ML20150E397
| ML20150E397 | |
| Person / Time | |
|---|---|
| Site: | 05000054 |
| Issue date: | 12/15/1978 |
| From: | George K UNION CARBIDE CORP. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20150E400 | List: |
| References | |
| NUDOCS 7812190091 | |
| Download: ML20150E397 (2) | |
Text
~ '
p
)
, ->t a\ l
~1 k ' UNION CARBIDE CORPORATION h p MEDICAL' PRODUCTS DIVISION P.O. DOX 324 TUXEDO, NEW YORK 10987 '
T E LEPHONE: 914 351 2131
-I 4
December 15, 1978 U. S. Nuclear. Regulatory Commission Division of Reactor Licensing Washington, D. C. 2.0555 Att: Mr. Robert.W. Reid Chief, Operating Reactors Branch No. 4 Gentlemen:
R,e : Dbcket 50-54 Forwarded with this letter for Commission approval are multiple copies of the. final draft of the Technical. Specifications for the Union Carbide Research Reactor (License R-81).
+
These specifications represent the consensus arrived at in dis--
cussions with your staf f that were started following our Initial submission in December '1973, and that have continued until the present date. We are appreciative of'the many helpful'recommenda-tions that have been made by your staff in the intervening period and especially of' the recent assistance of your Mr. Dominic DlIanni. -
As these specifications will require in many instances a massive ,
revision of existing procedures, we will need a' period of grace l before achieving full compliance with the new specifications. It is therefore requested that compliance with the.new specifications be required within 30 days of the date of. Issuance of the imple- '
menting License Amendment, with however the exception of the fol-lowing:
- a. Sect.-3.8.1 : effective 4 months after-Lic. Amendment date
- b. Sect. 3.8.2 : effective-2 months af ter Lic. Amendment date
- c. Sect. 4.0 : effective 1 month after Lic. Amendment date
- d. Sect. 6.0 : effective 4 months after Lic. Amendment date Tng 00GuiST conimus '
L, FCOR QUhUTY PAGES
~""~~--
-781219009/ 4 a
. 3 ,, , . . . . - .6- . -, , ~ + , ,, . . + .-. ..-#
.k
i
- Mr.= Robert'W -Reid- Page 2.
U.. S.~ Nuclear Regulatory Commission December, 15, 1978 ;
c The required 22 copies are being forwarded in two separate envelopes,
- as indicated below.
- Very truly yours, y
UNION CARBlDE CORPORATi3N -
Medical Products D' vistors - i
~
/[ ,
,/ K. D. George . .
Senior Development Scientist- l KDG:aw
' Encs.: 4 copies cc: Mr. Dominic DlIanni, w/ encls. (18 copies).
STATE OF NEW YORK)
)
COUNTY OF ORANGE )
On' this 15 day of December 1978 before.me personally came. Kenneth. r D. George to me known~to be the individual described in and who executed the aforegoing instrument and acknowledged that he executed the same.
i l A. < QU Notary Public /
JtJT T \ 71. .TO' ' 21 MOTW P!: ' Et: o ef New Yod N , : t ? ?3 P.-r! ,I:ng 3 Or. - - Cwnty Ccmmw.un &;io Erch 30, 1779 I ,
I
\ '
I \
i
. .._.,.._.....~.-m. . . , . . _ , , ,, ,,
4.' ,
x .
UNION CARB1DE R E S E A R'C H REACT 0R TECHNICAL SPEC 1FiCAT10NS APPEND 1X A i
License No. R-81 Docket No. 50-54 g
Revislon Date: 12/12/78 !
7g
' }
t (
TABLE OF CONTENTS Page 1.0 DEFINITIONS.......................................................... 1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS...........,,....... 4 2.1 - Sa fety Limi t s of Reactor 0pe ra t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.2 - Limi t i ng Sa fe ty Sy: tem Se t t i ngs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.0 L I M I T I N G C ON D I T I ON S FO R C" E RAT l 0N . . . . . . . . . . . . . . . . . , , . . . . . . . . . . . . . . . . 9 3 1 - Reactivity Limitations......................................... 9
- 3. 2 - Con t ro l a n d Sa f e t y Sy s t e ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.3 - Radiation Monitoring Systems................................... 13
- 3. 4 - E n g i n ee re d S a f e ty Fea t u re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
- 3. 5 - L i mi t a t i on s on E xpe r i men t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.6 - Fuel........................................................... 20 3 7 - Pool Water Qua11ty............................................. 22 3.8 - Radioactive Releases (Airborne)................................ 22 3 9 - Radiological Environmental Monitoring.......................... 26 3 10 - La n d U s e C e n s u s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 3.11 - Bases for Envi ronmen tal Spec i f i cat i on s . . . . . . . . . . . . . . . . . . . . . . . . . 26 4.0 SURVEILLANCE REQUIREMENTS............................................ 31 4.1 - General........................................................ 31 4 . 2 - S a f e t y C h a n n e l C a l i b ra t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 4.3 - Reactivity Surveillance........................................ 31 4.4 - Control and Safety System Surveillance......................... 32 4.5 - Radiation Monitoring System.................................... 32 4.6 - Engineered Safety Features..................................... 32 4.7 . Re a c t o r F u e l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.8 - Sealed 1aurces.................................................. 33 4.9 - Pool Water..................................................... 33 4 . 10 - C o r e S p r a y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4 . 1 1 - F l u x D i s t r i b u t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 5.0 DESIGN FEATURES...................................................... 34 5 1 - Reactor Fue1................................................... 34 l 5 2 - Control and Safety Systems..................................... 35 5 3 - Rod Control System............................................. 37 l
)
4 TABLE OF CONTENTS (Cont'd.)
Page 5.4 - cooling System................................................. 38 39 ,
5 5 - containment System.............................................
41 5 . 6 - Fu e l S t o r a g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
41 6.0 ADMINISTRATIVE CONTR0LS..............................................
41 6.1 - Organization...................................................
6.2 - Procedures.....................................................
47 6 3 - Experiment Review and Approva1................................. 48 r
6.4 - Required Actions............................................... 49 50 65- Reports........................................................
6.6 - Records......... ............................................. 53 55
7.0 REFERENCES
.i-r- v -
i e A '
l.
1.0 DEFINITIONS ThetermsSafetyLimit (SL), Limiting Safety System Setting (LSSS),
and Limiting Condition of Operation-(LCO) are as defined in 50.36 of 10 CFR Part 50.
1.1 Safety Ct annel - A Safety Channel is a measuring or protective channel in the reactor safety system.
1.2 Reactor Safety System - The Reactor Safety System is a combination of safety channels and associated circuitry which forms the auto-matic protective system for the reactor, or provides information which requires the initiation of manual protective action.
1.3 Operable - Operable neans a component or system is capable of performing its Intended function in its nornal manner.
1.4 Operating - Operating means a component or system is performing its intended function in i ts nornal manner.
1.5 Channel Check - A Channel Check is a qualitative verification of acceptable performance by observation of channel behavior.
1.6 Channel Test - A Channel Test is the introduction of a calibra-tion or test signal into the channel to verify that it responds in the specific manner.
1.7 Channel Calibration - A Channel Calibration is an adjustment of the channel components such that its output responds, within specified range and accuracy, to known values of the paraneter which the channel measures. Calibration shall encompass the entire channel, including readouts, alarm, or trip.
1.8 Unscheduled Shutdown - An Unscheduled Shutdown is any unplanned shutdown of the reactor, after startup has been initiated.
1.9 Reactor Shutdown - The reactor is shut down when the negative reactivity of the cold, clean core is equal to or greater than the shutdown margin.
1 1.10 Reactor Operating - The reactor is considered to be operating whenever it is not subcritical.
j Revision Date: 12/12/78
"' A '
2..
2- .- ,
1.11' Reactor ' Secured - The : rea'ctor is secured when:
a'.' . The.c' ore contains insufficient fuel to attain criticality
~
under: optimum conditions of moderation and reflection,' or
- b. .The moderator has been removed,' or
- c. (1). ' Minimum-number of control rods ' fully . inserted as required'by Technical' Specifications, Land (2) The console. key switch is in the off position and '
the key is removed f rom the lock, and
'(3) No work' is' in progress ~ lnvolving core = fuel, core structure, installed control ' rods or ' control rod drives unless.they,are physically decoupled from the control rods, and (4) No In-core experiments are. being noved 'or serviced with a reactivity worth exceeding the maximum value allowed for a single experiment or one dollar, .whichever is smaller.
f
~
1.12 True Value - The True Value of a parameter is its actual value at anyl Instant.
1.13 Measured Value - The Measured Value of a parameter is as it appears on the output'of a measuring channel.
1.14 Measuring Channel - A Measuring Channel is the combination of' sensor, lines, ampli fiers', and output devices which are connected for .the ,
purpose of measuring the value of a parameter.
1.15 Reportable Occurrence - A Reportable Occurrence is any of those conditions described in Section 6.5.3 of this specification.
1.16 An Experiment - An Experiment is an apparatus, device or material, placed in the reactor core, in an experiment facility, or in line with a. beam of radiation emanating from the reactor, excluding devices designed to measure reactor characteristics such as detec-tors and foils,
- a. Secured Experiment - Any experiment, experiment facili ty, or component of an experiment is deemed to be secured,'
Revision Date: 12/12/78
s
-( o 3 ,
1
.or in a' secured position, if lt is held in a stationary position relative to the reactor ' core.- The restraining forces must be substantially ' greater than those to.which -
]
the experiment might be subjected by hydraulic, pneumatic,- '
or other forces which are normal to the operating environ- . ;
nent.of the experiment (or by; forces which can arise as
- a resul't of ' credible malfur.ctions) .
- b. Movable' Experiment '-' A movable experiment is 'one which _
may be inserted, removed .or manipulated while the re-actor is critical.
- c. Untried Experinent - is a single experiment or class of experiments that has not been previously evaluated and approved by the. Nuclear. Safeguards Committee.
1.17 Experiment Facilities . An ~Experinent Facility is any structure, device or pipe system which is intended.to guide, orient,Jpost-tion, manipulate, control the envi ronment or otherwise' facilitate a multiplicity of experiments of. stmilar character.
1.18 Control Rod'- A control' rod is a rod fabricated from neutron absorb-ing material which is used to compensate for fuel,burnup, tempera-ture, and poison effects. A control rod is magnetically. coupled to its drive unit allowing it to perform the safety function when the magnet is de-energized.
1.19 ON CALL means a senior o'perator is assigned and can be contacted within a short time.
1.20 Scram Time - is the elapsed time between the instant a limiting safety system set point is reached and the instant that the slowest control rod is fully inserted. ,
1.21 Potential Reactivity Worth --The Potential Reactivity Worth'of an
~
experiment is the maximum positive value of the reactivity change that would occur as result of intended or anticipated. changes or credible malfunctions that alter experiment position or-configura-tion.
Revision Date: 12/12/78
l i
4.
1.22. Safety Limits - are limits on important process variables which are i found to be necessary to reasonably protect the integrity of'certain physical barriers which guard against release of radioactivity. The principal physical barrier is the fuel cladding.
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit's of Reactor Operation 2.1.1 Limits in Forced Cooling Mode
- a. Applicability - This specification applies to the variables that affect thermal and hydraulic performance of the core during forced cooling. They are:
(1) Powe r i n MW.
(2) Flow in GPM.
(3) Height of water above the core.
- b. Objective - To assure fuel cladding integrity.
- c. , Specifications -
(1) The maximum steady power level under various i
flow conditions shall be as shown in Figure 1.
(2) The pool water level shall not be less than 20 feet above the core.
- d. Bases - The analysis given in Ref.1, Sec. A1, forms the j basis for this specification. The superposition method of Cambill is there used to derive the burnout heat flux as a function of primary flow rate. A safety factor of 1.25 is applied to allow for uncertainties in the corre-lation. Pool temperature (or core inlet temperature) is not included in the specification as this variable changes very slowly and has only a minor ef fect, e.g. , a 10*F change results in only a 5% variation in burnout flux.
The latter, however, is evaluated conservatively near the high end of the pool temperature range that is ex-pected in practice. A de-rating factor can be applied for pool temperatures in excess of 120*F. The relation-ship between total power and peak heat flux is derived Revision Date: 12/12/78 4
1 p- ,
5
~for the core situa'tlon'with the greatest peaking factors, viz. a new fuel element adjacent to a central in-core flux trap. Reactor power, primary flow rate, and water level will be maintained well within safety limit speci- i fications through Ilmiting safety system scram settings.
(see 2.2.1) . ]
i l
i 1
l i
l l
l i
Revision Date: 12/12/78 l
e.
" $O ~ - ~ - ~ - * - .-
- + - - ' " - _ -
" yq - -- - _ '- -* - - - 1 _
f
' - - - ^ - _- ,
7, *. .
1 en s
t _
~- - -
Po
- ~. - . -
-~ *
+ -
rr '
~ " *" ~-- _.. +~ .. - ~
-rw t e .
- * *~
- : t
- 4 MO w -+ "T' ' t
- "r +
- * ' t-
'" m:. ** ?- b4 t.,I '
- i?,
cit
,."II 4
" . . !'p h9Sd q +4M et+i
- l
" - - , -3 mem ',
j*T
- - 9r +I'ti " " t i, -
t mIiN ' t .,,i;i n
i
' L a ' '"'t m. t,< e
+
- r i
- t' - .
- 4 Mkt i
- a ~ I; '
- 4, tt in ;t **i'ti i i. i
':4 ? I' i 4:tcj!
, 7 it avft@h e
i WO mumm ';4 ji t,P h ' i ' ,t p ivi v"iIt i r t
,, tp h t 'r
- tT 6!h#%i h de f *' ]n r[ 1 by {
t4m.fh .1 l
- p im *j ~ n n;n{j tIf' ai ?t n nqif [? I,+ifi h nfs N +im
- -t fto1f rn iut .
t .
rno i
'eN,+j
' i
, "ii - r eI 'ti ;nie 4f
+
tt lIi ii nijH j It tf +
+ '
" jT H i fjiiu
+I!
f
}ii 5 +1t
+t} :inp I*I l t I t1 i t
- yi
- neGi ftN tj i+ dH
- j .
h gt{t i! f
+4 n 6t H.f tr; i. i t tii ftt 7 kT r eO a
tt i,l:
I
- %@ij t
i
. tfI ?[,i j]j
- iI "i fitf f 7imW 1t
- t h.:1 t
i ,J. t 7!
t7f
+I fh i
t 1iy 1t itl,.t
. r iliNp afy tt1t
+ 1 j
- i tt tyiH Ii it f +i
+ trio h'if tt - >r * " ! t fiti tIfI i .
t,iie ,;i,. +*t'ti H4y k jr ri tr ttm 7 t}f - - i i ft:[
tk iIt' ty .
t r({}
h]+7M 7 l In d. +1p -
t j t1l tf! ]- !tH tii ,it tun 1 fI} . ji fr.
tt rt li t
ypfif 1,iytth ij
+T tt fY
+T jt I tn I, '
, yH Iip ;ioi .l, p
- ii ifii yiri I i 4T { tttf t.jh .,
-tgf
'@t, i4 j
,ty} 44
~
- j. tt pO it iplii j
y]j,y t ; tl' pk i tt ttag t t i[i th tf t
i
}t kt 7}*T ip j jf j.- ' ,t l
e -
i 1T yi i4 p ] .- -,
frrh
{+i, ;
rftt tt T i t, ith
> I a
!mlid 0{ T trtf i
tty ~
Y9 I t t l
{i a
! i i b
M M,f }tpn itN i
jm,rh irtfti diTg vyt' -
' k +y- i4 j I
ep hL
- t{i t}i titl TTrm *rr
'iT +tT -
- r' - ft i -
- - ~
~
- f r1i {
l pir
- tT ~ j dtti g ~Ic i rTT ia' T
t1mta p. 1ii
+ t tt
!.-1 g} T^M -
1 f'- cf
~ - '
- h+ +4 t' +trt pTtH e?'f@
{
i tif 7N ytt rtH rm,yi i! -
g - ~ ~ '
- ,T - ' -
~d 1t,i T,f^
D rli tt
{-
V~L f 4._ff+++t 1 37f i ti t i
- ~- -T - - .' t+ ' - - - -
7- -f4
- j - -
ti ;^
6iI 1 } 1 t r
aw o ttf t t}
- - TTr - ' - -
~o 7itW t *1
- - n1+} ,t
, i.
f* t tt j tii 1t 4 i
- ftf n } ;}ji ;t, t n} ' }g[;ott
- i ,
+ilN
}i h
- ttt +t t
,if4 a p .1 jti 7nt mn4i t} j } r tt iI t f h'Hh::tnh y ii fi"m iIf b0 k 7 rp I i
- 7 t
- T ti 4i t } nI' ;t }'
1 _ i ii i tH fiy
,)
1y k T 7 t7 t l r
.d+t$ t4,;i i, t} !
jit, ,'ttfH b. .- d,- ti 7fi - ftt }[l ttft I
.It t' i j, &t1 -
m.iL{i M B t inl1 p t i pht{nf tt t t i + i
^ tiTt t tH ti Ii Yf
.ry1i , .e TfM+- '
+ 1 i, - j jr -
B fp, t h ti !i f j }f jt d5 t l' 4 jt
- ,- f tith ttt tIii tI
~ : tjii ti i fr t +.j W
Mp;7t ij j' 4-f% t.tmt!Q,77fr 1t y f4:i pf F: I --.{ i1iTg gp**
- I m 4 1T k F c, 1
ch -4 pTittrH r! - tT
~
t {tt
;dntt 1 1!
~ i' c%;fb g i -
tf'+ Ii1 - ~ O t 1 4} tri. ,l i 7vH fTQ tN;,,
'!I4 t +T ir H, I,{i ,7 b ~fb I i
e *yfQ {tt tr1 +-
- ' j t Tr + r~'-
f dI . rtP} ,{-tH 1
,7-r T' t f iT tt
}
i t i i1i$iid.' .m fT ' tjI I
+tt:
4 - -i
; ~i Ni4 i f- +
titk t k-i, +i, c
,}' 4 -
I}- i IIi- t. 4 - + tb !
=g - - - -
o
- - + + - :
s m
)t
'- - -
- e i trg4#
ti t e
^ \'Y_
. w D* ,
- b-- .
- r-L * ,
t 9*- 6 ** b
'i b GHN ,
a
, e "7
' - mr-
+Il I o ,
t O
"O' '
,r* ' , r1,1 a
C'u s m 4
~i ' ' .
i '.
'"i. -I r '
M" - ni tA
-;"j + (
-pr ,y' m; j + s- m!f%
- i
#- swS i t G ,t:, !! e t .
W0 N 34 o , o e phj.
'"*d h i!!I - ,
t'! i t- d
' 'I'
' 'h :n ]!if! +
- : n t
n
' A, t
* "n1 4
u '
#I t
Hax iaS ts s A nu Ji. P1 ", t
+'J
?t - o l4 l+i -
'?
'1t:
&}! It }
i
'y fi h i'n tjj i -
I',
' , I'fi, Ut t, i3 I t
1f ni0 t a""nH t s t'I i va3 1 ;t umiq i
-mL1 - ra ++3 t
. i '
rti*4f i
;imi r ,Ili "Iit hiJ i t
r;r}fN i,#
*f 6 t; t y;;; +H l= alii l t
iI t
- i 7 ,j f
,yi:
?ttt ri m.pi r e,j ttt i
D4A , i
- '+ t ttd tj1 .
t t ii:t iH t lt i - t jtii O3 123 t44 Mn a htHm ni t:
'*vd+. t t f ,4 ii
!f y7{
t;-}'t + Hief
!fnf i
*l
~
ii, t
+Tr tT *i t
H t [f i i 'i e, { 1 ttg ii 4h lu i.
'tf i
i i t.- H !i) : tI i Ii 7I%[! IIi Infd I i tnif d,f
)1i
}
i jv.!:p t tni t! t t t 4l 3a ' i.t i*U Itf; ]{ ?ia . -rt,+t,% Sm
.[tH IrI:
+p t tr - tIij t rft a i, :+ 7y i ir i tf -
rmi. ; rih} ?tr%.ti4i ti1i tt + tfi - 1 { . Tj + +YM h Q FMv4 ;yTf
; t{ f +
;p i iT'[i -
1!tN im 7 +gt i i
- t i I[ i i4 i 7t -
7 g t f-
, t t ; '
77 '!if 7d tr 4 t l}i l - - 1pg+ * ; ' tt,f tri. t h +7 7 !?: t . , b to 31
;m.:,
1i e Triq yMfR : i ' f -t7 i I' l IiMyQ, - +i 1;TrH tr - . !I:f , b tr-4 nth i
- ~ '
-+
1l1i M T tJ t i * : ii rr ! t7?I ,ti rHv{}d - ~ irt
- t,
*; - ttf ~
- Ir c ^ ~ ti rt *i ' ' Iti 'i
' - e tI;
/+ ;- r.
+!( YTf1
- t
*,, *-++
it% 1 ti 4 h *Tt tlii - i + d14Tq - - -
.' 7,uIf 3 '
irr rqvA
!3tf * - + + ]
el! ,
" t: ?
&p.-- H j!l
- ma ;r]i tI' 4.- 4 -i
[h'1 bi+j tt t 41 - r? f r- - rf' 4T + tiw + +- + - tI'3f tI' Ii t:p t'1
?t;e -
ii'i
/ -
r/ -
'n '-
tt;i 1ti '* t t . ri tiIi I
- i
- i I,} }t {
- ftH -I W'j!tfi b'M1 ' *g ,.t{
tr l nitg !*
- T't i i t -jm t +
- I' -
prH t 'f "t ; I ,. itw! k I i 1+7 f H! lfti I i fi:
,1 + t t hpt i i
- i
,p t td i1 Ii t] ji [tr p'fd tt k ,t r'}
t 4I ' ! l clM t +1i i Tu1 ttg 7 rt% N[fdi1l1f W T.'
!:'w' l'!.
) t ti t7 tT Ii 7 tt i i t .
.'pj i rlv!
T {p ' tptl t! t i!i ttii ! t f-t' i T I1f @p}! It , ip f;7i3 f/, 1 ,imli ftt T tj r! H t f4 i ;i i ' ft I t t trt; ', t ji
!. *f7 tl.f 1t -
Hai!i ' ftji I'T f r ta j 7i ' T tI y +j4 tifO *I Iii! E t, M tti1
,.r rW
? , t;i il, . 'Ii
!o!A !1 4;,t tl tft 2rft
- iit - !
t*m }'f
!?u! ta -
T t[i jnim T!i it 1; iiM r i t ;
!. tltI !I .j t1t+
t+fn y' 4 I It t i1j tt t::f jJti iii I)tH hr}
-!, Ii I
- Tt-i
,r j i ii1 j t! t f' ,j : !Iij tt 1 7
- FF
^
r.r,T ?i
-- c-1!!I tij1 ti14 tI'ii yi l I -
1ti -h - jj
- r}i -
~ -- -
I' i - - tr
] tty ' I iI' ~ (
i7~ A Ef4 1ft& 1f}f, t t{ t:
] - tt' ic k
- - 'i_- - - -
1i ^ ~ ~ p i
- l :t
' - - ~ Tttk -
i
'- - - - h# - T- ~
O8 o ~ 4 Oo doo Q S0O g a1 a hJo
- v# $ , u.
_ ,,_ C , g,,, uQ %nL. l !1 lll', '1 l
7. 2.1.2 Lim!:s in Free Convection Mode
- a. Applicability - This specification applies to the thermal and hydraulic variables affecting the core duririg natural con-vection cooling. They are: ;
(1) Power in MW. (2) Height of water above tne core.
- b. Objective - To assure fuel cladding integrity.
- c. Specifications -
(1) The maximum reactor power level shall be 6.7 MW. (2) The pool water level shall not be less than 20 ft. above the top of the core,
- d. Beses - The analysis given in Ref.1, Sect. A2, forms the basis of this speci fication. The homogeneous method of Gambill and Bundy, used in this analysis, has been employed successfully to predict natural convection burn-out in ORR and HFIR fuel. The former fuel is close in design to UCNR fuel. A safety factor of 1.24 is applied to account for random variations and uncertainties. A pool temperature near the high end of the operating range (120*F) is assumed. The safety system settings on power and pool level (2.2.2) assure adherence to these speci-fications.
l 2.2 Limiting Safety System Settings 2.2.1 Safety Channel Set-Points in Forced Cooling Mode
- a. Applicability - This specification applies to the set-points of the safety channels,
- b. Objective - To insure that automatic action is initiated that will prevent a safety limit from being exceeded.
- c. Specification - For operation in the forced cooling mode the limiting safety system settings are:
(1) Power level at any flow rate sFall not exceed 7 5 MW. Revision Date: 12/12/78
.5 . .)
s 'r g ,.
'(2) Power level settings 'for various con'ditions of ;
flow and of pool temperature shall be in accord-- ance with fig. 1.
,(3) ~ Coolant flow shall not be less than 1800 gpm- '
for powers above 250 KW. (4) Fool level shall not be less than 20 ft. above the top 'of the core.
- d. Bases - Safety ilmits have' been shown previously - (Sec. 2.1 and Ref.1) > to lie - at a low flow-to powa.r ratio. To pro-vide adequate assurance that these Ilmits are not approached too closely, the LSSS ~ are chosen conservatively so as to minimize the' chance of boiling in the core. This results in a much larger flow / power ratio. In Ref.'1,1 5ect. A3, power . levels derived using conservative correlations for ,
incipient boiling are tabulated for various values of. pool temperatures and flow rates to illustrate the result-Ing temperature margins. Through a comparison wlth experi-monts at ORNI. (ORR) this method is shown to be conserva- ', tive. To preserve- the desired temperature margins for i all combinations of variables, the LSS settings are a com-bination of two fixed set points, viz. scrams'at 7 5 MW and 1800 gpm, plus an adjustable on' that provides auto-matic power reduction at a setting that depends on the pool temperature and flow rate.- Rates of change of pool temperature are very slow - a few 'F/hr. at most - and thus allow adequate lead time for adjustment. , For a reactivity transient the case considered is the step ' insertion of 0.25 % 6 K positive reactivity with the reactor operating at a steady power of 7.5 MW. The - analysis given in Ref.1, Sect. B3, shows that the power at the-end of .75 sec. (the scram time, Sect. 3 2.1 - below) .will be no more than 11 MW. This is well below _ the safety limit for this mode of operation. h Revision Date: 12/12/78
. .s he No. automatic scram is associated wl th pool temperature as .this parameter varies very slowly allowing ample time for _ appropriate operator action.
2 '. 2. 2 Safety Channel Se't-Points'in Natural Convection Mode ,
- a. Applicabili ty, - This speci fication applies ' to the set-points of the sefety channels.
- b. Objective - To insure that automatic action' is initiated that will prevent a safety limit from being exceeded. ;
- c. Specification - For operacion in the natural convection mode, the limiting _ safety system settings are-(1) Power Level 1 250 KW.
(0) Pool Level b 20 ft..above the core. t
- d. B The set points are chosen to avoid boiling'in t- during routine ' operation with natural convec-ti allng. The analysis given in Ref. 1, Sect. A4, shows that a power of 0 35 MW is needed for incipient boiling to occur. To allow for uncertainties a safety -
factor of 1.313 applied to this, resulting in a safety system set point of 0.25 MW. The latter is well below the safety ilmit of 6.7 MW given above (Sect. 2.1.2). In the case of reactivity transient, a-step insertion. of 0.25% A K positive reactivity at an initial power levei of 0.25 MW will, following' the analysis of Ref.1 (Sect. _B3), result in a transient power of 0.38 MW af ter l 1 second. The latter is well below the safety limit of l- 6.7 MW for the natural' convection mode (2.1.2). l 30 LIMITING CONDITIONS FOR OPEPATION l ..
-3.1 Reactivity Limitations l
3.1.1 Shutdown Margin ,- The minimum shutdown margin provided by control rods in the cold, xenon-f ree cobdi tion with the highest-worth rod fully wi thdrawn t Revision Date: 12/12/78 ,
- - ~ _ _ . . .._ _ . . _ _ _ , . _ . . . _ _ . . . . . . _ , _ , . . _ , - . __ ,. .
.- . 10. and with the highest-worth non-secured expcriment in its most post-tive reactive state shall not be less than 0.5% A K. This specification ensures that the reactor can be shut down from any operating condition and remain shut down af ter ccol-down and xenon decay even if the highest-worth control rod is stuck in its fully withdrawn condition. 3 1.2 Excess Reactivity The core shall not be loaded with an excess reactivity of greater than 10.2% A K when located in the stall position and 8.2% A K when the core is located In'the open pool position. 3.1.3 Exper!ments Reactivity limits on experiments are specif'ed in 3 5 below. 3.1.4 Regulating Rod The integral worth of the regulating rod shall not exceed 0.6% 6 K. This ensures that a malfunction of the control system cannot make the reactor prompt critical. 32 Control and Safety Systems 3 2.1 Scram Time The scram tire shall not exceed .75 second and the control-rod magnet release time shall not exceed .05 second. In the translent analysis (Ref.1, Sect. B3), these values were assumed. 3 2.2 Heasuring Channels The minimum number and type of measuring channels operable and providing information to the control room operator required for reactor operation are giver. as follows: Channel No. Operable Operating Mode in Which Requi red
~ Power Level (normal) 2 All Power Level (in te rmedia te) 1 All Period Channel 1 All Revision Date: 12/12/78
, , 11. Channel ik) . Ope rable Operating Mode in Which Requi red Count Rate i Startup Coolant Flow 1 Forced Cooling Core 6 T 1 Forced Cooling Rod Pos i tion 1/ rod All Pool Temperature 1 All Pool Level 1 All Note: a. Operable below 50 W. Bases - The normal power level instruments (" Level Safe-ties") provide redundant information on reactor power in
- the range 25%-150% of the' normal operating power level of 5 MW.
The intermediate power level instrumen'. (" Log N") provides usable reactor power Information in the logarithmic range 4 10 %-300% of the normal power of 5 MW. The count rate channel covers the neutron flux range from 4 the source level (% 1 cps)'to 10 cps on a logarithmic scale. It enables the operator to start the reactor safely from a shutdown condition, and to bring the power to a level that can be measured by the Log N instrument. Coolant flow rate and A T instruments allow the operator to calculate reactor power and calibrate the neutron flux channels in terms of power. Rod position Indicators show the operator the relative positions of control rods, and enable rod reactivity calibrations to be made. Pool temperature information allows the operator to adjust the cooling system to keep pool temperature within a pre-ferred range, and to adjust the overpower reverse set-point (see 3.2.3) . Revision Date: 12/12/78
12.
.3.2 3 l Safety Channels The minimum numberrand type of channels providing automatic action -
. that are required' for reactor operation are as followsi Chan'nel - No. Operable Function Operating Mode - l
- Power _ Leve11(normal) , : 2 Scram @ > 7 5 MW All- ,
Power Level -(intermediate). l' Scram A < 3 sec. period 'All
-Reverse @ < 10 sec. period All inhibit @ < 30 sec. period- All: t c.
Reverse All Count Rate, l' ' Inhibit @ < 2 cps 'Startup j Inhibit @ < 30 sec. period 'startup.. ! Pool Water Level l' Scram @ < ' 22 : f t. All . Pool. Temperature -1 Alarm.@ > 120*F All 6 Coolant' Flow 1 Scram @'<-1800 gpm Forced Ci rc. - Manual Button ~ 1 Scram All Bridge Lock I ' Scram. All-Guide Tube Lift 1/ rod Scram. All , Flapper Valve 1 Scram (above 250 kW w/ valve open) All Keyswitch 1 Scram All j Notes: a. . Ope rable below 50 W. b.- Operable ~above 250 KW.
- c. Overpower' reverse set-points shall be set so.that the- i relationship of pool temperature, flow and power -levels ,
i shown in Figure 1.are never exceeded. , Bases - The' power level scram provides. redundant auto- l
.matic protective action to prevent' exceeding the safety ;
limi t ' on. reactor 'powe r. , The period ' scram, assisted by the intermediate level i period reverse and rod inhibit, limits the rate of ! increase in reactor power' to values that are contro11 - l abic wi thout reaching excessive ' power levels or tempera- , e ? ture. .These functions are not limiting safety system settings. p I' ' l , Revision Date: .12/12/78 ~ , 4 dE.
._,,_.-.y=.,...,...eJ.. -.-,-.--,..,o--. d<-,.,,....,h , , ,..,.-_w. ., , . . , r'- - - y
.' a .t
- 13.
The two inhibi ts on the count rate channel prevent inad-vertent criticality during cold startup that could arise-f rom ' lack of neutront information.or f rom too rapid re- ; activity insertion by control ' rods. . , The scram on pool level provides an adequate head of water above the core and guards against11oss of coolant
~
and Lloss of, building containnent. The overpower-reverse on the intermediate power channel provides. automatic action to-reduce power and minimize the chance of incipient boiling-in the: core.
-The' coolant flow and flapper valve scrams ensure adequate coolant ' flow to prevent boiling in the core.
~
The scrams on. bridge lock and guide . tubes prevent unplanned l reactivity changes that could occur through core and con-- trol element movements respectively.- The keyswitch scram prevents unauthorized operation of ; the reactor. Bypass is permi tted on those parameters that can be moni-
.tored by alternate means if the initiating circuit mal-functions.
33 Radiation Monitoring . System _s The minimum acceptable monitoring instrumentation required for reactor operation is as follows: Type No. Operable Function i Excursion Monitor 1 Detect high radiation: ! Alarm and isolate at- > 5R/hr. Exhaust-Duct Monitor 1 Detect particulate,. gas'and (" Stack Monitor") lodine activities; alarm in Control room. Building CAM 'I Detect particulate. activity in reactor building; alarm.
.l Revision Date : 12/12/78
,- _ . . - . . , . . . .. _ . ~ _ - , . . . - -
.t_ &
j g, Type No. Operable Function-Fixed Area Monitron 3 Detect radiation (y) in key-locations; alarm in Control room.
. Evacuation Swi tch . I Alarm and initiate evacuation' -
sequence. (manual). Note: For' maintenance or repair, required radiation" monitors (except for excurst on monitor) may be ' replaced by portable or substitute instruments for periods up to 24 hours pro-vided'the function will still be accomplished.- Interruption for bhief periods to' permit checking or calibration is per-missible. i 3.4 Engineered Safety Features , 3 These specifications ' apply to required equipment for the confine- [ ment of activity thmugh ' controlled release of ' reactor building air to the atmosphere. 3.4.1 Excursion Moni tor
- a. Specification: see 3.3
- b. Basis - This monitor-senses excessive radiation at the reactor bridge and automatically initiates the "evacua-tion sequence", which consists of a distinctive alarm, closure of damper valves in the building ventilation system and hold-up tank vent, and starting of the emer-gency exhaust fan (see 5 5.2).
l 3.4.2 Emergen'cy Electric Generator
- l. ,
- a. Speci fi ca tion l
I Equipment No. Operable Function ; Electric Generator 1 Upon loss of utility powvr, I
- j. start automatically and. l supply emergency exhaust fan ll and ventilation system controls. l l-t- i Revision Date: 12/12/78 i
15
- b. Basis - Controlled release confinement requires the ability to run the emergency exhaust fan and to close building damper valves. The latter are pneumatically-operated but are electrically-controlled.
3.4.3 Containment
- a. Specification (1) The emergency exhaust fan shall be capable of sustaining a negative pressure within the re-actor building of at least .01-in w.g. at an exhaust flow rate of not greater than 200 cfm.
(2) Filters in the emergency exhaust shall be HEPA and charcoal with tested efficiencies of 99 5% for particle removal and 95% for iodine removal respectively. (3) Depth of water in the canal shall be at least 10 ft. This is equivalent to a water height above the core of 22 ft. (4) At least one door of the double airlock doors and the truck doors shall be closed while the reactor is operating.
- b. Bases - To effect controlled release under accident con-ditions of gaseous activity present in the building at-mosphere, a negative pressure is required so that any building Icakage will be inwarc. Reference 1 (Sec. C, 2) contains an analysis of a hypothetical accident resulting in release of airborne activity to unrestricted areas.
The assumed exhaust rate is 200 cfm and the filter effi-ciency for elemental iodine is 95t. In the design of the containnent building (5.5) the water _ seal in the canal is effected when the water depth is > 10 ft. 35 Limitations of Experiments Revision Date: 12/12/78
- . - 16.
3.5.1 Experiments
- a. Applicability - This specification applies to those experi-ments installed in the reactor and its experiment facilities.
- b. Objective - The objective is to prevent damage to the reactor or excessive release of radioactive material in the event of an experiment failure and also to prevent the safety limits from being exceeded.
- c. Specification - Experiments installed in the reactor shall meet the following conditions:
(1) The potential reactivity worth of all experiments that could affect the core reactivity shall not exceed 2% A K. (2) The total potential reactivity of all non-secured experiments shall not exceed 1.7% 4 K. (3) The reactivity of any single experiment shall not exceed 0.5% A K. (4) An experiment worth less than 0.25% a K may be moved when the reactor is critical. (5) An experiment worth more than 025% 6 K but less than 0 5% 4 K may be moved with the reactor sub-critical by at least 0 75% A K. (6) All material to be Irradiated in the reactor shall either be corrosion resistant or encap-sulated within corrosion-resistant containers. (7) Where failure of the pressure-containing walls of an experiment container can cause a hazard to personnel or to the reactor, the container shall be designed and tested in accordance with the appilcable pressure vessel codes. (8) In-core experiments exposed to reactor water shall be designed to prevent surface boiling. Revision Date: 12/12/78
. . = ,
'I
.- -i.
"s, ...
(9) ~ Experimental appara'tus, material, or equipment to be inserted in the reactor shall'not inter- t fere with the safe operation'of the' reactor. (10) The total primary coolant flow utilized by alllIn - core experiments shall be limited to the same asi s that in six standard fuel elements. , (11) Experiments on the grid plate extension are
~
~ limited to a total reactivity of 0.2% A K and a total. load of 100 lbs. ,
(12) Each class of experiment ~ Irradiated in the reactor , must have been previously reviewed and' approved by-the Nuclear Safeguards Committee (6.8). 4
- d. Bases (1) See Ref. 2, Sect. G6.
~
(2) It is shown in Ref. 3 that the reactor can safely self-limit a step reactivity insertion of $2.14. This corresponds to an insertion of 2.14 x .81 = , 1.73% 6 K. (3) The method of Ref.1, Sect. 83, shows that a step insertion of 0.5% a K.with the reactor critical at 5 MW (or 0.25 MW, in. natural con- - vection mode) will result at the end of .75 sec. j in a power of not more than 14 MW and .4 MW, for natural convection. Each of these power levels ~does not exceed the corresponding' safety ; limit. , (4) Simliarly it is-shown that a step increase of-0.25% A K will produce a power level at the ; end of the scram time that'is much less than the safety limit in either mode of operatior in addition .25% 6 K is well within the auto-notic con t rol capabili ty of. the reactor cont-- rol system. Revision Date: 12/12/78
18;
'(5) This specification ensures that,'even with'a 30%lerrorcin estimation-of the reactivity of
- . an experiement, ~ the reactor will not tnr made
' critical. Even if the reactor were cri tical, the resulting period (S 3, secs.)Lwill automatically initiate corrective control action.
'(6) This requirement guards against release of- acti-
~
vation products in the' primary coolant or chemi-cal interaction with core components.-
-(7) This specification ensures that there will-be no. ,
mechanical damage- to the reactor core nor hazards to personnel due to' failure of experiment con-tainers where pressure exists or builds up dur-Ing i rradiation. In the case of fueled experl-- ments, it further ensures, against hazardous and-uncontrolled release of fission products into the reactor building or the environment from the . same cause. (8-9) Ensures that no physical or nuclear interference with the safe operation of the reactor will. occur. ;
~(10) This condition is assumed in the analysis given in Ref. 1, Sect. A.
(11) These limits ensure that movement of these ex-periments will not-result in reactivity changes ' in excess of that in Sect. c(4) above. , (12) Ensures that all experiments are evaluated by an' independent group knowledgeable in the appropri-ate fields.
' 3. 5. 2 - Fueled Experiements
- a. Appilcability - These specifications apply to experiments contain ng nuclear fuel that.are installed in the' reactor or i ts e xperiment facilities.
Revision Date: 12/12/78
4 * , 4 19
- b. Objectives - The ' objective is to prevent damage to the reactor, prevent excessiw release of fission products in the ~ event of an experinent failure, and also to en-sure that safety limits are not exceeded.
- c. Specifications - Fuel-bearing experiments in the reactor shall meet the following conditions:
(1) All fueled experiments are to conform to the specifications listed above in Sect. 3.5.1. (2) The inventory of solid fuel bearing material being irradiated in the reactor core at any time shall be Ilmited to 200 gms of source and/or 750 gm of special nuclear material. (3) The inventory of solid fuel-bearing materials in a single irradiation capsule shall be limited to 200 g of source and/or 50 g of special nuclear material. (4) The fission power of an irradiation capsule containing special nuclear material shall be Ilmited to 13 KW. (5) The lodine inventory of a single capsule shall be limited to 500 curies 131 1 dose equivalent for a doubly-encapsulated capsule and 70 curies 131 1 dose equivalent for a singly-encapsulated capsule.
- d. Bases - These specifications place limits on the fission product inventory in a fueled capsule such that capsule failure and the hypothetical release of all contained !
fission products to the reactor coolant will not result in excessive exposure to personnel on and off site. I The detailed analyses that form the bases of this spect-i fication are given in Ref. 1, Sec. C 3 The total amount of special nuclear material permitted in the core at any time has been increased to 750 g. This increase does not Revision Date: 12/12/78
.3.
20. affect the consequences .of- the release from a single- , capsule as analyzed'in Ref. 1:for It has.been estab-lished' (see License' Amendment No.- 10). that failure of a single cap'sule wil1~ not' initiate failure in other - neighboring capsules. The core limit of 750 g is bn ed-on approximately 15 Irradiation positions,.cach holding. 50 g of SNM. The limit of 15 positions-is; dictated by . availability of primary cooling capacity. - 3.6 Fuel Applicabillty -'These specifications apply to.the number-
~
- a. -
and condition.of the fuel elements present In'the core.
+
~
- b. Objective - To ensure that power Is distributed in the-core among a sufficient number of fuel elements to ,
.. k avoid excessive peek / average ratio, and to avoid ex-cessive release of fission products.
- c. Specifications (1) The minimum number of' fuel elements in the core shall be 30. Each control element shall count as 1/2 fuel element for this purpose.
(2) Control rods shall be kept'within 10% of their mean position whenever the reactor power exceeds 500 KW. (3) Fuel elements exhibiting release of fission products due to cladding rupture shall, upon positive Identification, ce removed from the core. An increase in the normal gaseous fis-slon product release (due' to system contami-nation) by a factor of 100 shall constitute initial evidence of cladding rupture and. , require IJentification of the cause. (4) Fuel element loading and distribution in the , core shall be such that peak / average thermal f flux will not exceed 3 3 Revision Date: -12/12/78
. < 21.
(5) The fuel plates are composed of enriched uranium-aluminum sandwiched between high pur-Ity aluminum clad. Fuel' plates may be fab- I ricated by alloying the uranium-aluminum or by the powder metallurgy method where the starting Ingredients (uranium-aluminum) are in the fine powder form. Burnup of the fuel 21 assemblies shall be limited to .94 x 10 fission /cm . Fuel plates may also be fabri-cated from uranium ox!de-aluminum. (U3g 0 -A1) using the powder metallurgy' process and the 21 burnup shall be limited to 1.5 x 10 fission / cm3 .
- d. Bases (1) A minimum of 30 elements is assumed in the analysis'given in Ref. 1, Sects. A 1, A 2.
(2) This specification minimizes flux tilts that could cause concentrations or shif ts in power distribution across the core. Such shifts are only significant in power operation, and thus this limitation is restricted to power levels above 10% of the normal 5 MW. (3) Release of fission products from the compact fuel plates used in this reactor (Sect. 5.1) due to a localized cladding defect is con-fined to the defect locality. A relatively small defect thus cannot release large quan-tities of fission products. There is a normal small and varlable background of fission product release due to uranium contamination in the coolant and on fuel plates. It is thus-safe to specify a recognizable and substantial increase in this background as a possible indication of cladding rupture. If the Revision Date: 12/12/78
{*,
~
22.
< t - ..
l;
~ '
rupture. were. extensive, there woul'd be no doubt
' +
at all of this condition. t
)
(4)1 This' peak / average value Is used in the Ref. I analysis. (5)1 Amendment No. 12 and Ref. 6. a 37 Pool' Water' Quality
- a. .' Applicability - This specification appliesfto. primary coo, ling. system water ..in contact with fuel elements. l b.' . = Objective ~- To minimize corros ton of the aluminum: cladding :
of fuel plates 'and' activation.of dissolved materials'. - ll
- c. Specifications-
.(' 1) Pool water temperature wil.1 not' exceed '130*F. -
(2) Pool: water specific resistance is to'be not less than 200,000 ohm-cm,.except that for periods'not greater than 14 days it may be ;
+
- 3 70,000 ohm-cm.
t (3) The pH o" the pool water shall normally be maintained between 5.0 and 7.5 3.8- Radioactive Releases (Airborne) l s 3.8.1 Ai rborne Stack' Release Limi t I
?
. Maximum yearly release rates for noble gases, radiolodines and f particulates of half-life greater than eight days shall be limited
- by the following expression: '
{Q g (X/Q)/MPC,--< 1/6 5 where-Q; = The average -yearly release' rate of radionuclide, I, In gaseous effluent from the stack in CI/sec. - i MPC,
= Activi ty concentration of radionuclide,- i, as given in Table II, Column 1 of Appendix B to 10 CFR 20, i J in p Ci/cc.
, . Revision Date: 12/12/78 yi., ym.,- - .= r,e,5 -wr,-3 a.w ,e , w ,--e, w my, i,w--,,, ara r s vre es e W atre-1rz w er.a- a se a# Pee e-m> ka be'* ar -r- sr w or+v= ar 4 4 e er T- wev* +
;p . . . . . . . .
; ., 0: ,
" 4 23, l
J
.X/Q: = Shall be calculated from measured values of iodine ,
concentration sampled at or above the tree line 380 e
-eters NE of the exhaust stack. P
--3 ACTION: Should the ilmit'of this Section be exceeded the' licensee ,
!shall notify the Commission within 24' hours, and take action to reduce the release to wit' In the limits Innediately. ,
8 3.8.2 Body Dose In Unrestricted' Areas
- a. The total' body dose to any Individual in unrestricted areas due to noble gases released in gaseous effluents v from the site shall be limited to the following expres-sions: ,
l I) During any calendar ~ quarter:
-8 '
3 17 x 10 r M i G74) QI-< 1 2.5 mrem
- 11) During any calendar year: ,
r n 3.17 X.10" I M; (i74) Q, I- 5 mrem i where: s Q,
= The release of noble gas radionuclide, I,- In pCI. Releases shall be cumulative over the j calendar quarter or year -as . appropriate.
M; = The total body dose . factors duel ot gamma emissions j for each identified noble gas radionuclide, mrad / -; year per pCI/m3 from Table B-1 of Rev.-1, Reg. Guide 1.109 i i A
~1 , , Revision Date: .12/12/78 , . . . . . . . , , .~ . _.m... , . . . . . .-,-.. ..- -- . - . . - , .
t 4 - s 24. X/Q -= Shall be calculated from measured values of iodine concentration sampled at the environ-mental monitoring station in Laurel Ridge. This measured value shall be increased by a factor of 2 when calculating the body dose. limits. ACTION: With the calculated air dose from radioactive noble gases in gaseous effluent exceeding any of the above limits, prepare.and submit to the Commission and New York State Department of Environmental Conserva-tion, within 30 days, a special report which identi-fles the cause(s) for exceeding the limit (s) and define corrective actions to be taken to reduce the releases.
- b. The dose to an individual from radiolodines in gaseous effluents released to unrestricted areas shall be Ilmited to the following expression:
i) During any calendar quarter:
-8 3.17 x 10 I (R WQ) 5 7 5 mrem, and
- 11) During any calendar year:
3 17 x 10' I (R; W Q;) 5 15 mrem i where: Q, = The release of radiolodines in gaseous ef fluents, I, in pCi. Reicase shall be cumulative over the calendar quarter or year, as appropriate. Revision Date: 12/12/78
'. * , 25 -l 1
~
3 lR[=The; dose. facto'rfor'eachradiolodine, i, inmrem/yr. LperfpCl/m3 (from Reg. Gu1de 11109 or calculational I values accepted bv the Commission). ] WI ..= .The annual average dispersion parameter for es'timat-
]
ing the dose to an Individual in the controlling ,
, location from radiolodines In gaseous effluents-re-Lleased to unrestricted areas.
TW = , (W for the'. Inha1'ation pathway, in - sec/m ' (as determinid 'in 3.8.2.a)
~
T/ =. (D/Q) for the food pathways,. In meters ACTION: With~'theLcalculated dose from the release of radio-
'lodines exceeding any of the above Ilmits, prepare
~
and s~ubmit to.the Commission'and N'ew Iork State Department of Environmental Conservation,'within .
- 30 days,'a Special Report which identifies the cause(s) for, exceeding tne limit and defines the.
corrective actions to be taken to' reduce the: releases. NOTE: The present controlling dose pathway is via Infant inhalation at the Laurel Ridge Residential site.
- If the. Land Use Census (Section 3.10 of this speci-fication) Identifies a location.or pathway which yields a calculated dose or dose commitment greater than via the presently. calculated dose pathway, the '
dispersion' parameter (X/Q or D/Q) and dose factor. '
~
(R;'): for this more restrictive pathway shall be used {
.in this specification.
l 4,
=t Li i
Revision Date: 12/12/78 4
'a.. * '; .
.: . '26.
- . , i i
'3.91 Radiological Environmental Mc'nitorino.
The . radiological environmenta'. moni toring' program shall be -con -
. ducted,as specifified In-Table 3 9 1. The;results of' analyses ;
performedf on 'the radiological environmental monitoring samples Ishal1 be-summarized in an Annual Radiological Environmental.
~
Report. 3.10 Land:Use Census ' A land"use census shall be conducted'at 'ieast once perf12' months between June 1st and Oct. Ist, and shall Identify the location of the nearest.mlik'aninal, the. nearest residence and the nearest garden :of greater. than '500 square feet producing . f resh, ;1eafy
; vegetables in each' of the 16 meterological: sectors within a dis-tance of five. miles. ,
- Action With a land use census identifying a location (s)' which yleids a calculated dose'or dose. commitment greater than _ at a location
~for which ' dose is ~ currently being calculated in Specification 3.8.2.b and from which samples are currently being obtained in
~
accordance,with Specification 3 9, prepare and submit to the Commission and the New York State Department of Environmental Conservation, within 30 dayr, a Special Repo'rt which identifies the new location. The new location shall'be'added to the radio-logical environmental monitoring program within 30 days. The sampling location having the lowest calculated dose or dose com-mitment (via the same exposure pathway) may be deleted from this monitoring program af ter (October 31) of the year in which this land use census was conducted. 13 11 - Bases' for Envi ronmental Specifications - a; Speci fication 3.8.1 is provided to ensure that the dose at the exclusion area boundary from gaseous effluents
- k. .
I 4 Revision Date: 12/12/78 4 ~$
,* . - - - . _ -v,.-,,o w . + *r -. . ,+m . . ,m - - -m- .* , - ,
u - - + 27.- from-the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The specified release rate limits restrict,. at all times, the cor-responding gamma and beta dose rates above-background to ~ ' an Individual at or beyond the restricted area boundary to 5 (500) mrem / year to the total body. The'se release rate limits also restrict, at all times, the correspond-ing thyroid dose rate above background via the inhalation pathway to 5 1500 mrem / year.
'b. Specification 3.8.2. ls provided to demonstrate compliance -
with 10 CFR 20.1(c) which requires rei6ases of radioactive materials released to unrestricted areas' to as low as rea-sonably achievable. The action statements provide the operating flexibility and at. the same time impicment the design objective of minimizing the release .to unrestricted areas ' to as low as reasonably achievable. The specifi-cations for noble- gas releases are based on111miting the total body dose at the limiting populated area.to less than'5. millirem /yr. The specification for radiolodine is based on'the assumption that the limiting; dose path- - way for these radioisotopes is via infant inhalation at the Laurel Ridge Residential Site, and limits the-infant thyroid dose to less than 15 millirem /yr.
- c. The radiological monitoring program required by spect-fication 3 9 provides measurements of radiation and radioactive materials in those exposure.pathwaysLand for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitor- ,
Ing program by' verifying that the measureable concentra-tions of radioactive materials and levels of radiation ; are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure
)
Revision Date: 12/12/78 v--+'s s +---e
& s y }
4 t.1 ....' <
;28..
1 pathwaysi This ' monitoring program may cha,nge based on-
~
pcrational experience and.results of'the. land;use census..
- d. Specification 3.10 is provided to ensure that-changes.
Ih the'use of unrestricted areas are identified andi that mobifications to the technical' specification : limit'
~ "
of dose via the most:restric'tive dose pathway land'the I monitoring program.'can be made if required by,the.results.
~
- of:this census.
l 1 t t . t i , 69; J Revision.Date: 12/1'2/78
- r. .
c - t
~ -
7 i ..
^
TABLEL 3.9.1 l' , .
- ^ RADIOLOGICAL ENVIRONMENTAL' MONITORING PROGRAM L
Number of' Samples Exposure. Pathway and Sampling.and. _ Type and Frequency; , 4
- and/or- Sample' Sample' Locations Collection Frequency of Analysis j l.' AIRBORNE
[ a. Radiciodine 1 Sample f ran 380 Continuous operation ~Radiciodine* canister. P - and' . meters NE of of' sampler with sample.
- Analyze at least = once Particulates. s tack.' collection as required . per 7 days for;I-131. -
F by dust loading but at 1 Sample from Laurel .least once'per 7 days. Particulate sampler.
. -Ridge Area. -Analyze for gross beta-
- I radioactivity 2 24 hoursi
[ following-filter change. Perform gamma isotopic
' analysis on each sample t .when gross beta activity
~Is > 10 times the.mean ofL
[ control.' samples forHany' 1 medium. Perform =. gamma ' isotopic analysis on. composite (by 1ocation)
~
l , t , sample at least once per-- 92' days..
- n . 12. DIRECT RADIATION. Same as #1 above.. At least-once'per 31: Gamma dose. At11 east
! LE days. once per 31 days.-
-m
~
1: o At least'once per 92. Gamma dose. At least L: ? days. '(Read-outi once per 92idays.~ < E-
~
! frequencies =are deter- [' LE mined:by. type of dosi--
- meters selected.) ' U$
c ; - ! .- .v -- t: h 12s. e _ n- - _ _ _ , _ ' r _ _s ,_._ s-. mr- .w & - m'- w, -
,, 4 m -
,__.;_.-'3 ea-.--.__ w-r-- -s #-~v +,1 r-a--,
4 3 TABLE 3.9.1 (Continued)- RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM-i Number of Sampies - -
. Exposure. Pathway. and .
Samp1tng and Type'and Frequency ~' and/or~~ Sample Sample' Locations Collection Frequency of Analysis i INGEST 10N' Food Products Location to be At time of harvest.: : *I-131 analysis . u. determined from One sample of broad- 1 Land Use Census.- . leaf vegetation. 4 i t . 8 5
-~w g ,_
1 r.
~
8 i s'. , 5 *The maximum values for the lower limit for'I 131 are 7 'x' 10-2~ Ci/m3-airbornei concentration and 60-pC1/Kg,. wet weight leafy vegetables.
~
G $'
~ -
5 W
=
c ., m,- , .v,,e -w -
.a , n,r, n,, e e - , +a w g. -wn --su -
31. 4.0' SURVEILLANCE REQUIREMENTS
. 4.1 General The requirements IIsted below generally prescribe tests or in-spections to verify periodically that the performance of required systems is in accordance with specifications given above in sectionsf2 and 3 In all instances where the specified-frequency'
-Is annual, the interval between tests is not to exceed-14 months; when semiannual, the interval should not exceed 7 months; :when monthly the interval-shall~not exceed 6 weeks; when weekly the
-interval shall not exceed 10 days; and when daily the Interval shall not exceed 3 days.
4.2 Safety Channel Calibration A channel calibration of each safety channel shall be performed annually _(see Sect. 3.2.3) . 4.3 Reactivity Surveillance (1) The reactivity worth of each control rod (including the regulating rod) and the shut-down margin shall be determined whenever operation requires a reevalua-tion of core physics parameters, or annually, which-ever occurs fi rst. The rod worth will be determined using the reactivity period or rod-drop methods. (2) The reactivity worth of an experiment shall be esti-mated, or measured at low power, before conducting the experiment.
.(3) Boron / Carbide rods shall be gauged quarterly and any dimensional changes reported promptly to the'Com-mission. Si lve r/indi um/ Cadmium control; rods shall be gauged annually, or, in the case of newly installed b rods, at the end of the first six months. If any Ag/-
In/Cd rod should be found not to meet the acceptance criteria it shall be removed from service, in addi-tion all other rods manufactured of the same batch shall be inspected. Revision Date: 12/12/78
.._m
'l
. '32.
'4,4 -Control and Safety System Surveillance l (1). The.' scram time shall be measured annually. If a control rod is~ removed from the core temporarily,.or if a new.
rod is installed, its scram time shall be measured before
~
reactor operation. If the' bridge Is moved, the scram time will be~ measured before subsequent reactor operation.-
.(2) A-channel test ~of'each measuring channel in?the reactor safety system shall be performed monthly or prior to each reactor operating ~ period whichever occurs first unless the preceding shut-down period is 8 houri 3r less.-- A channel test before startup is, however, required on any channel receiving maintenance during the shut-down period.
(3) A channel check of each measuring channel . (except for the pool level) In the reactor safety system shall be per-formed daily.when the reactor is in-operation. 4.5 Radiation Monitoring System (1) The excursic stack, and area monitors shall be call-brated annually. (2) The excursion, stack, and area monitors shall receive a channel. test monthly. (3) The excursion, stad, and area monitors shall receive a channel check daily during reactor operating I'eriods. 4.6 Engineered Safety Features
' 4.6.1 Excursion Monitor: see above 4.5 4.6.2 Emergency Generator (1) ' The -ability of the emergency generator to start, to run normally, and to generate 440 VAC shall be checked weekly. ) H Revision Date: 12/12/78.
1 33- j (2) The generator :shall be tested for its ability to accept,
~
'via'the automatic transfer' switch, the reactor electrical q
load once every six months. LA commercial power outage and subsequent pickup of load by the.. emergency generator wl11 count as a successful load test.
- 4. 6. 3 ' Containment (1) The efficiency of the charcoal filters and of the abso--
' lute filters in the emergency exhaust system shall be measured annually and the flow rate verified.
(2) The operability of the evacuation alarm and containment isolation. system shall be tested, and negative pressure verified, semlannually. A utiilty power outage may be used.to in'itiate such tests. 4.7 Reactor Fuel (1) Upon receipt from the fuel vendor, all fuel elements . shall.be visually inspected and the accompanying quality control documents checked for compilance with specif f- I cations. 1 (2) Each new fuel element will be inspected for damage and flow obstructions prior to insertion .into the core. 4.8 Scaled Sources The antimony-beryllium sealed source shall be leak tested in accordance with the procedures described . in the application for license amn idment dated March 21, 1963, except that the frequency of Icak testing will be in accordance with 10 CFR Part 34.25(b). The strontium-90 scaled source shall be tested for leakage and/or contamination semiannually.
'4.9' Pool Water (1) The pH and specific resistance of the pool water shall be determined each week.
Revision Date: 12/12/78
34. (2) An analysis of the poo1~ water. for rad'ioactive ' material shall be-done at monthly intervals. This analysis is to include 124 Sb as an indicator of Sb-Be neutron source. t r.teg r i ty. (3) Activity of.the pool water will'be. measured weekly..
-4.10 Core Spray The core spray in.the reactor operating position shall be tested for operability semiannually.
4.11 Flux' Distribution in order to verify that power gradients = amongst feel elements do not cause peaking factors to exceed those used in the bases of Sect. 2.1, the radial neutron flux distribution will be determined whenever a significant core configuration change is made. [ 50 DESIGN FEATURES-Those design features relevant to operation safety and to.Ilmits that have been previously specified are described below.. These. features shall not be changed without appropriate review. 5.1 Reactor Fuel Fuel elements shall be of the general MTR/0RR type with thin plates containing uranium fuel enriched to about 93% 235 0 and clad with alum 1num. The fuel matrix may be fabricated by alloy-Ing high purity aluminum-uranium or by the powder metallurgy method where the starting Ingredients (uranium-aluminum) are in ; fine powder form. Fuel matrix may also be fabricated from ] uranium oxide-aluminum (0 0 -AI) using the powder metallurgy l 38 process. Elements shall conform to these nominal specifications: Overall Size: 3 In. x 3 in. x 34 in, i Clad Thickness: 0.015 in. Revision Date: 12/12/78
l . . 35. Plate Thickness: 0.050 in. l l Water Channel Width: 0.12 in. No.'of Plates: standard element - 16 fueled plates (min.) window element - 16 fueled plates (min.) control element - 9 fueled plates (min.) partial element - 9 fueled plates (min.) Plate
Attachment:
swaged or pinned Fuel Content (Total): 200 g 235 U nominal. This value may be increased or descreased provided speci-fications in 3.6 above are saticfied. Fuel Burnup: The fuel burnup shall not exccJ .94 x 10 21 fission /cm3 except for U 0 -Al which shall 21 38 not exceed.1.5 x 10 fission /cm 3, 52 Control and Safety systems Design features of the components of this system (3.2.2, 3.2 3) that are important to safety are given below. 5 2.1 Power Level (Normal Channels) For this function three independent measuring channels are provided, two of which are required to be operable as a minimum. Each channel i covers reliably the range from about 25% to 150% (of 5 MW). Each channel comprises an uncompensated boron-coated Ion chamber feed-Ing an ampilfler that controls electronic switches in the DC cur-rent that flows through each control rod electromagnet. Each chan-nel controls and scrams all control rods. Each channel is fall-safe. The " fast" scram (N 5 ms) from each channel also produces, and is backed up i y, a " slow" scram (N 20 ms) through interruption of the AC supply to the rod electromagnet DC power supply. Each channel indicates power level on a panel meter allowing channel checks to be donc dur;ng reacto. operation. Each chamber can be changed in position, ovec a ilmited range, so as to allow the channel reading to be standardized against reactor thermal power. I Revision Date: 12/12/7.8 I
a . 36. 5 2.2 Power Level (Intermediate) Channel For this function a single channel is provided, covering reliably
~3 the' range 10 % to 300% (of 5 MW) with a logarithmic output Indi-cation on both a panel meter and a chart recorder. To cover the range under all core condi tions a gamma-compensated boron-lon cham-ber is used to supply a logarithmic amplifier. The chamber can be changed in position, over a limited range, so as to allow the chan-nel-reading to be standardized against reactor thermal power. Rate of change of power information is also derived, in the form of a per-lod, that can produce a fast scram (and backup slow scram) in the same way as in Sect. 5.2.1. From this channel are also derived control and inhibit actions, v12. bypassing a count rate channel functions, bypassing of flow and flapper scrams, reversal and in-hibit of control rods. To negate the effect of overcompensation in the ion chamber, which can occur under certain conditions even in an initially undercompensated chamber, provision is made to supply an adjustable small current to the channel amplifier (up
~IO to 1.5 x 10 A) so as to facilitate startup.
523 Count Rate Channel A fission chamber is used to supply pulses to a linear amplifier and logarithmic count rate circuitry. Pulse height discrimination selects pulse amplitudes that correspond to fission events and rejects those from alpha particles. Count rate on a logarithmic scale is displayed on a panel meter and a chart recorder. In addition, count rate period information is derived and similarly displayed. The channel covers a range of 1-10 cps, correspond-ing roughly to 0.25 mw - 2.5 W, but the upper limit can be in-creased many decades by repositioning the chamber. The motor-l operated chamber drive is operated from the control room, the u.Ive position being indicated on a meter. To prevent control-l rod withdrawal when the neutron count rate information may not l l be reliably indicated, inhibits are provided on count rate and period, and when the fission chamber is being repositioned. All Revision Date: 12/12/78
~
l
. + l
<- .. 37 ,
th'ese inhibits are bypassed at a power. of > 50 W. A' scaler is also provided for: obtaining accurate, values at -low count rates i f needed (c' g. , approach to cri tical wi th new. fuel 'or new cor'e
- configruation).
5 2.4- FNeutron Source For obtaining the. reliable. neutron information.necessary-for startupLfrom a cold shut-down' condition, an-antimony-beryllium
- neutron source is: provided for Insertion into the core as needed.
This source, ncminally. 50 curies of 124 Sb, is renewed by neutron activation in the core, its presence in the core is not essen- , tial except af ter extended shut-downs. Integrity of the' source: d is checked by periodic sampling of pool water (4.9). [ 53 Rod Control System 5.3.1 Control Reds Up to five control rods are provided for the control of core re- , activity. These rods may be either. boron-carbide or silver-in- t di um-cadmium (see 4.3 1) .- Individual integril worths vary from , about 1-4% a K, depending on' position and core configuration. The ' rods are coupled to drive shafts through ' electromagnets 'that allo; release of the . rods' within 50 ms af ter receiving a scram signal. Position indicators on the control console show the ex-tent of withdrawal for each rod and a digital readout can be switched to any or.e rod. To limit the rate of reactivity increase ; l upon startup, the rod drive speeds are limited to 5 in./ min. and no < more than two rods can be withdrawn simultaneously. Switches on 3 the guide tubes attached to the control fuel elements are arranged l to produce a scram if any guide tube is lifted This guards against lifting of ;he attached fuel element. l l 5.3 2 Regulating Rod One' regulating rod is provided to aid in fine control and main-j .tenance of constant reactor power for long periods. The rod is-b < non-fueled, is limi ted to a total worth of 0.6% a K for safety [ Revision.Date: 12/12/78 .
-i
J
* *~
38. reasons .(3.1.4), and can be either manually or servo-controlled. The drive speed is 24'irt/ min. . Coarse and fine position readouts 4 are provided. i 15 . 4 - Cooling Gystem 5.4.1 Primary Cooling System Core cooling is effected by gravity flow of demineralized water from the reactor pool-to an underground holdup tank that provides 16 an approximate 10 minute delay to allow N activity to decay. ; I The water is then pumped back to. the pool through the primary side of a heat exchanger where heat is transferred to a secondary cool-Ing system. The holdup tank is vented to the building exhaust duct. The driving force for the coolant is the fixed head between the. pool overflow gutter and .the water . level in the holdup tank,.
'the latter being fixed by the total . volume of water in the system.
Flow is adjusted to a desired amount with a valve in the core exit 7 - line. Core cooling is not.immediately affected by pump failure as flow will continue tl11 the water levels equalize; neither will , the pool be drained. To prevent leakage of water through the pool
~
walls, a continuous steel shell is located within the concrete pour of the pools. All embedments penetrating the pools.are welded to
- this shell. To eliminate corrosion of' inaccessible piping, the I
' embedded portion of the reactor primary cooling piping under the pools is stainless steel. To change over automatically to natural convection cooling at low flow rates, a weighted flapper valve , seals the core exit plenum. This-valve, held closed by the core pressure drop, opens by gravity when the flow drops to approxi-mately 700'gpm. Leakage at the flapper valve seat, or in the plenum,-is monitored by a plenum leak detector that senses plenum pressure increase and alarms in the control. room. Primary flow is measured by taking the pressure drop across an orifice plate in the core exit 11ne, Indication being both in the pump room and on a recorder in the control room. Temperature sensors. in the pool (above. the core) and in the core exit ilne allow the i Revision Date: -12/12/78 f o-- , --- , e. e e , , # -v --+ - , ,n. -
39 core 6 T'to be measured. These are resistance thermometers having-alternative recorder and digital readout in the control room. Float switches are provided to monitor pool level. Normal pool level (at the overflow gutters) provides 24 ft. of water over the top of the
. fuel.
5 4.2 secondary cooling system Reactor power transferred through the heat exchanger is dissipated to the atmosphere via a cooling tower. To minimize corrosion, the exchanger has stainless steel shell and tubes. To prevent water from entering the secondary system should a tube leak occur, the static. pressure in the secondary is made higher than that of the , primary through the relative elevations of the two systems. 5 4.3 core spray For backup in the event a hot core is exposed, two spray nozzles are located at the two alternative operating positions of the re-actor. They are controlled from a manually-operated valve located outside the reactor building. 55 . containment system 551 Physical Features The containnent structure consists of the reactor building, with 3 a free air volume of about 7700m . This building houses the re-actor, the primary cooling system including holdup tank, and the heat exchanger. Personnel access is via double ai rlock doors or sliding doors with inflatable seals. A 12-feet deep water-filled canal penetrates the building with containment provided by a 25-inch deep water-seal weir. A water-tight gate wi th inflatable seals can be used to shut the canal off from the reactor pools when needed. Ventilation access to the building is through pneumatically operated damper valves that can be used to seal the building. ' These dampers are fall-shut upon reduction of ai r pressure. Revision Date: 12/12/78
a . 4 . 40. l 1 I
.2 Emergency Sequence While negative pressure within the containment building is not a ]
regul' site for reactor operation, it is required in the event of a release for the controlled-release containnent of airborne radio-active material . The emergency sequence is initiated either auto- . matica11y by the excursion monitor (see 3 3) or manually by the console operator. The sequence is that all air supply ducts and the pool sweep dampers are closed immediately, followed later by the exhaust duct damper as soon as negative pressure (-l in. w.g.) is attained in the building (but not more than 7 sec. later). Closure of the pool sweep damper prevents activity released above the core f rom reaching the exhaust duct before it closes. Also closed immediately are the isolation valves in the vent line and the air purge to the holdup tank. This prevents activity in this tank from reaching the exhaust duct. Upon closure of the exhaust duct damper, the emergency exhaust fan starts and maintains a nominal negative pressure in the containment building. This fan exhausts building air at a low rate (5200 cfm) through absolute and charcoal filters before connecting into the normal exhaust duct. The latter discharges to the atmosphere through a stack at a high elevation. The entire evacuation sequence is fair-safe upon loss of utility electric power. It will operate with either utility or emergency generator power. 553 Exhaust Duct Monitor (" Stack Monitor") Air in the exhaust duct is continuously sampled for particulate, lodine, and gaseous activities each being read by separate de-tectors. The relative proportions of each type of activity can thus be determined. The results are indicated on chart records, with repeaters in the control room. Detection or indication of a release is not dependent on all three detectors being opera-tional, for any release will have associated wi th it all three types of activity or will af fect each detector to some extent. Alarms when set points are exceeded are given at the monitor and repeated in the control room. 11evision Date: 12/12/78
41. I 5.6 Fuel Storage l l 5.6.1 New Fuel Unirradiated new fuel elements are stored in a vault-type room security area equipped with intrusion alarms in accordance with I the Security Plan. Elements are stored upright in metal racks in which the separation between elements is a minimum of 2 inches. With such an arrangement, subcriticality is assured (Ref. 5). 5.6.2 Irradiated Fuel irradiated fuel is stored upright under water in the storage pool within the reactor building in critically-safe racks. Each rack accommodates 16 elements in wells with center-to-center spacing of 6 inches. Ref. 5 states that an infinite number of elements so stored would be subcritical. Each well has a bottom hole to allow circulation of water for cooling. 6.0 ADMINISTRATIVE CONTROLS
6.1 Organization
6.1.1 Structure
The organization for the management and operation of the reactor facility shall be as a minimum the structure shown in Fig. 2. - Job titles shown are for Illustration and may vary. Four levels of authority are provided, as follows: Level 1: Individual responsible for the facility license and site administration. Level 2: Individual responsible for the reactor facility opera-tion and management. Level 3: Individual responsible for daily reactor operations. Level 4: Reactor operating staff. The Nuclear Safeguards Committee shall report to Level 1. Radia-tion safety personnel shall report to Level 2 or hie . Revision Date: 12/12/78
42.- f
'6.1.2- 'Responsibi1Ity: ,
i Responsibillty for the safe operation' of the reactor fact lity- shall be-within the chain of comma'nd shown in" Figure 2. Management. levels' i in addition to having responsibility for the policies and. operation f of the. reactor facility shall be responsible for safeguarding the
~
public. snd faci 11ty personnel. from undue radiation exposures and , for' adhering to all requirements- of the operating IIcense and tech-
~
-j nical specifications. In all instances responsibilities of one ,
level may be assumed by designated alternates or by higher levels, [ r conditional upon appropriate' qualifications.' g i I l' l i Y Y i a l .: Revision Datei 12/12/78 =* 7
43 i' FIGURE 2 BASIC ORGANIZATIONAL STRUCTURE l 1 l l Level 1 Site Nuclea r, Safeguards ! Administrator Committee Health, Safety and' Environmental Affairs Manager Health Physicist i Facility Level 2 Manager eact r Level 3 Supervisor Assistant Reactor Supervisor Level 4 Chief Operator , Lead Reactor Opera': ors i Reactor Operator.;
~
i l . Revision Date: 12/12/78
-- , - - ~ .-.. ~
"o -
' g, .
6.' t . 3 Staffing: ta. '.The minimum staffing when the. reactor is not. secured shall be: (1)- A IIcensed. Reactor Operator in the' control room.' (2) A- second' person' present at the site able to carry lout instruction' of' operators and summon help In.~ case the'.
-operator becomes Incapacttated. .
'(3) A licensed Senior Reactor Operator shall be' readily.
~
available on. call.
- b. Events requiring the presence' of! a Senior Operator:
(1) All fuel-element or control-rod alterations within the unsecured reactor system. (2) . Recovery from unplanned or unscheduled shutdowns unless ;
' they are' of a . type excluded by the Level 2 authort ty. . l Such exclusions, posted in the control room or in the. .
appropriate procedures, may include but are not Ilmited f to the following: (a) A commerc.. s le' trical power outage or fluctua - tion exclus. si failures of the internal power supplies of the reactor control and safety systems. (b) Inadvertent manipulation of equipment in a manner that does not adversely affect the safety of the ~ t reactor. (c) A verified malfunction of an instrument, circuit, or component, including false signals. . (d) An Intentional shut-down, not related to reactor safety, made by the reactor operator. s L Revision Date: 12/12/78 > ,. ~ . . . _ . ~ . . _ . _ . _ . , , - _ . ...,,,... .._ ._,- . . . . , . _ _ _ _
45.
. 6.1.4 Selection' and Training of Personnel: ..
The selection, training, and requalification of personnel shall meet.or exceed the requirements of'ANS-15.4/N 380 and Appendix A of CFR Part 55 and be in.accordance with the requ'alification plan : approved by the. Commission. A comprehensive written examination , will be given to each Ilcensed operator biennially (interval not- ; to exceed 27 months).
. 6.1.5 Revi ew and_ Audi t:
The' independent review and audit of reactor facility operations shall be performed by the Nuclear Safeguards Committee. ; 6.1.5.1 Composi tion and- Quali fications: The Nuclear Safeguards Committee shall be composed of a minimum of three members. The members shall collectively provide a broad
~
spectrum of ' expertise in . the appropriate reactor ~ technology. Mem- , bers and alternates shall be appointed by and report to the Level 1-authority. They may' include Individuals from within and/or out-side the operating organization. Quallfled-and approved alter- . r nates may serve in 'the absence of . regular members. t , 6.1.5 2 Charter and Rules: t The committee shall function under the following operating rules: i
- a. Meetings shall be held not less than once per calendar year, >
and more frequently as circumstances warrant consistent with
- effective monitoring of facility activities.
1 I
- b. A quorum shall consist of not less than one half the member- ,
ship, where the operating staff does not constitute a ma-
- Jori ty.
- c. Sub groups may be appointed to review specific items.
- d. Minutes shall be kept, and shall be disseminated to members and to the Level 1 authority within one month after the meeting.
Revision Date: 12/12/78 . y<,- c. g , -+e * , . --'ry vw er - '
- r avv r- --
t
- v=-<3"--*~ e n ++ ww - -+ * ~wer * - * + - *
(,- 46.
- e. The Committee shall appoint one or more quali fied individuals to perform the Audit Function.
6.1[53 Review Function: { The following items shalllbe reviewed by the review group or a ! subgroup thereof: *
- a. Determinations that proposed changes in equipment, systems, !
tests, experiments, or: procedures' did not involve an unre - ;
-viewed safety' question. I
- b. All new procedures and major ' revisions thereto having safety signi ficance, proposed' changes .In reactor facili ty equipment, ;
or systems having safety significance.
- c. Tests and esperiments in accordance with section 6.3
- d. Proposed changes in technical specifications, license, or charter.
- e. Violations of technical specifications, license, or charter. !
Violations of internal procedures or instructions 'having safety significance.
- f. Operating abnormalities having safety significance, and audit reports.
- g. Reportable occurrences listed in section 6.5.3 6.1.5.4 Audit Function:
The audit function shall include selective (but comprehensive) examination of operating records, logs, and other documents. L Where necessary, discussions with responsible personnel shall l- take place. in no case will the individual innediately respon- , l sible for the operations area audit in that area. The follow-ing items shall be audited: l
- a. The conformance of facility operations to the technical [
specifications and applicable license or charter condi- , tions, at least once per calendar year (interval not to exceed 18 months). Revision Date: 12/12/78 w - y- -v ,,n. w, mw,-
'l . o . . 47
- b. The retraining and requalification for the operating staff, at least once every other calendar year. (interval not to exceed 30 nonths). i l
- c. The results of actions taken to correct deGciencies occur- I ring in reactor facility equipment, systems, structures, or methods of-operations that affect reactor safety, at least once per calendar year (interval not to exceed 18 nonths).
- d. The reactor facility Emergency Plan, Security Plan and imple-menting procedures at least once every other calendar year (Interval not to exceed 30 months).
Deficiencies uncovered that af fect reactor safety shall immediately be reported to the Level 2 authority. A written report of the findings of the audit shall be submitted to the Level 1 authority and the Nuclear Safeguards Commi ttee members within four months after the audit has been completed. e
6.2 Procedures
There shall be written procedures for, and prior to, initiating any of the activities listed in this section. The procedures shall be reviewed by the Nuclear Safeguards Commi ttee and approved by Level 2 or designated alternates, and such reviews and approvals shall be documented. The procedures shall not preclude independent Judgment and action, should circumstances warrant. Several of the following activities may be included in a single manual or set of procedures or divided among various manuals or procedures,
- a. Startup, operation, and shutdown of the reactor.
- b. Fuel loading, unloading, and movement within the reactor.
- c. Routine maintenance of major components of systems that could have an effect on reactor safety.
- d. Surveillance tests and calibrations required by the technical specifications or those that may have on effect on reactor safety.
Revision Date: 12/12/78
o , l 1
*' 48.
. ./
- e. Personnelyradiation protection ~, consistent with'appilcable l regu.l at i ons .
- f. Administrative controls for operations- and maintenance' and t
~for the conduct of-Irradiations and experiments that could- [
r
. affect reactor safety or core reactivity..
- 9'.
. ImplementationLof the Emergency Plan and the Security Plan.
Substantive changes to. the above. procedures shall 'be made only after documented review by the. Nuclear. Safeguards Committee and
~
approval by Level 2 or. designated alternates. Minor.modifica-tions to the original' procedures which do not change their ort-ginal intent may be made by the Level 3 authority (Reactor l Supervisor) and must be. approved by Level 2 or designated al-ternates within 14 days. . Temporary changes to the procedures-that do not affect reactor safety may'be made by a Senior Re- o actor Operator and are valid for a period of one month. 6.3 Experiment' Review and Approval:
- a. . All new experiments ~or. classes of' experiments' that could affect reactivity or result in release of radioactive materials shall' be reviewed by the Nuclear Safeguards Com-mittee. This review shall assure that compliance with the requirements of the license, technical specifications, and applicable regulations has been satisfied, and shall be documented.
- b. Prior to review, an experiment plan or proposal shall be i
prepared describing the experiment including any safety consi de rat ions.
- c. Review. comments of the Nuclear Safeguards Committee setting j
forth any conditions and/or limitations- shall be documented l- in Commi ttee minutes and submitted to Level 2. t Revision Date: 12/12/78
4 Ld. . All new experiments lor classes of experiments shall be approved - in writing by Level 2 or designated alternates prior. to their-Initiation. ,
- c. Substantive changes. to approved experiments shall be made ;
~
only after review by the Nuclear Safeguards Committee and ! written approval by Level 2 or designated alternates. Hinor ci changes that do not significantly alter the experiment may be approved by the Level 3 authority (Reactor Supervisor). ;
- f. Approved experiments shall be carried out In accordance with established approved procedures. .
6.4 Required Actions: 6.4.1 Action to be Taken in Case of Safety Limit Violation:
- a. The reactor shall be shutdown, and reactor operations shall not be resumed until authorized by the. Commission.
- b. The safe'ty limit violation shall promptly be reported to j the Level 1 authority or designated alternates.
- c. The safety limit violation shall be reported to the Com- ;
mission in accordance with section 6.5.3 -
- d. A safety limit violation report shall be prepared. -The report shall describe the following:
(1) Appilcable' circumstances leading to the violation.
~
(2) Effect of the violation upon reactor facility com- 4 l _ ponet ts, systems, or structures. i (3) Corrective action to be taken to prevent recurrence. The report shall be reviewed by the Nuclear Safeguards Committee. A follow-up report describing extant activities shall be sub-mitted to tlie Commission when authorization is sought to resume operation of the reactor. - Revision Date: 12/12/78 t t
)
9 - - - , . , ,v.,- * ..wr- y e-w-, e*w ,e ---s-! *sy -- , - , s ,er-=.- +%,-*.*_a--.e = _ ~m . -
P 50. 1
'6.4.2- ' Action : to be taken' in the event of an' occurrence' as defined in .
section 6.5 3, a-1, 3: Corrective action shall be taken to' return conditions to a. normai; otherwise, the reactor shall be shut down and
. )
reactor operation shall not be resumed unless authorized by the Level 2 authority.or designated-alternates.
- b. All such occurrences shall be promptly reported to the Level 2 authority or designated alternates. '
c .- All such occurrences where appIlcable shall be reported to , the. Commission in accordance with section 6.5 3
- d. All such' occurrences including' action taken to prevent or reduce the probability of'a recurrence shall be reviewed by the Nuclear Safeguards Committee. !
6.5 Reports
In addition. to the requi rements of applicable regulations, , reports shall be made to the Commission as follows: 6.5,1 Startep Reports: Three months af ter completion of requisite startup and power- i escalation testing of the-reactor, or nine months after criti-cality, a written report shall be submitted to the Commission. The report shall include a summary of the following:
- a. Description of measured values of operating conditions
< or characteristics obtained and comparison of these ,
values with design predictions or specifications.
- b. Descriptions of major corrective actions taken to obtain satisfactory operation.
- c. Re-evaluation of safety analyses where measured values in- ,
dicate substantial variance from those values used in the Safety Analysis Report. t Revision Date: 12/12/78 l l l
*
- y y v , w -,,r,, ,.- ,-, + .,-,- .-~.-=*-.*-,w, - . - - . ,
i-51. 6.5.2 Operating Reports: Routine annual reports covering the activities of the reactor facility during the previous calendar year shall be submitted to the_1icensing authort ty within 3 months following the end of each prescribed year. Each annual operating report shall include the following inforenation:
- a. A narrative summary of reactor operating experience _ including the energy produced by the reactor.
- b. The unscheduled shutdowns including, where applicable, cor-rective action taken to preclude recurrence, but excluding those of the types listed in Sect. 6.1.3. b(2) above. '
- c. Tabulation of major preventive and corrective maintenance operations having safety significance.
- d. Tabulation of major changes in the reactor facility pro-cedures, and new tests and/or experiments significantly different from those performed previously and which are not described in the Safety Analysis Report, including conclusions that no unreviewed safety questions were -
involved.
- e. A summary of the nature and amount of radioactive effluents from the reactor facility released or discharged to the en-virons beyond the effective control of the licensee as de-termined at or before the point of such release or discharge.
The summary shall include where practicable an estimate of Individual radionuclides present in the effluent if the es-timated average release after dilution or diffusion is greater than 25% of the concentration allowed or recommended. I
- f. A summary of exposures received by facility personnel and visitors where such exposures are greater than 25% of that allowed or recommended.
l Revision Date: 12/12/78 L
'~ 1 a .c 52.
l 6.5.3 special Reports: (Reportable Occurrences) 1
- a. There shall be a report not later than the following working day by telephone and confirmed by telegraph or similar con-veyance to the Commission to be' followed by a written report within 14 days of any of the following:
(1) Release of radioactivity f rom the reactor above allowed limits, as provided by section 3.8.1 of this specifica-tion. (2) Violation of Safety Limits (3) Any of the following: (a) Operation with actua! safety-system settings less conservative than the limi ting safety-system set- ' tings specified in the Technical Specifications. (b) Operation in violation of Limiting Conditions for Operation established in the Technical Specifica-tions. (c) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during tests or pertods of reactor shutdowns. l l (Note: Where components or systems are provided in addition to those required by the Technical Speci-fications, the failure of the extra components or systems is not considered reportable provided that the minimum number of components or systems speci-l fled or required perform their intended reactor safety l function.) (d) An unanticipated or uncontrolled change in reactivity greater than or equal to 1% AK/K. (e) Abnormal and significant degradation in reactor fuel, and/or cladding, coolant boundary, or containment i Revision Date: 12/12/78 ls
+. - , ~ n .. . .
( *. s. . .
, 't' .
'?
# m; .53, ;\
>: >l
[ bound'a'ry1(excl u'd i ng mi nor, leak's) . whe re . app l i cabl e which could resultnin exceeding prescribed radia -
' tlon-exposure limi.ts of personnel and/or envi ron- ~J ment.
-(f) L An observed inadequacy in 'the Implementation of'
~
adm'inistrative or procedural controls such that : the' inadequacy.-causes or could have caused,an un- , safe condition with regard to reactor operations.
- b. .A written report within 30. days to the Commission of': ,
(1) ' Permanent changes in the facility organization structure.
-(2) Significant changes in the transient or.ac'cident analysis-as described-In the. Safety Analysis. Report.
e ,
,t
6.6 Records
u Records' offthe'following activities shall be maintained and re--
; tained for) the periods specified below. The . records may be in :
I the form of logs, data sheets, or other' suitable forms'. The re-quired information may be contained in single, or multiple records, or a combination - thereof. : Recorder'ctuarts are not considered to- , be operating records,'but. may be used as. such where appropriate. - i 6.6.1 Records to be Retained for a' Period of 'at Least Five Years or 5 for the Life'of the Component-Involved whichever is smaller.
- a. Normal reactor facility operations (including scheduled and L unscheduled shutdowns). Note: Supporting documents such as F
l . checklists, log sheets, etc. shall be maintained for a period ' L E of:at least two. years.
-?
- b. Principal maintenance operations,
- c. Reportable occurrences, j
- d. Surveillance' activities requi red by the Technical Speci fications.
s l ? ,
-t Revision Date: 12/12/78. (
4A:
-.~I._Yi '.
i s
. . _ ._. _ m _ .. _ _ . _ _ . _ .____ _ _ _ _ _
, , - 'l b ,
p is . (. y,i z ..;1 3 -54! I a 3
~
y,, 'i
- y ' . , , . . . . .
~'
e.:fReactorifacili tyl radiation and' contamination surveys 'where
~
-.. -; required by' applicable regulations; ,
.j.
).'Experimentsperformedwiththereactor.
- 3. ig.. Specla'lf NuclearL M aterials- (SNM) Inventories, receipts, and i F. shipments. .
T h. 'Approvedichanges in . operating' procedures. , l.: R'ecords.of'meetlng and audit reports'of'the Nuclear Safe-
~
t ards Committee-i J . . Sealed. Source . leak'. test resul ts. I
~
6.6.2' Records to be ' Retained for at least'One Requalification' Cycle' or ! for. the: Length. of Employment of the- Individual whichever is' Smaller: b
- a. Retraining and' requalificationt of licensed operations per- i
- sonnel. -However,. records'of the most recent complete cycle
~
shall'be maintained at-all times the Individual is employed. :i
' 6. 6. '3 Records to be Retained forLthe Lifetime of the. Reactor Facility: [
(Note: . Annual ' reports may be used where applicable as records ' i in this section.)-
- a. Gaseous and liquid radioactive effluents released to the t l
environs.
- b. Off-site envi ronmental-monitoring surveys requi red by the Technical Specifications. 6
- c. Radiation exposure .for all personnel monitored. ;
- d. Updated drawings of the reactor facility. [
I 1 i. ( r l l' I 4 f
'P
; E
'" - Revision Date: 12/12/78 ;
yp g= g e M' 4 f a' wt - - + rc *fg g g AT f " 15 -'rY"E - * * **N' ** ww **
t !, e .. : g . < 55 4 70
REFERENCES:
T ~ l..
.- Supplement Ho. 2 to Final Hazard Sunnary Report (Aprli 1977).
i
' 2.- Final Hazards Summary Report (May 1960). ,
i 3 Reactor Power Excursion Tests in the SPERT. IV. Facility, .lD0-17000 (Aug. 1964). ,
- 4. Supplementary information to Final Hazards Summary Report j (28 Apr. : 1961) . [
i
=
L 5 Critical Experiments with SPERT-D Fuel Elements, ORNL-TM-1207 (July 14,1965), by E. B. Johnson and P.'K. Reedy, Jr.
- 6. Supplement No. 3 to Final Hazard Summary ' Report (Dec. 1977).
~;
7 Supplement No. 4 to Final Hazard Summary Report (Apr. 1978)'. h
*e i
)
.t l
k b
+
i f
, 4
' Revision D' ate: 12/12/78. l
?
% . . . - ~ ,..... , . _ , , ~ , - , . . .}}