ML19319D886
| ML19319D886 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/13/1973 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-T, NUDOCS 8003270543 | |
| Download: ML19319D886 (100) | |
Text
L m-4~.~<
w
,w + %.,u.~,w,m%.m, e,owww.e a
_,sw a y e, yan.pw
+.A.,snswwasMr.mpw.pe
>.e u-sey Mv ' Mrs.Qhsh. JcMSe Ed-.MO.Gih.4T L MNMMil/$+Ced.MUk.*dW NNi,etr@gw.,awemM'J.!5bvMid y'#F/yn64:
.c
.wa.m..v
- ,,s. soy p
%"N.pa %w.P.WM30rV9M *9M6W)6W??s!a.la W '<' y4"MWr-p*.'#W.$a,.9%um.d W
4 J
.. S:1, u mm,.Msw..w.m...ma.m2:wa,m N,a w,.::.m ww as,:uw.egni.w%,,..s s
lg n; ~.
w-
- ~
.%..c.
..m w
..,W>
9,.-
.s"* r u. w:
r r
- > o.
.s,-
-r m
- x. gg
- m..
- y. xm. :cg" <4.: 2 ; n. n.
c1 v.m w m-4
.q, p -
g
.t.
4,
,e v
. ~,
s
- w.. <
, x-I UNITED STATES ATOMIC ENERGY COMMISSION t
)
)
o cs nc D
D m s r m.f M :
n
. n ir%.._. ?s.n.
%,,A' y. f,..
o rm 7 m n nn r,.
( s. m u
%. M.
.e
.R 4
n w.w.,.
4
-yh"<.(<D);3 lic h.g d a, e
~
].
.w.
w..As
...,1,m u. ;.,..
. ~.
..sm. g
~..
,, m.., c w w..
-u We, kgry.S. g..r '
-:V.w q..a u..
M. :
3:.. -
.n. m g q yy
,.-+
2 m. # -..
g
. %. Q:.e n %f.d.. Ac.T w'3e eM4/d;W+,. w y c. m, rv. _=- W A w.. w.e w w 4, w,.<.- m m e.e
- %,.m:.;o.e..W&.Ms ep.;
. $; M
.-,3 s
t
- w.,.,,.wyy 4 >..,q + 4.,M IN THE MATTER OF:^w
. 3.c <
,,. w s:;...u..w WM.....,~ w,.
s.4 W s.9. ;q y,%qy m
, +.. ~
mMe*h 4% w
...?- N';f'^*. Wi.:,n. Q* Q y g '& g f %,. m&.,? g,.*@r?-.R
.{ Q. y.,% %; e, W
()
%y r
a
-a
,,~..,..m,7
, j.,&.t.q.p&%, _.y.&Qsr;m,,h;&J$flpir.hs+&y&n w%.
- w -.. u
. 5er],k.,sW4,
.. - m s
- .m
' $.,,% W l;,;. Q L &.& y A. K.L.l$kj &,-s.
?h y '
$p
.y a tu.. +. ss o, y1'"h.:.s 2ee.%cg u A.s M..;: w A b
x, ipf>..c.,
r
- v..;
e.
anA y.
s,MQw m.i.,w.. %m awQ.y.,
m.en.w :.n
>. ~ m.i.:y. w..;u w % w c;vv @
.i,,,Na d k q v e.w.
p.c.p+.s,w, nS.ACRAMENU,.. MU..N. ICIP, AL, UTILITYW
+%~ 'vNsW W M. *MW m %..-
* * ' ' ^p,, n w s
.n c a c
~
~ ^
~, M.m. ws s.
e.
+.~* DISTRICT -
~ * * ' * ~
w-
. y:+M ),,}.,,
NWWc ~c wi W. j {g -
- .4.w,.f..,,. g r @.g }Ms,, a M.
w6p.
fdl W-ns ~.:.3. y-. ---. a.. m; r. p$,,.4 -3 +, g w~..
..,. r
%.#* w.
t. g m..,. <,. -
u.r e erg o
),,.
- 4. m
,.e,..u..ew w. m.m 1* 4
,. #.a.s,s. w* (R.ancho. Se
.a
- e w
a.
- v.. m s -
?.,a:y:q&lM[.; '.
.L.q n,co Unit: No. '. l) 9,,, '. 'A a M.;.r.a,,n.4.yv4 4
- n 6,T h e v *n &> <..
s m NWhi h
e.q ke w
g y w.Q.e.w::. i.p. xi p
y T.
- t..
.c<
.y 4, t.;. G w,:.,9 4.; s.. ;.y g. 43;.pyrug,.Qsk&A.m my w
,n s.
,%,:.p.
g y4. p.4 9 w;o w,yas p a m y. 9.w. 4mi.,, e.,.. e,
.N f. m x.w aiw/.%,, g% x m g q
py n 4
- 'n,1 e
prquW tmp.
.. /.
.y,:.,,4 5-e
.g 4 g A p. g : w i,g. w. w w w :;;. m. y y - g y %, q r e g y.
.pge;g o
w
., -.. w.r*a m...gge j! o a :..
J#.x..Lw-y.
w,..Lwq..u..+..w.+.,na.. u..p.#.. :/a..,...w
,. n.,,.. m.
p.
. t.
3,
,,y v
m r;
w.m r
2+ 4@h..tv%gp, W.Q,M.{"MV. o %7hin'Q: Docket...No',,.50 -312 ' '
4 j.,_
N)..a;*,,
F
"'W3 W.
2 v
Q-Q QL&-
g s.M@-,,9. a;.gwM
~ w %y, ag 1
.., a....:u.~.g :
, %...,n< m.
. m x,,..
- v..
p, 1
'A 3
n %.8_
u
.a - vJ,
.py, gg
.. ] g,_ y Q
,jg
-+.sf,<.4.. y+. y-~&.
- n.,... m-..- W%w
. pngn a w
,. m.
n, nm. q;.yu, ~.,c.p. ~..n-m,.
M m
.w
. s.
w w., n., m.mA..y
" d
- t. W...g w,Q_ 2 d %.m k y M,W ^
Q
~
vB Ma,,,,*[ N aie %.- &.~ -
n p
iW '
..^W M M#
?W V.
~
- gs.;,
g5 Date...~ S cram;.Aw... caw n.W,v.p mo
,.g. 3 y-e;
%,:..mw.
,n y fL; s :n < a s yg% % u%a.wwwl.c/w.1 C,.alifor%ia.Mw,yp@y* Pope *s s\\
J..
e.9 v%eWa pM+c o
n m
,e ento n
mQ:
w,
,n
.. w w,m. m %m y % ), b mw.,.
rn.hyx
- A y ednesday L 13; June:lS73 h h n g 4,.kw.
m.
+
4 h @ >ji.idff!M M.W, 9 w
y u,.
a %. n.mdy;. M %md.M.~p4;,%m,w:4M mp,@w%c,p
~
~,
g m.
XM v;
n h
M ?c -
%3nQgMONg
. g Mp,p>fN@[,YNrk@MMyMWM*MQ~rr NW K
mfMW&
D4 O*
&l~Q[t h MA E
&%. m.n~
m s
..,~,.,n,.._r.
,.- w. r.-.,,.
h h'b
$hNM h,N L4 W @ii$k'N F
- M.
' '~
r
%- W '.M M S,tk
+
sk wyfof sx wp m.
3.&4
&n&9.Wn.ggw $$
r
~.k.n
- w [.. y
- a.,.
?
lQ n $
- h. I h, @k;W.
f Q
Nk2h]xhh fY $b& W.
1
~
g^" 3 :
b gy.%,yggg.h - %..
r 4
nWo m@
p.
'3 m y md.wr d.ib f? g. p g g g g(Xg g4
.,i A ' M.
gy
- gt,,
h3 MSN Q h. M ' @A.,.7;A, _M M. D_f k & ~MEWh_MM M
QW e Code 20tl
'A' o
3 9,,,.y fds -
y 4.~ @ %, V y k3d,% @$g.nf.,n r..h Q.'; & * % + 4 9;:%-
.~
s m
f m
%g.N' 71 N..A }. }@p>{ +3M
% Q-
- y. -
f.,
4C W,y
..g p-d@@g. 0fftczal Reporter.m@E.o,hW@w n WQ N@W..$ f.
g y
.zy
- ~~- @ + 9
- )-
M Nyh'
$gt m-W,,,izq,{
. ; M"WJ415 Secomi Sereet, N.
Q W
my)%. p@M.cA.upp m7pf W.W a.o x w
M W'
~
a e
Ph ?5h'"We D C.A20002 W r "s % %* W @ M.',
W W. M.
r'.y% iW ). w, pM m
7-l @ m,a.3 %.;' % W Q Q % g s % %x, NO- &@$h& @p%gkh9"Qr.::
- f~p$T
.M %q&*"'Q~
.z c
- M ClNATIONWIDE COVERAGE 6PGM.W,.Qg '~~ _ M;jMy 4
M U
bWEh5hhf'Q4WM W$O&
g eqpa wmpg
. W M
1 PUBLIC NOTICE BY THE USAEC ADVISORY 2
COMMITTEE ON REACTOR SAFEGUARDS 3
e.
Wednesday, 13 June 1973 4
5 The contents of this stenographic transcript of the 6
proceedings of the United States Atomic Energy Commission's 7
Advisory Cummittee on Reactor Safeguards (ACRS), as reported 8
herein, is an uncorrected record of the discussions recorded 9
at the meeting held on the above date.
10 No member of the ACRS Staff and no participant at f
11 this meeting accepts any responsibility for errors or inaccu-12 racies of statement or data contained in this transcript.
13j 14 15 16 17 18 19 s
20 21
(.
~
22 23 24
- e-Feder;t Reporte,$, Inc, j
25 9
\\
5' -
i
DENNIS 1
CR 1752 dd i
UNITED STATES OF AMER 7CA 2
ATOMIC ENERGY COMMISSION 3
In the matter of e
4 SACRAMENTO MUNICIPAL UTILITY Docket No. 50-312 DISTRICT 5
(Rancho Seco Unit No. 1) 6
x 7
Federal Building & Courthouse 650 Capital Mall 8
Room 2545 Sacramento, California 95814 9
Wednesday, 13 June 1973 10 The meeting of the' Subcommittee on Rancho Seco 11 Nuclear Generating Station of the Advisory Committee on 12 Reactor Safeguards was convened, pursuant to notice, at 2:10 p.m.
14 BEFORE:
l'5 W. R.
STRATTON, Chairman, 16 Subcommittee on Rancho Seco Nuclear Generating Statior 17 DR. H. O. MONSON, Member.
18 DR. C.
P.
SIESS, Member.
19 PRESENT WERE:
20 on behalf of SMUD:
21 R. Wilson D. Kaplan
(
22 D.
Raasch 8
V' J. Mattimoe 23 F. Anderson R. Moore 24 L.
Stephenson ice -Feder:I Reporters, Inc.
R. Colombo 25 R. Rodriguez 9
6
2 dd 1
On behalf of SMUD - Continued.
2 C. H. air R. Pc'.<ers 3
R. Snider
^
M. Davis 4
J.
Lee L. Keilman 5
P. Oubre' J. McColligan 6
J.
Hiltz J. Dunn 7
On behalf of Bechtel:
8 W. Bingham 9
J. Dempsey L. Lepisto 10 B. Aley L. Brown 11 R. Norry W. Stinchfield 12 W. Brandes 13 l On behalf of Babcock & Wilcox:
14 G. Glei R. Schomer 15 R.
Reed R. Turner 16 J. Janus J. Johnson 17 On behalf of the Regulatory Staff:
18 R. C. DeYoung 19 R. W. Klecker B. C. Buckley 20 D. K. Davis A.'
I. Johnson 21 On behalf of the San Francisco Operations Office:
(
22 Frank Walter, Public Information Officer 23 24 ke - FederCl Reporters, Inc.
25
3 EEEEEEEEEEE CR 1752 1
CHAIRMAN STRATTON:
Good afternoon, Ladies and t:n 1 Reba 1 2 Gentlemen.
The meeting will come to order now, please.
3 This is a public meeting of the Subcommittee of the 4
Advisory Committee on Reactor Safeguards.
I would like to 5
make a few comments in regard to the purpose of the meeting 6
and our plans for proceeding this afternoon and tomorrow.
7 The purpose of this meeting is to develop additional 8
information for consideration by the full Advisory Committee on 9
Reactor Safeguards in its review of the application of the Sacra:n-10 ento Municipal Utility District for an operating license for 11 the Rancho Seco Nuclear Generating Station.
The need'for tnis 12 seview is specified in the Atomic Energy Law as amendad in 1954.
(
13 The same Act created the ACRS and defined its 14 responsibilities.
This meeting is being conducted in accordance 15 with the provisions of the Federal Advisory Committee Act, a 16 law which went into effect only at the beginning of this year.
17 In attendance at the meeting today is Mr. Mort Libarkin on my 18 left, the designated Federal employee who earlier approved the 19 agenda for this meeting.
20 The rules for public participation have been announce i 21 as part of the notice of this meeting previously published 22 in the Federal Register on May 23,, 1973.
Copies of the Federal Na 23 Register notice are available to anyone in attendance today.
l 24 If there is anyone whoidoes not'have a copy, we will be pleased 1.ce Federal Repor ers, tne.
25 to give you one at this time if you will so indicate by raising
4 D:n 1 I
your hand, perhaps.
R2ba 2 2
(No response) 3 CHAIRMAN STRATTON:
Ms. Dundr at the door has the n.
4 extra copies.
In addition to the publishes rules, we have some
~
5 minor rules regarding cameras and tape recorders.
We ask that 6
these devices be self-contained and not interfere wi~ch the 7
conduct of the meeting.
Movie and television cameras should be 8
used only during intermissions or breaks or at the close of 9
the meeting.
10 A transcript is being kep and will be available on 11 or about June 16th in Washington, D.
C.,
for a small fee.
12 Because a transcript is being made, I ask each
(
13 speaker to identify himself when he first speaks and to speak 14 clearly and with sufficient volume so that the Reporter as well 15 as the Subcommittee may be able to follow what is being said.
16 The official reporter on my left may interrupt 17 from time to time to confirm an identification or a difficult 18 expression if he does not get it the first time.
19 The public notice in the Federal Register established 20 the rules for submission of written statements or for requests 21 to make oral statements to the Subcommittee.
Because the press 22 release was inadvertently delayed, as some of you well know, b
23 the rules as published have been relaxed.
24 We have received written statements which I have
.ce-Federal Reporters. Inc.
25 in front of me here, and we have received ten requests to make
^
t~**
~ = - -
5 Den 1 I
oral statements.
If there are among you those who wish to make Rrba 3 2
oral stat.ements today instead of tomorrow, or find it more 3
convenient to do so, the Committee would be happy.to oblige you, and to those so wishing to make their statement today pleas 4
e 5
raise their hands so I may make a count and estimate the time 6
required.
7 How many people would like to make their oral state-8 ment today instead of tomorrow?
9 (No response) 10 CHAIRMAN STRATTON:
I see no hands.
While the 11 Act does not -- does provide that members of the public with 12 notice may make statements !.n writing or orally, there is no
(
13 provision for public members in attendance to interrogate 14 either the Committee or 'che Staff or the Applicant or repre-15 sentatives of the Applicant at this meeting.
16 We do not expect any questions from the public.
I 17 wish also to advise you that another meeting of this.e.ubcommittee 18 is planned for tomorrow, June 14th, in this same room, beginning 19 at about 8:30 a.m.
20 We also expect, if the Subcommittee's review is 21 completed tomorrow, that the Rancho Seco Project will be
(;
reviewed by the full Advisory Codbittee on Reactor Safeguards 22 23 in July during its 159th meeting in Washingto.., D.
C.,
at 1717 24 H Street, on either July 12th or July 13th.
.ce -Federal Reporwis, tnc.
25 We have for today's meeting the review of the Appli-cation of the Sacramento Municipal titility District for
-~n
6 D;n 1 I authority to operate the Rancho Seco Nuclear Generating Station'.
R:bn 4 2 For future reference in the record the nuclear steam supply 3 system is being furnished by the Babcock and Wilcox Company and 4 the Bechtel organization is providing construction services to 5 the Applicant.
6 The AEC Docket Number is 50-312.
7 The first item in the processing of the applicatich' 8 will be the report from the Regulatory Staff of the AEC on this 9 project.
10 I cal'1 upon Mr. Buckley of the AEC Regulatory Staff II.at this time.
Bart, would you proceed, sir?
12 MR. BUCKLEY:
Good afternoon, Gentlemen.
My name 13 !is Bartholomew C. Buckley.
I am the AEC's Project Manager for the 1
14 Rancho Seco nuclear facility.
15 I would like at this time to briefly describe the 16 Rancho Seco nuclear power plant.
The Rancho Seco plant is 17 located 26 miles southeast of Sacramento and is approximately 85 18 percent complete.
The Rancho Seco plant W.*ll use a pressurized 19 water reactor with a design core power level of 2772 megawatts 20 thermal.
t 21
.An addition 16 thermal megawatts will be generated 22:by four reactor coolant pumps.
Q The power and conversion systems 23 aave been designed to handle the combined thermal load of 2788 1
24 megawatts thermal.
n-rwerai neporwis, ine.
25 The heat generated by the reactor core and the reactor t
i
7 Den 1 I
coolant pumps will be removed by tvo once-through steam generaterc.
R:bn 5 2
The steam from the steam generators will be used to drive a 3
conventional turbine to generate approximately 963 electrical n
4 megawatts.
Heat energy from the steam cycle will be removed by 5.the condenser circulating water system which in turn will dissi-6 pate this heat energy to the atmosphere by convection and 7
evaporation in two hyperbolic, natural draft cooling towers.
8 Makeup water to the condenser circulating water 9
system will be supplied from the Folsom South Canal to compen-10 sate for water losses due to evaporation, drift, and blowdown 11 occurring in the cooling towers.
12 The reactor core will consist of 177 fuel assemblies,
(
13 with each assembly containing 208 pressurired fuel rods which i
14 will consist of resintered uranium dioxide pellets, approxi-15 mately 95 percent theoretical density, clad An cold worked 16 ziraloy-4.
The reactor core will be enclosed in a reactor 17 pressure vessel which in turn will be contained in the leak 18 tight reactor building.
This concludes the general description 19 of the plant.
20 There are certain modifica... I that we will require 21 to be completed prior to issuance of an operating license.
22 These are:
(1) modifications tosthe instrument guide tubes k_/
23 whi'ch penetrate the lower head of the reac.or pressure vessel, 24 (2) modification to the control rod extension tubes, (3) a-rwerri nepames, ine.
25 installation of a flow limiter in the core flooding tank line, 9
., j.
.-r
- + -
8 Dan l' and finally, the installation of a flow meter in the decay j
Reba6 heat removal system cross connected piping.
2 CHAIRMAN STRATTON:
Would you repeat those for us 3
again?
4 MR. BUCEEY:
Sure.
5 (1)
Modifications to the instrument guide tubes 6
which penetrate the lower head of the reactor pressure vessel, 7
(2) modification to the control rod extension tubes, (3) 8 9
installation of a flow limiter in the core flooding tank line, 10 and finally, the installation of a flow meter in,the decay heat jj removal system cross connected piping.
CHAIRMAN STRATTON:
Thank you, Bart.
12
(.
G ahead.
13 MR. BUCKLEY:
Okay.
14 All these modifications have 'Jeen reviewed by the 15 Staff and have been found to be acceptable.
16 There are several areas where SMUD has committed j7 to provide additional information of where SMUD has recently j8 submitted information within the last few days where we have j9 not had sufficient time to perform a thorough review.
These are:
(1) meteorology, (2) spray ponds, 21 (3) fuel densification, (4) emergpncy core cooling system, 22
.,b (5) high energy line rupture outside containment, and (6) 23 diesel generator schematic drawings.
24
. --r m.... 9j z,cu1,11x, ec,,1t3 your p.,m1,,1cn, ay, c3,1,m n, ae4 M
9 Den 1 discuss these matters in the sequence just mentioned.
1 2
CHAIRMAN STRATTON:
Okay.
Go ahead.
R ba 7 3
MR. BUCKLEY:
First is meteorology which is discussed em 4
in Section 2.3.6 of the Staff's Safety Evaluation Report.
5 We have reviewed SMUD's meteorological data and have 6
concluded that the atmospheric diffusion values proposed by 7
the Applicant on the basis of their data are acceptable.
How-8 ever, we believe that the onsite meteorological data collection 9
program should be improved to provide 90 percent data recovery.
10 SMUD has committed to upgrade its meteorological 11 program to conform with the provisions of Regulatory Guide 12 1.23 and will, following implementation of improvements in its
(
13 program, submit collected data in the near future to demonstrate i
14 that its meteorological program complies with Regulatory Guide 15 1.23 and to also confirm the conservation of the selected 16 diffusion values.
17 The next item, spray ponds, is discussed in Section 18 2.4.9 of the Safety Evaluation Report.
Rancho Seco will utilize 19 two spray ponds to dissipate decay hea.t during normal shutdown 20 and during emergency shutdown.
21 At the Staff's request, SMUD has conducted a com-22 prehensive preoperational test on,these ponds and is currently
(,
23 evaluating the results of this test.
We will re"lew these 24 results and the Applicant's evaluation to confirm the adequacy ice-Federal Reporws, tne.
25 of the spray pond heat dissipation capability before issuance 4
l
- ' ~
eo ~e
10 f
Dn1 1
of an operating license.
R;ba 8 2
CHAIRMAN STRATTON:
What kind of tests were these, 3
Bart?
e 4
MR. BUCKLEY:
Tests to measure heat balance, drift 5
and losses from the spray pond.
These were conducted approxi-6 mately two to three weeks ago.
7 CHAIRMAN LTRATTON:
These were merely putting the 8
spray ponds in action, so to speak, to see how much would drift 9
away?
10 MR. BUCKLEY:
Yes, Mr. Chairman.
It was pretty 11 detailed.
They had quite a bit of instrumentation on the ponds 12 and the various areas of the pond to measure drift and tempera-h 13[ ture increase and flow.
14 The Applicant will discuss this later.
15 CHAIRMAN STRATTON:
Fine.
16 MR. BUCKLEY:
We will look at these results before 17 we issue an operating license.
The next item, fuel densification, 18 is discussed in Section 4.2.1 of the Safety Evaluation Report.
19 The Rancho Seco Unit 1 plant is one of a series of Babcock and 20 Wilcox plants of which Oconee 1 is the lead plant and prototype.
21 Our review of the fuel densification phenomenon 22 has concentrated first on Oconee 1 and has recently been 23 completed to the point where acceptable operating limits have 24 been established for that plant.
ce-Feder:l Reporters. Inc.
. l 25 Meanwhile, SMUD has furnished us,an interim fuel
_-~-,
~
11-Den 1 1 densification analysis of Rancho Seco Nuclear Plant, Unit 1, Reba 9 2
in which they have proposed a reduction in allowable flux
,2a o 3 imbalance limits in order to meet t he 12340 Fahrenheit accep-4 tance criteria for the loss of coolant accident.
~
5 We will complete our fuel densification analysis 6
for Rancho Seco Unit 1 in a manner consistent with our analysis 7
of oconce 1.
8 CHAIRMAN STRATTON:
Did you say the Oconee Unit 1 9 review has been completed?
10 MR. BUCKLEY:
Yes, sir.
11 CHAIRMAN STRATTON:
Would you -- I would like to 12 have -- I would like to compare the two cores sooner or 13 later, today or tomorrow.
Perhaps it would be better if we s
14 just keep this in mind for a later point.
Would that be all 15 right?
16 MR. BUCKLEY:
The Applicant is going to discuss it.
17 We could respond later.
18 Several days ago, we received SMUD's final report 19 on this issue.
We have not had the opportunity to perform an 20 in depth review of this report.
However, a preliminary. review 21 has been p.erformed and indications are that the operating 22 limits approximately the same as those proposed by SMUD will be 23 acceptable for Rancho Seco Unit 1.
24 We will conclude our review of the fuel densification l.ce-reenl Reporkts, ine, l
25 matter for Rancho Seco and a lot of our technical specs before
12 Den 1 1
the issuance of an operating license.
We expect this review Rebc 10 2
to be completed in a few weeks and we will confirm the accep-3 tability of the design power operation for Rancho Seco with 4
respect to emergency core cooling.
.and 1
5
- CR 1752 6
7 8
9 10 11 12
\\
13 14 15 16 17 18 19 20 21 22 C
23 24
.e. - reserai nepor ters, inc.
25 I
4 6
m w.
13
~
CR 1752 1
The next item, the emergency core cooling system, D:n 2 2 is discussed in Section 6.3.1 of the Safety Evaluation Report.
R bn 1 3
On this matter, we recently requested SMUD to provide 4 additional information on the effects of a large line break, 5 a small line break, and a core flooding tank line break using 6 conservative thermal hydraulic assumptions.
7 SMUD has submitted this information and we are 8
currently in the process of evaluating this information to 9 confirm compliance with the emergency core cooling system interi: a 10 acceptance criteria.
11 The next item, high energy line ruptures outside 12 containment, is discussed in section 6.5 of the Safety Evaluatio n 13 Report.
Our initial overall review of the Rancho Seco Unit 14 1, for a high energy line rupture outside containment has been 15 completed.
16 We are satisfied that SMUD has used the criteria 17 provided by the Staff for the assessment of the Rancho Seco ig plant with respect to this problem.
19 It has identified areas where design modifications 20 will be necessary.
The Applicant intends to submit further
- 2) information on the results of his analysis in about a week.
We 22 expect to complete our review of that before the July ACRS 23 meeting.
24 We have informed the Applicant that we will review j co-FHeral Repntkts. Inc.
25 all design modifications for acceptability and will require their
+
t
_ mmW< '
~
14 Den 2 1
implementation before permitting Rancho Seco Unit 1 to operate R;bs 2 2
at significant power levels.
3 The last item concerns a review of the diesel m
4 generator drawings which is discussed in section 8.4 of the 5
Safety Evaluation Report.
We are currently reviewing recently l
6 submitted diesel generator drawings to obtain assurance of 7
compliance with IEEE 278 and IEEE 308.
8 We will complete this confirmatory evaluation prior 9
to issuance of an operating license.
We have concluded that 10 after satisfactory resolution of the foregoing items, Rancho 11 Seco can be operated safely at the designed core power level 12 of 2772 megawatts thermal.
i 13 This concludes my discussion, Mr. Chairman.
14 CHAIRMAN STRATTON:
Thank you, Mr. Buckley.
15 Does the Subcommittee have any questions of the Staff 16 at this time?
We will certainly have another chance.
17 DR. MONSON:
The last statement made by the Regulator y 18 Staff left me slightly confused.
May I ask this question?
19 Has the Regulatory Staff yet made a determination that the
~
20 Rancho Seco plant can be operated safely, and if so, at what 21 power level?
I am R4 chard De Young.
The Staff v
23 has made a conclusion that the plant can be operated at full 24 power safely.
We are doing some analyses to confirm some
! a - r *,ai Repo,t.,,, ine, 25 remaining considerations.
We expect these to come out adequate:.y '
H l
MMM
15 Den 2 I
so that the plant can be operated on full power.
Reba 3 2
In the event that these analyses, which have to do 3 with the fuel densification and ECCS requiring any change from n
4 full power will be expected to be limited.
5 DR. MONSON:
At this time, is it not correct that 6
the Regulatory Staff has not accepted the fuel behavior model 7
proposed by 3 and W, particularly -- well, in respect to fuel 8 densification?
9 MR. DAVIS:
I am Don Davis, Regulatory Staff.
10 We have reviewed the B&W's model for assessing the effects of 11 fuel densification, and with a few modifications to the methods 12 that they have proposed, have accepted it as adequate for asses-(
13 sing the effects of fuel densification.
14 DR. MONSON:
So you believe the newest model is 15 adequate; But then is it correct that you have not yet seen 16 the calculations of the effects of fuel densification employing 17 the newest model?
IB MR. DAVIS:
No.
If I can follow you through a little 19 bit on this, we have reviewed their methods, and they have 20 recently, on June 8th, submitted a report using those methods 21 and evaluated the Rancho Seco fuel.
Now there has been one 22 departure from the requirements of the Staff, and that is that
(
23 the as-built dimensions and properties of the fuel are not 24 available yet, and so the evaluation was done upon the specifi-
.ce - FWeral ReporMrs, inc.
l 25 cations and limits to the specifications.
w
16 Den 2 1
In other words, a conservative approximation to R:ba 4 2
what they expect the as-built properties to be was made, at 3
least B&W felt it was a conservative approximation.
4 The Staff will review this report and then require 5
B&W to furnish -- or SMUD to furnish -- the as-built properties 6
and will confirm that the specifications used by B&W to assess 7
the effects of fuel densification were in fact conservative.
8 There are three major as-built information data 9
that are not available.
They concern the initial densit? of:the 10 fuel and the standard deviation from that initial density; 11 the main pellet diameter; and the a sorbed gas content of the 12 fuel.
B&W has made what they feel are conservative approxi-13 mations for what these parameters will be in the as-built 14 condition, and after the results of the measurements are in, 15 the Staff will make an evaluation to assure that they were 16 indeed conservative or else we will require them to reanalyze 17 the fuel, essentially go through the model again, using the 18 as ' :llt information.
19 Is that clarifying?
20 DR. MONSON:
That answers part of my question, 21 but aside from these effects, in your Safety Evaluation Report, 22
.on page 4-3, you say, "The Regulatory Staff is continuing its 23 review of the fuel densification phenomenon and the associated 24 effects.
Presently the Staff is reviewing the B&W evaluation Ace-re rai nepo,wis, ine.
25 model for fuel of the type to be used in Rancho Seco Unit 1.
e w-
_tw-
~
17 "After development of an acceptable model by B&W Den 2 1
we will determine if any operating revisions will be necessary."
Reba 5 2
Now do I understand you correctly to say that since 3
~'
the issuance of this report, in effect, the Staff has become 4
satisfied with the B&W evaluation model?
5 MR. DAVIS:
Yes, that is correct.
6 DR. MONSON:
I see.
Now the nature of the operating 7
restrictions, that you may feel it necessary to impose, is what?
8 MR. DAVIS:
The Applicant will have more elaborate 9
discussion of this, but basically, they deal with the imbalance 10 limitsontheplankwhichaffectsthemaneuverabilitysomewhat.
11 CHAIRMAN STRATTON:
Imbalance, is that the same as 12 axial offset?
13 MR. DAVIS:
Yes.
14 CHAIRMAN STRATTON:
Same thing?
15 MR. DAVIS:
Yes.
16 CHAIRMAN STRATTON:
Thank you.
17 Which phrases are we using today?
18 MR. DAVIS:* The word " imbalance".
19 (Laughter) 20 CHAIRMAN STRATTON:
Just so I know.
21 Thank you.
r 22
\\#
DR. MONSON:
Is there any possibility that these 23 restrictions that may appear necessary to assure safe 24 Au - FWeral Repor krs, inc.
operations Would be unacceptable for some reason other than the 25 t
t
"-n
^m
18 Den 2 1 reason for which you're imposing it?
Reba 6 2
MR. DAVIS:
Anything is possible, but we do not 3
foresee any change such as that.
4 DR. MONSON:
Well, what-b'others me is that you seem 5
to have been able to draw the conclusion that this reactor 6
can be operated safely at full power; yet, you are awaiting 7
additional information to enable your determining whether or 8
not some operating restrictions may be required over and above 9
those presently proposed.
And ---
10 MR. DAVIS:
The information is confirmatory infor-11 mation.
I think that is the best way to cast it.
We expect 12 that'the information will confirm the assumptions made that 5
13 the as-built properties are conservative.
14 They are based upon B&W's experience with measuring 15 these parameters and what the values would be from past fuel 16 loadings.
It is not that it is something new or it is the 17 first time B&W has made these measurements.
18 They have a good idea that these, in fact, assumptions 19 they made are conservative or else they lend themselves to 20 the penalty of having to re-do the analysis again.
21 DR. MONSON:
Do you contemplate operation at 19 kw 22 per foot peak?
8
,(_.
23 At 100 percent power?
24 (Regulatory Staff conference)
Ace-Fede al Reporters,Inc.
25 MR. DAVIS:
We have not completed our review of the S
e
?
" ' ' ~
~
- - - - - - ~. ~,
_~.
19 Den 2 1
ECCS in the fuel densification report or in the additional R;bn 7 2
items that Mr.Buckley discussed.
I think we will have finished 3
our review by the full Committee -- the time of-the full 4
Committee meeting.
5 DR. MONSON:
But at this time it is 19 kw per foot, 6
the value proposed by the Applicant?
7 MR. DAVIS:
Yes.
8 DR. MONSON:
And you have found that operation at 9
full power, employing 19 kw per foot, I suppose, is acceptable ~
10 and safe?
11 (RegLlatory Staff conference) 12 MR. DAVIS:
The 19.1 kilowatts per foot is the peak 13 power that is -- it is not expected.
In fact, for Oconee 14 1, a similar situation, they would not achieve that power.
15 DR. MONSON:
The proposed permissible power, 16 however?
17 MR. DAVIS:
Yes.
58 DR. MONSON:
Since that is not one of the restrictions 19 that you are considering, apparently you propose accepting 20 19 kw per foot?
21 MR. BUCKLEY:
It is possible, yes, sir.
(.
22 We still have to look'at the fucl densification 23 report which uses 19.1, and we have to see the effects or 24 review the results of the SMUD submittal on the ECCS to confirm Au - FWerrl Reporkrs, lnc.
25 that 19.1 is acceptable, but indications are right now that g
20 Den 2 I
they will be.
Reba 8 2
We jutt got these reports several days back. We 3
have just glanced through it.
Indications are that they will 4
be acceptable.
We will confirm this.
5 DR. MONSON:
Suppose that the additional information 6
fails to confirm this.
Might you then feel it necessary to 7
lower the peak permissible kw per foot?
8 MR. DAVIS:
That is possible.
9 DR. MONSON:
Is there any question that the number 10 to which you would have to lower it is less than an acceptable I'
number for any other reason?
12 (Regulatory Staff conference) 13 MR. DAVIS:
I am sorry.
I don't understand your 14 question.
1 15 DR. MONSON:
If achieving this reduced peak kw 16 per foot required the use of a total peaking factor less than 17 you could be assured.the Applicant actually could achieve, and 18 be certain of achieving, would this not represent a problem to 19 you?
20 (Regulatory Staff conference) cnd 2 21 C
1752 22 v.-
23 24 Ace-FHerai Repor ers, Inc.
25
DD #3 21
. I ty 1
)
MR. DAVIS-What you say would probably be a 2
problem to us.
We don't anticipate that to be, in fact, 3
true.
If you were to compare the peaking factor that is 4
5 used in this plant, it is a rather large factor compared to peaking factors for pressurized water reactors which the 6
7 Staff has approved in the past, if that goes somewhat 8
towards answering your concern.
DR. MONSON:
Has any plant yet operated with a heat 9
10 flux as high as the peak heat flux presently proposed for this
[
jj unit, which I think was something like 770,000?
MR. DAVIS:
The heat flux you are referring to, 12 is that a hundred percent power heat flux?
13 DR. MONSON:
Yes.
It is in one of the tables in j4 Y ur report somewhere.
I just don't have my finger on it.
15 MR. DAVIS:
I am not sure.
16 Si DR. MONSON:
Page 42.
776,400.
j7 MR. BUCKLEY:
I am not sure at this time.
18 We can check that.
39 MR. DAVIS:
We can check that out for you.
20 DR. MONSON:
Well, let's go ahead.
21 chet?
22 u.-
DR. EIESS:
One of the outstanding items that 23 y u mentioned was the meteorological data collection.
24 In arriving at the relative concentrations in your 5
~m
22 ty 2 1
Safety Evaluation Report, did you use the available data that' 2
was coll,ected, and, if so, did you use it in some conservative 3
fashion to account for the --
4 MR. BUCKLEY:
Yes, sir.
We used the data's data.
5 I don't know the specifics of it but we did come up with 6
consservative estimates.
I checked this with our meteorology 7
department.
He is satisfied that the diffusion values are 8
conservative.
9 DR. SIESS:
If their data collection system, with 10 90 percent recovery, gives essentially the same results as 11 they have obtained so far, this would establish your calcula-12 tions as conservative?
(
13 net. BUCKLEY:
Yes, sir.
14 DR. SIESS:
HoS A-Ryour;_ relative. concentrations.,.
15 fof "i---reasucompace withrthose7atrawssume&atwthewen:
16 strar.t.i.on permiL, stage--at whichstime-:I~ suspect *yoti*dit not 17 h=c-e..f-eLAbath 18 MR. BUCKLEY:
I would have to review that.
_l~5A ^ --itMINielike=to 19 DR. SIESS:
We ' A 20 knew _la_t.er.
MR. BUCKLEY:
- Yes, 21 s_-
CHAIRMAN STRATTON:
knother question?
22 i
23 DR. MONSON:
In respect to the high energy 24 line rupture outside the containment provisions, you co-Federal Reporters, Inc.
25 indicate that complementation of the fix will have to be
23 ty 3 1
completed before the unit may be operated at significant 2
power levels.
3 Is that significant meant to be 5 percent of full 4
power? '
GGenerally we have required it to 6
be about -- you know, about 5 percent of full rated power.
7 However, in the event that there are one or two major modifica-8 tions that are rather difficult to design and to install, we 9
would accept some interim period of augmented in-service l
1 10 inspection for such logations.
11 We expect about 5 percent to be the guiding light 12 for this plant as it has been for all of the others.
13[
DR. MONSON: Thank you.
14 CHAIRMAN STRATTON:
T; _ _, __ r f m
.,_ mumi yy,
u..
15 Wo? :=
==L
- - E" "" ^ %h== 2 alar'Di udZoTt W '"'T~Oon't 16 ci- "lch<1ey-thc ' report %ich-leda to,,t.hextor.nada 17 cr W ir eu~ maxim s =+inda g eed which-isarG u1Zu M esign 3
18 cr Q ?,
19 MR. BUCKLEY:
We would like to do that tomorrow 20 with your permission.
21 CHAIRMAN STRATTON:
Fine.
22 DR. MONSON:
With respect to tornado protection,
(_.
23 you indicated on page 3-3 of your report that all systems 24 and components necessary for safe shutdown are now designed to-Federal ReporWrs, tnc.
25 to withstand a sustained wind velocity of 175 miles per hour, v.ws..
24 ty 4 1
or a suitable redundancy has been provided to minimize the 2
risk of loss of functional capability.
3 Can you tell us in what areas redundancy has been 4
provided in lieu of designing for 175 miles per hour?
5 MR. BUCKLEY:
This is discussed in the FSAR.
I 6
could specify a table where we could look at it specifically, 7
if you would like to do that.
8 DR. MONSON:
Would you?
9 I ask this because I have in mind cases such as 10 occurred at one plant wL ;; ;- v uaua c v u r. vuu av v a me m _;g 11 t -.-..
, _.. la m -fivu u ous.u s uon - unes uv. m..y i.. t ;-
12 M.
13 In other words, redundancy did not provide the 14 equivalents of designing for tornado protection, and 15 hence, I would want to see that these systems don't need the 16 same degree of protection as those for which you do 17 provide the 175 miles per hour design protection.
18 MR. WILSON:
Could I explain our little philosophy
}9 on that to assist you?
20 I am Robert Wilson from SMUD.
21 When we reviewed the facility, the basic 22 philosophy in looking for redundancy was that the backup s
23 supply system or co:.ponent was designed such that it just 24 was not available for tornado or tornado risk damage An co-Federal Repor ers, Inc.
.25 example of this is in one case we have the, borated water
~~
qp
~ _
25 ty 5 1
storage tank which is out in the yard and could be available 2
for damage.
3 A backup to it is a buried tank which is buried belc w 4
ground in concrete and, of course, would not be affected by a tornado, which has concentrated boric acid.
5 6
Another example is a backup to the condensnte storac e 7
tank which is above ground.
It would be underground piping 8
systems going to the reservoir.
9 This was the general trend of our review of backup 10 redundant systems for this philosophy.
11 MR. BUCKLEY:
For instance, spray ponds was 12 another issue, where you have two spray ponds.
If you were toloseone,by)$o'rtofatornadomissile,youcouldtransfer 13 14 h coolant or the water in the spray pond from one to the 15 other.
This may take a little time.
16 DR. MONSON:
Is that Table SA-7?
17 MR. BUCKLEY:
I think it is.
Where you have 18 various missiles and --
19 (Dr. Siess indicating.)
20 MR. BUCKLEY:
Yes.
That is the one I am 21 referring to.
22 DR. MONSON:
Now this table indicates the systems s.__.
23 for which you haveprovided design protection against wind 24 velocitics as indicated in the table.
Is that correct?
O-Fehral Reprters, Inc, 25 MR. BUCKLEY:
Yes.
h
ty 6 26 j
DR. MONSON:
It Anoc -^* "ac s a m"41 > indicate go:ridcd cueh-seesgn 2
systar # r ti^ rc; :.ou
..v-P;D scti^" but Oc"r^"na d redundancy?
E 3
4 MR. BUCKLEY:
You say it does not?
5 DR. MONSON:
I am asking.
6 DR. SIESS:
The last column which represents a 7
wind speed of 101 miles per hour, which was the original 8
design wind speed for a plant, is the only column under the 9
75 miles per hour tornado wind, and does this table include 10 all of the Class I systems?
And the X's in the last column ij represent those that would be vulnerable to a tornado missile 12 and are otherwise taken care of by the redundancy?
13 MR. BUCKLEY:
We are looking for the table while 14 y u were making that statement..
DR. SIESS:
The last column, 101 miles per hour 15 includes some X's on tornado missiles and two on wind.
16 j7 Ar m m..
-J._
r: 1;.r-~ r r.c a tr ' w_ _= " a
- "_1_a ^ r ri '_^
18 to W75r Ittire~s par hobr wilic'ana afeNerefere"taken cara-c"- b y,i,. u u W s trc y r-39 MR. BUCKLEY:
To the best of my knowledge, that is 20 21 correct.,
DR. MONSON:
E:_12 Lh:t t: h^ enva nad 23 vuuva v..
22 23 1 "? " " " 21 20 2 7 2!!' ;; ;t th" f"Il 2 2-- # ^ ' t ?
MR. WILSON:
Let me summarize this for you tomorrow 24 0-Federal Reporkts* Inc 2j and I will give you a complete rundown on the analysis.
t m
.-m- + -
e
- m. z
ty 7 27 1
DR. MONSON:
Very well.
2 CHAIRMAN STRATTON:
Does that complete what you have 3
to offer for us at the present time?
4 MR. BbCKLEY:
Does that complete my portion?
5 CHAIRMAN STRATTON:
Yes.
6 MR. BUCKLEY:
Yes.
7 CHAIRMAN STRATTON:
Thank you.
8 I will ask now for the presentation by the 9
Applicant.
10 Who will be the speaker?
11 MR. WILSON:
Mr. Chairman, my name is Robert Wilson, 12 I will be the primary speaker for the District.
13 I have with-me at the table starting at the far 14 end Dallas Raasch, who is project manager for Rmcho Seco; Dave 15 Kaplan, who is our general counsel; John Mattimoe, assistant 16 general manager and chief engineer; Greg Glei, from B&W; 17 John Dempsey, Bechtel Corporation; and Bill Bingham, of 18 Bechtel.
19 I may also call upon Staff members from t ?
20 audience and introduce them at that time.
CHAIRMAN STRATTON:
Very well.
21 I would like, if you,would, to -- let's omit 22 23 Item 1 on the agenda that I think was given to you, a descript:.on of site.
We can catch that tomorrow.
24 co-Federal Reporkts' Inc'
~
25 Let's move to Item 2, the history of the project k,.
ty 8 28 1
including the changes in the FSAR, the expected dates of 2
fuel loading and use of power.
3 MR. WILSON:
I do not have the agenda that you 4
identified.-
5 CHAIRMAN STRATTON:
I am sorry.
6 MR. WILSON:
I have some subjects but they are 7
not in any order.
8 MR. BUCKLEY:
Mr. Chairman, I will give him one if 9
I can find it.'
10 CHAIRMAN STRATTON:
I think we are on the same agenc a 11 now.
12 We would like to try Item No.
2, history of the 13 project and so on.
14 MR. WILSON:
On the history of the project, the --
15 studies performed by the District's long-range planning staff 16 in the early 1960s, established a need for a base load P ant in approximately 1973 to 1974.
It was also determined l
17 18 at that time that a nucLlar power plant would be the most 19 economical and would have the least adverse effects on the loca.1
~
20 environment.
21 A thorough analysis of the possible sites 22 available at"that time showed that a dry site was both 23
. feasible and in ouur particular situation was preferred.
24 The current Rancho seco site was selected and u-FWeral Reporkts, ine.
25 procured in 1964.
e m
-g
-.=en-eWNm,
--iw
.m=
29 ty 9 1
The District awarded the engineering and 2
construction management faces of the project to Bechtel 3
Corporation in 1967 and the design criteria for the plant 4
was established shortly therafter.
A Westinghouse turbine generator contract was 5
6 awarded in April of 1967 and a nuclear steam supply contract 7
was awarded to Babcock and Wilcox Company in August of 1967.
8 We filed our application for a construction 9
permit in November of 1967 and received our license for 10 construction in October 1968.
11 The initial site construction commenced after the 12 rains in March of 1969.
13 The constructionwork was separated into a number of 14 small construction contracts covering several items on-15 site preparation, structural work, cooling towers, mechanical 16 and electrical installation.
17 The mechanical and electrical installation contract 18 was awarded to Bechtel Corporation in June of 1970.
19 Zhe total site manpower, both manual and non-20 manual, peaked at over 1400 in 1972 and there are around 21 a thousand personnel on the site at this time.
22 At this point, all major components hdve been v
23 installed and over 70 percent of the systems have been 24 turned over to startup.
The remaining work exists primarily ce - ree,ai nepo,w,3, ine, 25 of a number of electrical connections with some minor cable
,s
ty 1a 30 I
pulling required.
2 The installation of a few small valves and related 3
welding to close ';. the piping system are also required along 4
with the associated cleanup of the piping systems.
5 Electrical tests and instrumentation personnel have 6
maintained the work to move along these activities.
Of the 7
over 500 line items that are required for operation of a 8
plant, approximately 50 percent of these have been checked 9
out by startup and are either in functional testing or 10 are being prepared for their functional test at this tfme, il The next subject, Mr. Chairman, is changes since 12 the PSAR --
13 CHAIRMAN STRATTON:
Please.
14 MR. WILSON:
Significant design revisions from the 15 PSAR have been identified in Section 1.3.2 of the FSAR.
16 I will summarize these at this time.
17 CEAIRMAN STRATTON:
What is that section?
18 MR. WILSON:
Section 1.3.2.
19 CHAIRMAN STRATTON:
Okay.
1.3.3?
20 MR. WILSON:
1.3.2.
21 CHAIRMAN STRATTON:
Got it this time.
End #3 22 (Laughter.)
23 24 ce-Federat Reporters,Inc.
25 c-
31 jeni 44 i
I 1752 MR. WILSON:
The first revision, core power rating, 2
has bee,n increased from 2452 megawatt thermal to 2772 mega-3 watt thermal.
4 The initial proposed core power rating of 2452 5
in the PSAR also had associated with it an analysis for 2568 6
in the PSAR for some core analysis and also for accident 7
analysis.
8 This upgrading to 2772 is defined in the FSAR with 9
supporting analysis.
10 The next revision, the control rod drives, have II been changed from a rack and pinion type to a roller nut, 12 sealed roller nut, lead screw type drive mechanism.
13 The fuel assemblies utilize inconel spacer grids 14 cupported by the control rod tubes rather than the stainless 15 steel grids which had been supported by an external stainless 16 steel perforated cam.
17 All fuel rods in Rancho Seco core will be 18 internally pressurized with helium to minimize clad fatigue
~
19 due to power and pressure cycling.
20 Eight of the sixty-nine control rods contain 21 neutron absorbers for a portion of their length to aid in l
1 22
- controlling xenon oscillation. 'These are the axial power i
\\s l
23 shaping rods.
24 The next revision is associated with reactivity ice-r.6 rai a.po,J, ine.
. l 25 ;;
control and that is burnable poison rod assemblies that have 9
jon2 32 I
been added to the first core fuel cycle to reduce the 2
magnitude of the beginning of life positive moderator 3
temperature coefficient.
4 CHAIRMAN STRATTON:
Is that -- what is the poison 5
in those rods?
6 MR. WILSON:
That is B C, a sinter B C, an aluminum 4
4 7
oxide absorber-material.
8 CHAIRMAN STRATTON:
Excuse me just a moment.
?
I wonder if we could have only one conversation 10 going on in the room, please.
11 Go ahead, Bob.
12 MR. WILSON:
The next revision was the incorpora-13 tion of ability for monitoring the in-core detectors on 14 auxiliary readout of the in-core detectors as recorded in 15 the control room.
16 There are a total of 36 of these detectors that 17 we would be continuously monitoring.
18 Next revision was a modification in the reactor 19 protective system which included the addition of a power trip 20 based on an imbalance and flow function.
21 The reactor high pressure trip was also added to 22 provide a reactor trip on high building pressure.
23 The start-up rate trip was deleted.
24 In addition, each protection channel has two
- ce-Federal Reporkts, lac.
25 key-operated bypass switches and channel bypass switch and
, -., _ ~ - _
~~~ jcn3T 33 1
shutdown bypass switch.
2 Another modification dealing with reactivity 3
control was the incorporation of the ability to feed and 4
bleed the boric acid for chemical control.
This was 5
incorporated to provide control of transient xenon instead of 6
using the transient xenon control rod group.
7 The initial design of the facility we assume we 8
would have cooling towers system using five sealed induced 9
draft cooling towers.
10 The final design of the facility incorporates two 11 natural draft hyperbolic cooling towers.
12 The nuclear primary piping system complies with 13 B31.7, 1968 Code, and the balance of the nut ear piping 14 complies with ANSI 1937, 1969 version.
15 This is in lieu of the 331.1 which was the code 16 in existence at that time.
17 It should be noted at this time that the 18 incorporation of these codes was at the discretion of the 19 District and is not a regulatory requirement.
~
20 Since the code requir ements are defined in 21 10 CFR 50-55, Paragraph A, on codes and standards for 1
22 nuclear power plants, the early, adoption of'these codes 23 provides more conservativism in our design.
24 The final revision is to the nuclear service bus c - rw.,.i n,,,,,, in,,
25 arrangement.
We previously had the nuclear service bus inter-
~
O 9
f
-w,
34 j:n4 I
connected with a bus tie such that power was normally supplied 2
to both buses from one outside source.
3 An automatic bus transfer was provided in this 4
case.
5 We have separated the systems and the design has been modified to eliminate the bus tie and each nuclear 6
7 service bus is supplied from a separate power supply.
8 In addition to these design changes that I have 9
just identifi'ed, the reactor vessel internals have been 10 modified to reduce flow use vibrations which was mentioned 11 earlier.
12 This modification is described in B&W's topical 13 report B&W 10,051.
~
14 I will provide to the next --
15 DR. MONSON:
Let me ask a question in this area.
16 nr. the time of the construction permit review you 17 were considering initial operation at 2452 megawatts thermal, i
18 nad you had mentioned the figure 2568.
I think the engineered 19 safety features were evaluated at 2568.
Was there any higher power level mentioned at 20 21 that time?
22 MR. WILSON:
To my Anowledge not during the PSAR 23 review.
~
24 DR. MONSON:
May I ask the Regulatory Staff does a-Fehral Repotkts,Inc.
25 it have a policy in this respect?
How much of an increase
jon5 35 I
should be seriously considered in respect to power level at 2
the operating license stage from that which was contemplated 3
at the construction permit stage?
4 MR. BUCKLEY:
We haven't made that assessment at 5
this point.
Again, a hydraulic analysis has been performed 6
at 2772 megawatt thermal for which the plant is designed.
7 Not at the PSAR stage.
8 DR. MONSON:
But it does raise the question as to 9
the significance of a construction permit to pose the 10 question in an extreme manner to make the point clear, if 11 an applicant proposed construction of a plant to be operated 12 at 500 megawatts thermal and got a construction permit for 13 such a plant; but then came in for approval to operate that 14 plant at 2500 megawatts thermal.
15 I am not sure --
16 CHAIRMAN STRATTON:
Let me try a couple of 17 clarification questions if I may.
18 What was the size of the turbine generator 19 ordered at the time of the PSAR; and when was this changed 20 to a' higher number?
21 In connection with this, I believe part of the 22 higher power will be -- has been justified on the basis of 23 larger pumps.
When was this change order made and change in 24 design and when was the Regulatory Staff notified of these a - Federal Reporkrs, tnc.
25 changes?
jen6 36 1
MR. WILSON:
I would like to ask Mr. Mattimoe to 2
address that.
3 MR. MATTIMOE:
To the best of my memory, 4
Mr. Chairman, we notified the Staff immediately when we 5
made our decision to go to 2772 and notified them prior to 6
submission of the FSAR that we would be coming in 2772.
7 We notified them when we confirmed the order on 8
both the turbine and the upgrading of the reactor cooling 9
pumps.
10 CHAIRMAN STRATTON:
Do you remember when that was?
11 MR. WILSON:
Three years?
12 MR. MATTIMOE:
About 1969, sir.
13 CHAIRMAN STRATTON:
Thank you.
14 Dick, you were about to help us out on this.
I was about to respond to the-16 increase by -- from 500 to 2500.
I think we would have a 17 Problem on such an increase.
But we know these plants do 18 increase in power as they accumulate more experience with the 19 components that they have provided, as they obtain more 20 information from R&D programs and fron plant operation at 21 other plants.
And we expect this to come in on those plants 22 that have provided conservatively designed components.
23 So we don't view this as unexpected.
But we would 24 view a very large increase of the type we have, where we Lu-Federal Repo:Wrs, Inc.
~
25 have not performed the study to determine where that line is, 9
- - ~ -. - - * - - ~
-n.--.. - - -.
jon7 37 I
where we would have a real problem.
2 We do review the plant at the OL stage for the 3
proposed power, review all components, all accident 4
consequences against that power; and if it is acceptabla 5
we approve.
6 There are plants that have come back with the 7
same power level, of course.
8 DR. SIESS:
I don't believe the Applicant answered 9
the question hbout the turbine generator capacity.
Was it 10 originally 2772?
Or was that changed?
II MR. MATTIMOE:
No.
The original order for the 12 turbine generator, as mentioned by Bob Wilson -- in fact, the 13 turbine generator was ordered before the nuclear steam 14 supply.
The firming up of the capacity of.the turbine 15 generator was done afterwards, in 1969.
16 DR. SIESS:
At the time of the construction permit 17 you didn't know what capacity turbine generator you were 18 ordering?
19 MR. MATTIMOE:
No.
At that time there was a lead 20 time of five years for delivery.
We did not have to confirm 21 the requirements until three years priot to delivery of the 22 turbine generator.
e s
23 DR.
MdNSON:
Where then did the number 2568 come 24 from in the PSAR?
co-Federal Reporkrs, Inc.
25 MR. MATTIMOE:
That was the power level for the v~*.~-
jon 38 1
Duke Plant, the first B&W plant that was similar in design 2
to ours.
3 DR. MONSON:
I believe the PSAR identified this 4
plant as being in respect to power and power density 5
identical to Oconee; is that right?
6
( Applicant conferring.)
7 MR. WILSON:
They were into a combination of 8
tedhnical and commercial definitions here.
As I recall, the 9
guaranteed design on the core at that time was 2452.
The 10 ultimate expected core capability, which was not a warranted 11 value, was 2568; and the core calculations -
,ad because it 12 was -- at that time was our stretch number for that existing
\\
13 core, the ECCS calculations and the environmental calculations 14 were based upon the 2568; although the initial expected power 15 that we would operate would be to 2452.
16 DR. MONSON:
Is it correct that the power density 17 and in addition the peak KW per foot presently proposed are 18 a peak percent higher than those values proposed at the time 19 of the construction permit review?
20 DR. STRATTON:
Eight percent higher than the stretch 21 power.
22 DR. MONSON:
Than the stretch power.
23 MR. WILSON:
I would like to defer that and perhaps 24 bring the PSAR in.
I don't recall what has happened to peaking
~
6ce reoerai sepo,ars,ine.
25 factors in the interim to know what to say on the peak linear
jcn 39 l
heat rates.
I don't know they would be directly eight percent 2
proportionate.
3 CHAIRMAN STRATTON:
We would like to get into m
4 peaking factors in just a few minutes.
5 Go ahead with the other items that you have.
6 I think you were going to give the expected dates 7
of fuel loading.
8 MR. WILSON:
Yes, sir.
The expected date for 9
fuel loading.- Over 70 percent of the plant has been 10 completed and turned over to startup, with most of the 11 remaining equipment scheduled for completion within the 12 next thirty days.
13 We are currently planning to perform a cold 14 hydro sometime in August which would be followed by hot 15 functional testing and fuel loading by the end of the year.
16 Based on this schedule we plan to be in commercial 17 operation July 1st of 1974.
18 The next item is somewhat related to what we were 19
_previously talking about, the initial power at which
' ~
20 we intend to operate the plant.
^
21 The generating unit will operate at cold power 22 level up to 2772 megawatt thermal.
All of the physics 23 calculations in the core, thermal hydraulic calculations, 24 and the safety analysis have been performed for this core a - FWeral Repor urs, ine.
25 design and +his power level.
f 4
- -gen heumedsun,me=M h w e as.
---e-N
jen 40 I
It is expected that the nuclear steam supply 2
system will have a capability of 2788 megawatts which 3
includes the 16 megawatt contribution for the reactor 4
coolant purposes.
5 All of the power conversion systems are designed 6
to accommodate this power level and produce a gross electrical 7
output of 966 megawatts.
8 CHAIRMAN STRATTON:
That is all in this area?
9 MR'. WILSON:
Yes, sir.
j 10 CHAIRMAN STRATTON:
Thank you.
11 Any questions, more questions in this area?
12 DR. MONSON:
In respect to the changes'that have 13 been made since the CP review, you mentioned the modifications 14 in'the. reactor vessel internals.
Presumably those are 15 identically the same modifications as have been made in 16 Oconee; is that correct?
17 MR. WILSON:
Yes, sir, that's right.
i 18 DR. MONSON:
The modifications are not yet 19 complete on your plant?
20 MR. WILSON:
The modifications are complete at 21 this time.
22 DR. MONSON:
They ane?
23 MR. WILSON:
Yes, sir.
J 24 DR. MONSON:
So they are identical?
Your present
- a-FWeral Reporkts. Inc.
25 plant is identical in re.spect to the vessel internals -- and 9
9
~ - -
w~'
wm..am--.-mm-+s-m a,.
_n,.
jcn 41 I
that includes the instrument guide, noz=les and the guide, 2
thimble.s, and the changes made in the thermal shield, support, 3
and so on?
4 MR. WILSON:
Yes, sir.
~
5 DR. MONSON:
Yours is identical now to Oconee 1 6
and 27 7
MR. WILSON:
Yes, sir, that's right.
8 We have been submitted a response to AEC's 9
questions 'similar to that which would, I believe, be in 10 Section 4-A, Appendix 4-A of our FSAR, that the internal 11 design will be identical to that, as modified, at Oconee.
12 DR. MONSON:
I know that the test results from 13 Oconee are available.
Has the Staff completed its evaluation 14 of the results and the Applicant's analysis of~those results?
15 MR. BUCKLEY:
Yes, sir.
16 DR. MONSON:
Thank you.
17 CHAIRMAN,STRATTON:
Any effect from the greater l'8 flow of water through the core?
19 MR. BUCKLEY:
Would you repeat that?
20 CHAIRMAN STRATTON:
Is there any effect in this 21 regard from the -- a rising from the larger pumps, 4th more 22 water flowing through the core?
23 MR. BUCKLEY:
Five percent?
24 CHAIRMAN STRATTON:
The pumps are larger?
tce-Federal Reporters, Inc.
~
25 MR. BUCKLEY:
Yes, eight percent.
We are not
.==m.---~,-..w.
m-...
42 jcn I
sure of this.
Maybe you could ask the Applicant.
2 I had asked one of our technical people.
He said 3
it wasn' t significant.
Maybe the Applicants could answer 4
that question.
5 MR. WILSON:
We have received that question from 6
the Staff also, and we are checking the reference for you.
7 We did respond to it indicating that we expected no 8
significant difference.
The difference in design of five 9
percent flow is quite small.
10 As I recall -- I think during testing at Oconee II they had another percent or so higher flow than what we had 12 expected, so that relative difference is quite small.
13 We didn' t anticipate any changes in the vibration 14 characteristics or flow characteristics of the internals.
15 DR. MONSON:
Of course, the purpose of the 16 testing is to find out whether what actually happens is what 17 you expected.
18 Did the testing at Oconee go only to 100 percent 19 or 101 percent of fuell flow or did it go to a high enough 20 value to include Rancho Seco's flow rate?
21 MR. BUCKLEY:
It could have been a little higher 22 possibly. Some of the fuel -- the flow could have been -- two 23 or three percent higher.
24 I am estimating at this point due to the small ice-Federal Reporters, Inc.
25 resistance in the primary flow path.
I could confirm this i
j:n 43 1
if you wish.
2 DR. MONSON:
Ti :_12;.
11 '
r^-"' *" tar *:'_ " 45 3
Ac 7 "a<4"'-d J,7ne vwration instrumenta'tte.. ir. Hi s 4
p'-"' "i M ': :..
1 as b.a..dve.-as4he--instaaunechtion 2
5 employed at Oconee because Oconee was to be considered a 6
prototype for this type of plant.
Isn't that right?
7 MR. WILSON:
That's correct.
8 DR. MONSON:
Now, it.is important, then, to know 9
that the test conditions employed at Oconee included the 10 conditions to be employed at this plant.
If it fell short 11 of them, in respect to flow rate, you are still making an 12 extrapolation.
You are not testing.
You are still using
(
13 only the analysis.
14 MR. BUCKLEY:
Well, there will be confirmatory 15 preoperational tests run for Rancho Seco.
Not in the same 16 detail as Oconee 1, but there will be confirmatory 17 vibration tests.
~~
18 Incidentally, the stresses that resulted on 19 Oconee 1 were substantially below the design stresses. I 20 feel that the -- I could check on this point.
2]
DR. MONSON:
All right.
You are saying that at 22 Oconee sufficient tests were rup to demonstrate conclusively 23 that the analytical methods employed were correct and that 24 those same analytical methods are used on the present plant; te-Federal Reporters, Inc.
l 25 that in addition there will be confirmatory tests, actual
~
44 jcn I
experimental evidence shown as to what the vibration might i
2 be in this plant to show that it conforms with the predicted results?
4 MR. BUCKLEY:
Yes, sir.
That is exactly correct.
5 (Board conferring.)
6 MR. DAVIS:
Could I correct something I said 7
earlier with respect to Dr. Monson's question on peak kilo-0 watts per foot?
9 I think one of B&W's people. Bob Turner, would 10 probably have some*.hing to say further on this.
II The 19.1 number, kilowatts per foot, that we 12 Lw discussed was not their anticipated achievable per foot.
13 That was what B&W considers the LOCA limit to produce that 14 peak clad below 2300 degrees.
That was the LOCA limit.
15 If my calculations are right, they would 16 anticipate an achievable peak kilowatt per foot of something 17 like 18.1 or 18.2 kilowatts per foot.
They would have a 18 margin below the LOCA limits.
l9 DR. MONSON:
By achievable, do you mean the 20 Applicant is proposing to operate the reactor in such a manner 2I that a peak KW per foot, higher than 18.5, does not exist?
22 MR. DAVIS:
That is.my understanding.
23 I believe there will be more discussion on this 24 from the Applicant.
ce-Fe7c al fieporters, Inc.
25 LR.' SIESS:
Could I clarify another definition?
45 jan
'I When you say the LOCA limit, you mean working back-2 wards from 2300 to find out what you can take with the interir 3
acceptance criteria in mind?
4 MR. DAVIS:
Yes, that's correct.
5 In this particular case 19.1 was 2286 degrees.
6 I think the B&W staff said that was close enough to 2300 and
~
7 just established that as a limit.
O CHAIRMAN STRATTON:
That is close enough.
9 I think it would be worthwhile if we all took 10 more than five but less than ten minutes to stretch our legs Il perhaps and find a cup of coffee.
12 We will have a short coffee break, if you wish, 13 for seven or eight minutes.
cnd4 I4 (Recess.)
1752 15 16 17 18 19 20 21 22 23
~
24 ce-Federal Reporters, Inc.
25
~
U '
" * * ' ' * ~ * -
46
- 5 Orl 1
CHAIRMAN STRATTON:
May we como to order again, l
P ease?
2 3
I would like to announce one item that I neglected 4
to earlier this afternoon.
The written statements that we 5
receive, we will -- the subcommittee will review this 6
evening; and the contents of those statements will be given 7
to the stenographer to read into the official record of the 8
meeting.
9 I think ow we can go aheed, Bob, if you will; 10 let's move on to a description -- description of the safety 11 systems and analysis.
12 MR. WILSON:
Yes, sir.
h 13 In this, do you desire a brief rundown on what 14 the safety systems consist or, or are you referring to the 15 analysis only?
16 CHAIRMAN STRATTON:
Primarily the analysis,but 17 you better cover what they consist of very briefly as you go 18 along, please.
39 MR. WILSON:
Yes, sir.
20 The safety systems for emergency core cooling 21 are divided into three basic systems.
Each particular system is associated with the various accidents that have been 22 23 analyzed and are for different portions and different types 24 of the accidents, depending upon the temperature pressure Ace-Federal Reporters, Inc.
conditions.
The high pressure conditions and the low pressure 25
47 ar2 I
conditions and the core flooding provide the full spectrum 2
for all systems.
3 Separate and independent flow paths are provided 4
for the emergency core cooling systems and redundancy of 5
active components are provided to insure the required func-6 tions will be performed in the event of a single failure 7
occurring.
8 Separate power cources are also provided to the 9
redundant active components and separate instrument channels 10 1re used to actuate the systems.
11 The high pressure injection system consists of 12 two high pressure injection pumps, each in separate high 13 pressure injection trains, injecting water into the reactor 14 coolant system in the event of a postulated reactor coolant 15 break.
16 This system is actuated on a high building pressure 17 or on a low primary system pressure.
At 1600 psi primary 18 system pressure, the high pressure injection pumps would 19 inject 300 gallons per minute of water into the primary system.
20 The low pressure injection system,also actuated 21 at 1600. psi primary system pressure and also a high building 22
, pressure, injects water into the primary system under low 23 pressure conditions such that at approximately 100 psi 24 designed flow is 3000 gpm per low pressure injection train.
Ace-rece,ai nepo,t,3, ine.
25 The core flooding system consists of two tanks with
48 dr3 I
a piping system which has check valves in it so that it is a-2 completely passive system.
3 The on_1y thing required for this system to 4
actuate is for the primary system pressure to reach 600 5
Pounds.
The check valves will then open because the core 6
flooding tanks are pressurized at 600 pounds and the core 7
flooding water would be injected into the reactor vessel.
8 CHAIRMAN STRATTON:
The differential pressure 9
across that valve can be very, very small?
10 MR. WILSON:
Yes, sir.
A few pounds and it will 11 OPen.
It is a 14-inch -- 14-inch line, I think the cross 12 sectional area, equivalent diameter of 10 inches.
13 CHAIFWUW STRATTON:
What kind of a valve is it?
14 MR. WILSON:
It is just a standard check valve.
15 CHAIRMAN STRATTON:
Okay.
Go ahead.
Okay.
16 MR. WILSON:
The analysis for emergency core 17 cooling for a postulated loss-of-coolant accident has been 18 performed and is discussed in Section 14-2 and in Appendix 14 (a) 19 of the FSAR.
20 Included in this, in Appendix 14 (a), is the responr-
)
i 21 to recent AEC questions on performing the analysis for the 22 large break, small break, and core flooding line break in
{
23 accordance with the interim acceptance criteria and with 24 additional conservative restrictions placed on the analysis.
Ace-Federal Reporters, Inc.
25 The results of the core cooling analysis for the hw-^-
-e
49 ar4 I
postulated large reactor coolant piping break and the small 2
reactor piping break and the cold flooding line break demon-3 strates that the emergency core cooling system will terminate 4
cladding temperatura transient and limit the course of the 5
loss-of-coolant accident in accordance with the AEC interim 6
acceptance criteria.
7 CHAIRMAN STRATTON:
Why did they have new calcula-8 tions?
I didn't understand that.
9 MR. WILSON:
I would like to leave that up to 10 the AEC.
11 CHAIRMAN STRATTON:
Okay.
12 Maybe, if you would, you can go through -- I would 13 like to get a feeling for some of the assumptions that are 14 involved in these analyses.
15 MR. DAVIS:
The --
16 CHAIRMAN STRATTON:
I am sorry.
Go ahead.
17 MR. DAVIS:
The -- basically the calculations we 18 asked to be redone were a large line break with anasumpt-iun 19 of more conservative assumption on entrainment of reflooding ne<
20 liquid using a carry of-a rate fraction correlation that ~
21 B&W developed and presented in the ECCS rulemaking hearing.
22 We requested that they analyze the small breaks, 23 r a small break, for the Rancho Seco power level in place 24 of the 2568 megawatts thermal which was analyzed in B&W's Ace -FHeral Reporars, Inc.
25 topical report on small breaks.
50 ar5 1
CHAIRMAN STRATTON:
Excuse me.
Back to the entrain-2 ment,,did you say that this entrainment assumption was 3
one that was developed by B&W?
4 MR. DAVIS:
Yes.
It was a correlation developed 5
by B&W in the course of the ECCS.rulemaking hearing.
6 CHAIRMAN STRATTON:
And the correlation was examined 7
by you and was deemed better than what had been used 8
previously or just more conservative?
9 MR. DAVIS:
Better and :aore conservative.
10 CHAIRMAN STRATTON:
Okay.
That's good if you can 11 get it both ways.
12 (Laughter.)
13 MR. DAVIS:
Also the analysis of the core flooding 14 tank line break was asked to be reanalyzed at the slightly 15 higher power level for Rancho Seco, over that supplied in Gd 16 B&W topical report BAW-10,046'.
17 CHAIRMAN STRATTON:
The small break you wanted to l'8 have done at 2772 megawatts instead of 25687 19 MR. DAVIS:
Yes, that's correct.
20 CHAIRMAN STRATTON:
What was the reason for the 2) core flood break?
22 MR. DAVIS:
That was 'the same reason.
23 CHAIRMAN STRATTON:
Okay.
24 MR. DAVIS:
We analyzed it at slightly higher power.
~
%ce-Federal Reporwis, Inc.
25 DR. SIESS:
The BAW's - - the topicals you referred
51 ar6 1
to were for the Oconee reactor?
T'> l 2
MR. DAVIS:
Yes.
At 2600 megawatt thermal.
3 DR. MONSON:
Is it correct that you have no reason 4
to think that the results of those three types of analyses 5
are likely to change your present determination that it is s
6 safe to operate at full power?
7 MR. DAVIS:
Well, they have supplied the requested 8
analyses and our review is not yet complete.
Our initial 9
review position is that it would not' be for a 100 percent 10 power operation.
11 DR. MONSON:
This is a change'from the Safety 12 Evaluation Report in which it said the information had not yet 13 been submitted?
1 14 MR. DAVIS:
Yes.
It has just recently been sub-15 mitted.
Several days.
16 DR. MONSON:
That's as long as we have had this 17 report.
They write the report a little.
jo earlier than we received it.
20 CHAIRMAN STRATTON:
Go ahead.
21 MR. WILSON:
That was my statement.
Do you have 22 more specifics?
23 CHAIRMAN STRATTON:
I would like to hear a little 24 bit about the assumptions that go in, and particularly those l
Ace-Federal Reporkrs, Inc.
25 that involve the core barrel, steam binding, and so on.
How
52 ar7 I
do these affect the results?
I think it is only the large 2, break.s that matter.
The others are related to load temperature, 3
aren't they?
4 MR. WILSON:
Yes, sir.
In most cases, almost 5
insignificant 1y low.
In one case the core remains flooded; 6
so that the temperature never increases.
It starts decreasing.
7 CHAIRMAN STRATTON:
What about the core flood tank 8
line break?
What is the final temperature on that one?
9 MR. WILSON:
That, I believe -- I will check the 10 statement -- I believe that was one where the -- when the 11 accident occurred, the flow was suffi.cient to where the core 12 remained covered in that the temperature calculation itself 13 was not calculated because the temperature decreased.
14 I will confirm that.
15 CHAIRMAN STRATTON:
- Okay, I thought there was 16 a small break --
17 MR. WILSON:
In the small break I know that was i8 the case.
The temperature calculations were not plotted 19 because the temperature remained constant or decreased.
20 I will check this.
21 So on both cases -- this is in Appendix 14 (a).
22 For -- in Appendix 14 (a), we responded to AEC 23 questions on this, for both the small leak analysis and the 24 core flooding line analysis, the core remained covered and Sce - Federal Reporters, lnc.
.25 the temperatures were therefore decreasing.
1
ar8 1
Particular interest on the core flooding line 2
nozzle -- core flooding line break, the restricting warp 3
that is in the nozzle is basically why we see such an improve-4 ment in this case, with the core remaining covered.
5 CHAIRMAN STRATTON:
Okay.
6 Chet?
7 DR. SIESS:
I am confused.
I am reading something 8
that says for the core flooding tank line break, the core is 9
not always covered; but the water level is high enough that 10 you have the steam cooling above it.
11 Is this new information?
12 MR. DAVIS:
If I may clarify a little bit, the 13 description in the Rancho Seco report is that af the Oconee 14 analysis; and that was the only analysis that was available 15 at the time that was written; and as it turns out, for 16 Rancho Seco, both the higher power level produces a higher
.'. O-17 mixture live and also a somewhat less conservative assumption l'8 was made in the analysis -- still conservative, but less 19 conservative than that made in the Oconee analysis; so that 20 the end result was B&W's calculations for Rancho Seco indicated 1
21 that there was no uncovering of the core.
i 22 CHAIRMAN STRATTON:
Ny question about the vent J
R.-
~
23 valves.i.s on page 617 of the Staff report, addressed to 24 either one, I guess.
The steam generator in the core, together l
%ce -Federal Reporters, Inc.
25 with the entrained liquid, is assumed to flow only through
54 ar9 I
the vent valves within the reactor vessel.
No credit is taken 2
for steam flow around the loop to the breaks.
3 Is this the way you did your evaluation?
4 MR. WILSON:
That's correct.
l 5
CHAIRMAN STRATTON:
At the Staff's direction?
6 MR. WILSON:
I don't know where it originated, but 7
I know that's where it was done.
8 CHAIRMAN STRATTON:
All right.
9 Suppose you didn't have the vent valves,as some 10 similar reactors don't.
How would we calculate it, and what 11 difference would it make?
12 MR. WILSON:
I would prefer not to answer.
I 13 don't know we have done that on our particular project.
I 14 don't know if B&W wants to comment.
15 MR. GLEI:
I might comment that the problem in 16 relation to steam binding with -- if you don't have vent 17 valves in a plant is -- it comes about in relation to the 18 piping -- the loop piping arrangement; and I believe that if 19 you are talking about other -- let's say, B&W plants, which 20 are non-vent valve: plants,-they have a different loop 21 arrangement than Rancho Seco.
CHAIRMAN STRATTON:
Gould you help me out on how 22 23 this loop is different, one to the other?
I guess I don't 24 know.
Ace-Federal Reporters, Inc.
25 MR. GLEI:
Yes.
55 ar10 l
CHAIRMAN STRATTON:
Could we have a blackboard 2
sketch or a slide?
3 MR. WILSON:
Basically the pressure vessel is 4
lower.
5 I could draw it up higher, I think.
6 I don't know how you can get this into the record.
7 (Laughter.)
8 B&W can correct me because I am drawing their 9
systems here.
10 One of the B&W units has the reactor vessel lower 11 down into the cavity; and there is a very short lag.
12 (Indicating.)
13 The -- SMUD's unit has a higher -- this is the 14 unit -- okay.
15 This is good, because he has it split, one on 16 one side, and one on the other.
17 (Indicating.)
18 on one type of a plant -- let's say this is the non-19 vent valve type plant, reactor vessel is lowered and comes 20 out to the reactor coolant pump.
2j (Indicating.)
22 This is out a very small change in elevation s.-
23 from steam generator to the reactor coolant pumps to reactor vessel.
24 Ace-Federal Reporters, Inc.
25 (Indicating.)
4 mm i
56 ar11 1
The other design which is similar to Rancho Seco 2
comes out to a reactor coolant pump, which sets on a very 3
high leg for the suction side of the pump and goes into the 4
5 (Indicating.)
6 The difference in this case is the loop 7
characteristics for the cold leg.
In this case we have the 8
vent valves -- and our analysis did not get credit for any 9
steam flow through the loop when we do the loss-of-coolant 10 analysis.
11 In this non-vent valve reactor case, the reactor 12 vessel is relatively higher with respect to the steam generator G:
13 and the full characteristics are different.
14 (Indicating.)
15 I think B&W could give you a more elaborate 16 description of the hydraulic characteristics of the two 17 systems.
c5 18 19 20 21 22 23 24 Ace - rede,ai nepo.wes, :nc.
25
~ = - - - -
-.,we,,
57 CR 1752 1
CHAIRMAN STRATTON:
I guess I am not convinced, Dn6 2
but it must be all right.
R0ba 1 3
MR. DAVIS:
Maybe I could help clarify.
I think it
~
4 was the Staff that probably forced B&W to make the assumption f ',w ay 5
there was no steamfitting through the cold legs and hot legs i
6 of the loop.
7 The problem is that if you assume the lower cold leg 8
from the pump suction does not clear with water, then you do 9
not have a path for
- eam to flow.
In other words, the water 10 in the cold' leg seals the., system, and you cannot force steam 11 through it because the height of the water that is driving 12 steam through the system is roughly the height of the downcomer, 13 and the seal pipe is greater than that height, then and there-14 fore, water or steam will not flow through the loop.
15 With the vent valves you do not have to flow through len 16 the loop.
You can flow to the up or downcomer annulus and 17 out the break.
18 CHAIRMAN STRATTON:
I guess I follow.
Okay.
1 19 (Board conference) j 20 CHAIRMAN STRATTON:
Let's go back to the break l
21 assumed in the core flooding tank line, and you say the core 22 does not become uncovered.
23 MR. WILSON:
Yes, sir.
24 CHAIRMAN STRATTON:
In this calculation.
Au-federal ReporWrs, Inc.
~
25 Are the new assumptions ~that the Staff requested less r
58 Den 6 1
conservative or mora conservative in this case?
Reba 2 2
Is that the question?
3 MR. LIBARKIN:
At one point we are talking about r,\\
b 4
changing the position of the blowdown.
Is that what you had 5
in mind?
6 MR. DAVIS:
No.
The difference between the Oconee 7
calculation, where there was a slight clad heat up, around 8
1200 degrees, was what was calculated for Oconee.
For Rancho 9
Seco, since the' core was covered with a mixture, a calculation 10
-- clad was not calculated to rise above its initial temperature.
11 The differences are two-fold.
One is the higher power level 12 at Rancho Seco, 8 percent higher, because a mixture heighth 13 to be -- a higher mixture height to be generated for the same 14 volume of water left in the core.
15 This is a cwelling of water or frothing which is 16 familiar to most boilers.
The other assumptions that were 17 different -- I have to refer to the analysis -- we can get 18 back to this in just a moment.
19 CHAIRMAN STRATTON:
Okay. Go ahead.
20 DR. MONSON:
While you are looking that up, may I 21 ask the Applicant, isthedOOpsicorefloodingtankpressure 22 the same as the initial pressure' proposed?
23 MR. WILSON:
Yes, sir.
24 DR. MONSON:
Okay.
Ace - FHeral Reporkts. Inc.
~
25 CHAIRMAN STRATTON:
Another question while we are O
e.
59 Den 6 I
waiting for Mr. Davis.
l Reba 3 2
I noticed in the FSAR the analysis of naximum 3
temperature as a function of break size assumed in the line was 4
not too much different from the largest breaks, say a half a 5
square foot, to six square feet, but then it started to drop 6
very rapidly as the break became progressively smaller, and 7
the break size which corresponds to a couple square feet, it 8
was really quite small, the maximum temperature.
9 Is this correct?
Do you have that?
10 MR. WILSON:
We have the figure.
11 CHAIRMAN STRATTON:
I don't remember the figure.
12 My_cuestion relates to that figura 4"-tha-env ^-that,whaF M d 13 th_e cha_r,acteristics-of--the-respumive-rectiesbe ikMf~you.
14 show1A. assumer-wruptuwH= ^ran += 11ti.veJ.y-a.k+1yese=t Aat 15 far tha f 4 rc:t c:b a l l ue
?y cnnpla of M h u na. h A o
16 or--sMiYgMTToWd 1ike tne smafis'r~TireaR~hite7areti:en 17 eventr:lly li. epene&upto a langer -heanic_ _sW7 18 9"
4-t' i
.. 1,---^'^#'-
..ima - r u rzMs 19 gopg) tmpture7We r iE =is yL1+tn.runture.
20 MR. WILSON:
Greg, have you looked at that?
21 (Board conference) 22 (Applicant conference) 23 MR. WILSON:
I ~.. ' L k..bw'*thtr"dhTwes-ta.that.r Qr. _.
24 Sisesatton.
I"'---
Ace - Federal Repor ters. Inc.
25 CHAIRMAN STRATTON:
Fine.
60 Den 6 I
MR. DAVIS:
Apparently the only difference was the R ba 4 2
difference in m.e power level.
They present another calculation 3 ' with a slightly different assumption, slightly less conservative 4
assumption.
It would yield more liquid in the core, but in both 5
cases the core is covered with a two phase mixture throughout 6
the transient, so the other less conservative assumption that 7
they use predicted a higher water level but both assumptions, 8
the ones used for Oconee and the one used for Rancho Seco, 9
were consistent.
10 DR. SIESS:
You are confusing me.
You say water 11 level in one instance and two phase mixture in the other.
12 The figures on page 6-25 for the Oconee andlysis, the Oconee 13 reactor analysis, those figures are also based on the two phase 14 mixture level?
15 MR. DAVIS:
The results ---
16 DR. SIESS:
When you say liquid level in the Staff's 17
. Safety Evaluation Report, you mean two phase mixture?
18 DR. DAVIS:
No.
I mean liquid level.
Two phase 19 or swollen height is the same meaning. Would you give me a 20 specific and I will see if there is a typographical error?
21 DR. SIESS:
I will read it again and see if I am 22 still confused.
23 CHAIRMAN STRATTON:
Review for me again, Bob, the
_'y,3, %
~
24 purpose of this connection of the low pressure engine lines Ace - ree.,.i a.po,a,3, ine, 25 that you showed to us tnis morning.
61 Den 6 1
MR. WILSON:
The cross tie?
R bn 5 2
CHAIRMAN STRATTON:
Yes, the cross tie.
3 The reason for it and how it was done in this s
4 instance.
5 MR. WILSON:
We have a cross tie and have always 6
had a cross tie in the discharge side of each low pressure 7
injection train.
It ties the two low pressure injection trains 8
cogether at a point downstream of the decay heat coolers.
9 Under normal conditions this corss tie connection 10 has two normally ciosed valves.
The -- this particular system 11 is being modified slightly as indicated in Mr. Buckley's 12 carlier statement by incorporating the flow element in the 13 cross tie.
14 This provides us additional flexibility for the 15 operator in the event of a postulated loss of coolant accident 16 of any nature, including core flooding line rupture, in which 17 one of the low pressure injection trains is una'&ilable for v
18 some reason, such as a single failure of a motor or diesel 19 generator.
20 In the event this should occur, and the operator l
21 sees that one system is unavailable, the cross tie valves would
[
22 be opened and the flow would be balanced using the remaining 23 low pressure injection pump, discharging them through both 24 trains into the primary system such that there would be at Ace-FNerat Reporters, inc.
25 least 1500 gallons per minute of flow going through each train.
-e,_
62 Den 6 1
This would assure, for long term cooling, that Reba 6 2
there was abundant cooling water available to protect the 3
core.
4 CHAIRMAN STRATTON:
I guess I missed the basic 5
reason for putting'in this cross tie.
You are assuming a single 6
failure where in the system?
7 MR. WILSON:
One of the items that is postulated 8
is the decay heat piping, the low pressure injection piping, 9
discharges into the core flooding piping, and if you consider, 10 as an example a core flooding line rupture, and impose upon 11 that a single failure in the opposite train, then you can use 12 the cross ties for long term cooling, even if a 1500 gallons 13 per minute goes out the break.
14 At least 1500 gallons per minute goes into the 15 alternate low pressure injection train and the associated core 16 flooding tank piping to provide adequate cooling.
17 MR. MATTIM,OE:
Otherwise you would be just pumping 18 through the break.
19 CHAIRMAN STRATTON:
My difficulty is I don't have 20 a clear picture of that tangle of pipes in my mind so I am 21 having difficulty visualizing the geometry.
22 I will try to review this.
23 MR. MATTIMOE:
We can make a little sketch.
24 MR. WILSON:
I can provide you a simplified sketch
~
Ace-Federal ReporWes, Inc.
25 of that.
s
~.
63 Den 6 1 CHAIRMAN STRATTON:
If you would tomorrow perhaps.*
R:ba 7 2
DR. MONSON:
Does this require valve realignment 3
after the event?
4 MR. WILSON:
Yes, sir.
5 MR. MATTIMOE:
Yes, sir.
6 DR. MONSON:
Within how many minutes after the 7
assumed event?
8 MR. WILSON:
It would be required, operationally 9
desirable, to have it accomplished before you would go into 10 recirculation.
You have on the order of 40 minutes, 45 minutes 11 to accomplish this.
12 This particular procedure has been walked through 13 and we have performed it with an operator, and it took about 14 ten to twelve minutes to perform.
15 DR. MONSON:
But if you should get a core flooding 16 tank line break close to the reactor vessel, and you have 17 a single failure in the other LPI train, the operator must take 18 this action within approximately 40 minutes, is that right?
19 MR. WILSON:
No.
Asffar as it being technically 20 required to be when it is flooded with respect to the core, 21 I don't know the answer to that.
22 I think the imposition has been placed on us by 23 the Staff that it be accomplished prior to going into recir-24 culation.
From an operational standpoint, this appears to be Ace -Federal Reporters, Inc, 25 more desirable because once you go into the recirculation u
64 Den 6 1
mode, and you are taking. water. -- instead of from the borated R:ba 8 2
water storage tank, you are taking water from a sump, this 3
could be a fairly high radiation.
4 DR. MONSON:
You have to have -- at sometime after 5
this particolar operation, you move at least 1500 gallons per 6
minute LPI flow in order to cool the core properly.
Is that 7
correct?
8 MR. WILSON:
No, sir.
9 I would defer that to the Staff, but I think it is 10 on the order of 500 GPM that is technically acceptable.
11 DR. MONSON:
But the -- let me clear it up -- is 12 the 500 GPM, however, more than you would get if the operator 13 did not realign valves and you had this CFT line break and 14 the opposite train, LPI, single failure?
15 MR. WILSON:
That is correct.
16 DR. MONSON: Okay. Then I come back torthe question, l
17 within how many minutes after this event occurs must this valve 18 realignment be accomplished in order to properly cool the 19 core?
20 MR. DAVIS:
I think there was a m. stake in your 21 answer.
The 500 GPM approximately is supplied without any i
22 maneuvering of the low pressure lnjection system.
This comes 23 from the high pressure injection system.
24 So if you were to look -- for example, the analysis Ace-FHeral RepotMrs, Inc.
25 that is presented in the application in Section 14(a) does not e
65 Den 6 1
assume any low pressure injection water at all.
R ba 9 2
It only assumes injection water from the one good 3
core flooding tank and the one high pressure injection pump.
4 DR. MONSON:
It is the high pressure injection pump 5
-- is it set up to operate on a long term basis?
d.wk 6
MR. DAVIS:
I would have to -- I notice in some 7
plants it is.
8 MR. WILSON: It is in our case.
9 (Simultaneous discussion) end 6 10 CR 1752 11 12 13 14 15 16 17 18 19 20 21 22 l
23 l
24 l Ace -Federal Reporters, Inc.
l 25 I
DD #7 66 ty 1 1
MR. WILSON:
In our particular case we have a 2
crossover from the high pressure and low pressure.
We can 3
operate on the recirculation mode using high pressure 4
injection as well.
It stays -- it continues 5
DR. MONSON:
Is it correct in the event this 6
particular accident should occur that even if the operator 7
never realigns these valves that we have been disucssing, 8
just so long as one high pressure injection pump is in 9
operation, the core will be adequately cooled?
10 MR. WILSON:
That is affirmative by our calculations.
1]
DR. MONSON:
On a long-term basis?
12 MR. WILSON:
Yes, sir.
13 DR. MONSON:
You earlier, I believe in your 14 discussion, have indicated that there were two high 15 Pressure pumps; but there are three; is that correct?
16 MR. WILSON:
No, sir.
There are two high pressure 17 injection systems and there is a makeup pump which is our 18 normal operating makeup pump.
It can be tied in to either 19 high pressure injectionsystem, but normally does not 20 receive its actuation.
In the event one high pressure injection 21 Pump is out of commission, we would transfer that makeup 22 Pump over to mechanically and electrically into the --
23 DR. MONSON:
You have two plus a standby?
24 MR. WILSON:
Yes, sir.
ce-Fedes11 Reporters. Inc.
25 CHAIRMAN STRATTON:
And the makeup pump would
ty 2 67 I
be running most of the time?
2 MR. WILSON:
Yes, sir.
It would run continuously.
3 CHAIRMAN STRATT N:
Okay.
4 The temperature that you get out of your line 5
break analysis depends upon the peak kilowatts per foot.
6 Could you give me any feeling as to how often, if ever, you 7
expect to operate the plant so as to see, say, 19 kilowatts 8
per foot; how often would you see and how long, 18; how long 9
would you see 17, for example?
10 What is your --
11 MR. WILSON:
Are you talking to the Applicant 12 as an operator or --
13 CHAIRMAN STRATTON:
Yes, I am.
14 MR. WILSON:
In that regard, the technical 15 specifications that presently exist, that we have proposed, 16 do not include the modified densification technical 17 specifications.
They have not been developed at this time.
18 In the modified -- if I can call them densified 19 technical specifications, the limitations on the peak 20 kilowatt per foot will be defined.
21 CHAIRMAN STRATTON:
Will be defined?
22 MR. WILSON:
It will,be defined in there as an 23 operating restriction.
I do not know what that is yet.
24 Bob, do you have anything tha't you can add to co-Federal Reporkts. Inc.
25 that?
www
68 ty 3 1
MR. TURNER:
Nothing that I can add.
2 MR. WILSON:
Perhaps we can cover that tomorrow.
3 CHAIRMAN STRATTON:
We keep coming back to this 4
sooner or later.
5 Is the relative density of the fuel in the B&W fuel 6
a proprietary matter?
7 MR. WILSON:
No.
It is in the FSAR.
8 CHAIRMAN STRATTON:
Right. This is stated as 9
being.95.
10 MR. WILSON:
Ninty-five percent theoretically, 11 yes, sir.
12 CHAIRMAN STRATTON:
And the Staff required you to x
13 calculate as if it had densified immediately to.965; is that 14 right?
15 MR. DAVIS:
That is right.
16 CHAIRMAN STRATTON:
This is not a very big factor.
17 I guess the technical specifications as now written would 18 be almost right for the point I have in mind.
19 What do they say about operation at some given linear 20 power now?
I haven't read them.
I am just trying to get a 21 feeling.as to how often and how long the plant might be 22 operated with a fairly extreme power density.
23 MR. WILSON:
The peaking factors that are defined 24 in technical specifications typically provide the extremes tham
~;e - Federal Reporters, Inc.
25 are acceptable from the safety standpoint, and these take O--
69 ty 4 1
into account abnormal plant operations that would get you 2
into that peaking condition.
3 Once you get through the transient that got you 4
through this peak and you get back into steady state 5
operation, you wouldn't see these peaking factors.
~
Typically, the great majority of the time, 95 6
7 percent of the time, you would be much lower than this 8
limiting of peaking factors that the safety analysis was 9
based on.
10 CHAIRMAN STRATTON:
Rancho Seco will be operated jj as a base loaded plant?
12 MR. WILSON:
"es, sir.
v 13 CHAIRMAN STRATTON:
So we will not really have 14 changes of any magnitude often, at least not for some time?
15 MR. WILSON:
No.
16 CHAIRMAN STRATTON:
And the peaking factor in this 37 high kilowatts per foot come really only at the new core?
18 MR. WILSON:
Beginning of life for the new core, 39 that is right.
20 CHAIRMAN STRATTON:
Do they return when you load
~
fuel?
21 MR. WILSON:
I believe it is lower.
I believe that 22 s
23 has been defined in the topic.
I don't know if we have 24 that in the tech specs.
~
co - rw.ci n.po,w,3, ine.
25 Bob, could you state what happens to the peak
70 ty 5 I
linear heat rate at the beginning of life peak for the 2
first cycle core compared to what an equilibrium core would 3
be, as a change for the equilibrium core?
4 MR. TURNER:
The Rancho Seco plant is B&W's first 5
plant that is maneuvered on boric acid on a feed and bleed 6
system.
Inherently, the peaking factors are considerably 7
lower in this type of operation than it has been in our 8
previous plants where we have maneuvering on control rods, 9
and the peaks in the Rancho Seco plant, as we have evaluated 10 them, 'really are rather constant on the lifetime, after one 11 gets equilibrium xenon built into the core.
12 When Mr. Wilson was talking about the peak linear 11 heat rates during the maneuvering operation, the reason 14 that we defined our limit at this 19.1 level is because that 15 is the -- that is our LocA limit.
However, this limit is 16 modified by 9.2 percent in order to account for a 5 percent 17 quadrant tilt.
In other words, when one is operating at a 18 particular time, unde,r tech spec regulations you do not --
19 you are allowed to operate with 5 percent quadrant tilt.
20 Therefore we take the LocA limit, reduce it by 9.2 percent, 21 and then that gives us -- this is the number that Mr. Davis 22 mentioned, about 18.2 kilowatts'per foot.
23 However, what you see is that the operating limit 24 is still 19.1.
What you do is go back and reduce the
, l
- e-Federal Reporters, Inc.
25 operational limits by administrative procedures, rod
71 ty 6 1
indexing -- rod indexing system such that the maximum peak 2
during operation, during one of these transients, would not 3
exceed 18.2 kilowatts per foot.
4 As he said, during the lifetime of the plant, 5
under base load operation, the peak kilowatts per foot are 6
considerably less than 18.
The safety limit or the LOCA 7
limit is still 19.1.
8 CHAIRMAN STRATTON:
What would you expect as the 9
peak kilowatts per foot during operation?
Considerably less 10 is pretty --
11 MR. TURNER:
It is on the order of about 15 12 kilowatts per foot.
13 CHAIRMAN STRATTON:
You would expect to see 15 14 quite often?
15 MR. TURNER:
Yes.
16 DR. MONSON:
You say the difference between the 17 15.1 and the 18.2 is. represented by the dif ference for a 5 18 Percent tilt?
19 MR. TURNER:
Yes, sir.
20 DR. MONSON:
Have I not read that. ycu believe the 21 sensitivity of your detectors is only about 5 percent?
~
(
22 MR. TURNER:
The sentitivity, the error that 23 one associates with the detectors is also included in the 24 imbalance trip system that we will talk abor.c tomorrow.
- e - Federal Repor ters, Inc.
25 In other words, we calculate allowable core 9
..--,.. -. m
_.-n~.mnn+.-
.,_m
72 ty 7 rl 1
offset limits based on fuel melt criteria, D&B criterion, 2
and thia LOCA limit, and then these allowable offset limits, 3
as one would see in the core, are modifield by what one would 4
expect to see on the outer core instrumentation.
5 We apply an instrumentation error between what goes 6
on in-core and what goes on out of core and that would be 7
demonstrated in the morning.
8 DR. MONSON:
Don't these calculations assume the 9
actual existence of a 5 percent tilt?
10 MR. TURNER:
Sir?
11 DR. MONSON:
Do not your calculations actually assune 12 the existence of a maximum tilt of 5 percent?
13 MR. TURNER:
In the limits that we set we assume 14 a 5 percent tilt because in the technical effects --
15 DR.,MONSON:
Is that an actual tilt or that is 16 what you measured the tilt to be?
Which?
And hope you have i
17 measured the actual tilt.
18 MR. TURNER:
That is an allowable limit that we put 19 in there.
20 In other words, I can't remember how far back this 21 started on radial tilt situation, but we got into a discussion 22 with the Staff on establishing driteria, and then turning 23 around and trying to find out, well, how do you know that 24 you don't have a radial tilt.
i to-Federal Reporters, Inc.
25 We decided then that the.5 percent radial tilt was
__.J,__._,
73 ty 8 1
very reasonable to see on your outboard instrumentation.
2 Therefore we used that as the factor that we would subtract 3
from our allowable limits to assure us that we were not 4
operating outside of tech spec limits.
5 Now if one drops below -- and we will have to revieu this -- I don't have the numbers exactly -- but typically if 6
7 the radial tilt were discoverede one would have to reduce --
8 let's say it is greater than 5 percent, then the tech specs.
4 9
will reflect the power level to which one will reduce to 4
10 cover this power -- this tilt.
11 Then you see this protects us, this covers us again 12 against our LOCA criterion.
13 DR. MONSON:
Well to measure the tilt in the 14 reactor, you measure the flux level in effect on one side, 15 you measure it on the other side, you take the difference 16 between the two; is that correct?
17 MR. TURNER:
No, sir.
jg DR. MONSON:
No?
19 MR. TURNER:
I will have to review that this 20 evening.
I haven't reviewed that in some time, but I can't 21 answer that question now.
22 DR. MONSON:
All right.
We can take it up later.
23 The thing I am interested in finding out is how accurately 24 y u really know what the tilt in the reactor is,.
Your m -Federal Reporters, Inc.
25
.inrtruments can only measure just so well, and what you are o
-A..e j
~
74 ty 9 I
doing is taking, I think, the difference between two flux 2
levels to get at the tilt.
A little bit of inaccuracy in your 3
flux readings gives you a problem in determining with 4
precision what your tilt is, and so if you could be ready 5
to address that tomorrow or sat the full committee meeting.
6 MR. TURNER:
I will look at at least three of these 7
outboard converters, most of the time four.
But I will 8
address that tomorrow.
We also use our in-core system.
9 CHAIRMAN STRATTON:
Thank you.
10 More on the loss-of-coolant analysis.
11 DR. MONSON:
The 2281 F peak clad temperature, 12 which I believe relates to the 8-1/2 square foot cold leg 13 split is calculated without including any densification 14 effect; is that correct?
15 MR. WILSON: That is correct.
16 DR. MONSON: Thank you.
17 CHAIRMAN STRATTON:
Do you have anything more on 18 this, Chet?
End #7 19 DR. SIESS:
No, thank you.
20 21 22 23 24
- e-Federal Reporters, Inc.
25
= - - - - _ _ _. - _ _
75 DD,#8 mm1 cr1752 1
MR. WILSON:
I might add on that particular one, 2
the recently submitted fuel densification report includes the 3
revised calculations for the LOCA with densification included.
4 That is in.the report that was submitied just a few 5
days ago.
The Staff has not had an opportunity to review that.
CHAIRMAN STRATTON:
What was the number of that 7
report?
8 MR. WILSON:
What is the number of that?
9 MR. TURNER:
1593.
BAW1593.
10 DR. MONSON:
Is that proprietary?
11 MR. WILSON:
That is proprietary, 12 CHAIRMAN STRATTON:
We will, I expect, turn to these 13 matters tomorrow.
14 Would you discuss the containment pressure calcula-15 tions now, including some comments on the factors that matter, 16 break size, and talk about the presssures within some compart-17 ments, if you would..
MR. WILSON:
Yes, sir.
19*
In summary, the reactor building pressure transients 20 for the loss of coolant acidents have been evaluated and are j
21 included in Section 14 of the FSAR for a complete range of 22 rupture sizes with the reactor initially operating at 23 power levels of 2772 megawatts thermal core power.
24 A Bechtel computer code has been used to perform La-Rwera Reprbers, Inc.
25 this analysis which assumed that the core cooling is provided
-+--=-.._.___.--.._,.,._m,,
76 I
by two core flooding tanks and a single emergency core cooling mm2 2
system train.
3 The ECCS train is assumed to operate on emergency 4
power.
5 In addition to subcompartment differential pressure 6
in addition to the reactor pressure analysis, subcompartment 7
pressure analysis was done for the reactor cavity and the 8
steam generator compartments.
9 For the limiting cold leg break, which is the jo 8.5 square foot break, the peak building pressure is 48 psig 11 occurring in approximately 16 seconds.
For the limiting hot 12 leg break, which is five sugare feet, the peak building pres-(
13 sure is 52 psig occurring in approximately 21. seconds.
i 14 The reactor cavity compartment pressure calculations 15 f r the 8.5 square foot hot leg break, which is a limiting 16 break, is 240.5; and for the 8.5 square foot cold leg break, 1
j7 it is 230 psig.
18 The steam generator compartment, the worst break is 39 a 14.1 square foot hot leg break provided 14.3 peak pressure.
20 CHAIRMAN STRATTON:
Where is this?
21 l MR. WILSON: This is steam generator compartment and 22 this is for the hot leg break.
23 These compartments are differential pressures.
24 The complete analyses for these cases are defines Ace - Federal Repor ters, Inc.
~
25 in Section 14.4 with appropriate transient, curves, pressure s
m,.
77 MM9 3 1 included.
The design pressure of the reactor building is 2
59 psig with a temperature of 284 degroes F.
3 CHAIRMAN STRATTON:
95 percent?
4 MR. WILSON: Yes, sir.
5 The net prevolume 9 18.illion cubic feet.
)K 6
CHAIRMAN STRATTON:
Now the reactor cavity that coul d 7
see 200 psi, what is its design?
8 MR. WILSON:
230.
9 CHAIRMAN STRAT!ON:
Designed for 230?
10 MR. WILSON: Yes, sir.
}j CHAIRMAN STRATTON:
And it could see --
12 MR. WILSON:
For the hot leg break it cuuld see
(
13 204.5.
For the cold leg break it is the design break, 230.
14 CHAIRMAN STRATTON:
So it could see the same 15 in the accident as it -- for which it is designed?
16 MR. WILSON: Yes, sir.
j7 CHAIRMAN STRATTON:
Is there an engineering 18 margin on this?
39 MR. WILSON:
It exists in the design of the 20 cavity itself and in the calculations, conservatism in the 21 calculations computer codes.
22 In addition to compar,tmem pressure, design criteria 23 is also imposed on'the structural design requirements on 24 the system for straight seismic and so on.
Ace-rece,ai nepo,ters, ine CHAIRMAN STRATTON:
What sort of engineering i
78 l
mm4 1
design factor is there that you have judged?
2 MR. WILSON:
I am getting a little beyond my 3
capabilities here.
I would like to refer that to someone 4
from Bechtel' Company.
5 John Dempsey?
6 MR. DEMPSEY:
The exact figure I will confirm, but 7
I believe it is 15 percent; and that is in the yield stress 8
when you combine all the loads including the -- this differen-9 tial pressure.'
10 CHAIRMAN'STRATTON:
Try that again.
I missed 11 something.
12 MR. DEMPSEY:
The loads are combined, the thermal 13 stresses and the earthquake loads, the seismic loads, in s.
14 conjunction with the differential pressure loading; and the 15 margin is in the allowable stresses compared to what these
~
16 factored load equations would come up for the reactor cavity.
17 h:
iL
- 15 e
- xnt, 1
ill u= v m Lo un 18 con 6iethat,.
"; rill have that fc: y m omorroy.
19 CHAIRMAN STRATTON:
Okay.
20 But you didn't perform this calculation until 21 recently, did you?
22 or, did you perform this calculation during the 23 design of the plan't?
24 (The Applicant conferring.)
Ace-Federal Reporters, tnc.
25 MR. WILSON:
No.
9
79 1
These calculations have been part of our design mm5 2
since the beginning.
The analysis was done on a preliminary 3
nature during the PSAR days.
It was recalculated for the 4
FSAR and recently we confirmed by very recent analysis 5
within the last ten or twelve months.
6 CHAIRMAN STRATTON:
What was the design in the PSAR?
7 MR. WILSON:
The design was 230, I believe.
8 CHAIRMAN STRATTON: What was the calculation?
9 What was the result of the calculation which led to 10 the design?
))
MR. BINGHAM:
I don't understand the question.
12 DR. SIESS:
Was the question -- what is the proper 13 term for this -- reactor cavity, was that determined at the 14 time of the PSAR?
15 MR. BINGHAM:
Yes.
16 DR. SIESS:
It had been determined?
17 MR. BINGHAM:
Yes.
jg MR. DEMPSEY:
The calculation was redone because 19 of the eventuality configurations changed in the penetrations 20 through the biological shield.
21 CHAIRMAN STRATTON:
What was the calculated pressure 22 at the PSAR stage, 230?
23 MR. WILSON:
I will check that.
24 CHAIRMAN STRATTON:
Okay.
Ace -Federal Reporters, Inc.
25 What is the margin in the steam generator -- is
-..m,~_.._.--a-
80 I
that in the csmpartment?
mm6 2
MR. WILSON:
Steam generator compartment.
3 The design is 19 on that one.
Calculated is 14.3.
4 CHAIRMAN STRATTON:
Calculated was the hot leg 5
break?
6 MR. WILSON:
Yes, sir.
7 CHAIRMAN STRATTON:
Are there any other cavities 8
inside the structure that are significant, that we should be 9
aware of, where the design pressure capability is close to the 10 calculated pressure following an accident?
11 (The Applicant conferring.)
12 MR. WILSON: No, sir.
13 The ones around the side compartment, structures, 14 add a greater margin than these that have been defined here.
15 The compartments on the outside have a greater 16 margin than what has been defined here.
I don't have those 17 numbers with me.
18 CHAIRMAN,STRATTON:
Okay.
19 Now, in the calculation of the pressure containment 20 itself, the design is 59 and the hot leg break you get to 21 be 52 and the cold leg 48.
22 Is the largest breakesize the worst case for these?
23 MR. WILSON: No, sir.
24 An intermediate break size, which is the 5-square Ace -Federal Reporters, Inc.
25 foot break, is an intermediate break size.
D b
--.-o--+w
--e.aw.,-
s,._..
81 mm7 1
CHAIRMAN STRATTON:
I thought you said that was the 2
hot leg.
Or, does it make any difference?
3 One was a cold leg break --
4 MR. WILSON: The most limiting cold leg break is 5
the 8.5.
6 CHAIRMAN STRATTON:
Okay.
7 The most limiting break of all is the hot leg 8
break?
9 MR. WILSON:
Yes, sir.
10 And it is an intermediate break.
11 CHAIRMAN STRATTON:
Okay.
12 Now, how important is the break size?
13 What is the sensitivity?
14 MR. WILSON: We have a somewhat -- somewhat a 15 Parametric study on this with a curve which presents the 16 Peak building pressure as a function of the rupture area.
17 This is in Figure 14.4-1 of the FSAR.
18 I will pass this up to you.
19 (Handing document to the Board.)
20 CHAIRMAN STRATTON:
What factors in the calculation 21 are most important in terms of the heat sources and the heat 22 sinks.
23 For example, suppose you should just take the 24 fluid in the system and flash this into steam within the AO-Federal Reporkts, Inc.
25 containment; would it make much difference?
l
82 mm8 1
Or, does it really matter that you claculate the 2
design -- which I assume you do -- the heat sink and the --
3 MR. WILSON:
Yes, sir.
4 Figure 14.4-4 provides a breakdown of the energy 5
inventories for the 5-square foot break as a function of time; 6
and provides a comparison of the energy for air coolers, 7
sctructures, sump, vapor, and total energy of the system.
8 (Handing document to the Board.)
9 CHAIRMAN STRATTON:
I apologize for not having read 10 everything I should have.
11 DR. MONSON:
While Dr. Stratton is looking at that, 12, may I ask you:
Have you conducted sensitivity studies to 13 determine the effect of changes in volume of the containment, 14 changes in the heat sink areas, changes in the heat transfer 15 coefficient for the surfaces of the heat sinks as a function of 16 time to detertuine the effect on containment pressures?
17 As I understand it, you have designed for 59 psi.
18 You are going to test at a higher nubmer, which is what?
19 Strength test?
20 MR. WILSON:
115 percent.
It is 69.
21 DR. MONSON:
You are going to test this 22 vessel at 69 psi.
You calculate a peak pressure in the 23 event of a worst accident to 52 psi.
To get a better handle 24 on how much this 17 psi differential represents a true Ace-Federal Reporters, Inc.
25 additional margin of safety above that calculated needed, have
[..
83 e
1 ycu conducted such sensitivity studies?
m9 2
(The Applicant conferring.)
end #8 3 ar1752 4
5 6
7 8
9 10 11 12 13 14 15 16 1
17 18 19 20 21 l
22 e
23 24 Ace-Federal Reporters, Inc.
25
~
84 CR 1752 D;n 9 l
MR. BINGHAM:
Yes, we have conducted them.
I am R ba 1 2
sorry.
I am Phil Bingham of Bechtel.
3 We have,:enducted all the stadis's Inclucr1Y:g film 4
thiqkn_ess_on..the... wall, of..the.c.orttainmeharl if i-
'-h, 5
toA v a,
- coeld present 20= p: ::c.Limw yau an 6
ddiea of-the-ca-rgins.
7 CHAIRMAN STRATTON:
The sensitivity of the calculaticas j
8 to these parameters is really what we are after.
9 DR. MONSON:
You have what appears to be quite a 10 substantial margin, but it would help us anyway to hear such 11 numbers tomorrow.
12 MR. BINGHAM:
Okay.
13 CHAIRMAN STRATTON:
How soon do you expect to be 14 testing the containment?
15 MR. WILSON:
You recall the date for containment 16 pressure, Jess?
17 MR. RAASCH:
Probably September.
18 MR. WILSON:
Approximately September, immediately 19 after cold hydro.
20 CHAIRMAN STRATTON:
Will you get a measure of the i
i 21 volume and the free volume at that time?
Can that be extracted l
l 22 from the experiment?
Do you havd a check ---
l 23 MR. WILSON:
We will'..have somewhat of a confirmation.
24 I don't know how accurate that is.
Ace -Federal Reporters, Inc.
~
25 CHAIRMAN STRATTON:
It is not easy.
e
- * ~ ~ ' " ' " * ' ' "
N+
85 Den 9 I
MR. WILSON:
I know.
R;bc 2 2
(Laughter)
TS 3
MR. WILSON:
Based on the volume of (arez required i
4 for pressurization.
5 But the calculations were not only based on estimates 6
of equipment, but they were recently modified after we started 7
getting equipment procured and as-built dimensions on equipment 8
so that the early calculations have been reconfirmed and data 9
input based on as-built dimensions of equipment, structures 10 in the containment building, to come up with the net free II volume.
12 CHAIRMAN STRATTON:
And your allowable leakage, 13 your tech spec leakage, design pressure, is what number?
14 MR. WILSON:
Point 1 percent.
15 CHAIRMAN STRATTON:
Point I?
16 MR. WILSON:
Yes, sir.
17 CHAIRMAN STRATTON:
It seems to me this was relatively 18 lower than some other,s I had seen.
19 MR. WILSON:
Yes, sir.
20 CHAIRMAN STRATTON:
It is?
21 MR. WILSON:
Some have been point 2.
We would be 22 happy to have that number.
23 CHAIRMAN STRATTON:
I am curious as to why point 1.
24 MR. SIESS:
What was it at the PSAR stage?
i Ace-Federal Reporters, Inc.
25 MR. WILSON:
Point 1 at that time.
Nee * - ~ ~-.
-+-.ww--
86 Daa 9 I
CHAIRMAN STRATTON:
So you have been consistent Ribn 3 2
at least.
3 MR. WILSON:
Yes, sir.
4 (Laughter) 5 DR. MONSON:
May I ask what effect the rupture f
~
6 of a small number of tubes in the steam generator might have on peak containment pressure in a pent-up LOCA?
7 8
(Applicant conferring) 9 MR. WILSON:
This is not a criteria that we are 10 required to design to and we have not analyzed that type of 11 a condition.
We do not know what the contribution would be to 12 the building pressure for the combined -- what you are referrinc 13 to would be a combined primary rupture and secondary rupture 14 with blowdown of the secondary system.
15 The inventory in our particular case with the 16 B&W steam generator is quite small compared to other types 17 of steam generators, so that the energy input from the steam generator would not be as significant as it may be in some 18 19 other units.
20 We have not calculated that.
21 DR. MONSON:
It -- has B&W made such calculations?
22 CHAIRMAN STRATTON:
Yaou can rephrase the calculation 23 to be how much -- how many BTU's can you add to your calculatio n 24 Ace - Federal Repos tets, Inc.
25 MR. WILSON:
To reach design pressure?
87 Den 9 I
CHAIRMAN STRATTON:
Before you reach design prcssure R;ba 4 2
and th.en relate this back to a number of tubes breaking, I do 3
believe.
t 4
MR. DEMPSEY:
In the FSAR, figure 14.4 -- ten shows 5
you the available energy both superheat and the saturated 6
energy input between the design pressure of the building and 7
from the design basis accident for the pressure transient 8
which is the five square foot hot leg rupture.
9 CHAIRMAN STRATTON:
That was too much for me to 10 catch.
We really need a projector I think.
11 MR. DEMPSEY:
We did an energy balance which is 12 presented on table 14.4-7 in the FSAR to show the changes in 13 the energy distribution from time zero being the inception 14 of the LOCA and 21 seconds, which is the occurrence of the 15 peak pressure for the design basis break which is the five 16 square foot hot leg rupture.
17 That resulted in a_ peak containment pressure
~
18 pressure and temperature.
19 MR. WILSON:
If I may clarify, figure 14.4-10 of 20 the FSAR provides information that you can determine the net 21 BTU's additional to be added above the primary rupture to 22 get you to design pressure. Thateis what you are asking for.
23 CHAIRMAN STRATTON:
I am asking you to tell me, 24 so I don't have to look it,up, and you tell me what this means Ace-Feetal Repor ers, tnc.
~
25 in terms of the steam generator.
_ _~
88 Den 9 1
MR. WILSON:
I can' t do tnat without an-analysis.
Roba 5 2
It dep, ends upon the type of input it is.
The curve presents 3
two forms.
One is in the form of saturated steam and the other 4
is in the form of superheat.
5 The different energy levels for each one are there.
6 For going to a case with the saturated steam would be approxi-7 mately -- looks like approximately 35 to 40 million BTU's addi-8 tional energy input.
9 CHAIRMAN STRATTON:
And what percent is that?
10 MR. WILSON:
That is approximately 15 percent.
11 CHAIRMAN STRATTON:
Okay.
12 DR. SIESS:
15 percent of what?
13 MR. WILSON:
Of the DBA total integral at the time.
14 DR. SIESS: Not of the steam generator?
15 MR. WILSCN:
No.
No, sir.
16 That represents approximately 15 percent of what 17 the total primary coolant energy input to the containment 18 building is at that time.
19 I would have to determine -- we could determine 20 what the energy input -- or energy is available in the steam generator, but I think this would be an approximation type 21 22 number as far as trying to determine the effect because the 23 steam generator does have some superheat.
24 A detailed analysis would have to be run to determine Ace-Federal Reporurs, Inc.
~
(Applicant conferring) 25
89
~
Den 9 I
MR. MATTIMOE:
The time constant ---
R bn 6 2
DR. MONSON:
How many times in the accident do 3
these ruptures occur?
It also depends upon where then in the
/
4 steam generator they occur.
5 MR. WILSON:
That is right.
6 DR. SIESS:
In that five square foot total for the 7
hot leg break, I guess it is, the one that governs, you do 8
get some energy input from the steam generator, don't you?
9 MR. WILSON:
No, sir.
For the cold leg break we 10 get the input from the steam generator.
11 DR. SIESS:
Not for the hot leg?
12 MR. WILSON:
No, sir.
13 It is the hot leg that is most limiting.
14 DR. MCNSON:
Even though you, as the Applicant, 15 are not required to assume the rupture of any tubes in the 16 steam generator duri'ng a LOCA, have you made any attempt':to 17 determine the probability of such ruptures occurring?
18 MR. WILSON:
I think there have been studies on that.
19 Pardon me.
20 (Applicant conferring) 21
.(Board' donference) o r':d 22 MR. WILSON:
I think phe subject has bddn addressed
~
in previous topicals, bmtmif=we'-could ~ch~eckethdt' toni'ght, 23 24 wa-will get_ m expl:.;Morr-bachte--yonetomoarew.
Ace - Federal Repor ters, Inc.
25 I think the question of the integrity of the steam
- l I
l 1
90
~
Den 9 1
generator tubes, in the event of a LOCA, has been addressed R3ba 7 2
to where the failure would not be anticipated.
3 DR. MONSON:
Qualitative descriptions of reasons
~
4 why the integrity should be higher are often given, but what 5
I am asking is have you made a study of the probability of 6
failure,. to come up with numbers?
7 MR. WILSON:
No, sir.
That hasntt been done yet.
8 DR. MONSON:
May I ask this:
For sone purposes 9
you do assume the rupture of feedwater pipe or steam pipe on 10 the secondary side for the heat exchanger, inside the contain-11 ment.
12 Have you analyzed the forces, the maximum forces 13 in the worst locations inside the steam generator insofar as 14 their possible capability for producing steam generator 15 tube ruptures is?
16 MR. WILSON:
I don't believe I follow that.
Could 17 you repeat it?
18 DR. MONSON:
If a steam pipe or a feedwater pipe, 19 going to or from the steam generator, should rupture inside 20 containment, close to the steam generator, forces which are 21 abnormal would be developed within the steam generator acting
^
22 on the tubes.
23 My question is, have you analyzed such forces in 24 an attempt to determine what the possibilities for failure Ace -Federal Reporters, Inc.
25 of the tubes in such a circumstance might be?
L
91 Den 9 I
(Applicant conferring)
Roba 8 2
MR. WILSON:
Again I think that is discussed in 3
the topical report that was submitted quite sometime ago, and 4
it is not very clear in my mind exactly what was done.
5 L thi@ hothpccmdation,4-analysisy-looiei-ng 6
ar tsh%stra cf a e -on-the-tube 2nd-perhaps-some -testim We 7
wM4--check. that. topical-tcai,ght and give _-you.=
=""==ry in that 8
arefor -both "Wh'aE7srteGiiGvuran<bwhat.we telt_ shou,1,d be m
9 ' don ~e.
10
-DR. MONSON: Were there any probability numbers?
11 MR. WILSON:
I don't believe there was any probability 12 analysis, sir.
13 DR. MONSON:
What is the axial distance between 14 supports or guides in the steam generator?
15 (Applicant conferring) 16 DR. MONSON:
Just approximately.
17 MR. GLEI:
I am going to say three feet, but I could 18 check it.
19 DR. MONSON:
We can go ahead while you are looking 20 that up.
21 (Board conference) i9 22 CR 1752 23 24
~
Ace -Federal Reporters, Inc.
25 9
~~
.ma
n.
92 410 l
Orl 1
CHAIRMAN STRATTON:
I would like to ask a question 2
on containment sprays, if I can.
3 You have sodium hydroxide additive, is that cor-4 rect?
5 MR. WILSON:
Yes, sir.
6 CHAIRMAN STRATTON:
Is this under the control of 7
the operator?
Suppose you do have something which fills the a
containment with steam and it looks like an accident?
9 Will these spr'ays go on automatically?
10 MR. WILSON:
The sprays come on automatically 11 with a five-minute delay.
12 CHAIRMAN STRATTCN:
Can he continue the delay if 13 he thinks of it before five minutes?
14 MR. WILSON:
If he does not take action to 15 prevent the sprays from initiating within the five minutes, 16 they will initiate automatically.
17 CHAIRMAN STRATTON:
But he can prevent it?
jg MR. WILSON:
Yes, sir, he can prevent it.
l 19 CHAIRMAN STRATTON:
Can he prevent the sodium 20 from being injected, or does this come automatically?
gj MR.' WILSON:
He would prevent the entire injection 22 f sodium hydroxide and the spray water itself.
CHAIRMAh STRATTON:
Can he allow just the spray 23 24 water if he should so judge, and not the sodium hydroxide Lu-Fehral Repotkrs, tsc.
25 additive?
9
93 ar2 1
MR. WILSON:
Not with the controls we have.
He 2
would have to operate some other valves.
3 DR. MONSON:
The sprays don't come on automatically
(
4 at the low overpressure level?
5 MR. WILSON:
No, sir, we have increased the pressura i
set point on that to include the system being actuated from 6
7 a secondary line rupture.
It is 30 psig.
8 CHAIRMAN STRATTON:
30 psi before they would 9
start the five-minute count?
10 MR. WILSON:
Yes, sir.
11 CHAIRMAN STRATTON:
Okay.
12 Are you and the Regulatory Staff in reasonable 13 agreement on the virtues of the sodium hydroxide additive 14 in removing the halogens from the atmosphere?
I --
15 (Laughter.)
16 Maybe that's the wrong question.
17 MR. WILSON:
Okay.
18 CHAIRMAN STRATTON:
Let me try it again.
19 There has been a recent report from the Regulatory 20 Staff in the matter of credit given to the. sprays, and spray 21 additives; has this report been applied to the Rancho Seco 22
. project?
In the analysis?
23 MR. BUCKLEY:
I would presume it is.
CHAIRMAN STRATTON:
I'd rather have something 24 ice-Federal Reporters' Inc.
25 better than " presume."
k--
94 ar3 1
MR. BUCKLEY:
We will check.
2 CHAIRMAN STRATTON:
It would change some things 3
by a considerable factor.
(
BUCKLEY:
The numbers that are used in the 4
MR.
5 interval have been prepared by the same people who reviewed 6
the sodium hydroxide sprays.
I will verify that.
7 CHAIRMAN STRATTON:
Have chese numbers changed 8
in the last several months, I guess is what I am asking.
I think you may be referring to a nwh l report not on the spray factor, but on the fraction of mety 5 10.
10
))
CHAIRMAN STRATTON:
Yes.
Right.
Okay.
This has not been factored in.
13 CHAIRMAN STRATTON:
Pardon me?
This has not been factored in.
15 CHAIRMAN STRATTON:
This has not?
16 This means what in terms of the off-site dose, 17 iodine dose?
The predicted dose is higher than we 19 would predict if we use the new guidelines.
20 CHAIRMAN STRATTON:
Are these guidelines still under review?
What is the reason they have not been applied 21
~
22 to this plant?
It takes time for the transition h
P ase.
They are still under review.
We will begin to apply 24 sco - Federal Reporkrs' Inc.
r 25 them for all safety evaluations that we do, I think, D
- e mesapM. 'd w - t%en --
.q.
95 ar4 1
beginning in August.
It takes time to convert the machine.
2 CHAIRMAN STRATTON:
I understand your administra-3 tive problems.
c i
4 Is the Applicant aware of this report?
5 MR. WILSON:
No, sir.
~
6 CHAIRMAN STRATTON:
It
-e ivailable to the public?
It is not a public report yet.
8 It is being prepared for the public.
9 (Laughter.)
10 CHAIRMAN STRATTON:
I am sorry.
I apologize.
jj (The Board conferring.)
12 DR. MONSON:
In an associated area, you have 13 provided four air coolers within the containment, two of 14 which must be operable after a LOCA.
15 If the ducting associated with those units were to 16 collapse, it is possible that the units could be rendered 17 ineffective.
18 I think it is for that reason that you have 19 provided some overpressure dampers in the ducting.
Whether 20 or not that's the reason, you do have the overpressure dampers 21 and I have read that they actuate at about 2 psi -- pressure 22 differential.
23 My question is:
A.se -they cler1%to-operate
- at 22ei, based -on-a. -static + analysis y-or-does 4 henna.Lys.is 24 ce - Federal Repor tets, Inc.
25 t^-
- "*^. A copunt _
"" * ^ -
5.- rise- %yra"ure jn the the racid
96 ar5 I
building so that a developed pressure differential across 2
tgsucting, more^than -2 psi ~,' cannot exist?
3 (The Applicant conferring.)
[
4 MR. WILSON:
I think we had that same question 5
from the Staff and I believe we have it haswered, but I 6
can't find it.
7 The pressure calculations were based on shock 8
analysis, not of the slow buildup.
9 DR. MONSON:
Not necessarily shock, but rapid 10 rise, I presume?
))
MR. WILSON:
I believe it was instantaneous shock 12 wave.
13 DR. MONSON:
All right.
Thank you.
14 (The Board conferring. )
15 CHAIRMAN STRATTON:
Any more on the ECCS?
16 MR. WILSON:
I can refer you to question 6.89 of 17 Appendix 6 (a).
18 DR. MONSON:
What you answered was correct?
}9 MR. WILSON:
That's correct.
20 DR. MONSON:
Thank you.
CHAIRMAN STRATTON:
Any more on the ECCS for today?
21 22 In that case, I would like to make a couple of short 23 announcements before we adjourn for the' day.
24 I Propose to receive the statements from members
%ce-FNeral ReporNrs* Inc.
~
25 f the public, the oral statements beginning at 1:00 o' clock
97 I
ar6 I
tomorrow afternoon, after we break for lunch.
2 For a period of about half an hour, I ask that 3
the oral statements be held to about three minutes, with
(
4 some flexibility; unless'you should wish to speak to me 5
beforehand and request a longer time.
~
6 I have 10 people and it will take a good bit of 7
time.
We will receive oral statements at 1:00 o' clock 8
tomorrow afternoon.
9 I also wish to announce that beginning tomorrow 10 morning, when we convene, we will go into a closed meeting, 11 first off, to discuss proprietary matters mentioned in the 12 Public notice in the Federal Register.
13 These will include the matters of fuel densifica-m 14 tion and possibly discussion of plant security plans.
15 We will have a closed session beginning tomorrow 16 morning for a period of about one hour.
17 If there is nothing more -- yes?
18 (Discussion off the record.)
19 CHAIRMAN STRATTON:
The public -- I am advised 20 that I should suggest that the time for the public session 21 to begin, I would expect that the public session would not begin before 9:30.
It will most likely be.a bit later than 22 23 9:30 tomorrow morning.
24 I will now declare this session adjourned until Lu-Federal Repormes, lnc.
~
25 tomorrow morning.
'=m
~w~~
~me.-
+-~
ar7 1
Thank you very much.
2 (Whereupon, at 5 :15 p.m., the hearing was adjourned,
3 to reconvene at 9:30 a.m., Thursday, 14 June 1973.)
I cn 4
5 6
7 8
9 10 11 12 13 s_.
14 15 16 17 l
18 19 20 21 22 23 24
'ca-Federal Reporters, Inc.
25 F
w-- -.
~