ML20235A063

Marked-up Draft Hazards Analysis

ML20235A063
Person / Time
Site:	05000000, Bodega Bay
Issue date:	05/27/1963
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML20234A767	List: IA-85-665, Discusses Seismic Design of Nuclear Power Plants,Including Const of Nuclear Plants in Seismically Active Zones ML20234A761 ML20234A769 ML20234A792 ML20234A798 ML20234A803 ML20234A812 ML20234A816 ML20234A827 ML20234A830 ML20234A838 ML20234A841 ML20234A844 ML20234A847 ML20234A851 ML20234A855 ML20234A858 ML20234A861 ML20234A864 ML20234A868 ML20234A871 ML20234A874 ML20234A893 ML20234A896 ML20234A900 ML20234A911 ML20234A922 ML20234A929 ML20234A932 ML20234A937 ML20234A944 ML20234A956 ML20234A965 ML20234A968 ML20234A976 ML20234A984 ML20234A993 ML20234A996 ML20234A999 ML20234B005 ML20234B008 ML20234B023 ML20234B032 ML20234B035 ML20234B040 ML20234B062 ML20234B066 ML20234B076 ML20234B084 ML20234B098 ... further results
References
FOIA-85-665 NUDOCS 8709230197
Download: ML20235A063 (35)

Text

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EAZARDS ANALYSIS .

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MYISIGr or I.ICElisIso AxD REGULATIGr g

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. _ PACIFIC CAS AND ELECTRIC C04PANY

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, BCDEGA EAY ATOGC PARK .

UNIT NUMBER 1 CGGTPUCTIGf PERMIT --

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TABLE & Cdrrurr3 I. Introduction

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II. anekground

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i III. Description & Safety Analysis A.

A. Eite & Erwirrm="t heters (Newell) 1

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3. I Special Earthquaha Dedign Considerations '

C. Containment i

D. Reacter Nuclear Syr,tems

1. Primary. systems & R V

2. Core Design (Mech. features; Nucl., Thezzal, & H Araulie x, -,

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I 3. Reactor controla - mechanical & nuclear facters '

4. Centrol & Sefety Instnmentation i

I f . Emergency & Safety Bystems a

l F. Pcwer System Equipment & Plant Auviliary

.! 0. Radiatica Monitoring P i

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• - 4 I. Vaste Treatmant, Storage & Discharge

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g 7 $/ IY. BhD Program ,.

K - Y. Safety Analyses 4

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YI. MCA Braluation

__ -VII. Technical 4alification of Applicant

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• VIII. Deports for ACRS '

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,: Iz. summary <{

I. Conclusions l

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t Nay 27,1963 APPEllDICES c

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A. Geology & Sei - M (Design Criteria) -

,1 B. Effects of Ecutine Discharge of Endiomative Materials to the Marine Ikrrircrament +

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May 27,1963

. Introduction he Pacific Osa & Meetric Company (PG&E) proposes to construct

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. . and operate a anslaar power plant on Bodega Esed in h== County,

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• ' California. Aceerding to the proposal PG&E will design and super-

' vise construstica of the unit, whilte the Sua General Moctrio -

Ccapany will furnish the nuclear steam supply system and the turbine generater.

The proposed plant, designated by PG&2 as Bcdega Bay Atomic Park Unit Number 1, vill be a direct cycle, forced-circulation PN boiling water reactor producing W st the zute of

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e 1,008 megawatts (Nv). The gross electriosi generating espasity A ~ y. .b I . ...,e e m' . ..

will be approximately 3ll!$;900' Mr.

The Bedeca Plant is similar in many respects to bciling vater power reactors now in operation. Features of the plant which eJ. bl , p.- =-t require research s ' --'-g n' ' c'* . y. order to provide engineering information necessary the design er evaluatien of the nuclear plant will be discussed in Section IV cf this analysis.

. II. Background On - _

70&E sutaitted an applientica to the AEC for a ,

eenstruction permit and operating license pursuant to Title 10,

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-Chapter 1, code of Federal Regulations, Part 50 (10 CFR $0). The u[

~ applicatica, which includes a "Pr=14=4==v Eszands Summary Report",

dated December 28, 1968, and Anenaments 1 sad 2 to the application Q

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dated March k and April 5,1963, respectively, hee been reviewed

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I by the staff of the IRvision of Licensing & Begulation in secuaance

[9,. with 10 CFR 50, 1he appliestico has also been considered by the ,

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ABC's Advisory Ceaunittee en Reestor Safeguards (ACBS), as required by the Atemic Ebsrby Amt and the regulations of the AEC. The l

views of the ACRS, as tapres' sed in a letter of April 18, 1963, l (acopyofwhichisattachedheretoesAppemfir ) von also considared in the regulatory staff's anraluation.

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At this time there an a number of features cf plant design and i

operation which have not been definitely resolved The Cessaission's  !

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, .. . ._ regulations provide for the issuance af a construction posit owm

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lg cases 3.as this in ,p -atdek some. aspects of design t h, N *. -

have not been completed. A provisional construction pomit may be issued, acccrding to Section 50 35, 10 CFR cn the basis of findings,

- 3 saangothers,that(1) the appliesnt has described the proposed design cf the facility, including, but net limited to, the principal architectural and engineering criteria fcr the design, and has j

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,Q , , identified the major features cr components on which further teobnical

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. infomat'en.fs aquind; (2) the cuitted technical infomation vill l ba supplied; (3) the applicant has propcsed, and there vill be een-

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ducted,a research and development progrea reasonably designed to

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... resolve the safety questions, if any, with respect to those features-

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or w ts which require research and development; and 'that (4)

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on the basis of the foregeing, there is reasonable assurance that (1) such safety questicas vill be satisfactorily resolved at or be-fore the latest date stated in the application for ecapletion of

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ocastruction cf the proposed facility and (ii) taking into eensidess.

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-- . . . thee.site criteria containal in Part 100, the

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een be constreetad and operated at the b;,-- ' location without

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undue risk to the health and safety of the publie.

The peamit, it granted, would authorize ocastruction enly.

The Cannission would require as timely reports fremt PG&E with respect to results of research and development and final design of the more significant design features. The 2 staff would continue

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its evaluatica of the safety of the plaist in light of this informa .

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.I (tien. An opersting lisense vill act be issued until the final design

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.has been evaluated by the M staff 'and the ACRS. In addition, the * .

l definite plans for operations would be evaluated by these two grcups.

Pursuant to a Botice of Nearing published ,

the issuance of a provisional eenstruction permit to PG&E will be considered at a public bearing to be bald ,

, at a.m., PDT, on 1963 6 * . ' - ,

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. 4 1 Se staff's evaluation of the pamposed Bedega nuclear power ,

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. pians ana its position en the issues as the fortheaming me. ring -

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based cm all the *mahn4==1 information submitted as part of the l P'

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's) applicant's request for a ecastruction permit and the report

.. fram the ACItS. All of this information is available for inspection I l

and review at the Cc mission's Public Decument Room in Washingten, l D. C., and at the Cassissica's San Fraccisco Operaticus Cffice, i

. San Francisco, Califernia. his ,

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_evalastics' and proposed ree===adati.:;= is subject te modification '*: -l

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.fa the light of any further informatim which may become av=41=M=, . ,33

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including the evidence introduced at the hearing. % der the Cennission's regulations, any person whose interest may be affected may file a petition to intervene and, if granted, may participate in the preeding. Se decision of the Ccemaission vill,be based upon tbo entire record in the proceeding. J s

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III. Description --' *it; .' hr , I

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.the Bodega Reactor is a direct eyele, forced circulation f

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.y .: y.e boiling water reactor with internal steam separation,Iuclear l

Jr'? energy released. in the reactor at the rate of 1,008 magavatts .

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will be transferred to the bili ..

water eoclant ~h *--"'-'*- -

c.o t%, ut:.A.i N through the reactor.Msebenemem4syHh stea:a generated at 1,075 onM h u=cb- A ^

. g paisflowstoa'ibrbinegeneratorjwHh-agrosselectrical r

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generating capacityg about 325 megavatts$eactor m.11 - cu.6J Wa:A..in- s s (teirculated through four lcopa each +=4-4

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p j 5 a pump rated

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- , at 29,000 gym.[ After passing through the turbinejthe steam vill 1

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, be ecedesse4 _

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,. , tb 41 '-W  % N.$t u ;4 A.e ' .j returned to the reactor vessel. The remotor coolant-nter, l

-1 which wd$ contain,ssmo radioactive materials, vill be ]* e ^-'e' h '3tal. e.3.

within a closed system frce which the only acmal eff5. tub 111 be the sentinuous discharge of noncendensible gases. This gaseous v.\ A d 6\ m & A 4 = St a+ % ~

materialAvill be menttered and released frem the :cencter stech.

As in conventional pcVer plants, the condenser vill be cooled by

.a water drawn fram nearby sources. In this case water to cool the

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4 ecadenser will be taken from Bodega Ny end discharged into the

• Pacifie Ocean ,fm.. - w 1;Ipgu.tatedquantities radio-f,'4

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.., y active liquids udt1 be discharged den the condenser coolant water.

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number of individual esfsty considerations. In T,no u.ana, analysis

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p, :, ..In the final analysis many of these' considerations require

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,a the detailed study and emb.uation of design features and ~

ara M  ! w operations which - -

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this case,PCdJil has presented a hazards report which contains general descriptive material concerning the conceptual design b#

of the proposed plane, known facts concoming the proposed site and its environment, and an analysis of the radielegical effects of. normal operation and secidents en the surrounding area. The

. w y c; . y present evaluatien by the Cemaission staff is therefore based

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upon the principles of design rather than upon detafla of design 4

themselfes. In the case of features which are of particular I importsace to safety, the staff expects to receive and intends to require information on final design of thsae features before PG&E i l

has expended any substantial amount of effort in the construction j of those features.

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In the following sections of this report the staff discusses

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the more impor ant safety considerations which have led to its  :

i eenelusions trith respect to operationf af the proposed plant

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without,. undue risk to the health and safety of the public. . .g

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t May 27, 1963

. w.

r#;

x, ;< . .  ;

III.' E. + .

,Septial Earthquake Design Considerations le ," ..(To be written after sempleta study of eensultants' reports)

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a May 27, 1963

c. Containment b eentainment system proposed for this facility is

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. cae hhich d:; ' Ogthe pressure suppressica concept. 36s.h py;[

,, . ., ~

. g . , - . .. c. . mantfz design is si=41=w in many respects to that used at ~

, y~^ -

af

.' Bumboldt Bay. . Significant fen.tures of the Bedega Bay Plant s: ,. c . . ,. ss 4

' design include the fon orings 4

1. Plans for the Bodega Bey Plant can for a dry well having a 60 ft, diameter spherical lover section and a 26 ft.

~

diameter cylindrical upper section. "TK %4 M '^ -@

I f.

kA Your w.a.M recirenlation A

_.$M:.:._ni M-'

. Id.

.u s fo<. W 4 u h 4L A k ~

.,- ., the.-oonobergueset-w&L katup located within the dry

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3 'Ble dry well vill have an airlock entsence. Personnel entry is mot planned durin6 reacter operaticn, but is een-templated with the reactor het and pressurized.

4. h suppression chamber will be in the form cf a torus and vin have a major dizmeter of 93 ft and a cross section i %:/

dissater af 26 ft.

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F Both the dry well and the suppression chamber will be designed

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,, and ecastructed in accordance with the ASME boiler and Pressure -

. .: Yessef Code,Section VIII. Piping restraints will be previded .;

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at ace,tainment $.htions to assure that failure of the pi;e >

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- . will'aot .eaune con *=4===nt suy"we. A concrete refueling . !,

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-9%

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bimms will conthin the dry won and suppressica chamber. +l

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. g _ 1 Pressure and leak rate specifications for these sentainment n.3 T

t g ents are as fonovs! -

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2 leak Rate Ccapenent Design Pressure (8/, et volume in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) i Bry well di2 psig 0.5 (at design pressure)

Suppressica shamber 35 psig 0 5 (at design pressure)

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' ' . ".12 in.' E20 100 (at 1/4 in. R 20)~ .

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In order to proof test the Bedean Bay pressure suppression ,

. . . . ~

. o. . i design, Pacifie Gas and Electrie is cenducting a test program at its Moss T=ading t. As in the Humboldt Bay case, the applicant has constructed a fun scale.sespant of the suppression system. In the test for Bodega Bay,a single 24-inch diameter vent pipe dagzza from the dry ven to the suppassion chanter e

6.y .y

.was used. SLaes the full size plant is to have 112 cf these

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.. . n e . . .

a w.. -

.dk .N

. Qf .g mt ,thetest,equipmentrepnsentsa1/112segmentof y: .a 8

s' the contaiammat. "

- b Tests were conducted with this mock-up to simulate various accident conditions. Aflowecaparableto1/n2thoftheflow  !

resulting from a complete circumferential bank of one of the 28 M

in' recirculation lines (with ficv cut W.h sides of the break)

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was taken as the *=M== endible operating accident" (MCOA).

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• Righest ecstainment pressures obsereed in these tests were 52 7:q

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W in the dry well and 30 psig in the suppression ebsaber.  !

Rose pressures were observed vben the mock-up dry well was pre-(9 _,

dp 4

M - *

- - , heated to 255' F and whom the mock-up reactor vessel water was 4:

O l - L, , q . ,faubcooled 350F. Tests,at higher and lover dry well temperatures , ,

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.i y e 'l' 4 and at higher and lower resetor tutor subeling yielded icwer dry yI t

well and suppassion thenber pressures. ,

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~.r m e , m.a In another test a break area 2.5 times that of the MCOA was

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e, simulated. In this test)the peak dry well pressure observed was

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.. 63 psig. ,Further. .. .

y Moss Isading testa are being conducted to determine ubether baffles are needed in the suppression chamber, C,, -

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f As~another significant eenta4= pant design feature, Pacifie a sm

, } Gas and ELeetric proposes that in a ausbesi-efg instances a single a

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isolatica valve vill hat be installed at the containment wall in

, b

. I pipes or ducts penetrating the con +mtn==nt. The applicant states,'

l gh M%  %.4 2 n %d O

l hoverer,L.. that p.x asch such line vill have#e isolation valveh-ene-

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. , . .. ~ J-Q-r *a ope 9+

process valve N .

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j<u n' . . ,, ly v.;

Notes ( .Additiemal remarks en isolation valving1 vill be ,

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.9ande after 708iB suhaits ===ad===it.)

' 4 4- ...<

p isolation valves located at the dry well mall in each main stesa line are to close on a manual signal or autcaatically en the occurrence of any of the following:

1. Low condenser vacuum

.. . ..~,

2 Main steam line leak (in the pipe tunnel) m w B'. . 3 Low rosator water level. , -

v. .

. ..;; . ~ .y. . .

'", the Bodega Bay design is such that during refu& ling, the spent fuel storage pool vill eaaneet directly to the shield water above w

,g y the reactor, thus permitting direct underwater transfer of fuel

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u without the need for a % transfer cask. This feature prevides

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~yg.. , , . , . . .,. . .-

• . g ; . . .4- =

u .W N- for both 6cntinuous shielding sal

...;v. ...

'$ cooling of spent fuel during treaster and storege.

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May 27, 1963 e

Tbe refueling butMingg which the drywell and suppression chamber system is located is provided with j

a scatrolled release ventilation system which discharges '

i to the plant stack. The W ilding and ventilation system design is such that the refueling building can be maintained at a .

. + .

wgative pnssum while et "- ___

% d'Sd s4

-- discharge frca the A .

bu41Aing passes %hrough ^1-- T  ; '; -

_ -... halcgen et . . . j a, ~,.J...A and radioactive particulate pric'r to scha de to the stack. f 4 k i M h 4EIE!has indicated that in accordance f with the reconnendations

=8ventof the ACRSj va the iMd.~r system vill c.bmq@ed be des 1 to permit

"-- ,_ _^, testing cf the ,t" *Y 5.: ""j parti &tes and 4e- ju m penere icdine at specified efficiencies. ' '

a .g;: .v .;pr; . . ~ ~ - - , ;L- .: .

For a reactor of the. type proposed the staff believes that j

the general contatmwnt scheme g ;:::1 is adequate. We believe , l l h '

i l hoverer, that some impcrtant criteria fcr the design of the (

e containment features have not at present been specifically pro- 1 posed. Such additional criteria, including thcae menticned

)

l explicitely by the ACRS report to the Scanission, are necessary j 1

i to r"5sure that the contaia==nt as proposed can, be reasonably

[

i expected to provide the high degree of integrity.paapaaed at any {

~

j time that it might be called upon to contain the rummegwwww.ir

H e%

eensequences of a ===Lrc ... '.A1. accident.

[

4 these eriteria involve containment testing, penetration design, I

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g May 27,1963 s 5

1 'the design should permit initial integral leak rate

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testing at the dry-well and suppressicas chamber at their

. . a.., . . . ,. .

. respective design pressure after the ins' t allation of all '.. .

,. j sg penetrations (including piping cenduits, electrical ecn. -

ducters, and gasketing closures) and subsequent periodic M

testing at ---.;.-

-4 design b

pressures, In the initial testing, the leakage rate cf the centairment cystem l

should be detemin'ed as a functica of apour pressure up to full design pressure.

f7 2. .The deafgn of penetrations should take into account, in '

lly ,

, . , ,- ..  : '6 .. < :. .. g

' .ng i - - .- .: ' ? additica to the ' pressure loed, the landa or deformations c."~T g

., imposed by thermal expansion, impact of missiles,

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f\ reactions cf ruptured pipes, and disturbances incident to A 3

R -

2.f installation, maintenance or repdirA

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Penetrations should a

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be shielded frca missiles to the extent practicable. All

,i penetrations should be designed so as tc, allow frequent i.f . .

periodic leakage rate tests of the penetrations only U b'h, 'u, , ,.

..;;i' c

. (inclurb.4 points of at+mehment to the containment shell), ..;

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- - - at ful.r. design pressure. .

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• 3. All pipes and eunduits which communicate with interior of < '

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3

. 4

..a;. .

3 the primary system er the contakaant system, and other '

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. .,A W.#

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.r.p'i ping (such as instrumentation and control piping) which U*i

1. 4 .

. ~ . . - .. .

1

_ . Veannot be adequately protected a6ainst accidental rupture,

, l ' ,',

y should contain double isolation valves. All valves per- ,

}'. fo1 ming the function of inclation valves should be provided ' t.

l' t y,W r %.

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May 27, 1963 with protection against materials in the system which proper mightprevent/ closing.andshouldbeprovidedwithreliabis 2., .

,.. ey. d - . . . . ..

_ autematic and manual actuation features. Isolation valvin g

, ;::' .ro . . -,

should be designed so as to permit periodic leakage rate

+.,

~ ~

. j tests. Appropriate closing times for isolation valves should be detezzined on the basis of analyses of system ruptures

. , which veuld release fission or actiYation prCducte Cutside I the dry well while the valves are not fully closed. .

i i i

j The Staff believes that PG&E shculd submit for l t t Coussission review the results of further developmental

.x .

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,y

.  ; '. ., ? ., ,- - tests of the suppression pool concept and final design  !

m f. , , , m... ..~,. .,.,. # . m,n ,;g.;;.; ; ,- y .. , ,. ,.

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, plans for the cont =%t as soon as they can be made j i

I

'I ' sym41mh13, I

(Note: Additional cements vill be made en tM effects j of earthquakes en containment desi6 n.) )

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Donctor helear Syo' tees :

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the features of a smelaar power 32mst which are.af seet iaqpartant to reestor safety are those inhich provide .

. ,;. s for the ocatainment cf sedicactive fission products in s ,

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normalcperatismam&'f

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- those systems for instrumentation and scatrol of the >

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neutron eboegt zes.ctice. Generally speaking the N

==

• mical design cf the reactor fuel elements and the design cf controls and the heat rencval systes abould )

te such that in ncruel cperations sad under weny gect- m ceivable occident situaticas radiesetive fissica products cm.a

. : . . v. b4.: Q g. j

• a.edd-be leML A164 eenfined MS C,owdhL to the taal^ elements A c e4 -themselves, N-1 y +%

o.J. @ M j. m .. *- -" *" the prissary systemr iterM eerres as a secondary coa +mh=ent system.

l .- Prirrry System l

M,1.L. c. .m ~

The reactor m; A ri"W. L r.it preseure x====+* vessel designed, built and tested j in acecrdance vith Sectica VIII cf the Boiler sad Fresears Vessel Code of.thee ameriesa Society of

. l k

na Mechanical 2cgineers. .!

.. - . . ~

The._ .50..ft. hish.br 15 ft.. . .. ;

diameter vessel $111 be ecastructed of carbon steel M G :. .s . . , M ,.

• intari .J .mL will te clad with%tainless I A. M Sypose % N koa d. I steel appM by veld overla metheds. The design pressureM th . . .,3 vill be 1235 peig at 575'7.

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.. S May 27, 1963

. i'. L: .

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Steam generated la the reactor

% !_J.L,WMt  %.~ f . ._ t.m- inside L e4 the &gresater- a. Wvessel .

.9.n -ts-Q-I" '"*""

4,, , , .

.,4 too 20-inch stema lines to the turbine 4

shieb is loested la a separate striseture.

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% m_

_ k, _-g tL.. . _ ___m._

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• r^ : :^r- "--^t

+b @Afour 23 inch passes f-- w-- --%h A

pipes. Fer.r recirculation p.:::pa, cne in each icep,

, provide the driving fene for circulating water through the reactor core. Feed Sater;-tc_iq frem the condenser is Lajected into the reactor vessel by a pump driven ty the main turhimog ehaft' . '

9 ,-, _. .;,: . .

?_::: x: r r::1 ".6piping loeste&~vithin the dry tell- -  :

. s .,

vill be designed, tested end constructed in accordance with applicable requirements of the ASNE 3piler and Pressure Yessel Code. Piping outside the dry well will conform to the reg'uirements et the American Steadards Association Ccde for Pressure Piping. Selve safety valves, w.rranged to discherge into the .suppressica chamber, are provided to from protect the zweeter and primary sys cwer.yressure.

2.' Cere DeAs. -

The reseter ecre vill to = 3-::r'. of 592 feel assemblies each of which provides a vertical chaa_ael thrcugh shich the t,.irture cf .ter.n and veter pasces, i

4 i

b

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j mar 27, 1963

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~. She'eese will home eqpyrosimetely the form at a

.h r18ht airoslar systmaar 14d inekes in dinaster and 225 inches high. . Que'hundast and forty-five ecstal rods will estar thA cose.from below the

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fuel asamMtas .:h$ah 2 _-- -

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h mhg %-

s.

M e , T1,, ~ ~ ^^"

_ .. . . n ., , s ,.

. . .. n:y

• .  ::w -. , ,

..broud.

E '~

..~.

,. . .. .

m__.__ _ . . .. ,__

-w n -w _ _ - -

4r +k-

._ _-- M Bcte

, _ _ . -_= Cenclusiens I en structumi destga vill be stated as seen e.s seisaclegical reports have been meelved and evalx.ated.)

Koch fuel assently will te camposed of 49 fuel rocs -

, in a square array. Fuel pellets et D% enriched to a.5% E235 will be eestained within stainless .. steel tubins. +sas- K(.4.

.s .. s. . . . -

.M, MW has tentatively propcoed that' this tattag veuld bare a acainal thickness of 0.011 inches and would te able to withstandanesposureof15,000 MWD /TGI. On the basis cf present inferuation and cperatinE experience, cr.e car.nct i1 t

• ' l 5 ~<

p, : 4. :r e - - *

.N h i

m.. D 3....- N .If*Y .n'. N' .,kNf5. . f5,IN. N...P. . YU9@E.Y.. 3.n.... N k.k. . *. N.,**.*"**.**"*'.

• E. O'.

. . . . . - . . '/ . 1 ,

.., . . .. .~. . .~ . s e ~~ . i < m n <. ~. > n. - + : .; . ~ .e .~.w":..~ ~e.ra ~ - ~ --~ ~ e w :.n+ -. ...

• LM. '.6,Q.".1X!'.RW 7* % % ' "'O A3W'W'*"'. ?&L:,% -%

• Ms'l-$ENL M.% ? Si 'ib %$&M*Q;(h.f'lhQT

.~ .. t I-j

e. "

a .e .

n ,

n Nmy 27, 1963

' ~

. .- 10 m . ..

~

(.  ?

be sure that fuel with this type of cladding can be irradiated for the expceures stated without experiencing artensive rupturing of aksas *1=dd'apM,v111 . Gra'm.e ot be f'--"4 Ae * --I

,' %_. .,_ s P_ A until further data from a Gameral Ebsetrie 2* search rd y .e .

~

develcyment prsgram are available. This pregn a is

.- - -lydesigned to provide an engineerin6 lasts fcr A, path.

fuel element design 3 In any event, extensive experience A,m.-

sithyr reacters provides reasonable assure.nce thet e se$e&y suittble design can be moder k+_d :M .

Theresi and hydraulie factors, which altimately determine the permissible pesar level of the remeter, .

hes caly been triefly described by P0bt. The data.

x Aa.,

presented 1- C . ,that 70ht interds to extend the limits generally in tse im zwacters cf this type.

Acelyses made at this time .tagakt together with ed f 'A ,

cperating experience do not form a sufficient tasis fer I

determining appropriate thermal lix,ita at this time.

70&E has had established b design eriteria for establishing

%t\

jf detailed thermal and hydraulic design factors they use '

i. .e me re.ctor, man.1,, t,at e m .m -

- sitheut less of claddics integrity ever the design expcaure h period at the maximum heat fluxes peasitie 1dthin turn-cut limitetiens. Cherating experience at other telling 1.eter reacters has indicated that this criteria een be .M '

.[.d.

w~,w I ,,.w.;w.h,;

r >k ., . - m. v n.ca.ww.w~.

.. .. c.u :.Ks , G .

J J u., ,. . ,;.e m -m:rN=

~. . '.> o .s A ,

o%.

i %2 e '- ' * - * . ~ ~ + '

~~

ww.w

+ f.

~

. . .,.# *E y 3.-m,.,e .a ~,, ,x.~,J. .3 *e? w. * ' .e-~m-- w' b * . 3 , +' e. v> ~, .'. . , , . . . .. * . .. -.#.

, . .. . . ~ ~. .. . . .a. ~ n ..

. - , +o i

\'s U.'**L'?Y.? O."w[,j W ?, BEN Y ? O $1*J A N. ?tiLD $ $ ? N ^"T'S? W k W b '5'b $ b,v, y ,s.w.s -n,m.,~ t - '

w 7,3 c,.,

kONYW5'U[Y']?  : ,

1~ j .. , so 3. s

~

_ p , . ,

/ C

'.~

t m

(l 1 !

a ) l

..\

l 8187 27, 1963 '

n ,* , .. ,

. 11 . . ., , . . , :-

p .- s i. , . .,.

. -r - -

( , ,: ~

u..

the power distributes which is sapeeted la the. .

litriAmma ggggggy gggg'ggg gggg ggg{ggggg ggy ggg'ggggggg 4.4. ,.

cf detaz=<=*== the therumi margins which wes14,sb4aer .ts;Q.

A a 4 A- i

1. th bott=t - m e.e e.ti e vin be

,. . .~ . . , .. .

refined by detailed enhlations of power distributes ..

g .. s in the ecurse of,4 wiga.e n . L. M t u .. L.,

o- S .

th ;..;= # 1-' . ih' S;tvill be moniter.ed centinucusly , i L.. mm.

.T 9 ..t by a system cf incore fins acnitorsh 'Bd ;./E

• mLk have #

been successfully used in other reacters and A.me_ *g ,

..u. .oo_

4.4 4#'s d

.mw__

3,

,,,,, y _ _9uwd,%t ,_ d_ i . , J (

a w -

s I

g .tt*e 1 thermal margin 6'i t n ;

. _ m ..

., <> . -  :.r"--!

. . . e , g..nr.

S :;--d:_.-a' . . ,. .

- M,s.A .eM.

- ~

' s , ,

,. :f.

)

Pre 1*=t===y calculaticas indicate that at rated ',;< 4--

4

.. n . . . . ..

. cyereting eenditions the rteam valisms freeticas would -

l be as fcu ews:

Average Core Veids .,7%

AversgoExitVoids.$8%

M PO&E believes that analeg computer studies being made sill show that the plant saa be designed to exhibit satistsetery dynamie perforumane with such high veia ,. i

9. -

., q content. 23m the staff knows of no substantial operating

'Ia experience that would eccfirm the acceptability of cperating at veid fracticas this highj ft S t r, besever,-thet sith appropriate limitatier.s high veid conditiens cza be safely apprcached in reacter tests

, , .  % j4 8 j , ,

,o , .y ,

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,y% , ...

t$$

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. f4 G p=C.. yew

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y.,.,,_ . ,,,gv ~,.. _ ~ . .. - m . __ ..

., , ,.y 3, . . u. . - a.. ;,. . _m .w. ._. . ,.,, , ,, . , , .

. , .; y,;;., .y :.o . -

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i

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• % p J

1. ' , thy 27,1S63

. , ;w ,

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w. .. ., .. .

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3. A-A. -

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h. n . /'. eartse&-se-ee 4e deteaudas the p,._ . .,,. 4 vota

g. < =

3 s a s' s .

contest for mamm3. operstica. L , . % -.

r- s 1 5.,,,

4 .. s.

~

4 . ,, . ... g , . _ . ~ . ,

Rosetor Cm trols ' ' '

3 -

n w.

h*hme estetr requires that them be as11able.

.- s . . ..

. . . ..~ .

~

means for anotrailing the suaetivit.y of a nuolear : .

reactor. ivity esa be considered a measure of 7

'the spidity with which the neutrca chain re==*n

\ .'

i

( ~

, reaction chemges. Een reactivity is positive the I t> w.a.6.k I

i chain reactie grews, that is the ante at which {

j mMeme energy is released by fisaism is imezwased.

y- s Carversely when remetivity is negative the chaia  :.

qm:p

- , . , , ., ~ /

., , ,o .,.-.y u ~. , ' ,

, diminimbas and pesar falleJ Se aparettag eemaittaa 't i . ~ .

v- -

x~ /

s... ,

cf the zesetor, its .tampenstarei presouro, power level,, .

HW voia conten and 4mme=ge:-effect-ear reactivity.

.%. ,,, t

_ L u k 4.%.,,A...*m. w - -_F- - i ge _- e _..- . . . . - . _

______e _e . ,.,m

• - - 'u.a.ek&4 da.a., O

. known, and the theoretical and experimental metheae

-M-fer investigating reactivity effects are sufficiently A

. e developed to permit design at the kr seester ocetrok .

. .; . ;. ,n; r .,

system to proeest with confidemos.4,. .

. . , .-.s,...

' '-- M It is espected that within the songs of operating variables w ^_ _ lated for this reactor the tweeter cv A c v L .m .a Z e. f . w a s r.hould tha 1 ;, m ^a, m h be stable) m .t is- n "? - .,,-

&. . 7m. s._ _ _ _ __ _ _ _#w,a %_ _ .p%,,, m m% 4

=u.- -

,-y ~ , . . -

- 3 t

o-.-A t~.A"E b dt have a strong tm eavy to decrease reactivityg h--

W 96%wJew- .=. . 'Tka asin purpcaes cr the control systes,therefore,ere

.. 1 I; d.% 'Y 4 ,. 4 , , .; g , N / 4,,, *

, ,'c'w'c 4, pr -

• lJ g .(gg,*p ' //,; p *e> . , c, -

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t. .

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. . v ,.- -4!.m. . f..,w,.,.3,..,.y.

+ , ,, , , . . , , , , . . , _ , ,.,g, . . . .

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...m.,p

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.,q.

4

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'MM s . S h f 44SnPts.44 aJsWop W .~,

. y > ave 6 apes 49* .sawg(.d@.s.se,e*,,

t#ey , gg d.c**7,

' ......j r., g,em .v .a ..y qt

. S.arJhnmag ,s.Jo.a.

. . . . g .y,.= c, m = w .. . r m . ,, n.

h +f nedresse e4 4,ase.me.n.

a.n , iii

, ,..,. g. e.- me %64,,s, s.A

+- - . - . .

4,p'A w. . . ,wi

.. . ~ g.6;..a,...,,. ,,.,ys.p

,,#.2,,....i e,,e .4 ,4- .c.r.-._

!.,+,.t.% w .;

. w .en

.%....,4t.,.,.=r*-:7..... c .**..w --++ eq

_..__c,.._..

. -._.- u ..,~..,._2; w._

. ~ r 4. .- e.-, ..~ a .~m ._m . .. s ,

..,...m.,

._. . a .,m .,e.,.,c.,. #. m.A_,l ..%

m 1 ,

,e .-

w.s m

v, May 27, 1963

. i

.U. ' '

l to provide a mesas cf precteely adjusting **setivity

~

t, cet J., f(t.2,, -M % & A-en". .. A ,Jrr and to provide a feet and reliable -

seens cf terminating .e :9f_- rn:tien shocid

. A & . . .. . ~

e t-e ww LA t.%' ::i .u e- m c: 2 - rid ; r r @zu.;r1r: 4Lcd"

., ----- U - ,

. e_ 4 .p- - A M 1,. ...,b emm. o 2,g*ga_n .. .~. - gfint prodret. b k l 7'

from the fuel elements er damage

~~' g the prime.z;y system.

h h -i:.ri,P "" +' r:

fr: ; r_  %

x& .L., u o w-

.f'T$ecentrelsyste.b

%M ita.

' (

e n.y es thich haWCpufficient G.A(centr [1 ret.ctivity worth to het. the reacter safely shut devn

,4h o_f..

even thcugh eney a.ight be ctuck ou.4 y %*

• 3 c m,

%,,,,, a. _ -. ._ . : : 28 -" ' ++- +M 1 *f - l Mh. .

bet e .?. .

A a A.iiFr The kett, ith all'ecutzel-retts 3 in the

e s S*

reacter,is calculated tc to 0 97. h ecubinei' worth tB G cf the 14) control'retrs is enleulated to be 0.18. h contrcl r.aterial vill te beren cartida ecutcined & *---*

gh

^ 0.175 in. O.D. stainless steel tubes. Mditienel con.

tztl is provided with 316pQA ecatrol eurtains '.hich will b o resi.pe2r.anently leetted isi...;r.i selected fuel elements.

h verth cf these curtains is esaculated to be Axi$

-(k.w.,, c. . . .

• O.12. dikm A centrol curtMna vill be constzveted of 0.1$

1 bcron stair.less steel.

l l

l 1

4 me .t

=-eg4 . .w e.c3 co.w. ..g. . we ev 9r W-* * *. = P*Te'9='.m*.,*'*.**-- .. - . - -  ;,.*v-*r**

-- * ~**.re-*N'**** ****** ** .*

w tap .r. 7. . 7 .o > yi ..J..,,...

9-pw .a76,vt --. . >..g.g7pf.ty,ywwvg , y.. g'* =se p. . , . m. + . ")w."a, ~.4 . ,.J s .

C_____________.______._________~

. . . . . . . . . . . . . ...-.z.,, s - , . . - - -

._e ..

. ". .Qll .s %;;Qif.Lgf%e~& 'v :&+R + _+,-+j- & + ~%G '&O$%.Y%eNh-db .

h  ;

.....-. v . y t

9 Nsr 27,1965

- 1A . '

. ;l . -

'the sometce design also immtes a liquid poison -

system that can be used to in, Met suam seditsa pantabarate

, . .y )

into the cozw in the event esagdets shutdown osanot be achieved hy use of the ecatrol rods.

Mae a scetrol aws drives "iE -2_ '

b ^ ' M *-

q 3;(%: -t :- '- % designed using the same basic concepts, % $ N

-Asu. O A % M.

as have tees employed in drives in use in bciling % ster reactor plaats at Dree6en, Big Beck Point, Nostoldt Bay

- %yW L and the EET! Plant in Italy. haterg no the hydraulic

%.L fluid can 3 be applied to either side of a piston which is rhia="v *~f =4 to the ocatant rod, thus providing for

. v .x. . . .  ;. . ./ .;

aither upverd or dowawaze ret metica. Qtly ese red' san be .

noved outwaat (increaming zweetivity) at a time and it may:

be acted a1ther continuously er in 6 inch steps. 3:4 speed is controlled by cri11ces which regulate the flow cf water awy trou.the icy pressure side of the piston. All rods P

can te inserted simultanseusly, shutting the reector desa.

Reds are scrummed upward by applying pressurized water fres either the remeter er frem accumulators to the bottom side 1 of the drive pistcas and simultaneously relieving the kr '

volume abeve the top .txt side of the pistcas to the seren dump task. The drive is locked in fixed positicas by )

collet fingers stich er.585e Groeves spzeed r,t 6-inch 1

1 interrals alcog the acveble index tube. The ecliet fir 4ers ]

suppcrt the stight cf the red and the downward fcrees due to

. 1 Istetor pfesSRre. s . . ,

a P ' l'8 ,

, , . _ _ ----8'*%er9- , _. . .._g_. .

D 839% 8 * *TD. 9, gem g .s,g pagg,pa g g, g g.qw g,p.gr

• ; $3

, . . .e . . - _. _ P f W{4._.'*..._,,_..-.,.._.t.r._._..-

A 98*.8 0- 9 W h q,. y, gain g,,gempo.M e ,

u - .. . . . . . _ .~. . , , .

.__m__ _ _ _ _ _. y

~*"I h *.$ s,J,.w w ( ",; ;j Y,.,adet*D.[.o. . y.h.,

i s+

u  ?. t & ! . . wS - ,

$ ,e.*'s- N fr @8+A'y.==,w ,, . b sa w *

. wn

%e .w.w. -: v m.Lw'u.a+M/*w'-ke-*'e,.d- %oNM yr -fe*W, +N '4.

W9wp

% T**I b f***xv :

a ,._.. n,, ,. g ,.a. m .c,.u. 7 w w, :a w

,l......-

. . ~

~

Q m m q,

c .,

May27,IM3

.n. , , ~

+, , ..

' ~

51aes drives stailar to these have been ase& at other plante '

as important part of our evalastian et these drives is based om -

previous esperiamee with ttsee drivas. Shae-Ametatre-9 sus 5Es

% +

__ . . . . . . .m E'

u ,, _ % . _ _

I ' h at Mg Book Peist, there home been toe idsdnted esourrences

~

cf red " drift-out".' In ene of these, the cause was attributed to an inadvertent miease cf daaineraliter Msins result 1:g in the eclist fingere being Janned in the open pcsitica'so thet-the red us free to crift as influemeed by the forces due to gravity and lodreolic pressure. In the second eene it was reprrted that s hs24 particle h=== tmpped betaaen the cellst

- i

)

pietaa and a sleeve obish is leested beskeen the es11st and the; indes tube. 'Bris assia is believed to have seused the eellet i fireers to be Jammed in the epen peoitice, thus permitting ice drift. The hard particle as never fe*.:nd. It should be netmd, hetever, that in r.alther cf these cases ncr in any c.ther et.se j has there been any apparent significant tapairment et screa esps.tility. Detailed design of drives for the Bodega reacter I

has act been made. General. Elmetrie is seasidering me41timatisme '

.. .. o 4 et earlier designs that will adminime the possibility at.

• fersign asterisi aceaunlating in the ved drives. The appliennt has also indicated that fumetional and endursace testa vill be  !

ande en the pxtetype trearmoryon Scdega meew- , but the detailed procedures for these tacts and the accepubility 1

criteria have act been deterniaod. '

i 8

4 4 L I y g 6 g ,

p,', W, h

• [ .e,*,.,g '

3

• t

=

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.e

• ,.7, .- .,

c? , w-

, . , [h ' T ** *[

-r.--

h' * * . *-

- - . . . . .. - ---**=-. . - . .. '.wm ey 'e di ~ . . IsmY.5=k M m*[.k *qt [ 4. k[ . h [._ J [_ II .* 8 4* * '"# J - *'  !.

.- :. ~ 2 ;,.c.,..:.

w_ y ww ~ . -u ,:+. ,pusee n- -.vn.a u G+2+

L a.Le O 2:ica h ... aid.h a u-~. x e sv-c w. ~ a. -~ w r tn ~ m ~a..m. >;,mmage; . &.n. w.w-,.s 4 ,

m .:m.

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l- .,

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\

16 Mar af,1963 i r ,

h !h 'e

""**'==8a== et the possible eescognances et a rei drepeat i .

, meesdent invoiring a stasie rod is 41senseed etee nere la this eb e e report under Accident Evelmaticas. It is our opinica that the sudden dropeut at more then cae rot at any given time is 13.-

eredible with thap the proposed ersten.

Castrol systemo 6 are designed to poect rapidly to demands for shutting a reacter deva genert11y have some potential for accidentally incrossing resctivity as well. This aspect cf the PGEE centrol systes design is discussed later in this +dJ j report (Sectica Y), dere eennequeness of a rod dii, -.1, are , .' wE DL i

ecosidered. PG&E bas indicates that devices for ifmf ting individuni red worth and for impeding the fall of a set are g.

3 . . .

<.-m. . , m .. -

under developeast. Seek devices emeld enhenee the easety of opesetien and stay 11ty the . ::M that are paesently teet with similar drives te provide secursace that rods cancet drop out, j

the staff believes that 7082 mk=1d schmit timely rescrts t

to the Cemetasien en develcyment, design, sad testing et the J oestrol erstem. ...- -

. , t (Ilotes Addittenal ecuments will be unde en een;; ~

effects.) , .

4. Centzel end safety Instressetation ,

, the instrumentation necessary for sarety in a nuclear power

~

plant geners117 inycives a large munner er semx:rs threcgbeut the verican precess systems. tese sensers measure a variety cf

, p 9

i* '7 T j

.. . .~

qg i+ . *W e

. . $ M.'44'A4;iTGpli';uaqM4WO.?;W.E4F4ihETng,IME994*rhKWgaW4.%GyA.MJ4A.M4 EAT #.'lii foe.

• % . . +'.C &. .f k 4 W M .O. h. La & S G 'u.nia- J Ms. A M % X M

• L M *a N * * *'a

• A " ~ " i ~ ' ~ " -* '~~ ~~

.. a

. m .

t + .%..... 'f .y '.).,

-m..

v.._f y N.4$. Q

..._ ,,;k.f k'f , %.%.9 > .~ull,'.:.grdl , .

• { --' , * [zO h ,:M,2::y i,f k ;QJ._

- p., .,

, 7 c.

,.,.:.. - W ..

]

j 1

by27,1%3  :

. 17 m.

, y ,

variaMas, such as neutres flant and gamma radistica levels, _ ,

and temperatures and pressumes of varione fluids. Informaties callected by the measuring Lastruments is used to guide the operating staff la oestrolling the plem$ and to astuste matenstie ccatrol devices.

.the instri.zents, cites.its, and control devices which are

\

cf aca.t impc.rtance to public health and safety are .(s) the,ee necessary fer and ecatributing to stable remeter everstion, (b) these twed 'in contivl cf asdioactive fidds and effitente, and (c) those toed for control of emar5ency equipment.

A general descriptism et Amstrumentattaa La passented. in [ ,- , ,, .

.... j the typlicaties. At this time, those is not suffietest lateaun. J 1

4 tion available from skich ene can determine stether instruments.

I tien previsione have 1,een made fcr all escential functices, ner I cem cre determine the degree of relishility that should be attributed to the reactor pattection systes, which is described by PG4B as " fail-safe". These, however, are design problems which appear ta require only the appliestion of well-kneum 1 .

!- engineering methods. The staff intends to evatunee the tweter . . - ..

contret and safety instrumentation la detail. Particular attention will be given to the need for autoestic functions and the reliability of safety instrumentation.

(Additional comments will be added on earthe.uakes.)

.ss . . ..,-  ;. .

i .

fj

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, Enestemey systems provide means either ger esfaly

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contieeleg spessties Aa the event of eene ognipment"'

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features of the facility =hteh have been provided for the 1

primary purpose of making plaat epwrat. Lea more ceavenient. '

roliebte for messemie er, in effect, amarsency systems.

Other festeres are designed primstily as amargency systees.

The principal  : 'es fasilities proposed are

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cooling syntes; ($ Bleed end feed systems (O Core l' and aprey systeng/(7) Centaineest system.

/ Some of these systems have already been distuneed in

, this maslysis. PMW features of other ammesney and .

t safety systems ese dia====ad ta this mesties. . .

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reliability of the M. m.to portens peeperly la adverse ' _

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, aireematenses. '2his regates not sely earefal desism of ,

the estatanding featuses het aloe attenties to seek related equipenet se signal and control circuits, power sag > plies, -

and is instrumentation. Maintamance and frequanc testing of earrgency eystems peep provides the final assurances of

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he taken inne mesammt in timmt design.

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i severst 1svels, as described in resa's M 'li I

application. The plaat is tied into the PGGE distribution ,

aposam by toe 220 Ev circuits to Isaacie bobstaties. l l

All plant assiliary power ogstrements can be est Ii il i

by either a tremefermer tied te the staties generator

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er by a N ==== F time to the 229 Ear 11 ass. An i

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2. puermeer coolier Systems 1 A sambaar of -- diffesent --- 411

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A be psev&ded for remadag-the-heet generated in the

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previsions will imeleds: I I

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(1) Tbs mensal eendemaste-feed-water systeng. ,

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[F- (2)' An emersency h- utdeh een be pet inte .

eparaties in event the reester smet be f l tooleted from the main esadenserg k3) A low prammere shut W oooling systems see- j i

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(4) A bleed-and-feed system which roleseos steen et a centrolled rate to the suppression pool. a4 (5) CL Q p - m e m e f"' capacity)

= MQQ.'f A seaber of sources of watet: and humping {

will be available to restore water lost throu6h occidental ruptures er threesh bleed-and-feed operations. - Both  :

d. l  !

hisk head and low head pumps will be provided with back-  ! ,l up pumping arrangements. .In the event of a asjer.

rupture of the primary system, amerseesy acties shem14 he capable of reducias to a great estest the amount of fuel damage sad fisates predact release from the reactor. i

, I (Additional comments will be made en earthquakes.) j U

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..I V. Barety Analyses

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. . . .l The design features of theplant have been described in the

, previous sections and'in many cases the ' safeguards provided by a particular design feature or the operational limits imposed by k.

c E335 safety considerations for a particular feature were discussed. In t t' '

genersi the criteria for plant design should include (1) means

-1' to control radiation hazards (including radioactive discharge) during normal operation, (2) design features to minimize the probability of having an accident, and (3) design features for mitigating the consequences of an accident should one occur. )

l The means for centro 111ng radiation hazards during normal

. operation vill be provided by suitabla shielding and radiation j monitoring in the case of' direct radiation estitted frcat the reactor and by proper monitoring of radioactive vastes which are i discharged from the plant site. For vastes discharged from the plant, the release rates shall be such that they do not result in personnel exposures in excess of 10 CFR 20 limits.

The sdequacy of the design features that are incorporated to 4

. mitigate the consequence of an accident in the =14h17 avant that one should occur are erraluated in the following section ca the )

, l

==v4="= credible accident. The consequence of this accident to the i health and safety of the public is presented taking into considers-tion the cafety features afforded by this containment system and the environmental character of the site,

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QJ l To evaluate the design featums that an incorponted into ,

the plant design to minimize the probability'of haafing an accident  ;[.. ,

a number of np n sentative abnormal conditions, equipment un1., .

'I functions and operator errors were postulated and evaluated by ]

l the applicant. Those which ve n p a sented in the Pn14=in= v

., +

Eazartis Summa..r Report includeas. -

, i

a. Manging pressure regulater handwheel ' setting *

'i

b. Continuous' control rod withdrawal or insertion

c. kss of electrical load

d. Control rod drive malfunction

e. Recirculation pump failures f.. Main steam valve closun s

g. . Failure of reactor safety valve to fement g .

h. Failure of reactor safety system ,

1. Fuel cladding failure  !

j. Loss of feedvater i

k. k as of condenser vacuum

1. Loss of anv414mg pover

m. Instrument air failun

n. Pn ssure regulator failure

o. Energency condenser tube failure  :.

..j

p. Reactor system ruptu n s inside the dry well l l

q. Failure to replenish cooling vater in emergency condeneer r/ Startup accident 1

'l 1

s. Fuel loading and h="A14ng accidents  !

1 l

t. Cold water accident.- '

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In addition to those conditions listed above, three equipment failures termed "Ma.jor Accidents" were evaluated by the applicant.

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'Ibese accidents includent -

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a, Control rod drop accident .

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b. Main stesa line rupture outside the dry well j

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c. Desctor system rupture in the dry well. ~- - I

.I' '

In some of the malfunctions.and failums presented, the i

evaluation is not yet completed, however, the applicant has stated that when the ans1vais is coorplete, the results vill be used as criteria in the detailed plant design (for exsarple, to i size the pressure relief valves and to set the isolation valve closurespecifications). In our opinion, the evaluation a sults - . .

which are complete and the stated design ob.jectives for plant 4-systems and components vbs m evaluation is inecaplete are satis-i factory emwA., s. kk f au.6 &qs he k s. r k  !

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v witt os,e exception. In the "Ccatyc1 Rod Drop Aceitest

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calculations Iry the applicant indiente that the most reactive 7 centrcl tlade ecu14 have a resetivity worth as high as .036.

Additicaal calemistians shew that if this blade sere to trop free sf the more e slaimum period,of 3 milliseecada eenld result, and the averese fas1 tampenetare seule renek $$ cot? is'the us. '

• ecatrclied fuel reme,, Tne ceasequences to the reseter vessel i in the etect cf this accident ere not er,tirely clesr. The applicant has indicated they are devoleping analytical models Ier scre accurate pudietics of the ccr,sequemees cf such a nuclest excursten and that the fcrthecair.g EPERT destructive test vill be used to cheek the model,that is developed.

... a -

f In medities to the analytical eerk, a red eerth mi=1=4=ar >

ocaputer and a red drepont velocity limiter are betag develspud fcr pessihie use is the 3edegs Plant. The red certh ecaputer sculd ec tinually acnitor contrcl rca patterns te zwinteree i

procedural ccetrels prcvided te insure that patterns causing individus1 reds to asswee undesirotly high mactivity verth are uct used. Comesyttal designs for flee restricting devices that j wec14 limit pctestisi sentrol red dropout valecities to safe

, i seines are also being develeyes. In the abeameo of experimental verificatice cf the appliesats positica that e red drepeut accident of this type vill act endanger the roseter veneel, se j i

le11ere that other design features, cuch es the red verth~

misiciser ccquter er the rca drepcut veiccity limiter, abculd  !

J le inecrpcreted inte the plaat design.

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