ML20234F560
Text
_-- - _-_.
p, ., yQgym mm%wfuf;mmyym&&.wsy:rpY i vx-m -w .. lSt r~wwn.::~7._;:=.
.q , ,.
h- i.d.h . MM- .
' *N. i S.. 1Y ,
I W W&
.. e, y.w m
- n. g e.m.-.m..M.+^.:hidNkf.."fMD_
k.s y
. y.a:Msm.
a- r w
. u i~
~
\.
- .. .xy.
, - ,, , .f.,_ne . .. y .
l:. - - , ,
1 n .
t
.s . . .,.
.;.g..
.:y ..e. .- ,
r ..
t .9.,.t ce? .t .,- .
,, ; ' , .y .
. :.ug;.. :n ..
e ..t.
.s vg:
- 6 - /s. ... f. ?
e -w-. '
. r w..: .
- 7. c4- 0 ,.,c..4,.'n.0.,,9. g.p ev, .+- w + < ~ HAZARDS ANALYSIS . -
gw.. v-. ,
a, y x; ,a .-aso .w &..:t : *:u. _.
.~
T, , _ .-E m .1. ?. s by the - -
f.2 3 .., . , . . , , , +., py ,, . ,
. g;...
m ; n .n., DITIBI N & LICENSING AND REGULATI M in the matter of PACIFIC GAS AND ELECTRIC CGTANY ,
BOIEGA BAY ATNIC PARK i
.i
, :s UNIT NUMBER 1 !
. 1 y . , 1..w.,g.
-i-t E
- _.,i m,. - .;,.. < v CGtSTRUCTION PERMIT j
4 .
i = te . . ' . ,l - . , . . ..a, , , . > . . ,. . . s , . p,g .
.s......i., s.
. [p I
% -205 .
. e '
,, .. r., ,
".g,.'..t.,,..- . '
- . , . .= _". / r. ' %:e 3
, .Ejn N. .. l'g ' ~ h* * *as .. wg. ,. /.' ...,,, p. ., 'g.7,
. . . , .p,yg .* ',*.V
' ,, " 4 s 4
.r .
8
.q., p s
.a.-...
- Ae c.. .u o
..L ..
r
-j- -
h.
l I
\
d g
a
.e e
I
..p1 a.Y e
- 1 4 .,a..,e*,
- +.'4 ' .
- y mJ .. . j. . . . ..% .. . -
+;
i 3
-...a . . ,
i e r ., . y. " w. y,. p. -.: .
m,%. /.a i
1
- A,
., .,. e . . .r.
. . e.,,,...+.
. . . ..u .
,...y . -
s.
w, A . .. 2...
<g
. *,- c..
t' ., ,4
" is*"
t l .
- . . . - rj .
,.. o,;-, <
4., A g.; .
g _ rs= :y . esse. .....,*'
4 i * ).i '..
..5-",
.. W. .
-y' 8 .. sg + r n., a.. g,'.4c. f .
y-e, . y yt.
6
'e'g,',
~ - * * ~ "' ' ' " '
4 . . * $
./
-e .
- g .,*..),,. e i s
1., $ . ., v
. ".( ,! 'W< . ' .
. ".J.. .*.r, .
-[ d8 3 ~ . ,l
,\
i; ' 5
.#.. wi ~ , ~. *
,,.9,<
J,s*' " *,'
4,
. ,4 *
..'s. '7.._.
~ + .-
- *3**
+-
- 3
-.. *}.' 9, L. '
.J . .
..E.., .
1 .
.. .. .,.,f,}...,
.*g.
, .a .
t_
.y 9,. .
S.p..
< .., O..;; '
.s 5. , , . ,
q
. . e .- - * .
i - t. . 1
.e i ..-
c-
..- , ,.. .o, . ,
8709230170 B51217 . . - .,
.- . .j FOIA e 4il *PDRFIREST 085-665 PDR
- + @w ~ -,, - -
e
<.- .-- 4 M
_o.7
. .[,'er,4\ ,, : .. w..e%'N % @ ftfy g 4.I,P %
I 4 *
+
- Pi k
'**9* 'b
. .,#, . m . a - -. - , ~i-w .. ~- *s m m"
- a w'I~~~4.m"'.'O< +,4
- I \
m e
,,Y 9 .
- r Me .'4 . d', A .Ye ' g
- ~.
- _h u ~1@ % %.&,';:X;_ndat*cbW W? "n:.& nim 'T!?m " W"I* _ *ht*'T':n*W O W"WM M C"" i E r- .s F A </ . ._ ~,
g I
r;g m- . ... 11 ,,
.p. , * * - '
- , 4 , f
- .e s.
- a ' ,p )
l
'if )
- 4 TABLE OF CONTENTS '
. :p
. c.:
< [ ,,
@grr -n I. Introduction ',
r
}
T,ig
%f II. whn ound . ,,
Description & S&
' ffb '
rl tr. .
III. Safei,, imdpi:
(?&
A. Site & Environmental Factors' F;,vf'f t:,+
I. Ple 4 L ocal/m .
'? '
~ .,-
__=...n_.u--- \
.g%$ - _ ___ ~
% m' , B, c,L. .
ww.u#1-/
. a c~.>.5M*st..s
- 7. on nt w
.q< .)i C. Reactor Nuclear Systems v n ,
t N
-' .f 1. Primary Systems '
f I
a
'[
- 2. Core Design ! -
i
't
!I/[.'
y;;, 3. Beactor Controls
- 4. Control & Safety Instrumentation
? 96 -
';g .D Bnergency & Safety Systems
~
., t
.i E Radiation Monitoring -
a K
] Waste Treatment, Storage & Discharge 4
IV.
R&D Program j V. Safety Analyses
-1 we VI. Mn4== Credible Accident Drhuation -
.mu ..
P,os VII. Technical Qualifications of Applicant
.~. s'ay y
y,5 VIII. Report of Advisory Committee on Reactor Safeguards
)
.DC..
g II. ,
Summary
- f d1 I, Conclusions
'nlM q;?'sK . r ' '
,S9s *
.a.
- rs' r s 87, t%
3:. .
.- n
..).
. e e
,l.
pa'. ,i - -- =
~
= - - = = -
f
- {.
> i ,,
4 %yd%u:m.-wee wnw.~>mn&+ .- r+i- rm "&w wmr>%hn & m~+e:tw ' -
~MW
.fM .fj J$ q . ' g" ?. ~ y' . ,
y .,,-g
/. . .. ~ n :/. .
,a - .
iii i , .
,1 , -
,i : , , '
4 /'. .
t + + i.,i c j- .
~ ,s). * . ~ 'J 1 ,
j L '_APPENMCES
- /f. r s
. WC..
.h ,4
.w' m
.m .m .
l 9
. 6-%. r.. r . 1.,
mrn.
i .. , . .
1 v,s .. n , . .
...a.,-
..o.....
2 v.c . .t.. w. - - ..
.,.,1 .
. ,, .. . . .t.....
w, ., .
.;.....a ,,
k:Q,*
v;
- w's ,
- .; y 7 J;.; 4 . Beport of ACRS, dated April 18, 1963 9
i.
i
- . e. L.e&c fea pe,a,/- a , / . /nhiae .
1-e, .O >
- l. r ,
t.
4 l^
)
r ,
,,a" g
- .{.-k J..,
,) " -
,\
t f.~__ .'f'. . . .
idth, . ,,
...f. . r.,
, 'i C . ./
- inc
. '/
.i
, , -, e d
.-,*fc ,,
4
., f
+ <., ' ~ i s
(,
I.
y
( j .. ,
i
, X '1; '. 4 *
.M p,'..'~,, .. .
hLW^. ..
kVf 4a F+
/
u;v r.e.
=.
- o .-->;m ! (
. ' .rew 9
p,4 K.,1 3. i .
,( ,
. .u' .' - ,
Y!? t' m
VM*A
>v+ n
&L}r
,Wj -
nl - ' '
'$y&% . s 4:++ '
- p/fA l/. *?.N. .
h'j -
/
o .
9 9
e
- t. =
4 j,
'y, 1 g , 4 a I',['_________
'g
- Rm , [ 4rk .-
.h * '+ *
, a a . = k
-&_,, f n A5 m
---= - ----
- fQyv.-:6?&GRh&&Q.m.qhwr:s+&:hs&nk. int'MwmMvw vMW'"OONN!'%N~t*'50*
.y.
.r . :# n< , .
.I; r . hh g. g -
eja ~, '
4
, w ,
_L *-
. :> sc.
a 3
, i,; ; D, . - Introduction .
, s The Pacific Gas & Electric Company (PG&.E) has proposed to M
Q$!..
construct and operate a nuclear power plant on Bodega Head in-M:NI p Sonoma County, California., PG&E vill design and supervise con-5 M., ~ '
. struction of the unit, and the General Electric Company (GE) vill l
~
v furnish the nucitar steam supply system and the turbine generator.
he proposed plant, designated by PG&E as Bodega Bay Atomic
.c,
- & Park Unit Number 1, will produce nuclear energy at the rate'of-e w ,
.+R 1,008 megawatts (Mw). . he gross electrical generating capacity
. s.l
'~i vill be approximately 325 Mw.
he Bodega Plant is similar in many respects to boiling vater I q*.
jQ power reactors now in operation. Steam vill be generated in a
.e "d
,d[b(i ,
.,g _ - direct cycle, forced circulation boiling water reactor. Its c . e Q.u detailed design will be based on operating experience from the ~
Vallecitos Boiling Watel ..eactor and the Dresden, Consumers, and
- .y <
'1 Humboldt Bay reactors.- Those features of the plant which require research or developmental effort in order to provide engineering >
q-
.j information necessary for their detailed design or evaluation
. m, 4 .
(
j .
v111 be discussed in Section IV of this analysis. -
g- .. - - . . . . . . .
Except for a discussion of seismological and geological factors this
.. . 2. . . -. . _. :... :..
4
'.$(#:a rep wt contains a discussion of first, the significantLfeatures of the site '
[4,# and environment which have a bearing on safety and secondly, the si6nificant 1 5)
{-[ j features of' design which affect the probability of or consequences of j
,e 7 -,
].tt y accidents or occurrences of safety significance' to the general public.
up i;
,$1 j_ he radiological impact of normal routine riperation of the plant, including g;g '
+ - . . . . . . - - .
]
w .
i N ol 4 the dischahge of radioactive materials are considered. 1 i
_...__._..._..___...__.._..._.... - . _ . . .._ ~ .-
L* dl
't j f r. > n.; a %,. .Jy'Q. ~a;$, W' *s u h-Dp; Mre ?. % ha.W Y" :: * .*ka r W ~; e r n n ..A ? - : hi!W bu E f .
.N.
+, .p
. . y;,-- -t-:- :u r 9-& ~ + wm W!c *~'rP " U n""*^@*M?"M m*" ~' "* * * *"* ' "* "'*'"" ^
, wt. t *,
, . YC;1 , ;*. ,
a ,
. .u l , . .
o
,5 -
.. , fi- n .
II. Background
$ d,; Ch December 28, 1962, IJ&E submitted an application to the r.n ,
s m -
);^h ,,.
AEC for a construction permit and operating license pursuant to
_J ,g- %g
, ' f,a;W Title 10, Chapter 1, Code of Federal Regulations, Part 50 (10 CFR 50).
) ct r .
- @y.m - - -
The application,.which includes a " Pre 14=4n= y Hazards Summary
.u .
.g 3
%j 14 Report", dated December 28,1962, 'a6EJAme' n dinents 1; 2;'isd' 3'to b
- w. the application dated March 4, April 5, and dwe /3 ,1963,
,,c-respectively, has been reviewed by the staff of the Division of
'fl >
n, s -
g %. ,
., . . . . . . . - - . - . . - - - - - - - - - - - ~. r -- - - - -- -~~~ - -~~~ ~ ~ -~-
l} .
.['l
$. ?p',
u.i
' I "*$1 !? , . .
i L
4.'.ffj t}-
w . -
- s. . .
s
'I %,,
f-
,( '
F
~
l l
1 1
4 9
.I
- .2 .l .
. /W
.;j
.py * . . ....- ,
e .
Y 'f y ,
1 .
g
- pd
,
- v. S. . ,
. ip h 8
- si . .
- i [*.l- ,r 3
'(j'[f k im
$7' a f ".
p,ji e '
- ' fry; '
n.
!..*t,
?
y
- 1 ,
1g,ff' .
w
- > Q- -
g'. ' . .
ph. ,,..s' Mi ._. - ,,
. %- a, -- , - ,
~ ~ ~ ~ ' - - .a
-s e.. . 1,
~
. g2: vlUO / :3.W:"J"EM .4 C I. 5 ~If TQ'RfMQ T?~:i' T"4WiWWUf5*SUF D IF kk .
19 .
@ O
,,/ *
$ Licensing & Regulation. Technical consultants in specialized
)
,y areas also advised the regulatory staff. The application has
')
also been considered by the AEC's Advisory Committee on Reactor
. r.
)
Safeguards (ACRS), as required by the Atomic Energy Act and the
.(.] 2'egulations of the AEC. The reco=mendations of the ACRS, as un
.Q. expressed in its report of April 18,1963, (a copy of which is
. .R
' C. ) : .
attached hereto as Appendix C) were also considered in the regulatory staff's evaluation.
'l As is cust envily the case in reactor facilities prior to
! the commencement of construction, there are a number of features t
of_ plant design and operation which have not, as yet, been i
j definitely resolved. The Com=ission's regulations provide for the pc
- j issuance of a construction permit on a provisional basis in cases A
~"" ' "
such as this, in which some aspects of design have not been com-pleted. A provisional construction permit may be issued, eccording
-) to Section 50.35, 10 CFR on the basis of findings, among others, i . that (1) the applicant has described the proposed design of the i
i
- facility, including, but not limited to, the principal architectural i
.! and engineering criteria for the design, and has identified the J
- e. 'J - \
q ;.l major features or co=ponents on which further technical infor=ation
.g
{
is required; (2) the omitted technical infoz=ation vill be j
+ ~
- J supplied; (3) the applicant has proposed, and there vill be
. :.)
d[j conducted, a research and development program reasonably designed y 1
- y to resolve the safety questions, if any, with respect to those
- 3. .d
&l -
features or co=ponents which require research and development; and T.]
- ' s
}
i g, ., '
NM[f}S f?M3L%N MfddbW%W38 hYW= ST!3YEi EbYW2Y *h
. W H4My@; g;'~ggg.QMr.:mn"mmk+rWcW&@20W?b55.vddenTMhCzif&P%?'O@h"-
e, .ra-
)* }o
. ;l < _ .I;
,:q -
[ .
9 l
, that (4) on the basis of the foregoing, there is reasonable ,
assurance that (1) such safety questions vill be satisfactorily a.
,', resolved at or before the latest date stated in the application
.i
- tl, for completion of construction of the proposed facility and (ii)
.q~.
Q' -
+= king into consideration the site criteria contained in Part 100, l2?.F
- p. 'the proposed facility can be constructed and operated at the
,%;f 7;.;y, m proposed location without undue risk to the health and safety of .
, .i j
the public.
,1
- The proposed construction permit, if granted, vould authorize l construction only. The Comnission vould require timely reports h from PG&E vith respect to results of research and development and
.d -
pg final design of the more significant design features. The AEC y t;- .
.y staff would continue its evaluation of the safety of the plant in light of this information. An operating license will not be
,! issued until the final design has been completed and evaluated by t
the AEC staff and the ACRS. In addition, the definite plans for 2
operations vould be evaluated by these two groups.
. Pursuant to a Notice of Hearing published ,
3
- t 1
1 the issuance of a provisional construction per.it to PG&E vill be j
@,Yd mg .
considered at a public hearing to be held in the hearing room of I
. ,rr y.
the Board of Supervisors of Sonoma County, Santa Rosa, California,
..g,.
at.10:00 a,m., PDT, on w.,? " 1963 before l l
ff
- j{3 an Atomic Licensing and Safety Board appointed by the AEC. The 4
':1 9
g 4 issues to be considered at the hearing are:
^:%
M .
/ j
, . l h . s.::..g f.h .~ ; :;c q k ~ ., : ,jy & & V c y s = ; .'.-:- sV , ; . ,":' 'c, % ' :: helNJiu. .. 0 k
8.gjp.t_gj, jh.gpe'lb@L'OP_3Nr_"'ctonZf7s9 .dsf.d(di O'!Smes'""',"?- i* l.PS"'Y"P ~ _.
v.w. i .
(. mcc .
..c .
(
~ h- .
. ut . .
j 1. Whether the applicant has submitted sufficient x
i information to provide reasonable assurance that a
.. i
, Q facility of the general type proposed in the
,.e,0 application can be constructed and operated at the
- ,m:?.
.g -
- {,,;.;,
proposed location without undue risk to the health
.,n . .. . ,
4@d :.
and safety of the public;
%k{;11 2.
. j.. Whether there is reasonable assurance that the -
,.4 technical information omitted from and required to i
- )
i cc < ete the application v111'be supplied; I j 3 Whether the applicant is technically qualified to design and construct the proposed facility;
[ .
- 4. Whether the applicant is, financially qualified to d(, ,
design and construct the proposed facility; and u.fi
/ " '
.1 F. 'Whether the issuance of an authorization for the construction of the facility vill be inimical to the
, ; common defense and security or to the health and safety f of the public.
he proposed plant er constructed on Bodega Head would be sub,$ect to I
.l '
severe shock from earthquakes. It here is also a possibility that earth-
, 7' -
m 4] . quakes in the vicinity might cause faulting beneath the plant which would '
cause severe dama6e to the facility. Se possible effects of such n
11 -:V ;
'. deismic activity on the proposed plant are still under study by the
(
Gil Arc Commission's regulatory staff. Be staff has not yet determined whether a.
7;,j 5 .
. . or
,_ not a plantacah .
...--.~
be constructed and operated safely at this location.
1 1
$d,
- y l. . e, '). .
4, 1I k 9. li ia e g/ ' '8, wk' Y *4"* g , ,
- 48 L$ ' '8 W 19 #
t A
. .&-6m~f5z:w%i,pMp,gg.pf.x%g~;ng,%y: sap y vv~:.~s.1. r.e. gim. :. m-r.wp -tm...m_ew.~::ep'~n v? w?~
.= = -
,u..Q,
+. c, , . , .. ,
. m. , ,
yy.g ..p.g.. <g.
~,s., ,
.g .
W p:2.
w 1
i At this time further infomation which must be based on exploration of the q ,
....4
- J,.*. ..
site, among other things, must be obtained. Further consideration must hiso
., 4 ?r. ,
, 1:--,
.,.. be given to criteria which must be applied in the design of systems or
.:.~,gg ..
9 p components of the facility which are of.importance to safety, egaecially
.m.;
c.@ thos, systems,whichmustberelieduponinaneme$gency,suchasasevere
" s; . ' ' .
earthquake, to prevent personal injury. -
3 4
') ' This analysis and the conclusionscontained th rein are made witho\ut m
a .
regard to the special safety considerationsvhich must also be taken into l
account in view of the seismicity of the proposed site.
.. '.l 9 hop.** s . &
l, .
1'y' ..s t
. son .
V ,s .* ( c. . . . > y ,. ..
. . .s
'"h$Y ;' ct i ?.~.;n d-q <T ' . .d.
' 1*.' .x: %~.'^ -> - ~,. ' r ;J: ;;, -
q'. l
.*1 e
\
1 l l 2
.. {
1'# 1 -
6
. ' -)
s-n 4
a s
.J v . l .
. %. . .. . - .<,4-s - , < . . .
i 4s, .- .. ,.- .%. 4 l 8 .
'e .
- e s#
o i
,i,f_.-M'.
d[(? d
. .. ' s ; * ' *
- . + ._;4 .
. y , 4. .
k . f %'% . \ . 1 I h'
w;;'.; ' .,
~- -
- ^
.;, ( *
- j, *;- , =
,- = . ' , , -
- i.n, O
,. t e a
- f g
ie e
- + i' w v.
w ,Q,
.nk.
- h .. , : n ,. o ':e w.v!.U. .. . &c.sm, G-::.:;&t:y.,.;9.; 3m-gl@ 1%., :. -
.x :~..ps:s ' .
, , r ,.
. .w .p
r s.
. .p,,, y a; 7q.n.:gg__w qygkny.gt % wm~hme.w+.b,mt: gy~g_,g;~nn.:n n:nen.~m e cp- m:&~M&:y: l -- -- e-- - - - - - > - -
- g. . i
-w . g ,r . . ,
.. 4.;l
' 'e
/
- =-
y,9:,y,
.. s.
- ,,
- n .
Q
\;J 5-
- . The staff's evaluation of the proposed Bodega nuclear power plaat described in subsequent sections of this report and its position on'the issues at the forthcoming hearing are based on s
all the technical information submitted as part of the applicant's y
- g
. request for a construction pemit and the report from the ACRS.
..4
_.. s . . t .; .
,n? .
- t. .All of .this info 21 cation is available for inspection and reviev
.t ;."
at the Commission's Public Document Room in Washington, D. C.,
and at the Commission's San Francisco Operations Office, ,
! 2111 Bancroft Way, Berkeley,. California. This evaluation and proposed recommendation is subject to modification in the light
- of any further information which may become available, including
,s
/- .,the evidence introduced at f,he hearing. Uml.. th: Ced eku's 9 . .
. s ..
. .1 n.
c.m , .:
. / 2:agdetice;,
.+:y: . ~ ,, .
,7 ye..cn ch; . Intr::t re ,1: .:.ff e-. ;,ma _,, ayyem y} -
f .s;. s.. % w fue ; ;; :12 _ ut=a; e, u _m, 1
,l r, -tici;.te la & y m eel -. The decision.of the Commission 9
vill be based upon the entire record in the proceeding.
1 I
i o
- T .
.#, 4 ="- / m .
-- . , , ~.
f-9 ., gg , ,%"b.'
hp % b A -
% ., - ~ d vl? -
g 4 q 'le n
.n:,
r ,
a.n .. . .
.l ,
')
1
. . 1
'~ . .
l 1
F q
' F- #
h 9 ! 8 j *[ NO P p ,,p :
g r
'.s 5' % .~ ' - - 4 g' f k wy* - ~ ~
p_,.A d
_ _ 4
h, q,;p,hgy,gr e_-erwrmrq;ygtsrnnysngjfg;;geg,-g Egg,gsggg. g:gg, ,,}
>@ 4 ,. .: - X .c. .
- -Q . . . , . <
- .w
, .. m ~. . . ... .
. . 5 ,,
3 1 III. Description and Safety Analysis *
,4 The Bodega Reactor is a direct cycle, forced circulation boiling water reactor with internal steam separation. Nuclear
.qp. energy. released in the reactor at the rate of 1,008 megawatts will
&G ~ , M .
Jyp6 m h tn..ir: ' " the water coolant which is circulated through W
. .. y Q- '
- 22%.
the reactor. Steam generated in the reacto.r at 1,075 psia flows 4 to a' turbine gfigerator with a gross electrical generating capacity of about 325 megawatts. Reactor coolant which has been c.
- separated from the steam is recirculated through four loops each
.'l
.J containing a pump rated at 29,'000 gpm. After passing through the 7 ej turbine the steam is condensed, and the condensate after p;. .
g%_ -demineralization is returned to the reactor vessel. This water, which
%p s
~:4.yg ,
, ..m.-
e q,, a :>e.-will contain some radioactive materials, will be circulated within jO %.
' ' a closed ' system from which the only normal effluent will be a continuous discharge of noncondensible gases. This gaseous material ,
i will be monitored continuously and released from the reactor stack
(
if the contained radioactivity is below permissible limits. As j
{ in conventional. power plants, the condenser will be cooled by water
. p. drawn from a nearby source. In this case water to cool the .
y.. , .
?5 .. A .conde er wJ1l be take'n from Bofe a Bay a d discharge into the j '
.' W ;2%s< H - & M hu/me Z.Me w
/ Pacific Ocean. From time to time, regulated and measured quantities u- 4
% Lt
,g,
~....
-of radioactive liquids will be mixed with the condenser coolant water g -
.m , . e%,,,
- g,. , and discharged to the ocean.
q:p;,S An overall judgment concerning the safety of reactor operations "I t.gey v.
' +
+
mA .
.M or the acceptability of potential hazards must be based upon a
- m u v m !
number of individual safety considerations. Our judgements at this
, , 1
\
C- \
9 time are based upon an evaluation of the design details, design criteriew- l 1
a m5 34q *-
$9 %
Did ,,...e u. w' * .- , , , . w a r . J. y e .. . *
,'p - '$** rs fe $)(( ; w h $4'k ? ,2 'e ,y ,! J f* {g [ ;lsf.). *; Q' ' '
, fhs <
, *i6w "j. ;
kUMGM6NRTiF#1W'a%JiWWeveAW 4 I
1iSNN D GEGi2Dl5W$ MIIEEEE?C-ICI#-@UN U M CC'-
N~ I'M- C1 -
@Q l. >; . >- . .
- * -7 4
.,j. and design concepts described in the PG6E application, the known d facts concerning the proposed site and .its environment, and an analysis of the effects on public health and safety of normal
. <.:;. ~
_ y.
i .T ?. i operations and of potential accidents during these operations, R .'Y.
! $;% As noted previously, detailed design of[ost features of the ,
~
7 facility is not'yet completed. For those features, the present "l9 .
\
- 5 4 evaluation by the Commission staff is based upon the principles J of design rather than upon details of design themselves. In the case of features which are of particular importance to safety, the J' '
staff intends to require and expects to receive information on final design of these features before PG6E has expended any substantial ew amount of effort in the construction of those featurcs.
J .- . . . . .. . .
.. ._h . . . - - . .-
ind ~
Be gesert.1 objective of nuclear safet is to revent or limit the
. & ff. wa g
- t4. W - ^^ '= 'D M
? p , ^-1 that the followingW -
exposure of persons to ionizing radiation.gt _f Aws $ck general features,be' 4 provided to achieve this objective: ,
(1) he fitat and principal safeguard is found in thedesign featurexoftheplantwhichcontribute[totheconfinementof all raC.ioactive materials to their proper placeJ, C " :- i
~]
,' Confinement of fission products and the products of neutron .
]
activation is attained by the three fold containment afforded ,
l by the fuel elements, the primary coolant system, and the
( containment bu m 4ng. Be isolation of the plant site provides n., , ,;
f,.i..'.3, . a fourth measure of protection of the public against the radiation from
- .9 fin materials routinely or ac61dently released from the plant.
- Q .
1,,; he design of the fuel elements and provisions for reactor 4.,. .
control and wer have.the purpose, from a safety standpoint, of
.? .A j retaining fission products within the fuel elements, in which location lh
~
!" ' ._ __ __3 i _
~ ~ - 2-2 ' :"' n' :k ' u-~ $
4 L.
.o % - *. e- m Q.W~. Ass:sAs,M M wiai % M s + a, v ~, w,.. w'sW anf W h " 0?1~....,"" z
% ,*.r7 ~M.T"' '~ g""%' ~." . -'" '
e - -
v ff 's _ L 'l. 1 f. "
- y :n v., G , >. Ac cu * .F y
~
O .
i the-shielding of their radiation is most easily. effective.
+3 T
$.l The primary coolant system, consisting of various F
Q'v vessels and serves to retain any fission products 1h which might be released from the fuel elements and any
- G
] 'other_ radioactive materials formed in the course of operation.
",g&j*j
. o. .
.a ~
z.. , > . . .o , ,
n.
FM*l # '- '5 Condenser cooling water which is discharged continuously to
.Q;'y, g the ocean is completely separated from the primary coolant.
- 4 .;y .
i!T The reactor vessel and the recirculation pumps and
%w . . .
,J ((
piping are located within a structure which is capable of qm S safely containing the' radioactive contents of the plant
,, in the event of an unlikely major accident which might
- i M Q) cause a release of fission products from the fuel and the iTM -
primary system.
i.56Gi .- a ,os . ..
.4% . ,
- +w,g).Radioactivefluidsandsolidsvastearecontr'lledand o i'EIE these processed in a manner which confines ihm hazardous materials to systems which are designed to permit their safe preparation
'.'. for disposal. Paths through which radioactive materials t
w are or might be discharged are continuously monitored by
.e .
radiation detectors. Storage, processing, sampling, and w; ,.
gg yg .e -
- c. , , , - -
.a-monitoring are employed to assure that quantities of radio-active materials released to the environment will be within h.'.!3 limitations established by AEC Regulations.
'7&J to JW
-(3) shieldingandarearadiationmonitoringareused/ continually d, ,i 1 ,
, .c .
?g,g assure the safety of workers from radiologiesi hazards.
.n .y
' VMc 'h ' -
' ' Such monitors'also serve as an additional safeguard against M'
% the release of radioactive materials to the public.
-j Y' '
. I r
. {':
w . . .
, 'e -M.* 'j ,Q.]g . 'f, {.[ j,;"(y g = y =- ;- -, -- - - - -. a[. , , / [:= s)., ,
& L'*
, o
_ _ _ _ - - - - - - - - - , 4
.as. - !g:n::ey.* dan-rMnhi~ # d1MmhA5'LWk&"U?%%%":;m MB? "~' *T"T'5" ' T ' '~ ^ '"'X' Y'?'Y.
.M.., - . , , . .
s.; ~
- y;.,g p
~-
.,x : ei',
- ~ .
m ... .
t
~t-(4) Energency systems are designed to ' prevent or reduce the hazards to plant personnel and the general public should accidents occur.
.p 57,
..---..-~...-... & m a .e 4'] f G d [ l-- W Y ,
, ,5 .
, .In the following sections of this reportgh::: ;;; discussed the
. .e ., . , . . .. . . % .
- .?@
-M;;r
'more important safety considerations which have led to the staff's
- .o g(gy; conclusions with respect to the safety of operation of the proposed
. .A.s C elc plant.
,e- I
. j I.i Y
a __mn if
?
- (.$.#q Af j(pt 1 spr%
.At " '
. .- *,p Qgh) . . , . .- ;.t.;' .,., ,
=; t
-' +
is,y,, ,. , ...
w . :ym .
...t ?
.. 1.j! * *
, .g : .
y
- " .Y'
,:[, *) '
+ '
->p,.j y,' e . . . e -N.m ,',a 6 4, [ ,
-.A+ *.
M
$t .
.'.b i ,
4 kid a
%*; 4. e Idq,ih
? **
.. e.
, . ,..]
- ;. . g e47 t 4 -
, , a.
^ ' '
M.h;] s. +
., , j
, g , . ** - -+ *
- _,.c ^. c ' - ~
~
i (pg - - ,
}.
, g. *
,". i,',) .
~, , i f
P 4
.'u., g O
- r* ,.
,, m e- >
..e.. , . , . , . .~..-s: ., , ,. .
J.
4 '-
EeN ' [.] ,'I,, k ,,.;k8[ 'Ne***k=
.F"% ^. ' s . ,. J'Af -%. l j d1 > 79 1 l," ***'
s"'d.
g -e ,
- [' % ' f ; e i
.#o. , *a
- 3 +
- 8& . (#. '1
- ]
4,W%ks:nhn~- Yi%-h&%%h&nAk%fi"WWMM Y " W"W$~*'FNC'ST W
=
y _, . ,
t e mg
- t
, a , ,
Q e
g -8=
L ,
' "./)j A. ' Site and Environmental Factors s
,, The site proposed for the Bodega Bay Reactor .
l
' ' .- : !, is located ~on Bodega Head, a small peninsula along I g .. i
..s. .
$.4,, 1 ,
the Pacific Coast' in Sonoma County, California, approxi-
. . . ..e .- t . ..
- ~
i .
n u * : mately.,
50 miles northwest'of San Francisco. Bodega Head is bounded on the east by Bodega Harbor, by Bodega-
..f o
Bay.on the south and the Pacific Ocean on the west. A as pQd ' sand spit known as Doran Beach or Doran Park extends towards Bodega Head and forms a natural breakwater for
, c:c Bodega Harbor. In the orientation of Bodega Head and the '
?.M proposed plant site with respect to the main California yy -
.7y v/+:n f '
-coastline as shown in figure 1.
Mb
?h' g
, . Oi,;.h: . ,; The environmental factors which are considertd to be
".".,)
significant with respect to the safety for this site and which 9 have been examined in detail includet (1) the location
[
.nI with respect to the nearby population, (2) the meteorological !
' 5 lj factors and (3) the marine environmental factors. As noted
, previously, geological and seismological factors,are signi-Mll . ., ficant to safety and will be the subject of further considera-PYi
$3 *
""'" tion by the Regulatory Staff. ' These environmental factors
,.'N
@ g -
are important insofar as they might affect the normal
-, g , .' .;
, operation of the plant or cause damage to essential equipment 7/$ and with respect to the bearings, the population and other i qg -
- t
]fyf' environmental factors have on the course of accidents.
&N _ ,
$5d . -
a t= 4
.i
[g
. Q% . ~
. , e a.[,. .S}M[d 'U. . = I J.;yn .
L. , q, g , ,7 }.s.'.ifh*f.', ,2 , , . 2,
, . y {. . , ,.., ,, ; s , , ,j
l ~
YMSX'EiXE*4WW%MMR&WCWWW ' " >.Mxff ~P '"~: "y' 27 L 773:;.3mq; gy_y.c.
-,c,., , 7 :_3. . .
,d , . ,
i L ' y '. ,- f\ ? y ,n
- , . _1
- n
- u.
's ~.t ** .
~
y,as' sy ,
= .
.g
@g.
'9
'7 I
'/ I.=' Plant Location L
' X.2) ji.6 '
The proposed location for the reactor plant is on !
kil fj '; . , ,
a 225 acre tract of land at Campbell Cove near the Southern
.m.. ,
+
'f
. .-,,4 3 i. ,;['l ~ , ,, ,f%.end of Bodega Head. The property owned by the applicant
- 'l
. u .. , ,,
. h .. .
includes the entire southern end of Bodega Head.
y ...
?, r,4 y;-
a ,
The' adjacent property to the north is under acquisition 1a n; 1
. .!,$e . by the University of California for a research facility.
Mi The reactor would be located on the east side of t'w,N,.
4 ,
.t,:.,
~
Bodega Head near Campbell Cove and'across the entrance
- m.
. ,. ?k channel to Bodega Harbor from Doran Park. The nearest J.py '
edge of Doran Park, which is owned by Sonoma County and'
'h3 .ns s
4
- % 77 contains no residences, lies approximately 1,300 feet m
$,y$ .
- y .
- ,;. , ;;;. . .. n , ,
T '- " east of the proposed reactor.
.r The traffic through the entrance channel to Bodega
, ;. . 1 Harbor consists primarily of commex1al and sports
,6 jfl j fishing boats. Usage of both the Channel and Doran
> .. )
Park could be controlled under emergency conditions if 4 J
this should become necessary for protect: ion of the
,g , i
.:a.a
,m
., , . , . public. Accordingly, the exclusion distance for this 1
, @4@9 .
3
, g;
,i:6M site can be considered to be the distance to the '
jtR \
g northern site boundary, which is a minimum of approxi- .f mately 2,700 feet (0.5 miles). ;
r[h #.
a; MI$ - - The population data submitted by the applicant based g.t,j . . .e , , m - ,:*r w '
- 3. ;
(waft ,
on the 1960 cent,us shows no population larger than about '-
i
. .%Q 200 within 10 miles of the site, and none larger than M.. ,
- Q;,: ,
M -
about 3,000 within 20 miles of the site. The nearest !
E j 3'.. *
~ , ,
}; R ,
- j. "
..-t+.- .s - . . ,
. g y ,y ri,, .g ,.
i p.4g, ,*..a.A M ;,O ,. 4.. . Cp. . #,,wr : ._yh a.n,.9,,.- i ,_Qi;g,4, 1.g. r. ,' . ..; . 4 , , ,, g-
,y ,
/
1< . $5
- f Gy
- '*MX***:rgr$*i.%C_.m:MM%**7,'r?u'*"MtT.
h Lw ?* *$~7*. :?**.,T?TQT?TW
- .a%sm -. -u: ~- V"jairi
- 1 "'yfCAQ-- Qy?Q;&'C_.fk!".m i
i
- ~., ., ,,
j '
f; .,
c s .
.Q
-2.0- .
_Q cities of more than 10,000 are Santa Rosa (31,037) and Petaluma (14,035) which are 21 and 24 miles
- 2. . .
- , respectively from the site. The population density a -Q.hi ,
@@s
-ry - -
within 25 miles is as follows
, ~ . ," '
Persons / Square Mile Total Population d y;,. '
Distance
- 7
- <, .
of 1.and Area' In Ares w
I
.? y 0-1% 0 0
..- 1%-5 21 5'0 0
@...o.'[ -
5-10 16 1600 10-15 81 15,700
, ~.. 15-20 97 30,000
.s ,
20-25 180 66,700
$ l...u j, 4,j The above tabulation shows that the propo'sed
! ' '. ':]",y .
f?
J.,
reactor site is favorably located with respect to D
' " ~
population distribution and density. The location on
.gy
- l. .
l .y a peninsula provides natural barriers agsf nst the n
j ;g future development of housing within at least two miles
- w. . . ,
of the reactor plant. The tabulation shows that the
- 7.
- ..
.j population density is low essentially out to the o
a distance of Petaluma and Santa Rosa and these cities ij$ '
Q account for about 75 percent of the population in the
.p.c annular area that lies between 20 and' 25 miles from the 3 m-j7,q.; reactor site. -
.Ws1 f g. , Based on the population data the staff believes that FE
- g the icw population distance and the population center an gjti - .
distance for the Bodega site can be considered to be
?$)N{;y 24 miles, which is the distance to Santa Rose. l
@; =
G'31 dI.'.' -
- e n
- '
- g/m.
D * -
W.
g
- 'ID.l#J s
. A;, w\* M:. -.--..N. ' . N f '(-~.,-n i
14 ) Y. E) dp r' D,6U /J'- ** ' e,b > ' 4 i,
. L*A . ' $ d
. = '*f. -"' ""#' "
- 5 ? 5 - ' ,- - ^ i .
i, : ' m? .,..m _.w ~ m m u 4 . y _,1
[% A m m m y-lp 9%,sggyne*~kn y n my m.n:w sw :v.ky F-:gm w g4 &.,._p.y;.m
- - s-
.g!'j h ,
.. 4 9
~
E -
j..::o ='.- -
q, y: w o . ' ,, - . ,
A.N.q E gy
~ Nj , 11 II 1L
-2. Site Meteorolorv *
- ' . In general, climatology of the coastal area at y
- ;;
, Bodega Bay is typical of the central to northern 4
s J.
. .- .. F coastal area of California and is characterized 4 '
.F by a wet' season extending from about Nove$ber through
.mi.d
.n w
- +IS,.,
March and a dry season from about April through
.an gu h October.
- gj; The topography of the area inland from the site
' ' ,"5 is characterized by a series of hills and valleys,
.4 W
-with the hills rising to elevations varying from
>4.l
.% J 'approximately 400 feet to approximately 1000 feet.
.,.,.r, ~m qxp
- v. -
W1. ewvu.m6:u,. s.,.The roughness of this terrain would be expected jk '
to enhance the atmospheric turbulence and result.in
- v. .n -
hi rapid dilution of gaseous effluents that might be d.FI
- transported from the site to inland locations. On M,M
}iy the other hand under strong inversion conditions, the range of hills along the coast, which is approxi-m
- /+fM
. . :.a -
mately three miles from the proposed site, would -
g'my Ag ,
u . -
~,x.,.' tend to restrict the transport of airborne materials .
[ $i ' 1 * '
A&
4., - to the inland areas.
ye --
NJ m
, Detailed meteorological information for the I f[_ y.
MQ; -y -
-
- proposed reactor location is not presently available.
v
. s,s .gF ,
nM Ay . . .,
~
,<s~.-, .
However, observations of wind directions and r W). . 1 -
. velocities o* the coastal locations at Point Reyes, e, ~u
- S
(,g approximately 22 miles south of Bodega Head and 39$ .
dIw? '
J,'W 1kkhc .. ,
- 1. -
be .a:x . . ..
. . . . ,, . ..;.,......, . ~ . . . , , _
Ik' d a=Ndk.EII' 'Ig' * , '
'N- SlPY$1} Ig . k "1I) h [ '-h*
, J,5 + ,t %' b,*6.: '- "
- * ' I = "# .- ' h. * * # 4'* e'# > I '+ - -
) # er i*
b
.y+.,--~u w.m
.~
,w1mwxyw
. . - - Ay- my~ ~~7m ::T-~; . ',7.*'2 L~=3~:r?_mvmOR*PW T:"W m.m
,v r. , ,
Q, - s-
..i .
.:m .
J.
12
. .i
',j Jenner approximately 12 miles north of the Head,
.c
- t shows the wind direction towards the coastline f...
i"% approximately 60 percent of the time with the
.$5 I.D , , ... ..o,.lr~.. . prevailing wind direction from the northwest.
- ., , ,u
}gff .
We The information indicates that the remainder of C,$ r ~
kI,3 the time the wind blows either offshore or sb.
Nh! generally parallel to the coastline.
,%S
<W ug Based on the meteorological infomation presently available the applicant has proposed diffusion N' parameters for use in the Sutton equation for ;
sg -
yg , estimating the smount of atmospheric dilution under d[d re ; , ..f. 1 various meteorological conditions, as follows.
r m. .n ....y,. .. 3 .
. tr, N- ~
Meteorolonient Condition Strong Moderate Lapse Parameter Inversion Inversion n .5 .3 .22 Cy .2 .21 .6
- Cz .02 .07 .2 y ' tr -(miles per hr.) 5 5 10 is.
W. The staff believes the parameters provide 1
. --:.y.
- .m ,
1
-f adequate conservatism for estimation of the possible
- w g
exposures under accident conditions except with .
. ]j
~'d1 ,
., respect to the wind speeds proposed for the inversion ap;t . . :. . .
,gl conditions. In our opinion, a wind speed of one W .
]' meter per second, which corresponds to about 2.2 miles
- j
]-[Wa q
- 1.J. '1 , .
.j l
e.@d w
)
l ,
v.a
, :@i f -
. v.* ; .
...;.g; f, ; .
g.p g,;,,, ..o , ,, , _, ; p , .
3 J6lihMMIF13XIMWMa*s.WAnw a mmua.caRMhstezw.-wr.w"MT5hl .
s
[g p# q &
?-
y_ p-s__gt p!VlgsjpQ*ny;,]_i+ygjy y4[7.r%-$h_6gt:Q ih_;,;fypyyfs y&. n &.y Q g-Q.p:._
lii?). ,- l !. , :6h;. [ ,, [:e i I
- . c%: . .::wM n.-s; +.m u - n- ,
.v.
4
. -3 i
per hour, would'be more representative of a-
~:
, , coastal location during periods of' slow diffusion.
r; , {
j,i3;; '
iTable I. Appendix III of the applicant's Pre-ipv .:w.-
F *""- ..
{pj { E j f,] ;'liminary Hazards Summary Report, which shows a '
w ,
g * ;f:T.. ,
relatively high' frequency of the'1ower wind speeds ,
l 'fs in the zero to' three miles per hour range confirms .l
]c..
our judgement in this respect.
The meteorological factors discussed here have
, j
.j j .
been used in the analysis of the maximum credible j. L f 1/W f ,
accident evaluated by the regulatory staff and 7*/I
..' " described in SectiIm VI.
o:.. _>
@)3 ; ,
, . e 4 Mg, .' PGSE has constructed a meteorology station with '
- c
-a tower approxisr.ately 250 fest in height to collect meteorological data, such as the frequency of wind i
_- .1 speeds and directions of various atmospheric and
- d. stabilizing conditions which would be appropriate for i
t-L use at the proposed site. It is expected that sufficient
~
2: . data vill be available for estimating the capacity M , ; .c .
- h.(,, ' 7 7 ~of the atmosphere to safely dilute radioactive gases m ..
D,M - " that might be released from the site. These data should
. u.1 ]
,4 j-jQ- -
be available before the reactor plant would be ready
,m
. , . hi + ,
_..,to operate.
- '3
. k ..,y; .
h.A
- p. ~e . .
. I'*
s:eg .1
{.y.f-> ., .. 1
., 1 i3-p , .e . ~
> ; j
' b * , ;.- ,9 1 . ;;. .; .i * '
, , . ,. , . ; J ,, [;k
l- ,
- ' ? *y(y?N~/p+ YJ **g.y*++'49, %-rp irf.4.~~rj b+/~b epive ignitpg,;rq,'pg g<vp ,,dg. f p,y . - _
- f. r
(.. .. 1.. ., i
', ..? -
hh$$ , .s, M;s u '. - -M - mu ~ .
L
= -
_u
- - i c
- 3. The Marine Environment w
- i. -
Approximately 250,000 gpm of cooling water for y< a
. M!n, "
the condenser will be withdrawn from the kin ^ _"
n?.h ....: .
g;.u.. . ,,
-]y - 'yatCampbellCoveontheeastsideofBodegaHead ~~
M Iw.k -
and discharged to the ocean on the west side of the v.3;;
Head with an estimated maximum temperature rise of 18 F. The condenser cooling water does not pass through the reactor so, therefore, does not become' 1-radioactive from exposure to neutron irradiation.
l This cooling water, however, will contain liquid u .:d ;- " effluents that are periodically released after
- y. s M,. ' c. I
. . ' .;. c . - monitoring fro.n the rad waste fbcility. The con-1 centration of radioactivity in the condenser cooling i vater before discharge to the ocean would be controlled l I to meet the requirements of 10 CFR Part 20 of the
\
Comission's Regulations. j l l
,J At the request of the AEC, the U. S. Bureau of j d.' ..J 2 ,. . .
Comercial Fisheries of the Department of the Interior
.gi:
~u a . .- .. .'
Mj , has reviewed the effects of reactor operations on the l;
- rU
)
MTW; marine environment of Bodega Head. The Bureau's I
.-.m'-
,~ +y . .
report, which is attached as Appendix II, indicates i N-5 - ,s
~,.~' potentially significant effects from the discharge ab b
c; _
~
a .
v -
of this effluent to the ocean ast Q
$@ i a #
.rc
- .c-e;
. . :.4
,;l_ %
- l-;; *- s
</2
.kk . ., ..
- i ....~.v p&y....a{gfg;m,a, .&.. . apg.s. .
. t. . . . . .
.a,7., z . ;;. p;.
._a,,;[. .....;, , ,,. . ._ p ; x , y ; .-~;; 9 s . ..
,3
y~."'$7:lT?--tp.tY.??9:&'hgpi;;2W%n:d.,7g?@g?.""'h&T'."::W cm _ .
Mp7"*;fR$'3CMQQ*/
.. ,4.,, -'L <
. Aw 3.o r . .
/ . 5:
~15-t
- i. ;j (1) Possible concentration of radioactivity
. , y.. in seafood, a ^' t Qp) , ,
4 (2) Possible concentration of radioactivity ip;:gi . v .. .,...i'. .
yf t . 1h ' along b'esches used for public recreation.
. tr e; 1 .
,?,e The Bureau of Fisheries states that it is well i
i
.i t
established that certain levels of radioactive wastes
.,I can be discharged into the oceans without adverse
. effects on fish and wildlife, since circulation insures mixing of radioactivity with large volumes
_, of water, quickly diluting and dispersing the radio-
- m. .. -
e f.* ;-- ,
activity. Their report further states that the v% - . ., . %g
~ys permissible levels and rates of discharge of
~
radioactivity are difficult to determine in advance for any specific area, and recomends an extensive monitoring program to insure that concentrations of
. radioactivity in the marine life do not exceed pre-I
.1 et determined levels. ,
a, ,
.94
_ :.y . , .
The experience at Windscale, England, where
~
,,a .; radioactive wastes have been discharged to coastal ,
-:4A ,
- . .] waters for several years was cited by the Bureau of
+m .
7;$ Fisheries as an example that shows no adverse effects
.-u b.-
JI'( , on fish or shellfish from reactor operations. The e i,n. )
,7 English have determined on basis of a monitoring program
- p .
~
'k carried out over,several years that 3,000 curies a
v.
- ' d ..
no.. . - . ..
- s. < . . . ,. . .. , ,
c . _. [d I
~ ~ ' ' ' ' '-
F l
... +e.. . m.gv.-:y- r7 giy.- zyg.mys ,y z p,i + _~m. ~:q ~xy ~;s; ~,n~~ry-p'.:y n& cum Q?) ,
. t
.e J.dk '? ?"!! , *',' *
, 2
( W p s i
per month of strontium 90 could be safely released to the coastal waters at Windscale.
Q -
.PG6E has specified that the concentration of
. ,m : ,
.q
. - n: . . o r . . ... ,.; ,. , . . . . , , .. radioactivity in the condenser cooling water discharge
-x; .. - ,
"Juf; .- -O , ., y ,_i,
.N 'c -
will not exceed that permitted by Part 20 of the n,"
Comission's Regulations. Based on a discharge rate i
of 250,000 gallons per minute the regulatory staff j
] has estimated that approximately 4.1 curies of
)
i I strontiura 90 per month could be discharged to the i ,
I Pacific Ocean, if all of the radioactivity in the
- condenser cooling water consisted of this nuclide.
- Ejg
.., 2 . 3.2 . .
This value appears extremely conservative in h
comparison withgpermissible release value determined by the English for Windscale. ,
j The applicant has initiated extensive studies of I
oceanography and marine biology to evaluate the
-l marine biological aspects of the proposed reactor
.7, operation. These studies are as follows: '
.a,~ y ,.d An oceanographic survey will be carried out 9 -
-1. .i, to determine the circulation pattern of
.i.).).
% the ocean in the vicinity of the outf all, Stt, -
c't)A
%g and the capacity of the ocean to dilute 80 the condenser cooling water discharge.
@D 2. An ecological survey Aid .:iliN an
.' J inventory of the marine organisms in the
'""'t.1 .g
! eyi , W. t.. .cy, ,
.g: . r,9,, % fec, ;,tyer gy(2 .'$) ,' , ..
v.,,,3 :4 , 4,,,.3y;,7_,4.,
,, , l
.g ,. . ,
n ; 2 7 2*.; Q '; f .: " M t*M M & 9 t M Z n = % M & ;' w ' M & " ~* c; M '* M & _ T "'
- M i' W W W 'n s C !.TT; E~ i
. ' J' .
-17
~i
,, vicinity of the outfall will be carried
, out.
.:[
- 3. An environmental curvey of the marine Qt.
environment will be conducted to determine 4.
g .,
existing levels of radioact.ivity before i
- ' , the reactor becomes operational. This program would continue af ter the reactor
, is in operation to determine any i
tendencies for reconcentration by the marine environ =ent of radioactivity
.i
<h released to the ocean from operation of the
<v o. , s.,
. plant.
'The staff has considered tSe information submitted by the applicant and the coments of the U. S. Bureau l 4
of Commercial Fisheries and concludes that the liquid )
effluents produced by operation of the proposed reactor can be disposed'to the Pacific Ocean without l J
] exceeding Part 20 limits, and furthermore, in view '
{
of this it is extremely unlikely that adverse effects -
on the marine environment itself will be observed.
1 .
j
\
.Q .
.The staff further believes that the programs proposed !
1 I
'by the applicant for maintaining vigilance over the j 1
l
- v. . <
marine environ:nent will be adequate to discover any (
e.
anomalies that may occur so that these cot.1d be
.)
! corrected. -
a , ' e,, 6 . r.
- e..*s.',
.* 1 4 W' q*
, e. j Q Ilf . ' , / =" + , N tg" j . n.ly ,
x, .cy ?"::c2t: tm: ;sm:';am;'1M nm- =.m:r" W:yn&:@m em:vw a%hn Nymppb ey ,
~Y
- C. dondo *ame.d th lY &
j The containment system serves as a principal safeguard against the N
release of radioactive materials to the environment in the event of a major rupture in of the primary system,17 . ' ~ > - - ^ " '-
m er
. ..ee;
,.~ a~^re; ;y os . L- m , 6:- ~a11v.nrovided. a= #- thi: ;/ 0, tr ei er prev t
/
,0 .,
'P ,
gor et e release of
/
ssion roducts under/y A. ous acc ente .' c tions.
.gg;! !< ..c. .
. 1: . rrer various nergency systems opeya e indepe ntly t ent
- .qj?,j .
/ /
. lg- rele r,e [3 r;di Live materials be y d the co - a n es vi ay ylpm. Con-
> J
-1 4 i
tainment M Oth:r : ~ ~ e ' : ;.rc discussed in this analysis in the g ' "
context that th;, nr: r;;;i.m. i t; safeguard operating personnel and. the public in l'---"
+ the event that the best effort to design the plant to the highe }atdardsandtoconstruc}andmaintainandoperatetheplant
> ) .._
^_
- 4-- i & t
- _, nw~
O should fail.
r e proposed design kN w -
o g e _ _ ._ j as J_2_1..J in
"[ this sectionjixxxxmit '- ~ .;r 2::2 ir e analysis of the - 4m m credible
_u,y; s.. . , '
1 accident, which is disedssed in Section 6. ^^*
- n'ysi; ic th h;;i; f.,4
- a;-~~ :15;; important d arameters such as drywell and suppression chankerdesignpressures/ $-< - -
Q w f ,$),
4 -.em a
,9 w
- sgsl) 'N,
- _ .' d's
,,,....y.. . . - ,.
.h) w, i t sp4
$.s o
~. Pg
( , *
, : N:.* ~
4y ., -
.; e i
s .) *
- 2 - . -
4
,4 A; ,, s , . ... , , . * '
h _ ?! _' 0$?*. ' * ' - ~* * * - ~~* * '
.w me w smmmw;e,m.~Bismo:gyn~2nyt,~.&m+dernmyn&t9*riTW'M:W 'w: ? .'&:~M3 m _
"f,
,,' 5f f .
3 n= n v -
y . ..
.g -
9
, , , .. 4 .
n ;;
.The containment system proposed for this facility is one which utilizes the pressure suppression concept. Itsj'. design is similar in many respects to that used at Humboldt 49.. =
;gd.
,.g
- +
. Bay reactor facility.
Significant features of the Bodega
,x(:n .,,
Bay Plant containwnt design include the following: h/ '(1) Plans for the Bodega Bay Plant call for a dry well
.,' having a 60 ft. diameter spherical lower section and a 26 ft. diameter cylindrical upper section. The over-1 l ,
an height of the dry well is approximately 100 feet. J I (2) The reactor vessel and four reactor recirculation loops,
^
,.. each with a pump, vill be. located within the dry well.
~ ~
m .. - - (3) The' dry well. vin have an airlock entrance. Personnel entry is not pjanned during reactor operation, but is i l , contemplated with the reactor hot and pressurized, t (k)' The suppression chamber vill be in the form of a torus and vill bave a major diameter of 93 ft. and a cross-section dia=eter of 26 ft. It vill contain about k65,000 l i
- . 4 gallons of water and have an air space above the water Q ~,J a.
w . -. , s '
,s. -
of about 80,000 cubic feet.' L 1. C. .: s Both the dry ven and the suppression chamber vill be designed 3
, _ and constructed in accordance with the ASME boiler and
%..n Pressure Vessel Code, Section VIII. Piping restraints vin
~
wn
,79;,-
';, .. ...,4 . .-
be provided at contai e nt penetrations to assure that failure f
, of the pipe vill not cause containment rupture. A concrete i, #
.'a jit 4 *,'f;
. f
,e >.
-b * } *h .:(vT' s.
d, ,, [$ ., T [-I he; G- #
. 1 - = ,.,,,,,..q ,
- 4%%%%mr+wwhminvu wmdM"mamwm a mrM
Y % $ M iE2ii$bN 5'M! W M Y M k$$*???@ M y h 'Th Q :^s$ $ % g5QQkyyq'W);;dyqh :
.q _ ' , ,
. i 'a .r " . . - ..
- [9 -
1- . j- . refueling building vin contain the dry ven and suppression 1.j J chamber. Pressure and leak rate specifications for these
; containment system components are as follows:
- . 4 Laak Bate M:;
- Component Design Pressure (% of volume in 24 hours)
O$
%i.i _ m .. .
. .. ' Dry well' s 62 psig 0 5 (at design pressure) ei^G
.hf.$ Suppression chamber 35 psig 0 5 (at design pressure) y Refueling building. _
12'in. H 2O 100 (at 1/4 in. H 2O) A In order to proof test the Bodega Bay pressure suppression MMM design Pacific Gaa and Electric is conducting a test program at l .its Moss landing Power Plant. s in the Humboldt Bay project, the m 11mm ,x m m nw w.v- xixrw111m 1.xn w m x m xnaxu m
.) :. r
, _ 7;.q applicant has constructed a fun scale segment of the suppression
.y .
system. . In the test of Bodega Bay design a single 24-inch diameter vent' pipe from the dry ven to the suppression chamber was used. Since the fun size plant is to have 112 of these vent pipes, the testequipmentrepresentsa1/112segnentofthecontainment. 1 l Testa vere conducted with this mock-up to simulate various I 1 l i accident conditions. Aflowco=parableto1/n2thoftheflov I I
,, resulting from a cc=plete circumferential break of one of the 28-
- .?.
., u,{ J e, in, reactor c.oolant recirculation lines (with flow out both sides G,! of the bree'r.) was taken as the " maximum credible operating accident" 3
+g
,F (MCOA). Highest containment pressures observed in these tests were C,j -
52 psig in the dry well and 30 psig in the suppression chamber. mn ' M,mD -
' These pressures were observed when the mock-up dry ven was pre-
' i.g. . heated.to 2550 F and when the mock-up reactor vessel vater was
'4 .._
subcooled 350F. Tests at higher and icver dry well temperatures and at higher and lower reactor water subcooling yielded lover
. dry well and. suppression chamber pressures.
.v , .. . . ...s. . +,- . ,. .
%,, ,, . . , we g , .ft y N l .' \ .n - '
4, G7 g 4 *f~ '* l'o *,f 0% f f-
i$! M Q %?4' & Ib W 55 W.St Y 5 Y 5 iN N WA D $ $ $ $$ 5 $ 5l b S S ^ M YI rfa- .,
- . t
- J. [ - .. .
_g_
~
/*,
, In another test a break area 2.5 times that of the MCOA
' ~
, was simulated. In this test the peak dry well pressure observed j t
, ' .',r,);,. was 63 psig. Further Moss Landing tests are being conducted to M ,. '
determine whether baffles are needed in the suppression chamber. x.s@.l eff b [p G
.As another significant contai-nt design feature, Pacific y 2.. , 5. . :.~ e ,' .
Gas and Electric p oposef/that in a number of instances a single '
'sv.o.A"Q g
#1 97 isolation valve weGil e installed at the containment vall in P;$$
y
' '. : pipes or ducts penetrati the containnentp # M
~:m .
y L_ _ ___ ~, J .g es. - N however,A ach such line vill have two 1 olatie valves /w m+ of * *q s@bt hr = . - c.-f,&Lau 1., l=d p uuu V' ,' =:: :c2 ~ -- .
.x._
'I
- M
. ; ,3..
{L . 1- - = i: .11.-12 e- -ee
~
d',6)
.3
--= - ~ - -
a-- m )
- e. "
q.g isolation valves located at the dry well vall in each Rg'? y
- 9j main steam line are to close on a manual signal or autenatically on.the occurrence cf any of the following:
1 (1) Lov condenser vacuum (2), Main steam line leak (in the pipe tunnel,)
~
'n e (3) Low reactor water level.
j;g . - .- .. .-_. - . - - u_ , _ _.:. , _ - -- . _ _. . .._ . .;;r . . ;= c -- - ,-- . k.f,f .. 4 84a G @ W og
~,u Y.% PG&E is giving consideration' to providing/lika protsetion m:x the main
.C g;+ steam line isolation valvecFEainst forei 6n matter, which might interfere
>?
. with proper valve action.
.p/k sf ... - --- _ . . . '
um .
;* ,,.;,', d
{(
.' d (c.e .
.p -
S l r e
~
N. A .
*'Q<,.'.-p, * - D .a
;Q. . [3.; , -
4 '
.......s "" '
.s' a . .
*.vsc e ;,3
. ,, .yy.. . l; . ' H n. . , ,
" . _ ,: , ,c, s . y_., ~ W,f'
,,,,;,,, m-
~w.p: m*45.i &sM~2'r?h:C d'9t W O N iki? ? Y T ' W W M W W ? # . (~ -
,F . .g.g,. . , - . c ~ -
- 2m -
yng;( '* '
"j ..n. d2,$4--- '
.. . . ..;; . a, . .:c - -
- .. s...
5 . The Bodega Bay design is such that during refueling, the spent f' . 5 fuel storage pool vill connect directly to the shield water above
=
i .
"f/;>;
;g.-
- the reactor, thus permitting direct underwater transfer.of fuel
. M,. .
b17,3, d vithout the need for a transfer cask. In our opinion this. feature Ti,1
. . +,
. .. .....provides[asimpleandreliablewayforbothcontinuousshielding
, , , . ~~.s (u., . ~ and cooling of spent fuel during transfer and storage,
?%d '
.; re -
1,7;'f *
/
8
) 2n e
e s, , s e ee g'i j '
;y ?.ni ,
^
-i
-?.%p . . e
,:y,k.
yn - yg4 'm .;,.y . . c . .
. ; t'.
en ,1 - W. i . k e 4 6
' *h g 3'l
%wh ,
.$'$ 5? -
.w._ e4 +
# ."g.ef' *
-.67r y. .
L'i
- .W
' ($?s
^/'
r,,.
.'!j' ' -
- a' s k.~$h. .. '
(Q.1
. ,l e..-
i y aR., ,,(w.
- .m.
sn pas
- el h% 3
...o-* ] .
b *
~ "' ,."^
g, . j' ig ,' % M, , e . g i
". I9ii N[ .. E i$ 7.' .1$ _ _ I ~# ' ' '
-*E* ' ' * * - 5' .
.'.' P > N'
t
^
_ hd$ hNk [kbh (($ 2[hM SEEEM:SNEY fk (( (hN2[$$h j .c ',. ,
- M v , . y w I
- 20 - '
a . Nh j,
. The refueling' building in which the dry well and suppression chamber system are 'ocated is provided with a controlled release ventilation system which discharges to the plant stack. The h building and ventilation system design i$ such that the refueling
"$h
{gg -;
- building can be maintsined at a negative pressure. Discharge
.9.*
'.,j, from the building pcases through halogen and radioactive particulate
> q: .
F cleanup equipment prior to discharge to the atack. PG&E has in-
' ']
J e dicated that, in accordance with the reco=mendations of the ACRS, the system vill be designed to permit frequent testing of the
-t i
ability to filter particulate and to remove iodine at specified efficiencies.
-(The folloving 2 paragrcphs v!.11 be revised to ieflect the
.s .
.c
, - expected In.andment 3. ) -
~
-j Thestaffbelievesthatthegeneralcontainmentsche=eproposeE 2
by PG&E is adequate for the proposed reactor. It is our opinion, however, that sc=e */.rpon; ant criteria for the design of the con-tain=ent features have not at present been specifically prcposed. Such additicnal criteria, including those mentioned explicitly by
,, the p report to the Co==ission, are necessary to assure that w l
.. the containment as proposed can be reasonably expected to provm
~
jfh? ^* y.,y the high degree of integrity proposed at any time that it mir).c \ y , 'ee' h B* '- y. ce
, _. be called upon to contain the consequences of a maximum credible (
.. w 1 f
W.,d accident. .
/ j g4 -
.. ~
kb The regulatory staff, in reviewing the details of design,when they ere
' ~ ~
. i p%
developed, intends to assure that the folloving ' criteria on contairmant l yp testing, penetration design, and isolation valving are mer:, s . -- . - - - . . . . . . . . l
;4 ,.
.) ~
. i *% ..t , .#,y b,,
s' en g4T7 **i. 6., ,. ([ S Ip' . ' . . I
\ ~ '
y
.ie. .> . .... - . .. s ... .-
. " ha
'**F
- d.f g.>.y 1,. .
~.
' . T* :
']
Q 3 Jhy--07,1 %
; y .
- 21.- ;
q
.,. u
.d 1. ' he desi5n should permit Initial integral leak rate q
@, testingafthekg-vellandsuppressionchamberattheir ik g[jy - "
' ' ,
- respective design pressure after the installation of all
- 3 :,
y , penetratio$s (Jacluding piping conduits, electrical. con-
'N f duct 6rs,andgnketingclosures)andsubsequent periodic e
o .,
,] testing at .supprensi".:6 pool design pressure. In the I
g initial testing,)t.be leakage rate of the containment system should be detez=ined as a function of spo:a pressure up to - full design pressure. ft "'
~c ,
i k 3
- 2. S e design of penetrations.should take into account, in I i4 \;
' addition to.the pressure load, the loads or defor=ations imposed by thermal expansion, impact'of missiles, reactions of ruptured pipes, and disturbances incident to
, I j installation, mdntensnce or repiir. Penetrations should
.4
.i be shielded from missiles to the extent pre. ticable. All
. penetrations should be designed so as to allow frequent 1
j .; periodic leakage rate tests of the penetrations only "1:
"i
~(including points of attachment to the containment shell),
, at full design pressure.
- ;1 a .3 All pipes and conduits which e-nicate with interior of y
.y J the primary system or the contaient system,' and other Wl)
- f. ' -
piping (such as instrumentation and control piping) which w
,q ..
cannot be adequately prctected against accidental rupture, j ' should contain double isolation valves. All valves per-t q forming the function of isolation valves should be provided
- )
o . . u. e ,. 0. " ' 5 . e'.0 5fm em' l , Y*. , . , Q *, < , 'l
, , ~
- Q >fe.h) f f ;' , _ ,
l ,,f. ' 'k ', q , '~ 3, < .
%aj $ ' CUE Q $e~TfEZ$? M MTW N! W Y W N'O*Mb M W6WiWW M M EN$$1?1'h@
v #444 e ,, t
,, j; 4 .,. p'.. .; , . .
9 ;. ;g. . ,;p. -se. . g g .
".. .m,,
o r, , c4_ _ . n , .- 22.- r ' w,
.f' a a 1 sjs with protection against materials' f.n the system which MJ proper "
thp might prevent / closing and should be provided with reliable '"l l w
,w 1 W.0 .
, -- , , , automatic and manual actuatior,sfeatures. Isolation valvin g
. My*.. ,. . ...,1..- .. r - -a- >
cg p
- . +... c ."
should be designed so as to permit' periodic leakage rate J n.z+;. l .
,2g.{ , thats'. Appropriate closing times for isolation valves should
- m. .
', +
be determined on the basis of analyses of system ruptures 1 ,
~L which would release fission or activation products outside the dry well while the valves are nor,<fvily closed.
'l ,
,. [
The Staff believes that FGes ahould submit for
,u. .
- 1.
. Cotuisaim review the results of further developmental
+^
. .' .i ;j a t.;<t
. o.n. . tests of the suppression pool concept and final' design w.f ;;v . , . , 4y. p: ny .3
,t +.. . ' t,
.3...- , , plans for the containment as soon as they can be made
.. 1 a
, ..] available. !
6 l a 1 S, _ .
. o , ,. _. .__ ,_ __ _ . -_ m .- - . - , - , .
- se w zne erroc a- ;
; . . . . ..j _ _ . u, . _ ,. _ ___ _. -_ _ . . _ , . _ ..m . . ,s
_ = e f, ( r. 1 ,3 ,i . \ --
~.
,b y' '
1
.'/. . .'. . . . -
..r. -
,/1 ,
..f.) -
W
..xyj . ,. , - o . .. .
l
, ;n, -
- n. ,. .< e
\
e . I
, , e I v -
$$ '. .,9 -
g, > c. , A. i . eii p
- j.
. > e l' $
l l
.-i -e4 4, * ,y a A *
.J i4,*..
9 ,, . . , ,
. 5h k-
, -. e *...s,
. M'd;9 d' Q -g.h W.rL.3h' {.91-h,'y.QK C ii4 941 6 ;. *~ '& , W' .:
h pnire.;.
.-~%
f
}W& ?lA WhhW?MbWM!d:M T.% % %fd O 'M Y.M C B G IG O M G M @ i.74 C K & g ,l5 Q "Q l& , ;; % ... ,,
^ ,..
22a &
) :;}
j
- n. ea.x-wss- .
As indicated previously the design of a reactordiould be expected to 3; inherently provide a large rnsure of assurance in the safety of operation.
.;'y
,a.,
That is,the features of the plant which are necessitated by considerations h w'
,.,,,.j'of convenience and economics do, incidentally, provide safety and reliability, Y en With. sdditional consideration' for safety values which are involved,
, .' . i the features of the reactor core, reactor controls, and primary systems 4 3 4
]' >
V.dch are necessary for normal operation can be deslened to provide the
.] principal protection against the development of radiological hazards h resulting from reactor accidents, e
To safeguard a reactor core against damage that vould release fuel i'.G.ml$ ,
$y xg3 and fission ....n products
.a, from the confinement profided by fuel elements, one
- +
;h hj must adequately provide fcr (1) control of the chain reaction, (2) removal
~d' of the heat ge.nerated in the fission process and the radioactive decay of fission products, and (3) protection of fuel elements a6ainst mechanical damage. Control of the nuclear chain reactor is necessary to limit the
~J rate at which energy is released within the fuel element in the form of )
,,. heat. Excessive heating or inadequate heat removal vould cause overheating x:7]
gg-
. c.v of the elements to the extent.that cladding is weakened or subjected to
-92 g ax excessive internal pressure. Mechanical forces, such as reactor pressure y:A
'n and hydraulic forces, which impose loads on the fuel element structure, y
g: 2ihg
. must'also be taken into account in fueltelement desi6n.
jh . - ce w r '. m : .v . CilCenerally speaking the
?L'i" . .
<J- mechanical hesign of the reactor fuci elements and the
- ,s..s r;i,
'r l
design of controls and the heat removal system should be. such that, in normal operations and under many conceivable
]
i e,3., 9 4 .= ,#' .. =9 ,
- 8- '
f rl' %.' .'$3
*U ' 44**. " k. k [;h h i c h *I N* UU*I, N* /
*" * ~ *
' b Y 5- "* Y l ' ' '
f
l 1 3 "W83M M55?#bI.N f3 N$ M D II# YMEMs7dlEiAY4iDN N DESN36IY5%dC. t< A
. ?.4L. l.. ".-
'e , ,
... c. .. -.
.. m 23- -
accident situations, radioactive fission products would
.{ 4 l
be confined to the ' fuel elements themselves. If this m i
.i. objective is met then the primary coolant system serves as a 9
'**'g
, .4.. secondary containment system. -
'M.
. t
- 1. Primary System 4,;;
.j.g , .- The functional integrity of the primary system-__is necessary to the a
. $j . integrity of the fuel elements themselves. Its integrity is also neces'sary )
i&
"' '- 1
' 'y.. in normal operations to properly confine the relatively minor quantity of radioactive fission products which are expected to leak frcm fuel elements. 1 l '1 As peern*4 yY (34E>"the reactor core will be located in a reactor 1
] pressure vessel designed, built and tested in -
' UN,: accordance with Section VIII of the Boiler and at e '
l .2$ $
'. l l Pressure Vessel Code of the American Society of 1
-A 4 B.,;u. -
.i. .
n, c . ., y ~
.,j
, . . Mechanical. Engineers. The 50 ft, high by 15 ft.
,s- .
1 diameter vessel will be constructed of carbon steel
! approximately 6 inches thick. The interior of the I a, ,
' :j . vessel will be clad with stainless steel applied by .
welds overlay methods. The design pressure will be
.,i 1235 psig at 575*F.
.v.-
s 72; Steam generated in the reactor is passel through
- p. . . ,
. -..a,
..9- axial flow steam separators and driers located and then db !
',u i inside the reactor vessel through two 20-inch steam lines
~;
l l l N).1 i.gj ' ' - 1
, p.N < . . ,
'M s-b__. . . _ .._. _ .. . . ; .._ _ .. .___ .. _- _ . . , . _ . . . _ . _ _ . . . _ _ . . ., i l '.bi?
;,y
- ~
I , i, Le -
, N.
l
^3'
'.' ; I l
,w l
"I"'d[ y Mkf s.* f.?fl:.
b'e Q[?b k* 4 ; % )* ^ ^ ,,,\ 3J i. M*. L ,' , q. ,. -
, ..-;..' f t ,
' &.qM ~h,ty$r&
J Ei M J I NjdSchtkS $ $b5 & g;& ff,f$hf-t ju$h.i$$k.$Q - M,. . *3 '
, . . - .s . .
s:- ~ . n,tyr .n - . . sf .? ~ *
- 24 -
e-
-i to the turbine, which is located in a separate 1
J structure. Water, after separation far from steam in the reactor vessel, passes from the lW g . reactor vessel through four 28-inch pipes. Four , gt ., . . . . . -
. gk.
am
= ..
,~ y r w ..
-recirculation pumps, one in each loop, provide the ,
a; r. , s#ps ' < driving force for circulating water through the _
??$1 -
= reactor core. Feed water from the condenser is
-,q'M injected into the reactor vessel by a pump driven by the main turbine.
, Pressure vessels and piping located within the
'l' dry well will be designed, tested and constructed in 4x Q
g 'accordance with applicable requirements of the ASME
- n. , ,
. . . m . , ..
@g p.
- f. .- e egjd,a.
. B, oiler and Pressure Vessel Code.
Piping outside .. <. m@ u- ".
, the dry well will conform to the requirements of 3 the American Standards Association Code for Pressure L Piping. Twelve safety valves, arranged to discharge
. . .y into the suppression chamber, are provided to a
, protect the reactor and primary system frem over-
~
.d
,g ; _.
- pressure. -
[l)$ he general, concepts and criteria of primary system design proposed are, 4
- y,$
', {
in important aspects; similar to those used in several other nuclear power *1 1 !
- ht plants. Se staff believes that a system desi ned
- - 6 according to these plans I
..'E,) can be expected to fulfill its safety functions.
; . . ,. - . , .:. .;u % . ;.a,
. - - - , - - - . - - ~-
t - . --
', q < C ' '[ . h. D. - * -
, ,4 . ' -
I 'W.-j k 4 *
.,tiq j
. . l e
.h ..,a. s.- s ..*m Q gr ,6 *s. * %.. .. ,..,.4 .v ,,,.s .* s ,
*l[a Q . , ,
(([ k k, TQP4.4g,*,Y-i7((gf f,f,f?g ',.i> s hC[%*d'$'$il, A".!,4;'4 jC *
. . . . - . . /
'(, p' M,j c [.'p,E ' f .1' y'L'i/ 6 dh ' h.[N r g p. .v ,.C [4' * ]
i .
.grhy!Cs'422Omejn.TCp*!*Mylv MCMtlM'"KC!TTTl:2;24?$1$yiMQ':DR&'X '
1- ; J. c c. .
; y . , . . . . - . -
1 . o:
. ,: r . . . .
. . 25-i ..
.j 2. Core Design
'l
- Since the nuclear characteristics of a reactor are dependent upon
( \ .. j thefuelconfiguration[smongotherthing the structural stability of I' Y core components is important to reactor safety. forces of mechanical 7.',lg . or thermal origin which exert stresses upon civilians. components of the i;;.3i . . Spf.] core must be taken into consideration in thet: design. D.xia Neutron absorbing q+,, .- control blades which are'necessary for the adjustment of reactivity and
.j l ,
control of the nettron chain reaction must be provided with channels in
.] .
which they may move freely. The power output of a reactor, which is of prime importance to the l operator depends upon many factors involving nuclear, hydraulic, and thermo-i
..[ -a dynamic phenomena. The objective of thesafety evaluation is to determine
. ;lj.
_ ; the effects on core and fuel element integrity'of operation under the
. w;g -
s ; 3- . >r .. o. .- e
- Pj proposed modes as they affect power distribution and af, ability, flow rate
'l I and thermo^
i v dynamic properties of the coolant, and fuel te=peratures. j 6 4 1,
)
^1 I
'a l
.w% .a a 'e
' he **g P i .*F g "I'8
, (O h4- 1 48. *79's
.. ~-
N"l } . t
. '. j g D 4 s,.". .
A
.m. . , .
['* *yij - 7q . i
.! t i k.U v. ,,
- r 5- *
. =
- .m . , -
,q'* If
,.*1. .. j i
-s I
p.c i , j
*a2.' y ,a s 'sr k ay ._* ,,
a *
. . . .o
,'5 w , . I NvN[ ,! 9.' I'< N *~ .[Wb .," **7a. tM
- S $'.W4kY'. '
, a = ..$ YME't
.I * '
' b' ' '
s g ?r&aw*cwnwswrt&&.u sw w w w u w M vi h it d d$3 m *.tt% m M.v'M Y*="Y'm D W yy , .' f ~ 6'mc y4:Lc: a;;. w ,:y;7 :. y. :;
% , . . :.c,;.. . . t .. ;; .. ;
. 3.+,, .
' L.& ;
...;. g y . - +> , ,
L , . . .,_.. .
-c' -
- j The reactor core will be composed of 592 fuel assemblies 1
, each of which provides a vertical channel through which .M) the mixture of steam and water passes. The core will Q:* v '
- ff . have approxiniately the form of a right circular cylinder a.a.a .
.:s .. . ; ,
w a ~. '. q4. * < 4.4 : ,. Lyj
;;. n
. . .. . 1.y .;* ~ 7' 1'40 "iriches'in' diameter and 125' inches high. One hundred
...,ec i .
7;g .., , c. and forty-five control rods will enter the core from sk; , below the fuel assemblies through control rod' guide tubes. The fuel assemblies are held in proper position by upper and i lower grid plates which are attached to the cylindrical l t q , core shroud. The weight of the fuel assemblics is borne s.,
., by the control rod guide tubes which extend to the lower head'of the 5eactor pressure vessel.
' ~
".: : :: C..: .t
,'Q ,
~. . $ ., 1 M .
. w. :. . .. x , . m .t...' S 6 :al,..:;_ 7;,,.y .n 4;;,3,._.... n v .n ,s , ., n , .c . ..
,,y ;
~
g
. ,3 - - -
.x .4 ;
- .3, (q
Each fuel assembly will be composed of 49 fuel rods in 1
,j a square array. Fuel pellets of UO2 enriched to 2.5% U-235
'1
; will be contained within stainless steel tubing. PCGE hus
] tentatively proposed that this tubing would have a nominal
- c.I CJ- . . . thickness of 0.011: inches and would be able to withstand .l
,%:q < .. ...
M , s .,,.. - , u %. . dw an exposure of 15,000 MWD / TON. On the basis of present A
)y?)' . information and operating. experience, one cannot be assured that fuel with cladding of this type and thickness J%. .
,g s
=
,*g ~
'(, , ,R 68* *"'
- &* *i
, " . . x ' *: .
y,: - . - . .
- n. .
. r:
- #.T.] * '
y' .:
- s. . .
4
,. j,; t '
*; ' +s
~ vl 'w h s ,
',y ~ . .. . , . e;;, s . ~. ,, .
l h,e* 'o'~* *kN; -h . , p . ~6MTf72 $ W # kin "b '? d J' * @NI/N'% . ... , .. ,dMS'iEM.1. 3 * ' ".O "#0K I *d."
" *.it M, -
M.:Y'S ,PI.c ,
f
'E'd f# 4 N? M rG FZ 8 t$TFS E F i%dE F#f M A N G5 N iN N 4?i @ h M17'Y 5 bif @ N MJ 7?O M l $ f. .s,. ,
ef;;.h -t . x y,C ' can be irradiated for the exposures stated without
. -l
- experiencing excessive rupturing of the cladding.
- .1s
- .Wh However, the design of fuel for the Bodega reactor will 5$
Pd* not be completed until further data from a General
-j Electric research and development program are available.
s, 5,e s
. . . . - ... . ,. g : , ,, . . , _
;,,y '
' 'In our opinion, this program is reasonably designed to
- n.g .
hj - provide an engineering basis for a suitable fuel element 45"C '
. design. In any event, extensive experience with other g
power reactors provides reasonable assurance that a l design suitable from a safety standpoint can be developed, l
-t Thermal and hydraulic factors, which ultimately
; determine the permissible power level of the reactor,
.e, - , .
fl '," ,
- have only been.briefly described by PG6E. While there is
,3 . .u.
.w m m. not at present a firm basis for establishing thermal limits
. .h as high as is suggested by some of the parameters specified
-l j in the application, PG6E has established a criterion for I
I determining the proper detailed thermal and hydraulic design 3 factors for the Bodega reactor, namely, that the fuel will
,i operate without loss of cladding integrity over the design
' d,. ..;;.4 g";;l exposure period at the maximum heat fluxes possible
- w. . . . . . .
'y .
'within burnout limitations. Operating experience at other !
4 W % boiling water reactors has indicated that this criterica (( , can be met and that it is acceptable from a safety yyy .. ; d; standpoint. i.d.V
, , ,~
The power distribution which is expected in'the
;.'d
- 1. .gn, Bodega Reactor core has been estimated for the purpose d of making a preliminary determination of the thermal margins y
e
) .
, .~k ,a b _0 L$4.f h"Y$'$ 5 '
-N'D' Y : s_ .
^~ " ih ' Yh - $* ~<.+'
e' ~N
Efa[Mne:t%E~=E3Ng72$$@dd5%hSIND5010iTasA1514EOME2MA l L V y " "~ -
. .. . 2. ,
7 * ~1.' ,
- 27 -
d- l which would obtain in the hottest fuel assembly.
;2[ These estimates will be refined by detailed calculations of na W' power distribution in the course of design of the reactor.
4lG . [% s.After operation commences the power distribution will be R.
"4 > monitored continuously by a system of in-core flux
$.M L$j monitors. Such methods have been successfully used in other
.. reactors and should provide a reliable means of establishing the theimal margins that are experienced in operation.
The stability of reactor power is affected by operating conditions
' ,;, which in turn ham: transitory effects on the neutron economy of the chain A;
4 reaction. The volume of steam within the reactor coolant is one such 4 gv
. d.?/'I
, pm
. m operating condition
< r -
which has an important bearing on reactor stability.
.Q '
e Preliminary calculations indicate that at rated s operating conditions the steam volume fractions would be as follows:
) Average Core Voids-37%
l Average Exit Voids-58%
- a. Although the staff is not aware of any substantial .
~;
v32M MC . ..; , .,o , 0Perating experience that would confirm the acceptability
- . e N of operating at void fractions this high, PG6E believes N
, u;) .
that analog computer studies being made will show that the g
. . .a plant can be designed to exhibit satisfactory dynamic
'M _. performance with such high void content. In any event, .I
,%js 1 ,
however, we believe that with appropriate limitations high f.g d i$p void conditions can be safely approached in reactor tests n T,. 1 g'f devised so as to determine the proper range of void ill
, , content for maxzza normal reactor operation.
, l ,
h l' ,
- k a %
s *
- ihjh gl,;;g+L p y. '",'
""f ,h, '[ ,
- f d[ ,6.r,- f[y .47 4,
, .g.f W s '(. .g s a 4 , 51 1
_ j.72:' w G w w w a m sff Q Q 73.hjd M V.Q G y qq');GRK;*ki; % $ # 35 2 $ % V TiGU W S.M $
~
. . s,i:; . c.,*aW.WQ -
,, c - -.
'& khD'OddtN:4-
,, 'n ; , . . : . '-+ ,
+
e K74,fd %
.4
~'.,c. W :
.g-
- 3. Reactor Controls
.,d ,
Nuclear safety safety requires.that there be
;d M
#$y+bk -
reliable means for contro11,ing the reactivity of a
' )?ltB >a" -
. nuclear reactor. Reactivity can be related directly to ifM
.,. ; a e..
,,. J. w . .;...the rapidity with which the neutron chain reaction
..n g, -sygg.;bg,..g.; . ..
g ,
.... .,... . un. w ~ . ... .
.' changes.When reactivity'is positive the chain reaction grows; i
s..
~ ~
- w e~ ,
f _
~
r that is, the rate at which nuclear energy is released by fission
@f.
jr .w .is increased. Conversely, when reactivity is negative the e y. 19... '
. . chain diminishes and power falls. The operating condition
,n ,
of the reactor, its temperature, pressure, power . level, i..g .
' void content, and fuel exposure all affect reactivity.
\
~$ - -
_ In our opinion the general way in whith each of these factors l 5$ . : ' ' hy i;# ,A A ~ c.. I ,W.'.:d ;a
, a .4%%:pp,q
; ffects reactivity is quite well k' own, n and the theoretical
+
. +
., y 3A ,
, . . .o
. c,eri; and experimental methods for investigating reactivity, effects . ]
<a are sufficiently developed to permit design of the Bodega J
- 7. . - .
reactor control system to proceed with confidence.
% It is expected that within the range of operating variables contemplated for this reactor the reactor 1.
J.,g s , should be stable. That is, an increase in reactor power I
- b. causes changes in operating conditions which have a strong gy
. . a. ,, .. .
-!gdf . Atendancy to decrease reactivity and limit the power -
i.! fy$ increase. The main purposes of the Bodega reactor control
$l d . .,,-
, ng ,,
. system, therefore, are to provide a means of precisely and rwusMy
'ik;
' (ON _.i. _. . .. . . . L m adjusting reactivity ..- .
r .',
. . . p(.f Necessary safety ob,1ectives for the control system are to provide
~ ~{
hsw T s
~ means'for (1) shutting the reactor down by a safe margin under any w
circumstance, (2) starting the reactor and increasing power at a safe rate,
*k,.
and (3) main +=4ning stable level of power safely within the capabilities of o }.3 ; , w m e w o m w m e s.:
, ,g ,
g% :. . : .~ . . . , . .. . . . A
, - , . s . .. . . . -..x. .. . . . .
- m. ....,....,-c,._m -
. . o
'. nM., Jmt;*r#7R2d'.#iif M -yf.'13 5% # i@i?f M M "mPZidi$ib;b%..&te:DM4r?W$lTaCWNNtME.TG"A 7
', >: ; ;;. , y.r ,
*~
, , 4:.; f: .
p(@$,
.; han * % ><%V, z.- - . :: .
g,
- 29,-
c+ .t . .
,... e . .
i,M.
. .I
- n. -
~x . c mv :. . -.
, n.,;
. Prepose:4
'js,y , The control systenA f or this reactor is an array of we
,[T p
..n' +gj jf
, [145 movable control blades, which have sufficient reactivity
,(w/ -
~ 7 worth' to keep the reactor safety shut down, even though one A-
- ?'*:
)iQ* .
frJl& , of these blades (or rods) might be stuck out of the reactor di,$
- ~; - core. -
t
., q 9.
gg p, he k-effective of the reactor with all control blades in the reactor core i a . is calculated to be 0.97 he material in the movable blacies will be boron carbide contained
.[, within 0.17 Pin. 0.D. stainless tubes similar to those presently in use at a.y , . -
'c-; Dresden. Additional control devices, removable only through a inaMnr
- b .
Wi - ml? procedure,- are provided by fast control curtains of 0.1% boron stainless '
$M .
b __ , steel which will be semi-permanentlylocated between selected fuel elements. -
. "1 he fixed control devices provide a flexibility in the reactor design so )
t
. y, y.. that reactivity of the core can be e'asily adjusted to attain the shutdown .
w margin required fromsafety considerations. r------ -=-- - 2- :- 2 -- 2 --- -- I xitiabcxxxixxiitzhax l
#iei -
a, .s twysj hw.., " *~ m., . 6 t?$ o..g . . . --.t-. 4%, - t ,e. e'A -- -
- y 7, f W,,.m > -
ap - . s g4 ,
.. M 4
~ y;41 L '
;9'f ,
p
,W ,
p.r i .f i >14 . F e
~ ^
< t -
, 2 m.e, ct 1 . J
, [7 1
,.4 h, s 6
.N ,-
...t .. .; - , m . t ,, :, . 9S , a c .,'.,,..4.-e,...
- c j > .[ NL X g , t . . . , ,.*
h 4 f ..- z. #,
' '.WrQ_ [j }g.to.y , si.6 0,b}.,b .+ ni j. ,W
- 0 ** y?b.5, .'.,' >.Np .>
lYw..'d , \ *5*tt isit ' V T * '
'a
- s
- W.b'?'.
- I ^ W."*If ? %**L 4 N' '1%
^
E
?"&7kT;dg
$fdMU'T5MSW1-MKWWNWN'TTG@D!MMM~&Nd!W9 ' --
v. _ g, , .
;v , , y . . n . - .n - . 2 . . .
m * ** The reactor design also incorporates a liquid poison c.p
.g .i n, system that can be used to inject marcs sodium pentaborate y ,
into the core in the event complete shutdown cannot be achieved by *use of the control r . IE *'# * '" V' ~ ' ' "
# * *~* **- ~ k ~ * ~~'
. ;. . .or e. . -a vA,.af W i., Ao.. M s *~><
>g .,
hydraulic control rod drives to be used in the Bodega 3 q% ,, _ s ' Bay plant are to be designed using the same basic concepts
,g$
.- t< .
as have been employed in drives in use,in boiling water
.ffy .
reactor plants at Dresden, Big Rock Point, Humboldt Bay
!M _ ,y and the SENN Plant in Italy. Water used as the hydraulic fluid can be applied to either side of a piston which is
,'h mechanically coupled to the control rod, thus providing for '
either upward or downward rod motion. Only one rod can be mj . :.,4
>)y,H ,
~.movedoutward'(increasingreactivity)atatimeanditmay j
- A.*4
.'::e
,: be moved either continuously or in 6 inch steps. Rod speed m
is controlled by orifices which regulate the flow of water
,j
~~
away from the low pressure side of the piston. [ _ . _ . - Therodspeedfornormalwithdrawalvillbecontrolledsothat/themaxi=um _ __ at
, :1, rate of vithdraval the reactor power would not increase at a fxxt rate __ .
t q All rods
,_, _ _ fast enough to lead to serious consequences.
can be inserted simultaneously, shutting the reactor down. l M '
'Roda are scrammed upward by applying pressurized water from g- , ,;
A:p either the reactor or from accumulators to the bottom side' *
.u .
',y.x
+m N -
of the drive pistons and simultaneously relieving the kx ( .1Cff.c. 9 volume above the top .txt side of the pistons to the scram ,
#mk - . %M dump tank. The drive is locked in fixed positions by M,.
w '~
. collet fingers which engage grooves spread at 6-inch 4o ,
intervals along the movable index tube. The collet fingers
, support the veight of the rod and the downward forces due to c.1, .
. reactor pzessure.
y , , . _
- Wdd A ,
%* s Y '*O 4, f.s , s' o.V5s ?* } s 'W ** %, *4 ** f Y *** 'Y *; ' 6 ' . ' S. } ? I, ~ M*; ht ~ oa ^*4l*,.
s* ,s f & .' f f['.'~; .};' Q Y. * *TtRY l]-
"QE % A r w is r d M kR O,n e n k k & it.;4 ~j % % .H_a + p y y .p g n & pg ey S h g .w fg;M..y. . y' .;* : < y M., . .o . . y . . ;. - '
6
- r n.:: .[,*.; n ,c * .; g,
. ;.x w . ,.. .
s; w ~,..,,~,,a..,
, , j g-t ;
l
. '.' w,..d. . d.m Since drives similar to these have been used at other plants p.} an important part.of our evaluation of these drives is based on 3.$% 4 .
pg . ,
,,~ -
previous experience with these drives. This includes Dresden -
@',y . c '"
wr7p . r s;,; . . . ,
.+ ...m. _
g
,@ P" "
experience as well as initial Big Rock operations. A iW ,
- At Big Rock Point, there have been two. isolated occurrences
(- 1~
~ -
. n. ;w 3 of rod " drift.out". In one of these, the cause was attributed
' ' .s sd~ -
m.. oy p s.: to an inadvertent release of domineralizer resins resulting in J., ,7 the collet fingers being jammed in the open position so that 1
- :I i 1
m the rod was free to drift as influenced by the forces due to !
@ .4,, . -
' f" ihM x_w gravity and hydraulic pressure. In the second case it was
- d. .Ag ,
.'Ib y.w ._
a .ds> ,, ... . reported that a hard particle became trapped between the collet mm m ;
- z".a - .
~ ..
i,E . s . piston and a sleeve which is located between the collet and the
+-
T inder tube. This again is. believed to have caused the collet l i fingers to be jammed in the open' position, thus pe mitting rod
,1 drift. The hard particle was never found. .It should be noted, however, that in neither of these cases nor in any other case d ,
g-has there been any apparent a4puh8* mne impairment of scram j
. , c:.
, , i uf?.; - . capability.g Detailed design of drives for the Bodega reactor !
p
.,as v.. ,
{. i J:tM - has not been made. General lilectric is considering modifications
#ce ,
..xf4 ^
fw h .of earlier designs that vill minim 4ze the possibility of
~
'-- fcreign material accumulating in the rod driveg. The applicant s J l&et e o tw
- 6:My..
4;2Pff- ,:. ,. . , . . . , 7
~has also indicated that functiccal and endurance tests vill be
~
Ym.yy@ - , _ . _
- .W:p,- .w.. W. . . .' >
s- , ;
, ;n made on the prototype * '--- ;-- Bodega mechanisms, but the jd.Q.
;~y.
+
9y4 - detailed proceduns for these tests and the acceptability
~~ g ,
m criteria have not been pecFo**E w. t*,'. L .. s
. l -w*. g, a .- r ,
c
.W * . a;.+ n. .a. . ...," ,
e a .y.7*....e,~;,1 pr A3Sf y g ,A, 9, Q;.# ,.,vr, . ,
- n. c . , .
j @. ."A . . . .. , 9 m. e%.Q'..;f%*(8M. p.14DfS,,Wi{tp ' j cak S j J d}k $ ,g,yi.- g , . [, ygy .,5) . . .
., 7 jg!
.s m_________ _ __ _ _ _ _ _ _ _ -
_ _ _ . . a
.. ;. ~ .. .
. . . .. .. .. . . . .- .. . . . , ,. . . . . . . . . , . :.s :. .. . ..;w... ..
.. .h [h_'[1- D . $P85YsDN#2Nh.IYEDNEY 8"-NYdW...NNYbMN-N~ . NIEYh5C'N
. D @@$
I 5($ @:h:yf*: , M .6%:.: ' Ai:a.h'. : ~: t&:tp,
$ , ::<; : w 2u:Q: .y.t;isa s d W
R.. . w.
- g,f , Control systems which'are designed to react rapidly to
;j ..
demands for shutting a reactor down generally have some potential I
.'~ j for accident &1ly increasing reactivity as well. This aspect
%l:^
.dn < ,
of the PG4B control system design is discussed later in this . .
\
k M;h ;. -
. J.m , , i , ,, , ,,p, report (Section V), where consequences of a rod drop 9ut are y * . .. . . ;g;7 V
~
.n I
isU . considered. PG4E has indicated that devices for limiting n - - fjr. individual rod worth and for impeding the fall of a rod 1 i .f- .a are under development. Such devices could enhance the
/
, safety of operation and simplify the procedures that are pra presently used with similar drives to provide the necessary I a.ssurance that a rod dropout accident cannot cause a serious c
s.u --ipublic hazard.
.x .,
s ., - S q,: - .< ..
.A.'c, .,.n A,. yew.In :a view of the importance to safety of the detailed .
- m. .
"7 design of the control rod drive system, the staff believes
- that PG5E should submit timely reports to the Commission during facility construction on development, design, and testing of the Bodega control system. -
. .,,,u _ . . .
. g..-_ .__..,.......i.,....-_,..-w w.e w G f 4.6 6% wean w - - -
. s
.; .-.f.rwwba.J ..
O' '
-I
.y**
- . p . , 4 -- s3, ...
* . - > ,
- e
, , s p.y. . s... -
. .m.y- .; o t .- -
. ;:q . .w 3,
.y ;q ,e 64 *. ,
, . j) '*
.~. -
# )
**e
,'4( _
g r
- g s .
.dg
.x. m j , ., u. .
.i yD '& , , ;
.;,_ q
?:f.,;q )
- mn .s.'
< ry +..
- %;e -
Nb , i s' , ~.m . l. e
'N l
s ., L
- e. .
s ;ta-
..- 4_ ,
~
,u I c ,
o( , I-
,y * .
4 3 h
. ,4
, 'A1
$ se ,
- g j
( . ,b**r *k=NT? * ,-ea ( }. h,.( . ' ' ',5* f N '[. k? ' I h ., ,' . ** 'I' * ;a ' 2 ' ' Y#
* 'f4
- a'f*. ' *' NP *
- y QY5&WwdhE%iQ9fSh5 S$5 YihY$5 O h h N Y & Y N W 5 % ?$"H' QQ , tylj . g . ,*q, , ':y : .? w . < ,
$ 9, . "?;; 6gd '! 4&re': s:..
,. ,, A "
~,,,. l 3 p; . . . .. <. .
g, . . . .
- a. .-
, 33 ,
+ A
- 4. . Control' and Safety Instrumentation
-' .{ j 4
' ~.
~
The instrumentation necessary for safety in a
, i x,h.l nuclear power plant generally involves a large number of r .;
W *. sensors throughout the .various process systems. These y%
. t .. s $
s v - y g m.v.. y l ."r".. % < -- ~
- g. .~s,.
% n .O .i sensors measure a variety of variables, such as neutron flux W.
. E.- and gamma radiation levels, and temperatures and pressures R -
' N,$lj ;
of various fluids. Information collected by the measuring t
.1
, instruments is used to guide the operating staff in d
. .j controlling the plant and to actuate automatic control d
,j devices.
3 .
@ The instruments, circuits, and control devices which are W .
ggj
~
.i1
- rie of most importance to public health and safety are: (a)
. . - u. .
4 3;.,p -g , o;. those necessary for and contributing to stable reactor operation, 1 :
. .t.
7
.(b) those used in control of radioactive fluids and effluents,
,a
; and (c) those used for control of emergency equipment.
'j- A general adequate description of proposed instrumentation
.1 wMie % desce W on h *&M is presented in the application.3- this time, there is not j sufficient information available to determine whether 7h .
.s y .
instrumentation provisions have been made for all essential Q5 functions or to determh$e the wegree of reliability that should
@,f 3 be attributed to the reactor protection system, which is
- E A.
, , _ ,, ,' described by PG4E as " fail-safe". These, however, are design
~
.' problems which appear to be recognized by the applicant and
,.gs;<
+ ,
'f
, 'which require only the application of well-known engineering
'w c..
. methods to provide an acceptable design. The staff intends t
9-g 9 *- N ' *
}.'
~ q:,,k ,
$.$$*QW.$k$&Y$,, . u.f kf t lh&hh.h$kt N
$ $hhN ~; d *Vo*1 sS* U:t ' ' &, E$ D ' .
..7.t m, ,- w..%. .m-1vce.-a.wW:W.mm,a . : t . .~ . . . . . ; -m es m e:m r e .er_~c r.e.me -w.. n. -
nWp...u -wov.wceme .. 7.: :
- . :u,
&>. ,, 1
~ s y . . . e,.
~,..
;,$.p.d
; i y .. ,Id l'F,.e, y-i['<f 'f pf,3 g
,E ' 3 -- , - M.i .
4 I.i
, y, .
4 a' ' D1 to evaluate the' reactor control and safety instrumentation i in detail prior to reactor operation in order to assure that I, proper attention has .been given to the need for automatic
- 4. ..
l'
~$,g ~.
.. functions and the reliability.of safety instrumentation.
\t
,.. a. m ,
Q4 .. 6 - 3 i
- n. +
_q ,L.
,+ .,w %3 l ,
x,,,,.;
, 4 .: ; -
+
., , .4 cp , ,
, , 3 :. + - 5i.,...p.r.;. .,r ., ,
' Y *: I
,"W A4+'4^""' C ada wuA De RGGed on earT,nquakaa.)
I a;t . tR f , Q" * [ i i! . l-
- Q p. .
mo-au . i lY. , . . ' l
. b; - g Q
w
/* <E 1
; .': ,t, .
'a .f a
. t
,' " s J'
J, i 4 ,I* *
~ '
[ r s,e,.- '.
~
.. 3?. ,- . a.
't.
+V.% A g 'O
'as f ' T7h,, se - g *c ,
9
, y. ,
m 1 t i
? !
' 1, l
l b 4 1 4
.t e .
*M..
4 tk %i5 9 a s 4 -' q
.r.; .%.d <? , . > us. .
~ ,; +c. c " .
, ~
, ~
..t . ' '
=
rw .- I* ' f,,. .'", g , , l
,-- i.a .
..M r .*
+r'.,
. I 1,
..e. garg t 5t
;'..qdj -
A ,j, , , ' * ,t- -
.m * .'
.Iy . u A i .f e.' e+. ,
f d d'*ig $ g g b JW5
,. , ~ $3 k. .
*s's '- Ij 95 iu g
re . u l 'g , .p. ". f G l
. ag
{' .p
** 4J t , ' . De . .a*, l .- h =[. .N, .,
s ,J d " '
,(# g ,, ** #
j
.',A .
'=, *= y,
.e v .c
't L m,
',r, g f ~'.f J ',
- 1
,a- 1 +
l
. agi 9
.s .
4 t 5: , e
' f'a e I l 1
i l % i
- ", 4 9 e
"' 99D f , , ,.4
$ p g
w .
. e' : 3,,, 6[*
% . ' , . ' . y p V *,
r.rs.f . . ;& s.m m'.w'g. . ' , +$ ,-.- . c 't I
. cgg,y 6.W* ,,.,.. Qs,..w .r :. v .' .%. : ww.
.g g, u g, .p.m a'f%. 4.p.g *f* I '8 'M ,
- s. > * / *
. v1;..% .
,.g , y',
a..
e~ 8 a. ,. ., . .. $e W ' s c, ein v ., . , * . . - , . - -. .. "s c < ' ...._I ' . a5
# m & 2$;T X': m n @.Ph?P'=&TnTOWF7:W.~DZ'*WW!'"*;'* WJ';9 W' * '"W' W W.'* * **'?*GY" ' ""
Iy - fg ..a >
,- 'f.,,- , -: , & ,
i l
%.r. w
;dp;t W % m r. .
> n:,: *
;,1 - w.
d .. . . . - . f. o
.m ,
s ., t
.- *'I' -
E. Energency and Safety Systems i
~% Energency systems provide means either for safely -
~
djdj . continuing operation in the event of some equipment failure
'QU% ~
,? or operator error or for limiting the extent of Mamage and
.. ... a y ~ ..c..+.,-%.E ; 7.. .. ,
resultant hazard. In many instances, design factures of the 7w%
- .b '~
f6-6 ,T ',- ^ facility which have been provided for the primary purpose of W,9
- w. .
rnaking plant operation more convenient, reliable3 or economic
/- are, in effect, emergency systems. Other features are y
. designed primarily for the purpose of providing emergency '
functions. 3 , The principal emergency systems and components proposed
- an .
.'-L
@eq,e ".'. , ,: e for this reactor are: .
......m,
.-. ,. . t
...,u.:,;:. o " ,' : ..:a e
"(1) AltArnate power supplies for critical electrical
- Ny
; loads:
1
.,l i
(2) Reactor control safety devices and circuitry; l (3) Liquid poison injection systems (4) . Energer.cy cooling system
,!l
.a ;j (5) Bleed and feed system i
.: ~
@ .,.;.....,y....as . -
'(6) Core spray system
. m% , ,'. n.
{iy ;" . 4 (7) Containment system 5h+1 Some of these systems have already been discussed in 7.l4
.m.,- - ,,
? ., ,
,,r, .. -
pd > - Principal features of other emergency and
. fy . - :
. ' v~,o this analysis. . a. ,
l@ L ' . f.' r,. safety systems are discussed in this section.
.:QQ
.. c. .p . . ,.. ,.. . ;;.
. ,. . f., .F . . ~
's . .- ,
m
'In 'all sucli systems a high degree of reliability must -
jll."g . 5, g.c % ,- . c.i
.c' -
.* be provided so that the system vill perform properly in adverse circumstances. This requires not only careful design sf.i
.n 7. . . . . . s - a. q_ , . , . . u:, -
- a. . :. .
j, ; gj V4MEPcb. c z;.a M xs.T.itI .... @ $;FO .. x @ $1td c.#,P?' W G. W e'/. a ,,.dfr'.$Nj. , , ,n.Q ? NW .'d' cW NN i m.'MD.". 66, .' '2 T5 4 ,
- iW/?M.;
_ _ - _ _ _ _ _ - _ _ _ -~ - M r@ t h d d tis w h- WP4M~r E M M ITI M *$ 9E'F: WP51N96N^TN D'h"N l g l..% .c.? .n . : .L l..s. : . . . -
.,.s.:
~
m.:3 , * , . cx .
.. v. m ' :, _ w. ,_ .
- Y il'f9w Q 36 -
f' 1Vg s:
,f l - of the principal features .but also attention to such related W
M equipment as signal and control circuits, power supplies, and instrumentation. Maintenance and' frequent testing of
.y . .,' emergency
.. systems' provides the final assurances of readiness -
m.. ...
'~ '
Mj
~
m of emergency systems to response to the demands placed upon - a , y9 them. These factors must, therefore, be taken into account mu . .
;.,%.M -
- in final design.
iM The applicant in his analysis ofoaccidents has taken credit for the c? , . effectively
/ actionofthecorespraysystemto/reducethecalculateddosestosersons m
offsite. As is indicated in the discussion of accidents in Section 6 of p . i;his report, a number of these emergency systems, mainly cont =4==nt and 9 7, u ,
'C emergency cooling systems with their associated water supplies, power ,
., 31,.. ~
.. *,,d ,e- .. p 4g , supplies and controls, must be relied on to limit the consequences of
. :,7 4/4 serious reactor accidents to an acceptable level.
.} - _ _ _ . . .= - .=.,, . , . . . . . . . . .
1. g.! Power Supply '
. , ' Protection of power supplies is provided on several levels, as described in PG&E's application. The plant is y tied into the PG&E distribution system by two 220 kv circuits
- n. .
h <. - . -m. , .
~ to Ignacio Substation. All plant aw414ry power requitw-3,.o -
' ments can be met by either a transformer tied to the 2.,0 N
1 istation generator or by a transformer tied to the 220 Ev i -. y f'
- 4" ['
- i. lines. .An additional external transmission line and m , a . , . . - 1 31
">(transfozueroflimitedcapacityandanengine-driven
'J"' generator provide emergency power to equipment necessary Gl,.%.; . .
$J0
% for safe plant shut-down. Station batteries will supply
>:3 , .
JL the electrical energy for the more critical loads, yp . ._ _.. _._; _., ji.-- - i r.. .y. W':_
- j **
. w.m,
,w.. } -
, . ,v .. . ,a.. <.. s , , . .
p ,
,' % .* & % i. R.] =$ ,Y,, f Y.W&?# h'rhy*,W'WxYpsWS e A.%h* fey kY*? $ ' *'dh . , 'Th
- Y.U* I",*t Af l V' ' '
+
- i "
- 1** O Y' u
m&-*:-*%m W- mndMmvym%v %&:'sw.rm5>WSN5&W'M ;'T 'r"? """:'
- p .y,;l _ , ,
e.Y : , .r r ..,D .;
-'*N;., ;g. " . ~ . ,.,_ y :l'
- n . , .g.>,.f 37
- c, ~)
, j
. ,. 2
~
)
- 2. Emergency Cooling Systems
'1 A number of different means vill be provided for t.- .a removing after-heat generated in the reactor core as a 2,oa c w.,
mp'. y,Q .*; s .* consequence of radioactive decay of fission products. 1 Dld ., .. .
.-e.u./ :. .d Such provisions' are necessary to remove decay heat after fg ; ' . . .w Pih -
reactor shutdown to prevent' melting or rupture of fuel
$N . - '
pgjj elements, which would lead to the release and dispersal p r 4 a ., __of fission products. These provisions vill include: c;.
.r..-
?'"
(1) The normal. condensate-feed-water system 1 (2) An emergency condenser which can be put into lef,' operation in event the reactor must be isohnted
- kh g ,,.-.: ;.W from the main' condenser ifgg .. ; p; - 4 , .. .
My ~ 2,t;.4 e . (3) A low pressure shut.down cooling system ' 2l:a '
'st (4) A bleed-and-feed system which releases steam at a o.
'd .
controlled rate to the suppression pool
'l' (5) A high pressure core spray system
'i (6) A lov pressure core spray system.
i
.t
. .t A number of sources of water (and pu= ping capacity) vill q.; ; be available to restore water lost through accidental ruptures
.;;;g.
p ?'-
.w
.))
1or through bleed-and-feed operations. Both high head and low
;, . some '
head pumps vill be provided with back-up 3 pumping arrangements. Q} W.t
. In the event of a r.ajor rupture of the primary system, gg '
emergency action'should be capable of reducing to a great l gg . YTM d extent the amount of fuel damage and fission.. product release' .
;$w'. '
from the reactor, s, . . . . . ..n., n , . ~ . . . . _ _ . , . . - - . - ~ - ~ " * - ~ ~ ' * * * ** " ~ ~ ..
= . -}if.'} .
l
,O *
'\ wF , .
n l;s 1 , ,. 2 - El , Sh. F e. b e ..+=6d, ,h / Ykh>G,se OS;Y'N0?:mA*WA ;g.H Y." j
' ? N *c y ** G L.bS* ' ' ' E M*W # * ' ' ' _ ;
jn.*d. -446x,- &.wrA%%%:?-u&M~:$%&A4WhW%% W ~ ~ - 9 M~+ ?%iNAU = ,. ,
). .
~ ',. ~
?.S 7a y% wt(pulmW:n M:.2:.gS ;;.3 1 f :O'N -l- % . l.{
Q u o; , v w u. 4 . ._ . . ... . . .. 4.h%- . PG&E has indicated in Anendment 3 to the Application that the a .:.s
, ' emergency cooling system vill be provided ziyk with pump backup s.
yi beyond that'provided by the aurniary ' - L-- -.--_- L (startup)
$g e e 3
W
..teedwater pumps, he' staff believes that the plant should be provided c x.a .
],$ , with both high pressure and low pressure reserve pumping capacity beyond
.- .~ . y . . . ~ . . , ,- a. , , ,
'), % .
l{ ; ,
'that descri N.d specifically in the application. These features, in'sddition
^
. h.. . - .to the final design factors already' discussed must be carefully. reviewed in' a
n.c, detail. There is no reason to believe, however, that the suitable final ,,
.c . .
f.0,5 design for the emergency systems cannot be made so as to provide for the gh,gn - - ..- , -.._, .. necessary functions usually encountered in nuclear power plants.
; ~ - . - . . . . - . . . .. . _
l'.,.
,) 'f
.4, g' RA' % g b ,
- M Q .j
.,x ., , ;9 ;~ p -
gi, .
.. +lp' ,
,a
. 't j'
.n'$+p --
me
,6 jLt'~y. ~a44 ~ .,c.. ' .
. d 3, , ,,
- o ji , -
i . p ' i @/'. %5A 6; f* * ,e c ', 'a j g2y- . , , .
~ '
e.g . <
, . . .p . ..
, i
.Y e( ,ng ah de
!* ~1 '
f 1' '..
, [r';, g . * * , . , , , , g 0, ,- %,, . .s . .,3.&'* 'x' . st ', f. % , *
- y ,, ,
s.p
- 6. .,,,.
J 1
.,gi
,q.v
, .J y .. .
,1 ) '," 8 4g . + ,
,(,,,(. *
+' *** h: '
k 21 %.w.c :. +* -r e-a ', e '
- 3c m ]
-T. '
4
.b.
-Q;:a:s
'9
) .
, ?, ,
g I.. vN'$ b - 4 p: ,' , .' ,
-.s.q..3 ..
a- s ; S. h.3 L , , , . i, , b hf , 4 , , 5 *
.g . C *
- h. *,
., s ., - .
'c ' . a ,, a tl ,
*y*.7.M.1 &
2.,t d .
' %;. .
- i
,s
__j y , e
- t , ,
g
' 9 ,PD9%. g b 6 , 3 , ,
h ,, ,1 " ? t
- M .:N,p M.pk .Q E n - ". - , 4 /- .
>*s , .
- - ,. e , ,.t.; - Ai ,
t ."'dfMkMM~MrMf@N7.A:,5^7$M
R .
pw hi 6NW""
-% m s fNvuw'*?ra "m .s* * ^ 'M'.'~"<* --
m s 14' '"? M..? ^l#.~wu t.- -~ , sc ~rWC':"QMW-- i l
..r.
, , . . .' 4 *
- m L*-c.) ,u.<:w .v k *i...?:. -
4 4 ,~
@4, 3 .
K 6 4. Radiation Monitoring .
- w. , .
Badiation monitoring equipment be provided for two gp purposas involving safety of operating personnel and the
< .v. .
H~y
'M general public: .(1)formonitoringofradioactiveeffluents s~ gi
_ . . , , , . s .. -
- and-(2) for deterining levels of radiation in work areas in p.
_,y., .. 2~j the plant. - A n ., . e , , i:dfr . . 4Wi ,
.~
> r,
.. e .
t ".'.0Pf'.
}
'7h The following measures, which are typical of other
.:j ~
. E.u. . facilities, vill be employed at Bodega: _ 1,; (1) Batches of liquid vastes vill be analyzed p g .; v- - g
, ". radioche=ically prior to discharge to determine v.yw . . . .
[fh W, -
,, $ $ 1l. d . /..~ $ $ [ ' , i. quantities and types of activity present; .
(2) Primary coolant and fluids in various auxiliary 4 systems will be monitored by radiation detectors 4 i
~
j4 .,s and sa= pled for determination of quantities and
;M.; types of activity present;
, u.
(3) Continuously discharged gases vill be continuously I
'. k
. j,. ,% monitored to determine g quantity of activity discharged; F.t
$d ,
(4) A program of radiological monitoring of the environ-4
@. n,.; ment vill be conducted; and
.1; i 3?i
.. m .
(5) Fixed and portable equipment will be used to measure j
%h *~
radiation levels in occupied regionso[N F land.
.;q, p , .
- - ~ . . - --- . . . - -
The type of plant proposed and the environmental conditions do not *- -- - - - q;Q..., s.- v . 1
..s W.
.3M. . pose any unusual requirements for monitoring methods and equipment. We ' I
..a
,,': /.I believe that proper monitoring equipment and methods are available to 9.;
fulfill the requirements of safety at Bodega. . M 'y
,',g.;
f f . n ,:c % % % @ v. N w :. x 6 % A M w '
- g 9MrnE*tmnMwi-rErd1%*;h59W~rt m,_
e+tm'"'M *M tme m EnmetMiar.,v,w w
&i 9 -*>< e.: :l ' t.m c.;,' .W i) .
2% m: ..,. o 1,
'n :n C I, . !G. Waste Treatment. Storane and Discharme'
- q g
The applicant has de, scribed in general terms the i&.,
.s radioactive wastes that would be produced during operation
- . . t
, ig% ,
oftheproposedreactorfScilityandhasproposed' general nh ,e . sp . , . . y:;g [~ methods for management of,these wastes in order to 7.Ag ,=
. [. '
d 44.. /,/ ,4 4Las # , 3.QQ , . , , meet3 10 CFR Part 20 of the Commission's B,*guistions. MA i .9 The sources and general' character of these wastes and N i S- the general methods proposed for meeting the health and
~
7,( j safety requirements are summarized briefly in the , w:3 . following paragraphs. ' s y '
- 1. Radioactive Liould Wasten y%j -
. :c.
.. .y ;a The principal sources of radioactive liquid g.,
w.,y m . . # e,9 ,
- v c .: ., w. % '.#.. -
d wastes from a plant "o f the proposed type consist of
,, 'email amounts of leakage of primary coolant from a
4
. valves and equipment when maintenance is performed,
.A.. .
e,1, r , < > and wastes from decontamination procedures. Other
.u.
- . .,.[ sources include laundering operations for contaminated
,, clothing and laboratory operations which are carried
.;y out as a part of the reactor and power plant operating
.m.3e 69:
control procedures. , q;%
<.d4
- The amoung of radioactivity in these liquids Yj pd *aly ,
fd ' Mu
. . *<* is variable and will depend upon 3 the concentration {
1 qu
. s,- t . . ' present in the primary coolant water system. The J
[h-
. .t *
. . > ' radioactivity in the primary coolant consists of ' N
;l'M ~
- u. '
' fission proddcts that may be released from the fusi
%g;i i .
. fa s%
. :.M through minor imperfections in the fuel cladding. .
/ .
)
4.s a
, i j.% r, 'J'
,, f.C
'v%
f.) s '. \ ws y a * , . s' . ' . , .. ,: C l
'c,'
hos k f:'_ S ( pp V', ?'. N,*
,, ' L'\ ,
- Q , c 4 -- ' e s
*;.11. lx3 r.* a n* l6
o I h!. h645$5$i+$$Mi&?-n?MV5 a .
% %*O W & M? % *=W M*".!OfC4% % +d"& %c'*iST & 1
+wy ...m.e.. 7 . . . ,e n. n w. rg.y . . n_e.. .....-
.,v .. ..
s..v i
.;W,';'lb-k,.,,r,'
-l 4 d*
j .*l,. fr.y.f .*?.l,Zf , %>;? g.
; J 5- '
'N. 3g
- . ,.y
- ~ .. .
q .. 9 . p'" .
- (6 -
, m: ,I .
{h.
-3 i.
;W-. and of irradiated impurities that may be
.r :. . . , .
~,.:hh '
present in the celoling water. Such' impurities
.I. ;.;,. u., . . .
p .... . gi e, .
- c. ... .' ' , ' .,[
~
~i " '
i,..would include co'rrosion products from the coolant 1.... g .a ,
. . . ; s . ,.~ .
- a. m. .. .
Q sq O D.0
? '
- f system and fissi'$n products from traces of
.e ', .
gpp J s.. :" uranium impurities that may exist in the fuel
'E
.hT$ g ,
Nj ?n cladding surface. 94
,h Present expeilence with this general typ's
- g. 6 g/w of reactor at Dresden indicates that the range 4
J'e g.i of radioactivity' concentration in these l'iquid
. ./.;.3 1
d.3: -
. -wastes may vary Eiom values which would be low
'.gg('d.g .; * ; . .
t *: 4.y g, y . ..,,. wm. . A L, ~.4 ' ',#;;enough to meet sth' drinking water requirements
* .c b ' -u
.# 7x-
. .eu - +
- .e.o.. 4 p ' m "e.v a .-
. . of 10 CFR Part 20 for the public with'out treatment
.m tw V' .
,;Y- . to as much as se}veral microcuries per cubic centimeter. .
M. The volume of th'e'se wastes may vary from a few 2C . WR 3w tens of thousands,of gallons per month to a few
g* f.#' e
~~ l hundreds of thousands of gallons per month.
pM. _- The applicant is proposing to construct a pm .p1 k;g
~
'special system of drains and tanks for collection q
M;g 7 ,,,
..h , 'of the radioactive liquids from all potential
&& ~-
r
$'M<, .W >
' sources, and to provide a Radweste Facility for ' '
e . . ,. . 2..+f . monitoring and decontamination of these wates. wp - _, a -
, {bQ9
- ~
, ,. f Radioactive 11guld waste will generally be disposed ,
y 1
'y ,;
- f
" - ~~ '
@ph ~
. f by injecting it into the condenser cooling o
1- n - 4%@ wrt
. iwater affluent stream after monitoring for c , , . .
9'* ' radioactive content. . This disposal will be so
.BM -
- o- -
, f6. -
.) 4* f 4 (' e. . & 4 ,g[*. .e I V -
g 3 I c..- -1,,- i - Qyygg,.4 .fg 7 . .r +.w. .~... ps.tsw, 4.
.c.%w,h,w.'.7.: . is
. 4. ._.bo we ). m wet .. , ,n
~p. _ . -+N - ,
- , .r
.fgjyr m.9'xPW':'e:4W WT*?^%-%4 9 & P '$vte :up W M :% s w 4 m pr' @ nto g
& -i
.19' ^ y n % .qy& lg 7, . :,g.m.:.y w;y:v.4 yy. n,y... .
- e.
.;w-+
, D.
.b
- :W
-n
, . , . w
..e
. 5.~. ; . . ..
.it
.C'
- m. .
.c . .n c ,
41
/
j d.
.M'. ,.
W. g 4, controlled that concentrations in the effluent pg'J ., w, . W hg 7 . a v.c; .'b.u w
,m .e m. stream at the point of discharge to the Pacific
~
i..' .g - - -- ex .- - i. ijj@g gq s 0cean will not si.cceed those specified in 10 CFR 20.
,. u ', yw hiiiX.
;e,:%f,U ".Z/@.L;:.y. .%:gggpg. d. . W./53 .
g f%q -
- v. ,
~
,y- .'" The staff has reviewed the general concepts iff y J.:,L .c;V ,JnQ;.
u EDj, . ,
. ;1 r4. . S M" , proposed by the 'a;pplicant for the design of .
v . ,i , this system and believes that the concept is adequate
@q. A.
l]
. y', . i.
to meet the requirements of the Commission's Regula-tions.
. 8 di, 2. Radioactive Solid Wastes S! . l .
. w h. n -
$$f at o,. . *
..The principal sources of solid wastes inc1Bde :.
.? w.e , c;:,
;te,4. c.y;dSteg. px; .,.r .
oy t'
. ' . . . ... s p.c ? A y k , ,
*i
.W
- .u - y -
. .f.m. . .
N - m . A.g' 'rg; q.g. f.e' .-r.
. .. s.b'g; . ,4" .os :d.es.wTd.w ppent'domineraliser resi p
. ,2 m
. .f-M E JF.,* equipment and' miscellaneous trash such as hand i 3 .
J - tools, laboratory ware, etc. s .4 The applicant proposes to collect the radio-active domineralizer resins in storage tanks .
.i j which would be located in the Radvaste Facility
.g@; . .
and to dispose of these materials by shipment nw;
.w.3
. . s.
[.l
-;f.g y.;s
. .: s
- .t
~,h>.
- c . , g,,Cf, f. ! G 4 , w..
% to an AEC licensed waste disposal facility.
.The shipments, of course, would be required to
. y - 1 4;
44- sy . -
,, n '
meet the appropriate AEC and ICC standards for
% s. .u
. . ~y .
.:,4 . . .
i Fm). u , -< < ., g ., .
- . .h .y: shipment of radioactive materials. All other
. ,- 3 Y "~ ' ' '
l rY1 s' ~ , r!
' ,' . 6 gg '
y, v. '% k E !.1 '
' solid waste would be. collected and stored in a .e .
. ry v.%
.. . : . b ,, .
x,.nt,- c - n>
. . ff.. ,
. . I,.. '.,4:,,.'y g vault constructed for this purpose pending s
m n t ,,.
. packaging and shipment to the licensed waste
'I9'O
.) .y- ,.
x, . . ..x.
~
" ' * ~ disposal facility.
'. ? ? - 2., .
- 4. 4 ,
6 df4 . %- 'A- -
/ . ' J, F9 I3 e-4 g, _#
' .? tim * .. ~. s .'%. v a k .s .n. c ' ..yep,*,s .. . &. .g s -
, . ,j -
l ;: Q .$ h.. * . f y. 4,.*:y 5sQ.4%%;p,.,,,f, .j*e : .fpl.it;,v
,;.y;l' yQQ:.Qi;,(yj',4 , , gy + , g- '* . -
[$26YEd is.k + :,'.*. ..
- EMDENMOM I M fE;>;, D E-IGB ~ ,*,
EG5 t324 N F ".? %.T N ME T Q
- ::.a. :.O w ., .. .
. ,ay , 9 g. .n., q:
.x . , ~ . -
w,. z a.g .
- c. ,,
- 3. . . ,
e .. .. t-
-l , .
'q' ' 'i l q
Thestaffbe51evesthattheconcept s.v
'
- proposed by the applicant for handling these
,dy .
sw s=q .,., g.y 4.$ 'si solid wastes is~ satisfactory.
...;.,- ,, a .y. .;y.
A' sin
- J,.m m
' ,....: , s ;, '- :: 9@,C... ..w l
3.. Radioactive Came,o,.us Waste
'."'.jj N' ,
,'. O "B.. t .
m4.c. 1 l
' y@.
' Gaseous wastes from a reactor of the type ;
proposed consist principally of non-condensible radioactive gases that are removed from the main 1 condenser by th5"' air ejector, and from the turbine
~t gland seal system. The process areas, laundey
[d*il -
,,. . ..:,' .a.a.? . . .
4;; and laboratory w'111 contain trace amounts of
,co ,
n.v ..,4 3: . Further,
$w d,?d$Y;' .
w
.s a , ww .%.g.2_?;R. -
.3 u.a..,;r c 3
.?. .,:.1.; contaminated dusts, mists and vapors.
,. W f c p ,i , i. , - , the dry well will contain radioactive argon i.
q as a result of neutron irradiation of the air within this cavity, altihough this air would only be
. v
- s. released to the stack on an intermittent basis ,
l 4 whenever personnel access' to the dry well would .
,l *
,--oh. ,
be requiredd for, maintenance purposes. , j o
.[#N t.;.bQ m,.;;m; .; < v;
.;. ,f .
,# . 7 The air ejector off-gases, which comprises
+ . . 3,1,. ,
"l 'N. ~ .
.4-the bulk of the gase60s radioactivity to be I; cK .- Q :
1
< s. . r released to the atmosphere through the stack,.
(.l - } } , m -
.M.y .g)
~ ' ' ' '
' consist of various isotopes of nitrogen and j.
o t 16 Y* 1 ...4,. '
.^.
M , .
, oxygen. The off-gases also may contain non-yQ'j .
,j W s.y y - .., er 'f .
h.j m
. 1 .. k qa condensible fission product noble gases, mostly c ,
;y.~ , xenon and krypton, that may leak through minor s - 's 4 w4, . , a-
e.-
M; , ,,..
+ ~ s',
=
l
~
+
?E '
)
me. w- -=. '['q .-
] ,*s ;
+ I, . ],_
s ..' 7 , * ?. CC e p , _, !,':,6 n.J m;$ $ $j45Y$l.If $d R '$,fij & l,:h; f ,)l. & if.7.,.Q, Q :g ,,:;?,;; p,, gg . ;' , [z ,,
i v.,.n... w ep.m n r~,,eg~ wec:.e x.r. g ~ n~ m- w A:1.emr. n. eym-~:. v.w.5~ vim > wk + i ,e
.c ~. :.
. 3 .
- g. ...
, - s- . W. , . .s.g Ryu :,y. . , ;. ..;, x p .e ,g M$g 42 'i4W 'Wlii%r,W$svG:&iQ '* ns. :n O
e
. .h ~
, , 'I.
i Ej 1 imperfections in the fuel cladding into t ,, :I
." .' > the primary cooir. ant system or result from
' gg - -
.. y
;w <n. 3
.. .7.-
y .
,,, i
,A,,y s - u. y.f. ., ;g ;;. .f . _ p ; ap% .. .: .g irradiation of trace quantities of residual j
,y n .,g . . .,..
c g.;-
.w . ,. .
.; 3 . > ~ ..r % . w % w
,,]p.,. m ,e .
. ;.~n l3: uranium contamination on the fuel cladding surface.
m.m
' ' '" ~.i;,> .
. .- .~ 'e.. :- .\
n ~3 .
~.~p" **
cl n .
. c.,,.g,p
..3 n,:n . The applicant'has listed the isotopes of
.j .,
" e
)If nitrogen *and oxygen which would be present in 1' the air ejector'off-gases as follows: N-13, i
l N-16, N-17 and F19. Based on the experience I at the Dresden Plant the staff would expect that . [ 7,n, c E r.'.. a-J .- a x. ..e '..% A f n Nitrogen 13 wou1N be the major contributor to 43x .~; : . - o; g i
;n 3[ . . . . . . .,a.he total gaseous radioactivity release to the.
- a. .,mg
.. ..m.~ . te. ,m:.n~ 9,.. -..
t
.e . :v:.{JMA. . s - me . .4% ,4.i,.9.m..... W , y .
w: n
,M./.. . ch - ' atmosphere during normal operations. In
' " a this regard, experience at Dresden has shown thatconcentratibnsofradioactivegasesreleased a ,
~" ,
are well within the limits established by the Commision's Regulations. J,7I . The average annual rate of gaseous radio-a.. e.c.v WL.M,l . ~.. . .:. . m
. . M, ctivity release from the stack will be limited c +.,9
. %. '. i. 4 .'.~%
r
~4.f. 4
@q$ ,_ Y:5 W: !.to a specified quantity at the time of granting ,
. .e . . . ;.~
v.
. u.
. of.an operating license to assure that the
- .u.w
- 4. .~ p. . - .
eu 3 : M
.. g, .. .
.y .
7.. E .,. f n..
. at.4-*
~
..,..],requirementsoftheCommission'sRegulationsare i
l
*':., . , y glc. ,
.. :4 , , ) met. The meteorological and topographical condi- ,-
?fn. % . Q :,e z .)i.$ !M :::% g" :;p - .i st 5 ..
,py .
.y ~
fe.tions at the site as well as the engineering we, .. w
-s. ,s
.. ., , c ,. s; -
.w ? e ;'
- .g'., ; . /
- m. w
, design of the, plant will be taken into account '
7 ,
.r
/ *.$ - -
.when this limit is established. The applicant ,
. :.7 - ,a y,p . . >
"(; . .- ,f , <
y ( ,. . _ ' .,( fej. s.. ;g. ss ,
, - ~ .&h.$ l $. sk &{Q: ,-ls$"*Ejf';t[h($l *?hh. $k'. ' h 'i.._. L '
. , , "' L } .
4
- . s .- ' -^ . * . * . >, , ., g a .,s s. . ,1. ,,a s . . . s s. -
g f,D %,-'iC . , M' l>Mnu' $h. r:2M't9N
- v m va'%P W,' * * < W W W rrwm'+'
- e "43-7 id>'er" ; ' N' ~ '
. , . . , , 1 .
g,' - vy: ' .
- , , y t:q -
.e a .
N v - .
. , ,. s % .dg . , . . . , . , .
. , P j -! #.;' 3 4.
- i 1
proposes to meassre the amount of radioactivity
' ~
A. l released from th,e air ejector and instantaneous
,...c&.;
t.
. ., , 7s',.j. .-~
7,M
./q m. - : ; ., . s, . - +- radiation level ko provide for an alam and for .
.m s, . u, ie closure of the off gas ' vent system valve '
I
- )
.w.. ;;m. .n ,
.. ~':
if th<t radiation $. level would exceed limits pre- _J 8
?
, determined by the Commission tolbe acceptable.
s
-T ,
Theconceptproposedbythe,japplicantfor l measurement and Tontrol of the radioactive gases appears to be an application of conceptions jE.a , . . c.
,. . , ' , .c
~
e , that have been previously approved by the Commission p g=% , ,, g. . .
'd
^
~
~,!. V . # .._.for other reactor plants. Accordingly, the staff .
M:%ec . m '.u.W .;J % . :.. i" a believes that the methods proposed will be ade-Q:.' . .t . - . j quate to satisfy the health and safety require- f
~j ments for operation of the' type of reactor
- '] .
.i facility proposed.
'i .
'% .6 m
- 4 l3 G ' ~
. o
..v *. ' . > ~ -
~ ,
', j' i% W. . 4,s ., r , gl.. .... ,, , .
7,3 .,,..- -
'.; 'f C -U -
p
, is f5 ..
p ;d ,
,,,. +. <
,,- (p,94; e
- e*
.f o p y n.
e
, ..d . n f <
lQ;
., .. .m ,
+,. gr . . .
, , , 3;, '
f s
' .')
.l )
- u. -
- .i
- 5 3r 4
*;J,- { . ',' ,, ,gt,n ;.
... p. . . . . .. . . .
~
W :!#c -
- jr
.W) , . < r. . . :te ,
- 4. .... .. . .
, . g
. y 5.*b kp p
- e e, .
h .k. ' j i -- ., ,
' I '
+8 t
(. ,N * % 't n' e 'M P' ,h 4 7,'M....., . . . . , at#. ..,'.,.l.'.. . . . .3 < ;- p. . u _ , . .; , (. 't i , , . . ,, ..m ..,
- Y. _ *" _ . Y___
W* Y& Y : ' ' Y ~ *U ^~ ' ' ' ** ' ' * ** ' '" ' -' "' s'*".?* '
~b
$,MDS'YC.@NTfd/3N:$"GTMDM#dds;,tWMdgmMgy$p S@dsjbl+Q.@r-y
&&p, : n: . : s,g:
- l ;! , , .
g qw;, ..%yn . . . .qq gssgw. g , g . 7- ~ u . , a p ,7 - J
.' ,W I IV. Rosearch and Developmen.t.. Program
=r 1 ,
,e
-i . ' While the design of/ this nuclear power plant is similar in many
! - E respects to other plahts now in opNration,;certain features of the design p . ; e. ~
6 -
<:., . require research and denlopment v6rk to establish complete adequacy of fy .
- 4 7 3 _' . ,,
% ~
' ,'[7^ $~'., intended design:and operational parameters. In recognition of this need,,
Pacific Gas and Ei.ectric Cogagr . [4 i conducting research and development ll.-[ m . n . . - > .
' '; [.. .
.. programs'as out m ad below M (a)~ Materology. ' A meteorological" facility is beirg installed at the
' site to provide necessary data for atmospheric diffusion studies, q
, ? Instruments will be mounted ai.three levelte on a 250 ft. tower and
.. 9. i U will measure temperature and wind speed and direction. All readings
] ,
~
,. . i,,s i
5 ?;.. /o r evill be digitized and. recorded on paper tape. Sp%9..
.am ., 4 a ...u .
l j7[., . .
' Q, (b)i. '; Oceanography." Th,e capacity of the ocean to diffuse the conden'ser <.
. . w, , . . ..
a - cooling wath and*4=4ze thi, effects of temperature and radio-1 .
.: activity on the marine biota is being investigated in a series of tests a
d conducted at the sita. These tests include use of drift polos and - s
. .i Ti i 1 uranine dye as well as measurements of temperature and salinity. ,
i i
; They will continue through at least one annuti cycle of oceanographic.
i1 l 71 'and meteorological conditions vn ,
. .- )
;y
..~t s.6;. e) ,
m a. .y (e) . Marine &iology Survey. An ecologie:1 survey i. s being condv, oted to
, , ,j e, , ,, <
' ~31. , . .
prepare check lists of the marine fauna and . flora of Bodega'53ad Land
.>..r, ,
s ;.
.2:1-4.cq ,
. Harbor.. ,. . .
{ H _
- x. %- 4
~
. /M} l 1, a- , -
, " ';(d) Radiological Survey. A preoperational monitoring survey of the site
,t . - s.. , , . . . .
.:;i.ig%,;;- -
l c:e.; ,and - its envirorn will be initiated two years before commencement;of tgl' - e : operation of the reactor. Thidetailsofthisprogram'havenotbeen
- ? _
, u .
. completed for Bodega Bay. However, it is anticipated that it will be 4
- similar to that conducted forlho Company's Humboldt Bay nuclear unit.
W, ..
. l b h3$.* .g 9 g NY' . g, g " 'hh6 Jiu, 4 $d$MEM.Nhpic.Iv54%$kWjOh.QMNbb .d;.m. .,,a y, e gge_ gy 3,, ; AL g , ;. .. h. ',
;y; . .
0]C,;,MiGc.9""idmWNM'I*Nenn@' t T*f'-W?.WdW996**'*"WFWaiEY.W:# . j g n f. - 4 7 1
,g kg }% .{. .-. , v h ,
j
~,.4
; -g4- ]
bT ', fj lj
./ th .
a (e) Pressure Suppression Tests As dncribed in the applicant's hazards 1
) im t< (
,' y' summary report, Appendix I, extensive tests of the pressure suppression ') .
'/" . 1' ., ~ , - concepthavebe'enconducted."1dditionaltestswillbeconductedat 7
gt y "A- the Company's Moss Landing Pow $r Plant to determine whether or not
- s Q
.'%,d"h baffles between vent pipes are required in the suppression pool. l Q
In addition to the research and development work being carried out by
".yf a M lhly' s8,
', the Pacific Gas and Electric Compa!Ur the General Electric Company is t
carrying out a number of research h aSd development programs of safety l ' '
'r 1
significance that will influence t14 design of the Bodega Bay plant. These ) l th 'j j , are: ny (a) Fuel Development Results frodi uel f vienent development tests and experie'n ce with fuel designs $w esiployed in existing reactors will 5 1 .. . M . i' ore the basis for the selectSton of the Bodega fuel. -- i N >
.' (b)x Instruaentstion .-
Development N-core startup range neutron detectors
,x t
/ are being developed as a posd!,$le substitute for the previously A
,o
! planned out-of-core detectorss e
.uc
<i t (c) Cont,ro_1 System Development kyprototype Rodega control rod drive is 3 e l currently being manufactured.idIt will be subjected to extensive
,: vc developmental testing before the final drive design is released for M. m :..
ny
& manufacture. Several devices which could reduce the likelihood or an
^ ,.
y a. t magnitude of a control rod dropout accident are being developed for !
- n. ,
', , possible use in the Bodega control system. ' '
)
'W is l
Q",! '(d) Nuclear Excursion Analysis Development Analytical models are being l 1 hr~ ' ist deveioped for the more accurate prediction of the physical consequences q$1 . l 14j "$
'c 1 of nuclear excursion. t q lo
.-_. wc
^l It is the opinion of the Staff that these research and development programs \
6 I r*
'*e are properly oriented. As provided fo{by paragraph 50.35, Title 10
. .., , . . , , .ove
.u. 4 .. . - ae,,,, m -
.. ,.s -
. s e .'E e$ $l 4.e -' ~~ 'Y ~~
' ?- &Sa.S " h y mbo A+ - -~~~~*O -
Y'
- c. 3u a ,
l- . . . . yMa . . .
.a .r.
qqpm.
- b.d._u. e w.@n< -r.. ,- - mpeWN.,p,Wy .. s s.w,w.m. . , w<*4 . e. - ..-. . , .~yWre~^fw
, -. .. 1.v e ..- -%iwa'
.m . ..g ;,ds**yc
. , -a, v., *6 .*,W.e .-., . e
. W M a h.*W ,* .
. .7.'.. t '-e**hi .m _
. .m.-.
m ..%. --. _g a.sm, mce
-. . u r_
' y Ori'.) ^.
. .,
- p. - , .
. .y.sj s ,39 "1
. , . ..." . ,.:. w >
< . w w,- ,
. - s.
+.w ,,...f..-, ;a,.m 7_ . ._,,A, , ,.. , . , <
,7 o. .
a
'3
. .(
~ - g ,~
, ;, , x:
o l' i
- 45
.i of the Oce'.e of Federal Regulattins the Staff should be informed of the
.I
._ results of these zesearch and development programs and will condidor these 4
' $; results in its evaluation of the propriety of, granting an operating licanse
- p. >
e . i E@
- m. , ~ ,
n
- ' to this facility.
, . , , . . , - . >.. me-
. f+ ", -
w
. , ~ : ;, . .n , i+.,c.: ,
w..,- . -
;Of , , . ~ ~ . p- T.y a, k* '. .'.. f?' N
,r
+'4 t q- ,
su l l 4 h*
- l s
/ y n
> r., J,. .,. , .
9.; , <. .. . 6 '
..a... ...
- n r e, ,,[l [ a- ,'d,.. #
,a
. *' e.
. n' . men ...c.c.v.
w 9- A wj; . . . . .c . 4'N - w- - y 4
. s *
.l .n u y I
i I 1 I i 1
. *k s g
..,. ., a i . ., ,
. .v ..e - . -, . c. , . . y.. .. . .. . 4
,'g s
's
'1 *
.y x .
s c , ,
.]
c
- tr..
. ,- , c . .
; .- . .>+ .
Q:'. k 3.,u r
- ~
q. e s * ' .! ...; ' ,
.ps:y .
. _%-. . ~ . , .. s;
, 9 ,,
; .y ,
m....e , Mmi.I. d s* 3 '* . . att
,. .., O fo i,*
= * '}' .
. -]
.a -+ =
0 e 4,. 5.. 4 4 ,#^ 4 4 i - . [, . L' y . 'a- * ' 8 '* ~ ' *' 3, hl . - - _ - _
. 9 K '
-g .
55 [. # ##
*$ k *
'k -* *'.a "
(( .
7 MdwG . mm..;cwei6:de- +'*M&%wm.%e. WiMT :'r MOyWN;%@t?'9^#e' :-s. kefW"' Q; ' r:y:t& %d ? e. - 3.m. ,6 .
~
' ~
.i,jp% j.,,7 - A . . ;
g '
; g - Vffh.ipY$.9~$%M gh.A)yE , , , , ,
. - ~ -
.;; ; ?c , ,
- %~
, ,, q .
h} V.
- Safety Analyses '
c!:
., :(
. he design features of the' plant have been described in the
, , previous sections, and in many esses the safeguards provided by se
,. 4 mf;4,., , -
~ 4 Particular design feature 6Y the operational limits imposed by
!g.f;. ,
7
, .. +
.Y n..:,%y safety considerations for a particular feature were discussed..
$p ~ ? 1 ,q'.i",IA. ,
..gu.R '.,;,,3
.@4,~. .. . r. . ,
(Q .+
" In ' general the criteria for plant design should include: (1) 8, 3 .., z.
.,, . -means to control radiation hasards (including discharge of radio-
, J activity) during normal operation; (2) design features to minimize
. i '-
the probability of having an accident; and (3) design features for i
; j mitigating the consequences of'an accident should'one occur.
!, The means for. controlling' radiation hazards during normal.
,e b, .
. +
5 (Operation vill be provided by/ suitable shielding and radiation
*p. .,r. . :s < ,
ilM l .
, ,?e- ..' , n w..m . wd.b,monitoringinthecaseofdirectradiationemittedfremthe
- y. v.s m. ,
r .- , , , . , . - y.. '" -
<" reactor and by proper monitoring of radioactive vastes which are NW]
m discharged from the plant site. For vastes discharged from the
',I
.. plant, the release rates shall be such that they do not result in a'j . ,. . - - ~
I personnel exposures in excess'of 10 CFR 20 limits. 1/ j! The adequacy of the design features that are incorporated to
-s .. .;
;O J.d
;w.fy m mitigate the consequence of an accident in the im14kaly event tf2at w .
*fw:, !,.y . 4
. r. .<#;:sM..r,f,,y p one should occur are evaluated'in the following section on the-i v.w(
'l'.t . 2~ 7 - ==vinnnn credible accident. 'Ibe consequence of this accident to
.j - m ;.
4
' ..q. , J,.
z . the health and safety of the pu.blic is presented +aking into
.w,, ,
. ss , . . .
My u.e , , o' , . , . ' ;<,W . . . ... . . .
,y consideration the safety features afforded by the containment
,' l [.npnd enfker emanyany
. $Q,f
.yT: 2 &. :+
3py +
~.1,v. y ,?,o ..Asyttimaand the environmental character ,
+ .
z of the site. ,. vm .. . ( p%.'i , ' * ra 44 , o[ l? i
" e 6 * ,
- m i
, 4 .
.,, u
,W i q ,~ . . , ,.. q.w. . . p. O ' - , -= a:, * . - .
.I .e ( ,; $44', J , 8
~
kj 2h ,' V < ? a .* h' . n * '. i . h '. p \ *#=.' ~* '
*~ , . S *.4 4 46; k# a '%d. I *"'" *
['* 'N 1 Ce by ?*
- M * ' N Nj *;* (;#
1 7 2 e 1."%= a me?'S # >e** 19. La * /'# ', *Mt * *Nr / * #
- 8*\ '
/ I N T,dt"I M E C M G T E N P P NE M T M TM TE fE* @'?'WiIN W s WI N 5 N M NI-PE D M E C -
} [ C. 7~[ < . _ , . ~L 0- '
elir
.. : ' A V * * .' mL ,
~.4, . n. ., .
,U r
+
1
- . -- To evaluate the design features that are incorporated into .
2:: -
@, the plant design to =4n4=4ze t'he probability of having an accident 4% .-
J'+ tf J. . .
. ,, a number of representative abEormal conditions, equipment mal-
. : . . .. '. Ja.6M,t/ ? functions and operator errors'inere postulated and evaluated by
.h,.y?jf ,
. . . .y .
, . f.{ ~
,the applicant. Those which veh presented in the Pra14=4"*7
.h. . . ~.
.f{c7
/ Hazards S m Report includeid:
j s. Chitnging pressure regulator handwheel setting I.. . 1, b. Continuous control rod withdrawal or insertion
.i*
;. c. Loss of electrical load i
- t d. Control rod drive malfunction S.,
N. is ,,,...
's e,:Ree.irculation pump failures
..c,. t p4p c^ -
1 - , w x.n .p-- . . ' . , ,Y . . 7 f., Main steam valve closures .
'M g. Failure of reactor safety valve to reseat
- h. tailun of reactor safety system t
j i. Fuel cladding failurei" 4 Loss of'feedvater I I
- k. Loss of condenser vacsum h*
. ,. u ..
- 1. Loss of awiliary power
.. a-. .
$.6 . ." - m. . Instrument air failure
2.9 n.
- Pressure regulator failure jj, , .o. Energency condenser tube failure dj.s
- p. Reactor system ruptures inside the dry well ,
esp., - pgj
, q. Failure to replenish cooling water in emergency condensei
- 4. , ,.7 i7 '
@ c'., i .
r/ Startup accident
# s. Fuel loading and handling accidents
+ t. Cold water accident. .
..r.
,i
..y- .i. -
M {;i.$hsh d;/5h,8;,*ddDU MI;S[Nji 'fd d. ,w e,' . l Wil b b fM kNe N4"f*-E.e'6 b "'s ' A d' -
. .Md#-E-a ,W%.m-~~.,*r'-sw ',~*/~ms "45 4MS# WINO NEMAN'r4N '**'S*'-
g - p 5
.... m_. .s
- a. ~ 3.,, -. .
4.p'. ' % .7,*,"% e d pQQ ,
. ~ *
. %f ., .
>w 48
?p yp
,% ~..
- Q
, $ t ,,.
r
.e. . . . In addition to those conditions listed above, three equipment
. u. .z .
2,:j'j(fj failures termed " Major Accidents" were evaluated by the applicant.
* $w'f f.t i A "'~ These accidents included:
C ., l
..s ,
.~ .
. Sir -
. 4 . .
- . .;,.w 9. 2 . . . . . .
h 6 ,1 w . . ,. p . , y. ,, g < ,.3} .. g a
- > a*f* . . /. - * * . . . .
n .. .e..; m w .e:=. . 1 Control rod drop accident 9.j-
+
r' 3~
?
u y m - _
.- A . ; b .- Main steam line rupture outside the dry well -
- i '*, .*% $ '
ify c. Reactor system rupture in the dry well. m.o% qV4 In some of the malfunctions and failures presented, the mts evaluation is not yet completed; however, the applicant has , stated that when the analysis is co=plete, the results vill be
.f.S. used as criteria in the detailed plant design (for example, to 74.p. .
- s Ea
..,.n size the pressure relief valves and to set the isolation valve
- p ; q n. .
; u . . - .
4 62.. u..
, - In our opinion the evaluations,that N*been completed at this time have
.,.. formed a satisfactory basis for deter'"ining 'the nature and consequences .
Dd :. ' ' a of the maximum credible accident and fe establishhgproper perspective
,n. - - of the necessity for emergency systems, except for further considerations
'If 1 which should be given to the control rod drop accident.
gag ~ , . _ apq ,,< .s , V :~ 4 - , ' - v ,. v ,F e w J .' J y } p ed , **
- t
- r #
. .. . 1 -
& .} .
^
;g.!,"g~ -
w
,A y,ag 4..
,y . 4.<j. 4 ,. -
.= 9 ;
i
~4 . -- , .
3 >
, , . W. *'
- r. -
g . - * * ,
$ ...i ;. *
.i 3 ,
y,,.e ,
.s - ---
1 i . 7. c(4 . ' ' '
.c . g , .' .
L'
}
l\w $ .
. . .> g e=
(
,q.Q we e-c f. [*s ,
,< : ..'s .
4.g' ' ' . . , t .g, , . , , t . r .n ;4 .,.C...
'bi'rl W3., $'ON t J, ".N.S.C.328 E 44,Ni'8.% %'A. * '* d'b t ' T.'"
At-*'.,.,. .
' , ,/A . ... 3 he~2MD. .
( *' ' . ,i '
.Y .f' t(.- ~
**'# 4 5 I - J' ^**#'"
' * - '-~
a.
. x- w. +. +1teswefne.e: Ne. .v. & ww k, awn .~er+r .Mzmwamene"wmemenw~'1*~5'? n.- a n g:g . M?MI L 6,s* m ..
u ... .x... . a . : . , e j?p; a. b y , ~ s.; . q_.h .=i.M - 'm
;yh el{.(M;}h . ;c. :
- q;;y . u g.,4 3 ), .,. '
"g .
%6 , ';
- V'
, 49 , .
. + ,
- j .6,.!
, . - a.
. c.
p_f, 91 kr thkr McWMt',
, m;j.p. ,.m g calculations by the applicant indicate that the most reactive ik 2w control blade could have a reactivity worth as high as .036.
'Aiditional calculations show that if this blade vere to drop
.h y
d ,
.,a.
, ., n. . .
, w .. .. . s .. 1
; .j..-
3 ; j'
" e free of the core a =4n4="= period of 3 milliseconds could result, f 5 and the' average' fuel temperature vould reach 55000F in the un-b ,5
?E .
_ controlled fuel zone. The causequences to the reactor vessel
.fiy d in the event.of this accident are not entirely clear. The 1
applicant has indicated they are developing analytical models '
~
- ' ~
for more accurate prediction of the consequences of such a w '/ nuclear excursion and that the forthcoming SPERT destructive
- 4;E . . . :w 4' 1< - ,
2. test vill be used to check the model that is developed.
- j,p'% 7 d dt 61* 3
...r..
i- n .1 x J ,- . v. , gy .,. ~ '" ' " ' In addition to the analytical work, a rod worth minim 4 zer ' ^~ m,.sy; - o f@u,. ~ computer and a rod dropout velocity limiter are being developed
.w.
for possible use in the Bodega Plant. The rod vorth cot.puter g.. Would continually monitor control rod patterns to reinforce procedural controls provided to insure that patterns causing Qz individual rods to assume undesirable high reactivit'y worth are g 'J..:*, g;. p.,,g not used. Conceptual designs for flow restricting devices that
~.t .c . . , , .. . .. - 4
[.jI would limit potential control rod dropout velocities to safe
.w:n. . . , ,
by values are also. being developed. In the absence-of experimental - u:; (s.9
.: a g
. verification of the applicants position that a rod dropout
.\,
g., gd7
.S et:i accident of this type vill not endanger the' reactor vessel, we kf[g s
believe that other design, features,'such as the rod worth F~ni! y$.w O minimizer computer or the rod dropout velo' c ity 1Mter, should W
- We believe 2:/ be incorporated into the plant design.
p, ...
; ,r s ' . ,
i y ,. u 7 .. :;+en , - , ~
.- N hd'*h.Y *
- Ur.N$ 'N[$b b h . k* *M ')"-'
4 . ..
> ,'. % -w_
.ic- .hs4mMpirs+N$$pj;yd4?fut@Oh=#rt.mWr-@mmnnhd4fMwv'md . ,MP%dar:tur
.E. IN. -E .r 4g9 g.
s i a
, i.H.
. f.
a.. .l o-Q 3 1 w .. ., .. < 1 1- ' f/ however that by u'se of these alternatives one can obtain adequate
' fr' . . assurance that the control rod drop accident would not have consequences
- 3. . ; .
YEII4 k- ' - as serious as the mav4m m credible accident discussed in the next section.
, oy - - - , . ~ . . ~ .y s .. , - ,.. >.
.a-,
$4bN * * ' ,.-------_..._
.Mt ,
. . m:.:d
, ..'e',=*y' . ~
~
dE
A.,
*i a
# 6 l
0 1
'n .
?~. -
s.
-h A *U ue
~
j' M ' g, 4
',W 5
.'y'e t g t '".$ *Ma.. 4 **f '
- r
,w 'e ,s
. i.
1
,.s 1
L f ,U.- 1
..E# q * +
- Y,*.,tk
. y s [
, . 1 . , ,
.,.~. .
..%t- ..
- r. t .-e ,A.
.Wh yQ' j '( *
, A&
a
.-y.-g., . ..
[ h% I
; +.r.
E A9h ej 4 W1 *
.?,O & $. , ,
*Mg s ;
v,.4. .M . .* , i ' Yh . e *
'g h D .
~ ' * ,
e.%:S. s ,
. .[.e .
a*
'.U, r .
3.s.g ;. 1 s'.iem .'. 4 Tf. %
' k %f
'?.,.,
, .f' 4> .' .. '... ? l9 O ;7' _k ' -b ,m. ' y,$ .Y i - - ^ *
* ~' ~ ~ O* V '
*' - b
4 * * , t .s . .
'"';j y-?d stPUNEE M N % N i$td%$8M5 .E3 k N M d5' $ iM N U M h M 5 M M M W!'M W C W C5 kg . n* -
st> -
- 4..
ft/s f $UI kn y ( f d U U M b r k g$t h y:g VI. Maximum Credible Accident Evaluation
.d For the purpose of evaluating the adequacy of the proposed '
,n. . ' .
" m), ,.
- $1 containment concept, the applicant has hypothesized a major 0;05 GM accident involving a substantial release of radioactive fission products i y], .
* '" from the reactor fuel,' and has estimated the consequences of this accident' 7,65 '
I W.< in terms of potential radiation exposure to the public, taking into consi- .
. M. !
, '. $. deration moderating effects on such exposures of the containment system
.and the environmental characteristics of the site. The maximum credible
.i accident chosen by the applicant results from an instantaneous complete
., rupture of one primary coolant line inside of the dry well af ter reactor
. y operation at rated power for an extended period of time when the fission
. s.
m.,n ry product inventory is at a maximum. The pipe rupture would release g$.@ ' ME'the pressure in the reactor system (assumed to be at 1250 psig), resulting in
"+ r. ,
- all of the reactor coolant system water flashing to steam. An inmediate
, ! buildup of steam pressure in the drywell to about 62 psig would ensue and 1
pressure would increase in the suppression chamber to about 35 psig. The < ,_j s j pressure in thr; suppression chamber would be reduced within a few minutes due to steam condensation in the water contained within the suppression pool.
<go,,
JM Other assumptions concerning the magnitude of the accident and the 1%b M1.1 effectiveness of the engineered safeguards systems for alleviating the q, ,
$./,,,'
4 severity of the consequences are as follows! ~~-% k.w;. .(1) The loss of coolant from the reactor would result in i
. k.
J f.g melting of one-half of the reactor fuel and damage I
.: p, !
;.jv to the cladding in the remainder of the fuel. (The -
..; 6. . ,
. g:Q
'qj applicant will provide an emergency core spray system q'.H 1
for preventing such significant damage to the core under T, 1 p. J these severe conditions, however, for the purpose of this I
,s. ;, ..,
,f*:
Ai! c :D _ }k;.7_f $ A.!%;,l $ nl f 55$ h i h Q.Q % W . ~
'%_' .s ~ .
~
~ Ns , . . . h 6. Q ~ *e L
- s" h ~L; ;G j. E + +-,R s M r s & :-h a b h ss. %_ s . kinchm-2&
~Br &. *M i $mw% . . x.~ . .. . . .m?-
n % w & &..# M..Y 3
%. f. - a .. , : ,'
.n *.
(e- - . . , 3g i
- :.u ' ,.
aQ
=;g. w . m .n m,w v ,- 1 - > -
, - .h
, ,' ;7 .2 . , ,
6 . '.d
/ analyses, it is assumed to be only 507. effective). !
.,.$s . (2) The fission product release from the fuel was assumed
..g., , .
QM X- *i as follows:- -
- n. >
; - '. FISSION PRODUCT RELEASE FROM REACTOR CORE (PercentO
$1 w
~.- From Damaged .From Melted Total Release
, e,
..~% _. . % %
..~.,,.g....
,, ,. ., , . . - . . ~ 1.; ..." Fuel Cladding Portion From Fuel
,' ,e .
eNoble Cases 20 100 60
-{.y .g '. Halogens 20 100 60
,. e Q Other Solids 0 1 0.5 a 1 i, .
y] '(3) One-half of the halogens would be removed in the dry well and suppression pool by siste out and scrubbing action of the water. The remainder of the halogens and all of the noble W e. .
;,, y,y .
.a gases would be "available" for leakage to the reactor building-w.<
-n;;t.
s M+ , r, s .g% a. .:
- 'at a. rate of 0.5 percent per day at design pressure. As the
%M . m.;, . . ,, .r N' ' ' ~
I pressure in the dry well and suppression chamber decrease, there would be a corresponding decrease in leak rate. As noted previously, the applicant will be required to demonstrate i
} the assumed integrity of the dry well and suppression chamber
.1 j prior to reactor operation by suitable leak rate tests.
.ci4 gj; (4) The reactor building air which is maintained at a slight Tr.q ~
3NG y ,
- s. negative pressure (k" of water) would be exhausted to the -
.p' 3 .
'J!$.h a atmosphere through filters for removal of particulate -
,a:p#
. -~.
g:
. and halogens, at a volume flow rate equivalent to 100 percent 16 .
.of the reactor building air volume in 24 hours. The particulate
, pz y% w U "j v
nn, j
'and iodine removal' filters are assumed to be 95 percent effective.
4p#;
~
- As has been stated, provisions will be made in the design for
,% verification of the effectiveness of these filters on a periodic basis. .
I l , f ,
; * ~
;* , . " . s
; Jh t5 ? *h ,,'* h N,f [*.d'h * *- *[ > I*' , U" I N *' # -
? ['.'*, b e 't *-
;.~v'
- e. 5A e. w .y . < /, *t ..* '- .J4 'i, ,.gr. r. D a
- t. .4 % , .e e! . t ~ .,,e- e .* ..--e.&
,#,..,...m.
.c . -
- TW TW?rgywn2d!u*kDh:~~5%~Egrogrgw,rb& %f&tnx:gryg.~s939.gnm&:rit!.nezW:'?'p~: i.
- a,ma na.;06 l&, a ;;i: '
. x.; . ..
- ' $2-4 w;c: D -4. .d.w:: ':~.x ;&.
, , v.- -
o, 5 , !* 1 l 4 'l
- L ]J - , 1 gf:J The applicant's evaluation of the constiquences of this accident !
. ;un ,
' > :1
*l j assumed release to the atmosphere through a stack 300 feet high, 33(M ,
$m$$.$ although st the present time no specifications have been proposed' for the stack design.
M
- e. m. 9
~ Assuming that the wind direction and velocity were constant during i; - -
.w@w s,.y.
e
~
.,jI.C[.. the course of the accident, the applicant calculated exposures as
'"s-,.,- -
follows8
- 1. For bood meteorological diffusion (lapse) conditions and q .
..l a wind speed of 10 miles per hour the maximum exposure rate at
' t' ground level would occur at a distance of approximately 0.6 miles
~ 1: . ~ - '
%w/4 Jfrom the stack. -
. . . .c..
ns? .n.m. , . .
. The maximum potential dose rate to the thyroid was Ig m; < .
s
.:, a.
, l.
~m 2 ^
esiculated by the applicant to be approximately 8 millirems per hours (0.008 rems per hour) and the total potential dose for the duration of the release 4 4 .
.'h, #
t
' . I'W ss g +
n.<-, , ,. e , , O '#P. v N
. A "g
' ..w y a n we.: g > .. , . + . . ... .. .. :s
' r je*f e* *
.. r vae.~. r.
,i
. i
;4+ .,
. '~~% ,
,f Y ~? ,, s i,- ;d ,g ,
vy *
* $%fQ , ' -
1
~
'C2 Sr.. 3.;.
gpi,l . o A i w e.
.~
0..-..,
~ r e
'g,
,e
.yw .,
.,< s. , - .,
'(Yl. i
, , *,- *- ' s-
- e ** ' g,
'Q e
- 7 ' 4 ,, . .; : .,,' ~
3a.
~
.., *s
. e.c
..$9'il ... .
s
._.o
*9 Y g.G: ,
y+e *'.
.. 4 i.,
,l s' e,br . e. " y *, *= . L '.':
f -lt $*i$ .;f f$ * .N f l9,5, $jijk(.'$';aYk.h &'),h i%f.N l'.,Y _3 *'k*,I.s " O ' 1, .',.l.,',' ,% j' .-;: . y. ,:- ',
o .' f, , y. w ,a .. ., -,3,......,,.....-
*~l-r.tw%. 5. .-r rv&Wt .shu cmx.cim?'& .aw :v & ~ve h m " M M *****3 .%
. A.~-s,.&vy;W.
~
w _
.~ Z .- ' ' .'Q& - [l.y ,4*r
\ {:& .m ar;
< s an: -
35 - 1
' i . e' to the atmosphere was estimated to be approxi-w
? mately 1.5 rems.
41,,
- fk!! b. Se ==v4m"= potential whole body dose rate due to it ,
f .
- g. g
. 4 ,
. . ~ .,,,...m -
noble gases was calculated to be approximately 2 f;[Y
.: Y@p: ' milltrem per hour with a total potential dose for the f'hN ,
duration of the accident approximately 0.024 rem. -
." Sd 2.
I For moderate inversion conditions with a vind speed of
] 5 miles per hour the applicant estimated the maximum
- j; exposure rates at ground level vould occur approx 1=ately i
3 miles from the site. Under these conditions the mav4-m I
'i E . . ',+' ' ' potential dose rate to the thyroid and total dose for the j w.. . ~.. ' .j;$ . L.,i :,,.; . , 7 duration of release was estimated t'o be less than 40 l
millirema per hour (0.04 rems per hour) and 7 re=s,
.l respectively. Se whole body potential dose rate and
-1 integrated dose vere estimated to be less than 10 If>
miliivems per hour (0.01 re=s per hour) and one rem,
't l respectively.
4
;y As previously stated in the section of this report describing a'*h
'i7
. ~ ,
the meteorology of the site the staff believes that vind speeds t-'y during inversion conditions may be somewhat lover than assumed by
. 0, i ~ the applicant in this evaluation. In this regard, the staff has i,j, made calculations based upon'the above assumptions which take l 1
cd . i
'.'9.qI . . into account the possibility for the accident to occur at a vind i T T M) ~
* /, f'f speed of one meter per second (2.2 miles per hour) under either
.+.
= !
l l l f l 4,b., '
]
h.h
- L. . ., . . .Slh Y '?SO : $5W .
- % % U: % <
a E
, + ,. ,
Q,fr~;"W &%6sm&$ OWNTAO6% % M3%WD3tN** M C"W W W5"B^Y YI'.*3j ; .p. . , . 2 m> w .d - ,' c v. 9
. , .,n.
-.-l w . s., . .... , .
p iM ~.;N;l,- d.S N5 5 5 % & w & ?r < - s. = ~ '~ .
- w - ,
,N, 5h . i
. ,4 , l
,;v l
' ' o., ..
lapse or inversion conditions. Using an effective stack vy , I N,,ag . height of 300 feet and the one meter per second vind speed '
.y . .. <
..s / '
-[,,j ,
the staff estimates that the =av4== potential whole body and g-;,4 . ..-j .. .E 2 -1 , ' gas..a r e .p- App 'ff2. thyr,oid dosages for the duration of the accident are 0.94 rem'
. :.,n -
g? . ..p> - -- and 53 rem, respectively.
- k. y
~
he staff has considered both the applicant'scassumptions concerning t,he postulated ==v4=m credible accident and the
.i n
general concept of the safety features proposed for mitigating the consequences of such an accident. As indicated in the above analysin, the amount of fission product released from s , . . My
,*. nE,i.
r . , . . -
*:0 the fuel depends on the extent of core damage, which in ,
.h. h . .$d:Nh.. turn, depenks on the effectiveness of the emergency core spray.
- ;..~ . '
' he staff believes that a suitable core spray design which would provide for an adequate supply of emergency cooling water would l'
subs %.tially reduce the extent of damage to the fuel, even to the point of preventing any melting. On the other hand, if i the emergency core spray failed to function at all the inventory l l~.
. O. 4 . . . .of fission products released from the core vould be increased by ..i}p .. . - . :c -
q -e -~~ apprmimately a factor of two. In this case the dosage values .
.m -
jy vould be increased to approximately 2 rem whole body and to 100
.n, rem to the thyroid. - - .
t;g - .l.Dh;. *' * ' ' From the abovet. analyses the staff has concluded that the
%'n c.
h ., .0- :
.j.g
,7 y . . tengineered safety features proposed'for this facility should be ss.
373) O. + ' capable of limiting the degree of' harm that could result if an accident
~ .y '
?.11 such as postulated should occur. herefore, since it is believed
- 1 1
i Qf -'yy.-o,+ , , ,
,*' S 'n*
*4
& . C&.; c,. f.X:g &fQ#Q&2;. l$:ih' , ., ' *.40 ~:lQ' b , ; W. . '-N : . ,, . L. - a ' . > - :P -. h I
. , + o
- c. ,ts.. ... . . . . . . . . . . . . . . ,. ~ n.
vj a ( 4a Les* Mi-*p a,h_. ""f *"21* d**8* b er t?*;ed -Wh_,.m- ...Mr.-Ng'aurytt erqW de........_p..,yg,1,8.Mt9% Os[.y
, .iby
, y.*f,ggggy gp. p[g 3
- . . . . .. . . . , . . ~ .. .- .. . .. - .. .
6* 4*f,,4L?=**yQd5 y_ . **y[Q {p ,w.47 *
...v W l. o rr 2.
.g F /,}:' .
*f' ,
- -,.r- .,4 . . - . . . .
a.; 55 - I t ,, r that the occurrence of such an accident is highly unlikely, the
, 2 ;' . - -
i,,.t
,s
'O '? '
staff concludes that operation of the proposed reactor at the
~, .a .- .
Wai .g . -
/ Bodega site would not represent an undue risk to the public.
M
%.4.j,
.. .gi ~.. I l .,
A n;4 . ., . , , , . . . - . - .; . . . > . , t q q ,' < o g n 's. Y er -
.mt M. w.[
.MT T@i A%
kt
%*p;p ,qy p
g- i 4
- '1
' h, 1
( 0 a
-1
!U* s _y -
%:;fyl1 ihas . ".~s'.* '
.J.
4.
*
- r 4. "
- p. .e g- '
;Q $ . '
*{ h.i.. Q ,, _. .
h.b"* ,.K; . ' +>i. S ' #., -
.* e 4 0.P . , " 6 4 ' + + + ,$g 5 ...". -
- '!y" , ' 4. + ,
s c< L, ,
.s>
R
.l
,1
?s
.a(
#, % "1
/ '
.. ; s
.v '
e s* O i n.A. .
'"&tp 2 +.. ca .. c + 4 ,. ,,
q.- ,.
~
,*?, ,. J't' i i 'y- .
,. .? ' e d
.: .1 ,
.fa. s >
3%
. ;r;.-( .
ew%Ady : . .. w- -- , rir (
. +; ; , .
i*
, s,s i
** t- 3 e 6 9", f a
;0' )s?!" , a
?,$,?J
_ !?.i .r
.g , '
,.pl ,
- r ,
e q e -* 4-* s... *a
% . s 4.' ,
1#p." 9-
.e*p 6
*{ 4 7 A "y.ap,, -
4'] g ' , l g .* Jg.. % g,g g,, , .j k q,_} , " ,,,,4 ..,' et.g, , W ,,7- 2,,e 4 , %.,2 f Jt * '* p d.Q > .
c, s. . .
. ,'s ~?MDL*MiK"T@ SM'MVWUDZ9DX$M5"llC5179M 3'TYM%M&**":!'*"WYK'"Ti.
; /;.:
.p ( , > > e e :v f3.. "
l $y,,Nj.'$.~ bhl? & ' ; : F: ', kh, -
. g, .
't
, ! VII. Technical Qualifications of Applicant
;j i he technical qualifications of PG&E are described in the q-
' application for a construction permit. PG&E has constructed and
- / . a ;b i..du is nov' operating a boiling vater nuclear power plant at Humboldt q.;r;g
.i t wm o.,q: .' . :..sr.' Bay near Eureka, California., Their principal contractor for the M--
7;,qq.,, -
,, o, .
W@q / Bodega construction, the General Electric Company; designed and a:,% n , furnished the major components of the Humboldt nuclear steam supply t
.M'g
, system, including the reactor with its controls and instruments-tion. GE has also designed and furnished sim h O21pment for r .. .- -
several other boiling water reactors in this country and abroad. 4
. e
~
. f m.
p'
, Is
***q>,*h,. .O q" *@';'t n .
. .. . ..i G . v.g, > J, p., s ,
; i
. p.
.j+E&
-l ,, ..' t-i .y - N
- qm .
a"t., ,a,
. , .- , , m. .
; rs._.d
..a" ' r.
, s
,se;,,e'
,h'-
,Z 4
Y
- lA J r ,
d am 3.ft]
- a. . . > 4 . ,.
, n:W:s .
, ,. %. .- , . - , , . . , . . ... : - )
-JM:
s.>
.Q '
' ' ' ~
m . ,. i
- n. s'
%e.ti k [. '5g sy @ . 6 3 s #
4 .
..+ -g'.- , , ,
. g;
,3; :-
r *
@j Q
, '?
n ;. ~ . ., _
-y % ,'.; -e :;.: ,
s i 4,:t . .n t s
'5
'c. t ,'
*: 1I ,
- t.
? .- l a
. s s
'.Q
~
, , u. ,
j
. g l:rY \Y' N*,ihjoi '
l ' ', ' ,0f * , ' ' " l f ** {. ' ' * * ,* ' ' < n .
' ^}}