ML20148M747
| ML20148M747 | |
| Person / Time | |
|---|---|
| Issue date: | 06/29/1977 |
| From: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| To: | |
| Shared Package | |
| ML20148M724 | List: |
| References | |
| NUDOCS 8012240023 | |
| Download: ML20148M747 (66) | |
Text
. _ . . _.. .. -. - . -.
m- - ,-. . . - . . .. ., . -.
l g ,
'g !;UCLCMt REGUf1 DORY COT!ISSIO'1 .
GENEf'AL STATEMENT OF POLICY ON '
I STAND.VtDIZATION OF NUCLFAR IV.ER PLMTS .
e p ;
The initial policy statercnt on standardization of nuclear power ;
~
platits va . issuea by the Atomic Energy Commission (AEC) in April 1972.
In March hs73, the AEC announced the regulatory staff's' readiness to a icplemnt the standardi:ation policy utilizing three distinct concepts; u=i. :
nam 21y, the manufacturing license concept, the duplicate plant concept,
~
' and the reference system concept. In August 1974, the AEC announced -
that the concept of reolicmtien would be acceptable as a transitional ,
step toward standardization. The AEC was abolished and its regulatory responsibilities assigned to the newly formed Nuclear Regulatory Cornission ~
(t510) on January 19, 1975. Currently, available guidance on standardi-zation is contained in IMH-1341, "Programnatic Information for the .
Licensing of Standardized Nuclear Plants," dated August 1974, and ,
l supplementary docum:nts, and in published speeches given by AEC and ,
t1RC Co.missioners and senior management representatives. .
The record shows that the standardization program has progressed ,
.in a n.eaningful way. Since the standardization policy was announced: -
9
- 1. Seventeen applications for preliminary design approvals under the reference s9. stem concept have been received. Ten preliminary dbsign u
m approvals for reference system designs have been insued to date and #=
decisions are expected to be reached on three others this year and ..
4 m
{.*
h e
s.....
80122400n . .
. .m, H... i s
j hi b:-
two others in 1978. The review of the remaining two applications ~
rg has been deferred or terminate,' at the request of the applicants. .
- 2. Ten construction permit applications for a total of 25 units ,
l l
referencing five of the reference system designs have been !
l received. Construction permits for nine of the units have been issued. Decisions for 12 others are expected to be reached this year and the renuining four in 1978.
- 3. One application for a manufacturing license for eight floating nuclear plants has been received. A decision on issuance of the -
manufacturing license is expected later this year or early next F year.
- 4. Eight applications for construction permits, for a total of 15 units, have been received under the duplicate plant concept.
Construction permits for seven of the units have been issued and the decisions on the remaining eight units are expected later this .
year. ,
. 5. Three applications for construction permits, for a total of six .
units, have been received under the replication concept. Decisions-on construction permits for four of the units are expected to be reached this year and for the remaining two units in 1978.
The Nucicar. Regulatory Commission continues to believe that tno advantages of standardization are significant enough to warrant its i
continuation and' extension. An icportant adva,ntage is the enhancement E
e-9[
L
% m/
. . n ,.-
a ; ::.
. .-(u:
gj w ,
e.y
- v:- .
r k: '
n f.5' w
i!!
(d: . of public health and safety due to the concentration of staff and industry [.; ,
.a=*
. efforts on t he in-depth review of standard designs. - As a companion result, e ES . . . . . ..
there is a reduction in the time and resources needed for the licensing 7=
review of a utility power reactor application which is based on a standard ....
,- -E i design, with the extent'of the reduction dependent upon the degree ro which the?
plant ~is standardized. In addition, construction benefits can be realized ,
through earlier availability of final design docu.:ents and through construction.':
experience. We firmly believe that stan'dardization of the design of nuclear-power plants continues to be in the interest of public health and safety, '
and of effective and efficient regulation, and we reaffirm our strong-support for its continued and expanded use within the Cornission's regulatory ' ;
activities. However, the full benefits of standardization will only be
' realized if both government and industry mane.gement are firm in their ;
cem.itnant to limit changes to an approved standard design to those clearly ne&d for public health and safety reason's. '
[i:
In a related matter, the Commission has adopted and published
- I effective rules establishing procedures for the early review of site :
suitability issues associated with sites that are under consideration ! '
for location of nuclear power plants. This review could be conducted '
prior' to and separate from the detailed review of the design features ;
for the facility. We believe the early site review process could pon-
{,
tribute sign'ifichnt:1y to cutting down the time needed to plan and cone.truct . . .
i a nuclear inter plant when combined with the use of standardized plants.
l 7
l_:4.'l- \
I
,. . . . . . I".Z.:
' ' d
, ' p=
m
'The Co: mission ~ staff hos completed a preliminary assessment of 5 )
E: 3 the standardization program-M to determine what further definition and a support of the program is needed on the basis of the accumulated exper-
--icnce to dote., In addition, the staff is planning to conduct a more detailed study for presentation to the Comnission in the near future. ,
.The purpose of this detailed study is to examine and reco:nmend to th?
Com.nission various aci:tinistrative steps, including possible changes in t!R" regulations, for encouraging continued and expanded industry support ,
for and participation in the standardization program for nuclear power plants. The staff will consider and evaluate public comr.ents and suggestions
~ '
in the development of this more detailed study. The Com.nission has
. previously recom. mended and is also now considering possible legislative
. ch inges which would encourage and allow fuller benefit to be realized from the concept of pre-approved sites and standardized facility designs.
Based on its preliminary assessment of the standardization program, :.;is
- the staff has concludad: *-
- 1. Tne reference system concept of standardization is the most widely used of the concepts. The present guidance is directed mainly to the proliainary design approval phase and has been shown to be effective. !!owever, further definition of the concept is needed with resp:ct to the final design approval phase. Two alternativo final dcaign approvals' for the reference siystem concept are being contemplated. ;
=
5
. T/ copies at the report : .ay ba obeained' from the oircetor, office of
~
. tJuclear Reactor ib.julation, U.S. Iluclear Regulatory Co:r.nission, Ha:hington, D. C. 20555 ,
4 . h.
- 3 u
_5- -
p:::
A. A final design approval (Alternate 1), designated FDA/1,
-(g tis f which would te: -
. 5 i (1) ' Based _ on the preliminary design on which the preliminary -
[ ..... ,
design approval (PDA)' was based except for those necessary ,
. ' changes incident to converting a preliminafy design 5: .
to a final design.
(2) ' Subject- to the Regulatory Guides in effcet 'as of the time ;
the staCf positions wera issued in connsction with the rcview of-the PUA. However,'this cutoff date will not apply in the_ case of new significant safety issues.
(3) . Acceptable for referencing by operating license applicants ,
r who have previously referenced the PDA on t;hich the FDA/l '
is based, and remain in effect until_those referencing applications have resulted in the granting of operating I t
licenses or have been' disqualified for good cause as - I reference applications. An PDA/l may not be referenced -
{
s :
by construction permit applicants after the PDA on which it is based has expired. i
. i B. A final design approval (Alternate 2), designated FDA/2, i r
{
3 which would be: .
1 (1) Based on the preliminary design on which the PDA was i
baced, except that the applicant may maw a limited
. . !s nuthor of changes which it consi6ers to be desirable beyond those incident to converting a prelirainary design )
b:~. I to a final design. .
a p
c_ .
i.. .
ls y
$7Tll i.:..
7.?;;: ;
Y: k
. .. - me i
. - ~ <
G. ~,
lE i P
(2) Subject.to-all Regulatory Guides in effect at the' time . "
1
.} , the FDV2 application is accepted for doci;eting. "
9 L....
-(3) Acceptable for referencirig by applicants for co>: tined
- f=
- s construction permits and final design approvals for ! j purposes.of issuance of operating licenses2 / from the
-i time of doci;cting until five years af ter -issuance of .
I the FDA/2. !
(4)- Acceptable for referencing by' applicants for operating . t licenses t/no have previously referenced the PDA on which ' "
it is based, and have conformed their designs to the ;
= ::..
design for which the FDA/2 has been issued. -
It is the staff's view that the FDA/l can be a useful m2chanism to -
f permit'a single review at the OL stage for those facility applications '
. that referenced the PDA on which the FDA/l was based and thus serve to.
(
roduce'the duplication of licensing efforts. The staff believes that more ,
I significant 'bancfits can be derived from the FDA/2 in that it will per-mit maximum utilization of FDAs in both CP and OL applications and advance toward the goal of a single review by the staff of a facility application. ~
- 2. The exp2rience with the duplicato plant concept of standardization has been favorable and no changes in the definition or use of this 'E con:cpt appear ' to be . neer.I2d.
3/ Un;ter 10 C N Sectiona 2.303(c), 50.35 Note, and 50.52, the Co. nission I may issue a co:nbined construction parmit and final design approval for purpon;s 00 issuance of an operatira licon:n. IcJi :1ation to
. spacifically authorize issuance of combinnd construction p.2rmits and oparating licenses has boon proposed' by .the Cc;.".nission in the ~
ilIth Congress. .
7.d r.", .
~ '
~. 's-
......e'e,
- "O "#
.T............ ...~a-.- "
-- -- --~-
- . 'g--'e
l k ;
1
. hI
- 3. The experience with. the manufacturing license concept of standardi- c;.
- ':2;:' I zation has been acceptable and no changes in the definition or use if ,
of this concept appear to be needed at this time. ..;
t=
- 4. The replication concept was developed' to serve during .. u onti-tion phase of standardization and can continue to play a useful role in that regard. The concept has been utilized but not to _
the extent expected and its need appears.to be. diminishing. ,
No changes in the definition or use of the concept appear to be t
needed at this tina; however, it is expected that this concept will eventually be discontinued, and the staff plans to evaluate this concept further to determine when this should be accomplished.
The Commission would appreciate receiving con =ents and suggestions by
- on (1) the proposed changes and additional definition of the Cor, mission's standardization program developed by the staff and discussed herein, (2) other matters that might be considered and impler.unted in oraar to provide further naaded definition to the F C.
. c Commission's standardization program, and (3) other stops that the ; r
[7 Co: mission might undertal:e to further encouraga standardization.
Coatents and suggestions should be sent to the Director, Office of v
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Cont.insion, [
Washington, D. C. 20555, in order that they miy be considered and ;:m:
90 days atter publication of this notice in the Federal Kocister.
4 9
- M%6
l
~8' ,:.
l
(
iE evaluated in the staff's ~ detailed study of the standardization program for nuclear po,ter plants. Copics of coments roccived by the Comission
- N,*.
E may b2 cr.a. mined at the Coraission's Public Documnt Room,1717 !! Street, N. W., Washingt.on, D. C.
- Dated at Washington, D. C. this 29th day of June,1977.
EOR THE NUCLEAR REGUL!\ TORY COMMISSIO.'i l J. '0ifi k -
. Secretary of the form:ission ,
9 I
S T .--- [.:.:.
=:: .:r
.h
3..
s::
- ii 5 2
USE OF STANDARDIZATION - SU:IIORY' TOTAL USE OF STANDARDIZATIO.1 - FACILITIES (UtilTS)
APPLICATIONS IM UFACTURil;G REFERt;;CE LPERIOD FACILITIES (UNITS)., DUPLICATION LICENSE REPLICATION CESIG;l- TOTAL 3/73 - 12/73 - 16 _(35) 2 (4) 1 (8) -3l(12)-
1974 19 (42) 6 (11) 1 (2) 7 (19)- 12 (26)*
1975 6 (9) 1 (2) 2 (4) 3. (6) _
1976 '2 (4) . 'l-(2) 1 (2) 2 (4) .
~
1/77 - 5/77 -2 (4 ) 1 (2)'- 1 (2)
TOTALS 45 (95) 8 (15) 1 (8) 3 (6) 11 (27) 21-(50)*
- The total is less than the sum of the parts'because two applications for a total of six units used both the reference design and the duplication concept .
O
~
_ e*.
-n
- n. ... - - ,
- ti: :!w , cijp an ai! en
l F
pi.
EARLY SITE REVIEW b
F LI k p
On June 6,-1977, the NRC amended its regulations to provide procedures 5 which will encourage and simplify early consideration of site suitability issues associated with the construction of nuclear -
power plants. l pg l Early decisions on siting of power plants are'a key element in reducing the time required to bring needed. facilities into 1 operation, and in assuring more effective public participation in the licensing process. =
Under the new regulations, applicants for pemits to build nuclear '
plants my request an early review, public hearing, and a partial decision on site suitability issues as much as five years in advance of submitting the remainder of their application. ,
l The regulations.also establish procedures for State Govemments interested in advance planning for plants to request NRC i review of site suitability matters. l Assuming the partial decision on site suitability is favorabic, and unless the Commission, Licensing Board, or Appeal Board determines that there is significant new info mation affecting
- the decision and that the hearing must be reopened -- the partial decision will. remain in effect for five years. -
If the applicant has submitted a complete applicati.cn, the partial decision will be effective Vatil the construction ,
pemit proceeding has been concluded. .
If good cause is shown, the Czmission may extend the five-year effectiveness period for a reasonable period of time, not to .
exceed one year.
M The amended regulations take into account the fact that, jis within the last year or more, a number of utilities have fou".d it necessary for financial reasons to cancel or postpone b plans to build nuclear power plants. The new procedures will b q a llow such utilities to request a site suitability review on -E '
i a postponed plant, without commiting them.selves to construction.
The staff's site suitability report prepared under the new regulations does not constitute an IGC connitment to issue ,
a construction pemit or operating license, but it may be referenced in future construction pemit applications.
In order to provide additional guidance on early site review, '
the NRC staff has prepared a report entitled "Early Site ;
Reviews for Nuclear Power Facilities" (Nt;* REG-0180), h g
5 Is)
=
STATUS OF NUCLEAR POWER PLANTS - JUNE 30,1977
~
Number Rated' Capacity i Of Units (MWe) -
- 65 L! C E N S E D TO O P ER ATE .................. ........................... ........ 47,000
- T/ CO NSTRU CTIO N PERMIT G RANTED ................................ 82,000 25 Under Operating License Revie'w.................................... 25,000 -
'52 Operating License Not Yet Applied For .......................... 57,000 60 UNDER CONSTRUCTION PERMIT REVIEW ...................... G7,000
- 13 Site Work Authorized, Safety Review in Process............13.000 47 Other U nits U nder C P Review........................................ 54.000 .
9 O R D 5 R E D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1,00 0 j
19 P U B LIC LY AN N O U N C ED ..................................................... 23,000 l 23 0 T OTA L . . . . . . . . . . . . . . . . . . .'. . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 230,0 0 0 l
l 'To date there have been 364 reactor years of operation. Not included are two operable ERDA-owned reactors with a j
combir3cd capacity of 940 MWe. ,
- ** Total of units authorized construction (Construction Permit Granted plus Site Work Authorized)
- 90 units. 95.000 MWe.
So.srce: MtPC NP1-77 '
i i
i l.
J l ****"**.**'8 . , +. .. . . ...
.i! !!!i f' -!!!
- = gr.
I e l o 5 t IE h-(-l:;;
c, 4 4 .41 { g I
. . .. g
\en
. , . . . Y, .
sa e r -
t
' o te
<< o as se
+
,T
- e. p << = 4;
- .< *x . . . . . . .
m-i
, << a p ]
b 4 4
M w
o.<.% ,
1 *hh 4 I 5 3
So
- l $ <
t ". '
~ ..
D M 4
.-
- 4 a j; . < ,
g 1
=i r
V . ; "! ! <
- % 44 .ce . .< t ,
<n .= , <
w p M
- n 4 ,
~ 7. I 4 e
' =~ ?:
g k O H
a e ___s *
,a N
- i[ <
, =
.j" J,1
^ ^ "
..- : a. L1 c ;;~ w
- O < cc !/ f* g 4 w E'd W h .
f- O O
CO CO ,
O
, l}
h, s .
~
p
- I m
~
O i
a . e
/
f j -
55 j
O La fm o
w n J m. . .
& L O Y U ] i a
< b
- [ 4 . a ,
. < . ., s NJ * !s . i I .i. .
</.
- j ; a I I i 1 r'
, << l ;g } .i a 4 . 4
< e .#
E h ...
!:i Ts '= ::
-i E:.
,r
~
~ 4
- e. j
)
,r s._.. ,
w.;;..
l
. g.
. w l 1
l r.pr
(= v, 0?ERATl!IG REACTORS BY STATE h= '
s.
p y %;w ,
, caa$oP' = >w Er '
9
- ' s j
4
,, , , w*
, .' % y s << ,
s .-
% .= ,j . l?
, s s
y<, , , s< S )'es
, , * 's
, , g i
. , , %ij J.Q - M - I ' '
',' . j a.
, ?r >
r , .' & ,
. , &y.yi:.)
, ',, ';p.i ' ' <, Jgs , ::::.. .
j^" jw.
- qyy
. , ,, l', y '
~ '
-_, , ,, , i
.,ts,,_ m: , - - s
, ; >; ; y.t. -
...;-. s ,
2 m.
y , ,
- '**se.,, - ' %s g
'e...
, , , s, M ~,
~
e
.. s
,e- . .
.s .
ay,. ., m.s.s .
S* e ...-
- s. .
m,
. y . ." ...
. -e - - ..~ -
., , .tk. .p; - .
s
., * '==*a* "*.- ,
,,,, q 111 e ,v 4 .
%. -~~ 1 l
...<, i
"" *.'="
,,,,,e * , .. +e ' ' , , .M:.
.ss s 0 ,
u .. . s -
. 1 g:, --
n, '
.. . .. .a v. . ,
' .p._ _.
= = - .-
aA _
m 3...
s 1
/ .- ++*** . p- .: ..,, ,,. -
i
-- / . p:.c_. -
L f1 '
., ".".*"J.
, s
.s s s '
s I
hh l
., OS sY s , ,
y s
,s, .,
gg , CECPCIA P.!fC5ffCAN I -
e, a pe,,y t. 2 6 3 Itcss Mwe ***) **"'"""
g
}
IL L f *JC'1 "d ',$'m*'"Q"*8 pas n D****"' ' im WH ung , ......
A' tans.ee i 1350 Mm) o,e.co i r2co Mwe M ..5
' " ' i Osten 1 b 3 094 vuve weni CALF 80P.*M .
t g't g 3 Abt.sstie t945 mwd haw M i & 2 p pe M WWAq lQM, @L
&.a %e.kJt 34u n.ae t. maSMwe16.1 MWet Onof,n.ay th Somat 2 & 3 iW5 Mwe e=al e 114JO Mee) 'a
- CQ,g eg Owene A,noki(533Mne) gg hhY((g ,
b'*" WiTM* m Ii IW isis AN Mwn CfM f AROLIN A ;
5 L St. Vee a 1330 MWe)
Ma.ae venhee 1790 Mweg qQ3*g,$I,' f .Q Ora,er Ce.eti (660 wwW k'A9Yt.A Q #ao 1 0he 3( t h. , '
Coaaectiewe Yashw (575 MWel swis ae Pu.i e isu Mao cCalvett cm.2isas Cf HiI (C.s3 Mwe Mwe M41 stone M.at 2 (323 Mwel p.,,,,,,ia igr, M., ' ""**' Y*d" 13 ' d ""*$ f. .
- EM ggacomsg r73 C noa 1490 wwd losJian Posat 1 PCS Mwd yto0g 4' si. t i isio Mwei las n e Posit 21973 MWel $
P'* i ess5 Mwo g,,,,,Y I & 218?2 MWe mn)
Twh* Pasat J / 4 (69J Mwe e.ch) *""'N'** I NO "M ta+.a Nas's is73 met c,,se n e, s ists Mwn '
ta*J't cAnous4 + UDD t m,,,,,,,,,,,,a 641,1 CENSED l'O OPtAAt E 8'we*=ak 1 & 2 4871 Mwe seent ,%'0 5 W
'*'*8 h...?.
Y: 7 i F."
6 10 '
i q $' .
. ~ . * ,
, , , , , , . _ . . ------**'**'~~~'
ei
,. as4.'r* *
. t E
...w..... _ m - _ _.
M' n-
~
m.
, . , I y c:. i tn
[._" :b::
- :: l REACT 0ftS PLAflilED BY STATE m um g
^
- e .
- j$ j ...
s
, s y
o e 3 . . . ' :p; , ~
, y;
,,3'
_.,,, e' s
- - * ~
s , ...'y
., ' / 4 Mj-
).' , . . , , , [ ( h:.
. ~ _ , ~ . .o 3 ,., ..
j,,,. . I.Y
, . ==.e. ... ' 5 :.. J 6 y' ',[; ,
< *a j +
df.
} ,
I %% ** jf. s..
j .- '
~
.=f;: ,oses 4 4
'I -_ eseest .,..L .g [, k^ ;
404 so.
- '*a -' 'r -
c , .;gaP. ,,().
, egg ,
p.g * * /
_i- ,
'f
% , A ,,
j-- ..
e==e
,be b, ' 'g ,
'[M
+' -
- 'g'I :;.2 -
s new 4
L g,' =. , sa k. ,' A' .' 's' s 's
,s
- i,/
w'aa
- ] s' s <
.. ,,. , > !: C s
>' e.. . - . - ,,,,,,,,.[ ', , , ,
- < y- S.
{ '7 , s
, s
'; , + s4
,L--" 4 ,, v < {',,, ...:H
- 7 ==- ,, ... j.; pc;
- g (g P*,
. k ., ^ ,' <
[ll's .'
t -
s
, , I.. .. , s s et *41 - ** < ea - nY'J'[j.; , li
,. . , . . jp.. ,g.
.e .- "
' z.s ;
e ,, / 29
[f .
/ , , ,,
, i> r W' ,:s , <
.n<s...s.
e s
.v. v ;
v..> s gq.
x- a s s.>:'s s
s < :;e s . .E
. w .' v s s %.:; L.w y .
f' '
L' ". 6. . M.C . . , ' '
Nf W Putato m'CO^ff y
< < s . c+x, ,, + 1 s wp,;q.;;. ,,3.,. jg,3 ,. , ,, Q;L:;g;,;.;yy.
AL&dauA hilCM'OAN C E L AMOM A fa8f09 l L I tillt lasMt gettne000 2 ( 1 tl200 tatut stata 90s l C 3 title gaCMt CALIF 0musa MissimPPt omgcog ea=C=0 btCO P 4 8 t*38 vtLLow tatta i g 3 4133% tatut plaatt $saln;l i E 21L260 4 ACMB .
gan acatwies t=4 41300 f atal 6f amiltavn I & 2 Il20J tagne esOm fM i:AROLfreA PINN57tVANIA =
lu%:llleg 3 g J 107,asCas ,
gattasm Ostlig g g 3 a ge.)( t aC a t ptnet=3 3 Canotina 3.3) 4t 81>85 tise tachta;ne8 FutTOss 8 8 2 til60 EAC"8 s
pa Nt 99 A$8L A sue 00415L ANO 444taat 30 4 uge t t 4 82001 s tad t C atu@.g ; gg33gl Ntv thCtah01 & 2 title R&Cm) 8LLINot6 %tamalaa L g g got s TINNi&&E 8 Cata0LL & 4 215170 to Mt W JERSM CLfhCs4 alves l'Ild t Nosaga a*6 antic & L t Intgg tacel I S I III U I'0"I aftsasf4C 3 g a littu sacus 9ggag passtt esILg i g 3 tit)3 84Cnl
""" ALList Catra g egtsse Sa t s se (094 t 7 lassis SLb4 MILLS 4 & 2 tvle facna ,
Punf aJul i L 2 4 41%0 n atus JaailPWI I L # 41160 f aCo*' W ASHINGTCPI e it.el* P a l t-; l 5ftatstes 3 (;sgap ente vtas t C i ellJO (4 Cat laatif I & 2 4 83FF tacut a'AR Y LAP.D
- !9-auD'.3se t a n t 4 & 2 a llCS isch .
' S'# del pa a me%f I & J 4 & 300 t atut %
(JWCtal DJ1%f 8 & 1 titta satub *ble law *t4LE l 4 7 (1833 taCnl We%CONSIN
- t 0*MieD ni.* et l f tilais ,
a0tH404094 8 C 2 t900 faCne OM10 f v80%d t fillet t#ll 8 4 7 187e3 faCat 2 l * *' # # IIl70t 10 f al . . . . . . . . . . . . . . . .. . 79 UNgis put R f C RICO N3dfM CUall 8981) 14 t'L Afdato .e L04 A f f Ma s.0 f F i nse.
6 12 .
I
. . ~ . .
i i !
)
- .=. . ,
_ REACTORS UNDER CONSTRUCTION BY STATE DATA AS OF' JUNO.18U T" s -
x t ..
, U-
...5 - .:
- .0 f' , S ::. W:. - [::: :.:
+',
- J:
,< W:,
o , .e.~-
- t. . .
- !g - '
5 . t
' i E
'I. +
<s .- y r,, *.
.:q:
j o .sa e m ,
grm
........ 7 **'**
I j2 3 '
- TG . , . .
m: :
- A :+:: :
y-3.;:.
l -
) ****'
, vasc. s.A.?;.. , iG u.s.
gn. q'y t of:i.i!; i, 1
<l.
l
.o. \
r'--
- y,ces. h....
a f :1 (.
va.
- 6
=4 Dk
- *t .::
avo
-W
' ' ** g .. ,, ) 8 .,l ' .j :.
L-k so4e k, ,
mi M.E i
l use i _ 0*"8 / m :: ' 7'u
,, a, .f 4.84 -
/
, -- vs. va- 3 ' !
l- , s= va.
, , u .
co. a 1
.., - --... 4, ,..,_,,,4, 3
,w...
m e.
. =. :
& $1 '
\ #**** t
.a f abb j %_
l e
,,,, r
- *' { =
. y% l i
I '
... j 8< . .u ss.
aka. clo.
e ** ;
l s.
I L .
s
,e n.s # .. .1. e LA-
%:?, s b::. *
, %:!.1l:- t'"
h.,@ ..w **
I < ,
t q
w :! Y!YS I 1
i s
, I 4.:
i ,
- :+;. i ALABAMA If1.*1 OELA4AAf "ff. .!
LOUISIANA WM NEW YORF M *I le .eMate l '.13 s
.
- bati! t .50 SOUTH CAROLIN A VIRGINIA WM s 8 whend t 934 Mine kJe Pouit 2 1100 Ca:4eba l 1153 Nortn Aana l 898 j f e .ev enie 2
- 2 D pggy g. R4e Stad ? 936 Ca:4,ba l 1153 Aerth Anaal 395
- '*' $5eetham 319 Feel Jp 83 lII2 hetn anea 3 M7 9
) Fp'n2 n;, ithsel s!)
C5Oi'G$A MICt ?lCAN NORTH CAROLINA
- EAchee l 1230
+ + Cherokee 2 1230 harth ans4 4 907 i
= M ANSAS Caen 2 1%0
,,,.n ; .,, frmi2 ic93
- Martis t 900 *
- Che, g, rah,e,e
, 3 973 3 1220 1 4:sa%,i hn Cae vait 2 s .; y,c,3 ,;;; k diaad! 460 e Marns 2 900 Mhad 2 SIL 900 TE W SS M a A P'IO N A g.,, *, . .* 3 . Hari,s 3 1 1233 WASWNGTON i MISSIS $1PPI , ppg 4 900 Hartsville, y Wast. heisar Prel 1 121d pg,gi 1 Poe mse l 12is 'k ' \0'S g* I GIN CraadCWil 1250 Ps.e Vrese 2 123J NcCaust 1133 Hadssile 3 1213 wash. hadur rasi.2 110 3 . ;;,
ts<e wu 3 ills Id * * *, , *LI3 CrandCait! 1250 McCvive2 1180 liarnvin, 4 1233 *
- Was4. Nwur Pmt4 1213 L-(
OM 0 Sequoyait i 1148 *
- Wash. Nuctew 3 1242; '
CALIFORNIA tr* 2 tuo 2
Causest i 1120 . , g,emer,,.&ene 2 906 810 Sequerah 2 1148 Watts Bar1 1177
- o h har 5 m2 j Diaa*Caesa l 104 Cmea l ajj Ca'fa*Jy! 1120 . . c,,,s.6nse 3 906 0 49 Caw 2 1D* C *'m 2 Watas lar 2 1177 L all *
,1 L,0 ee.,2 11:; LLJe t IU1 P EW HAMPSMihE Per*)1 1205 TEXAS
$ 3,a :neer,J 1140 LL:lv i ICA g g gg g g g . ;
SealiroonI 1200 A Mace Pesh 2 1t50 "
.) CONN ECTICUT I
- I* 43 I O PENNSYLVANIA Satminasi 1250
?
P j Id"'P 333'** 641 NEw;gpsgy n '
""' I"b Beaeevva e72 852 Southletest 1250
,4 g a.,3Ag fosses sing 3g;g L**"31 1045 n 050 *
, ..ee -e d, 9,. . . _ ..d..,W.
I.
6-11 i
. . . . t 1
..k.- -... a p 4 k -l,c.a,- a ...4; a p.,--n - -.s,+,s +<.wsc- ..n: = 4 .- .- u,
- ::::q l
- .::. ..:..~?.
u= . LICENSING PROJECTION SU) MARY, FY 77 and 78, as of'6/15/77 :_ i 7
f..:.l.m
- :4 =
Licensing Projections FY 1977 .
=q River Bend 115 2 ......
St. Lucie 2 !? ~ j Perry 1 6 2 Hartsville-1,2,3'S 4 32 Wolf Creek'l - ~ ~ ,
Washington Nuclear 3 4 5. ,
Sterling 1 .
i=. .
Cherokee 1,.2,.5 3 . .
TOTAL: Sites.- 8; Units - 16 . !
- 7 !
i'9 Limited Work Authorizations 5. ;.
Wolf Creek 1 N:
Washington Nuclear 3'G 5 !
Marble Mill 1 G 2 '
Pilgrim 2 -
Phipps Bend 1 6 2 .
i.
TOTAL: Sites - 5; Units - 8 ~
i Preliminary Design Approvals
' ~
Yellow Creek ,
Light Water Breeder Reactor Resar - 3S =i '
Gessar - 238 "~"
i Swessar/Ressar - 3S l Gessar - 251 Sundesert l I
San Joaquin Fort Calhoun 2 I Fluor '
Blue Hills s. I
~ =. s::;..
TOTAL: 11 !
r v
^"..~...~'.
i
. !f, ;
+ '
9 0
I
- , , - . .- . ,, .L.,
._ _ - . _ ~ _ . _ . _
l
\:
' .nv.7f. ,
=:
- .1:::::
- i:.+:)
M:
t: N.*'(.'
~ Licensing Projection.- 2.' ...
','ii.ii' Ooerating Licenses r ECrystal Rivir 3-Davis-Besse 1 i.... ..= - .
""~
'Farley l' North Anna 1 =-
- . =. .
T0rAL: ' Sites - 4;. Units - 4 ...: .~"'.
Licensing Projections FY 78 i'
' Construction' Permits '
- Washington Nuclear 4 ;
LMarble Hill 1 6 2 -
.Phipps Bend 1 6 2 ; '
-Perkins 1, 2 6 3 ; -
Tyrone Jamesport 1 6 2 -
- Pilgrim 2 I=. .
Harris 1, 2, 3 6 4 Yellow Creek 1;6 2-Pebbic Springs 1 6 2-Greene County .
Skagit 1 6 2 ,
Atlantic 1 6 2- ::E:
Davis-Besse 2 6 3 -
Black Fox 1 6 2 "
=...
TOTAL: Sites 15; Units - 29 Limited Work Authorizations
. Black Fox 1 6 2 i Yellow Creek 1 6 2 ;
Sundesert 1 C 2 I o Es l
'IDTAL: Sites - 3; Units - 6 -
{
s l is
.h 1 .
e b
e + --
.r . . . . , . . o -.<m*, s-
4+
Licensing Projections --3 t Preliminary Design Acprovals
- BSAR - 205 '
Gastar - 6 Swessar/ESAR 205 Perryman FNP 1-8 Resar 414 TMAL: 6
- r:;.
Ocerating Licenses '
Tnree Mile Island 2 Arkansas 2 Cook 2 '
Sequoyah 1 Hatch 2 Zim.er 1 Diablo Canyon 1 6 2 North Araa 2 McGuire 1 L TOTAL: Sites - 9; Units - 10 I'
h e
8 f
i b
t 9
k a
l
= =&= ,
m :: )
v
- t==
k!!.
.+
(; g... .
, >; ; _ APPLICATION FORECASTS, FY 77 and' 78, as of 6/15/77 : i
. "Il .
NEW ADPLICATION FORECASTS, FY 77'
'5'.9. '
Constructicn Pemits tZ t
[, 1 Erie 1 6 2 c Sundesert l'5 2 :==. . !
"~". . . . . . .
TOTAL: Sites - 2; Units - 4
-[
Operating Licenses =d j
~
Midland 1 5 2
'!!.! j q
TOTAL: Sites '1; Units - 2 t
],;,", ';
-i
- Standard Plants e i
Gibbsar. (CE/B5W) I I
Site Reviews '
$:: Per p an . .
..~.~.
i Suecial Projects i t
S 8 G Ship Application'(ERDA Naval Reactor) '
SEN APPLICATION FORECASTS, FY 78 i Construction Pemits !
Sumit Rev 1 Palo Verde 4 '
l Perry: nan 15 2 == #
. TOTAL: Sites - 3; Units -.4 d: !
I p.E
. 8
,- i
..9
'i e * ,
e .,,.,-n _ .- - - , , - -4 _ . _ , m
y.=
r -
r New Application Forecasts - 2 Occrating Licenses , ;g.gf.
Bellefonte l 4 2 Susquehanna 1 4 2 Commanche Peak 1 6 2 . . .
South Texas 1 4 2 . . .~IL Braidwood 1 4 2 Byron 1 6 2 Grand Gulf 1 4 2 Waterford 3 KPPSS 1 TOTAL: . Sites - 9; Units - 15 Standard Plants ~
~
Bopsar Revision f Cessar System 80 Gibbssar (BSAR)
Gibsar (CBSSAR)
Gessar 238 h1
(~ Standard Bh'R B0P Gilbert Comm. - BOP TOTAL: 7 ,
Site Reviews Colorado New Mexico Boardman TCTTAL: 3 Special Proiects Advanced Submarine Core
G C F R PSID S 8 G Ship Bechtel hTGR Gas Turbine PDIS SLSF - MOD l
IOTAL: 6 l
l 1
t.
I;E *
- id ==. .
- lll:.
> ~
- c:.
i je REACTOR SAFETY TABl E '0F CONTD.TS .
- 1. -
Siting' - is
- dh- i
.a.. Locating nuclear power plant sites.
h"EE
'b. Reactor siting milestones
- c. Siting policy and practice ,
- d. Nuclear power plant design - site-related criteria
'e. Underground siting Sandia Study E
'f.. Faergency plans: .
- 2. Design,. Construction, and Operation
- a. Design and construction criteria for nuclear power plants
- b. Quality assurance
- c. NRC inspection philosophy -
=:
- e. . Routine emissions of radioactivity - ALARA
' f. . Training and qualifications of ' nuclear power plant personnel ,
- 3. Safety Iss,ues-
- a. Resolution of generic safety issues
. b. Generic Issues:
(1) Fh'R Steam Generator Tube Integrity (2) Overpressurization (3) Water Ha:mer (4) . D.c.~ Battery Failure .
- 4. Feedback
~
- a. Inspection and enforcement activities to improve. reactor safety
.b.
Systematic evaluation program for operating reactor::
- c. NRC research programs (including descriptions)
- d. Risk assessment in the regulatory process (1) The Reactor Safety Study
- e. The Brown's Ferry Fire l :. p l
t i
F L
. - _ _ . . _ - ._ _ . . . . . ~ , . . . , _ . _ __
I q
n+:-
l LOCATING KUCLEAR POWER PLANT SITES g
- g. 7;_ ,
e a NRC regulations governing reactor site criteria are given in-the \ ..,,,,
Co=tission's. regulations in 10 CFR Part 100. An application to
_ construct a nuclear power plant, either land-based or offshore,
'must demonstrate that the proposed' site is in conformance with these criteria.-
- =+
The exclusion area required for a nuclear power plant is .detemined in part by the potential radiation dose to an individual located at
'the exclusion area boundary during a release of radioactivity during an accident. Experience indicates that an exclusion area radius of-
- about 0.4 miles will usually. provide' reasonable assurance that the dose guidelines of 10 CFR Part 100 can be met. This is because current engineered safety features can assure that doses are within the guidelines even under extremely poor nieteorological conditions.
t av In addition, for_a designated exclusion area to be acceptable, an applicant for a construction _ permit must also demonstrate that he has
- authority to determine all activities within the area, including the ,
exclusion and removal of personnel and property frc, the area.- -
- Ostership of all land within the designated area is . generally how this authority is obtained; however, other arrangements, such as leasing, are acceptable.
10 CFR Fart 100 also requires designation of a low population zone .
O eround the site. Acceptability of the applicant's designated low popalation zone is based oil radiation exoosure computations and a determination that there is reasonable assurance that appropriate ,
protective measures can be taken on behalf of the population within the zone in the event of a serious accident. Excerience indicates that a low population zone radius of two miles is usually acceptable.
'Ihe regulations also req'uire that the closest boundary of the nearest population center of more than about 25,000 residents be at least 1-1/3 times the distance to the outer boundary of the low population tone.
hhere large cities are involved, a greater distance may be required because of integrated population dose considerations.
The boundary of the nearest population center is detemined
- j. upon consideration of population distribution. '
I.
.i
'a
! . tu
, +
b 0 e.+.. I'"If
- . < - - , ..* .. . - - - , , . e-a - , ., ., .- . , - -
REACTOR SITING MILESTONES X
Ii 1953 SHIPPINGPORT REACTOR ANNOUNCED 1954 ATCb!IC ENERGY ACT SIGED 1955 10' CFR PART 50 PROPOSED 1956 10 CFR PART 50 ISSUED 1957-60 GENERAL SITING CRITERLA DRAFTED 1961 10 CFR PART 100 PROPOSED 1952 10 CFR PART 100 #3 ITS SUPPORT (TID-14844) ISSUED 1963-66 DETAILED GUIDANCE ON COMPENSATION ALLOWED ESF'S '
1967 GENERAL DESIGN CRITERIA (APPENDIX A TO PART 50) PROPOSED l9'68 IhTERNAL STAFF GUIDANCE REPORTS ISSUED 1969 METROPOLITAN SITING QUESTION CONSIDERED
~1970 NEPA EFFECTIVE AND DIPLEETED (APPENDIX D TO PART 50); ALAP REQUIREESTS ISSUED AND SAFETY GUIDES ISSUED PUBLICLY.
1971 CALVERT CLIFFS DECISION (NEPA), GENERAL DESIGN CRITERIA ISSUE ALAP GUIDANCE (APPENDIX I TO PART 50) ISSUED AND SEIS.!IC A ,
GEOLOGIC SITING CRITERIA (APPENDIX A TO PART 100) PROPO 1972 GUIDANCE ON EVALUATING SIIES (RG 4.2) ISSUED #3 -
E.WIR010ESTAL HEARINGS INITIATED 1973 NEGOLD ISLAND POSITION (NEPA) AND SEIS'IIC AND GEOLOGIC SITING CRITERIA (APPENDIX A TO PART 100) ISSUED 1974 NEPA LICENSING REGULATION (10 CFR 51) ISSUED; GEiERAL SITE SUIIABILITY CRITERIA (RG 4.7) ISSUED: AND STAN"DARD PLANT DESIGN REGULATION (APPENDIX M TO PART 50) ISSUED.
1975 REALTOR SAFETY STUDY (NASH-1400) ISSUED; STANDARD REVIE7 PLAN '
AND SAFETY REVIEG ISSUED:
PRACTICE INITIATED. STAFF REVIE1 OF SITING POLICY AND. ,
1976 STANDARD REVIE1 PLAN FOR ENVIRON. ENTAL RE IB G INITIATED; FEDERAL / STATE COORDINATION STUDY INITIATED; STAFF REVIEf PAPER (SECY-76-286) CGIPLETED.
1977 SEAE200K DECISION; EARLY SITE REVIE1 REGULATIONS ISSUED:
FEDERAL / STATE COORDINATION STUDY CO:SLETED.
goe D.
C.
1
i 3
4 1
..-)
.. J. 1
- x. - SITING POLICY AND PRACTICE~- .h r .?. .
'~" , .
I._. . '
A'cor::plete' reexamination-of NRC siting policy and practice has . '.;
been in progress since June 1975, and staff analyses have been
- tm discussed with the Comission at approximately six-month intervals [r
-since then. ~
E 1l.
Key policy considerations in NRC siting activities. include: >
1--
treatment of. radiological risks resulting from severe I external natural and man-made events;_ ,
==
[;
conpensation of site / region characteristics through ;
safety design features;
- n ;
-- reasonable assurance of no undue risks from routine operations t and from potential reactor accidents; !
'} T d.
overall site suitability evaluation combining facility design features and site / region characteristics;.
^j t
cost / benefit considerations-in the environmental review; !
interdisciplinary approach in the environmental review;
?: --
treatment of public interests early in the siting review; i balancing environmental, social, and economic factors; , !
, 1 analyses and issues addressed in the environmental impact i
statement;
. . i roles of other Federal agencies in the siting review; !
public hearings and staff reviews in the siting process (
and procedures; +
coordination of State / Federal activities. -
Activities in progress .Sclude: ,
t development of alternative policy statements expressing , .
present and possible future practices;- ,
{.
preparation of policy analyses on specific siting issues, =*
(
including alternative sites, emergency planning, ,= !
accident evaluations, and seismic requirements. l
= ;
The Federal / State Siting Study issued for public coment
(see section on NRC and the States) i i
r l
l
. = ;
e ;
l
.t
._-____-_--._-_._______--_-,.r , - , , - , . -w,- --g,-- - , - , . , ,- _, ,. .. ,g , n , ,e,_. .-w.y , ,,.m
=-.::
=
Siting Policy and Practice - 2 I Current licensing cases involving controversial siting issues include:
~::"':
-- Diablo Canyon (California): new infomation shows added =r seismic hazard. (See additional infomation under Current Issues section)
-- Seabrook (New Hampshire) and Charlestown (Massachusetts):
Appeal Board (ALAB-390) reversed the Licensing Board and denied intervenor's motion to require consideration of possibility of emergency action outside the Low Population Zone as part of the site review. Tne Appeal Board urged the Commission to take up the issue in a rulemaking proceeding, and in June 1977, the Comission announced it would undertake such a rulemaking proceeding.
e s
1 l
I l .
L l
4 f.
(Dl l
t k
l
e 2;
.r NJCLEAR POWER PLANT DESIGN - SITE-RELATED CRITERLA . . .
'p' E.
[10 CFR Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants"] g.
r CRITERION 2: DESIGN BASES FOR PROTECTION AGAINST MTURAL PHENOMENA Structures, systems, and components important to safety shall be '
designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without p t loss of capability to perform their safety functions. The design ti bases for these structures, systens, and components shall reflect: k
-- appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient '
margin for the limited accuracy, quantity, and period -
of time in which the historical data have been accumulated.
-- appropriate combinations of the effects of nonnal and accident conditions with the effects of the natural phenomena;
-- the importance of the safety functions to be performed.
CRITERION 3: FIRE PROTECTION Structures, systems, and components important to safety shall be designed and located to minimi::e, consistent with other safety requirements, the probability and effect of fires and explosions.
The criterion does not specify that the fires or explosions referred to are only those associated with in plant events. Consequently, fires and explosions of external origin may be considered in applying ;
this criterion.
1 Ett CRITERION 4: ENVIRO.T.IENTAL AND MISSILE DESIGN EASES - -
Structures, systems, and components important to safety shall ba appropriately protected against dynamic effects, including the [
effects of missiles, pipe whipping, and discharging fluids, that U -
may result from equipment failures and from events and conditions outside the nuclear unit. s. ,
?
CRITERION 60: C0hTROL OF RELFASES OF RADI0 ACTIVE MATERIALS TO TrE EWIRO.YST There must be provided sufficient holdup capacity for retention of gaseous ,
and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose ,
unusual operational limitations upon the releases of such effluents . . .
[ to the environment. [ -
U it i! ,
E?! =
l l ,
/7 UNDERGROUND SITING--SANDIA STUDY RESULTS [
H
!L
"~ '
I:i his April message on nuclear energy, President Carter called for an examination of the underground siting of nuclear power plants, as ;;;m !
a means of increasing safety. M= ,
' WW Sandia Laboratories has been conducting a study of underground siting, and this study has been completed and is expected to be issued in
[5;
- July 1977. The study considers three basic underground installations
- .
t
-- cut-and-cover;
-- rock cavity _with vertical entry; _
j
-- rock cavity.with horizontal entry (e.g. into side of a cliff).
l In each case, the turbogenerators were not to be included in the .?
underground installation.
The principal conclusions reached by'the Sandia study are as follows: ?
- 1. CONTAIDENT OF PADI0 ACTIVE MATERIALS: underground siting has ;
only negligible advantages over surface siting for containment of @ (
radiological releases from accidents which do not involve a core melt. '
For those accidents which would involve a core melt, underground l 0 siting does provide an improved containment, conditioned on the ,
g reliability of penetration seals'and access closure mechanisms. ;-
- 2. GROU'ONATER CONTRIINATION: groundwater contamination may '
be more severe from an underground site because of the greater likelihood of contaminated sump water escaping. "
- 3. SITE AVAIIABILITY: the amount of land potentially favorable ;
for cut-and-cover siting is much greater than that available or <
favorable for mined rock cavern siting. Underground siting in '
general imposes additional siting restrictions, u ,
- 4. SEIS'IIC VULNERABILITY: there is a modest reduction in i seismic vulnerability at reasonable depths of underground siting. .l S. COSTS: increased costs of construction, equipment, and - )
interest required by extended construction schedules would make
- s. fE ,
underground nuclear plants more expensive than surface plants. 1 Estimates of the underground plant's cost differential range . t from 20 to 40 percent increase. :,
I r I i
i A
=
?
1:=
l l t
l l u-5 4ll.
Underground Siting - 2 f'
!(.7
- 6. FEASIBILITY: there are no insumountable technical problems which would prevent underground siting of a nuclear power plant using any of the three methods. The major design considerations gg would be the active and passive systems required to seal accessways, ..,,
but these are within the abilities of current technology.
- 7. OPERATIONAL CONSIDERATIONS: underground siting complicata =
the maintenance, repair, and inspection of equipment. Spent fuel handling is ha::pered. Personnel safety may be a problem, particularly in regard to mobility during an emergency.
- 8. PLANT SECURITY: underground siting provides increased sabotage protection only against high-strength forceful threats i or external attacks using mtmitions. These advantages are offset by a reduction in the ability of the security personnel to regain control of the plant and to control damage. ,
- 9. PROTECTION AGAINST EXTERNAL EhVIR0hMENTAL PROBLB!S: underground siting provides increased protection against external environmental problems, particularly aircraft crashes and acts of war. But underground plants would experience increased vulnerability to '
ficoding. p.
(f I
4 5
f l..
!a 6 o ,
t k
fii. .
I p
[5 :
- t L
t
[N !
i l
i I
/ Emergency Plans - 2 - !?:
b, C g
IE A continuing NRC inspection program is carried out to assure that -
each licensee maintains a satisfactory state of preparedness to effectively implement his emergency plans. ,.
The emergency preparedness site inspections are conducted on an annual basis and are divided ar.ong four major areas:
-- coordination with offsite agencies;
[
-- written implementing procedures; d: ,
-- equipment and facilities; };
H .
-- test exercises and drills, i
Each of these areas is covered thoroughly, and more than one site visit is frequently required to complete the inspection.
e t
P h
P P
,,' kI
{.:
. . . r
. N :.
]
- :.7.*.*
-1 :
he NRC's standards and criteria for the evaluation of proposed nuclear power plants include provisions for substantial conservatisms ,
in design and operating safety mrgins. Nevertheless, the NRC ~
recognizes that emergencies can arise in the operation of nuclear s:
power plants, and has therefore taken steps to assure the establishment of an acceptable state of preparedness to cope with emergency situations. .'
In 1952, the Atomic Energy Comission published its Reactor Site Criteria as 10 CFR Part 100. One of these criteria references a need for consideration of establishing a capability for taking protective measures, in the event of a serious accident, to -
protect the public within the Low Population Zone. L e scope and extent of advance planning for such measures, e.g. evacuation of persons or instructions to take shelter on a timely basis, is explicitly identified as one of the factors to be considered in detemining an adequate Low Population Zone.
In.1970, the AEC published its requirements for plans to cope with emergencies in 10 CFR Part 50. These requirements, along with the portion of the Reactor Site Criteria mentioned above, -
represent current NRC policy with regard to emergency planning that must be undertaken prior to issuance of a nuclear power plant operating license. Elements of preliininary planning are required for issuance of a constructica permit, and the elements of if, substantive plannlag are required for issuance of an operating license. ,
The planning elements required are set forth in Appendix E to Part 50. hhile many of the planning elements identified in -
N Appendix E are directed specifically to radiation emergencies, the scope of Appendix E.has generally been understood 'to apply also to situations which have the potential for becoming radiation emergencies, e.g. fires, floods, hurricanes, etc.
Accordingly, emergency planning might be required to encompass areas beyond the Low Population Zone to adequately protect the health and safety of the public.
Emergency plans are reviewed by the NRC staff and are frequently modified and improved as a result of this process. At the conclusion of each review, .the staff's findings are published in the Safety Evaluation Report for each proposed licensing action. ,.
Before a plant is licensed to operate, the staff muist determine that the emergency plans provide reasonable assurance that appropriate measures can and will be taken in the event of an emergency to protect public health and safety and prevent damage 3 to property.
)
..#.. n.. . . . . . e
- c. ' DESIGN AND CONSTRUCTION LRITERIA FOR NUCLFAR POWER PIETS
'(i..
Tne mininta requirements for the principal design criteria for ' .= . .
water-ccoled nuclear power plants are contained in Appendix A,
" General Design Criteria for Nuclear Power Plants," of 10 CFR Part 50.
Some of the more important and essential elements of the principal design criteria for a nuclear power plant are the proposed construction procedures. Many of these procedures are contained and codified in well-known design and construction codes which have been published -
and adopted by national societies and institutes. A few examples F of these codes are:
-- the American Society of Mechanical r_ngineers' Boiler and Pressure Vessel Code;
-- the American Concrete Institute codes;
-- the American Institute for Steel Construction Code for Steel Structures;
-- the codes and standards of the Institute of Electrical and Electronic Engineers.
G Tne foregoing listing illustrates that there are widely used standards and codes which can be adopted to provide assurances that acceptable construction procedures are followed during the constructicn phase of nuclear power plants. These national standards have been used in many different types of construction projects for many years, and have provided acceptable protection of the public health and safety. -
g In addition to the national standards, the NRC has established guidance for applicants for licenses to construct and operate nuclear power plants. This guidance is contained in the NRC Regulatory Guides, which are issued to:
-- supplement the national standards, or
-- provide guidance for acc ptable design criteria when the national standards are not sufficiently conservative for ,
certain aspects of the construction of nuclear power plants.
When possible, the NRC staff issues design criteria for specific systems or components of nuclear power plants. These specific design criteria are contained in Branch Technical Positions and reflect the knowledge obtained frem the review of previous applications for licenses and from the operating experience of licensed plants.
g
g Design and Construction Criteria - 2 ~
A= 4
(
Finally, the staff establishes ad hoc design criteria for the =am construction of nuclear power pI~ ants when the existing national -
standards, regulatory guides, and Branch Technical Positions do not address unique design features of a proposed facility. -
A ee W
f f p t
t 4
t t
ei. *,
F
-=
$?:
!?w g, _ QUALITY ASSURANCE 1 u~
i
- The essence of a good quality assurance. (QA) program is to assure that disciplined engineering and responsible management practices are applied to the design, fabrication, construction, and operation of
'=
nuclear power plants.
]
The htC's QA requirements are contained in Appendix B (10 CFR Part 50), '
" Quality Assurance Criteria for Nuclear Power Plants and Fuel I Reprocessing Plants." They provide' a basis upon which the hRC judges the acceptability of QA programs, and apply to all activities affecting !?
safety-related functions of nuclear power reactor structures, systems, F and components. '
hRC has severa1' specific QA responsibilities:
-- developing criteria and guides for judging the acceptability of nuclear power plant QA programs;
--- reviewing the QA programs of the licensee and his principal i .
contractors to assure that sufficient management and program '
controls exist; -
-- inspection of selected activities to determine that QA prograns'are being implemented effectively.
Each hRC licensee is reponsible for assuring that his nuclear power plants are built' and operated safely, and in conformance with NRC c".
regulations. Licensees are also required to assure that their suppliers meet the applicable hRC criteria. In this respect, the licensee is responsible for functions such as product inspection and non destructive testing of reactor components, structures, and - -
systems, even though he may on occasion delegate the actual perfomance of the activity to another organization.
The hRC QA program review process involves the following sequence:
- 1. Approximately 9 months before an applicant applies for a CP hRC staff personnel meet with him to discuss h2C's QA requiremen,ts and the mechanics of hRC's review and inspection activities associated with a CP application. Even prior to a CP application, a utility must demonstrate that it has developed and is implementing u an effective QA program for any safety-related ongoing activity, i
- 2. Immediately after receipt of the application for a CP, h2C conducts an accelerated review of the QA program description I f
contained in the applicant's Preliminary Safety Analysis Report (PSAR). y l This review detemines if the applicant has established a satisfactory [.5 l QA program in the areas of organization, program,. design control, ki: l si [5 I I::
p q I...
"I L.,
. . . . s.
1
- - - a -,u ...a., ..a.. e a , . . . u + n j'
I m*'l j,'..*:.***..,
.{ QA' 2 L r
,.w.. . . . .
-procurement, and audits, If? the QA program is found unacceptable j:53 during this review, the application is not docketed. If the QA .
prog- m is_. acceptable, an inspection is made to evaluate QA 2 ....
program implementation.' ~~ "
r.it-
'3. Foll'owing docketing, the h3R staff _ :onducts a detailed review
, e and evaluation of the QA program of the applicant and his principal n !
contractors, including the reactor vendor, the architect-engineer, ,
and the constructor. A conclusion of program acceptability is e based on defined acceptance criteria, and the staff verifies that: ==
the organizations and persons perfo ming QA functions have FN the required indep2ndence and authority to effectively carry <
out the QA program without' undue influence from thosa -i
- .s directly responsible for costs and schedules;
-- the'QA program contains requirements and controls which, i when properly implemented, comply with the requirements of Appendix B.
77 :
(
- 4. The hRC's Office of. Inspection and Enforcement '( I 4 E) conducts periodic : scheduled and unannounced field inspections of the applicants QA program implementation as.well as those of his contractors and suppliers. These inspections start prior to docketing of the application and continue throughout the construction phase, the preoperational testing program, and the operating lifetime of the facility. The field inspections cover:
-- review of the applicant's QA perfomance, including audits of the applicant's QA records and documentation; ;
-- surveillance of construction practices and inspection of the facility at various stages of constzaction; 2.t review of the qualifications and training of the construction personnel as well as those of the QA and quality control personnal (this includes all personnel t at the site, including the specialized subcontractors, and -
QA/QC personnel at the manufacturing facilities of the i vendors and suppliers. '
- 5. During the OL review phase, the staff reviews the operational !
QA program in much the same nanner as the construction Qi program was reviewed earlier. . The h2C maintains its QA responsioilities throughout the . operational lifetime of a nuclear power plant, '
through frequent and regular inspections of operations and records. ;
Also, the hRC staff must review and approve any change to licensed operating conditions. - '
b e e p Y
I
= :.
iE '
QA - 3 g.3
- (63..
h~nere inspections or events indicate deficiencies in the QA program or its implementation, the staff requires appropriate program upgrading or uses enforcement authority as necessary to achieve .. . . .
proper implementation. If a general problem develops, improvements
~
in QA programs are made on an industr/-wide br. sis. :- :
ARC looks to the power plant owners, the utilities themselves, to take the leadership role in assuring the quality of their plants and operations.
Tnis requires careful attention to the selection of engineering specifications and QA procedures and practices for each task and their implementation by the workers on the job.
Most importantly, there must be adequate and qualified personnel at the management, operating, and staff levels. The NRC places the highest emphasis on active involvement of top managemene in QA programs.
1 l
l 1
s.
[.'.
g p g.,
j::
l HEC INSPECTION PMIIDSOPHY f..
The hRC inspection effort consists of a planned inspection approach and a reactive ins?ection approach. Both approaches are based on the premise tnat the licensee is responsible for ensuring the proper design, construction, testing, and ' safe operation of the facility. .. j The h1C inspection program is the apex of a pyramid of inspections, audits, and controls within the nuclear industry. Beneath this apex ..
=
are the layers of inspections, audits, and controls implemented by the hRC licensee and his contractors.
hRC inspections are not intended to duplicate or substitute for these lower tiers, but rather to ascertain through examination of a vertical slice to an appropriate depth whether these " people" activities are adequate to meet the licensee's responsibilities. hRC rules and regulations and licenso conditions are the fundamental acceptance criteria for this detemination, and comoliance with them demonstrates that utility management performance is ehsuring public health and safety. -
Tne planned h1C inspection effort is conducted in accordance with a deilned program expressed in detailed inspection procedures and is accomplished at prescribed intervals by hRC field inspectors. 1 g- The principal objective of this inspection effort is to provide reasonable assurance that hRC-licensed activities are being conducted safely and in compliance with h1C requirements.
The reactive hRC inspection effort is conducted' in response to infomation received by hRC regarding conditions or events which have occurred during design, constniction, testing or operation of a nuclear power plant and which may affect public health and safety.
The principal objective of the reactive effort is to obtain sufficient infomation through independent in-depth examination to establish the significance of the particular condition, event, or allegation, and to effect corrective action commensurate with the significance of the problem.
L 2
1
=
v;;.d$
n 4 BERGENCY ' COP.E COOLING: SYSTBI -(ECCS)' N k
- Each' nuclear power plant licensed by the NRC contains a number'of k engineered safety systems, one of which is an Emergency Core ~
Cooling System (ECCS) . -
.s,
. Acceptance Criteria for' ECCS were published January .4,1974, a and the ECCS proposed by an applicant for a CP undergoes a ~
- thorough review prior to .is'suar.:e of the CP.
The ECCS consists of many redundant subsystems, each capable of cooling the reactor core under emergency conditions. This
. planned redundancy is part of the engineered safeguard design philosophy of licensed power reactors which generally requires that no single failure will be allowed to impede the functioning EE
. of systems that are essential to safety.
Prior to any fuel loading and issuance of an OL, the ECCS is rereviewed by staff to assure that, as designed and built, it ,
+=
conforms to the staff's Acceptance Criteria. ' .
During this rereview, the Technical Specifications for the completed power plant are carefully reviewed to detemine that the pre-operational and periodic testing that will be performed
- g. on'the. individual subsystems of the ECCS meet staff requirements.
- After issuance of an OL, the status of the ECCS is monitored via the periodic. testing procedures detailed in,the Technical Specifications. .
No licensed power reacto,r has had an accident situation requiring actuation and full-scale operation of the ECCS.
In addition to the periodic testing of the major subsystems of the ECCS installed in all power reactors, NRC is conducting ,
a confirmatory research program designed to study the performance of engineered safety features such as ECCS performance during p various' accident or loss-of-coolant mnditions.
9 4 '
e
':i.:...
4 i
l 4 f I
1r 7
l, ll2 r::~~" s I
I
'g::
- E
.The Co mission subscribes to the general principle that radiation s
- exposure of the public .should be kept.as low as is reasonably achievable,-
within established radiation protection guides. Tnus, operating ,=
licenses . include provisions to linit and control radioactive effluents from nuclear power plants. .
As defined, the term "as low as is reasonably achievable" (ALARA) '
requires taking into account the state of technology, the economics-of improvements in relation to benefits to the public health and :
safety, and the general relationship of these to the utilization of nuclear energy in the public interest. ,
'Ihe Comission undertook a rulemaking action to adopt numerical guidelines on routine emissions of radioactivity from nuclear power.
plants. Tne initial rulemak2.ng decision established numerical 9 guides for design objectives and limiting conditions for operation -
to' meet the criterion "as low as practicable" for thR effluents. !
Subsequently, the Comission amended the regulations to' incorporate ~
ALA % , not to reflect a change in the objective but rather to use !
language which more closely describes its intention. '
" The guides require that:on design objectives set forth in Appendix I (10 CFR part 50) ,
-- the calculated annual total quantity of all radioactive material i above background to be released from each thR will not result t in an estimated annual dose or dose comitment from liqui >
effluents for any individual in an unrestricted area in ,
excess of 3 millirems to the total body or 10 millirems to any organ. ,
-- the calculated annual total quantity of all radioactive material '
above background to be released from each LhR to the atmosphere ;
will not result in an estimated annual air dose to individuals !
at any location near ground level in unrestricted areas in ;
excess of 10 millirads for gamma radiation or 20 millirads ;
for beta radiation. However, lower or higher quantities '
may be .specified by the Cennission depending on whether the level released will result in an estimated annual external u dose from gaseous effluents to any individual in unrestricted =
3 areas in excess of 5 millirens to the total body or 15 >
millirems to the skin.
I h
t"::.~
1... i I" ,
,, _ . , , .- + ' ' - * *
, r .
I l l '
4 i
' a:
3 y+I:.
g ALARA - 2 g
. ;;o .
p ;
the calculated annual total quantity of radioactive iodine 2+ -!
-and radioactive material in particulate fom above bach;round l to:be released from each thR in effluents to the atmosphere will' not result .in an estimated annual dose or ' dose cormitment a for any individual in an unrestricted area from all. * ,
. pathways of exposure in excess of'15 millirems to any organ. ;
t the radwaste system shall include all_ items of reasonably
- demonstrated technology that, when added to the system can, ; ;
for a favorable cost-benefit ratio, effect reductions in dosa '
' to the population,within 50 miles of the reactor. As an !
. interim measure and until establishment and adoption ~ of
- better values, $1000 per total body man-rem and $1000 per man-thyroid-rem (or such lesser values as may be. demone rated :h I to be suitable in a particular case) shall be used in this cost-benefit analysis. '
The guides' for limiting conditions 'as set forth above apply in any '
case where a CP application for a Lh2 was filed on or after January 2,.1971. '
For each thR constructed pursuant to a pemit for which application a
.. 1.as filed prior to JanuaryJ2,1971, the holder of the CP or C,L . ,
i
- C.. .. was required.to file with the Commission by June 4, 1976: .
-- - such infomation as necessary to evaluate the means I employed for keeping levels of radioactivity in effluents !
to unrestricted areas as low as'is reasonably achievable, and '
plans and propose'd technical specifications developed for the purpose of keeping releases of radioactive materials ;
to unrestricted areas during nomal reactor operations, , ;
including expected operational occurrences, ALARA. !
i In summary, plant operations'are controlled by Technical Specifications !
' appended to the OL which keep release of radioactive materials ALAPA.
- The potential doses are controlled by restrictions on the rate of ;
radioactive liquid and gaseous effluent releases, and are further l checked by radiological environmental monitoring programs. .
- s. ,
s i:r !
l t
k" y ,
i
w
! 1 e '
^
f ,', *
- TPA" INING AND QUALIFICATI" ON' OF hUCLEAR PO'GR PLANT PF.RSONL j?. ..;.. 'l G.- i~~!
The IGC has issued regulations requiring that applicents for Ots
, submit information conceming their ' organizational stmeture and .l i
operating personnel qualifications. . '
.This information is reviewed in detail by the NRC and must be $
found acceptable prior to issuance of an OL. '
.m. .
+ The applicable criteria are contained in Regulatory Guide 1.8,
" Personnel Selection and Training," and in the knerican hational t l
Standards Institute's N18.1-1971, " Selection and Training of #='
I Nuclear Power Plant Personnel."- l The standard establishes the minimum qualifications and training for !
all functional-levels of the operating organization, ~ including '
managers, supervisors, professional-technical, and cperators- I technicians-repairmen, .
i
'Ihese requirements are designed to provide assurance that the. ', '
nuclear power plant operating personnel will: -
-- be capable.of safely and efficiently operating the facility;
- . understand the complexities of the plant design;
-- be capable of properly manipulating the plant controls;
-- will maintain and repair the plant equipment in an acceptable manner.
rs Before any operating personnel can manipulate the controls of any 6 operating nuclear power plant, they must obtain a license from the .
NRC which authorizes them to operate that~ specific plant. Aa applicant for an operator's license will be approved if the bRC finds that the' individual is in good health, and has passed a-
- written examination and operating test to detemine that he or she has learned to operate a specific plant.
The scope of the senior operator written examination covers 4 21 different aspects of reactor' operation, including:
l
\
-- fundamentals of reactor theory; '
-- general design features;
-- general operating characteristics; 1
-- conditions and limitations in the facility license ;
\
f5 e
9 4
0 gw
)
1
'( ' Training and Qualification -- 2 (b . ,
?
-- fuel handling facilities. l 1
. In. addition, the operators of a facility are required to demonstrate -
an understanding of'12 separate matters, including: a'
- -- the' required manipulation of the console controls; as
-- the.use and function of the facility's radiation monitoring systems; l
-- the emergency plans for the facility.
s=
In summary, the operating personnel of nuclear power plants are l carefully selected, intensively trained in a broad range of nuclear '
power plant operations, and carefully retested periodically by the
. NRC to determine that they can safely operate the facility for which -E they are licensed.
3= 1 1
+
e 7aNs'a
' .*IIIi
4
,._.; q Ya n
~:5.E i
A 5
e I
- -r,.---.- e ,,, , - ,,, m
- -. --- , ,.m- y e e-. - m * ---y 'w
- p' 1
J 1
L
, U.T.*l J j
. ' RESOLUTION OF GENERIC SAFETY ISSUES ,
((# *= ::. e :
A number of' generic safety issues have emerged.during the licensing J.s '
review of nuclear power plants. These issues have nomally been :
resolved on a case-by-case basis. "" '
Resolution of these issues on a generic. basis, however, could !
'. hela to.streanline the licensing process by providing for = ;
staaility, uniformity, and predictability. l
'Over the years, a backlog of several hundred generic issues 'has )
accumulated. :In May.the Comission implemented a program to !
seek timely resolution of generic issues. The main elements of '
the progra~ are as follows-Establishment of unifom criteria for grouping technical E !
activities on generic issues into categories. Using these criteria, the issues are to be placed into fou+-
. categories, with Category A containing the most important and Category D containing the issues requiring no regulatory attention.
l 1
-- Establishment of a Technical Activities Steering Committee i to assign activities to the priority categories, assign lead responsibilities, approve action plans for resolution ;
of the issues, and regularly review progress on tasks -
, assigned. j i
-- Assignment of task managers and their functions. ;
T
-- Scheduling of technical activities. !
i
-- Documentation of the final resolution and disposition E i of each technical activity and publication of a MREG l report on the highest priority issues as they are !
resolved. i As of June 23, 1977, 34 Category A items had been identified, }
Category B items are to be identified by July 11,197< . t All category assignments are to be co;apleted by the end of 1977. )
1 Examples of-important current generic issues follow.
{
h i
?
y l t
6 g.. ,
i
!!i.
OURPRESSURIZATION b
(F "
b Reactor vessels are fabricated out of carbon steels. These carbon F
' steels exhibit great strength at the high temperatures and pressures experienced during nomal reactor operation, but they do become ,
brittle at lower temperatures. =
A measure called the nil ductility transition temperature (NDT).
describes the point at which embrittlement of the reactor vessel begins to occur. Neutron irradiation of the s; eel causes a rise in the NDT.
This is of particular concern in reactor operations since " aging" of the reactor vessel through continued use subjects it to irradiation and increases ;
its NDT. . Thus, reactor vessel strength decreases in the lower temperature range, increasing the potential for problems during shutdown conditions. ,
Most of the incidents have occurred during shutdown.
-Temperature-pressure limits are established for each reactor, specifying the temperature at which a reactor vessel must be E prior to exceeding a specific pressure. Neutron irradiation ~
shifts the limits, requiring an increased temperature for the corresponding pressure. Thus, as the reactor vessel ages, -
p greater protection must be afforded to prevent the occurrence J' E of pressurization greater than than specified for a particular [
temperature.
E t
The Operating Technical Specifications for plants contain the f.s; temperature pressure limits for each plant, and licensees are '
required to report to NRC if these limits are exceeded during operation. -
Overpressurization events can occur due to operator error, such as inadvertent starting of high pressure injection pumps, or an equipment failure, such as a malfunction causing the dumping of safety injection tanks.
1 As overpressuri:ation events have been reported to the .NRC, the staff has required the licensees, as appropriate, to make specific remedial changes, either in administrative procedures or in hardware, to provide additional protection against the recurrence of overpressurization events.
As overpressurization incidents continued to occur, the staff realized that the probica required consideration on a generic basis as a Category A item to determine what measures ure required, beyond those being implemented on a case-by-case basis. ,
In August 1976, the staff wrote to each Ph2 licensee requesting specific actions to be taken to pe manently and effectively ^
reduce the likelihood of future pressure-transient events.
Among the short-tem measures being implemented are the following:
l l
l
- i e
I
I g; Overpresrurization - 2 $
minimize time while at a water-solid condition; 6
-- upgrade existing procedures to ensure that appropriata F
- warnings and cautions are included to alert the operator to the potential for overpressurization during certain plant evolutions.
-- removal of power from the motor operators of high-pressure injection valves while shut down, and while below nomal operating te:::perature and pressure;
-- removal of power from pressurizer heaters while water solid;
-- additional training sessions to increase the awareness of operators to the potential for overpressurization; pressure alam to alert the operator if system pressure ,
approaches Appendix G limits. ~ '
Tne current staff schedule calls for completing the evaluation of licensee-proposed long-tem solutions, including design modifications, by the end of 1977. Priority attention is
( being given to the older facilities and those facilities which have experienced the highest frequency of overpressurization -
events. An example of a proposed rodification is the use of dual setpoints on the power operated relief valve on the pressurizer; the lower setpoint is selected by the operator whenever system temperature ar. pressure are lowered below 275'F and 500 psig during plant outages. =
The schedule for implementation of long-tem solutions will be detemined at the end of 1977. ,
s 3
M
!= j i
r l
[l l
)
1
.cm.,
n.m
( PRESSURIZED WATER RFACTOR (PNR) STFR4 GENERATOR TUEE INTEGRITY [li
.G.~ rii
[:; l
~
In a PWR, the steam generator is a heat exchanger which serves >
as the heat transfer interface between the primary system _
(approximately 2000 psi, 600*F) and the turbine-driving cycle != * ,
fluid on the secondary side (approximately 1000 psi, 550 F). ,E, Tne primary system water passes through thousands of tubes . .
(either in a U-shaped or straight-through configuration) to 0; heat the secondary-side water on the shell side. '
i ifnen weaknesses (wall thinning,' distortion, etc..) are identified !
in some of the tubes, these tubes can be plugged to avvid leakage and accidental rupture, and hence potential offsite consequences, a
Tne steam generators have been designed with an excess of tubes ;
to allow some plugging without loss of efficiency. However, '
a point is teached where significant tube losses can cause a reduction in the electrical output of the unit. . In additica, those p who. inspect and repair the tubes may incur significant radiation exposure. :
i Steam tube degradation can occur through wastage, stress corrosion cracking, reduction in tube diameter (denting), and tube support
- inplane expansion and cracking. Westinghouse plants which have experienced one or all of the above failure mechanisms are Surry, !
Turkey Point, Indian Point, and San Onofre. Babcock and Wilcox's
- t Oconee steam generators have experienced at least one tube leak i
@ incident in each unit, and Combustion Engineering has experienced , ;
tube wastage in the Palisades unit. i NRC analyzes the potential accident following rupture of a steam f generator tube during full power operation. However, other i accident scenarios, such as LOCA followed by postulated tube '
rupture, have not been analyzed since staff has maintained that t -
the secondary side integrity is not breached during LOCA. I e
Steam generator tube integrity has been evaluated as a l Category A generic item, and NRC has-initiated both direct i corrective actions and appropriate analyses as follows: I
-- establishment of inservice inspection requirements I
-- steam generator rep hcement ?
-- establishment of secondary s Me coolant chemistry :
control requirements .
-- review of supportive data and establishment of plugging l criteria for various foms of tube degradation. ,
-- evaluation of the tube failure mechanisms and failure probabilities. . j
-- evaluation of tube failure consequences under accident conditions. -
l
{::
s
r i
g WATER HABER g :
l h'ater hammer is a phenomenon caused by a sudden surge of water :
through a pipeline that is empty 6r has been voided of its water ;
for some reason. Tne surge of water causes tremendous stress, ;
or " water ha=er," as 'it strikes the bends in the piping. >
Kater hammer is included in the Category A lis+. of generic issues i
for resolution.
Over the past several years, dozens of water hamer events p" t
have occurred in EEs and.P33, causing significant damage to pipes, valves, and pipe -supports for' safety-related systems, ;
including the Low Pressure Coolant Injection, High Pressure Coolant Injection, Core Spray, Containment Spray, Auxiliary 1 Feedwater, and Residual Heat Renoval systems.
- 3. '
The most dramatic event occurred at Indian Point 2 on E i November Ib,1973, when a water hamer cracked an 18" diameter , :
feedwater pipe, causing the blowdown of a stea:a generator through ;
the two square inch crack area. The most recent event occurred at the D.C. Cook plant on March 10, 1977, khen the same type of ,
water ha=er cracked an auxiliary feedwater line.
(
On March 10, 1977 the staff established a Water Hamer
^ Review Group to. consider this generic issue. The Review Group '
will develop a comprehensive position paper which will include: -
i
-- a review of operating experience and analytic investigations to date; ,
-- identification of the safety significance of water ham.er phenomena in nuclear power plants; and ' '
- . a su=ary of the current regulatory position regarding water hammer phenomena for CP and OL reviews as well as reviews of operating plants. i,7 i
i~
After the initial position paper is completed, the group will -
develop review procedures and positions for use by the hRC in dealing with the water hamer phenomenon during licensing proceedings and reviews of operating plants, s,
t t
-p
- i i
]
g D.C.: BATTERY FAILURE O
e .
D.c. power supply systems at nuclear power plants consist of banks of cells linked together to fom the batteries, and other equipment, such as the battery chargers, etc.
Tne systems are installed for a number of purposes, including '
the provision of d.c. power during emergency situations te power logic and valve controls for starting emergency diesels and power to other safety loads.
Elbert p. Epler, a consultant on instrumentation and controls to the Advisory Comittee on Reactor Safeguards, has contended that the d.c. power supply system is a safety flaw because the reliability of the batteries is questionable. He contends that the failure rate identified by Licensee Event Reports is sufficient to warrant design changes, as the loss of a battery system would result in a reactor trip. Following this reactor trip, loss of offsite power could cause reliance on the redundant battery system for all plant safety functions.
The NRC staff has studied Mr. Epler's contentico5, and acknowledges that the number of identified events requires a one-year study, but the' staff does not believe that they are of sufficient significance to require immediate regulatory action, for the i following reasons.
- 1. The large number of failures are not necessarily complete battery system failures. Licensees are required to report on individual cell failures, and failure of cells in a battery does not necessarily disable it since it is composed of many such cells.
NRC places stringent surveillance requirements on the battery systems, including weekly checks of such items as specific gravity, temperature, voltage, and electrolyte content of the batteries' pilot cells. Each cell of the battery is checked quarterly for electrolyte, specific gravity, and cell voltage.
Furthermore, battery conditions are monitored in the plant control room such that plant operators are alerted before either of the D.C. battery systems are degraded.
- 2. The loss of one battery does not necessarily force reliance on the other battery unless the loss causes reactor trip and offsite power is lost. The loss of a battery leading to reactor trip '
occurs in approximately 50 percent of operating uni'ts.
- 3. KASH-1400 (the Reactor Safety Study) quotes the probability that the reactor trip would cause loss of offsite power as one in a thousand.
1 g
{w I
. I
. -- .. -. _ ~ . - - - ~ -- -
I
. .i
+
( (i. D.C. Battery - 2 5
- 4. If the reactor trip causes a' loss 'of offsite power, the redundant t
j
' battery. is then needed for a period of approximately 30 seconds '
in order to start the emergency diesel which is the onsite a.c.
power source. Any d.c. power then required would be supplied by l rectified a.c. from the emergency diesels. l y 1
.5. If there were a reactor trip, loss of offsite power, and N= i the redundant battery also failed, so that neither battery could . l start the emergency diesel, the plant operators would have at- !
least an hour to take 'the actions necessary to prevent a core ;
melt. The probability cf the sequential problems occurring is :
very low, and the abilit'/ of the plant operators to take the appropriate corrective actions is high.
l
- 6. Mr. Epler states that there are parasitic loads on the f batteries, and that these loads should be removed. In the newer !
reactor designs, other battery systems now handle these parasitic '
loads. In addition, there are new reactor designs in which l
there are two safety-related batteries incorporated for each ,
1 redundant' division.
- 7. For .the foregoing reasons, the staff has initiated a one-year study to decemine the extent of the problem and what {
o'~ ..:. j regulatory action is required, if any.
9::.t::.. ;
W" i I
~
e I
t f
i I
i (k
I
. ~ . . .
++h. e w g e. m m. g +Lq~+-wq- m.er pw,-A4
- v. .
ii:: $
L p.
' ... INSPECTION AND ENFORCDIENT ACTIVITIES TO DIPROVE REACTOR SAFETY k ;
- (;i?:;p 5
% g:;; .,
NRC's Office of Inspection and Enforcement has taken a number of ., .-
steps to upgrade the scope and effectiveness of its activities. *
- 1. In consonance with President Carter's request, the Commissioners 3:
have approved a program to place resident. inspectors at power ;
reactor sites during both. construction and operation. Tne Office of I E E conducted a Trial Near-Site Inspection Program,. completed ,
in fall of 1976,- and the results of this trial program led to development of the resident inspector program which is planned to be fully implemented by the end of IY- 1981.
The Office of I 6 E is expanding.its vendor inspection 2.
program on a trial basis, to test the feasibility of utili-ing
~
American Society of Mechanical Engineers (AS'E) code inspectors e as part of a third-party inspection system supplementing the h direct NRC inspection programs. If the trial program proves E <
feasible, the NRC will-be able to expand its inspection of the Quality Assurance programs of licensee contracters and . vendors who supply nuclear plant components and systems. ,
- 3. A team of operations analysts, under the direct supervision - '
of the Direccor of I 6 E and in close coordination with the j.') I 6 E line staff, is undertaking a systematic study. of the.
inspection and enforcement program in such areas as: b
-- review of incentives system; l
-- procedures utilized by other agencies with I 6 E programs related to safety missions (both in U.S. !
and abroad); -
I
-- allocation'of inspection resources according to licensee perfornance; j
1
-- techniques for assessment of licensee perfomance;- l
-- confirmatory and independent f.easurements program. )
- 4. I 6 E is developing a program to assure that licensees l and affected vendors and contractors have implemented procedures for the reporting of defects and incidents, :.s required under NRC's new : regulations regarding abnormal occurrences, w s
p SYSTDRTIC EVALUATION PROGRAM FOR OPERATING REACTORS The purposes of the systematic evaluation program for coerating reactors are:
- 1
-- to identify and evaluate the significance of disparities between actual design and the criteria presently applied i l
-- to detemine the need for backfitting in accordance with NRC regulations; j
-- to document fully the results of the review, and thus in conjunction with procedures for implementing new standards, to lessen the potential for needing future re-reviews.
The program approach is as follows:
-- senior staff will co= pile a comprehensive list of generic issues -
from all sources and _ systematically eliminate issues of lesser J' -
safety significance;
-- a review team will establish an appropriate issue list for each reactor; L -- the licensee will be asked to analyze the significance of the issues identified for each reactor; .
-- the review team will evaluate the licens.ee analysis, document their findings, and make appropriate recommendations for change.
The systematic evaluation program ms approved in general fom by the Comission in January 1977. The Commission directed the staff to carry out the compilation and culling of issues phase and then to brief the Commission before proceeding to the plant-specific phase.
p Resources: p
-- Staff: FY 77,10 man-years, $0.6M; FY 78, 35 man-years, $1.4M; FY 79, 35 man-years, $1.8M. "
-- Industry: $2 to $4 million per plant over the three-year period.~
t l
l l
l l
. l lE NRC RESEARCH PROGRMS
~
E Light h'ater Reactor Safety Research Tne bulk of the LhR research program is concerned with a postulated loss of coolant accident (LCCA), generally assumed to result from a ,
pipe break in the reactor inlet or outlet water lines. Tne Loss of Fluid Test (LOFT) facility is designed to study the behavior of engineered safety features, such as emergency core cooling "I
systems (ECCS), during postulated reactor accident conditions.
-- LOFT is a well instrumented pressurized water reactor (PhR) at about 1/60 the power level and'1/5 the physical size -
of a cor=ercial PhR. Located at Idaho National Engineering Laboratory.
-- Data from LOFT experiments are being used to validate and improve the analytical methods currently used to predict the course of a LOCA. ,
-- The first four nonnuclear LOCA experiments have been successfully run and reported in det' ail.
-- Project is now three weeks ahead of schedule and within budget. ,
-- Nuclear core load projected for fall 1977.
Other 2..:ents -f the thR research program focus on: .
-- Primary system integrity: pipe and vessel flaw detection, failure mechanisms.
p
-- Physical damage to fuel rods resulting from a variety '
F -
of accident conditions. '
-- Semiscale testing on a non-nuclear facility for reactor design computer code testing.
-- Various efforts to study fundamental aspects of reactor phenomena.
0 (Backup information contains descriptions of the most important Lh2 research facilities and projects.) -
Advanced Reactor Safety Research
- 1. Liquid Metal Fast Breeder Reactor Program -- Oak Ridge National Laboratory and Sandia. President Carter asked Congress for $33 million for E8DA to continue research, but both House and Senate have
+
w
r h2C Research Programs - 2 O
taken action to fund ERDA demonstration plant and to provide NRC with funds for regulatory research, so long as application is not withdrawn or construction halted.
- 2. High Temperature Gas-Cooled Reactor (HTGR) - presently a small program effort awaiting industry-ERDA plans.
Other Research
- 1. Risk assessment
- 2. Health
- 3. Environment
- 4. Site safety
- 5. Transportation -
6, Waste Disposal
- 7. Safeguards L..
O t
e
)
b i
(
t I
t
- o. ! .
h t?:.
I NUCLEAR REGULATORY RESEARCH ;
f".
LOFT (Loss of Fluid Test) DATA ASOR m u.un l PURPOSE:
The tort facility is a een instrumented p eisurtzed water reactor (pWa)
- at about 1/60 the power level and 1/$ the physical site of a ceWrC141 3i FW7.. It is 1ocated at the Idaho hational Engineering Laboratory and is " s designed to study the behavior of engineered safety features. such as seergency core cooltag systea.s. during postulated reactor accicer.t conditions. Data from LOFT experiments are being used to validate and ieprove the analytical rethods currently used to predict the course of a loss of coolant accident, i STATUS:
- 1. The first four scheduled nonnuclear loss of coolant experiants have !
been successfully run and reported in detail. -
- 2. The esperieental predictions report was issued prior to running the fourth scheduled experiment and the esperfrient was run success.
fully on May 3.1977. three weeks ahead of schedule. The analysis ,
of this esperiment has been designated as a ' standard problee.'
for which any foreign and derestic researchers will calculate F e prediction of the results. Data from the esperirent have been ,
withheld while participants in the
- standard problem
- program use the initial experimental conditions to predict the outcome. When the predictions are submitted for study, by June 20. the results will be puBWhed through the norma 1 Charmels.
- 3. The newly repaired quick. opening valves functioned properly i
{
during the L14 experiment. .
4 The project is now three weeks ahead of schedule and within the budget established in October 1976. !
- 5. By reprograming some work and reserve funds the schedule for tasks preliminary to core load (mid October 1977) is being accelerated in an effort to comp 1ete this milestone in !
ear 1r September. ?
- 6. Procedures have been defined and are being 1.placented for the turnover from (A0A construction to Ntc operation for the LOFT i
(" facility.
4 PROBLEMS: '
Many construction item remain to be completed and considerable inspection work is yet to be done prior to core load and the first criticality tests. .
Contingencies have been allowed for potential problees and several are !
undet evaluation. :
SEMISCALE '
s PURPOSE:
If{i The Semiscale Nd.1 configuration is a one.dleensional, nonnuclear representation of the thermal hydraulic aspects of a pressurized water ,
reactor, de*1gned to strulate tcC$ perforrance at a small scale, it ,
is about 1/30 the power level and 1/5 the physical size of LOFT. with !
one operating steam generator loop and a second loos of strulated i c o*00nen ts . The No.l.1.6 segawatt core is a 5.$. foot long. 7 inch dismater asse@1y of electrically. heated fuel pins, j I
The purpose of the test facility is twofold: a) to provide separate and loss integral of coolant effects experirental accident data methodss computational for developing)and verifying and c to provide data l for optimizing selection of LOTT test parameters, assess reliability 6 of LOFT instrumentation. evaluate L0fT test results, and address LOTT design comprbmises. ;
STATUS: 4 The schedule for the Semiscale Mod 212. foot long core has been altered. . i' The Mod.2 configuration consists of a 12. foot 8cag core and active operating and broken steam generating loops. The upper plenue/ upper
- head is typical of conventional PWR nuclear steam supply systems. l t
At the reovest of NRR the design effort has been shifted to accalerate ,
the ava11 ability of the Mod.3 configuration in order to cbtain i experimental information on the performece of the Westinghouse l vpper head injection (CCS concept. Additional ridyear funding has :
allowed an acce1erated cor@letion date for the Mod 3 configuration from 1 the originally scheduled date of April 1978 to January 1978. The Mod.3 l configuration has an upper head to sirulate the bestinghouse Upper Need i injection for emergency core cooling. Simulated steam generator tube ,
t
.. j rupture tests are under way. l h PROBLEMS 1 hone. l j
I
! 52 ,
l l
I l a
c !
NUCLEAR REGULATORY RESEARCH WAASR m.m p"_
p=
PBF(P0wer Burst facility)
PURPOSE:
The Power Burst fact 11ty (P87) consists of a pressurlied tes loop in
- a pool type 30 regawatt (steady state) reactor capable of pulsed operation. It is located at the Idaho hattonal Engineering Lateratory ~
and is used to obtain data on the performace r! fuel rod clusters r under abnoral power. coolant flow and ener gy density conditions. . . .
t C'
This includes the PCM (Power Cooling Misatch) test. LOCA (Loss of Coolant Accident) tests. Rf A (Asactivity Initiated Accident) tests and inlet flow blockage tests, The data obtained will be used to develop or confirm analytical caoability which can be used to predict fuel f response in reactors daring off-norcal operating conditions.
STATUS:
Twelve programatte tests are planned for FY 77 including four pfgqytack tests. These include (a) fuel rod tests unser power-cooling mismatch 5
conditions to observe the post CNF behavfor of irradiated and unirradiated l fuel rodss (b) tests under LOCA blowdown conditions; and (c) gap conductance ressureawnt tests to provide needed information on fuel pellet stored energy.
The first ten P8F tests sche 6aled for FY 77 have been completed. Four were power cooling mismatch tests (three with pre-teradf ated rods) and sis were gap conductance tests (three of which were piggyback tests j with pre leradiated rods). ,
The first LOCA blowdown test is scheduled for late FY 77 and the first R!A test is scheduled for early in FY 78. !
k Preparations for planned reactor modiffcations will require a reactor l downtime of approsimatel/ four months, ..
i PROBLEMS '
RIS has issued a directive terminating the design and procuremnt of a PBF reload core because of technical pechte s limiting the perfomance f and increased costs. R!$ is presently exactning alternative cations for meeting the fuel red testing needs that the reload core was intended !
to provide. '
l The [G4G management of P8F has been reorganited. The new manage *ent team #
has been emphastains the accomplishment of hign priorf ty test cbjectives.
Reviews of all of the ongoing plant improvemnt projects and more j cogrehensive planning and scheduling for them have resulted in delays. . y However. It is expected that this effort will lead to improved management ;
ard better eSst control.
The raterial deliveries on the LOCA Mod project have been slower than anticipated;, therefore. the contractor plans to split the construction work into two phases with a testing period in between. This should >
allow deliveries to catch up with the construction progress and should still allow LCCA Mod completfon in time to rett test cocrtitments. t 1
ECCBYPASS TESI FACll.lTY '
PURPOSE:
Tne ICC Bypass Test facility is intended to provide data for evaluating '
the potential for (CC bypass and the type of two phase flow phencrena i which night occur daring the latter stages cf blowdown and during the early portions of reflood, following a postulated loss of coolant i atcicer.t in a reactor. rata from this program is expected to help establish when end of bvDass would occur.
STATUS:
)
Conceptual design effort for a 1/3 scale trergeacy Core Cooling Bypass Test Facility (t8TF) utillatng PG&f's Contra Costa Power Plant for stea9 has been terminated. A conceptual desfgn report was issued by
- Ato=les International in April 1977. A total progree estimated cost of $102 M was projected ($7$ M for design and construction and $27 M ,
i for esperleentation), along with a start-of testing estirated to be January 1983, assuming an autherf red program in FY ';/8. !
i The high estirated cost. Curtaf1 rent of available steam from PGit. '
uncertainties in funding authoritations. apparently small benefits -
to be gained from (CC bypass research versus other considerations such as decay heat and 2r-water rate reactions have led to the reco rendation to carry out a final conceptual design for a " nominal 1/5 scale' ESTF by Battelle-Columbus Laboratories (BCL). Current efforts are being directed at studying various options (including the purchase or lease !
of additional boilers) to provide a basis fcr specifying facility ,
requirements for the BCL final conceptual design effort, v p PROBLEMS: 5 g hone. t 5-3 :,
i
)
em h NUCLEAR REGULATORY RESEARCH DATAASE i
h PRESSURIZED WATER REACTOR BLOWDOWN HEAT TRANUER PROGRAM F i
~
PURPOSE: :(PWR-BDHT) j The 81ewtown Heat Transfer progran at the Oak Ridge National Laboratory i is designes to determine the tire to critical heat flua (CMT) at the !
fuel. coolant interface and heat transfer rates daring pee. and post CHF regiees of the blowdown (rapid depressuriaation and coolant less) phase
- of a loss of coolant accident as influenced by variations in evactor power. system pressure, coolant flow, and pipe break locations.
STATUS Fif teen blowdown tests. simulatteg a doable ended pressurized water reactor (NR) pipe break, have been co pleted using a 49 fuel rod ,
electrically heated test bundle. The full length bundle simulates a portion of a PW4 nuclear core. Eight of the tests wee performed with l four rods unpowered to study the irfluence of unhestt: FWR rods (such ,
i as a control rod thimole which contale no uraniur) e the time to CHF and the te perature rise during blowdown. Two failed rods are being used in unpowered positions.
PROBLEMS: ,.
Four large (2.$ We) generators, which are shared by this program and E j i
EADA's fusion program at CANL. need rebuildirg. Discussions are being .
carried out with the ER A Civision of Magnetic Fusion Energy regarding I the means for funding the costs of reiviiding the generators (ssso x). . ,
HEAVY SECTION STEELTECHNOLOGY PROGRAM (HSST) .
t PURPOSE-E ;
n ,
The purpose of the H55T program is to develop and evaluate methods of If '
fracture analysis of heavy sectiod steel itsht water reactor pressure b vessels for inclusion in codes and standards. Thermal shoct hydraulic and preu-atic tests of model reactor pressure vessels are carried out at the Oak Ridge hational Laboratory.
STATUS:
Mine tests on deliberately flawed vessels have been performed, pressures from two to three tieus design pressure have been retaired to cause vessel failure.
b June 1976, a vessel with a large flaw was retested using oneustic pressure to determine whether a dif ference in ef fects would occur between this test and a previous one conducted with hydraulic pressure.
ho dif fereoce in behavior nas observed. The results have been forwarded to NR4.
Two more vessel tests are planned foa lats TY 77. Each test vessel will be flawed in.a weld repair region. Ont will have a flaw estendin 90t through the wall and be tested at ductile shelf temperature the (g nearly temperature region corresponding to unleus toughness of the steel) while the other will have a flaw nearly halfmey through the wall and be tested in the temperature regtes corresponding to the transition between r.tnimum and NIirus toughness. These tasts will provide valuable information on the A911 Code-approved sothed of weld repair.
Four therm) shock experiments on test cylinders have been performed to sleulate effects of cold (CC$ water injections into hot steel pressure vessels. The fourth test was perforced on a metallurgically degraded cylinder with a long axial fisw on the inner surface. Rapid pecDagation of a crack initiated from this flaw at the predicted '
time and then arrested as predicted. Thermal shock testing will be i completed during FY 77, i An irradiation program on heavy section low :helf weld utal test l spec 1riens is underway to quantify the potential toughness loss caused by )
eestron irradiation damage. The irradiation phase was completed in F4rch .
1977 and a qualifiestion program is underway for the testing method to I be used to evaluate high levels of fracture toughness in fracture mechanics te m . from se ll specimens.
PROBLEMS:
war, prestresstag which has been postulated as a e.ans of mitigating tae effects of enerui shock, my not be feasioie en criineers that would be therN11y shocked. This problem has been reviewed with the french who are esperirentf ag with tRe use of li4did nitrogen for thermal shock generation. Analytical work is sont1 Ruing on this test at OM.
with esserleents usirg liquid nitro;en and vessel coatings to change the heat tra9sfer characttristics, FJterial properties are being evaluated to detemine the appropriate L
,. te9peratJre to be used in the experiments for the transition temperature b p test of a weld + repaired interMdiate test vessel. We now foreste a F' i
-l metale-f.,1 test following more careful material property characterization. '
]
l 54 e a, f..
2** !
l 4
e I
L..
.:C' NUCLEAR REGULATORY RESEARCH DATA AS OF: aunt is. is77 p, PRESSURE SUPPRESSION TEST & ANAL.YSIS R a
PURPOSE: .
The pressure Suppression Test and Analysis progra- is intended to provice experimental data on the Bdt Mark.1 Tocus Suppression Charoer for I analytical sedel verification and licensing. A three-dicansional. I 1
quarter-segment.1/5 scale model of a toroidal pressure suppression chamber has been constructed at the Lawrence Livermore Laboratory.
principally for studying vertical loads and other parareters on the torcidal wetwell STATUS
- m 'l The program is designed to study effects of air venting into the bi toroidal wetwell. variations in drywell overpressure, drywell pressurisation E rate, downcomer submeegence. vent pipe flow assy:netry and enthalpy flux.
The air venting test phase has been completed. Four out of seven issues of the Quick Look Reports have been sent to R34. The reminder are f expected to be received by June 24. The steam venting phase of the test program is being reevaluated.
PROBLEMS: ,
hone.
ACPR AMULAR CORE P'RSE REACTOR UPGRADE PURPOSE:
The purpose of this program is to upgrade the Annular Core Pulse Reactor (A0pR) at landia Laboratories by installation of a new high heat capacity core and an improved control syste= and cooling syste?. The pulse h c
fluence and steady state power will be increased by a factor of 2.6.
A fission gama-ray coded apperture fuel notion diagnostics system is to be developed and installed. The upgrade will tricrease the capability of the ACpR to perform advanced rea: tor safety experiments on pro 7% burst energetics and post accident heat re oval. The A084 upg*ade is a joint peoject with (ROA/D"A which is furnishing eagineering support, while ,
NRO/RSR is purchasing the new core and appurtenances and an essSciated fuel.netion diagnostics system, t STATIJ$ t i
l The ACpR upgrade is currently on schedule aed within budget and the design performance esceeds specifications. The $4fety Analysis Report for the upgraded AOPR has been submitted to htR for review and ccTent at [RDA's re'aatst.
Fabrication of the fuel, mechanical components and the new control console are proceeding on schedule.
Shutdown for installation of the upgrade core is scheduled for September 30,1977.
PROBLEMS ho problems currently exist. A pot.entially serious pecblem exists in delay in approval of the Safety Analysis Report which would produce a schedule slippage on a weet fr week basis, i l
m
[
t 55 a.. .
i p:
(:
N RISK ASSESSMENT IN THE REGULATORY PROCESS i~
E 5
The following is a chronology of the continuing evolution of risk' assessment and its application to nuclear safety.
-- 1957: publication of NASH-740, a quantitative analysis of potential consequences of nuclear reactor accidents.
Accicbat probabilities (i.e. risks) were addressed -
only qualitatively, as WASH-740 was essentially meant to be the technical basis for the trovisions of the -
Price-Anderson Act. (ORNL-3441, pu31ished in 1964 had taken a similar approach to certain non-reactor facilities.)
-- 1965: by means of a letter to Congress, the AEC Chaiman, reported the results of a review of NASH-740 in light of developments since 1957. The basic conclusion was that, while the most serious consequences of an accident were clearly greater than those estimated in WASH-740, the probabilities of such accidents were lower.
-- 1972: publication of 1GSH-1400, the Reactor Safety "
Study (co=only known as the Rasmussen Report) in draft ;
fom for public coment, after review by an AEC-regulatory -
task force and a few individuals outside the AEC.
-- 1975: publication of the final version of WASH-1400. '
The final version included responses to coments from 120 individuals and organizations, including the American Physical Society, the EPA, and the AEC regulatory staff (which had by then become the nucleus of the present NRC) .
-- 1976: the Udall Subco::nittee on Energy and the Environment ,
held hearings on WASH-1400. Although there was general consensus that the Reactor Safety Study was an important i advance in the development of risk assessment, nevertheless there remained a continuing controversy over the context of -
its development and release.
-- 1977: following a continued exchange of correspondence between Chaiman Udall and Chaiman Rowden, the NRC is now -
in the process of establishing an independent cc::nittee of technical experts to review peer coments on WASH-1400 '
and on developments in risk assessment so as to recommand future improvements, r
In addition, the Co:inission has emphasized that the agency focus on the development of risk assessment and its appropriate use in the regulator / process. It is identified as one of the eleven Agency-Wide Objectives. !
i 6
I
I F ._'IHE REACTOR SAFEIY STUDY -- hUCLEAR RISKS IN ' PERSPECTIVE -
WASH-1400, the..Rasmussen Report" provides probability estimates '
t.hich put nuclear risks in' perspective. '
RISK OF DEATH FROM VARIOUS SOURCES PER YEAR FOR AN INDIVIDUAL !
-- autos, 1 in 4,000 !
-- falls, 1 in 10,000 '
-- fires,,1 in 25,000 - i q
-- plane crash, 1 in 100,000 _
-- nuclear. accident,1 in 5,000,000,000 (assuming 100 plants) l From these and other statistics, WASH-1400 concluded that the !
risk _ to the general population from nuclear plants are very low, as compared with other low probability, high consequence events l
, j such as dan failures. The following table indicates such a ~ i Corparison; p:. f Annual
- 6 Probability '
of !
- Source Probable ;
Occurrence , Consecuence ,
Nuclear plant 1 in 20,000 l core meltdown No deaths or '
, injuries Dam failure 1 in 30,000 (Based on study 14,000 to 260,000 !
of 10 California up to 1 in 40 fatalities ;
, i dams) * '
t n,
J h
d
~
f
(
I w
(?
,.. 11E BR0hN'S FERRY FIP2 D.
l* ,
On March 22, 1975, a fire occurred at TVA's Brown's' Ferry Nuclear
. ' Plant Units 1 and 2 which resulted in shutdown of both units. -
The fire resulted from ignition of polyurethane foan being used to' seal leaks in cable penetrations between the Unit 1 reactor building and the cable spreading room (which is located beneath the control room shared by both units.
- Substantial difficulties in depressurizing and residual heat
. removal were resolved by operator use of redundant safety equipment, as the ECCS was not available because of cable damage (1600 cables were damaged).
' Implications of the fire:
-- Defense-in-depth approach proved effective in protecting the public;
-- h2C's regulatory program for fire prevention and control and related QA inadeqate;
-- criteria for cable separation probably inadequate;
-- separation of redundant systems insufficient to prevent ,
cc:=on mode failure;
-- emergency planning needed improvement.
Corrective actions taken by NRC: N
-- all operating power reactors were inspected by end of May 1975 with respect to installation of fire stops on electrical cable trays and' penetration seals, and adequacy of licensee's fire prevention and protection controls. ;= I
-- special review group evaluated fire incident and -
submitted detailed recommendations to NRC in February 1976.
-- hRC inspection program expanded to include fire prevention and protection for both construction and operation phases. -
o
-- NRC issued specific guidelines on fire protection p" (Standard Review Plan 9.5.1).
-- On May 11,'1976 liunsees were requested to describe steps being taken to imple.nent new guidelines.
. q ..
t :.
O I
, 6*** ***
l
,q. .4:
.3 !... i e
p Brown's Ferry Fire.- 2 .
m+
On September 30, 1976, licensees were requested to provide a fire hazards analysis and technical specifications pertaining to fire protection ~. ,
- .All licensee responses expected to be received by end of~ July 1977.t :-
-- Fire protection programs- for all operating plants being reviewed and safety evaluation reports bei.ng issued; completion expected l by Decerher 31, 1978. ;
-- After' investigations, hearings, restoration, and operational testing, Brown's Ferry Units 1 and 2 were authorized to resume operation on August 20, 1976. _ :
i L
b e
- D au h L .1 '
[4i
[:.
}i: .i i::: [
i:E :
, { ::
y he. !
. e i
- s. i i
- $, 1
?'
b::
I 1 d i
g c. ;
, " l- h
. t
, ./
i;: i.
y .
^
t... 9
\ce .
'l
~ . . . . . _ . . . ~ . _ - , . , . . . . . . , _
I
- 5
' l GEED - CURRENT STA'IUS
$ 1 1
' f[4 e
On April 7th, 1977, President Carter announced that reprocessing - ; !?
in the United States should be deferred indefinitely due to concerns-about the international proliferation of nuclear weapons. .,,
- The President also decided against providing any funds to assist f.! ,
construction of a reprocessing facility at Barnwell, S.C. [
[li! .
In early May, the Comission asked for:
J '-- public comments on the President's April 7th nuclear '
policy. statement ,
-- the President's views, via letter, on the relationship of t his nuclear policy statement to the GESSD proceeding.
Public comments have been received.in response to the Federal I Register notice. The nuclear industry and the utilities generally '
, favor continuing the GESMO proceeding; the environmental groups i favor haltfag it. There has been no reply as yet from the President, although a reply may be forthcoming shortly.
Pending completion of the Comission's assessment of 'the l appropriate future course of the GES)D proceeding: ;
-- the GESMO hearing panel has postponed further hearings :
on'H., S 6.E; i
-- the Draft Safeguards Supplement will not be published !
for public review and coment.
It should be noted that the public has had close to complete l access to all. documents involved in the GES5D proceeding.
Executive Order 11652 prohibits release of classified information, f but in the case of such classified documents, IGC has devised ;
procedures to permit appropriate access by GESMO participants. n ;
1 The only documents withheld are those containing proprietary ."
info mation, and predecisional documents, the release of which . !
would compromise the effectivoness of the agency's internal '
decisionmaking process.
ki .I, l
na E
+ ,
-i ,
. i i.1 .;
?
i
( GE9 D - HISTORY- t (G. ;-
it .
c.: .
In August 1974 the Atomic Energy Commission issued the Draft '"
Environmental Statement (GESD) in' connection with the proposed i E wide-scale use of. recycled plutonium (mixed-oxide fuel) in light
- utter reactors (LhRs).
q,:; !
Technical signifi6ance of plutonium recycle: ;
=,
f
-- permits. recovery of some of original fuel costs; !
- reduces requirement for uranium. input to fuel cycle; i
t
-- reduces plutonium inventory; [jj !
-- reduces spent. fuel storage at reactor. sites; !
+.
o
---introduces a. potential safeguards problem. i Safeguards became an issue and hearings on GESIO were deferred. !
t HEC reviewed the subject and stated its provisional views in -
May 1975, requesting public corr.ent. i NRC announced in November 1975: '
i
-- Full assessment of safeguards issue needed before !
considering the wide-scale use of mixed-oxide. fuel (M0X) '
in thRs. =n t
- Scope, procedures, schedule for GES'.D proceedings. [
-- Criteria for interim licensing actions. ,
i hRC announced on January 6,1976, detailed rules for public hearings: I
-- On final environmental impact statement on wide-scale l use of SDX (health, safety, and environmental aspects). l
+
-- On the (to-be-developed after EIS) proposed rules for wide-scale use of FDX. l s.
- x ;
-- Separate hearings after publication of the Final j Safeguards Supplement.
i I
J. ~ 1
,.\_..'
- be i'
= ,
jf-
. r- 1
.___m_ . _ _ _ . _ _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ . . _ _ . _
- e em"- 'r-
r
~.
l f
[5 7 1
r
/ GESD History - 2
(;n. .
1 11976 decision of'the Circuit Court of' Appeals:
I 1 :
(
--- Upheld' the NRC November decision on procedures for reaching !e "
k a decision on Pu recycle.
- Disapproved interin licensing of Pu recycle related facilities. t
- g
'-- NRC has taken .the issue of interira licensing to the Supreme
- Court for judicial review. '
U; t
- The 1976 decision also raised issues relating to consideration of Di !
alternative forms of energy generation, both non-nuclear andl nuclear. t
-In response to the Court's dicta, the staff addressed alternative forms of energy generation in the H, S 4 E portion of the GESD -!
proceeding. i f,
I NRC issued a final environmental impact statement (FES) on ;
. health, safety, and environment (H, S, 6 E) on September 1,1976.
The Safeguards Supplement was projected at that time for publication '
in the near future. (See separate discussion of Safeguards Supplement.) ,
The principal staff findings of FES on H, S 4 E are: i I
/ 3 c
-- the safety of reactors and fuel cycle facilities is not :
affected significantly by recycle;
- i i
-- nonradiological environmental impacts from recycle are slightly smaller than those from no recycle;
- plutonium recycle extends uranium resources and reduces !
enrichment requirements, while requiring resources for . ... f reprocessing and fuel fabrication; ~~!
-- despite uncertainties, recycle has a likely economic 1 e
advantage to no recycle; ,
-- no waste canagement considerations were identified that would bar recycle. i Tne GESSD panel reviewed in detail the FES on H, S 6 E and '
developed a substantial record. During the winter of 1976-77, the panel completed questioning of the NRC staff and-received
. testimony from other GESD participants. The panel anticipated .,
completion of its hearings by mid-1977. Mi
- i i
r ,
u .1
. - .- - ..- - - --. - .- . . , . , - _ . . . ~ -
I
!!? .
,w: i f.f, c . n!E GESMO SAFEGUARDS SUPPLEMEST nu >
bNb.
Draft safeguards supplement to GESMO addresses safeguarding of wide-scale use'oi nixed oxide fuel in light water reactors. ... :
p ,
Tne draft safeguards supplement has been in preparation since Novenber 1975. Release-of the draft has been held in abeyance i
pending the results of the Conaission's assessment of the .
future course of GESMO in light of President. Carter's April 7th ;m i nuclear policy statement. '
.+
9
. One possibility is to issue the safeguards supplement as a . 'f?
technical report, in lieu of an EIS, in order to provide the
- public.with the. information developed, while being consistent , ;
i with the? President's. policy statement. .
fE' g The supplement was prepared in response!..to comments on the '
t AEC's original draft GESMO OASH-1327, of August 1974). The Council on Environmental Quality was particularly concerned about the safeguards question. !
Tne supplement addresses the following basic questions: '
-- What would be the potential risks to society from the presence of large quantities of plutonium in ,
cthe commercial sector? .
?E -
5 t
-- Could MOX be'sufficiently well-protected as to i
. assure that the risks to society from malevolent -
acts would be acceptably low? .
-- Given that adequate safeguards could be provided, wuld their economic and other societal impacts be ~!
acceptable? 9!
i General perfomance criteria for safeguards systems: system ;
must be designed to deter, prevent, and detect: $
- unauthorized activities,' entry or introduction t
-l
' of contraband into areas where special nuclear material (S.41) is' located; j
-- unauthori:cd removal of S.4! from process lines or work areas,
[ E l
i l
.i i
k_
i
- 1. ... ;
hr !
i
),1 s, , - ,
e L :
.m
1 s]
Safeguards Supplement - 2 -
.f.
(g .
Pc,
(
~
Preliminary staff findings:
6 impacts of adequate safeguards system for wide-scale use of MOX in LhRs:- l
-- Civil'1iberties: affects relatively few people and '
. represents primarily extensions of existing practices. w; i
-- Relatively. minor potential impacts on existing institutions.
-- Economic: . nominal incremental costs; $150-$200 :million (1976 dollars) .' annually in year 2000 equating to about
$2/ year to, average electric power cus,tomer. Indirect government costs $1.7.nillion (1976 dollars) in year 2000. i m>
3
-- Little legal impact anticipated.
-- A-significant potential risk to society could exist if ,
threats to the MOX fuel cycle industry materialized and -
adequate safeguards were not provided. =
-- Use of M0X fuel would have no significat impact on the potential consequences of sabotage to nuclear reactors.
(
-- Safeguards system.s can be designed to protect a future b MOX fuel industry to an extent that, in the judgement of the NRC staff, reduces the risk of theft and malevolent use of plutonium to an acceptably low level. .
-- The incremental burdens placed on society by the imposition o . such safeguards would consist principally of relatively small increases in MOX industry costs, and in the number of individuals affected by plant security and safeguards
- systems. . ,... . .
s e
i e
- k. ,
. ::l "r.T i?
)
r.' < !
,3
- . i
- o. -
'_{-
, _. t i
$N! !
/z.u
, i' ,4 E!. i n: ,
ts .
( ICOMPARISONOFESTLIATESOFSAVINGSINU0 AND ShU FOR PU RECYCLE k?:
33 E. -
i
- i. U03g SU W
{
~, a
- 1. GESMO
'M* M . !#
t 1975 - 2000 ctenlative 22% 14%i '
507 LhRs in 2000 i $:q i
.30% tails. i
~2. ERDA MODEL Bh2: . LIFE CYCLE i=":
, t.
30 year life 22% 24%
l
.3%~ tails- I time variable capacity ~
factor (average 66%) :
e
- 3. ,'AIF MODEL LhR -
a , t .- -
e
. %.. n 30 year life 17% 21% !
.3% tails . '
70% capacity factor- i:..
E...
t-4 4 i ERDA 76/121 BENEFIT -'
i ANALYSIS OF RECYCLE ' i, .
I 1976 -'2000 ctaulative 25% --
i 500.LhRs in 2000 l r
I
. l t !
ie j a
I q ;
l ;
i F
.. '!