncerevery
months.
Amendnnwnt NO. 44. 44-. 2322~-$ v--5
T~frF.z63
5 e C- ; ý, -C4.4 Jr is .4.)t I
I
1222
1z 3. 9 -If
49.S e*
JAFNPP
3.9 fcont'd) 4.9 (cont'd,
1 . Every week the specific gravity. voltage a-n•• A.c
Vol IIm be maa: .
2. Every th 24 months the following m shalmn "e S•a. Voltage of each cal to the nosiest of 0.0 I V;
b." Specific gravity of each call;
C. Tempeature of ae very fifth ccll.'.,
4. Once every 24 months, each battery shall be • subjectqd to a service (duty cycle) test.' !
4. Once every 60 months. each battery shall be j
subjected to a performance test for modified performanco test). This test shal verify that the
battery capacity is at least 80% of the manufactuieol' rating.
Amendment No. -,, . 44.406. 444 232 2A2/
Mr-Vial, IU
(
q'LIl Ir%
JAFNPP
4.9 lcont'd)
F. LPCL
'2. Every liv.. months ,i,, oowing ,.,su-,et.shalf be made:
a. Voltage of each call to the nearest of 0.0 1v;
• I)." Specific gravity of each call.
•.C. Tempateltme Of every fifth call.
3. Once every 24 months, each battery shall be subjectqd to a service (duty cycle) teast.'
4. Once every 60 months, each battery shall be subjected to a performance test for modified performance test). This test shelf verily that the battery capacity Is at Wlest 80% of the mariuactwuer't rating.
Amendment No. -W. 44.406. 4; , 232
RA9.'t-2.
I
222@ 7' I,_ 3 .r C
1^F1CA'(C1A60 07.e
USE L WMV 9nd EndRnE awE; Snil
15 R -3. Ia. 4. -31
Ir, .+-5. -a -s A
-- 7.7
3.9 (cant**
F. ipaI W3 MOM mdtPwe am
MS:
13,Ae~.
4.9 fcantdi eA7,A
3, .9.6
.9V o each cal
b. Specifc gravit of each COl. el
'J" , Lit rsctW~d ioea sevko *m esv cvde aset.'
Onca emey 60 amonth. eachu battery d"e be mkojcted fto a perfomance IMs for nmowdi
peo mance tea.tW led.ta dual wify ftha the bater capacit keat Ien" 60% of the
AmndMKW lt No. a*B. 44.406. 44 * 23223A2~2&
SOe c; ý.- 3. 9. 6 (Dý
7--ki-e
cf, C-0,
3 App�iceA��
3.5 F(CC-3.Ci fC(C3
IF��
2. During power operation.
kc rIO A]- becomes unavailable, re end continued reactor of not to exceed 7 days to operable sooney r. SI
1n~able f any reas~ro
Amendment No. -44. 4-00.
jTrs: 3.~
7 Re.To €€e. fr$.:3,8. •L
.r5 ~I~
•jjjsij 1 F. LPCI MOV Indesendent Power Suanlies
j2 cli less both independent 5. Accelerated performance testing for modified
n nvorters and performance test) shall be conducted on any battery:
iated buses MCU ERM ww. i-d_.fla) Annually if capacity drops more than 10% fram its previous performance test for modified
If one independent power supply performance test). pakis sdal be made immediately
pratilon Is permisible far a period bi Annually if capacity is below 90% of
asa the unavailable tro• is made manufactuer's rating.
EflO u lunal-Mrm1 U aim5 plie Is ad oro found be c) Annually if it has reached 85% of its service
0, tfollowk woldaly life with capacity < 100%. of manufactorats
nt war supply Inc ng its gating. ' of m cu gtt y and assozte u is /dl Once every 24 months if it has reached 85%
of its service life with capacity > 100% of the •-• Di~facturor's s ting. 'a 2
Pcif ic gr ty and temper ture and -1 i: I f
go ago asued immed* tely and SR 6. Each battery r and inverter ha
rthe able indepe nt pewe 3,.'mnd a performance est ci- ucted. once every 24 months.
" a.nt poweMr sup be., 7. once/month: open the battery chager, A-C input oclated LPCI MOVW u, and s breakers one at a time and observe performance for
rswitched to its Eiewrnata po1..' proper operation.
232 REVISION D
19441
5/RE
Dt Sourtes - c) a..
(3.9 Ico.Vd)
17 IF]- I I Ti I iupMMn.Ij fK
Accelerated performance testing (or modified performance test) shel be conducted on any battery:
Annual!ifi c city i beow %o
Amendment No. -441.409, 232
cl Annually if it has reached 865% of its service J life with capacity < 100% of manufactweuis rating. " T
"-"t Once every 24 months if it has roached 85% 8. 4 . • of its service fife with capacity k 100% of the
Ivj rtVI"' manufactworxs rating.
27. Oncelmo : open the battery acger A-C in tI Weakera t ., time.rid serve perfr|©m a ef
ration.,
222b (QV4~l
t
r- -
TGe oWier inependn Pow., -W upyInckhqo -hq. lowest. -aey now aseociaed bus
a. cclmste pevs puleasence tesft lto modified
hI Anuauy, N capacity dros moelo w 10% loa
manulata~erseating ci Annuually Itl capctis bealhe 90% of t evc
Me with capacity < 100% 09 markulactimmrs rating.
di Once every 24 maonths lit t has reached 85% a £. "A- U .16 .. Za. MW ^ IM a"*h
manuascturers rating.
h. Pilot ce v~lag. speciic paVIDA end SORW stuos en ~as bettey w~lta e nwamewed bumedleel and 6. Each bettery chegern rvwWWb iu~
0 W Nel thersafter fai One @mewal hIndpendent PewF onspected weekly and a perlormence test conducted. sup*l beatey. 0F- Weve" 24 am~onh.
c. The huperable Indep enieea power supply dolal be 7. oncehnoth: open the battery cheger A-C input
lealeted hmker . asoinaalted Lloa MOV bus. end th" beakers one at A time and observe perfamormnc for
bua waill be omanumy swilched to its skoeternt power plope operation.
Ameandmeant No. -44..4"0 232 -? o -()922b*
C, f
JAFNPP 5re 3.3. 8 2
3.9 (cont'd)
F3. From and after the time both power supplies are made or found inoperable the reactor shall be brought to cold condition within 24 hours. r T _7
REACTOR PR ASSEMBLIES
(Acv.7I J A
ASSEMBLIES
Two RPS electrical protection assemblies for each inservice The RPS electrical protection assemblies instrumentation sl
RPS MG set and inservice alternate source shall be a erable be determined operable by: "- ,
•=•L-,., I• L1. Performing a channel functional testI ach time t hme te plar
1. With one RPS electrical protection assembly for an A ., 3. 'Z2. in dco s utdown or a periou of more than 2
inservice RPS MG set or an inservice alternate power unless performed in the previous isup 'ply inoperableeor th inonperable.canlt.
within 72 hours or remove the associated 2. Once per 24 months, demonstrating the operability of
RPS MG set or alternate power supply from service. / 1i 3,2,. ZJ.tover-voltage, under-voltage and under-frequency S(U protective instrumentation by performance of a chani
2. With two RPS electrical protection assemblies for an C 7, ) i onc ang stmu a e tomatic actu ion of inservice RP Mseoraievice alternate power e•' arrs• •2in
supply inoperable,( estore at least one to operable status breakersjand verifying the following A/I.M within I or remove the associated RPSI Mset RPS MG SET OURCE
or alternate power supply from service. _._ MG SETSOURCE
SOVER-VOLTAGE <1 32V
! <4 second Time Delay
UNDER-VOLTAGE
vs.r*". 4 co'Aar add 401e
Amendment No. 1, 76. 18.. 23 245
UNDER-FREQUENCY
Ž1 12.5V for "A"Channel Ž1 13.9V for "B" Channel !4 second Time Delay
657Hz <4 second Time Delay
(continued on page 222d)
222c
2_
6•;)lle- Trlý:f "%
t5k 5.112. 7-1
I
?'Q 7 e-
'JcriatoJ
Amendment No. 41, 76. 18., 233 245 (continued on page 222d)222c
P'tj e 17
REVISION D
a atlA 3 Jl&rJAFNPP - . I
From and after the time th nower u are made or found inoperable the reactor shall be brought to cold condition g, within 4kours, 5('
G. REACTOR PROTECTION SYSTEM ELECTRICAL PROTECTION G. REACTOR PROTECTION SYSTEM ELECTRICAL PROTECTION ASSEMBLIES ASSEMBLIES
Two RPS electrical protection assemblies for each inservice The RPS electrical protection assemblies instrumentation shall RPS MG set and inservice alternate source shall be operable be determined operable by: except as specified below:
1. Performing a channel functional test each time the plant is 1. With one RPS electrical protection assembly for an in cold shutdown for a period of more than 24 hours,
inservice RPS MG set or an inservice alternate power unless performed in the previous 6 months. supply inoperable, restore the inoperable channel to operable status within 72 hours or remove the associated 2. Once per 24 months, demonstrating the operability of RPS MG set or alternate power supply from service, over-voltage, under-voltage and under-frequency
protective instrumentation by performance of a channel 2. With two RPS electrical protection assemblies for an calibration including simulated automatic actuation of the
inservice RPS MG set or an inservice alternate power protective relays, tripping logic and output circuit supply inoperable, restore at least one to operable status breakers and verifying the following setpoints: within 30 minutes or remove the associated RPS MG set or alternate power supply from service. RPS MG SET SOURCE
OVER-VOLTAGE !1 32V
,se iTS: 3..8t 4 second Time Delay
UNDER-VOLTAGE >_1 12.5V for "A"Channel _1 13.9V for "B" Channel s4 second Time Delay
UNDER-FREQUENCY >_57Hz _<4 second Time Delay
I
JAFNPPpec KCA(dO%38
3. From and a er the time both power uppiies found ino rable the reactor shall ibrought within 2 ohours.
G. REACTOR PROTECTION SYSTEM ELECTRICAASSEMBLIES
Two RPS electrical protection assemblies for RPS MG set and inservice alternate source sh•except as specified below:
1. With one RPS electrical protection asser inservice RPS MG set or an inservice alte supply inoperable, restore the inoperable operable status within 72 hours or remo• RPS MG set or alternate power supply fr
2. With two RPS electrical protection assen inservice RPS MG set or an inservice alte supply inoperable, restore at least one to within 30 minutes or remove the associa or alternate power supply from service.
see 11r5 ' 33.
Amendment No. 41, 76. 180. 233 245
rema orM toc2 ondfition RAT
L PROTECTION G. REACTOR PROTECTION SYSTEM ELECTRICAL PROTECTION ASSEMBLIES
each inservice The RPS electrical protection assemblies instrumentation shall all be operable be determined operable by:
1. Performing a channel functional test each time the plant is nbly for an in cold shutdown for a period of more than 24 hours, ernate power unless performed in the previous 6 months. channel to
ve the associated 2. Once per 24 months, demonstrating the operability of om service. -over-voltage, under-voltage and under-frequency
protective instrumentation by performance of a channel nblies for an calibration including simulated -automatic actuation of the
mrnate power protective relays, tripping logic and output circuit operable status breakers and verifying the following setpoints: ted RPS MG set
RPS MG SET SOURCE
"OVER-VOLTAGE <132V •<4 second Time Delay
UNDER-VOLTAGE ;Ž112.5V for "A"Channel >1 13.9V for "B" Channel !54 second Time Delay
UNDER-FREQUENCY >_57Hz s4 second Time Delay
(continued on page 222dl
222c
5,0-r 'r5
,
I.
'reC. dv ¾ 3 .Z
4.9. G (conl'd) l
AI ATEIINATO SOURCE
22 2d
Amuendm~ent tio. 76I.
raa Ir e
ir
3.~~ ~ ~ ~ 18 410 MSAaiTIa
Applies to fuel andling and Core reactivity limita Applies to the periodic tostin of those interlocks ti°u'" and instruents used durin refueling and core alterations.
QowLti iL To assure t at core reactivity Is within the spaIblity of t control rods and to prevent crit allty To verify the operabil ty of instrumentation and during rtof•ing. interlocks used In req uire g and core alterations.
I I
1reactoroPrior 0 any fue handling, _i thAiheh d off *I p1osition ur core eItra! ona te *-th" refueling #p& inelock e '. reluifiiinterlocks shall be operab j'qA ,. 3 1 bhall be functionally tested. They shall also P-7-Or"y pecbe tested at weekly Intervals thereafter until
2.fruel shell not be loaded into the reactor cor nloess all control rods are fully ins2rted) 2. Whenever t e reactor mode switch Is In the ~zceut ~Refuel position and refueling Interlocks are bypassed, one licensed operator and a member of
C 3,10,6 the reactor analyst department shall verify
the corresponding control rod Is withdraw
Amendment No. 115 22711
I
_I -•L 93 ?2
JAFHPP
/4.10 5UR-KILI:AM"_QUIREMNTS
S_ a n y fU e . ... . _ i h h e e of-h.a. ( - ) shall be functionally tested. They s-las
ýbe tested at weekly intervals thereafter until noIon erreaired Ia riV;oy1 n i i ýpair Lwork associated with the Interlocks.... - -'-
2. • henever the reactor moae Owlcc I s a-in the
Refuel position and refueling interlocks are bypassed, one licensed operator and a member of the reactor analyst department shall verify
\that the control cell contains no fuel before \the corresponding control rod is withdrawn. a.
17 5s.3. /
227
/ý /
c C-JAFHPP
(5]3.10 L1MlLiNU
3 . 10
I AL
Applies to Lug
tions.
To assure the bility of the , during refuslji
Specifications
SThe Reactor Mode Switch shell be locked in the 1. Prior to any fuel handling, with the head off Refuel position during core alterations and the the reactor vessel, the refueling interlocks refueling Interlocks shall be operable except shall be functionally tested. They shall also as permitted by Specifications 3.10.A.5, he tested at weekly intervals thereafter until 310.A.6, 3.10.A.7 and 3.10.D. no longer required and following any repair work associated wi_.th t
4 all control rods are fully inserted . Whenever the reactor modi swich Is c I (~ze~ts prmitedby pecfictio i.U~a1. efuel position and refueling Interlock: are
bypassed, one licensed operator and a member of e JT.5•: the reactor analyst department shall verify , that the control cell contains no fuel before the corresponding control rod Is withdrawn.
L; =,-
Amendment No. ý4Airt 1AJA
I
227
/ \ (
I =S11 a -
5>'tAGA~., 3/QC4__ __jx NPP .. . .. .. .
4.10 1L1Ai"GCO0DITIjM 4-1
3.10 CORE ALTERATONS 4. 10 ,._ALTEIATLQ S
Appl ica biLy / Applies to fuel handling and or@ reactivity limita- Applies to the per odic testing o tions.
and instruments Red during ri Mf l alterations.
To assure that core reac vity Is within the capa- M e /
bility of the control rode and to prevent criticality To verity th operability of Ii during refueling, interlocks use In refueling and coi
A. Refusei- |Ttrlc
1. The Reactor Mode Switch shall be locked in the 1. Prior to any u handling. Refuel position during core alterations and the the reactor vessel, the rel refueling interlocks shall be operable except shall be functionally tested, as permitted by Specification. 3.10 I be tested at weekly interval* 3.10.A.6. 3.10.A.7 and 3.10.D. S--_--__,__ no longer required and fol.
F7.-uel .hall not be loaded into the reactor core work asociated with h 1 -te, "%•unless all control rn-a ful..Ily Inaert- 2. Whenever the reactor mode a ce spermutta e ec caton3*l0A - - ( Refuel position and refuelil
3 • Ia ' •that the control cell contal
ILCO "j,
f those inter. 'fueling and
with the head off fueling interlocks . They shall also a thereafter until lowing any repair
switch Is in the ng interlocks are ir au a member f
shall verify ns no fuel ý5 |®
unua control riod js wjth ajwijjn
M/Amendment No. ;4g #of, 115
227
74 .7 )
I
'L O ,0,
JAFNPP 3.10 (cost'd)
3 . T h e f u I g a p e h o i s t l o d s w i t c h sa " b e n oC A at I* than or equal a 650 lbs.
4. It te u rame-mounteauxluatry hoi the mono ra mutdaslr hoist,. r cel o ih onto used to b tdrawn h beIn reactorr•Iod he
c. t load e li tceor the hoistsoiat ed i all 0b net at less &an or0 eqa o/O~
cotrl el ontinn f laeopable. Refuelig inter o uks asoit edot
5.n eci r of control rods may be wbtya•daw or
an. The reactor mode switch 'in locked In the "Refuel" position) and
I S b. The fuel assemblies situated In the control celleha-e cell of the control rod to be withdrawn have . .
been removed; and
c. Refueling Interlocks associated with all control Celle containing fuel are operable. Refueling Interlocks associated with a specific. control rod may be bypassed after the fuel assembliesIs the control cell have be6a removed) and
d. Fuel on-loading operations shall be
suspended until Specification 3.10.A.2 is satisfied.
Amendment No.115
228
5-ý Pe ; (-..
3LIO
Cc)o3,1.lJ
S. 7). (. MA1, ý z ( JAFNPP
3.10 (coat Id) UeUPA)
3. The fuel grapple hoist load switch shall be set at less than or equal to 650 lbs.
4. If the frame-mounted auxiliary hoist, the monorail-mounted auxiliary hoist, or the service C platform hoist Is to be used for handling fuel with the head off the reactor vessel, the hoist load switch on the hoist to be used shall be set at less than or equal to 400 lbs.
S. An y numb er of control cod a may b. withdrawn or
,_removed from the rea___._.cctor core provided: -
" toeg e 1" posit[onl ad• ... 'a
/•b. The fuel assemblies situated In the control
cell of the control rod to be withdrawn have been removed; and
. KTue ng interlocks soclated wi all
!.contfol.coils contaona fuel are Refueling interlocks associated with a specific control rod may be bypassed after he fuel assemblies in the control cell have an removedo and
T-_ .3'
I c(O /.4
Amendment No. 1 1 5
228
"3,o0, -4
,
I L.ro 34.
.j Ac IO'A A
11, ) t t e) f
$6p-ect'I ( 0- 6d b,
3.10 (contd)
A spiral off-load may be conducted provided: 1-1
b. Refueling interlocks are bypassed only for
those control cells which contain no fuel; and
rt~o IN:
LL03.Wd1 .
229
7ý ('-2
6.
c�y7
C. Fuel Is removed from a control cell before Its control rod Is withdrawn.
7. A spiral onload may be conducted provided:
a. Refueling Interlocks may be bypassed only for those control cells which contain no
- t fuel; and 6\ , \') . --
O0. b. Ub spiral onload may c once at e t or Aore around a "dlnking type tector" or, ero id on. of/ the sourc range monitors (Placement/of the "d.ii ng tyl debtcto in the ore center does t violate the intent f the spira onloa ng patter . Fuel ma be loaded into his bundl location wlen the dunki gdot has oen removed.) and,"
Bc. efore loading fuel Into an empty control cell, its control rod Is fully Inserted, and
ýthe refueling interlocks for that control rod are operablej and
n(d. Refuel Interlocks are operable f an
bLi Amendment No. Ep'15
a d�
&
-T) )61, 7
C6' -- .3, S.c
1. ,I~ / /' , X3 .gg cor altsjtftlOnim two URN'S shall be~
operable, One In the core quadrant where fuel or -5,s. control rods &to being moved and one In an
-33.J.2.. adJacOet quadrant. for am SRN to be considered ,operable the following conditions shell be
I, (Tho g shl
l ~ ~0( maiz- uking ontype detectors during Initial fuel •(t •,12-t loading aind imajor core altelrations In place o
' (ormal detectors Is permlisable as long as the detector Is connected into normal SRN circuit).
The SRN shall have a minimum of 3 counta/sec SvWith all rods fully Inserted In the core da
3. Prior to spi &I unloading, the RN. shall ha an lantla count rate of 3 CPS. D ing spiral u onding, the coum rate of the RM's mnay dro below 3 CPS. - "
ru
Amendent No. >f, 0< 115230
'-
fr�a4�,ai) F
t
JAFMPP
1-5 3.10 (cost'd)
(• During Spiral reload, SUN operability will be G eified by using a portable external source •-• |, overY 12 hours until enough fuel is loaded to
"mintsin 3 CPS. Alternatively, a maximum of 9 1 3 ). four fuel assemblies will be loaded in 3- different calls containing control blades
around each SIN to obtain the required 3 CPS. Umtil those assemblies have been loaded In a given quadrant, It Is not necessary for the SRtH
A th t quadrant to Imdicate the inaimun count rate of 33 CPS. The loading of fuel near the SRM's doom not violate the Intent of the spiral
reloadig pattern.
Amendment No. 115 3( u 230a
1-11
,
r Cc NA C4
S Pee tA aý,V
3•, Spenet ruel ntor-,_ Pool Water Level
WiJAFNPP
P'7' )& SpanL~ul Storage Pool NatirKJjysj
irI
• mimimuf
SD. Control ld and Control a od Drive kaintenan"-
1. Two control rod* may be withdrawn from the reactor core to perform maintenance provideda
a. The Reactor Mode Switch is locked In the Refuel position and all refueling interlocks Are operable except for those necessary to perform the demonstration and maintenance described In Specification 4.10.D.1.
b. Control rods imemdiately face and diagonally Adjacnt~m to th-~~ A L -- & .. lADLWwiur
• ~ ~~~ re avbv gou au uwanurawn are fully inserted, electrically disarmed and sufficient margin to criticality demonstrated.
c. Control rods to be withdrawn are separated by three or more cells in any direction. (T24e specification does not apply to the control rods used to perform the demonstration required by Specificationj
3.10.D~l-bs
i beWhnever Irradiated fuel Is heor ant ,••,-. fuel storag a oo .- the pool water level shall be recorded •
D.Oan 0 o gocQ .Hd a
1 W" hen. two control rods are withdrawn from the
\ reactor core for maintenance, the following \surveillance shall be per formedst
\a. If the reactor vessel head Is removed, specification 4.10.A.1 shall be satisfied.
bh. Demonstrate that the reactor core can bet~
Smaintained subcritical with a margin of I ~0..38 percent A k at any time during the
Smaintenance with the analytically \determined strongest worth operable \ o~on trol rod fully withdrawn. Trhis merge
S\ ~shall be demonstrated after Specification S• IO1.D.1 has been satisfied.
r.wI.-uw- U
A/St
Amendment No.)h< 115
231 r7t ( 04Lr I
)LE o
(I.
7-Z
rA lict
b. Cornol rd mmea
S q the cont are fully inserted, O! S a uftIcient m demonstrated.
. Control d to be by tbre# or more (This srpacification/
4.10 (cont'd)
C. •-anLM.-'U Rm•i -_o d Dt e aint ,
Whenever Irradiated fuel Is stored 7 in the spent . fuel storage pool, the pool water level shall be recorded daily.
1. When 01c7ontrol rods are withdrawn from the reactor core for maintenance, thea.1M 1ig~
eAce4 ~oitch Is locked in the /V Nb. all refueling ine , the oactor ,,e91 head re.ueo,- ' for those necessary to • f.e ispeci lVcatio..4.10O•X.i shall
tratIoR and maintenance , , _ • --.~atiWn b.ODl b. Diemonstrate that the reactor core can be
__ _ ' I '- -maintained subcritical with a margin of t e alce n diagoiA-M I 0.38 percent A k at any time during the rol rods to be withdrawn maintenance with the analytically
electrically diarmed. determined stronqest worth operable ,argin to criticaii ty control rod fully withdrawn. This margin
ow This _, .,g
Sdoes not apply/t the AA
od to perf..rm tho/[ ] ,red b. Soc~,. / ti/ J- h/._ / •7
..2,3
toIj
/(
Pe e -2A
o2k( o•/..
4.10 cont'd)
2. When mio a than two control rod are with raw from 18 reactor core for intenanc , th: toll •ng surveil ance shall b performs s
"a. Specificati s 4.l0.A.1 and 4.10. .2 shal satiagt ..
2. 3. Maz tban two control red may be wit rawn\ f rom the sector core to perform main nance I provideds
a. Spec ication 3.10.A. Is satisfied-i I
Amendment mo. F 115
232
REVISION D
1P
JAPNPP
.(07
3.1 LIIIGCONDITIONS FOROEAI 4.1 1 SURVEILLANCREULEN
Aoolicabilitv: .. Aoolicabilit"
• Applies to the operating stat of the main control and Applies t a1• surveillance requirements for tt relay rooms, and battery ro . ventilation and cooling, control d relay room, battery roomn ventilatii Applies to emergency ser ce water system and.intake emer ncy service water and intake deicing h deicing heaters. ater. / / ' To verify the operability ot availability under c
To assure the av lability of the main control and relay for which these capabilities avan essential re
room, and bat y room ventilation systems, to assure f plant abnormalities.
the availabili of the emergency service water system / Sand intake •licing heaters, under the conditions for
whidh the epability is an essential response to plant • • •-...... abnorm lties.,- -'
t Main Control Room Ventilation2 A./• i Cnntrnl*'nnm- V ntilatf e-,•( •~~ -•1 Each of the. control room emergenc
_1. The reactor shall not hive a coolant• --. • air SUE&l fans and damprsr shall b , I I / I-Jtemperature rn tr thE And 11 "BA - -.hlmay opr -- it - --. r"a'= h; o-ery; ... m nh....,•-'". . .. ýfor l CC4 nQ, bel_•.J• hi|ln M nle-ss bGotfh of thef na•dcontrol room~ vetlton,,,o airsupply fan and reshThe reshairfiltr trins hallbeo,
/ l~air filter trains are available for normal every 6 months as follows: -" Sgene~~ration except that one eme qon.w " SI • _ I". -•- I ,,a. P'ressure drop test across eac,
I•.,•- the filter system.
-T -S S
Amendment No. 231 • --
237
i
Sf ký C 1'4-CCI 6 -1lfý
Aty' (2iAL-0-QwOIB!2Qm.&enLLI atlionc- :F SV
IF E a ;ch o
EL eo
t tZ dK_
aod PD Oe Vf) for MI
Z'*--ýýýýýýý' litFsk .3. 7 ;3,17
P,5t 0Amendment No. 231
I ý ý
�/
JAFNPP
Applies to the operaft~ status of the main control and Apiplies to the surveillance joqtkomamts far the main
arely tooms, and battery room ventilation and cookig control and relay room. battery oom ventilation systems. Applies to emergency srvico wat system Ondinblsae Gmegencys SWWO wter and Intake deicing heaters. deicrig homers.D
a~kwa o. . To Vrify the operability oa rvailability under condilions
To saweuv the &vlablty of the maincontrol nd relay forw w fit hese t aps bitUes me an essentialiresponse to oom. and bettery room ventilation systems, to sawep plant bnormaltioes.t
the evailabolfto of thee meOrgecy 1vlcw water ysteme and intaked hecing eters,w uider the conditions forw whldh tie B iity se •n ssentialresponse to piant. B-5 3.7 bnormalities.A S/ (~~A. Mein Cotol,, noon Ven-t . A. Nutn C-wUrel Roan Ventilation / \.
S. I • Each of thle control room emergency ventiletion . The reactor sheal not have a coolant fans and dtmnpag " be tested for
tempaeatwe groeatr then 212F a nd fuel mey Zo ity *my 3 manyn. not be handled unless both of the controlroome emegency ventilation ai suply f ans and fresh tThe rath ai filter trains shall be tested once \ir filter trains oe evaile o rn normale ver• tn• s f ollows: operation except that one emergency .- .a . re -we l e % J ( •'. a. B cl] a. . Fteusure drop test moosslrh IItF'
Onefilter s(
3.7...IS( est 41yoa- S8 //. lr
Amndet No. 231
I " '5)Cc'jkC041OW
,I
,-4
3.11 (cont'd)
(Enill~aon air supply fen andlor f svi.e .for 14 days.
2. 'ýThe main control
Xentilation air suj
'.4
JAFNPP
3,3�4.11 Icont'dl
ilter may be outofb. Di-octyl test f particulate • ~efficiency greater then 99% for particulate Wrester
S~than 0.3 micron size.
"•C. Freon-1 12 test for charcoal filter bypass as a
S~measure of filter 'efficiency of at least 99.5% for
halogen removal.
on monitor shall be 2. At least once per 24 months or 11) after any structural
gom emergency maintenance on the HEPA filteror charcoal adsorber
tar trains are reo ied t housings, or 12) following painting, fire, or chemical
tion of the control room release, that could adversely affect the ability of the Jte =• • •charcoal to perflom its Intended function, in any
ventilation zone communicating with the system, verify:
11) Within 31 days after removal, that a laboratory test
of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of
Regulatory Guide 1.52, Revision 2, March 1978,
shows the methyl Iodide penetration to be less than or equal to 5 percent when tested in accordance with ASTM D3803-1989 at a temperature of 30
degrees C 186 degrees F1, and a relative humidity ol at least 95 percent.
12) Within 31 days of completing 720 hours of charcoal adsorber operation, that a laboratory test of a sample of the charcoal adsorber, when obtained in
'3 accordance with Regulatory Position C.8.b of Regulatory Guide 1.52, Revision 2, March 1978,
shows the methyl iodide penetration to be less than or equal to 5 percent when tested in accordance with ASTM D3803-1989 at a temperature of 30 degrees C (86 degrees F], and a relative humidity of at least 95 percent.
' -J
P^5,z Z -Amendernort No. 411. 420, 02. 3. 269
238REVISION F
p
L
(
51-~;F2&rCdrfVj 3, 37,?ý
JAFNPP
4.11 (con
2. The main control room air radiation monitor shall be operable whenever the control room emergency ventilation air supply fans and filter trains are required to
Sbe operable by 3.11.A. 1 or filtration of the control room ventilation intake air must be initiated.
t'd)
b. Di-octylphtalate (DOP) test for particulate filter efficiency greater than 99% for particulate greater
than 0.3 micron size. e
c. Freon- 112 test for charcoal filter bypass as a measure of filter efficiency of at least 99.5% for halogen removal.
At least once per 24 months or (1) after any structural maintenance on the HEPA filteror charcoal adsorber housings, or (2) following painting, fire, or chemical release, that could adversely affect the ability of the charcoal to perform its intended function, in any ventilation zone communicating with the system, verify:
(1)
(2)
Within 31 days after removal, that a laboratory test of a sample of the charcoal adsorber, when obtaine in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration to be less than or equal to 5 percent when tested in accordance with ASTM D3803-1989 at a temperature of 30 degrees C [86 degrees Fl, and a relative humidity of at least 95 percent.
Within 31 days of completing 720 hours of charcoal adsorber operation, that a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration to be less than or equal to 5 percent when tested in accordance with ASTM D3803-1989 at a temperature of 30 degrees C 186 degrees F], and a relative humidity of at least 95 percent.
N
C
Amendment No. 114, 129, 102, 233, 269238
(6e2-,P V
l
3.11 1cont'd| S4eQ- fl "7 ,
YA?. iventi9ation air supply fan andlor filter may be out oT_ \.s•eervice for 14diay.
MU5*O. C- r. de &r i4~j% r*-Y 6Lr&&e 1. rz ~-e A 2,.L ~ v 760 0 ~o/10o vxm 'O
L{ (1) 3ithin lays af te emoval hat a laboratory test o a sample of the Charcoal adsorber0 when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2. March 1978, shows the methyl iodide penetration to be less than or equal to 5 percent when tested in accordance with ASTM D3803-1989 at a temperature of 30 degrees C 186 degrees Fl, and a relative humidity of at least 95 percent.
12) •ithlin 3da s orn c~petin '20 hours of charcoal
adsorber operation, t at a laboratory test of a sample of the charcoal adsorber, when obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration to be less than
NL4 f or equal to 5 percent when tested in accordance with ASTM D3803-1989 at a temperature of 30 degrees C 186 degrees Fl, and a relative humidity of at least 95 percent.
Amendment No. 110. 120. 102. 222. 269 1
238REVISION H
57ee c f4 MJd' hj 0 ',
JAFNPP
1 1 ( c o n t ' d ) "LP ' ) / 'Z4 .1 1 (
3. The control room emergency ventilation system shall not be out of service for a period exceeding 3 days during normal reactor operation or refueling operations. In the event that the system is not returned to service within 3 days, the reactor shall be in cold shutdown within 24 hours and any handling of irradiated fuel, core alterations, and operations with a potential for draining the reactor Svessel shall be suspended as soon as prati able .
3. Operability of the main control room air intake radiation monitor shall be tested once/3 months.
Temperature transmitters and differential pressure switches shall be calibrated once per 24 months.
Main control room emergency ventilation air supply system capacity shall be tested once every 18 months to assure that it is ± 10% of the design value of 1000
cfm.
-0. D ELET*E&
C. Battery Room Ventilation
Battery room ventilation shall be operable on a continuous basis whenever specification 3.9.E is required to be satisf
1. From and after the date that one of the battery roon
ventilation systems is made or found to be inoperabl
associated battery shall be considered to be inoperal
for purposes of specification 3.9.E.
ied.
e, its ble
B - D .L.ETED C. Battery Room Ventilation
Battery room ventilation equipment shall be demonstrated operable once/week.
1. When it is determined that one battery room ventilation
system is inoperable, the remaining ventilation system shall be verified operable and daily thereafter.
2. Temperature transmitters and differential pressure switches shall be calibrated once per 24 months.
•ee CT 7-/5y,
Amendment No. 48, 82, 126, 134, 118, 156, 231, 23.., 269
30 1 Cy (414i9 WI
3.
4.
5.
239
-7, .7, 77, 1
CT5 /" 1.
JAFNPP
Amendment No. 48, 82, 126, 124. 148, 166, 23 ,.233239
(cff(
f
•p�Cjf7(��J 3.7?
JAFNPP
3.11 (cont'd) 4.11 (cont'd)
tnY h-n-andlng oT irradiated fuel, core alteratioi V-Zions with a potential for draining the reactor shall be suspended as soon as practicable
F4. Temperature trans -tters and differenti r e s~sure itches shall be cali ed once per 24 nts
I-Mbati irsupply venntifat -ion ir suppl-y -ý
d once everkl 8 months ke design valu'l of 10001
B.
C.
Battery room ventilation shall tbeoperible o basis whenever specification 3.9.E is requir
1. From and after the date that one of th ventilation systems is made or found t associated battery shall be considered for purposes of specification 3.9.E.
:•pl ,, ,•0 /•,k r•ELETED
k oe.C0l Battery Room Ventilation
n n u Batte ry r o e tl ton equ ip en hall be dem onstrated
ed to be satisfied. operable once/week. ._
e battery room 1. When it is determined that one battery room ventilation to be inoperable, its system is inoperable, the remaining ventilation system shall to be inoperable be verified operable and daily thereafter.
2. Temperature transmitters and differential pressure switches shall be calibrated once per 24 months.
Amendment No. 49, 82. 126, 134, 118, 156, 231, 233, 269
N r�j
N.4
239
,,S
I
JAFNPP5-,c , ,; # . 3.13.7.3
a.Simulated Automatic. Once every 24 months
Actuation Test " a-
ft
b. Flow Rate Test - Each In Accordance with
ESW pump shall deliver the Inservice at least 1500 gpm to-its Testing Program respective loop. The loop. The pump total developed head shall be greater than or equalto te corresponding point on the pump curve, reduced by a maximum of 7%. for the measured flow.
C. Verify that each valve (manual, power operated, or automatic) in the system flowpath that Is
otherwise secured in position,,is in the correct position.
d. Motor Operated In Accordance wit~h v.alves the Inservice Testing
Program
Once per 31 Days
Amendment No. 71, 1 - 4. 0, 241240
/of. Z_
*0
I
/ l \
T&V _C_
*1 .� �
,•. Verify that each valve Imanual, power operated, or automatic) in the system flowpath that is not locked, sealed or otherwise secured in position, is in the correct position.
d. Motor Operated Valves
Once per 31 Days
Amendment No. 71, 14i. 223. 230, 241240
7, '2'
)
IVJAFNPP
3.11 (cont'd)
IS. 7.annelESW instrument 4 AIOinl onIZth channel calibrationel- OceI3monts
E-5& S. 1. 7. Logic System Functional Test
Once every 24 months
W er Syste made or foup to be ingerable iorny ,e.0 cn ued I.c•o aon Is pffmissib, I•._ efo not o exceed 7 dayi0 provided/ihat:
- the o •abl Emerge y Diesel Genera( Syste is)
darm strated to bq)opereble immedi ely and aeily -the eafter; and,
ds are verifie Operable immedi ely and d ly C
Amendment No. !48,r 4 B2 230241
PSW will not be supplied to RBCLC system during,-testin
(�jjjjJzZ/
7 ' t " L Z-2 (Z
Ii
r2.I
JAFNPP i�ht a
From and after the time that one Emergency Service Water System Is made or found to be inoperable for any reason continued reactor operation is permissible for a period not to exceed 7 days, provided that:
Instrumentation'", "check OncP/la
ESW Instrument channel calibration OnceI3 months
f. Logic System Once every
Functional Toot 24 months
2. W will not sup is to o toCi, dur,
esLAZ
'ACTno'J j1 3.
Sthe opereble go Voeh-'l~ ne-rstr tSys m is'• ... demonstretjt to be operabl /mmediately an daily LY/z.' thereafter-, nd,/ ./
all Emor ency Diesel Go rator System a roency L loads a verfied opera immediately a deil there er.
If specification 3.11.D.2 cannot be met the reactor shall 13. be placedin the cold condition within hours.
00
Amendment No. 441,40-Y 230
241 'Aey'2.1
UCA11OIJ 4,92
!
241
JAFNPP
AN* f a 56 I
1,72.
Armencnet No. 'tA 141242
?0, CýR
eel 6(eý 17d'A
soý s
4.12 SURVEILLANCE REQUIREMENTS
'10,I
7fhe reactor may be considered to be in Co It reactor coolant s RCS
'the reactor vessel not vented, and LCO
requirements normally applicable when RCS temperature is
greater than 212"F are not required, to allow performance of
Inservice leak or hydrostatic testing provided the following LCOs are met:
2. L0 3.5F CS-oldSA G lation." atmetys -lw pre~e _su~b~svsItels shall be op~erable,/
2. LCO a3.F........., "Standby Gas Treatment System";
11ý Inservice Leak and Hydrostatic Testing Overation S. I. . 'Perform the applicable surveillance requirements for the
F - required LCOs.
LCO SA. C OA C4_%-.aI -. (V
Amendment No. 34,. 90, 4. 135, 218. 267
REVIION
REVISION D
(
JAFNPP
>C 0ýnoeOlkq*
.LO S 1b. tL4 ,L !)t \(
LTU3,(,4.1~ACA
shall be to be
(� A e.IA(Im4:.v4�� .' .-
7
With the above requirements not met, immediately V
suspend activities that could increase reactor coolant
temperature or ressure and reduce reactor coolant
temperature to e n 12"F within 24 hours.
Amendment No. 267 244b
REVISION D
U0. 3,,.001
56 ),,'C,; % o•,, -IA , I
"L.. -O "S. 1. (P.,"L
.z " f 2
C .,e
JAFNPP
-5eDESIGN FEATURIES
5.1 .1 The Japs A. FitzPateck N lear Power Plant is loc,•,d on the Entergy Nuclear FitzP nck, LLC portion t Nine Mile Point site, approximately 3,000 ft. east of tN!ni Mile Point Nu ear Station, Unit 1. T NPP-JAF site is on-ake
ntario in Oswego unty, New York, apprý imately 7 miles northeast of Oswego. The pi t is locaed at coordin snorh4.8l9,545.0T2meas " 386,968,945 , on the Universal Tra~piverse Mercator Syszem.
, 5.1 .2 The near t point on the propert from the reactor building and ap points1 ,,,'
tof pottial gaseous effluents e eption lake relin is ý a ""
n n th
er
,"X _ -located at the northeast ;orner of the orop~e rtv.T-h' ,i..(.,.n rl 2 p
3,1 ft. an--srthe radisdfthe x n area s d , _••s a o
~-32Rf6CTOR71R -ý t cr ILL
,5"2.1 The reactor core consists of not more than 56 fuel assemblies. Each L _ assembly shall consist of a matrix of Zircaloy lad fuel rods with an initial
..-- composition of slightly enriched uranium dioxide (UO) as fuel material.i._F..el /" assemblies shall be limited to those fuel design approve bythe NRC staff-r I " • e In u v 0 S . .. . / . . . ;(_ .]- / ] / ,
.- ......... ... .!
Section 3.4'of othe ns In;~iT 1.H ~ ' A e2..2.2 The reactor core co *ns 137 cruciform-shaped control rodi~aesciiiný4 L1-a
/ :•ectioh 3.---- fthe-fSAR) -- • r - L- _ _" " ••
5.3 REACTOFR FIRESUEVSE / he reactor.pr.ssure.ve.sel.is he reactor pressure vessel is as described in Table 4.2-1 and 4.2.2 o the FSAR. The
applicable design codes are/described in Section 4.2 of the FSAR.
5.4 C•QNTAINMIENT / / / 5.4:1 The principal design parameters and characteristics for/the primary containment
are given injTable 5.2-1 of the FSAR. / /
•,/ / //
i,'" 5.4.2 The secondary containment-is described in Section.5.3 and the applicable * codes aie as described in Section 12.4 of the FSAR.
5.4.3 Penerations of the primary containment and piping passing through such penetrations are designed in accordance with standards set forth in Section5.2
Amendment No. 29. 42, 10, e i. 66, 71, 100, 117;, 1 62, 2.268
245 �7L.A (�
�D �
REVISION D
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Eq. ..- 6-.-&FUEL STORAGE [L1 .3.1] -5-.-'l--Criticality
[4.5-5(.1 .i The spent fuel storage racks are designed and shall be maintained with:
a. Fuel assemblies having a maximum k. of 1.32 in the normal reactor core configuration at cold conditions (20 C);
b. k.,f < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.3 of theJFSAR; and ;nc h
C. A nomina ceenter nteristance between fuel assemblies placed in thestr a rackse E-l-X7
[1,3,2-1 55.i The new fuel storage racks are designed and shall e ith:
a. Fuel assemblies having a maximum k. of 1.31 in the normal reattor core configuration at cold conditions (20"C);
b. k., :_ 0.90 if dry;
c. k,, • 0.95 if fully flooded with unborated water; and
d. A nominal 6.625 inch center to center distance between fuel assemblies placed in storage racks.
Eq.3.3 5•,5• Capacity
The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3239 fuel assemblies.
(. M7•.5.5 SEISMI DESIG,
SThe eactor building and aengineered safeguards are signed onea basis 7L dynanf• a alysis using accelerati~f response spectrum curve •#hich are normalized to a und
Sfotion of 0.08 g for t• Operating Basis Earthquak)e and 0. 15 g for the DesignjBa sisu
Earthquaken N. ,6
Amendment No. 68, 6-3, 74., '101, 1:76, 256
246
Poise, I/o- 7 REVISION D
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Amendment No. 60. 60. ;8, '344, 37;, ;:;a. 203. 2;o. ;2e,2261
REVISION H
6.0 ADMINISTRATIVE CONTROLS
Administrative Co ols are the means y which plant op rations are subjc. to
management co rot. Measures spec ed in this sectio provide for the a ,gn.-ent of responsibilitie , plant organization. affing cualificatit s and related re iremen~s.
Sreview and •dit mechanisms, pr edural Controls d reporting recui ments. Eac.n of these mea res are necessary ensure safe an- ffiient facility a ration
6.1 RSPONSIBILITY C -n't" -WPLYa. ovev.AY
The ~ ve iceis responsible for s operation of'te pant and shall
delegate in writing the succession to this responsibility dunng his absence. MZ
6.2.1 Facility Mangqmn n ehia upr _' • •.
Onste nd ffste rgaizaion shall be established for plant operation an0 corporatey
management, respectively. The onsixe and offsite organizations shall inclune tnee • •
positions for act.ivities that affect the safety of the nuclear power plant. (
1. Lines of authority, responsibility, and communication shall be establishedi and defined
for the highest management levels through intermediate levels to and incluaing all
operating organization positions. These relationships shall be documented and
updated. as appropriate, in the form of organization charts, functional descriptions of
department responsibilities and relationships, and lob descriptions for key personnel
positions, or in equivalent forms of documentation. These requirements shall oe
documented in the Updated FSAR,
2. The Site Executive Officer shall be responsible for overall plant operation, and shall
have control over those onsite activities that are necessary for safe ooeration and
maintenance of the plant.
3. The. Chief Nuclear Officer shall take any measures needed to ensure acceplacle
performance of the staff in operating, maintaining, and providing tecrnnical support t=
the plant to ensure nuclear safety.
4. The individuals who train the operating staff and those who carry out health physics
end quality assurance functions may report to the appropriate onsite manager;
however. they shall have sufficient organizational freedom to ensure their
independence from operating pressures.
6.2.2 Plant Staff
The plant &Taff organization shall be as follows:
1. Each shift crew shall be composed of at least the minimum shift crew composition
shown in Table 6.2-1:
11<�
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• , 'r -• -J A F N P P r6.0 ADMINIS;TRATIVE CONTROLS • - <q_• " ,- j • '•,
Administrative Cont~rols are the means by which plant operations are subject t.o management control. Measures specified in this section provide for the ass'911..'en~t Of responsib~ilities. plant organization, staffing qualifications and related requirsmen'•s. . review and audit mechanisms, procedural controls andl reporting requirements. Eac.n cf
these measures are necessary to ensure safe and efficient facility operation.
6.1 RESPONSIORLITY
The Site Executive Officer is responsible for safe operation of the plant and srhall
ddelegate in writing the succession to this responsibility during his an
6.2 ORGANIZATION
Z '. .. 1] 6.2.1 Facility Management and Technical Suoport
Onsite and offsite organizations shall be established for plant operation and corporate
management, respectively. The onsite and offsite organizations shall Include tne
positions for activities that affect the safety of the nuclear power piant.
. .. 1. Lines of authority, responsibility, and communication shall be established and defined
for the highest management levels through intermediate levels to and including all
operating organization positions. These relationships shell be documented and
updated, as appropriate, in the form of organization charts, functional descriptions of
department responsibilities and relationships, and job descriptions for key personnel
ire.C V" positions, or in equivalent forms of documentation. These require nts shall b
documented in the Updated FEA1. )&~ A
2. he &~m iv e c hell be responsible fork vall plant operation, ad shall
have control over those onsite activities that are necessary for safe operation and m a in t e n a n c e o f t h e p l a n t . • ; / - A . re-' e C o5 PorOrstIb l 4isP" F ',
I. r" •] 3. Th .ihiaf •u .at ificer •shall take any measures needed to ensure acceptable
performance of the staff in operating, maintaining, and providing technical suppor to
the plant to ensure nuclear safety.
4. The individuals who train the operating staff and those who carry outfel p ics
and quality assurance functions may report to the appropriate onstle manager;
however..they shall have sufficient organizational freedom to ensure their
independence from operating pressures. r "
.,.23 6.2.2 Plant Staff ý .( "q '
The plant &Uff organization shall be as follows:
1. Each shif crew shall be co posed ofat least th minimum shiftcr/ compsnosit-o
mendment No. 6O. 6O. ;7_., 1 , 13-. 4. .7 448. ;Ga. 2• 20, 22, 254
247
REVISION H
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oom u h an SRnr0a
A- SRa go F u a license Ieitelp fuel han nteg shall diremt tupewvin "hlou tof ho perfor shae-l h:6 no other gies during ac t om ie a
92 In the event of allnesd- unexpected absendce,up to two (2) houm t is allan cto pstoe the shift craw to the minimumm pmplemen rr -l1 nTheo m ralons wenager or issistant ioperations anager.'ýh a contr How everivnhM-eeLhan hold an SRO re en lenms rqe s sa t sr o over t tn A •NUeJa 1"ntr~l,.,o • tl'Pm sn=all nOl'a-n RO lioen• or an SIDlcns•{_ . ••
d straimpleentedto limit the w , iAmiini thatiWe ~ aT be developed and impmetd wrking
husof aj i~who perform safety-lte •ntos•~:,nior Pctor fperat-ors•
tob o auxiliary operatorsd anof mahutdown f oreaintenane ol- Mr
aquate sn verag shall mintained without routine hea use of overtir. The objective shapt to haveO:o rtn m snnlwrk a r ,8 hus nminal our week, wh h plant is pop ting.
!However in the event tJ~a unforeseen blems require stnial amo) ts of overtime L to be u'dor during e~tnded period ,•of shudw forl leling, mjreal!! nno o Lo , ;9.. *,u ;R,.!,,• 4", a; he fmllmuu, " ;~~Alin-me eh Ik fnllnwari"
shf turnover) meru a Anindividual s ud n tb e a to w r m than 16 h o r traight. 0
b. An individu should not permitted to w more than 1 ours in any 24 hour period, n more than 2 hours in any 4 our period, n more than 72 ho rs in
c'. A . ak o:rfrat least '=l ht hours" shOul be allowed betw' n work period sift•
tu over time can Included In the reaktime.
' d. cept during ended shutdow periods, the u of overtime shoul be cons' ered
Any deviation fom th guidelines shall bec a zeb t eit e or€ th naa,,•, lUl02no -r --- -- ' mtofls nhigher Ievj§ of ma la eelin accordance " w~stabrished •ocedures and with documentation of the basis for rnig[ th
ation.~~~~ ~ Off e n dd inr U :ocd esw(cthtiondi t al ovlertiprhat • d mon by the 6 Eecutor crJs o As d e
"Aa enodment o. .11 10 127 deviation from th17 2,2uidelines
Amendment No. 40. 1, 430 437', 48,• -2 .2 228, 244..7
247a REISIN4 H1
REVISION H
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D54 .2 b]
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6.3 PLANT STAFF QUALIFICATIONS "6
6.3.1 The minimum qualifications with regard to educational background and experience for f plant staff positions shown in FSAR Fgure 13.2-7 shall meet or exceed the minimumJ qualifications of ANSI N18.1-1971 for comparable positions; except for the radiation I 1
protection manager who shall meet or exceed the qualifications of Regulatory Guide V) 1._8, September 1975. f' _y •
6.3.2 -- a. _ I hal 4eet or exceed the minimum requirementki) Pr 1 ~jn I (Co~mbined SPUD/ IA P'osmon) or l;option z i•ontinued use 9r STA
[ Po I anudefined in the Commission Policy Statement n ngineenng tse on Shift, published in the October 28,,1985 Federal Register (50 FR 43621 . ; (i nv king (•Ip o n 1 tj~ 3 1 p ro 7 rn -e fi lled b y t* 6 S h idft M a n r o r (;; ,, m
6.3.3 A ny detviAs w11 be jusyified to the NRCrTor to an individ) al's filling of o pe of these)
NOTE:
(1) ie 13 individuals ýwho hold SRO licenses, and have completed the FitzPatrick Advanced Technical Training Program prior to the issuance of License Amendme
6.4 ETRINING AND REPLACEMENT TRAININ L A Atraining program shall be maintained under the direction of the Training Manager to"
assure overall proficiency of the plant staff organization. It shall consist of both
reCmaining and replacement training and shall meet or exceed the minimum requirements fSection 55 f ANS IN1 .- 7 .
The retraining program shallI not exceed periods two years in length with a curriculum •.designed to me et or exceed t er uaii ti n e u r ments of 10 CFR 55.5.,
6.5 /REVIEW AND AUDTS?& Cý-)-J 0LA.. + G,,-• ,t,
Review requirements are completed by using designat-ed tachnic.al • reviewers/qualified safety reviewer and two separate review committees. The Plant Operating Review Committee (PORC) is an onsite review group; the Safety Review Committee (SRC) Is an independent offsite review and audit group.
6.5.0 REVIEW AND APPROVAL OF PROGRAMS AND PROCEDURES
6.5.0.1 The procedure review and approval process shall be controlled and implemented by administrative procedure(s).
6.5.0.2 Each program and procedure required by Specification 6.8 and other procedures that affect nuclear safety, and changes thereto, shall be reviewed by a minimum of two designated technical reviewers who are knowledgeable in the affected functional area.
Amendment No. 22, .*4, 67. go, 11, 134, 137, . ;7. 0 21., 222.. , 270
248 . ('4" REVISION H
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6.3 PLANT STAFF QUALIFCATIONS
"6.3.1 The minimum quarifications with rard toducational background and experience for
plant staff positions own' FS -u- .2-Zhall meet or exceed the minimum , qualifications of ANSI-I o comparab positons except for the radiation
-T? Sc ' Z4-iprotecton manager who shall meet or exceed the qualifications of Regulatory ud 1.8, September 1975.
6.3.2 'The Shift Technical Advisor (STA) shall Meet or exceed the" minimum requirements of'•• /
either Option 1 (Combined SROISTA Position) or Option 2 (Continued use of STA Position), as defined in the Commission Policy Statement on Engineering Expertise on Shift, published in the October 28, 1985 Federal Register (50 FR 43621). When 40 invoking Option 1, the STA role may be filled by the Shift Manager or Control Room V
6.3.3 Any deviations will be justified to the NRC prior to an individual's filling of one of these
positions. r NOTE:
(1) The 13 individuals who hold SRO licenses, and have completed the FitzPatrick Advanced Technical Training Program prior to the issuance of License Amendment 111, shall be considered qualified as dual-role SRO/STAs.
6.4 RETRAINING AND REPLACEMENT TRAINING T .. S e e- 0 3.2
A training program shall be maintained under the direction of the Training Manager to assure overall proficiency of the plant staff organization. It shall consist of both retaining and replacement training and shall meet or exceed the minimum requirements of Section 5.5 of ANSI NI18. 1-1971."
The retraining program shall not exceed periods two years in length with a curriculum
designed to meet or exceed the requalification requirements of 10 CFR 55.59
6.5 RE~nEW AND AUDIT
Review requirements are completed by using designated technical reviewersqualified safety reviewer and two separate review committees. The Plant Operating Review Committee (PORC) is an onsite review group; the Safety Review Committee (SRC) is an independent offuite review and audit group.
6.5.0 REVIEW AND APPROVAL OF PROGRAMS AND PROCEDURES
6.5.0.1 The procedure review and approval process shall be controlled and implemented by administrative procedure(s).
6.6.0.2 Each program and procedure required by Speclflcation 6.8 and other procedures that affect nuclear safety, and changes thereto, shall be reviewed by a minimum of two Cdesignated technical reviewers who ore knowledgeable in the affected functional area. A
Amendment No. 22, 34, 67. . !1!, Q31, 137. 178, 1R0, 213, 222, 22B. 270 H 248
REVISION H
JAFNPP
see6.3 PLANT STAFF QUALIFICATIONS .... ... ...- • . . .
/6.3.1 The minimum qualifications with regard to educational background and experience for \ plant staff positions shown in FSAR Figure 132.-7 shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions; except for the Radiological-" and Environmental Services Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, Seember 1975.
6.3-2 The Shift Technical Advisor (STA) shall meet or exceed the minimum requirements of, ch• either Option I (Combined SRO/STA Position) or Option 2 (Continued use of STA L' Position), as defined in the Commission Policy Statement on Engineenng Expertise on Shift, published in the October 28, 1985 Federal Register (50 FR 43621). When invoking Option 1, the STA role may be filled by the Shift Manager or Control Room Supervisor. (1)
6.3.3 Any deviations will be justified to the NRC prior to an individuars filling of one of these
positions.
NOTE:
(1) The 13 individuals who hold SRO licenses, and have completed the FitzPatrick Advanced Technical Training Program pnor to the issuance of License Amendment 111, shall be considered qualified as dual-role SRO/STAsM.
/6.4 RETRAINING AND REPLACFMFNT TRAINING
A tramin6n program shall be maintained un the direction of the Training Manager to assure overall proficiency of the plant staff Mranization. It shall consist of both rmiminirm andl rm~anmmnt tminir nd oJl mm r excm.d tha minimnum I(-
require ~nts of Section 5.5 of ANSI NIS. 1971. ,The raning program shall not exceed lriods two years in length with a cumiculumi' " desighed to meet or exceed the requalion requirements of 10 CFR 5.59 -"
S Re"•v rsiquirme• are completd by, dgesignated technica.l re rirqualifiled safety reviewer and/two separate review comritteea The Plant
Review Committee (PORO) 4s an onsite review group: the Safety Review
".5.0H-Afte (SRC) is anu•id dent of•sits review and audit group.
6.5.0m A &JrA% APPROVAL OFPRC&AMS AND PROCFDRMI A7
I 6.5.0.1 procedure review and aproval shall be controlled and* mented by
instativ procedure(s).
8.5.0.2 cli program and proceduur by Specification 6.8 and procedures that ,.o. /. =
affect nuclear safety, and ch im thereto, shall be reviewed by a imum of two dsignated technical reviewer o are knoewledgeable in the fnctional area
Amendment No. , 11, 6... 99, 11, 4.U , 4.0,246.222, 228248
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6.5.0.3 Designated technical reviewer(s) shall meet or exceed the qualifications described in
" ".- - Section 4 af'ANSI Ni 8.1-1971 for applicable positions, with the exdussion of the
positions identfied in Sections 4.3.2 and 4.5. Individuals whose positions are described in Sections 4.3.2 and 4.5 may qualify as designated technical reviewers provided they meet the qualificatons described in other portions of Section 4.
6.5.0.4 The designated technical reviewer shall determine the need for cross-disciplinary reviews. Individuals performing cross-disciplinary reviews shall meet or exceed the
"qualifications described in Section 4 of ANSI N1i .1-1971 for applicable positions.
6.5.0.5 Each program and procedure\required by Specification 6.8 and other procedures that affect nuclear safety, and changes thereto, shall be reviewed from a safety perspective_ by a qualified safety reviewer. Safety andior environmental impact evaluations, when
required,"shall be reviewed by PORC~per Specification 6.5.1.6.a.
6.5.0.6 Nuclear safety related procedures and procedure changes shall be reviewed and
approved, prior to implementation, by the a rte member(s) of managementn
6.5.1 PLANT OPERATN(a R I•V COMMrl"TF (PORC)
6.5.1.1 Function
P C shall function to the Site Executive Officervon matters related to nuclear and environmental impact.
6.5.12 Membe siplbco.sdf
Te Plant .Review Committeshlbe copsdf a Chairman, ViceChairmn members designated in iwrting by the Site Executive Officer. Members shall collective have responsiblity in at least the areas of Operations, Maintenance, Radiologic alo m." nmental Services. Engineerng, Reactor Engneenng, Instrumentation Control, Nuclear Licensing, and Quality Assurance. -'
Amendment No. G4+-P-40-,n. 22824ft
Pc- I e 2 o-P I
JAFNPP PT~5Cv(Art-e 69.0
Alternative members shall e appointed in writing by the PORC hairman to serve on a
temporary basis.
6.5.1.4 Meeting Freque y
Meetings f be called by the Chairman as the oc sions for review or investigation arise. etings will be no less frequent than o e a month.
6.5.1.5 Quor m
A quo of the PORC shall consist the Chairman or a Vice-Chairman and fViv mem rs including designated alter ates. Vice-Chairmen may act as memb when not a g as Chairman. A quorum s!ill contain no more than two alternate
6.5.1 Responsibilities
The PORC shall be responsible r the:
a. Review of 10 CIF .59 safety and environmental* pact evaluations associated with procedures and programs required by Specifi tion 6.8, and changes thereto.
/ "e Ifpe ul a a e y
b. Review of'proposed tests and experiments that affect nuclear safety.
c. Review of proposed changes to the Operating Ucense and Technical Specifications,
d. Review of proposed changes or modifications/o plant systems or equipment thataj.ect nuclear safety.
e Investigation of violations of the nical Specifications. The PORC shal repare and present a report covering e evaluations and recommendations to event recurrence to the Site Exec ive Officer, who will then forward the re ort to the Chief Nuclear Officer, th. irector Regulatory Affairs and Special Pr jects, and to the Chairman of the Safety Review Committee.
f. Review of plant operations to detect potential safety hazards. //
g. Performance of 4ecial reviews and/or investigations at the re est of the Site Executive Of r. r
h. Review of I reportable events.
i. Revie of the Process Control Program and the Of ite Dose Calculation Manual (OD'A) and changes thereto.
A Amendment No. 2, 57, 94, 1:70, 210, 220, 222, 228, 240, 252
249
Revision B
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6.5.1.3 Alternates
Alternative members shall be appointed in writing by the PORC Chairman to serve on a temporary basis.
6.5.1.4 Meeting Frequency
Meetings wiul be called by the Chairman as the occasions for review or investigation arise. Meetings will be no less frequent than once a month.
6.5.1.5 Quorum
A quorum of the PORC shall consist of the Chairman or a Voce-Chairman and five members including designated alternates. Vice-Chairmen may act as members when not acting as Chairman. A quorum shall contain no mrom than two alternates.
6.5.1.6 Responsibilities
The PORC shall be responsible for the:
a. Review of 10 CFR 50.59 safety and environmental impact evaluations associated with procedures and programs required by Specification 6.8, and changes thereto.
b. Review of proposed tests and experiments that affect nuclear safety.
c. Review of proposed changes to the Operating Ucense and Technical Specifications.
d. Review of proposed changes or modifications to plant systems or equipment that affect nuclear safety.
e. Investigation of violations of the Technical Specifications. The PORC shall prepare and present a report covering the evaluations and recommendations to prevent recurrence to the Site Executive Officer, who will then forward the report to the Chief Nuclear Officer, the Director Regulatory Affairs and Special Projects, and to the Chairman of the Safety Review Committee.
f. Review of plant operations to detect potential safety hazards.
g. Performance of special reviews and/or investigations at the request of the Site Executive Officer.
h. Review of all reportable events.
I. Review of the Process Control Program and the Offsite Dose Calculation Manual (ODCM) and changes thereto.
J. Review the FirtPatrick Fire Protection Program and 4mplementing procedures and changes thereto.
Amendment No. -'2,,,,, ,4, 1:,"e, ;1161 ... a.. 22.. 240
249
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6.5.1.7- Awthority
The POR
a. Recommend to the Site Executive Officer, approval or disapproval, of those items reviewed under Specifications 6.5.1.6.a through 6.5.1.6.d.
b. Render determinations with regard to ýte or not items considered under Specification 6.5.1.6.a through 6.5.1.6.e constitute an Unreviewed Safety Question as defined in -10 CFR 50.59.
c. In the event of a disagreement between the PORC'and Site Executive Officer, notify the Chief Nuclear Officer and the SRC Chairmrn..or their designated alternates, within 24 hours and provide written notification by the next business
day. The Site Executive Officer shall have responsibility for resolution of such disagreement pursuant to Section 6.1.
6.5.1.8,, Records
• Minutes of all meetings of the PORC shall be recorded and numbered. Copos will "•be retained in file. Copies will be foiwarded to the Chairman of the SRC and the Chief Nuclear Officer.
6.5.1.9 Procedures
Conduct of PORC and the mechanism for irplementation of its responsibilities and authority defin the pertinent Admiinistative Procedures.
6.5.2 RAFT REVIEW CMUMVr9F (SRC1
6.5.2.1 SRC shall function to ftes in the areas of'.
review and audit of designated
Amendment No. 5... 76, 98, 94, 10, 2", 2.., MO 228
250
C shall:
I.
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h. Eo c an ica t e n g in e s r mn. Emergency
P nangi" g Sn. O ther approp rit e fields associated w ith the 'unique cl d 'a. oi stcs of a nu l a pow er plant / ' A/ 6 cond f theSRC will be in accordance with a'charter apprve by the Chief Nuciar 6. . . hOfficer. ~ nThe or w il define the SRC 's au hI it '2 establish the m echanism for carryin
out its rspon ilities.
/
k.CMfl VStrutum uai a harnannd
C. hair an. m be s e appointe d by the ounR e latr ya A ffais gn n
a nO a h e r atro p t i neld s a sso at e w.. , ih t i c a o f t 2 e e0ia e w pt a s h a l av e a
L 6.52.rTe cor fRtegltr ARCaeirs andcpcoranclwt 4ard approved by the Chief Nuclar Ofcr
Officer uTahe may "ueda detiemi y the SRC Chauirman esablis temcans m for cin the
6.5.6 The SRCs1 comet at leas tsex irths incldin a Cianac
A2n n o,1, , ha%
aproe b heCie ucea ffr SR mbr ad l251 ss hv
:n d i ore nenierigorapyscl orteeqialna hl4hv
JAFNPP
6.5.2.8 The SRC rIew fa at leat a majoriaote wi't Qui ualt (Auranc e Prognam, des) cred im Chapt"e 17 of td e JAF FS No. "a
6.5.2.9 A Sf facility activities a ee oredu ance wf the SRC and e
a with the Quality Assurance Program, descrbed in Chapter 17 of the
Amerdmeit No. .6,. , ..... i2g,210.251
Revision B
As
A I Affondment No. 66, 60, 65. din -ilb ý 40mom ".9 "N-
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Revision B
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2 1Amgndment No go, M, 20, 22, 25 125Ub
Revision B
2.o
SThe fohowng actions shal be taken for Reportable Events:
S(A) The Commission shall be notified and a report submitted pursuant to th
requirements of Section 50.73 to 10 CFR Part 50. and
S(B) Each• Reportable Event shall be reviewed by the PORC, and the results of this review
shall be submitted to the Chief Nuclear Officer, the Director Regulatory Affairs and •..._ Special Projects, and the Chairman of th SRC. j
SAFET' LIMIT ViOLtTIONIA
(~If-T a fl= z)faafetylimftidtis exceeded, the reactor shall be shut d parse n F :I97a~ro ce vsi so.lCRO3 IcIie; A
Amendment No. se.s..... 93. 11.. . 22... H .. ., 240
REVISION D
253
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(A)1 The Commissin SW be MO an a rexl ••i-rt. to the
miquirments of SectoM 50.73 to 10 CFR Pant 50, and
(1) Ea:h Reportabse Event shall be rfvewed by the PORC. and the results Of this revW shall be submitted to me Chie Nuclea Oftw' onDiecto Regui____Affars an
(A) If a sef" WAmis Isi ax•d.d the reactor shl be shut and reco oprtinshall
only e resumed in acodac with to prvsin of 10 CFR 50.3,6.()()
(B) An imme:.at* report of ealch safety kMit violation shafl be made to the NRC by, ie
Site E~xecutive Officer. The Chief Nuclear Office, toirector Regulatory Affairs an SpecuM Projects, and tie Chairman of the SRC wig be n ifted w 24 hours.
C The PORO shall prepare a complete kvweutgat"v report of eadi safety limt violation and include appropriate analysis and evaluiation of: QI) applical circumstanlces
\ systems or strucure and (3) cofctv at mc e.ii,,• ce.• • The SSite
Executive Officer shall forward this report to the Che Nuclear Offcer. the Director recuatorAffairs sa -- anP•" -';"- eSC nd tn.
.RC.I,-a.
WAI Written nrocedur$ I
.1
"sla be established, implemented.and maintamedaf
t. tireqwiriMep and reo&ffi of Section5flOf Z4 ANSI I6.. 199722 * Fac i Poices Pocdre
" ftrecomrnendebd in Appendix A ofRegulatoty Gukid 1.33, November 1972.
Wo)4 naeit the FiwProtection Progam. bL~q d1 1 ~GAe.pwcoavesiEJ
,*c~m m~ueprogras spedfid pendx the~ ~ i~
macf prmaumof peciicaion6.8.(A), and chaiages thereto, shall beaprvd Y!
AJq/dLC & 73') 9%r S. .54--a L ai-3
Amendment No. Se.--W. 65. i0 18, vI. s. 2. r. 4
253
REVISION H
0
• crý3A
11,nl
CrTSCki 6.c0
JAFNPP
The :lowir; actions sha ll e t ki _n for RaMkf l l Ev
(A) The CommIS'=o sMhatU be noti•WiAdnd a report suj.itted pursant to. *ltnaumetiof ecin5.73 t0~10 CFRPad 5W=~
B) E,,h Reo�'�t�f Event shaU be eviewe the "PORC, and theis su•ts of this review "shall be s bmifed to the Chief Nucfler O DirectorR Affairs and
S ecial P M,. ann the IChairm of t4e SRC. 6"• L ET IM rT Ze z s2 .
( A ) If a S a f e t y lim it ýe d, t A r e aM o U n g b s h u t d l aw r n r e co r o =e • s h a l11 o m erisll i ~ c m n w it th e p v i i n o f 1 0 C F R $ 0 .3 6 ., € ( 1 ( )
(B) An imrne: ale report of each safety ulmit violation shalt be made to the NRC by the
Site Exec.utve Officer. The Chief Nucea Officer, the Director Regulatory Affairs and
Special Projects, and the Chairman of the SRC will be notified within 24 hours.
() The PORC shall prepare a complete investigative report of each safety limit vioation
and include appropriate analysis and evaluation of: M applicable atcumstances preceding the occurrence. (2) effects of the occurrence upon facUlity component systems or structures and (3) corrective action required to prevent recurrence. The Site Executive Officer shall forward this report to the Chief Nuclear Officer, te Director Regulatory Affairs and Special Projects, the Chairman of the SRC, andmthe
() Writtn procedurs -and adminitatv oiie~s s"Nl be osiblse, imleened and maintained Vtht
1. meet or exceed One requirements and reconmermetkns of Sectio 5 of
ANSI 18.7-1972 Facility Administrative Policies and Procedures.
2. am recommended in Appendix A of Regulatory Guide 1.33. November 1972.
3. irnplement the Fr Protectio Program.
4. irdude programs specified in Appendix B of the Radiological Effluet Technicai Specifications. Section 7.2.
() Each promedure of Specification 6.e.(A), and changes thereto shal be approved prior to implementation by the appropriate responsible member of management specified in Specification 6.5.0.
Amendment No. Se. ... .. :7. .. .. C. pee.. .... . g e. , 240
253
S...... .. .r- -- _ o - 14D
JAFNPP
(C)Tm•aay to the procedures nKfuired by Specifi 6.8.(A)ma be
S2. the 0• is SW~rvWd by two members of the management staff, a
leas of whom holds a Senior Reactor 0 s license.
3. tim e is documented, reviewed and SW by the appropriate bet of pan managemet as req, red bypecifoation 6.5.0 within
day" of iniplemenitaton.
(A) ROUT1NE REPORT
The following reports Shall be submit'ted in accordance with 10 CFR 50uls
ootmwwise noted.
a. A summary report of plant strtup and power escalation testing shall be submitted following (1) amendment to the license involving a planned increase in power evel (2) installation of fuel that has a different design or has been manUfactured by a diffetent fuel supplier. and (3) modifications that may haw significarnty altered te nucler, thermal, or hydraulic performance of the plant. The Startup Report for the inikil fuel cycle shell address each of the tam identified in the "FSAR end shall include a description of the measured values of the opemrt conditions or characteristics obtained during e test program and a comparison of these values with design predictons and specifions. Any corrective actions that were requred to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. Starump Reports for
abeeQue fuel cycles shall address startup tests that are necessary to demnonstat~e the acceptability of changes and modifictos
Amendment No. 7. 2. 2,. -", 213,222
254a
o~ief Lik~IIJ U-rv- V I
JAFNPP
EWEQ�EW�
R.OUTINE REPORT MRT
The ollom rporsha" I be submitted in accordance with 10 CFRt
a h. ml or hydrauli of th plant. shettrtup Ri oer a es.shl
be submitte floi (1) araimn to the Ii i nvl ving a
ifort inicial i Well yl a ainch of the etas inti ie i "m•,lan shal include a deeign of the m easured balu u ther. pro nd ompar:on of • ths t d•n pcon and *3 oniati en co het sive e fcanu weerqae ow
obtkncr • o peatonsh a-so be u .bed •.•. Anyad itoa
comma en or indra u ded i- this re•ant. he Reprts for to ndemonstatete acceptbiit of chanes andx m ausodfctions
Amerndment No. , -22 :, " IQ 212.222
254aRje I .; 9
REVISION H
[-5.]L 6.9
(A)
-o
/
(A) ifi(iont ,rue,!
b Startup 3 eports I b. subaft vitals (1) go da olloviaq lomition. of the UP test preq *or (a) 90 day folioviaq 4i r tioa or camme at of eommerc L4e oeseo o wUi r iS earliest. If the Start-up sprt does sot co both eves L.e.. complotiou f startup test and4 reaumptios or C00es0c at of Commercial r operatioa, lemestagy reports sball be eted atlat r7 th~ree 010atba W both eveats are comiples -0 b
53.m:.ta
A tabulationes as man uual basis of the aumer of and other persomiael Ciacludmas cesatraavmra,n4
254-11
-ýkq 2- 04T
REVISION H I
Lw.
p
r
-5ec,+l Qv% 15-4
81
Ameadmat so. 0. ý 130
JAFNPP
(A) ROUTINE REPORTS (Continued)
4. CORE OPERATING LIMITS REPORT
- 1 o a. Core operating limits shall be established prior to startup from each reload cycle.
or prior to any remaining portion of a reload cycle for the following:
"* The Average Planar Linear Heat Generation Rates (APLHGR) of Specificationak
"* J njmum Critical Power Ratio MCPR n MCP lo flow ad ment
cgof Specifications Q.1.8 an••
The Linear Heat Generation Rate (LHGR) of Specificatio)
* The Reactor Protection System (RPS) APRM ow iase n of
9 The ow a Rod Block Monitor (RBM) rod block
Tab[ -and fqg
* The Power/Flow Exclusion Region of Specificatior y" .D
and shall be documented in the Core Operating Limits Report (COLR).
b. The analytical methods used to determine the core operating limits shall be
(..' "' J •those previously reviewed and approved by the NRC as described in:
1. "General Electric Standard Application for Reactor Fuel," NEDE-2401 1-P-"O
2. "James A. FitzPatrick Nuclear Power Plant SAFERPGESTR - LOCA
Loss-of-Coolant Accident Analysis," NEDC-313171if- ý ý ý
3. "BWR Owners' Group Lon-term Stability Solutions Licensing Methodology," NEDO-31960-,j:Mjý
b
+,oj•-ca .j % i•. , , ,', 1 G4s;t.A
-o-- Vr-r1'Cr 44,*- to LfL.. R.o4
Amendment No. 442, 2 ,2 266.254c
REVISION H
fIr.0
1%~
'4%
I
I1
0-"
5--e- 0 " ý-. (, E ý
.�Ee+IOmI 4�>c�
JAFNPP
I •r5c. The core operating limits *u be determined such that an applicable limits
(e.g., fuel thermal-meChanital limits, core tVhoem-hydraulic limits, ECCS limits, nuclewr limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis re met.
d. The COLR. inchlding any mid-cycle Miuons Or e thereto.
-5
Amendment No. ,v4 1.44a--, 3254d
REVISION H
.T
I
a 3 S .4 I La
U'
-v
mC) Co
iIT
-SoV
LW
N
N
IN
i i
JAFNPP
2. Reor nd =PM at pitnciai Rma f tehne acMies fupcirU.tp
3. 4.W Itoybl the c
5.React=. of ata au" and yearfoffft
(A. Thetoiwm of buihs ma edtoeratleasProceduesw.
7. Recorf eds k of mi~c2 W madtm c~pi aecUiit eUira
4. Records of seald ource laktas. aetla and rusut abmmi.db m
S. Records df arirWa physical inventory Of al source matetial orcrd.w
AMendmentNO. U-G.4404;. 218
2M9g
Revision B
JAFNPP
(B SPECIAL REPORTS
S 1. Fifteen copies of the Evaluation Report of the results of the first five years of
performnance of thefnon-destructive inspection listed in Table 4.6-1 of Tech~nical
Specifications 4.6.F, Structural Integrity, relating to te FitzPatrick in-service inspection program shall be submitted to the NRC, Director of Operating Reactors, within three months of the completion of the fifth year of the program.
2. DELETED
t.1O' RECORD RETENTION S .
(A) The following records shall be retained for at least five years:
1. 'Records and logs of facility operation covering time intervals at each power level.
2. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
3. All Reportable Events.
4. Records of surveillmnce activities, inspections and calibrations required by these Technical Specifications.
5& Records of reactor tests and experiments.
6.\Records of changes made 1o Operating Procedures.
7. Recorqsof radioactive shipmen~ts.
S 8. Records •d sealed source leak tests'and results.
9. Records of annual physical inventory of all source material of record.
Amendment No. a44-4 C 6, 218 254g
F4ee (I o" z2-
(8)/ The 901oWLovS Creors .shall be retaitned tat the dutation of the• Facilitry OperatiLng License:
1. Retords of any drawing chan a reflecting facility d ign mod1i4c1 ion& &ade to sySt.em and equipment described the rinal Satf, Analysis Report.
2. Recor sf new and irradiated fue* inventory, fuel transfe a and Assembly urnup histories.
3. Records of iliy radiation And co ,anation surveye.tfei 4. Recorsat toZ rdtin eipasure tot al~l I Ivtdua3•s entering tadi i~on
Control ateu a s
3.noad. at gaseoMuls and •liquid radioactive ntateial released to the
1 I0. Recordts of seo eva performed f~or _ebasg• made to procedure. • S• equi.pment or vli~ews of test. and esper atn.8 pursuant to \/
11 Records f meetia o tOeP-01"adf theV-01 f
2.2. )ecori. foc _mnilo ~nata Qualification vwtrhich covered under the\ p. 4oi n trofns ptanent h o6.2.T5.o
/ u~a~el nce s ida entifi •ed I •J~ -p e lteds fta thos facil .....ity
Re aec r of teraineviand ift.aof all hydfatlic ned ecasi afto ubbs1
whon stff.uecudaw eyafctayofeyrae ytm a. ago othe daetve at whi the petrvielf m en pu suan d t, sociat*
9. Rcedr ord peri ssnuradieaction pr trequired byl repare n
a.h Rcauos for *1 platop e tio n b. q sad* procedures sha r"So me to at tvies rafitestsand enpsres ci pursIant to
1asRe cr as of belinow the 1± C an s heif n 0~ 0a 1rct 1. ec T heto pmit cede Qualifcaionld pwhicht coveedacuo andeth
tanisg or topeai and maintena reod actofithe Tfh Iey s dalla ofti to0
*25 S '
Prcdt o esen :1to ttc~o hl oa* n ad ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~RVSO Hdt a 2patae lng hspoeue hlb
(a) The following zecoras shall be retained for the duraino h racility O~perating License:
1. Records-of any dra7ing changes reflecting facility design
modifications made to systems and equipment described in the Final Safety Analysis Report.
2. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
3. Records of facility radiation and contamination surveys.
4. Records of radiation exposure for all Individuals entering radiation control areas.
S. Records of gaseous and liquid radioactive material released to the
(ý6. Records of transient or operational cycles for those facilit " • components identified in Table 6.10- .
7. Records of training and quaT IcaT on for current members of the plant staff.
a. Records of in-service inspections performed pursuant to these Technical specifications.
9. Records of Quality Assurance activities required by the Quality Assurance manual. r
10. Records of reviews perfoered for changes made to procedures or equipment or reviews of tests and experiments pursuant to. 10 CPR 50.59.
11. Records of meetings of the PORC and the SRC.
12. Records for Environmental Qualification which are covered under the provisio~ns of paragraph 4.15.
13. Records of the service 1le of all hydraulic and mechanical snubbers, whose failure could adversely affect any safety-related system, including the date at which thbe service life commences and associated J Installation and maintenance records as of the effective date of this annedm*ent.•
ADIA?!ON PROTECTION PROGRAM
Procedures f r personnel radiation protec ion shall be prepared and adhered to for 1 plant operations. These ocedures shall be formulatea to intain radiation exposures iceved durinr operat n and maintenance as at below the limits specifi d in 10 CYR 20 as practicable, he procedures shall Include lannmng, pieparation and training, for ration and maintenance act vities. They shall a so include expos re allocation, radiation an contamination contro %techniques, a~d final debriefing./
Amendment No. •) .•(•, 92 Order dated October 24, 1980
255
?4$S-/ 0 f2 tgZ.
(a) The following recoros shall be retained for the duration of the Facility Opetrating License:
1. Records ofany drawing changes reflectIng facility design modifications made to systems and equipment described in the PrIal Safety Analysis Report. \ " 2. Records of new and irradiated fuel inventory, fuel transfers and
assembly burnup histories.
3. Records of facilityradiation and contaainao surveys. 4. Records of radiation exposure for all dividuans\entering radiationi
control areas. 5. Records of gaseous and liqgaid radioactive material released to the
envions.
6. Records of transient or operational cycles for those facility " 1_. c€omponents ident ified in Table 6.10-1. ,
. Records of t aining and qualification f-or current memb r- of the plant staff. "
a. Records of in-s"rvice inspections performed pursuant to t so Technical Specifications.
P. Records of Quality surance, activities required by the Qualit Assurance manual.
10. Records of reviews per r•red for changes made to procedures or ipment or :eviews of ests and experiments pursuant to
1cia 50.59.
.11. Rec eds of meetings of the C and the SRC.
12. Recur s for ZnvIronmental Qua fication which are covered under the provis orsof paragraph 6.15.
13. Records the service life of al hydraulic and mechanical 'inubbers, whose fai re could adversely affcc any safety-related system, including e date at which the service life cormences and associated-' installatlo and maintenance records as of the effective date of this Saasmendment.\
•
6.11 RADIATION PROZCTIoI PRoGRAm ( Procedures for personnel radiation protection shall be prepared and adhered to for all plant operations. These procedures shall be formulated to maintain radiation exposures received during operation and
maintenance as far below the limits specified in 10 Cia 20 as S pract ica ble. The p r~ocedures sh all include planning, preparat ion, and St~a innog fo r ope ra tio n and ma intenance act iv it ie . T e s a l a s
include exposure allocation, radiation and contamination control
Amendment No. 92, ed , 2 55. 7 Order dated October 24, 1980
25S
F~e I ~( z.
t-T - L'ý V"x"i " -4--
SeP-C --Y ;vi
JAFNPP _--••
[._7 6.11 (A) High Radiation Area
1. In lieu of the 'control device" or *alarm signal required by paragraph ;W03(c)
(2) of 10 CFR 20, each High Radi *on Area (i.e., z 100 mrem/hr) i vl*ch the
intensity of radiation is 1000 m/hr or less shall be barricaded PM conspicuously posted as a hW radiation area and entrance the to shall be controlled by requiring is nce of a Radiation Work Permit ( ). Any individual or group of I viduals permitted to enter such ar as shall be provided with or accompanied one or more of the following:
a. A radiation nitoring device which continuously *:icates the radiation dose rate• the area.
b. A ra 'aion monitoring device which cont• usly integrates the radiation do rate in the area and alarms when a set integrated dose is received.
try into such areas with this monito 'g device may be made after the dose rate level In the area has been blished and personnel have been made knowledgeable of them.
c. An Individual qualified In radietio protection procedures who is equipped with a radiation dose rate ma ing device. This individual shall be responsible for providing pos e control over the activities within sa and shall perform peridic diction surveillance at the frequency ecified by the facility heal p icist in the Radiation Work Permit.
2. The requirements of . 1.A.1 above, shall also apply to an igh radiation area in which the inte of radiation is greater than 1000 m/Ihr. In addition. locked doors sh be provided to prevent unauthoriz entry into such areas and the keys sh maintained under the administra control of the Shift Manager uty and/or the radiation protectio anager.
P%
"Rediation psonnel shell be exempt from th RWP uncu requirement during the performan. of their assigned radiation protection duties, aidd they comply with
pproved on protection procedures for entry into high iaton areas.
Amendment No. 22. 27. 29, 7B. _228 270 25-8 4- I" c RF"VI•IfN H
REVISION H
Amendment No. 38 9k 0 2 Do f 2a?257
ý17
JAFNPP
____________________ ) '
Q 3EMERG EN P N
THIS S CTION PURIP0J Y BLANK
Amendment No. -:7, -6,- .Be, 252258
po-'e 3 4 2-.2.
Revision B
Z11
I
6.15 ENVIRONMENTAL OUALIFICATION
A. By no later- than June 30, 1982 all \safety-related electrical equipment
in the facility shall be qualfied ft, accordance with the, provisions of:
Division.of Op aling Reactors "Guidelines for Evaluating "Envirommental
Qualification oe lass IE Electrical Eq.ýjpment in Operating"9ea.ctors'
(DOR Guidelines); '6r, KUREG-0588 "InterimS\taff Position on E ironmental
ualificatlofl of Sa fiy-Related Electrical quipfnent", December 979.
ies of these docwu e ts are attached to Or er for Modification
Li nse DPR-59 dated Oc ber 24, 1980. "
B. By .no tar than .Deceber 1 '1980, complete- and a itible records must Be aBan ble and maintined a 1a central location ich describe the
be av ........
envlT'o~e 1 qual~fication ,me� od used for all safeEC -related
eletrial ~uet n suffi cindealt documet*&dgreo compliance wi th •e DR Guidelines .r IiUREG-O88. Ther e'er, sucn records Should updated and mainta a ed current as equitent is t
replaced, furtherested ,or o therwise further qualt ifed. /.
o(e
Order dated October 24, 1980 2saa/-S- o-(' 7 Z-
4-4-'. .- '
I
-\ý OFF~DS ! f 0 CALCULA210H MANOAL (00CM)
A. The 02CN shall descrzitbthe methodology and parameters to be used in the calculation of off••te doses due to radioactAie gaseous and liquid effluents and in the "calculation of gaseous and effluents monitorinq instrumentation alar/triv set•oins/Aons ea w•t
ap ISe s08 ntaine -s-e••o echnica \
2.( • e 0 shall aaiatai 4 at theo lant an shall rbglect epted bodlo s and ca lation PC edutes.
Asendtent Ho. 93 28I iDjsrts~ T8b,
-n
ý:A 0e 4 22
REVISION H
'4
A. The PCP Shall bt 8 MAnUa1 Conta/inin operational informatict concerninlg the solidification of radioactive wastes from
liquid systems.
1. The PM shall be maintained at the plant COnsiAtentt wi± these ?echanical specifications and vith approved plant procedures.
C. Revisions of the PCP:
1. shall be submitted to the Commission in the Semiannual Radioactive fgflueAt Release Report fat the period in vwhch the revisions were sade effective. This submitta shall contata:
a. sufficiently detailed information to support the :ationale ft: the revisions without benefit of additional information:
b.-._a determination that the revision did not reduce the Overall contformance Of the Solidified waste product to existing criteria far solid wastes; and
c. documenta:tio that the revision has been reviewed and found acctptaole by the PORC.
2. shall become effective upon issue following review and acceptance by the PMRC.
6.16 PROCESS CONTROL PROGRAM (PCP)
A. The PCP shall be a manual containing operational informatic concerning the solidification of radioactive wastes from liquid systems.
B. The PCP shall be maintained at the plant consistent with these Technical Specifications and with approved plant procedures.
C. Revisions of the PCP: I
1. shall be submitted to the Commi~ssion in the Semiannual
Radioactive Effluent Release Report for the period in which the revisions were made effective. This submitta shall contain:
a. sufficiently detailed information to support the rationale for the revisions without benefit of additional information;
b.__a deterination that the revision did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
c. documentation that the revision has been reviewed and found acceptable by the PORC.
2. shall become effective upon issue following review / and acceptance by the PORC.
I./6.17 OPSIT7 DOSE CALCULATION MANUAL (ODCM) l
/ A.The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and effluents monitoring
n
I
insrumntaionaiarmitrip setpoints cnsistent wita tbe applicable LCOs contained in these Technical Specifications.
B. The ODCM shall be maintained at the plant and shall reflect accepted methodologies and calculational procedures.
kmeadment No. 93 239b .25.>2A'••
12&- Z I
A
I
- . .-- -
See-�o.y� 5:
1C Revisions of the ODCM: C (,.E,. * ,
[•~. c.3J 1. shall be submitted to the Cotm!issio te %)RDRoc:x Effluent Release Report for the period in which the revisions were
ZM5 .a .Z 7- made in eaf ontaion: Q~.u 1 .•8E a. sufficiently detailed informat~on to support the ration~ale !or
! enera tT auitizonaVii nor rmati ted shall 4bnsist of rlised page (of the
re numbere and Providod with an Fproval ,her with appropriate evaluations ustifying
b. a determiatýion tl
t ' .c.J,
6.18 MAJOR MODIFICATIONS TO RADIOACTIVE LIQUID. GASESOUS AND SOLID WASTE TREATEMENT SYSTEMS*
A. Major modifications to radioactive waste systems (liquid, gaseous and solid):
1. shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the modifications is completed and made operational. The discussion of each modification shall contain:
a. a summary of the evaluation that led to the .- determination that, the modification could made in
accordance with 10 CFR 50.59;
b. sufficient information to support the reason for the modification without benefit of additional or supplemental information; and
C. a description of the equipment, components and processes involved and the interfaces with other plant systems.
- Entergy Nuclear Operations, Inc. my elect to submit the information called
for in this Specification as f the annual CFR SO.S9 Safety Evaluation rt.
Amentdmen xo. oaya 2682Sc€."
REV*SION 4 H4.
REVISION H-
V
'xi -S
C. Revisions f t ODOf th,0
1. shall be submitted to the Commission in the Semiannual Radioactive Effluent Relese Report for the period in which the revisions were made effective. This submittal bs5.ll contain,
a. sufficiently detailed information to support the rationale for the revisions without besefit of additional information (informatioA submitted shall consist of revised pages of the ODOE, with each page numbered and provided with an approval and date box, together with appropriate evaluations usAtifyinq the revisions)j
b. a doeermination that the revisions will not roduce the accuracy or reliability of dose calculations or setpoint
c. docuentation that the revisions have been reviewed and 2 fo und a cc ep tab le by th e OI• 1 .M .
s a become e ffect ive upn I ssue fol llo "vJ~nq ýrelvevi* a mndd
.18 ML7OI MWI!F!UTTOffS VO uADoAk VE L!OU7!D. G)AROI AMD SOLID ' 1 L
A. or modifications to radioactive waste sytems liquid, gaseous and olid)s
1. 1 be reported to the Commission in the Semiannual Radi otive Sff luent Roel e Report for the per in which the
a. a- Is
modifi tioa is completed mad operational. discussion
dt\no ta te it od��I be m ade i f e
aoaach with 10 c u5 0. 5
b. eauhficaie n isoratio cc I tterao o h
de if aw o b o a mupplema inormtinan
a. a dscmmripio of the eauioentht copoensad 0/h
0pr o e s se s anv o l a n d t h e fit e r fa • e •s c h ob e r p l a n t
b. nhr myeett su bmiti informatioa oa r teralled for i this
Spe cificat io n as pa st of ti atnu 1 Cb f i of* S a fety dii oal ao rI
Ameadenta So. % 127
•II: 251cft f i
I rt OcF 2
em
'2. The following evaluations shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report, where such evaluations are vequired to be performed in order to assure compliano' with the requirements Of 10 CFR 50.59:
a. an evaluation of the modific tion, vk c sh s tjhe ~ \, predicted-releases of radioac ve materials i
liquid and~qaseous effluents and/or quanStiy of solid vaste that differ from th previously \ predicted in the license applicat n and amendments
thereto.
b.\'an evaluation of\the modification, % h shows expected *maximum exposures o individuals in the UNRSTRICTED AREA anA to the general pop ation that differ from those, previously estimati in the license application and.amendments thereto; nd
c. a comparison of the pr*ditted :elease o: rado`aý.cive \ materia'1s, in liquid and q&-e"ous effluents and in
solid was-to, to the actual riaeases for the period prior to ven the modifications are to be made.
Amendment .4o. 93 258d
Ia9 e /9--J -Z.
LP
(s. �D-
3. MSIV leakages raft e a tanc• criteria is I 1ý1.5 cdfh foreaach MSlV
in tho4 Pr •ontC Leakagje Rate T stingP•
The pRos of are aic a toftpmary Containment
Leakagle Rafts Testin rg ra.A
__4
SP4A.:fýi C0.tUO4h .5 6,111 be, C&~ &-e t* rnodt-ry-tkc te$V~.j e4e I-e14* >J vei by /0 C.'t S- A-pe..~ e .X
Tm-je 84 ci, -7
REVISION H
JAFNPP e drv%
POSTACCIDENT SAMPLING PROGRAM L
capability to obtam and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:
A) Training of personnel,
9) Procedures for sampling and analysis,
C) Provisions for maintenance of sampling and analysis
> PRIMARY CNANMENT FAArRAET m M rk P'JAw
ontainment as required by 10 PR 50.54 (o) and 10 CFR 50, Appendix J, Option 9, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, -Per•ormance-Based Containment Leak-Test Program*. dated September 1M. as modfWed by the exception that Type C testing of valves not isolable from the containment fre air space may be accomplished by pressurizato in the reverse direction provided that tesing in this manner provides equwialent or more conservative results than testing in the accident direction. If potential atmospherc lake pat (e.g.. valve stem packing) are not subjected to test presaure, th portions of the valve not exposed to tes p.esure shall be subjected to leakage rate measurement du•ng regularly scheduled Type A testing. A list of these
valves, the leakage rate measurement method, and the acceptance crtena, shall be containei in the Program.
f The peak Primary Containment internal pressure for t~he desig basi loss of coolant accldent (P). is 45 psig.
( I. The maximum alwable Primary Containmnt leakage rate (L.). at P=, shall be 1.5% of primary conanmetm ar weight per day.
(• The leakage rate acceptance criteria are:
1. Primary containmenrt leakage rafte acceptance criteria is < 1.0 L.. During unit startup following testi in accordance with tiFis program. the leakage rate acceptane crtena am I .W L, for the ype B and Ty C tests and I 0.75 L for the Type A tests:
2. Airlock test• acceptance criteria are:
a. Overal airlock leakage rate is 1 0.05 L, when tesWed at Z P,
b. For each door seal, leakage rate is : 120 scfd when tested at Z Pa.
JAFNPP
[.,'.1 3._3v CONFiGURATION RISK MANAGEMENT PROGRA&M L f•. v• M
The - - k Uiinao das a proceduralized risk
inf~x, mon~d- ant to manage tHe n associated with equipment inoperability.
The program applies to technical specification structures. systems, or components
for which a risk-informad allowed outage time has been granted. The program is to
include the following:
a. Provisions for the control and implementation of a Level 1 at-power internal
events PRA-informed methodology. The assessment is to be capable of evaluating the applicable plant configuration.
b. Provisions for performidg an assessment prior to entering the plant conflnio described by the Limiting Conditons for Operation IL.C O
C. Provisions for performing an assessment after entering the plant configuration decie b teLCO for unplanned entry
into the LCO d . •
d. Provisions for assessing the need for additional actions after the discovery of additional equipment-out-of-servi'e conditions while in the plant
configuration described by the LCO
0. Provisions for considering other applicable risk-significant contributors such e. as Level 2 issues and external events, qualitatively or quantitatively. • S-.>6ov.Je_
The progiram provides controls for potentiany explosive gas mixtures contained in
the Main Condenser Offgas Treartment System.re
ooram so include J nIts for concentration of hydrogen and oxygen in
the Main Condenser Offgas Treatment System and a surveillance program to ensure
rt-o thelimits are maintained. Such limits shall be appropriate to the system's design criteria (i.e.. whether or not the system is designed to withstand a hydrogen
explosion). . o 3. 3
The provisions of( 1j O.nd 4.0. are applicable to th ogram
furv a i elan c ) %q u e n cies . ? G•-"-k
-, ,t%4 0..• - .• 7 - -0,.
AmedmntNo. 1W 70 2;
2581 REVISION H
6
0 No
fla
oic
1ýR, o-+ - ' -S. 2-
c7IA
e STA posiin may be c iibined filled by/ady individual r(eeting th
ccordanq6 with Section . .2.
uring startu r planned shut dwn; both in Cont I Room.
SRO) - Liccised Senior perator L
I ]SA) Sj~hf Technic)I Advisor \(RO) - )6censed Reactor Operator)
Amendment No. 444,268 260a
REVISION H
4W -S.
X-
(,
)T5
I Reactor ensure Vessel M4 sal Startup (!00°i/I } ................. 120
? Ro, rh:. . ..... ............. 400
STetoL" Trip at... 251 Power...... 40
\~da .latr \y~d •. ..... ....... ,......... 7 L\ a of e wate Pumps.._......: ... 060-60..... 10, Tubn Ge atr ri . l. ..... .oo . . ... 40•
S\ \ ~ ~ r s,,,]e • .... ..... ... ..... s S • '\.,.sf..y vah ve ool do wn()O I F .......... .. ,. 2I
Al-O he S ra .... _.............o...... ".. o. \ 47 rj S a
m
ca~ 0=1
# " 3 ,7 :. - • / • .e . <-
PAGE INTENTIONALLY LEFT BLANK
\ N
Pages through 264 were deleted by Amendment . 37
The next page is Page 285
265 - 284Amendment No. 22
THIS
NOTE:
Pa Iv2 27- o" 2z2-
'/ý" I A (ý9
Ct - ,\ J-41-:01il-_ýCCAet-I IE-
<Z;Dose Eq~ian 1-3 0 concentration of 1-131 (microcuries/gramm) alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in International Commission on Radiological Protection Publication 30 (ICRP-30), "Limits ýf take by Workers" or in NRC Regulatory Guide 1.109, RevMP l4 0
1977.
S •eex A Techn~a cifications.
See Appendix chnical Specifications.
See ix A Technical Spec.i"cations.
See Apppendix A Technical Specificatio
F . M em b er s( s ) o f t 4 1 i Member(s) o-,(he Public include all persons who are not / occupat'QI lly associated vi the facilities on the-EntergNuer Fitz arick, LLC (ENF)/( Niagara Mohawk Power CorT tion site.
Is category does not clude employees of the compaed s, its - contractors or vendors. Also excluded from this catlegory are persons who enter the site a service equipment orto deliveries. This category does ildepersons who use porti of the site for recreational, occupational, or other pup s not associated with the
G.MotAae The 0 asTreatment system is esystem designed an nta, ed to: r radioactive gaseous a uents by collectin p~rima= coolant * te" of fgases from the condenser; and, provie or delay of
of fgas for the e of reducing the total ra ctvity prior to release: to the en onment. 4K21i
The 0001 describes the methodoloY and Parameters to be used in the calculation of of fiite doses due to radioactive gaseous and liquid 3I effluents and in the calculation of gaseous and liquid effluents monitoring instrumentation alarm/trip set points and in thecodt
t environmental monitoring program.
0endment No.g oa- _&r268
vcýC.. 1 O 1J., REVISION D
0 I
<0
..Y,
Am
0)1
C. Rnf l h L O G 5 Ž F F LEQ
the quantity and isotopic mixture of 1-131, 1-132, 1-133 1-135 actually present. The thyroid dose conversion fac
yrord dose as
this calculation shall be those listed in International Commission on Radiological Protection Publication 30 (ICRP-30), "Limits for Intake by Workers" or in NRC Regulatory Guide 1.109, Revision 1. October 1977.
B. Instrument Qh~amel calihration
See Appendix A Technical Specifications.
C. 1intm•nt rhtin.1 •,n xp1 Test
See Appendix A Technical Specifications.
D. TnStl-ufftot Check See Appendix A Technical Specifications.
E. Looiegyjto Funetion Teet See Apppendix A Technical Specifications.
F. Memberg(n) of rhe Pubticir Member(s) of the Public includes all persons who are not occupationally associated with the facilities on the .Entergy Nuclear FitzPatrick, LLC (ENF)/(NMPC) Niagara Mohawk Power Corporation site. This category does not include employees of the companies, its contractors or vendors. Also excluded from this category are personswho enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plants.
G. Ofaau Treaem~yt Svmtem
The Offgas Treatment System is the system designed and installed to: reduce radioactive gaseous effluents by collecting primary coolant system offgasas from the main condenser; and, providing for delay of the offgas for the purpose of reducing the total radioactivity prior to release to the environment. ,
SOf.flAto Bom t1a Ilatiem Manual (OalC)The COCK describes the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluents monitoring instrumentation alarm/trip set points and in the conduct of theaenvironmental monitoring program.
See Appendix A Technical Specifications.1
Amendment noe. n*y. akr,268I
~&je. 1422
REVISION H
I
I
l
r
"i ' iv,
-,Tic-,
o's
7p[,.e>'. '5-4 1 -OD-
(Z Ktý,r
J . ,:o-s Control ,Pr c ,, •.. . - •- "/ [ •Ra PCP is a do --ent which identifie-e a urn omls sampling
t hese act-itviies in such a -#• as to assure compl~iance with 10 ýCFR 20,-: 10
CFR .6 ,010 CFR 71 and othe applicable regulatory requirements goyerning
issoa fter atv waste. K/. :ae h Power/Apendix A Technical "peciications
Th siteB oundary i ta ln . 5cc.r__
or otherwise controlled by ENP, ENO or N1MPC. Refer to Figure S.1-1 fh ieBudr sta ie eodwihteln snt o rnd lea ed
!the map of the site boundary with regard to liquid and gaseous releases.
M olidifido
o tlid .onin of we s
N. .espoo / // "AnSourestickd arhe quallf aon assessmena of channel resi se whena
ofe cdiaduals froms exposre to raistourc aencesd radioatvemtrialiotay
Are as thinhe sitec bdday usedue foresidcentraia ~ q radtiractrifo
mat ial per unme ,•a reiestsed ti end/irorrere This inonudel supoe pcoesses as filtra :(on, evaporati F/condensation° .•tling/decanting, a~d
P / An U enitoynrestritede ed area ued i beyond the e buadoi
/ 0.3(a)s(17)/Te• unrestricted., •ea bondr may conside with, the / unr:est i dr os not coilled a Mr ea for purpodes. oTet4•ncpton of indi e sare frm exp ie to datio be•yond raeioactive maeunal, or anyi atrea i•hin the site b oundary used for pesidenil quartersop r folro nd riaý,•-commerýýA, institutinrAl and/or recreationaVipurposes.
X/Effluent Techni-al SpecificAt~inn •has been expanded er that in 10 ••
.P/ 20.3(a) (17) .,O*ne unrestricted ras boundary may cc,'cide with the/
exclusion Miy nced) area turzto0 7. dCPR 100.3(a), /lhe
Amen~t do. Me, 268
Ct
2
1ýOR-~L,
REVISION D
I,�i
"PC4,._Y
C. �\4�j�f' LU')
J. Process Control Program (PCP)The PCP is a document which identifies the current formu.las, sampln; methods, analyses, tests, and determinations used to control the processing and packaging of solid radioactive wastes. The PCP controls these activities in such a way as to assure compliance with 10 CFR 20, 10 CFR.61, 10 CFR 71 and other applicable regulatory requirements governing, the disposal of the radioactive waste.
K. Rated Thermal Power ýSee Rated Power, Appendix A Technical Specifications.
[The Site Bounhdario thi r Tbeyond hric th-e =and __e __tre b EO or'N?4!iRefr t
the map of the site bona [iý' ~ ,jA.
M. cation__ _____________
soliduifiain is tht converin~w~wse noa~r a meets shipping and burial ground requirements.
N. Source Check A Source Check is the qualification assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
0. Treatment Any process which effectively reduces the concentration of radioactive material per unit measure released to the environment. This includes such processes as filtration, evaporation/condensation, settling/decanting, and solidification.
P. Unrestricted Area An unrestricted area shall be any area at or beyond the site boundary access to which is not controlled by EKF or ENO for purposes of protection of individuals from exposure to radiation and radioactive material, or any area within the site boundary used for residential quarters or for industrial,-commercial, institutional, and/or recreational purposes.
The definition of unrestricted area used in implementing the Radiological Effluent Technical Specifications has been expanded over that in 10 CFR 20.3(a) (I?). The unrestricted area boundary may coincide with the
exclusion (fenced) area boundary, as defined in 10 CFR 100.3(a), but the unrestricted area does not include areas over water bodies. The concept Sof unrestricted areas, established at or beyond the site boundary, is
utilized in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant to 10 CFR 50.36a. -
Amendment Mo.....2...... ,.. : No. ea./
2 -
REVISION D
N\
ýQ
C- S 2_. I
lI.ITING CONDITION FOR OPERATION SURVtILLANCE REqUIREmENTS
2.
2.
Amendment No. 93
OLI ID EFFLUENT
I MqID zFFLu MopITORS 2.1 LI E/FFLUENT OMNITORS
Ay abiaii A picability
Applies to netrumontetion required tot maui- A lies to Instrumentation for monitoring ratortin red active liquid effluent discharges d oactive liquid effluent discharges.
chatres are properly monitored and recorded sdioctivo liquid effluent discharges are into theenv romest 0 specfiedctfTable2.1-1
a. T e li it ine8 co nd iti~ o for oper at ion o f the a . The al am /tr ip set points of th ele. h an na1 netruments tha t montor radioactive liquid efhall be detehmIned end adjusted fn ccor
Iffluente are given in Table 2.1-1. Nith a• dance with the methodology and pal meters in redios oactive liquid effluent onitoring ing the OffIeit Does Clculation HanI (ODCN).
d trumontatios channel alarm/trip set poin
lpees conservtive than required by the sp
wivth out d o lay su spe nd t h release of r ad -" / active liquid t bfluents onrimtoed by
n '
a ffected channel, or declare the channel noperable, or change the set point so i is
a cceptably conservat ve. /
a. a l ith lei than the minis uo nu mber of tdi- b. The su• oellance requ re p nts of the radio
S active liquid effluent monitoring insatmen- active liquid effluent nttoring Inetruen
S taioin channels operable, take the aection tation is ehorn on Tebl 3.10-2.
s hown In Table 2. 1-I. Take correct ye ac
tions to retu~rn the Instruments to o erable
3
Q{ �
/
2\
t
Amendment No. 93
RF-TS P, IC-T�
)
4
I
Fcr5 I.i)-,
Gross radio,• vity monitors providing alarm and automa c termination of release
Li d radvaste effluent line (a)
Gre beta or gama radioactivity monitors p ovidiaq alarm but ne•t providing automati eruieatioe of release
Servieo water system effluent I (b)
Flow rate measurement devices
Liquid radvaste efflue. liae I (l)
(a) Wi th the a r of operable channels less than the required minimum number, of ueat releases may continue provided that prier to initiati a releases
a. Two adepnadent samples are aaysAyods
b. technically qualified members of the facility staff verify the discharge line valvingi
Otherwise, suspend release of radioactive effluents via this pa
(b) With the number operable of channels lose then the requir minimum number, effluent releases in this pathway nay continue pr ided that, at least once per 12 hours, grab samples are collect and analysed for principal game emitters at a limit of detection. at least Sz 0 7 mnicrocurieeesol. The principal gams emit a for which the LL specificatiea apples exclusively are desce is Doe (C) to Table 2. 3-.
(c) with the amber of operable channels less the required minimum number, effluent releases via this pan may continue provided the flow rate Is estimated at least Once r four bouws duwing actual releases. P" curves or task level creases generated Is siti may be used to estimate flow.
Amenduent go. # 127
cT.S MT73 A
LINITINC CONDITIONS ptOPERAT'ION /
Amendment go. 93
2.2 CONCINTRATI OF LIQUID EFFLUENTS 2.2 CONCENTRATION OF LIQ /DEFFLUFNTS
App as to the concentration of ra cactive mp- Applies to th analysti of radioactive liquid t ials In liquid effluents. Weaste from a plant through a liquid pathway I
to an unres icted area.
To ensure that the coac trations of radieac- To a 1ur that analyses are performed and contive materials In liqui effluents ore kept to ce ation determined for radioactive liquid acceptable levels. / .ae.
Specifications Specifications
a. The concentra on of radioactive materials a. Radioactive liquid wastes shall be sampled released to the unrestricted areas ehali/and analysed according to the sampling and not eaceed a values specified In 10 CF analyses program of Table 2.2-1. 20, Appen x B. Table 11, Columi 2. F r dissolved or entralied noble gases he concentr tion shall be limited to 2 O-ta
b. Wit the concentration of radloact a mate- b. The results of the radioacti ty analyses ras released from the plant to u estricted shall be used in accordance u h the methods ar a• exceeding the above lii a. restore In the 000 to a@sure tha the concentrathe concentration to within t above Its- tione at the point of rel so are maintained its or terminate the release. within the limits of Sp ification 2.2.s.
7/
oh",as
atto
C-FS fý eTs 21 Z
of
NOM9,c-r FOR'T TAE2
()TeLLD (Loth eari Loeimit of dete ction , is deinedo abrose fi tihro
sypeiicaltaoes. of IV.I.ad ll t do n ioe of rain acthe e acu ation. i
a: saml dha b rc nised ta nt con bovD sysefied bas k tha Villr ar thdetact)d liith 5eprobui, aabilityan t ofl 5a meoasrmety of falsdl
mixeldil tohsue aep blnkav o soramrpl esntng.ral
For ah parincial mesuemettr stor (wihithe mL aiia tionud rappies ez
LL sthes aui are towber demtectd adton r s otd tefi paksted abv (imcromeauribes p d idetii abe taso olu ethrv)h;bv ucies las
be bde adad reportedion athe S ackgroudicoactine rfflentr theles Rapore. o o a blan so oup9 e-l, as aprorate (in zouns er.t)
Imen ien o. nin 93ica (i 8onsprdsnerto)
cf~rES
S(d) A compositreample i.s one. I which the quantlt7 of 11i ~d ampled is proportionm to0 the quatnctl of liquid vast*e discharge and in• which the
method o sampling employ results in a specimen t is representative of the quids released
(4) A c inuous release s the discharge of Uqui wastes of a aondiscrat vol *, e.g., from volume of a system that: an Input flow during e c tinu'oU:s rTeleas24
(f) o be represean t•ve of the quantities concentrations of ra oactive materials in quid effluents, samples 11 be collected coat nously in proportion the rate of flow of the fhluent streum. Prior a analyses, all sampl taken for the composite shall bet thoroughly ad in order for the c site ample to be repr sentative of the effl 1 t release.
Ame•dment No. 93 9
Pa It 4,9 f ý
C Ts F,-7-:.'
LIUITING CONDITIONS FOR OP1 ION S0VEILLANCE REQUIUIIEITS
Amendment No. 93 10
/,71lz
Z,3
2.3 POSK nOM LIQID EFFLU .3 POSE FROM LIQUID EFFLUENT Applicability, A /pplicability,
Applies to rad ties domes frno liquid a U- Applies to the c culation of the radiation eats containi radioactive materials. dome from liquid (fluents containing radiesc
tivo materials.
To as ro that the dose limita one of 10 CPR To .nlur thet the radiation dome from adloac50. pp.ndix I for liquids ore at. tive 1 uld effluents is deter.ined.
S// . .... ,s ifications
a. The dose to a member f te public (rm ra- . Cumulative dose contributi a from liquid' dioactive material releaaed fros the plant effluents shall be deter t ed In accordance in liquid efflue a to unrestricted areas with the O0-0 at least thly for the curshall be limited as followst rent calendar quarter a the current eaten
dar year. 1. Duiring an calendar quarter, limited to
less t or equal to 3.5 area to t whole dy and to less then or equal a 5 ore to any organl and,
2. Do ng any calendar year. limited loes t anor equal to 3 remtto thew abody nd to lass than or equal to I orem to
b With the calculated dose from e release of :rkdioactive materials in i 1id effluents exceeding any of the above 1 mits, the tolloving shall be donae
\
LIMITING CONDITIONS OR OPERATION SURM)I CE RRQUIREHENTS
Id ientify the causes for a cis releape rates;
3. :rep/re and submit a Fport toth R v,,hi. 30 days.. h
Amendment No. 93 I[
•?0e o Z _. _
rl- .
(7$ � j�L
I
Applies to t operability of radioactive iq- Does projections apply to 1quid effluents e
u lwd process a equipment.
leased to unrestricted are a.
Ob aectiv ye / Objactiv ye
To e••ns t liquid radus.t .tratme. system(@) To ensure 6hat actie levels to require operaare a rated to prevent exceeding a dose ila- tion of waite treat nt systems are determined. its,/ Specification 2.3.
SP .. .. , / ns Specifications
The liquid radioactive w to treatment eye- a. Doses to ndividuals in unreatricted areas tem shall be used when the projected dose due to quid releases shall be projected at from untreated liquid elesses, over a 31 least nthly In accordance with ODCH. day period, to a meab of the public would. exceed:
S0.06 ures to t whole bodyl or.
2. 0.2 ures to y organ.
b. With radloac yve liquid waste Seolo diecharged in c•me of the above limits, prepare and n sit to the Coamission within 30 days a r part that includes the following I n f o rm s ! O ak:,•
\ ,. /x nto flqidrdat a en
Amendment No. 93 12
/
CTS -
LIHITI
2.
4.
ING CODITIONS FOR OPE 5ION
---------- -/
SuRVI maI sEquih34EuTs
Identification of any Inoperable equipmeat or aubsyste and fhe reason for the Inoperability/
Action(s) tak to restore the inoperable equipment to parable statmel mad
Summary do ription of action(s) taken to prevent a ocurresce.
Amendment No. 93
Fq7f -*
LIHITINO CONDITIONS FOR OPERATION SURVEILLANCE EEQUIRFMENTS
Q��Ifl1Nm1DETAWS
licabi it
Applies to tanks located outdoors that do not have catch basins that drain back to the buildiog.
release of the task's contents. the resulting concentrations Iould be l1se than the limits
of 10 CYR 20. Appendix 8, Table IIl Column 2a
at the nearest surface water supply in an unrestricted area.
ve I on
The quantity of liquid radioactive material
contained in a condensate storage tank or any outside temporary tank shall be limited
to. 10 curies, excluding Tritium and dis
solved or entralundloble a
Vith the qupntity of lffId rndioactivj ma
Sterra1 ihn tank obo( this unit. dedue the tankb radioc~*'e contents w ithin
the lim$ within 4 houral and /
The quantity of radioactive material contained In a condensate storage tank or any
outside temporary tank shall be determined
by analysing a liquid sample of the tank's
contents w• • y enadoac qu Is aena a- d to the rnk.
joyts 4ý 3R '10.2- %m4 S7Z 3.
~V j,,Ply*4 rvI(LV"ce e~.4C .!C ;S a
Amendment No. 93 14 I I ~~TV- 5
REVISION H
V �
t[S-. 5,6D
63. ~
4q
fI-,.
'4
SURVNILLANCZ REqUIRDIENTSLIlNITING. CONDITIONS FOR OPERATION
JA, .. PP
I
Amendment No. -Ba, 213
-� -�
17
(~ra.,F
LIMITING CONDITIONS ROPERATION SUR7 ýNCE REQUIREMENTS
3.0 GASBEOUS EFFLUENTS A
These requirements pply to instauments which monitor the Ap es to instrumentation listed in Specification 3.1.o and gaseous redi y effluent pathways to the environment. ayses of gaseous effluent releases.
To assure t t radioactive gaseous effluent discharges are To ensure that instrumentation required for ga us effluent propedy tored and recorded during release. releases is maintained and calibrated and the ioactivity of
gaseous releases; is determined.
a. Radioactive gaseous wastes released to the environm t a. Operation of the gaseous of t monitors end recorders via the below listed pathways shall be monitored listed in Specification 3.1. 1shal be verified by recorded during release from the respective pathwa. performing instrument a uillance as specified on
Table 3.10-2. 1 . Main stock exhaust /
2. Refuel floor exhaust
3. Reactor building exhaust
4. Turbine building exhaust
Snd•rn•tA haildinn Axhaust
tr-rs /Te r5 '!).t
?,,,, ý, / 0-f
7-.5 /5 67:3
be sample or iodine and pprticulato radio filiteir for a pathway listed In Specif 1activit an a continuous 'basis during -cstloý 3.1. shall be changed out at lea Jesse FRo the respective pathway. ky
c. If pecifications 3.1.a and b.. ob a. Can- C. Crab a mpIes, when required, shall Colt be met. eff luent releases In" continue I ct d at least once per 12 hours nd one
via the respective pathway prow, ad gaseous ly d within 24 hours of collecti n. Auxilgrab~~~~~~~~~~ smlsreclctdi acaeoaipy samplers shall run contin usly and be
monitor out of service and auxiliary semi- hanged out at least weekly. piers are used in C .ase a prticulate or Iodine sampler is out of as ice;
1. Return the inutru t to operable status
2. Provide an ex anation in the next Semint' snnual Nadi ctive Release Report as o bhy the I parability was not curve ad
d. Alarm/tr set points shall be dota mned In accord ce with the COCI mand met o ensure that me limits of Specification .2 are not
ax d1 e d. Wi1th a radioactive ameous at
, I nt monitoring Instrument ion channel a arnltrip set point less to ervative than
aqired by the above specif stiont
1. Wlithout delay. suspen the release of radioactive gaseous fluent. monitored by the affected than 1; or
2. Declare the chenne inoperablej or
3. Change time set p at so It is acceptably ,conservativye.I
Amendment No. 93 is
Fl~r0 2-of
(
I *
CTS 7? R-FS
"4 LIMITING CONDITIONS FOR OPERATION /
3.2 GASEOUS DOSE RATES
Applicability
Applies to the radiat dome from radioactive material in gaseous fluent. from the plant.
Objective
To ensure at the dome rates at or beyond the site bou ary from gaseous effluents do not exceed e annual dose limits of 10 CYR 20 for unres cted areas.
S cifications
a. The dose rate at or beyond the site boundary due to radioactive "mterills released from the plant in gaseous effluents @hall be lImited as follows:
I. 5500 mrealyear to the whole body an S3000 mrem/yeqr to the skin from no a gases; and,
2. £1500 wren/year to any organ f£a Iodine131, Iodine-133. Tritlum an or radiosctive materials in parti ate form with halt-lives greater th 8 days (inhalation pathiway only).
b. With the dose r (a) exceeding the above limits. withoue delay restore the release rate to within the above limits.
Amendment No. 93
SU9IEILLANCE REQUIRHI ME"S
3.2 GASEOUS DOSE RATES
Applicability
Applies to the calculation of the dose ra a from radioactive materials in gaseous eft nts from the plant.
Objective
To ensure that appropriotio calculations are performed to determin the dose raten from gaseous effluents fr the plant.
a. Th one rate due to noble gases in gaseous fluenta shall be determined to be contin
uously within the limits of Specification 3.2.&. In accordance with the methods and procedures of the ODCI.
b. The dose rat us to Iodine-131. lodine-133, Tritim a to radionucldes In particulate form wi hall-lives greater than 8 days In
I
7 y / /" %
,/
3,Z..
)
19 ýrV-JS
Ii
cfr: 9(13 3z
gammons effentm. ihall be determined to within tj� 'above lmita in accordance the iM ad and procedures of the 0CH1 vill •e done by obtaining repre titj mas'em and performing enalyses n acec dtce vith the sampling and en sem progi specified In Table 3.2-1.
Amendment No. 93 20
72q Ca-e 0C- -
(1TkCrD
t
OF-S R : i 5r -3, ,
Main Stack and
Refuel Floor Vent
and
Reactor I
Vent and Turbine I
Vent n
Radas. Uuildinq event /S
le6ble 0en08 Gross $eta, or Gamma
PtIaeipal cam,
I I
p jr,5-1
TOLI C ý 0ý ý
61-5 RF, 75
)
TIOTZS TOR TA3LE 3.2-1! ()The LLD. is defined. for purpose of eso specifcations, as the
eoucantracion of radioactive mateFa t: 7altample that€ yL1 y.,
ctunt (above system background) hat will be detected with 95Z p Icy and with only 52 probab ty of falsely concludi•g that a b serrvadion representsa "re s signal.
For a particular mesa t . system (which say i.clu radio separation):
4.66%
smallest ld. a at troil1
•k ob
chemical
-*V e2.22 x10 *T aeap t) Where: LID is the riori lower lizmt of detec , as definad above (in micro
cuispýrats missorute);o t,- a-rd deviati of n backgound counting rate or of the 5un./ g rte, of a blank ample,• apqopec,4 (in counts per :•:•t.);
I tecounting efcec onsprdsn~rto)
istesample size (in ta ofSaso ou)
2.= ~ L g 06i he er of disintegrations Per Snt e i Cro
T is the frae: radlochsinal, yiJeld (when applicable
I is the tive 4e4y constant for the patst radionuclide; and
at for ;I effluents Is the elapsed time be the midpoint of sample .Collect sand time of counting.
Typi values of 2. V, T, sad at should e used in the calculation.
I should be ecogniaed that the ise defined as an a iorio• (before *-.faes) liat bepresenting thea ability of a messaun. oent system and
not as 0ean a "Steisiari (aft the fas) limit for a particular
(b) The principal gn mitt fo r which the LLD specification applies xelusively arm the fall g tadiomiuclides: "z-7, Kr-8S, Ze-133, X 33m. Ks-15. Z.-3Ma, and U3S far gaseous emissions; and. ft-54 4-59. CO-(86 Ca-4O, ZU-45 99, Ca-134, and Cs-L37 for particulate sains. This list does no, that only these nuclides are to be acted and "rep-rtGd• Other -Oak that are measurable and identif e, together with the above clides, sha.Ll also be identified and orted in the Semiannual oactive Effluent Release Report. The for Ino-99, Ca-L4l, end I"1 La 5XlO-11. For oil samples, Tab 2.2-1, Note (c)
(Cc) The rsti Of the sample flow rate to the sampled tream flow rate shall be far the tize period covered by each do or doese rate calculatamn de In accordance with specifications. is determination shall be made using design flay rates if fl~jy mete are not provided or are
?GAR q 4 17/
(7-5 3, 7,_,
r/ ~ ~NOTES FOR T LE -1(continued)
(d) Main stack gaseous sampling and a ysis shall also be performed foill ng shutdown, startup, or a thermal p or change exceeding 20% of r d thermal power in one hour.
1. This requirement appli only if:
"h Anslysis showt hat the dose equivalent 1-131 ce ntration in the prima coolant hu cre.sed more then a factor of and
"• The no• gas monitor shows that offlus alrctivity has increased more than a facto i1f 3; and ge•
"• •rrection for increases due to ch gsin thermal power levj have been i ado in both cases. /
"() M tc odn n ariuae pin hl.ls ep d daily following / ower in one hour.
1. Daily sampling is not r uired for thermal power nges if the off gas charcoal filters are in service. /
2. In addition, this uirement applies only i.
" Analysis ows that the dose aqum lent 1-131 concentratioin the primary coolant, increased more tha factor of 3; and a/
"f Tnhierae a be dischre voiao shoohewta t h effluestactiiy a anr3d buildihnga
fe o.. .3; mon. "* or-e--.- for ince u to changes in therrI power level have been I /made in both cases./ •
3 Daily sampling shoil perfo)r ed until two aconsecuteve samplen show gre i ncrease in cone an•:J but not to eoceed 7 'onsecutive days./
4. L. , ma be . ,. sod, byafcor ,of 0faaysi of, dail ,.,, @... 5. Analysis of do*•l and weekly sampls• •/11 be completed withi ¢48 hours of
(f) Incinerated oi may be discharged viaii ~ ts other then the m ln stack and building , vents. (i.e.,•~iir boiler). Release U@11 be accounted for llsod on pro-release grse sample d. /
(g) Samples of incinerated oil release shall be collected fr and representative of Sfiltered oil in liquid form. Whon r oil samples canno be filtered such as No. 6 bukrfulol, raw oi sample 11al be collectedod . rnelyzod.
Amendment No. $ 34. 203 23
Applies the air doese due / noble Iions to- / Applies to the calcul tfon of the air dose duo,
le ase d Fo s th e p lan t , . / to no b le g a s . ff lu e n
of &$sur that thea in&n do..mIe ntato- To ensure. t appropriae. c°,col-ti t. ar /sttO FR 0.Appndfl.oremet /performed to eterfnis the air dose tr• noble
1 C r T h e a p p do s 1 .a rea mt o e y dt. e r ,r ° C l tf ai pr d e o t t u o f r n b e
Speit.cahen OPCeco -h tonsen dgo th/clen clndr
I~ ~~~~~I tbtooda/ol n/
Ib ou,. nd ary m re ino l g ame s o le . 4f romn t h/e- se/h l e c l u a e e s o t l p e•lant In , so o s o f u n s A b in t d c o d n e w t the st..,, or. . the. cqn /
t~~~cl ne quarter and the' curren.ton calendar.]..• 1 D!n n aed r ur r ol s tha er
C
cr� �
LIHITING CONDITIONS 94I OPERATION SURWVILLAN
1. Identifies he causeWs for exceeding th limit(s); ad
2. Del iae the corrective.atin tht ae bean / ken to reduce the releases;
t nato b a oasr h abe ." abveldi/
Amendment No. 93
9 A3RETNS
25
�r�j
; 1, 0
e /-_5 A 8•:' 3f ,
LINITING CONDITIONS FOR OP'y*Alo" S$IFEIJ.LANCE REQUIRRHENTS
3.4 nose but TO IGOINE-131 1001MBN-I))- TRITIUH AND 3 DOSE DUE TO IODINE-133.lOUEi RADIONUJCLIDES IN ]PARTIP TE F WNDIOIUCLIDES IN TPARTICULAY VOR
A
Al
Ob
T"o
so
Sp
Repi cabSi ity Applcab.ility
plile to the c latiwe does trom Iodine-In3 Applies to the calc atnom of the dome due t dlne-133. Tri an. and radionuclides in p I- odiae-3131 Iodine 33. Tritium. and radlonuIculate form gasious effluent@. clides In patic at* forum in gaseous efilmI auto.
L15S.ive Obloctiva
tsaeura that the dose limitations 30 Cps To ensure hat appropriate calculat• a are ). App dim I, are met. performed to determine the done ON todlum-131 Iodine-.33. Tritiunm and radionuclides a particulate form.
afications !Pe ications
The dose to a member of t public at or a. nulativa dome contributions aill be celbeyond the site boundary roe lodine-ll* culetod at least monthly in ccordance with lodina-I33. Trittuan. and radionuclides in the O0CH for the current clender quarter particulate form with If-lives greater and the current calendar year. than A days released ram the plant in gaseous effluent& shall a litedi
I. During any cilenda quarter to loes than aor equal to 7.5 or m to any organ; and.t
2. During any calen ar year to leos then or equal to IS or& to any orgon.
3. sess than 0.1 of the limits of Specitication 3.4.a. and 3.4.a.2 an a result of burning contaminated oil.
Aen"eleant No. 93 26
/ a ý
I
?a51
C-4r -s /frr;
Aaendugwnt No. 93 2
I
27
r4ro
toe ('Ac h0,
MAIN CONDENSER STEAM JET AIR EJECTOR (SJAE)
, ' f-,,"/Ol~icability
&'•,,;•I•' . Applies to main condenser oa discharge rate for noble gases when the reactor Is In the rno, stbkpfhot standby or hot shuldown mode of operation and the SJAE Is In service.
/ 7 CO il•7
2. With the SJAE Monitor reading at 5,000 pCi/sec or greater, within 4 hours following an Increase of greater I
r- "a500/. (after factoring out Increases due to changos In thermal power level) in the nominal steady state fission gas release from the primary coolant.
0.11
Amendment No. W, X, 21128
?2"/46D-
!
3,7,65reC14-c to, *
JAFNPP
1311101$ FOR /ERATION
Except as spefled in 1. and . below, both SJAE system b. Operation
radiation monitors "h be able during reactor power be vodfiv operation. The trip time setting for closure of the specified
SJAE Isolation valve aell exceed 15 min.
1. A channel may be ih an inoperable status for
up to six hours periods of required surveilance wit placing the Trip System in the tripped condition vided the other OPERABLE channel Is monkt that Trip Function. Otherwise. In the event that of the two SJAE radiation
monitors is m or found to be Inoperable,
continued react power operation Is permissible
provided that t inoperable monitor is tripped.
2. Upon the losI f both SJAE system radiation
monitors, sit temporarily monitor radiation levels
at the SJAE initiate an orderly shutdown and have
the main at isolation valves closed within 8
hours. __
& IAmendment No. -". 2 4 9 29
UMITNOI CO
b.
I
ofhe SJAE radioactive offgos monitors shall
Sperfoming instrument surveillance as Table 3.10-2.
-p4,c 2o f
Revision B
$1.
c /S ,-r&
A~mdms.a No. yl' 127
30
(-RcT]5
?aa
(gcS 7-5 S. C
Amendent No. 9 12731 ?07 aZJ
£ (. 0Ju \ havim'0 I V ý
CAS.M (Si
rubo -. 1.5. I" lett1.11m coedwlefew e fe'ereatlem, are givem aSI Table 3.10-1.
Ammae~mmI no. 91 34
[l- (0)Revision Op
D. T 'hur
(!ilýl
~3 3,-P'-
b. It t limits of ble 3. 0-I to not t, the scuam urn 11 be i..
35
6727
/0
!ý rc i A'at ý Gv%
Pql.e /a ý
V b'(~f*1 Yý AIlt r( -5V,)S.pe.
I - -�
UNKITING CONDITIONS F03OR 0!TICK KVUULuCE mEQIR9m~41FIXTs
3.10 HAIN CONTROL NOONl VOOTILAT RADIATION HONITOR 3.10 HAIN CONTRkOL go/ VENTILATION RADIATION I4ONITO
Applicability Appilcb .IlIw
A p l i s o b e s e a c y a i s p p y r e t m n tA p l i s t e n s r u e n a t o n u h c h w o I o r
sysemto hemat Ptro ". hetoratromsplyar
operation are given1~7t 5~ 7 j~j....... ~ T i~mitin g c onditions for
on
At"
Amendment No. 93
requirements are
1171 3�3��? 'K�J
3'
IA.ZO "/�A444� �7
3,-3
Ariffut"VERwro,
if
3.5.6./
JAFNPP
lT� 210-1 1
JRAMNAT0 MOMdTOMh(G SYSTEM THAT OM~AEADP IOAVVTM
Minimum No. of Operable instrument Channels per Trio System
�Lp lie)
Trip Funcatiop 4d /9c/70 *07C/'I?0(
Refuel Area Exhaust Monitor. Reactor Biding Area Exhaust Monitors
Q), - SJAE Radiation Monitors
1 18l Turbine Building Exhaust Monitors
1 Ils Radwaate Building Exhaust Monitors
1k) Main Control Room Ventilation
1h) Mechanical Vacuum Pump Isolatimon
jjj'T S F MIA T/• a _: 3.I E l
(al A channel may be placed in an inopWeble status for up to six
System in the tipped condition provided the other OPERABLE € maintained..___
the ino"pdeve channel sha" be restored to operoabl status with
Uhl Tin l-vel sattni Is in accordance with the m•thod7/i'- pIoce
VTotal Number of Instrument Channels
S L el Provided by Design Action
<5o0.000 vcilsa• 2 () lb) 2 " '|i
lbi 2 (0l
s4 x 103 cpe I (g) •_,• Z13T
Power Background
oume during pwiode of required surveillance without placing the Trip i
channel is monitoring that Trip Function. that is, tri capbliyi
• ilathis would cause As• Tfin Function to oc•;• n those cases,
hin 24 hours, or t~he indicated actio_.n shadllb take-n
the94h-../Lf4I�7\ /
SAmendment No. 93, 1H1 . 2991. 249 'P--,3 c~ a 0 z(ý. R-67rjJ5 7REVISION F
it L .6
5;re C- I -A e4L 170&%r,, JL 3.3.4. 1
?'rjWWf14 15111-'h4l
:nrtvX4tzd'C ho k
Ta 6U C;ý -3, 3.
T, tWo -fA A
JAFNPP
L!a lr ý ORM SYSTEMS ir *NTIATE AND/OdISOIATE SVI(T;M) FRA-01,qjoN !ýM WO SYSTV-
Channels per Trip System
(a) A channel may be placed in an Inopeable status for up to six hours during period. of required swrvellance without placing the Trip
WK %4 filt. Systemr in the tripped conaditki provided the other OPIERABLE channel Is monitofrig that Trip Function, that is, trip capability is
t•cI~bLI 4• the •Inopble shall be estred to operable status wi;tOn 24 houre, or the indicat act sheil be taken. -- - -. •th e....
~(halsoe
Amendment No. 9,. 27.- 2 , 2 4 9 '?-,se- 7 j I S37
Revision.2'7
1)BigteSA ea rats below th4 mp level within 72 hours of isol-ate either " •A or all main steam fines within the next 12
%P 4f-~ 0 57JAFNPP
RAIATION MONITRINGSIYSTEMS THAT /TIATE ANDIII ISOLA•4 SYSTE
c May be...aced n--.... status for up to six hous during soF required surveillance without placing the T
System in the tripped condithm provided the other OPERABLE channel Is monitoring that Trip Function, that is, trip capability is
maintained. Ans lkpersbed cha.nel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these 4
the channel sleil be restored to operable status within 24 hours, or the Indicated action shal be taken.
Ib) Tai lve setting is in accordance with the mathoda and procedures of the ODCM. " . - r
Ic) Cease operation of the refueling equipment. J. z U.:
A& Amendment No. 99), 17 RA;110, -249
Revisionrp:
141. -0'
9 f .It IF stem 11ý61
f /fl."d Z46?
JAFNPP
PAM7ATOUN TORWNG SYSTM HTN IADO ISOLATE SYSTEMS
ion
Ila) Refuel Are Exhaust Monitor (4b 2 4c) or Id! I "se
S1 a) .Reactor Building Area Exhaust Monitors Ib) 2 ddl
Iii SJAE Radiation Monitors -<. 500.000 P/Il/ec 2 (a)
1i.) Tudpe Buiding Exhaust Monitori (b) 2 If)
I lRl , -dwaste Bul--n Exhaust Monitors lb) 2 1I)
1k) Main Coftr Room Ventlation 4x10c 1 10)
Mechanical Vacuum Pump Isolation o. 3,3.Nrml1ul (io 1~I~ A ~ ~Power Background ceo
IQ A channsl may be placed in an inopWeble status for up to six hours during period@ of required surveillance without placing the Trip
System in the trippe condition provided the other OPERABLE channel ia monitoring that Trip Function, that Is, trip capability is
maintained. An Inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In thes. cases,
the aberle channel shal be restored to a2Me status within 24 hours, or the indicated action shall beoa.
Ic, ceaseoperatin of the ,, e.-q u ntp.i ' eS:3
Id)_,_.soate secondary containment and start the SBGTS. •A t • i.rs: 3.25
r|el Bring the SJAE release rate below the trip level within 72 hours or isolate either the SJAE or aN main steam lines within the next 12
Z.°-f3 7
liro. j
I
10
Revision B
r -
& Amnmn o 9 i , ta 249
SýpeCu4- C, A -*A
CL a 11
.. LIATA.F un-Il
Ia A Xciii.... ma y e-p-a---d in an inoper.b.. status for up to six hot* during periods of required surveillance without placing the Yrip.
Syrste in the tripped condition provided the other OPERABLE channel Is monitoring that Trip Function, that is. trip capability is
nmulntaned. An inoperable channel need not be Placed in the tripped condition where this would cause the Trip Function to occur. In these cases,
the in able ch, nel sh. be restored to able status within 24 hous, or the indicated actio oho ,o b.t
, fb level etti Is in a the methods and procedures of the ODCM
operation of the retualinge-Wipmen tp$' . C.-I. 1c) cease ory r-ontainment and at pa A/ 83
- lking the SJAE release iota below the trip level within 72 OF isoiste either the SJAE at all main steam lines with
.4 C"..j
& S 1 Amendment No. 99. 12 ý, 249 R~es3 o0 f B Revision B
!
/, f
37
{1
':T5 •r e -3 ./
JAFNPP .
RA IONMONITORING SYSTEMS THAT INIT1E AND/OR ISOLATE SYT S
Instrumee tal Number of
Channels par In rument Channels
Trip System T Fnto/Trip Level Setting ovided by Design Action
Refuel Area Exhaust Monitor lb| 2 Ic) or (d)
I1Ea Reactor Building Area Exhaust Monitors lb) 2 (dl
III SJAE Radiation Monitors -K 500,000 pCilsec 2 Ee)
ela) Turbine Building Exhaust Monitors lb) 2 If)
1 t) Radwaste Building Exhaust Monitors (lb 2 11)
T k) Main Control Room Ventilation s4 x 10' cpm" 1 (gI
(hI Mechanical Vacuum Pump Isolation s3 x Normal Full 4 EhI Power Background
Se fi
NOTE--R ,Ao. T1,0-1, o. I cto
six 1 9 dui piods of required surveillance nthoui
placing tTrip
SSystem in the tripped conch provided the other OPERABLE annel is monitoring that Trip Function, •t Iis, trip capabili is
maintained.••"s SAn ;nopeable chainnel aed not be placed in the tripped co tion where this would cause the .Trip Fu tion to occur. In hese cases,/
t he =inup.,-., chanel. 7hall be resatored to operabile status lithi.n 24 hours, or thne indicateda ction sha betakxen__
(b) Tr~ip val antfing Is In- accdomdne With the meths an •drs of the ODCM.
(c) CeaseI operation of the raui' qupent. L- -C::e• • V:•••'• "
d) isolate secondary containmentad-~33 .
lBring the rs ease ate below the trip level within 72 hours or isolate either the SJAE or all main steam lines within the next 12 hours.jr•
l• i - '-q
& I Amendment No. 9. itP1, 203, 21i, 249#7
Revision B
O~f 6
4 1a)
(
f j-Tobb 3,0- A*
(h rrann ol-pbiycnafflit eltlto ihmanSemlnerdain Refer to j- dl A Tal 9 1formimm
SeeA LO .
and 3.5. 2. rrs 7.'.
IM A Channel may be placed in aanInoperable status for up to six hours during periods of requied survedlance without placing the Trip
""A6 Function In he condition, or the Ikndcted action sh•lf be taken.
AN Amendment No. ft ¶2*7-20•, Z1, 249 37a
MEj' T7
Re 16o.#
Revi sionxP•"
JAPNPP
MHTFM FOR cr5 jeS' IT
fig)~~~Reiso Bot o IoainI
JAFNPP
•-_.,.om.~m'ITEMS TRIVAT 1T14TE ANDORm ,SO'YF S.STE•
~~~ LCO SA d YCO 3.. an .5 ..
__ _ _
91 r is l t o Is
h scto" n the t ripped condtio wnnr th Ind gicate aton thigain b te a t ken rost .Roto toA T.732-.or1nku
SAmendomet No. inr t 127. 0 1 249 3
3 7r19 xy
CID ~ ~ ~ ~ ~ ~ ~ ~ ~ L oOSWýotr s 8 - iiii. e
(ýýS"~~~~eiso BNt ... 35.. Z
Revision B
cT5 9?(7- 3. /
JAFNPP
Refer to AM"hdl 8 LCO 3.V14'sod LCO3 3.1 .d.y ~
-Control ron Iodati"n Is,,i~in ,., •.''. 3. ,o n,.,,.
Uses owns sensors as primary containment Isolation on high main steam line radiation. Refer to Appendix A Table 3.2-1 for minimum
number of operable Intunament channels end action r ired.
Corwson v factor ,is . X0 c.pm- IP~ilcc.
(iiV ee RTS LC 3.5b.1 nd 3.5.b.2. Se ~ 'S___
1k) A channel may be placed in an inoperable status for up to six hours during periods of required surveillance without placing the Tri
Function in the tripped condition, or the indicated action be taken.
L~e•- .
L$• \ I Amendment No. ,t 127, 208, 249 37a 1-iCr53
Ta 7 e-'1 Revision B
I
0; or
JAFNPPA
TABLE 3.10-2 NIMU. TEST AND CFREQUENCY -I. QI•DTQNM MONITORMING SYRTIEMS"
nInstrument I Cnstrument Channel) Instrument Channel Logic System Instrument Channels hFntonal Test" Calibration ,Function Test"'
I Main Stack Exhaust Monitors and Recorders Daily Quarterly Quarterl¥, )Reluel Area Exhaust Monitors and Recorders Daily Quarterly Quarterly n -
ig. Reactor Building Area Exhaust Monitofst~jjý u(Mbii1~
RFadwast. uildino Exhaust Monitors and Recoders DailyOuaiImulv
Quarterly- 5
Quarterly fl'tnrmitil
24 Months eCT
R er r$.:
SJAE Radiation Monitors/Ollgas Line Isolation Daily Quarterly Quarterly Once per 24 Months
Main Control Room Ventilation Monitor Daily Quarterly Quarterly --
Mechanical Vacuum Pump IsolationO - -
Liquid Radwaste Discherl Isolation*,""'
Liquid Radwaste Dischar Measuring Devices"
Liquid Raeoaste Ditchari Recorder "
Normal Service Water Ell
SBGTS Actuation
WT: 1-7!5g)
ie Monatorl
le Flow Rate
)e Radioactivity
luent
Daily When Discharging
Daily
Daily
Daily
S-- Once per So 4 4 Months si'.l.
Quarterly Quarterly Once per 24 Months
Quarterly Once per --
18 Months
Quarterly Once per - e', 18 Months
Quarterly Quarterly
... Unce 24 Months
,5 .3
Amendment No. 93. 12?. 213. 2. 248 ;L,4A 4 2- ý,
REVISION F
ft
38
(rxo$72A1
,,
3.1
rAJ
CTS
$~, c~f~AA-hW~% 3.34Z
C,2 Notmal SmyriceW,
S SOGTS Actuation
Amendmflefl NO. 3382. 21. 23 38
Revision
6-5?eCA{1# :3.3.7.1
S~JAFNPP ., I .' • TABLE 3.10-22 -_
• IMUM~~ ~ TSANCA/BRATION FREQUENCY FOR IADI ,A1_•MONI'i
Instrum• -) Intr,.me",= Channel Inst, Instrument Channels • CheckW • Funcltionlal Test" ..
amn tack Exhaust Monitors and Recorders Daily Quarterly Quarterly
Refuel Area Exhaust Monitors and Recorders Daily Quarterly Quarterly ".
Reactor Building Area Exhaust Monitors, Recorders, Daily Quarterly Quarterly Once per and Isolation 24 Months Turbine Building Exhaust Monitors and Recorders Daily Quarterly Quarterly
Radwaste Building Exhaust Monitors and Recorders Daily Quarterly Quarterly -
SJAE Radiation Monitors/Offlas Line Isolation Daily Quarterly Quarterly Once per
Man Control Room Ventilation Monitor u t Quarterly 24 Months Mechanical Vacuum Pump Isolation- .. Once per
1 24 MonthsLiquid Radwaste Discharge Monitor/ IsolationklmN~4
Liquid Radwaste Discharge Flow Rate Measuring Devices'4
Liquid Raglvaste Discharge Radioactivity Recorder *,
Normal Service Water Effluent
Quarterly
Quarterly
Quar terly
Quarterly
Quarterly
Once per 18 Months Once per 18 Months
Quarterly
SBGTS Actuation .. Once per T 1 24 Months i
Daily
Daily
M37 .' 1, 2`-1Z
S2.1 /
3ee LTS �
Amendment No. 93, 12?, 213, 2, tf38
& t 73-7IA00
its :;- 3.4. 11NG SYSTEMS
Y. 4. q. 2. 3.4.413
ument Channel Logic System Calibration (Fugction Te`st ý"'
.1, 3. z A 2
S'l9 sAccu RM'. aL
Daily When Discharging
Daily
Once per 24 Months I
-T,7, 7
5rectf'C*A'Or -3,73, 7 7-
(91JAFNPP
TAB 3.10-2 X-:37 MINM TESAN CALIBRATION FREQNCY FOR RADIAION M0 ORING SYSTEMT - 3., s '~~~~~~- P. 6 ,,•T•• . .
Instrument Instrument Channel InstrumentiChannel Fonci S este Instrument Channels LCheck" Functional Test" Cairtion a ! Fnto Ts~m
a -1 S .7 , fl .. ........ .......... Puma .......
Main Stack Exhaust Monitors and Recorders Daily Quarterly Quarterly
Refuel Area Exhaust Monitors and Recorders Daily Quarterly Quarterly Reactor Bitlding Area Exhaust Monitors. Recorders, Daily Quarterly Quarterly Once per and Isolation "24 Montt
Turbine Building Exhaust Monitors and Recorders Daily Quarterly Quarterly
Radwaste Building Exhaust Monitors and Recorders Daily Quarterly Quarterly
SJAE Radiation Monitors/lOfgas Line Isolation Daily Quarterly Quarterly Once per 24 Month
Main Control Room Ventilation Monitor Daily Quarterly Quarterly -echanical Vacuum Pumo Isolation- ..... ...
Liquid Radwaste Discharge Monitor/ Isolationkt'e"""
Liquid Radwaste Discharge Flow Rate Measuring Devices'
4
Liquid Radlaste Discharge Radioactivity Recorder "
Discharging
Daily
Daily
Quartelly
Quarterly
Quarterly
Uuarterly
Once per 18 Months
Once per 18 Months
II...,, p r. -- .. 24 Months
unce perunce per 24 Months
Normal Service Water Elfluent Daily Quarterly Quarterly -
SOGTS Actuation . Once per
24 Months
5ee! 3.3ý ýJ,3£.
Amendment No. 9., 12?. al3, 23, Zfic
-e CM5 ,ev" 2.7
38
r'7 J-JPeeP/e
(
JAFNPP
TABLE 3.10-2 R MINIMUM TEST AND CALIBRATION FREQUENCY FOR RADIATION MONITORING SYSTEMS" c'(Z5
Instrument Instrument Channel Instrument Channel Instrument Channels Check" Functional Test" Calibration
JLII�Logic System Function lest"~
Main Stack Exhaust Monitors and Recorders Daily Quarterly Quarterly -
Refuel Area Exhaust Monitors and Recorders Daily Quarterly Quarterly Reactor Building Area Exhaust Monitors, Recorders, Daily Quarterly Quarterly Once per and Isolation 24 Months Turbine Building Exhaust Monitors and Recorders Daily Quarterly Quarterly
Radwaste Building Exhaust Monitors and Recorders Daily Quarterly Quarterly SJAE Radiation MonitorslOlfgas Line Isolation Daily Quarterly Quarterly Once per
Main Control Room Ventilation Monitor Daily Quarterly Quarterly 24 Months
Mechanical Vacuum Pump Isolation" ..... Once per 24 Months
Liquid Radwaste Discharge Monitor/ Daily When Quarterly Quarterly Once per Isolation"'1"'" Discharging 24 Months Liquid Radwaste Discharge Flow Rate Daily Quarterly Once per -Measuring Devices' 18 Months Liquid Ragtvaste Discharge Radioactivity Daily Quarterly Once per Recorder 18 Months
Normal Service Water Effllent Daily Quarterly Quarterly -
SBGTS Actuation - -- Once per 24 Months
Ers;
Amendment No. 9.. 92,, 23,. 233, 7. q?38
(~rK f
(Ti
r C;-ýCo-ýOA
33,4- Z
5pr-CC4ýIC(CA kr" 3. (. - -3
JAFNPP
TABLE 3.10-2 MINIMUM TEST AND CALIBRATION FREQUENCY FOR RADIATION MONITORING SYSTEM'Was
Instrument Instrument Channel Instrument Channel Logic System Instrument Channels Check" Functional Test" Calibration Function Test"
Main Stack Exhaust Monitors and Recorders Daily Quarterly Quarterlv
Rieuel Area txnaust monitors and Recoraers Daily Reactor Building Area Exhaust Monitors. Recorders, Daily and isolation Quarterly
1uurterly
Quarterly Once per -
Z ýfm' 3.
IT$ 3.7s,
S*e JTS
;ee 9. 3.7 -L
I.. . . . .. . .. lry R Urbino Building Exhaust Monitors and Recorders Daily Quarterly Quarterly -Radwaste Building Exhaust Monitors and Recorders Daily Quarterly Quarterly-
-5JAE Radiation Monitors/Oifgas Line Isolation Daily Quarterly Quarterly Once per | -24 Mnrh
Main Control Room Ventilation Monitor Daily Quarterly "lQarterl - .
Mechanical Vacuum Pump Isolation .. - Once per 24 Months I
Liquid Radwaste Discharge Monitor/ Daily When Quartery Quarterly Once per Isolationk't"Miff Discharging Qatryurel 24 Mon~ths Liquid Radwaste Discharge Flow Rate Daily Quarterly Once per Measuring Devices' 18 Months Liquid Radjraste Discharge Radioactivity Daily Quarterly Once per Recorder 18 Months Normal Service Water Effluent Daily Quarterly Quarterly .
, SBGTS Actuation24 Mnh
-<-e ITs.3, 3.3.C. 2-
I
(
Amendment No. 93., 127, 213. 2, , -, V.38
(P9
I
-Pk I I ý to
ni* * -l~h I ... _
<
F
5P C C ;iF'fd b 0 # a 7. ý
JAFNPP
TABLE 3.10-2 Z I MUM TEST AND CALIBRATION FREQUENCY FR"R,/IAiON MONITORING SYSTEM
Instrument Instrum Channel Instrument Channel Logic System Instrument Channe Check" Funicti% na Test" Calibration Function Testfi""4'
M i Srl
Main Stack Exhaust Monitors and Recorders Daily Quarterly Quarterly s Refuel Area Exhaust Monitors and Recorders Daily Quarterly Quarterly .anReactorlsltoBuilding Area Exhaust Monitors. Recorders, Daily Quarterly Quarterly Once2 Mnhper _Z, •
and Isolation 24 Months I Turbine Building Exhaust Monitors and Recorders Daily Quarterly Quarterly -
Radwaste Building Exhaust Monitors and Recorders Daily Quarterly Quarterly -
Q~l hm .JCCr. a '4 * &A :,lf Ins L..r .au I ;DaUiy uuarterly Uuarterly
s LJ24 Months I Main Control Room Ventilation Monitor Daily Quarterly Quarterly -
Mechanical Vacuum Pump Isolation' ..... Once per 24 Months
Liquid Radwaste Discharge Monitor/ Daily When Quarterly Quarterly Once per Isolation"AW01 Discharging 24 Months Liquid Radwaste Discharge Flow Rate Daily Quarterly Once per Measuring Devicesla 18 Months
Liquid Ra aveste Discharge Radioactivity Daily Quarterly Once per -Recorder IV 18 Months Normal Service Water Effluent Daily Quarterly Quarterly -
S8GTS Actuation -. Once per 24 Months
TJE ;' .
C! t5 ID
Once per
Amendment No. 93, 127. 21-3, , 148
See. � 4 q3
38
MclrsJ .(of" 6
/
" "- " -"V. ,,v,,,,n ,,,, • ,.,,,, gas ,,,,,,,,,,
CT5 9C7-5 4ý*
JAFNPP Al S....TABLE 3.10-2--- /10
MINIMU HEST AND CALIBRATION FREQUENCY FOR RADIATION MONITORING SYSTEMS" S/ / /
Instrument ChannelsInstrument Check Mj
Main Stack Exhaust Monitors and Recorders
Refuel Area Exhaust Monitors and Recorders
Reactor Building Area Exhaust Monitors, Recorders, and Isolation
Turbine Building Exhaust Monitors and Recorders
Radwaste Building Exhaust Monitors and Recorders
Instrumen Channel Functio ITest"'
Daily Quarterly
Daily
Daily
Daily Daily
Quarterly Quarterly
Quarterly Quarterly
Instrument Channel Calibration /
Quarterly
Quarterly Quarterly
uuarterly Quarterly
Logic Syster Function Te.
o . Once per 24 Months
SJAI Radiation Monitors/uOlgas Line ISolaidiOnl Udily QJUOalrtery U0auu a•,y , F 2 Month
ain ontrol Room Ventilation Monitor Daily Quarterly Quarterly - - ,
Mechanical Vacuum Pump Isolation - - Once per 24 Months
Liquid Radwaste Discharge nitor/ Daily When Quarte y Quarterly nce per isolationicieJ)(I.I Discharging 24 Months
Liquid Radwaste Discharge low Rate Daily Quar erly Once per -
Measuring Devices'4 - 18 Months
Liquid Radwaste Dischar e Radioactivity Daily Q rterly Once Lqi Ray., per /
Recorder i 18 Months
Normal Service Water ffluent Daily uarterl Quarterly -
SBGTS Actuation - - - - Once per
C-
5t0 TT", ,;. ,(
Amendment No. 93, 12:7, 213, 233, 2.'18p/"efCAr T
38
I
A
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(3\
(4\
(s')
(7\
(ci'
(�o�
('I)
J
I 1-i I
n st(1104,
er_. y L&O
3.4
3 I 3
C-r-5 PT T5 3.1
JAFNPP
\ islation-c"a) Liquid Radwaste Discharge Flow Rate Measuring Devices"'i Liquid Radwaste Discharge Radioactivity Recorder idi
Normal Service Water Effluent SBGTS Actuation
Amendment No. 93, 127, 213, 233, j Y.38
14 -pj 247C •1�6
/
,n
0
ýJ
S-,
*1
(
II'
I
(bl_ 1= 0;en! =chocks shall be Pa. Inedulearylk-Aw"y !dp insmAnents am to 'be opepM!./ý
(c) 0 check 00 -be p-erf w-ff -od p-rjo`rýtýc"ch release.
(d) Uquid radwasts aftent Im WWUWmntsticn surveillance r@QLIirSfn@M need not be
psrfcx i when the insvwnents am not reqLired as the result of the fie rge path
a on pla
:7 s
cha" our"
chec
(a) An ionuvurnent channel calibration shell be pedornwd with known radioactive sources
'Mrfstandandizedd on plant eauiwnent which has been cartlx.t with NBS vaceablo
(f) CS'--mWatsd autwnstic scumWorrihall be Perfwmed once oer 21ý4mýý
Xjhendment No. 92,4'1;1 23339
Revision Br'
(.
.1.~. [hjlfir ~ i
gil 3
iAiII Tl. ~ III~ I jii
39
PJ, Mv ý 1.0
ý,Pec
rs er r--, 9- 1., 17r5,' 3. 3- -f- /
3 5, a. L JAFWP 3.7 5
s9mmiatmMLEIMEr
to) Funcoonel tBm, caMbr*Wm wW nwwK 0*04 n"d not be perfam"mmed emm Wamwmu we not roojnW to be opwaWs of we VippW.
(b) InmunwK WWU MoN be Pa !a at We; per day *jrkV tMms np"sniods wl the Wwaumwn we r*QL*W tO bf OWGW@-
U*W radwom SMO. Irw kwwumwuadm swvoMwm fow*wrwft rwW not be Mlormed w tho wmmvnwM m not nKW*d w the rmft of the dodo go path not NWV udktot
is) An kwwumm chw,- mWastion OM be pwfo"ned with loww rodowo" --- \ suw4wdind an clonAaWwwn mll Ff with NU vw"ble
us-r- (f) jSkZnuMlwW g&L*Mwno%* ag;g =be wwýdfcýmisd ýornwme ow" ý24ffmwmýdhwfth*07
high ým=ý ýpwtofý
for
0) Tft&WwmwwnmionbwmomdftmgwfunadonolKdardiVwC-7ha Lotetkokm tM WN gffdi@t Of I i 1111 4 a sknWaM eWcvkal mWW bvo Ow emmuL rnam kivoymi p ------- sim wm be c o, - uwq Wgnals once ova V Om mordhL
-1. XJý
CTS
Affmonw*No. Oil aG;, 233
JAFNPP
f� �-�*�Y�1
33 4.?L te) Functional tests. calibrstions andl inswum~ent checks need not be performed when -
these instrants are not meuird to be oeable or ae tripped.
(b) Instrument checks shall be performied at leas once pe day during these periods Z*4I n theinstrumen we tobe le..
(Cc) A source check shall be performe prior to each release.
(d) Uiuid radwasts •effluent Hne instrumentdc rveillnce rqmuirements need not be Performed when the irstruments ae not reqred as the rsult of the discharge path Cs . 2.
not being utlized.
a) An instrument channel calibratiun shall be performed with known radioactive sources
sta.,da..d od an lan, im~m mfc u bee ca Na with'r~ Na vae ab¢.T
Simisted/a m aif c tuation shell be pormed once pr 24 monts re 0
ICg-) e9fer to A4&f or charne functiWa test anid instrument chaý calibration requiremnts Liabl 4.2-1). These requirementts are performed as part of ST- &n 3.37. 7L
main steam high radiation monitor surveillances.ýý ý o
(h) The logic system fwwdctIoa sell~a include a Calibration -of time delay relays and
11) This irtstrimnenttdon is excepted from the hnfntioa test definiton The fumctioa 3. ~2 tes will consist of. 9 -fecid a simdulatd oeletrca signal inro the measurmient channel. Then instrumenvt channel wil be cabrt td using simuitlated eolectri sgnasonce every three C TS Kell:~
3.3A.7.
tT ,r T5s V~el. I sj
Amendment No. 911,"O; 2333.
Ehe13]?or to
Qýrq
a 7 -) -5.7 'ý-
l(f) SirnW@W onOwtic WUNWM " be Polal -- d w: possiblo, M l9bic sys furcd" tan wN be ý;Tcrf;Wd i
fwwdonal fth-)-!Týhs nlotqtc *2
tinmn for tewr furcbonkv =00ýw ý ý T-
Affmwimem No. 03 '14; Z3339
t-RF- T-53 Tale-ý ,+ 6
CT5 p-T- : 2,1
JAFNPP
INOTTE FOR TAKLE 3.10-2'
rations and instrument checks naee not required to be operable or are
Irnstrume not pcks shae be performed at la once per day when the instrmet a ~required to beoper
(c) A source k shall be performed prior to release.
,d) Uquid radwa effluent line instrumentation illane requireme need not be
ptndrdizmed ohn plnstrupments r o wh qich h as be e n ucl wi the d' trace pat n ot -beil.\\
possibein allloic te m ucinltas ilb fre sigte a ak.
(g) Refer to Apndix A for inchmnt cha o s channel $e. - ... .
calibration reu irments (Tale 4.2-1). These reZuirments ar rf
main steam high radiation monitor sur mweilanc
h)Thelogics r m funcftional teats shall includte a rtrip no tim eea re and
standarized on plnt• uipme K whicthe functional definition. The functio
channel6. alTho aese functionaltests will be rfoirm'ed usin smltea lmat jaks
mai steam hi h radato monitor ,.
rmendment No. 92 417 23339
these
?a I *_ ,r- _50-
CTS ,pE-r5 3.(t
JAFNPP
(a) Functional tests. libation and instrument checks need e
(b) instrument chec shall be performed a tImoncePer do during these periods
when the instrumn are re *red to ble.
(c) A source check shall be performed prior to each release.
(d) Liquid radwaste effluent line instrumentation surveillance reQuirements nred not be
performed when the instruments are not required as the result of the discharge path
not being utified. .CT . "
(a) An instrument channel alibration shal be performed with known radioactive sources
standardized on plant equipment which has been calibrated with N8S traceable
(f) Simulated automatic actuation shall be performed once per 24 months. Where I3.3 possibl(i all I ic System functional tests will be performed using the test jacks,
(g) Refer to Appendix A for instrul''Iment channel111MM functional test and instrument channe 5 iaZ
(i) This i ntartion is excepted from the functiol test definition. The functional 'test will ClFm of. injecting a simulated eW-- ct4&gnal into .th measurement ,,
channel. instrument channels will be caliorated using simulated electrical sinals a• every three man*,&
Amendment No. 03,-"; 233 39
CTIJ
C 5 !,6-7-' .f<(KC
43
rRE5 / s" K z
,K
C r (S r5 G
LINITINC CONDITIONS FOR OPURATIy4 SURVILLIANCE 3"UtIlREENTS
c A subsequent teat verifies solidlilcation.
Solidification of the batch may then be
resumed using the parameters determlr d by the PCI.
2. If the Initial teat specimen from batch of waste falls to verify solid tcation. the PCF &hall provide for the collection and testing of representativ test *peacmen@ from each consecutiv batch of thi mam type of wet waste u til at leasn consecutive initial tea specimens d n-h3 strate solldification.
The PC? ehy41 be
Amenmenth p i N.t 944
strate soiifcton h{&Ps Ibt
LIHITINC CONDITIONS FOR OPERAION SC--
N.
(5.O
5.1
TOTAL DOSE
TOTAL DOSE F~fl4 URAlNI FUEL CYCLE 5.1 TOTAL ISE RON URANIUH FUEL CYCLE
Applicab ility A A p qc ab ililty
Applies to red tion dose from releases of ra- pplies to the calculation of total dose d dioactivity d radiation from uranium fuel releases of radioactivity and radiation cycle source. / uronlu fuel cycle source..
To aocu that the requirements of 40 Cli I To ensure that appropriate calc ationa are me performed to determine total do to a u
of the public. Spe tication. ecllaln
Specifications
a The dose or dose commitlment to any er of a. Dose Calculations Cu lativo dose cont the public* due to releases of red activity tiona from liquid and gaseous . ffl and radiation, from uranium .t cycle shall be determine In accordance with sources shall be limited as fol wat Ificatlona 2.3.a 3.3.&. and 3.4.a as
/~accordance wit Vthe ODCH.
I. Less than or equal to 25 a amlyear to the
whole bodyl and.
2. Leas than or equal to mrem/year to any orgam except the thyr Id which shall be limited to less th or equal to 75 area/year.
I. Wtth the calculated d sea from the release b. Cu ative dose contribution f am d of radioactive materiels in liquid or gas- rad ation from the reactor unit a and eous effluents exceeding tuice the limits ra waste storage tanks shall e deter
ribuumnts Specid in
Amendment No. 93 46
ember
/ oV I, ý,
UNITING CONDIITIONS j'ROPERATION SURVIALUAJCE ROQUIlR0Ilfs
Amendment No. 93 47
,/ /
of Specification 2.3.a 3.3.i or 3.4., In accordance with the metda calculations shall be e0 including an meters In the ODCH. Th•. metimate of direct radiatio contributions op kiable only under conditio to Xdtermine whether the Himit of 5.1 have in 9p ification 5.I.b. been \xceeded. If this is the ca . a report definih corrective actions to be taken to reduce a sequent releases to level within limits, a & with a schedule for ach vian conformance, hell be prepared and submi ad to the Comin on within 30 days. This r port, as def in in 10 CFt 20.405c, ahal Include estimates f the radiation exposure (dose) to a member the public from.uranius fuel cycle source including a11 efflueant pathuwys and dire radiation, for the calendar year that imci as the relase(s) covered by this report. shell also derb* levels of radiation a concentration
o radioactive materiel Iv ead. and the caus of the exposure levels concentrat s. If the estimated dose(!s exceed(s) \ th:e aoy limits. and if the rlee a condition rest tIng In violation of 40 FP 190 hai not alr r dy been corrected, the sport shell Include request for variance accordance with t po ons of0 CI 90.
Submittal of the \report ts coonsiders a mely request. and a variance IV grant d
un i1 staff action on the request is con:. plt
V I
I
IL
f-i
MG 26
Tcr5 Et o
~rwo 1,,, 3 .ý ý
NOTES To rTGURE[ 5.1- J
(a) •Mel stack (heig~ht is 350 feet
(b) VW2 stack (height is 430of
(c) JAFVPP stack (height is 5 feet)
(d) Building vents
(e) MII1 radioac vs liquid discharge (Lake Ontario, bottom)
(f) )f1P2 ra oactive liquid discharge (Lake Ontario, bottom)
(g) J P radioactive liquid discharge (Lake Ontario. bott )
C Site boundary
Mi Lake Ontario shoreline
Additional Information:
-- M(P2 reactor building vent s located 187 feet above ground level
-- JAFNPP reactor and tur ne building vents are located 173 feet above ground level
-- JAFVPP radwaste bu ding vent is 112 feet above ground level
Amendment No. 515 127
49
-7 7- 7
REVISION D
(a) ~ ~ ~ ~ ~ ~ r NM tc bgi 5 et
M NN) Pl2 stack (h h s 430 fleet).
CW) JAFUPP stac/ (height is 385 feet)
(d) Building onto
(C) 3.WI r /ioactiv, liquid disc r, (Lake Ontario, bottom)
CM) /1W radioactive liquid d mcharge (Lake Ontario, bottom)
(a) J P radioactive IL id discharge (Lake Ontario. bottom)
Mh it. boundary
CU Lake Ontario sho line
Additional ormation:
-- UZ reactor building vent is loce 187 feet above $round level
-- .l P reactor and turbine b1il a• vent@ are located 17 feet above
-- JAIIPP radwaste building v is 112 feet above g level
,Aaanet Is. 9" 127
49 •rrs]
LIMITING CO94DIT S FOR OPERATION "- SIIVEILLAWCE REQUIRDEHNTS
6.o RADIOLOGICAL ENVIROMEN A ITORING
6.1 MONITORING PROGRAM6. OT11 PG
Applicability The radiological envi mental mon-toring somplea shall be collect e pursuant to Table
At allU clas. • 6.I-I from the locations van In the table and figure(s) In the ODCH and I1 be analyzed
Ob ctive •pursuant to the requirements o able 6.1-1. and the detection capabilities u•red by To *va to~l the effect* of plant operation on Table 6.1-3.
the onvir6 and to verity the effectiveness of the cont rolsaradioactive material. /
Specification,
a. with the radiolo cal environmental monitoring program not tug conducted as specified to Table 6.1 -lI@ repare and submit to the Commission fto the nual Radiological
Envdrontental Operating .rt. description of the reagans fofr noconducting the •program an required and the aJ~ni for pro
aempi a schedule It samples are u btain:abLPG dd to hazardous conditions. so ~onal unvaialt. theft, uncooperative r I~deate or • matlfunction of automatic a qpln eup at. It the letter, flortb
.*al bemd ocmlte correct Iva actiton• prior to the end of th* next a =piing • perod.)
Amendment No. 93 51
(7;)'
,
C s x C, /,
. IITIN CONDITIONS FOR 0PoRATIOs'. SIUT,'-]LLANCK REQUIRFE.
b. lith the level of radioactivity (as t reaule',•Pf plant effluents) In an eavironmen • sampliuj medium at a apecifled location ox cesdint tae reporting levels of Table 6.1-2 when avera ad over any calendar quartere prepare and sbhit to the Commission a report within thirty (30) days from the end of the affected plender quarter or within thirty (30) days m the time it it determined that a report level has been exceeded. This rport ehall identify the cause(*) for xceadIng the limit(s) and define the corrective act i(a) to be taken to reduce radioactive efflue a so that the calculated annual dome to a eber of the public Is lees than the calendar ear limits of Specifications 2.3. 3.3. and 4. When wmore than one of the radionuclides n Table ".1-2 are detected in the sampling dlum.
t a report shall be submitted if/
concentrat (I) concentration2 , * & list level•1) - limit lev- l V
When radl uclides other than those in Table 6.1-2 a a detected and are the result of plant efflue t•. this report shall be submitted It the calculated annual dose to an Individual Isa ua to or greater then tile calendar year I Its of Specification 2.3. 3.3. and 3.4.
Wneport Is not squired if the measured Ilvel of radloactivi was not the result of plant effluents; hoe r. in such an evet,t tile condition shall reported and deacribed In the Annual R t ological Environmental Operating Report.
7
UTS
K I
Cr� �73c�('
KI.IHITING C ITIOHS FOR OPERATION SUII a LLANCE REQUIREHMS
c. Vith silk or Ir h leafy vegetable weaplee unavailable from a or more of the sop,: locations required bTableý 6.1-1, locations for obtaining replace nt Vamples shall be identified and added the radiological environmental monitoring ograq within 30 days. The specific locat ne from which samples were unavailable may t a be deleted from the monitoring program. a cause of he unavailability of samples a the new
I ation(s) for obtaining replace at Sample shall be identified in the nag Semiannua Radioactive Effluent aeas. R a0ort. Alsa cluded in the report shall a revised nture(s) and table for the
Amendment No. 93 5
(~wT
(
f "• TABLE 6.1l-1
OPERtATIONAL RtADIOLOGICAL 1ENVIRtOMNHET, HON•H RIN OCRAH
Eposure \Samp!ling dlk
Pathway (s\Coliectioo Type and frequency
and or Sample Number of Sam ln .a" Locations Fquenc7*a of Andlyeqsu. n\
Radiolod s Samples from 5 locations: Continuous eam- Ba olodine Canister.: and pie operation Ansa e weekly for 1-131. Particulate a. 3 samples from offelts loca one in dif- with sample col
ferent sectors of the highs& calculated lection weekly Particc te Samplee: site average DIQ (based on al licensed or an required Crais be; radioactivit b) site reactors). by dust loading, following i ter changel ".
whichever to composite ( locallyn) b. I sample from the vicinity of a c mity more frequent. for gams io epicp
&vin8 the highest calculated site ver- quarterly (as minimun). e DIQ (based on all licensed site &
a ors).
C. 1 s Ie from a control location 9 to 2 miles distant in the least prevalent wInd rectioo
Direct (a) 32 stations w h two or more dosimeters arterly Camma dome monthly or ladisat••"o placed as foll at An Inner ring of stations quarterly.
in the general es of the site boundary and an outer ring in he 4 to 5 nile range from the site with a a tion in each of the land based sectors of e ch ring. There are 16 land based sectors a the inner ring and 8 land based sectors the outer ring. The bnlntc of the .tt 9a3 (8) are placed in9 special InAre a ea uch as population centers.t ne rby reside a.shos n In2 or 3areas to srean oto ttos
Amendment Ho. 93 54
i
exposure Pathway and/or sample
ofI6. l-1 (contitlued)
Sam]
Number of so lee and Locations Col
"pins and leCtiooa) Type and Frequency quencys of Analysis
9ý5C./
WATERVORNE \. \, (f)
Surt c ) a. I sample upotr Composite man- C Isotopic analysis ple over one& monthl Composite for
b. I sample from the oif)e moat downtream month period . Triti) nalysis quarcooling water lots .terly
Sediment on I sample from a downstream area with existing c8per year. Cams faotopI analysis shoreline or potential recreational v tue. semiannually
INGESTION
milk Samples from • lch animals in locations Twice per nth. Camma Isotopic and 1-13 within 3.5 miles distant having he high- April thro analysis twice per month eat calculated sits aver•es D/Q. it December (sa when milch animals are there are none, then I somple It mwtch plea will b on pasture (April through nimals in each of 3 areas 3.5 to .0 collected in December); monthly (Janlea distant having the highest cacCu- January through uary throuth March). if
I ted alto average niq (kiaed on all March If 1-131 required 1It *ned site reactors) Is detected in
November and b. a s Is from witch animals at a contr I December of the
loca tion (9 to 20 miles distatn|•nd in preceding year). less •revalent wind directin'. \
Amendment No. 93 55
16
@9
15-- c"(_
(YT-5 ,Z6Sý
\ \) ad Locatios Frequeac,(a)
£;Iposur.
Pathway •,or I
Flab
Ifood Pro
Gamm eotop amalys of portions
camea lotop analysis of portions. (I to isclude I
Amendment no. ? 127
56
4-P
(
ample
rensre oallyTmportut sreiee r a. 1 Bu of each of 2 commercially or Twice per year.
foee treat ally Important species is h viasty of a ite discharge point.
b. 1 sample o each of 2 species (same as In a. above or of a species with similar feeding hab a) from an or*& at least 5 miles datn Efrom the alt'(d).
a. In lieu of the gardes census a" specified Once during is 6.2. smple of at least 3 different rvoest season. kin"d of broad If vegetation (such as vegetables) gr nearest each of two different ofiflt olocations of highest predicted site avdage D/Q (Jasod os all licensed it. &eac or.).
Ome (1) amplo of o ch of the simller broad leaf vegotati grows at least 9.3 miles distant in a least prevalent wind direction secto (
V
, -
/o'z
NOTES FOR TABLE 6. 1
7() It Is recognized that, at times, I y not be possible or practical obtain samples of the media of oicm at the most desired location o time. Zn these instanc.s s o I alternative media and location.s may be chosen for the particular vay in question. Actual locations (distance and directions) from the te shall be provided in the Annual Radiological Environmental Operati Report. Calculated site averaged D/Q. valku& and meteorological par tars are based on historical data (specif in the ODO) for all I. used mite reactors.
(b) Particulate le filters should be analyzed for gr beta 24 hours or ore af t sampling to alloy for radon and tho daughter decay.. If
gros t activity in air Is greater than I aes a historical yearly man f control samples, Sa Isotopic awa is shall be performed on the
vidual samples.
(c Game isotopic analysis =mane the dentification and quantification of Sama emitting radionuclides t may be attributable to the effluents from the plant.
(d) The purpose of these aes is to obtain background information. If4t is not practical to tablish control locations in accordance vi; the distance and wind rdction criteria, other sites which pr valid background data be substituted.
(a) One or more trments, such as a pressurized ion , for measuring and record S dose rate continuously any be used in ce of. or in addltion to, tgrating dosimeters. For the purpose this table, a thermolmin uet dosimeter my be considered to be phosphor and tvo or more pho ors in a pocket may be considered o or more dosimeters. Film b gas shall not be used for measuring t radiation.
(f The "upstream sample" shall bet take at a distance beyond significant influence of the discharge. The " -tram sample" shall be taken in an area beyond, but near, the g zone, if practical.
(C) Composite. samples should collected with equipment (or equivalent) whic Is capable of colleti g an allquoit at time intervals vhich are short (e.g., hourly) Iative to the compositing period (e.g., ma y) in order to as3re t a repreentatve sample Is obtaied.
Ma A milk sap oain a eurdi able 6.1-1 dfined as a location g••a at: least 10 ailkin$ cows present jra designated milk
of co rri:•ng with local fares ta of 10 milking cows is nece ayt gaate nadqaesupply• silk twict per month for
for0 smlsa areuto uki cav s an peios when the •sadulty
S. animals are dry. In the event th• 3 milk sample locations caimot meet I• the requirement for 10 I•€le )tvs , then a sample location ba~vin$ less/ \ than 10 milki.ng cows ca be aed if an adequate supply of milk can/
rea~sonably and reliatbly be obtained basedt on commanicat:ions with th•
Am•endment Wo. 93 37
K
C T-5 9 E-75 I
Amendment No. ) 127
!K
(17 eE7YC,/
Amendment No. 7J" 12759
C-xcr!J ? /?6ý le-
(7-5 FC75
sons3 FOR ?LMLZ 6.1-3
(a) The LLD is the smallest concentration radioactive material in a Sample that will be detected vi 9S% probability and with s% probability of falsely conclu that a blank observationrepresents a "real" signal.
For a particular measureme system (which may include radiochemical separation),
" • ~4.66 Bb
LLD is the lower limit of detection, defined above (in picocurie r unit mass or veline),
sb i the standard deviation of the b ound countinq rate or of the counting rate of a blank sampl as appropriate (in counts per
2 is the counting efficioeny a counts Per transformation),
V is the sample usie (in ts of mass or volmes):
2.22 is the anmber o transformations per minute per picocurieo
T is the fractio rldlochemical yield (when applicable):'
/is the r active decay constant for the particular r nuclide,
& t is elapsed time betwen eample collec (or end of the sample ollection period) and time of coun
Cal values of 3. V, T, and sheuld be used in the
it should be recognisod that LW Is defined as an a Mmxi (before the fact) limit rep enting the capability of a measurement
e(after the fact) limit for a particular measurenet Analyses shall be performed in such a
ganet ha th a LLW will be achiewe under routin ditioms., Oncasie ly backgrounad fluctuations, umvI sml $=BPle sizes* presence of interfering nuclides, or ruControllables cir tances nay render these LLD ssnsc cases, the atributiag factors shall b;ei ieti and. described i the ladiloegical Unvirommental prai Report.
(c) so nking water pathway exists at Usne Mile Point Site.nd operating conditions due to direction and distance of
&rest drinking water intake Therefore, an LLD Value of pci/liter is nsed.
Amendment no. q$ 12760
Q (P
I
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C73 /r&TS 6,
LIHITING CONDITIONS FOR OPERAY , ON SUtV/Z LLANCE RIEqUIRnIENTS
(be conducted dur
nst once par 12 moni that will provide a door to door aurve
mGulttjg local •t The results of $(e Ii
luded In the A 'uai I Operating Refort.
Amendment No. 93
�;y (�f- �-
6.2 LAND USE CENSUS PROGRMH 6.2 UAND USE CENSUS PROGRAM
Appl cability T use cesus e
At all time.lnfomtion beat results, such as y
Objective aerial survey, or by co ture authorities, a .
To Identify I ties@ of milch aninmal and Or- use census shall It inc dens of gre• r than 50 square moters Vi in 3 dlological Enviro ntal miles of th site.
te'fr
Speci fic ionis
a. A I ad use census shall be c ducted and sh I Identify the locations I all milch a mals. the nearest raaidenc * and all &sr
nsa of greater than 50 sq, r meters proucing fresh leafy vets n•a, in each of the 16 meteorological sac ra within a distance of 5 miles from th site.
b. With a land use conau identifying a milch animal In a locations) which represents a calculated DI/Q valu greater than the values currently beng usa In calculating Surveillance Requirement .4, identify the new location(s) in the "axt Semiannual Radioactive Effluent Rales. Report.
Broad leaf ve tation sampling may be performed In lieu of t e garden census as specified in L~•l 6.-1. •
I
61
CT S •eT-• •. 7__
(
t
LINITINC CONDITIONS FOR 0#RATION SURVEILLANCE REQUXEHENTS
J! I i I I I
Amendment No. 93 62
01 cl
i.
. -- J
With the land use anue Identifying a m ch animal Iocation(s that represents a c culated DIQ (via a same eipoeure pa way) 501 greater t at a location ft which Saoplee are rrently being obta ad in accordance vwh Table 6.1-1, add the new location(s) •o the radiological ea iron.ental monitor ogf program within 30 aey. The sampling •cation(s), excluding he control station aetion, having the e~aet celcinlated D• (via the same exp oreo pathway) may he deleted from this mon oring program after/4 tober 31 of the yea) in which this land mee cnarius is conduct d. Identity the new •ocatlon(s) in the ne) t Sem•iannual Radioactive Effluent Rtelsa e Report and include the additions in t ODhI.
!
Z o ý- z--
CT5 f?�ts c;.3
"Mendment No. 93 '3
0145CTs I
JAPNP9
7.0 A§MIN,173A!IVI �N1fl6LS
wryman procedwurs MWd ad1¶itrs~trev DaiciCss "h be aamisehed. imown'u-ed d
maintained Vst meet or exceed tre remerstIrlts and tecommefldamon of Swouon I
*aciiiv Adffurvetrag Polies" and Procedure . f £9451 18.7-1$72 and Requiatoe
Guide 1.33. Nov.ember 1372. Apoendix A. in addition. procedures "a te
established. wnolemet't and rmainewied for ire PCP. 0COCK and Oewit Conura
P!gtS for effluern and en*0uYnmemai moin using ve gwodance in Mlpi
contained in SoeaficavonS 2.3. 3.3 wW 3.4.
b. 5M~konff£kmd gweeffding Lbvit9x of TONS 1-1.1
When vie 6mits of Tale 6. 1-2 are eeeed refer to Speftsaflni 6.1 .b f o
sprepmU reuiremenms
c. grmiamwum R aslem pffho Ramesa Room
floudna Paedese~v 9ffluient felaeeae ftwts covering tos- sooesn of Vie wunt
duwvq to pveiou 6 mendfs of eeman . a be asub tted Widen 60 deWs aft
J**W I WW Mv* Io eeeJn "or. The penod of go fW*r~ toet UiIbeg -1l
I1. The Rsimdo i'. Effluent Reltolase Reout kchlude a SITnfiwY Of V
openipe ,of radiowthoe howod end gaseous efflunts and solid weats rebel
from VIS Wrt U"g as guiance Regulatory Guide 1.21. Revtoon I. - un
1974. *Mossurig. tvskuetrlS. end Ragaun" Raioectntv in, Solid Wastes
and ReleSeS of Radi*acve MaU~teria n LisWand SWGas~c Effluents from
LigM.Watst..Cele NasleW ftowe fatM. w11h data SUamrmnned an a
Quarterly bse toll!!!g V4 format of Appendix 1 VsieOI.
Amarwdmnt No. A . 2/254
er
Ad
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. .3o 3
REVISION H
Ly
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I
JAFNPP
70ADMN, STRATIVS CONTROLS IT •_•1S 5-a.t;O"A S',•
a. The Site Executive Officer shall have direct responsibility for assuring the operation of) ~the James A. FitzPatrick Plant is conducted in such a manner as to provide contienuing
protection to the environment and shall delegate in writing the succession to this
responsibility during his absence.
b. Implementatlo f the Radiolo al Effluent Techn ai Specifications the r-iI"I• responsibility f the Genera anager - Operati s, with the assi ance of the plant
staff organi tion.
Written procedures and administrative policies shall be established, implemented and
maWntained that meet or exceed the requirements and recommendations of S6ections
"FaSety Administratl e Policies and Procedures" of ANSI 1o8.7-1972 and Regulatoryr Guide 1.33, November 1972, Appendix A. In addttion, procedures shall be estabftsnedry implemented and maintained for the PCP, OfCth and Quality Control Program for f
fluent and environmental moltoring using the guidance in Regulatory Guide 4.1ciia
73RE•PORTING REQUIREMIVENTS 'e T
a. P hnned Liquidand Gaseous Releases shall nt
The limits for radioactive linuid and gaseous effluents ansl wste eleas f re
contained in Specifications 2.3, 3.3 and 3.4. J 7
b. Environmental Samples FExceedinq Limits of Table 6.1-2
When the limits wh Table 6.1-2 are exceeded, refer to Spelyfication 6.1.b for rellroing
requirements.
:.Semiannual Radioeactive E~ffluent RIesese Report
Routine Radioactive Effluent Release Reports covering the operation of the unit during
th'i previous 6 months of operation shall be submitted within 60 days after January 1
and July of each year. The period of the first report shall begin with the date of
initial criticality. I . The UlJ~ioiactive Effluent Release Report shall include a summary of the quantities t
of radioactive liqluid and gaseous effluents and solid waste released from the unit•
using as guidance Regulatory Guide 11.21, Revision 1, June 1974. 'Measuring,. Evalusting, an 101tn ai a tviyi oi a t san ee s so e
Radioactive Materials in Liquid and Gaseous Effluents from Light-Water -Cooled V SNuclearf a PofPwer peniPlants", withB hrefj-data summarized, onea quarterly basis following th f.
Amendment No. 03 9, 25q
E• • OVEVWl • ff
JAFNPP
-•_•. 7.0 ADMINISTRATIVE CONTROLS / ee Z*TE •e c-t ,, (4
a. The Site Executive Officer shall have direct responsibility for assuring the operation of t(he James A. FitzPatrick Plant is conducted in such a manner as to provide jcontinulng)
p the environment and shall delegate in writing the succession to this
b. Implementation of the Radiological Effluent Technical Specifications is the responsibility of the General Manager - Operations, with the assistance of the plant staff organization.
PROCEDURES O
Written procedures n adr inistiyve pQj;',es shall be established, implemented and manandhteet or ex z'ea the require n" n ana reco )•endations of !,!cli~on 5 ,
r~clv diitative Picies and Proce ;res" of ANSI78. -1972 -nd F;( ulet 3 t 13 _ve- _ A _oendix _ . In addition-_ _st i
L ,, implemented and maintained for th P , ODCM, and Quality Control Program for effluent and environmental monitoring i.ng t-he uidance iRe ulatory e 4.1
7.3 REPORTING REQUIREMENTS
"a. Planned Liquid and Gaseou Releases V- e- - I S 0oT',M.
The limits for radioactive materials contained in liquid and gaseous effluents are
contained in Specifications 2.3. 3.3 and 3.4.
b. Environmental Samples Exceeding Limits of Table 6.1-2
When the limits of Table 6.1-2 are exceeded, refer to Specification 6.1 .b for reporting requirements.
c. Semiannual Radioactive Effluent Release Report
Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.
1. The Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit using as guidance Regulatory Guide 1.21, Revision 1, June 1974, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of
LRadioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", with data summarized on a quarterly basis following the format of Appendix 8 thereof.
Amendment No. -2 2 2 -5,
REVISION H
c) A4
JAFNPP
" 7.1 RESPONSIBILITY At '
a. The Site Executive Officer shall have direct responsibility for assuring the operation of the James A. FitzPatrick Plant is conducted in such a manner as to provide continuing
protection to the environment and shall delegate in writing the succession to this
resoonsibility during his absence.
b. Implementation of the Radiological Effluent Technical Specifications is the
responsibility of the General Manager - Operations, with the assistance of the plant staff organization.
7.2 PROCEDURES
Written procedures and administrative policies shall be established, implemented and
maintained that meet or exceed the requirements and recommendations of Section 5 "Facility Administrative Policies and Procedures" of ANSI 18.7-1972 and Regulatory
Guide 1.33, November 1972, Appendix A. In addition, procedures shall be established, implemented and maintained for the PCP, ODCM, and Quality Control Program for effluent and environmental monitoring using the guidance in Regulatory Guide 4.1,
7.3 REPORTING REQUIREMENTS
.The limits for radoactive materials contained in liquid and gaseous effluents are Scontained in Specifications 2.3, 3.3 and 3.4.
Sb. Environmental Samples Exceeding Urimits of Table 6.1-2
When the limits of Table 6.1-2 are exceeded, refer to Specification 6.1.b for reporting•
C. e n lRadioactive Effluent Release Reports -a /coveri 'ng Atchee
Radioactive Effluent RaisesRelports covering th~e operation of the unit during'-)
l, th " evious(p6 shallsbe subratte hfll be submittays aT. y January 1
rlld Wastes an'd I Effluents from L
ýd on a quarterly,
Amendment No. 03, 203 254
&-v.d. 10 c.... ... , Se•t,&", .
?Ate- 4 9REVISION H
"4e
J*2
[ý'I'
y a%,..: : :.-: ' , .= ;.-n . . . .. : .
meteorological data -:act¢ed <ver týhe prericýs year• tn e,
- meteorological data is not cluded, ENO sha7: ret.: In 4z e On provide it to the U.S..%u ear- .egl':cr-y ::--r.:s -cn.:on :eze-:_. This same report shall 4.cllade an assess=en= o• /.:e raca'ti::cn coses*
due to the radioactiv iquid and gaseous ef' ýen:s re-'ease .
the unit or station ring the previous cla dar year to the pub:'c.
Al assumptions us in making these asse ments ,i.e., s ---i--:
activity, exposu time and location) sh 1 be inc•uoded in tnese
reports. The sessment of radiation ses snall be perfc.ned in
accordance wi the ODCM.
3. The Radioa ive Effluent Release ports shall include any ch ge to
the PCP the 00CM made during e reporting period, as we as a listin of new locations for d e calculations and/or env onmental monit ing indentified by th and use census pursuant t
Spe fication 6.2.
4. a Radioactive Effluent elease Report to be subm' ted 60 days after January 1 of ea year shall also include assessment of radiation doses* to t likely most exposed me er of the public from reactor releas and other nearby urani fuel cycle sources (including doses om primary effluent pat ays and direct radiation) dunn the previous calendar ar, to show conformance
with!4 CFR 19, , Environmental Radiati Protection Standards !o: Nuclear we Operation. This asses ent of radiation doses Is performedi accordance with the 00 ..
5. The Rad' active Ef fluent Releas eports shall include the followi q info -ion for each class of lid waste (defined by 10 CFR 61)
shi doffsit during the r rt period:
(a Container volume;
(b) Total curie quan ty (specify whether determined measurement or stinate),
Cc)..:-Principal ra onuclides-(specify whether det ned by measuremen or estimate),
Wd) Source waste and processing employe (e.g., dewatered spent
resin, ompacted dry waste evaporato bottoms),
(e) T of container (e.g., LSA, T A, Large Quantity), and
(f) lidification agent or abso t (e.g., cement, Dow media,. etc.)
6. Radioactive Effluent Releas Reports shall include a list an
scription of unplanned role es, to unrestricted areas of radioactive materials in ga oua and liquid effluents made du ng the reporting period.
The dose assessment secti of the Semiannual Radiological ffluent Rease Report shall be submit within 90 days after January of each year as
addendum to the Semiannua Radiological Effluent Release R rt.
Amendment No. 03,268 67
Y_( T 6sl(� .4
REVISION H
•- • . ..
-- : -. . . , . -°....t.. -
00
Is
Q
LA Z
7.M Isdjagctiy Stff Ieat ReleaSe lOPert shal Contain the cause for uMAawGAlabALt7 0f 407 OMW Z"etAA SOMle requileb by Table 6.*1.1 andd shall identify the O4,catims for oht-Lala ropI at samles. This shall al.e laslude a ?r sed flgure(s) and table for C=4 reflectianI the now l ocatl ea(s). Re to spe cifica tion
at~s
8. * m asoactive, af fluent ease Report shall contLan loeations %INoiam tied La the laud use seams La acouiames with Rpoeific Lets *.a.b
g * - eac~~~tive gtff mat tolem Ras" shall caaateeot ecu to th asditLes which resulted La 1 curios for ftanks specifiled La the Unitinga Conditiose ter Opration. Sectioa 2.5l.4. 6LC
4. -nAl tailain 3wip sn l 2mernvimi Rmrt ii 3 ta ire in Ata SLIRPRep ~Orts s won the Operation of the unit
daring the precious Calendar yew sa"ll be sumtc
~,, ~ The Annual Radielogical lnwirerneatal Operating Reports shall include &,sf *'CA041 sumaries. interpretations. mo as analysis of tranms ot the resuits of the 00;4,is; M 00 raddlelregical, eaviroemntal survoillpAce activities for
ygmls.5 W-.i.6.z1 I3, Op. Weala trols (as propr1te) mand eavirea ytal suzi.3~llaace rejprts j .. free the pow Uwo ye am andm esomat of \the obse'q hapacta of
the p t Opera a sa avirernent. Mi roeper"shl Lals caluae rsults efth Us e9"I Cams re~~rd if I dAtLeA' 6.2
TeAsmaus Radiological Zawtrerneta Opermatig Reports shall, laclugs the esulats of sa"alyis of a11 rdLelogleal eany rmeatal samles and of all
amimailo 610 tso-e te of these eudrmso an Measuremnats La the format of the table La the Radiological Assessment grase Techaimal Position. Revilea~ 1. Noember 1379. Za the oesot that Sam Ladiwiduaml results art not waallablo for Lacluales ia the report, the report shaLl sets and explaia the reasens for the missing results. 2M missing data shall he subimitted "s seon as possible is a ouplemestaqy report.
The rports shall &10lalsoIcld the follow'Ief: 1% @Unmm descelptiem Of the ladielegleal remtal NeaL clapn Programs at least tw qe~l U aps* Coerinig aUll ing le im hoe t iv= "a ois nc and
fer goater =,,e f eatar th toef Apart~ip~o the laterlah" u Comarison. a required by Upee~ficatiea (or
Late Wa gre peopn )*mad &Lseses of all asaslys\ LA ieh8 cWs egal by Table 6.1-3 met routinely ewable.
the mee dietas statiess. thst a second I Incluade
)mae osnt se.eP 127 .
?&i~4 4v
REVISION H
4L+e EtcALj1o'