ML20235A092

Revised Draft Hazards Analysis

ML20235A092
Person / Time
Site:	05000000, Bodega Bay
Issue date:	06/06/1963
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 8709230205
Download: ML20235A092 (68)

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HAZARDS ANEYSIS l

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DIVISION C LICENSING AND REGULATION

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in the matter of I

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PACIFIC GAS AND ELECTRIC COMPANY BODEGA BAY ATCMIC PARK

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I;j-I, Introduction

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II.

Background

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LIII.

~ Description & Safety Analysis

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Site & Environmental Factors I

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Special Earthquake Design Considerations k

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C.

Containment l

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D.

Reactor Nuclear Systems 1.

Primary Syste=s 4

2.

Core Design i

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3.

Reactor Controls i

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Control & Safety Instrumentation r

E.

hergency & Safety Systems 4

F.

Radiation Monitoring I

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Waste Treatment, Storage & Discharge d

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IV.

R&D Program V.

Safety Analyses

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M== Credible Accident Evaluation

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Technical Qualifications of Applicant 1

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Report of Advisory Committee on Reactor Safeguards r

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,Sumary

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Conclusions T Vfi M'i,@fl

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. APPENDICES I

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A.

GeologyL& Seismology (DesignCriteria).

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Effects of Routine DLacharge of Radioactive Materials to the

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%.u Marine Ihvironment.

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Report of ACRS, dated April 18, 1963

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Introduction 1

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The Pacific Gas & Electric Company (PG&E) has proposed to l

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construct and operate a nuclear power plant on Bodega Head in

1. ,vd Sonoma County, California. PG&E vin design and supervise con-

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- f struction of the unit, a d the General Electric Company (GE) vin

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t furnish the nuclear steam supply system and the turbine generator, 4A I

Tho proposed plant, designated by PG&E as Bodega Bay Atomic Park Unit Number 1, vill produce nuclear energy at the rate of i

1,008 megawatts (Mv). Tae gross ellectrical generating capacity,

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.f Vill be approximately 323 Mv.

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The Bodega Plant is nimilar in many respects to boiling water 1

-power reactors.nov in o,?eration. Steam vill be generated in a 1

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direct cycle, forced circuls. tion boiling vater reactor. Its

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l detailed design vill be gased on operating experience from the 3

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Vallecitos Boiling Water Reactor and the Dresden, Consumers, and n

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Humboldt Bay reactors. Those features of the plant which require I

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- p3 research or developmental effort in order to provide cugineering

, ner information necessary for their detailed des 1 n or evaluation l

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Background

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Cn December 28, 1962, PG&E submitted an application to the tm

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AEC for a construction permit and operating license pursuant to

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Title 10, Chapter 1, Code of Federal Regulations, Part 50 (10 CFR 50).

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'.g The application, which includes a " Preliminary Hazards Su 7

}mE,j Report", dated December 28, 1962, 16CAmehdients lj 2;"asd'3'to the" application dated March 4, April 5, and

1963, y$d

.s respectively, has been revieved by the staff of the Division of Vf k.

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Licensing & Regulation. Technical consultants in specialized J

areas also advised the regulatory staff.

The application has I

also been considered by the AEC's Advisory Co==1ttee on Reactor

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g[g$ l Safeguards (ACRS), as required by the Atomic Energy Act and the 7

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regulations of the AEC. The recommendations of the ACRS, as

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expressed in 1ts report of April 18,1963, (a copy of thich is g,

j attached hereto as Appendix C) were also considered in the 4

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regulatory staff's evaluation.

l As is customarily the case in reactor. facilities prior to

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l the co==cncement of construction, there are a nu=ber of features

• of. plant design and operation which have not, as yet, been I

definitely resolved.

The Co==ission's regulations provide for the

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. issuance of a construction pemit on a provisional basis in cases

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such as this, in which some aspects of design have not been com-i pleted. A provisional construction pemit =ay be issued, according R..

1 to Section 50 35, lo CFR on the basis of findings, among others, i.

. that (1) the applicant has described the proposed design of the facility, including, but not limited to, the principal architectural

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.+ycij and engineering criteria for the design, and has identified the 4-major features or cc=ponents on which further technical infor:;ation (2) the o=it'ed technical infor=ation vill be 3

is required; j

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C.i supplied; (3) the applicant has proposed, and there vill be

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Eg conducted, a research and development program reasonably designed j

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features or components which require research and development; and

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that (4)'on the basis of the foregoing, there is reasonable t

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assurance that (1) such safety questions vill be satisfactorily

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resolved at or before the latest date stated in the application

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for completion of construction of the proposed facility and (ii)

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taking 'into consideration'the site criteria contained in Part 100, j.

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the' proposed facility can be, constructed and operated at the

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' proposed location without undue risk to the health and safety of s

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i the public.

The proposed construction permit, if granted, would authorize

.The Cor: mission vould require timely reports' 1

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construction only.

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-from PG&E with respect to results of research and development and' aps j

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final design of the more significant design features. The AEC 7.!.d

. staff would continue its evaluation of the safety of the plant in D

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,. m, light of this information. An operating license vill not be 9l: ;1

, issued unti1~ the final design has been completed and evaluated by A

the AEC staff and the ACRS. In addition, the definite plans for

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operations would be evaluated by these-two groups.

Pursuant to a Notice of Hearing published

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7.,. g the issuani:e of a provisional construction permit to PG&E will be ikx y

"d% :p f considered at a public hearing to be held in the hearing room of Q;pd@;t

,spf EU T /"4.j the Board of Supervisors of Sonoma County, Santa Rosa., California,

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,W%y.1 1963 before eLy.:

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at 10:00 a.m., PDT, on

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-an Atomic Licensing' and Safety Board appointed by the AEC. The

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' issues to be considered at the hearing are:

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1 Whether the applicant has submitted sufficient J d n

information to provide reasonable assurance that a

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facility of the general type proposed in the

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application can be constructed and operated at the f

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proposed location without undue risk to the health Q3fj "l 7>-

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and safety of the public; M&,'.P

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2. ~Whether there is reasonable assurance that the p

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complete the application vill be supplied; l

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3 Whether the applicant is technically qualified to

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design and construct the proposed facility; 1 4

.$3 SQ Whether the applicant is financially qualified to

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Whether the issuance'of an authorization for the.

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construction of'the facility vill be inimical to the i

M. ;I common defense and security or to the health and safety l

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l" 1',0 of the public.

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, F.., J6 (Statement of position of staff at hearing)

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The staff's evaluation of the proposed Bodega nuclear power l

1 plant described in subsequent sections of this report and its 4

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',.. f"s M6 position on'the issues at the forthcoming hearing are based on

-l ms 1A all the technical information submitted as part of the applicant's

-u'e g)h s-request for a construction permit and the report from the ACRS.

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All of.this infozzation is available for inspection and reviev'

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> at the Commission's Public Document Room in Washington, D. C.,

i and at the Commission's San Francisco Operations Office, l

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2111 Bancroft Way, Berkeley, California.

This evaluation and 4

proposed recommendation is subject to modification in the light l

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'the evidence introduced at the hearing. Under the Co=mdssion's.

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regulations, any person, whose interest may be affected may appear

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at the hearing or may file a petition to intervene and, if granted, l

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may participate in the proceeding. The decision of the Com=ission i

i g.y will be based upon the entire record in the proceeding.

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-III.

Description and Safety Analysis L,; ;-

The Bodega Reactor is a direct cycle, forced circulation

,p2k ky boiling water. reactor with internal steam separation. Nuclear i

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(:%ag.x energy released in'the reactor at the rate of 1,008 megawatts will ubs 4p gg;a be tra'nsferred to the water coolant which is circulated.through rfw Q

the reactor, Steam. generated in the reactor at 1,075 psia flows Ww.-

fy,$$je to a turbine gneerator with a gross electrical generating m

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capacity of about 325 megawatts. Reactor coolant which has been

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containing a pump rated at 29;000 gpm. After passing through the

1; turbine the steam is condensed, and the condensate after 9

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demineralization is returned to the reactor vessel. This water, which

.h. yl(%..L,, n [n will contain some' radioactive mat.erials, will be circulated within t i r,

i, a closed system from which the only normal effluent will be a

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I continuous discharge of noncondensible gases. This gaseous material d

s b'l will be monitored continuously and released from the reactor stack

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in conventional power plants,,the condenser will be cooled by water

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In this case water to cool the

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condenser will be taken from Bodega Bay and discharged into the

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m Pacific Ocean. From time to time, regulated and measured quantities

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of radioactive liquids will be mixed with the condenser coolant water

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and discharged to the ocean.

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An overall judgment concerning the safety of reactor operations fW;.S 4

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number of individual safety considerations. Our judgements at this time are based upon an evaluation af the design details, design criterior; s s,.

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7 and design concepts described in the PGGE application, the known I

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analysis of the effects on public health and safety of normal my flM.'

operations and of potential accidents during these operations.

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- As noted previously, detailed design of a most features of the.

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facility is not yet completed. For those features, the present t

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evaluation by the Commission staff is based upon the principles t

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of design rather tha'n upon details of design themselves.

In the q.

j case of features which are of particular importance to safety, the:

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j staff intends to require and expects to receive information on final 1

dl design of these features before PGGE has expended any substantial l

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In the following sections of this report there are discussed the A d[0 more important safety considerations which have led to the staff's

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d conclusions with respect to the safety of operation of the proposed

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June 5, 1963 i

A.

Site and Enviromentel Factors

.t

.c The site proposed for the Bodega Bay reactor i~s

. e.

located on Bodega Head, a small peninsula along the Pacific Qf g,

Coast in Sonoma County, California. The environmental ne i'M I

factors which are considered to be significant to safety

.,n:s.

-l q{g;g.

for this site and which have been examined in detail

l. j includet (1) the location with respect to nearby population, j

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(2) the meteorological factors, (3) the marine environ-1 mental factors, and (4) the geological and' seismological f

factors.

1-1.

P1 ant Loeatien zu

.i

[>

d The plant is proposed to be located on a 225 acre

.9 a

'cM,., q q j tract of land at Campbell Cove near the southern end of

.e Bodega Head. The property owned by the applicant, which i

f.]

includes the entire southern end of Bodega Head, is a

/

A,A,

'ed bounded on the north by the land under acquisition by C. s i.AI F

the University of California for a research facility.

., t; 'ij The distance to the nearest point on the northern boundary

. &.- 3 q uw 4%d' is approximately 2,700 feet. The property is bounded on t #,

h the other three sides by Bodega Harbor, Bodega Bay and.the we.n

[%jf.' l Pacific Ocean. The reactor would be located approximately H

.GM;.s 3

1.300 feet from the nearest edge of Doran Beach (Doran Park),

wc w piu; g5 which is across the entrance channel to Bodega Harbor. The efl.,

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park, which is owned by Sonoma County and contains no

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/{*.j:j residences, consists of a sand spit that extends approxi-Q.a.vW mately 1-3/4 miles from the mainlend towards Bodega Head j

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.~h,3 -&.M y >l:,x ;Lu..

forming a natural breakwater for Bodega Harbor.

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There is a moderate amount of traffice in the entrance

% e ia 44R channel to Bodega Harbor, primarily consercial and sports

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'. fishing boats. Usage of both the channel and Doran p ]y y

p.

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~ Park s of a transient nature and could be controlled i

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under emergency conditions if this should become necessary e.: -

Q for protection of the public. Because of the transient w

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usage of these areas the exclusion distance for this v,

site can be considered to be the distance to the northern boundary, which is a minimum of approximately 2,700 feet 1

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l(0.5 miles).,

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.' i " ! The data submitted by the applicant based on the 1960

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h, 4..,,,

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. census shows no population groups larger than about 200 within 10 miles of the site, and none larger than about 9

I 3,000 within 20 miles of the site. The nearest cities of s.

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f.Q more than 10,000 are Santa Rosa (31,037) at 21 miles and C

,(. j Petaluma (14,035) at 24 miles. The population density e,:q

.. !!:l-ji within 25. miles is shown as follows:

.MG Persons / Square Mile Total Population

. we. J Distance h.,

. 9 4

of I.and Area In Area

,. a.:m:,V.

o.1 o

o*

.a. p,r.w.

e d W.1

..1 -5 21 500

..iac/At

.c. o 9.c, s...a 5-10 16 1600

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,..., ; ;g

. 10-15 81 15,700

";@w. ;Q -

15-20 sp 97 30,000 1m g, 20-25 180 66,700

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4

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m.

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% 91 ? ng,..q %,F $,r p e,..su j.;nkp !.l _y

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l

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Qqi.ja,R:H +..,Q gi, Mglc;{.M W /$ $ W,irQ' Lad:.m.. :L, x

~f;;.dL u a _,.E. n 4.,;.x,,;'hu,. ;;;;,;a ;;.y g ggg < ~& :4,g&jnJMWA.: Wmk;ce.n ggg .,, a., t.; 9: 'ign$.w Yg t,.; ?g e.- y-10 * ~! -!ci .The location on a peninsula together with the natural ~1 barriers between the site and populated areas provide , '.O ,I% ',{ uniquely favorable site characteristics. The staff e-M j considers that the proposed site is favorably located with

y 4 'k.;.

1 ~ 1, ; respect to population density and. discussed in a 1 ster n b J..l section of this report has concluded that the exclusion I n distance, low population zone, and population center distances are acceptable for the proposed reactor. i 2. Site Meteorology .. ;4 s.

d  %.p 1--

9; ~ x., u e 9 i 't ',;, 1 i i I There is a range of' hills along the coast inland from' I .:;} -

4.,c4 ?,

. Bodega, Bay with crest elevations that vary from about 400 .:US .fest to approximate'ly 1000 feet. Under stable atmospheric s;:.- 'f ; conditions (inversions) it is expected that these hills yh 1,2 would restrict the transport of airborne materials from .a fj, the proposed reactor site to inland areas such as the i populated Cotati Valley for. releases of materials at j 'l i or near ground level. During inversions the concentration a of airborne materials vould be higher on the seaward side 4 J'Q;. 'l of these hills than would be the case when unstable . v; OT C y e.. meteorological conditions would prevail. With. regard to .e.n + w r T. ' elevated releases from a tall stack at the site it is not I l presently known whether the plume would intexept the ' *q:i hillside under inversi.on conditions, although under - h.h u.y.3 unstable meteorological conditions the roughness of the j. terrain would be expected to enhance the atmospheric !..h)

Q>,.,_

turbulence which would result in greater mixing and lower I k:f concentrations of meterials released from the stack. ':)'NfiQ

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The applicant has constructed a meteorology station - Jk,',"; - 'A with a tower some 250 feet S .:eight in order to collect Ny ?dj ~~ F 1 meteorological data that will delineate the atmospheric c a.. 3.', f' ' diffusion parameters for this site. It is expected .y. J .NQ that sufficient data will be available for determining the DDd. appropriate value's for these parameters before an operating i I.T,. n .. y;p 1 / r

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u fy M~.. MdN M... M W S@i.UfutilhItib:jjfggdM?$2]hjyMff,tr $.pg@p g j g 283Y Offf

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.' I Ah D, ' i'y ~ ,,[.' ,v. N / g , h... h .o .{, i [,I I 12 - 1 { u 1 3 .e license would be granted. For the present purpose of j ./, s e. (.,: e s' h . evaluating,the adequacy of the proposed design features i NA" '.M the applicant has suggested use of the following parameters: %..>g Un Meteorological Condition f%,al Parameter' Unstable Stable x n" .22 ~. 5 4 4 '.,...; t. cy .6 .2 1-4i Cs .2 .02 s .1 .I u (Miles Per Hour) 10 5 "'I With the exception of the wind speed value under ...h,. , p.r.~ sd.K fM. ,(. stable conditions the staff believes these parameters provide ..m, e.. " y. 'v ,s'.m. adequate conservatism for calculation of the possible > - d .$f. f4 ,,r ,~_, t. 'M' exposure under accident conditions. With respect to the ' t, value of wind speed proposed for stable conditions,' the e i,.:. , j l staff believes that wind speeds lower than five miles per l* B

/,;f:

hour are more suitable for describing diffusion rates 63 'Y1 during periods of inversion. Table I, Appendix III of . ~ w:.%9;[, ::18 the applicant's Preliminary Hazards Summary Report, which n': 41

-shows a reistively high frequency of the lower wind spee,ds

'sjy.Qf." ' in the sero to three miles per hour range, serves to confirm ,m a.,d; M.iffj The staff believes the staff judgement in this respect. WW jN.-J'@b , that a wind speed of one meter per second, which corresponds W, _Q V 4 ~ 'i ;.& iD 1. to.about 2.2 miles per hrur' would provide an appropriate ,$h ff ' -,; ~ -je slff degree of conservatism in diffusion calculations for this -,.m Fii a a site. g .e ~ ' 4 ~ ",.basN2 2 .n A .. e 'gk' p.y@.; 'W " fig i@g]E;i $f-pyR,f.M1 j;p g( j4 .g v fe $ j g.j,[.,.., y.,[s

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[ ;q, ' 3. The Marine Environment s ;. The applicant proposes to withdraw approximately ?j < U. f.'.j 250,000 gpm of cooling water from the bay side of Bodega F .s w . so,.f.u % gg --.2..,.. Head and discharge it to the ocean side of. the head

..,y d@,..

with an estimated maximum temperature rise of 18 degrees F. 'l NM5 Radioactive liquid wastes released from the proposed plant . - jl 1 .G. af ter monitoring would be diluted in this cooling water 6 befor.e discharge to the ocean. The maximum concentrations. 4 '. :4 ? of radioactivity in the cooling water discharge would .;,: n.. ..dr,., ;; .not exceed those allowed by 10 CFR Part 20'of the Gef j . f;y Cournission's Regulations.

~x&aiiryi At the request of the AEC, the U. S. Bureau of Commercial C

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• /

.{.' Fisheries of the Department of the Interior prepared a report on the effects of reactor operations on the marine environment t i of Bodega Head. The Bureau has enumerated (See Appendix

3. :. ; '

J.:f, ' ?" to this analysis) four possible adverse effects of disposal

c. 3 of radioactive wa::e into the marine environments

_ ' ',. y S. $;de, 1. Possible hazard to people from eating contaminated wo, :;nc q g ..w. .s: m. jj j,yfj ,' seafood, m 'Q 2. Possible economic damage to the fisheries of the ',l1.'.;4. E,..yj

Ud J[jh area resulting from the stigma associated with 5%

radioactivity by the consumer public, ji.-a.g. 1g .:- n: .wc ~ 3. Possible damage to marine organisms from internal' ."J.d*

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W ~ radiation which could result in reducing a ^. 1. i wS

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'Q leg: 1ishery, and e+. dia 'ik g r . N,,in A ~' ,,+.5 El 3$M .,...%.W 4. 44,._, ~

ka%h h b?Y$9 I hh h&hYh hhhNh$ .[h,h h$;_ &,ypr+tj

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y gjg w,, t r- %,.4., e g E. l t'l 14 y ,4 - 4.. Possible hazard to people'from exposure to lA .;..q? radioactivity associated with sand on beaches gdF-e.,. y 3{ u-y.y. , q-p_ .7 j.4. or with sendiments in shallow water. m. ,,y..

3..,. -,...

To a great extent the area where such effects could .Q, jp. 3

occur will depend ori the circulatory pattern of the ocean

- 4 'I at the outfall. j ..}' The applicant has recognized these possible effects i 'I and, in this regard, has initiated research and development l programs in certain phases of oceanography and marine .3 g, ] biology to evaluate these aspects. These programs, s-

1..

w w v.-a described in the PG&E application and by the U. S. Bureau ~ of Commercial Fisheries in Appendix to this analysis, are sunnarized as follows! a> 1: 1 l I 1. An oceanographic survey will be carried out to n'. 1 determined the circulation pattern of the ..4 ocean in the vicinity of the outfall, and the ao ... f.,s. m. $*X capacity of the ocean.to dilute the condenser 4i # . y ,..... w. 'j +r ~ cooling water discharge.

.d

,f... '$,NM,, .d.

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. An ecological survey which will include an tJ 3 2. Eg

a

' $q.g$ inventory of the marine organismsin the vicinity q .wM ..x

N) ac.; +.S of the outfall vill be carried out.

' t ' 3. An environmental survey of the marine environ- [ j' l%f g a. 4 i' ~ ment will be conducted to determine existing . k;'.$ d A levels of radioactivity before the reactor becomes n p'? f.y E,j operational. This prc~ gram would cont'Anue af ter a-9 - ynffy e .{.tgg c. ec y,...,.,,....... ,.m Y- $:S ' 'c R'A h4h l $_ N ! ... gu 3... _ (f ~ .V J ~

w+-.. A,..>,..,4+,?R, ~w' J. i s ',,.....~.......,nn<-s..w.'

, m.,,g.. .v .., e :g.,.,. . v.,.. .:. ~.. ....-c,. . s,.n . y.. %ph +$.ise'"k $$m:%EddmWyl$r&bt@!~'E55IW Y'Ds'"$$'?1?WW$$W$,'l&bh,5.$D Q& &[W.&7 y 4f:t. ' t;)4 N e. .c p.- ,u.y 2., x. o f f e e t the reactor is in operation to determine any . g ]t n. h j tendencies for reconcentration by the marine 4@ environment of radioactivity released to the ..y .ec [ .y..s w.t; ocean from operation of the plant. %1 i As previously noted the applicant proposes to limit

1. 3.' '., i

,{ the concentration of radioactivity in the water discharged 1. '.j ]3 to the ocean to meet the requirements of the Federal 1 ~ Regulations for drinking wated as set forth in Table II, ~i i Appendix B, of 10 CFR Part 20. The staff has estimated i that the quantity of radioactivity that could be discharged ...t. q b ? T' monthly in the condenser cooling water flow, when ndther .p M the identity or the concentrations of the individual f3 radionuclides in the discharge are known, would be I '. J l shout 0.41 curies per month. If all of the radioactivity i. e. in the discharge stream were known to be strontium 90, i ..] iG which has the lowest MPC of any fission product isotope,

.t ii9 the quantity released to the ocean per month would be b.. j I

approximately 4.1 curies. The British have shown that a -d '5.i i monthly discharge rate for stroncium 90 of 150 curies resulted . #.4 ] ib.$ in seafood concentration below the permissible limits for

pe

- [h long term human consumption. The potential discharge Y.".@&M ~i rate of radioactivity from the proposed reactor is o;..k N. fhh..a substantially less than this. ' ~'d< The staff has considered the information submitted 1 ~ t: "'~Q,- by the applicant and the coc=ents of the U. S. Bureau of i ,k. Comercial Fisheries and concludes that is is likely i- . d. t

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. rp,,h;i.'; 5& Wj$[ J. n S&: . s '.' .( %ll.,f4Q :',.lk, & . ), A' h , ;*j "'*f:) ,,C.gf g; ~....

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e,.w....a ~ ., w,.. n9,.,.,,,... y'. fiL*'d. m.p.,,.n.g... ~...,..pp7 L. f 4 ' ^ i'l*:y.,; ..,1 .o .. n. w ,..a .r.. Q - 3 g '.w.,,,is. ",. . +. ,Q, *,' b. h' met 'M %*i M f A pM2bg,,.. . w. - .u.

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.. 3.y 3 2..dM....:Je. f.,.M.'.. M 4...,%. '. A, .m-4 ,y -..a> g-p..* j = A.q, a t i 16-t l I 1 i j 1 t j unlikely that exposures to the public will exceed 4,f'.e.,, Part 20 limits, and furthermore, in view of this it .~ fr>.,;'fj y .is extremely unlikely that adverse effects on the g g %.~,,.. . us..l. . :.:.a c ~ marine environment itself will be observed. The staff n; }. i i ~;rQ.y:s.1 ~ ',.7,. ' ~ further believes that the programs proposed by th; .c 4.,. applicant for. maintaining vigilance over the marine j 1. y j environment will be adequate to discover any anomalies i i that may occur so that these could be corrected. f i l.

j 4

Geoloeical and Seismological Considerettom .c. cd , d. ). This section wi11 be prepared after receipt and review m... [. 'Qy~ 11 .. y, of reports from the followings r .m. - u .G ~1. U. S. Geological Survey - Al Clebsch 1 . i.I 2. Seismology - Frank Neumann 2 Earthquaka Design, Consideration - Newmark and Williamson y 3. i I...r

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+ E... _jf = ?. s i f C. Containment ~ The containment system proposed for this facility is one which utilizes the pressure suppression concept. Its c.L T,i i design is simi k in many respects to that used at Humboldt A j 3.; ' ~" " Bay reactor facility. ' Significant features of the Bodega ey ,y, Bay Plant con +mhant desi n include the following: 6 ..r* (1) Plans for the Bodega Bay Plant call for a dry well 3 I having a 60 ft. diameter spherical lover section and t ~ a 26 ft, diameter cylindrical upper section. The over, all height of the dry ven is approximately 100 feet. 4 l (2) The reactor vessel and four reactor recirculation loops, 73 ' each with a pump, vill be located within the dry well. I.ATA sW j h -(3).The.drywell.9111haveanairlockentrance. Personnel 1 - f. ' entry is not p;anned during reactor operation, but is I contemplated with the reactor hot and pressurized. j ! ~ '.I ![% (4) The suppression chanber vill be in the for= of a torus 1 j l d! and will have a major diameter of 93 ft. and a cross- -!I j section dianeter of 26 ft. It vill contain about h65,000 . > {-p 5.Q,j . gall' ens of water and have an air space above the water

3. ~

"a.(h]t ~ f about 80,000 cubic feet, l '.'.7 ti.,: o -tsst:1 i Mk;a.,;): Both the dry well and the suppression chamber vill be designed l . :. e- ) .eg D h.

j -},;.i and constructed in accordance with the AEME boiler and

': 37 %5k; Pressure Vessel Code, Section VIII. Piping restraints vill h5f 's[4 ldk .be provided at contain=ent penetrations to assure that failure i l cc vg 1 ,2 'lofthepipevillnotcausecontainmentrupture. A concrete .7, s l .Y i \\ W f tj

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) YlM@&Q&_p[:;f+-jisy$h3yfinhjhk&.4$$-@sw;f4hN@., kc..Si&m ? $hb$..., +Y.. f:uch;f.wn;', d,- - u Q 2 w

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$i@ N-a ..y ~ y,%*E'4,t ...,,4' j tf [' - t l' \\ il . refueling building vill contain the dry ven and suppression 1 d chamber. Pressure and leak rate specifications for these I 4 containment system components are as fonovs: b i Leak Bate (% of volume in 24 hou gig Component Design Pressure . &.n -,i 3 .s.... .. ~,. . & Ara,%, Dry ve n. .62 psig 0 5 (at design pressure) 4E9.5

.gn.;,

' y1 & o..... s, s n .,.?l39.y .uf> Suppression chamber 35 psig o.5 (at design pressure) u M ,-m IM'1 Refueling building 12 in. H O 100(at1/4in.HO) 2 2 m l[ In order to proof test the Bodega Bay pressure suppression ,,.3 design, Pacific Gas and Electric is conducting a test program at ' f.. I its Moss Tenaing Power Plant. As in the Humboldt Bay project, the I'. ~ ' pr annininr x um u xmnnuinnru 1m xra x x x m*m inx mmwumssiha x *kg ,.e .,1 35 Qf '.-l '. applicant has construct,ed a full scale segment of the suppression g- .m s ! ?. ;

g. In the test of Bodega Bay design a single 2k-inch diameter a

m ' ' g... s - e.9 vent pipe from the dry ven to the suppression chamber was used. 1;. '.i a j h Since the full size plant is to have 112 of these vent pipes, the ). '. l testequipmentrepresents,a1/112segnentofthecontafnment. %g l 'f Tests were conducted with this mock-up to simulate various ' jh; accident conditions. A flow co= parable to 1/112th of the flov ',op ((j resulting frcim a complete circunferential break of one of the 28- ,ij&s9n' ly .4.,4. in.. reactor coolant recirculation lines (with flow out both sides m.&.9 s m, Nh , ~.:a of the break) was taken as the "= W credible operating accident" 1..T@3" (MCOA). Highest containunt pressures observed in these tests were

• ' j.y%11 c

,.' i +.,. y,Q:.,.4q 5.3 wR d'y 6 52 psig in the dry well and 30 psig in the suppression chs=ber. Av..;s.u N.h.y9 ( ',. These pressures were observed vben the mock-up dry well was pre-7.- . h.h., . heated to 2550 F and when the mock-up reactor vessel water was h^?~[.?j M,f ) *- subcooled 350F. Tests at higher and lover dry vell te=peratures ' V :, J. p{ ja (,'[l .and at higher and lover reactor water subcooling yielded lower N: M, d dry ven and suppression chamber pressures. ...,q .....s s. h, [ - Nf *, l. .s hM[. h*,1 P.

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' m $ %&:4 ';) h y yyp* qi_gyce:n. 9@s.u;.Nr.g: ii W A'4 *Q M.%y.h i+A % m.n.m.m sem + m W.y.+ ~ a, b-ews. 4.' f.WhCA*C-b[A -.~4e'.-a..r 9 4 6;i s emsliGa.eb dE9mv ' * 'A r

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6 4;% - . x. - t. s<: c n.pQ)?,tjh. *. ' 1 -I$idd. @ - - ~ '- d'h *. N:' ..- s i.p 3 D /.' - (g',c-Q _ 19 _' .c 4...b,7 r .9 o.4... .l. 'l In another test a break area 2.5 times that of the MCOA ~ s.-i, p' was simulated. In this test the peak dry well pressure observed , J,.; h was 63 psig. Further Moss' Landing tests are being conducted to . )

c,4

.v ). ...;f M,, determine whether baffles are needed in the suppress 1or.Ichamber. , ;,. wk.. ... +, - % g,J * . r.r e .-As another significant containment design feature, Pacific ~ q w 3v.rl.. 2M 4 a Gas and ZLectric proposes that in a number of instances a single ~ i fg. ~ j '

• ih,g 7

isolation valve' vill be installed at the contniment vall in ,y ?1 m Q4 pipes or ducts penetratizig the containment. The applicant states, .y~ ,' W however, that each such line vill have two isolation valves, one e + -Q ; of which may be a remotely operable process valve located elseu . !, a

J ~

vhere. ,,1 PM (Note: Aiditional remarks on isolation valving vill be j .e 3, to t ' 4/5 - O. ~ 4. madeafterPG&Esubmitsamendment.) ]& :. ' % ' L, ~ .'l u. WM. ..Two isolation valves located at the dry well vall in each T* Y,. main steam line are to close on a msnual signal or automatically ? 4 on' the occurrence of any of the following: s, .d I d -(1) Lov condenser vacuum i I v..w o 4 ', yj (2)' Main steam line leak (in the pipe tunnel,) . re: [ d;.i"!:) (3) Low reactor water level. .t r u ..co a.e:pm, sg We,.1 g3.. ,m..i, ' lThe Bodega Bay design is such that during refueling, the spent .c: Ret &f. ,n !1.. Y;;f.A . (' fuel storage pool vill connect directly to the shield water above Jy. Qp@:

• e 4

.t..

'the reactor, thus permitting direct underwater transfer of fuel rw py e. . $$$k,e.! a[.[1.vithouttheneedforatransfercask. In our opinion this feature - 4'p.; j q> ' G,i=:' provides in a simple and reliable way for both continuous shielding &jrd h i

A}r.,% m.,.:% F :.., -

s 0: v 3',and cooling of spent fuel during transfer and storage. M% % wa?is fv ,.,k ' ' n.,{[- s./?& n.. r. 9 f 'k g .. '%Q,.J .w e, ~ 9 t.. r -4 .. y jp 'l y y a. ,.u y,$ O f h [*f k $ h :b k,e n m,,. w.j w w,.g n.WY$hWk55EYh$$kD'N $O*? ' Y

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w w

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s..+.:.,..a.2,, ,a 2. g ga.c: J,. d.l,.;Q Nh L E %.m W x.~::';.+~sx c'~C AN: dn;:.,.A ::nm uu.... mE4M1N 3MMM-WDE5Mk MCML' J..r.b SMdlMa jf,4._,; QW.,. >% : u...

+

p. J., I ...r q. 'Si.

• 3 The refueling building in which the dry well and suppression 1

i ( chamber system are located is provided with a controlled release 4, j;h. ' "j ventilation system which discharges to the plant stack. The y bunaing and ventilation system design is such that the refueling q:,N INO M

building can be maintained at a negative pressure. Discharge (4.&NQ from the buiW"g passes through halogen and radioactive particulate t

hk I cleanup equipment prior to discharge to the stack. PG&E has in- .ne .s ', ( dicated that, in accordance with the reco==endations of the ACRS,

di the system' vill be designed to permit frequent testing of the c

-I i ability to filter particulate and to remove iodine at specified .i. j ,f. g s...m ~i efficiencies. (The following 2 paragraphs vill be revised to reflect the &g '{. expectedAmendment3.) C.p j ":^ 'q. J;j The staff believes that the general containment scheme proposed a j by PG&E is adequate for the proposed reactor. It is our opinion, si .h however, that scue important criteria for the design of the con-n. s f, c c... taimant features have not at present been specifically proposed. i A . w.D :: Such additional criteria, including those =entioned explicitly by A

w..

Q$ @M.> the CRS report to the Co==ission, are necessary to assure that

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g['( '~ the con +aimant 'as proposed can be reasonably expected to pr, ovide

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' the high degree of integrity proposed at any time that it might

k. $..

%;f' ~l W s .f. e( ?..y}}. 9 ~ be called upon to contain the consequences of a rav4mm credible n; f ~ accident. . WY. Q(gi ~ $ These criteria involve contain=ent testing, penetration design, ,w;.. a ~~ .:. 51 1 and isolation valving. . e +: 'J J . : cs,*; .,.g i ..-3 1

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1. 'The design should permit initial integral leak rate-m.. $.

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. testing af the dry-well and suppression chamber at'their y e ,. 3, o ..o jp ,w...,..,.. ctive des @ pm N h hhh of C f. , d,y ' penetrations (including piping' conduits,- electrical. con-4 .c. m y% ,. s. . duct 6rs,; and gasketing closures) and' subsequent periodic

4%

't ~ ~- o ..nn. testing at suppression pool design pressure. In the ,l initialtesting,thei.eakagerateofthecontainmentsystem } 0 should be determined as a function of apcxx pressure up to 4;':) th.I. ^ ^ full design pressure. 'G, M.7}p.E , y 2.- The design of penetrations should tInke into' account, in Qi,%%,4

.s&,

&..m.'<.w. yr.s,,4 w..;,y e-My: :pQL.90 addition to.the pressure lond, the loads or deformations Jyp r-a.p.Q'e ,. a.u', - e + + I. %, w.- imposed by thermal expansion, impact of missiles, y- .f,, ,'F.( t. o reactions.of ruptured pipes, and disturbances incident to <s .,.! ' 9 l . W.1 ; d installation, maintenance or rep &ir., Penetrations should n :v,o.a -

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' y $s.4 . be shielded from missiles to the extent practicable. All 3 pp", penetrations should be designed so'as to allow frequent ,q #. ,M %qiR periodic leakage rate tests of the penetrations only % W:h

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..r. e.,,* .+.2,.i(including points of attachment to the conta4=nt shell), . i'_ ' ~ ge e. w s i 'e..M e at full design pressure, g. r. t 6 .s .; w m. '[k..,,b. 4,.1 1 ??;C '2N

3. ' All pipesc ana conduits which communicate with interior of 4

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• Ithe primary system or the containment system, and other

's;i. N ohiping'(suchasinstrumentationandcontrolpiping)which y&

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~ ., t y..- y a.g 4(.9 'cannot be adequately protected against accidental rupture, l, d ',;. Tij .should contain double isolation valves. All valves per- , 4. A wi 74:Mg forming the function of isolation valves should be provided > r.3%@W) s m 1 y. dQf4M, "',Nj..% 2 ,~ . 1 , y 41i. Ny{l,y,bQ G %' % % $ G $if % $.th's +gWs 'iG.W@rJ;jiid'$. d 3, k..' %. a-., - 1 : e A. -i ~ 3 u,. Ngf6,@rb%yMEEi(Gp, i.W.-O d "

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&,...nt -+ +&:. .a. wp,2C 44ws%&Apmt.'e,e g-Y..:~ p es+ 4 M M M M4 4sr>:&....n M.en:k+w:&,&&P,%,:"... MYvm, #,C<ss. Q: ig m W y3 M. h~>wMe.g.r4.* :.%,e estas M. Q..n. . fi

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,,. c, ::,n u.;*. ra ..g. 4.. %.r %,7,,.N..: <a .,,. 5 wa .-4. t. s. u,, .r, e.., s., r; 0 i . ~.ywR..W:'! w ...e

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p,+ 22.. > / -Kj% f t pMG >@gji J proper , with' protection against materials in the system which j.@ might prevent / closing and should be provided with. reliable ,w. a.: . ns w

S ~ 4h m.l. automatic and. manual actuation features. - Isolation valvin g p g g..

s ~. w..q[ -should be designed so as to permit periodic leakage rate - Tgy, [ -l;, ~ r,., thats'. Appropriate closing times for isolation valves should 1 w p..;..g .t. f.h '. be determined on the, basis of analyses of system ruptures ',4@g .x. a .Q which would release fission or actin tion products outside .m, yy, the dry well while the valves are not fully closed. . a.. - ", ^ij 'lhe Staff believes that PG&E should 'sub=1t for d (g ;i; ~Ccamission review the results of further developmental 4,.. n@h y.

g

s.

7.. ,C d T.7. tests of the suppression pool concept and final design ~ ,;4> :. -m. ,L Q. "l.J;Wb ..i. .. s ":q; J plans.for the containment as soon as they can be made Q-available. .I '.; : -} (Kote: M.7i; J Additional cocments vill be made on.the effects of earthquakes on containment design.) .w,,1? _.b

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t;%.c.&c e a. r ~ t.t pp r %.p %p 1. A r.-l m.... ,,y ., +. c.o.a.. a <- N./.N. s. s t .m..- . n _ h.>. c.. o -. c. ' $,Ne., gj 4c P@@ pw..fy'j[zem.5Jc;*.':.y 1 My,:m.W&n.w' ~pp'.., S 4. ...y J, - d' A WN,.ff,j"SW,,q$c"@" g;;,Qy . th,J v.. :. e j.. p r!5 t J* eV d S; 4 7,- er

l .'.QQ'J;hi.,kff. *,d%%Gdnl%'bi%CQ.?y Vdhdiee-$fOj,5f&SS5AAS5N5 . n.r e~ ~ ~ 3@g~ AWh*.. ', '. : ' ' '.', i ' [- j y ~ ~ p -, g:. f., ..-, C1l .e 23 - s.I i (j ' j D. Reactor Nuclear Systems F.; The features of a nuclear power plant which are,of I ,,;.\\ Nf.' most important to reactor safety are those which provide for x. Ec p' the containment of radioactive fission products in normal f 9 Reoperation and those systems for instrumentation and control! a ~ %E of the neutron chain reaction. Generally speaking the n-J W @f W.p m 3.g. mechanical design of the reactor fuel elements and the s.m : m.4 A...,c,. design of controls and the heat removal system should be 'i i ~ i j such that, in normal operations and under many conceivable \\ q ~ j.l accident situations, radioactive fission products would vs be confined to the fuel elements themselves. If this a, v.,. w f;g objective is met thin the primary coolant system serves as a's ,c.., MditA secondary containment system. M. - /.s g Primary System i e 8 f,. 1.

1. 1 43 The reactor core will be located in a reactor 3

t g a pressure vessel designed, built and tested in. .A J gj accordance with Section VIII of the Boiler and 4. . 4 6.'. :w Pressure Vessel Code of the American Society of .d31$ 9 "*nr Mechan'ical Engineers. The 50 ft. high by 15 ft. 2@.hk k;yI& ' J'. diameter vessel will be constructed of carbon steel l) II i ~' i.9 .. y, q .i - M3 approximately 6 inches thick. The interior of the m'J@j;d. v. .m,,,.. ,g vessel will be clad with stainless steel applied by x!,p 'y

w. q%m

' ? welds' overlay methods..The design pressure will be o [1235psigat575*F. .} QQ +

ww.$

d' 3 W[w@$N lsM... Steam generated in the reactor is passed through W; j axial. flow steam separators and driers located and then

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',inside the reactor vessel through two 20-inch steam lines + [h M'* .w 't ,dl )1

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gg; % s a.,..e.:4.sM+@M=f.6-% i2G; :.:.i@a;;;w%%%;..;i%w$i,'r*5?'QM&S&&& W M% W*, r~% %w ~M%g2v q;::: .di.T WL.1pmiiShm.w ~ =.v & q-e i f. O. W; .Q g;

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4, 2 .:; WT; to the turbine, which is located in a separate L; 4 Af [. .Q, structure.; Water, after separation far from c. , (1 ' [ - 'M.... steam in the reactor vessel, passes from the l . rw'ta 4 SA f e a 1-g.Jj-reactor vessel through four 28-inch pipes. Four w,y. :J g%p . ~.,,.. f, w _:. A..-,.e -recirculation pumps, one in each loop, provide the y - ~ y ~ v.

p..

driving force for circulatin'g water through the eu

a i

g. reactor core. Feed water from the condenser is ~a.u 'Nk,h./l injected into the reactor ve'ssel by a pump driven w, a 1 a ..s. by the main turbine. i N,,c. %r. Pressure vessels and piping located within the. . N,y,' .z dry well will be designed, tested and constructed in 'y;d.. fMP'[. - - 14 W accordance with applicable requirements of the ASME w . w.., w I'

,,;l@w.'8]q'q'.~

s ~ ; .!,.,4 sen -Je g od.w.;d.WEc O Q Boiler and Pressure Vessel Code. Piping outside ..'r- - a p. wwwm1.w. -]h 3 L f .~the dry well will confonn to the requirements of "10 =I a..'1 the American Standards Association Code for Pressure .. :. " n Piping. Twelve sa.fety valves, arranged to discharge ',s - 7,.k ' "} =., e g,1 f q,j.'] into the suppression chamber, are provided to i. c. - ':S protect the reactor and primary system from over-(,t/qlp 9 , pressure. (Additional comments will be made on . 4 7.9.e . y<. ny s.,4,p; $a;. .- s..i. 4 N... $7@ pea' thquake effects..These inay indicate that code r d c,. e: e.7 s,. w>-. 4 +, M. ~- T,,!h d. 2,1 W g+ requirements are sufC cient or that they should . 3*., <.j s. m.

9. '

4, be a'ugmented by further. specifications.) s ~n .n..,. a.  % $

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b 25-2' Core Design f -

The reactor core will be composed of 592 fuel assemblics -z a, e. .(..., ? sach of which provides a vertical channel through which g a, gm ^g. fy the mixture of steam and water passes. The core will sg ....; - m gr f.; 7. _,,,,.,, 9, :.q:y. ..'M have approximately the form of a right circular cylinder a s e,u. ~., : 1, 9:. ki .,4 : ; p& . '... 140 inches in diameter and 125 inches high., One hundred

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.,i d, th ..y.; 2 -t, p. I g.,

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.and forty-five control rods will enter the core from ? e 145: + 4.$ i below the fuel assemblies through control rod guide tubes. En$ l ,,q4)[l). [ The fuel assemblies:are held in prop'r position by upper and .e [7, l lower grid plates which are attached to the cylindrical +Y core shroud. The weight of the fuel assemblies is borne l+M ~fG

.1

.j , 2(b'y the control rod guide tubes which extend to the lower ,N 3 sl,,.'.%" '. -l,' ') ii... i t.,1, a,\\. .A - pg yq- .r.o wre 4.s

2.... - ~ e v.w.%. 3,s,n.; head of the reactor pressure vessel.

(Note: Conclusions l 4 {, /. @,:. ^, .on structural design will be stated as soon as seismological] 'I r a y 4 reports have been received and evaluated.) ~ 6 "M. Each fuel assembly will be composed of 4D fuel rods in y n.,. ,, g a square array. Fuel pellets of UO2 enriched to 2.5% U-235 t e, Mj{ will be contained within stainle'ss steel tubing. PGSE has '$.E M.T -MQ; -...Mi. :. " tentatively proposed that this tubing would have a nominal w u.. ~....- s m.,. e 2.n:qfe - +%g:..;.:,,.tTthickness of 0.011 inches and would be able to withstand +ww, ..w., ea g-. a. My, .o. ~m.h..M.. ..s ." I tan exposure of 15,000 WD/ TON. On the basis of present . y .v '2*NN

• information and operating, experience, one cannot be 5..;

m, ,....g,:: ap."'. s,W W' assured that fuel with cladding of this type and thickness ...u .J. e,., D, p " '. 2i P '[4, .1 i*fty

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.a v " ).,,e Q' ,, n Q L.( ' 6a y, f,4, '._.'j+-.* U. * j ~ ..,e' s 3 3 ..g.._a,., " 3 y.;. , s., s

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e . ' M h d8 Ed.'93 s'ONM R.@9MUNNMb.Md%NdN M M*M O"i@'d J$tM .p 4.- o.. n.

• '"E ' D DUM,pa r.. - 3~ ~M;f.-

c c 9 s.;;g - M-5 ~ @W44N.b.;Me66si@&S3 e-m.n@hN9+.4@M$$. 6ikpff)g~Ed. sp:fN.M336$$$tSN. s mmm w ns n ~~ tw?~m~m--m~ 4'(TlK,m N' Ji M u 'n yg;; p::.,- p p : y.. m, x .v s ' , *3 .s e O 26 - } J 1 i f [! can be irradiated for the exposures stated without ] experiencing excessive rupturing of the cladding. d . Howev'er, the design of fuel for the Bodega reactor will l 3-q dj. not be completed until further data from a General l ~ ~ WY e / d-LElectric research and development program are available. .g 7,,. 1 s.,+ 1 e . 4,

g. ' '

'In.ourophtion,thisprogramisreasonablydesignedto ] ~ 4 gg provide an eng'ineering basis for a suitable fuel element j

u..

(.;' ' design. In any event, extensive experience with other i L 'f' power reactors provides reasonable assurance that a J t design suitable from a. safety standpoint can be develope'd. i d .1 Thermal and hydraulic factors, which ultimately i +

.g

..w, g determine the permissible power level of the reactor, .. 4 ~. ~. have only been briefly described by PG6E. While there is W 9 (.. op,, .....e. .u . wag,, x-. s. r:e, - , o;, - not at present a firm basis for establishing thermal limits ' 3 '. as high as is suggested by some of the parameters specified ,j .;.. j, in the application, PG6E has established a criterion for

. e;

/ 4 ~.F uiq determining the proper detailed thermal and hydraulic design

g. '..

1; % g factors for the Bodega reactor, namely, that the fuel will l ..g.ql i ,7 Ry/ci'l ..g 'fe operate without loss of cladding integrity over the design

u. n'-

g, g~1.;f, - ?@ exposure period at the maximum heat fluxes possible . j.:4 ...a.. m..w 4 ~...., d t hf**. "wikhin burnout limitations. Operating experience at' other l S&- % w.;, w. a% WW -'(boiling water reactors has indicated that this criterion 1 ,,,.Qt.t n .gsig tewf f >can be met and that it is acceptable from a safety f.;,fs k zl y a ,. ;~.. .n Gy;yl 5%* Q t i y< ~ standpoint. l $phJ j m- ..m - c W.Oe -The power distribution which is expected in the g .s. i -t . g g.,, Bodega Reactor. core has been estimated for.the purpose w:nk.. '99 q of making a preliminary determination of the thermal margins .ng i : .,.<-.'e .b .s-c- ,,ir kid 9 ,.4 g, pa k g 1 yIi

• 4.

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~ Q MiQ %,d.%g m;ehringmQr%d; Y&2Q.3~ c w a :. m..q u %.., 4 % ietgin. 4 ygyMa Xw%C,+u: 9 u- % ifn--W y g g#EQ ~ yyy g .w>m: W. ]{Q, 4,, l f.};_ -_ .s y > c. v m,. ' e, gg ..v * ( <.1e ~;. + which would obtain in the hottest fuel. assembly. i .j These estimates will be refined by detailed calculations of-s power distribution in 'the course of design of the reactor. i.g.. 5;+s - After operation commences the power distribution will be o. y*,#. 4Ph., s. !. M.t,oonitored continuously by a system of in-core flux .l.. ' ts.w*r .wn 7;.nw.j . monitors. Such methods have been'successfully used in other l t'w._, o .i Q reactors and should provide a reliable means of establishing a .Mj 'the thermal margins that are experienced in operation. l hi.. i

T !

-Preliminary calculations indicate that at rated q 1~j operating conditions the steam volume fractions would t- ..f 4 ' il be as follows: + L ;: y. I m ~ p- ,l', ,l,,:e"-y. fAverage Core Voids-37% Md,y y.f 'yee:- ^k'.. ~ df.i.%D.$s 1:.X. Average Exit Voids-58% a T 1,J .;6._ 4 ,, Y n ~ ; ~ Although the staff is not aware of any substantial c,. ' :j. ' operating experience that would confirm the acceptability [ ]) /< of operating at void fractions this high, PG8E believes ...2 >/% g.j that analog computer studies being made will show that the 1 ~Q ,.Q plant can be designed-to exhibit satisfactory dynamic performance with such high void content. In any event, I w 5 - .,e 2.r ' 43 fhowever,webelievethatwithappropriatelimitationshigh 1 l $ Qj void conditions can be safely approached in reactor tests .e y $h' ~ devised 'so as to determine the proper range of void i s ,a.x 4, ~...

2 n.g..

m ~ content for maxmax normal reactor operation. l ,gt ,.L .,, c.,, 4 ..j 'q.,.,,

c. -

.,, " i ' Y[ T ,"l,^ "t,ek *f. M ?, .+.., .ii ...ff u = A Q f Afr ,s ; <* " t, a i J c p h.'. , i. o p, ..,3.'1-j fss[fy*; & r m =.4 . ] '" t

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,h[ ' Id. h.f , 3h d !?f 4 TW' - l, 0 l

x ...r. i i A'? &aN_ &l k s $0h & YylY1 hkN e:nN**%e/. - -ik*k. ffQ gt* p".,J,; m,m-.m..a mum;%g'f. ow.v%,.,h:u..'w",,N+ ~,}'jf?l.%~~h j,' k?en y ~ Y (~*- wm'*gv 4 n-m %y :.- ww -%w p ,, ' egem .4 +. ~ f Q.= . t i e i l ) 3. Reactor Controls T j . Nuclear safety safety requires that there be i e j reliable means for controlling the reactivity of a l p. 4 J . ? ^.; - j nuclear reactor. Reactivity can be related directly to M. yr... i . M(.x ,.'1:'..;....l'.s.ithe rapidity with which the neutron chain reaction .\\ l , y'Ym' 'p,a. a I W changes. When reactivit'y is positive the chain reaction grows; i b ~ g .that is, the rate at which nuclear energy is released by fissiorl ~ is increased. Conversely, when reactivity is negative the I.! chain diminishes and power falls. The operating condition .l of the reactor, its temperature, pressdre, power level, i i .i. l 1 1 J void content, and fuel exposure all affect reactivity. I .I 4 1 .. 1. c_.,.. fin our opinion the general way in which each of these factors l ,.A 1 .. ~, y ; NCR

55. {.%$.e*t-affects reactivity is quite well known, and the theore l

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:p;9 "and experimental methods for investigating reactivity effects l

,~ j are sufficiently developed to permit design of the Bodega i uj

}

reactor control system to proceed with confidence. "dj It is expected that within the range of operating R,T variables contemplated for this reactor the reactor 2 w.a j.h . s should be stable. That is, an increase in reactor power .m: r ) .Q.{ g _ ,y,,,: fcauses changes in operating conditions which have a strong j i m {, oo .{l

J.

tendancy to decrease reactivity and limit the power <E, 9,1 increase. The main purposes of the Bodega reactor control .y .w. ,NE;' [7 . system, therefore, are to provide a means of precisely 4.;3 @?i])., E . ' adjusting reactivity so as to control reactor power and .x.ndh (- .a v. m. jff[ ( other operating conditions and to provide a fast and _ q.; ereliable means of terminating the neutron chain reaction N;. 98j should any condition occur in the system that could, if w.-m,,q .8 '.)y* q,.fd'e'

s

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m,, 7 4 ,e , *j yd.,5-b. phj f }.v,* y+x&y.%~ lin ' . b. J .w 4- + x.r:. :Y ^ ^. w: 29 - . t 1 . allowed to go unchecked, result in either the release of fission products from the fuel elements or, damage the . Q... primary system. c

• fM The control system for this reactor is an array of m

n* 1 v ..[ ;.< :. g;g. .g. :,, Jp145 movable control blades, which have sufficient reactivity

n M '.s

';.x, ~:. ~.-)-.v w... w ;, ~., - mw

• A.; *, U-G'!

- worth"tb keep'the' ' reactor safety shut down, even though one ~ -e. y.a ;- 3 .ft,f.g 5 ..of these blades (or rods) might be stuck out of the reactor 3 q na..e .,f J f 1 <t.

core,

r.,

f .The k V".. i - t off with all control blades in the reactor core &'j is-calculated to be 0.97. The combined worth of the 145

• i'

' control blades is calculated to be 0.18. The control .l. g.. material 'wl.11 be boron carbide contained within 0.175 in. i %,,.. 4 S c;.;?.b.N O.D. stainless steel tubes similar to those presently in . s. ~ s.4w n; yg + -v T.n

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[.pu:.x,;gg r ~- s.v ' ".. u..~ euse at Dresden. ' Additional control is provided with 316 f fixed control curtains which will be semi-permanently t-1 I4 i.i located between selected fuel elements. The worth of .td./ these curtains is calculated t.o be 0.12 The control curtains i, /

d

.- -pm W will be constructed of 0.1% boron stainless steel. ,' ':W

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30..

.L l.+ i So reactor design also incorporates 'a liquid poison v.@Qgg g.,4 7. system that can be used to inject xams sodium pentaborate .--.,,.m ~yg

7. Tints the n in W W W @ shutd m e t be h-achieved. by 'use of the control ro A.. M **

V' . ;m.2yf,

• • ~*- ~ d n ~*'

i e, a.. peaqd.as. i., b..w.s w e m}t he hydraulic. control rod drives to be used in the Bodega , ~ f .!.l ~. ' 'i: Bay plant are to be designed using the same basic concepts <- t f 8 .-p ."' I as have been employed in drives in use in. boiling water I reactor plants at Dresden,. Big Rock Point, Humboldt Bay n. 1 and the SENN. Plant in Italy. Water used as the hydraulic i I. ,? y<j ?. ~ Z'( . ? fluid can be ' applied to either side of a pistou which is-2-- ^ .Ql m: .; q: ~ ~'":

, u?';V.Waechanically coupled to the control rod, thus providing for "

~ 9 .s.g J,@ My .. ~

.4 s.2

'l "either upward'or downward rod motion. Only one rod can be n.. moved outvard (ineteasing reactivity) at a time and it may j .a ,be moved either continuously or in 6 inch steps. Rod speed v: a ?.sj .is controlled by orifices which regulate the flov of water ,.. 9 Hj", away frem the low pressure side of the piston. All rods . -.1.>.5.. m c.{ i g_ can be inserted simultaneously, shutting.the reactor down. < d.A ,. % $ i:lly.. l,,. s i ~ )- m p- , H'., Rods are scrammed upward by applying pressuripid water Trom .y _y n - 7 3 _-(a . :J - -,G reither.the reactor or from accumulators to the bottom side q, .t q ,.e. r

rr M,,

s l >yr, ~ Mwof the drive pistons and simultaneously relieving the lot h s,. e.

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, ~ w.h W ' ;j;.p, volume above the top ist side of the pistons ' y. o., . :. '.y ;.y: .9 - ~ .-G...C

• L/Ulf! dump tank. Se drive is. locked in fixed positions by gj.J.

.9 w e m. ., g - q.; c [: -\\.' collet fingers which engage grooves spread at 6. inch 'i ~ A: d , v, f . intervals along the movable index tube. Se collet fingers

g~.j

'. support:the weight of the rod and the downward forces due to g. _f . reactor pressure. n. m.. idh M -%. A L _NC.M M___M____M___jdI[S M $.- did,i?d M M / W #YI @'< Nd M i M I 6. /U@' YO M. M$ND....dM. k

..,,,,..w- . Nf0$ k!. EYb NAl" 5 N NE Yh b. We f."4 " M M, 9.,.y..,, M. '.. ' /,4%),, .g a.: e< Q?h ,,, 3. W, sh" ', ' ~. h .tJ '.f ' 3; 1 1 ,q i g, .m_ m, , nn ,..v ~.7 m, -m c. 1. u,., ,- U - m Since drives sim11rr to these have been used at other plants ... ~ g an important part of our evaluation of these drives is based on j e:, . g. ? g .q '. 3.. c ' pre,vious'experienceivith these drives.- This includes Dresden; .m e .:Y.;s ..v ' a.; g experience as well as initial Big Rock operations. g. , "Qg l [ At Big Rock Point,;there have been two isolated occurrences p.- of rod " drift-out". In one of these, the cause was attributed to an inadvertent release of.dominera11zer resins resulting in o t j. the collet fingers being jamed in the open pdsition so that .9-t' .gq% the rod was free to drift as influenced by the forces due to

• dQ gravity and hydraulic pr. essure. In the.second case it was

.,...w 'y' ~ i::3.:$ ' Y! >.. c ' ' y,, ] reported.that a hard particle became trapped between the collet .. J g @.QG ,% W Wi '} - ' 2,

1. c; ;

piston and a sleeve whichlis located between the collet and the ,{ , index tube. This again is. believed to have caused the collet ~ [i;f fingers to be jammed in the open' position, thus permitting rod a drift. The hard particle was never found..It should be noted, y y, p; . however, that in neither of these cases nor in any other case fd@i tr @v bg has there been any apparent W impairment of scram + capability.gDetailed design,of drives for the Bodega reactor

• 2dk

. ~,.. k A'c- .g 9.m -.1 }g '3has not been made. General Electric is considering modifications ..y u.d&,p s~d o' r *o-l~.s a pn 4.;My. 'Y ~.of earlier designs that vill minimize the possibility of g. , s h* -h ~.. i ! l [. foreign material accumulating in the rod driveg. The applicant a M'g~@[y;q& .3[.?,: m h has also indicated that functional and endurance tests will be 1 ~e

. L

@J.- g:.T b made on the prototype L :--- ; -- Bodega mechanisms, but the a 4i.[j . _. ~ detailed procedures for these tests and the acceptability ._ p -p ,a W criteria have not been determined. . a p ~ e i. ~'# N*N f

• j$*.), ',

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-g . 32 - b[ ',.I a 1 Control systems which are designed to react rapidly to . ; [,J demands for shutting a reactor down generally. have some potenti . g., ifI ~ for accidentil.ly increasing reactivity as well. This aspect sp .c ., a p:. .;~. a.A -. 2,.47.2-of the PG6E control system design is discussed later in.this e 3t:,.. ,y.Ma @;

,e,.; g;.es;

~,.. .rV ad.;u,' a-m r, eport (Section V), ' where consequences of a rod drop-out are " ' wr ...a .o -o i ,re ,W,.=,,,. ~ considered.

t.in...

PG6E has indicated that devices for limiting a. sy,: tt. individual rod woith and for impeding the fall of a rod ...j ( l are under development. Such devices could enhance the safety of operation and simplify the procedures that are pre { j i.., j presently used with similar drives to provide the necessary j L 4 - J.j. assurance that a rod dropout accident cannot cause a serious-i , '. ~,. 4 1 _ ~,:. ',),..public hazard. " ~ 4 . p...

x. ^,. + +

.s,' s. v. ~ - Th;..;) 'ms MM -W%$.',4In view of.the importance to safety of the detailed - n o.. , w .. d4 sign of the control' rod drive system, the staff believes - \\ "J J., ; > og. i i. that PG4E should submit timely reports to the Commission i / ..d. during facility construction on development, design, and . s!. testing of the Bodega control system. j'. (Note: Additional comments will be made on earthquake effects.) s-

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4.. Control and Safety Instrumentation s

1 3 The instruments' tion necessary for safety in a @'q' ,O .b e nuclear power plant ' generally involver a large ' number of. .q e$$d'- a I , sensors throughout the various process systems. These ,_{ ,,w p,. c.-. 7.-,... s. e-g f mv n.) J. g f. q ^ ) and ' gamma radiation levels, and temperatures and pressures of various fluids. Information collected by the measuring 'i [;. } instruments is used to guide the operating staff in f controlling the plant and to actuate automatic control f. devices.

...d.

, The instruments, circuits, and control devices which are 3 ) [,h,wk...' ;,,- < .j.? : [.fof most importance to public health and safety are: (a) , s.,g.. .4' v. w'3m m, e t

n. x;

those necessary for and contributing to stable reactor operation; r

, +M;.,.; 1; >1 (b) those used in control of radioactive fluids and. effluents, 1 f, and (c) those used for control of emergency equipment.

• [sj A general adeqf2 ate description of proposed instrumentation i...;

is presented in the application. At this time, there is not

3. Sj.]1 sufficient information available to determine whether

..a...

t

.t.2.%@ed ,'7. ,.1 . instrumentation provisions h' ave been made for all essential c /;& .c l c-.... '.:n. i 1 U u.ut ~ t functions or to determine the degree of reliability t' hat should J ua1 "dh 'be attributed to the reactor protection system, which is ) pu. r;

' described by. PG6E as " fail-safe". -

f.g ; These, however, are design -qq ; w we M1 ' problems which appear to be recognized by the applicant and

r..,

m m. + a An>g a k A. ;e. wm 9;.c

which require only the application of well-known engineering

.m y e.gy ', c .e.. -+ o [,NT

1. 1. q imethods' to provide an acceptable design. The staff intends E

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, w a w?, s 1. 1 34 to evaluate the reactor control and safety instrumentation .1 in detail prior to reactor operation in order t.o assure that J . &. e proper attention has been given to the need for automatic ,e. M..ma . kp 3. J ~ j=.4 f functions and the reliability of safety instrumentation. ] A ' t 2. it .D+ 7,N. *.Xp '",'"/, "' '[ ' y*? 5 .J.,jc' ~ ~

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- l er. g.,m. 7 . =,. a..p;. .m.w w.m.. ;u 1 ..c . '%,,] .I a. t(Additional comments will be added on earthquakes.) , 4.. .~ l

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~

egs 4 a i* 1 ' 1 i ) E. Eciergency and Safety Syste=s r

n-Energency systems provide means either for safely

..$? I m . continuing operation in the event of some equipment failure j ..h. ',i. s .s s. ' .. nor. operator error or for limiting the extent of damage and l " I resultant hazard. In many instances, design features of the ) + 'e k. facility which have been provided for the primary purpose of .s. l .t \\ mnMng plant operation more convenient, reliable for economic 4 I are, in effect, emergency syste=s. Other features are { o designed primarily for the purpose of providing emergency { functions. i /. ir$ The principal emergency systems and components proposed "i. (S ~.:. .'M -19 6 47 M ffor this reactor are: , w;, 1,' r.1 M W Wc - c l >] -.A .g. ~ (1) Alt'rnate power supplies for critical electrical e loads l (2) Reactor control safety devices and circuitry; 4 x < t-t ... r.. (3) Liquid poison injection syste=s

~, a

.I ' (4) Ibergency cooling system i,..,.,

h. u.d

. art (5). Bleed and feed system

f. e.s x

i } j Z .[. -(6) Core spray system . M{$9pw ]L (7) Contain m t system 1 ] C,fdf') Some of these syste=s have already been discussed in ,.Q{.{W ..,( d j . Principal features of other emergency and

)

),.[. -,... this analysis. 1 ' g. l k; s f/.IU p :. <, q. % ~ safety syste=s are discussed in this section. N ,. ;.a. 's 9 g '. ? - In all such systems a high degree of reliability must I c:%. , f.. f..f.:d., be provided so that the system vill perform properly in ^ N Y 's, 4 j ..{n3 ~ adverse circ a tances..This requires not only careful design , i g4 .+ .x; . sd i h MM s i! M k N s @ $ N M M M U N N N d b U f d ? A N.a: N M 5 N.MN 8

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o

,.,4% $ L. * : :.y' n: ~ '(Q, ,.36 - C 1 n y p, of the principal features but also attention to such related i j .1 equipment as signal and control circuits, power supplies, and instrumentation., Maintenance and frequent testing of 3"te N-emergency systems provides 'the final assurances of readiness ..g %,: 0 .:-.3. g,., . -.y -g. ..of emergency systems to response to the demands placed upon c.] i t. t . v 'h* a "Q.%.. them. These factors must, therefore, be taken into account ig,- >Mf in final design. 1 1. Power Supply I. " Protection of power supplies is provided on several ~ i t is 1evels, as described in PG&E's application. The plant is tied into the PG&E distribution system by two 220 kv circuits .;r. .4 l3 .,; g <4f, .J *. ' l.; i .to Ignacio Substation. All plant a,rriliary power requii'e-c e. g-J .f r 4:' i' %,y ments can be met by either a transformer tied to the station generator or by a transformer t!ied to the 220 Ev a t lines. An additional external transmission line and f ..i. { Vj$ transformer of limited capacity and an engine-driven i I c,..:i generator provide emergency power to equipment necessary

8 j

i , p. for safe plant shut-down. Station batteries vill supply j 7fl

, i the electrical energy for the more critical loads.

. %m., #M x* .2. "Energency Cooling Systems .Q3

/,$.f A number of different means vill be provided for y,4%;c M.

. (;;A. m removing after-heat generated in the reactor core as a 4 .s yp;.p. . consequence of radioactive decay of fission products. .,. a. ~.: v ;) ]8.'h 5% \\ 7 J Buch provisions are necessary to remove decay heat e.fter ,\\ ~ ffs 6 reactor shutdovn to prevent melting or rupture of fuel '~ elements, which vould lead to the release and dispersal .p' .1 of fission products. These provisiens vill include: }" r?

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( v ,~ n . ~.. - ( y.:ipmye.., . :r- ,a.c 37 - ,i. f 4 o /.j g;--r w;,. (1) The normal condensate-feed-vater system e

d. '[,

(2) An emergency condenwer which can be put into 3 gd N N.. e' Y' operation in event the reactor must be isolated Wd4' m fx +. m@ ,..p. , w?from the main condenser gi .i ~ 'i, n3 ' j, ma ) -l (3) A low pressure shutAovn cooling ' system 3 .i

r,+

y ^ "-9 (4) A bleed-and-feed system vtich releases steam at a ...f \\,. controlled rate to the st;ppression pool e' (5) A high pressure core spra,t system 8 m },; (6) A lov pressure core spray system. k 4,.. ' A number of sour,cea pf water (and pumping capacity) vil( J l c.p.' e s $....be available to rest.tre water lost through accidental 9tptures .jg Q4 g ' /

. w cr through bloed-and feed operations. Both high head er.d lov Q. Y ::,

~, W i L.l l ^ + 'IQ.d .:.y... e ', 'o . L.. n. N' ~.] . head purps vill te'rrovided with tack-up pumping arrangtm:ents.. i: 4 l {; 1 In the event of a major rupture of the primary system, ') J p.s; i

1

j. J.d l

ag, emergency netion should be capable of reducing to a great t %e.:;. e + s cxtent the anc'.i,t-of ful da= age and: fission; product release W;.,. 1 from the reactor.- ~. :..k y ~.d+.6Mei , %.Z(Additional ecu.ents vill be r.ade on earthquakes'aM .-sw-6.. - .,jg.s.a {5 'y.g [* 4 noncli,151Cd3 ) be$ stated.) ^ ', , m,%... m e / s < m- ,.n e~. A-y

,;p,g..c 4

q,, :Mgj . w. : 4. y, e 7 s. .. at 1.*.P'. ,'I.'*. 1 .. ~,../. i M k. 1 '.9.t. +'E .g A>.J ~.....g 'J + <r -... p a s.. e n -. w.3 .xu m ** [,.~, ~ w 'O f'- [+h ' J' d' '* M. N*d, ,. g p[Q{j G p. n g.w , pf...wl) (,* 1,** . .s.,,,: y. }ef g "

• 5

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g, d _j. s .. ' M. ?" L,1,.,. s e, :3.s.pv. e.:c...~ >.. m ". - y 9 o " ls. 4 K <T. Radiation Monitoring Radiation monitoring equipment will be provided for two ,g-i 'M L purposes involving safety of operating personnel and the I .: %, 1,, s

';fil general public

1(1)formonitoringofradioactiveeffluerts n

2.tly I g ].ll and"(2) for deter iniv levels of radiation in work areas in 3,M .:gy the plant, he type of' plant proposed and the environmental . NQ. iR conditions do not pose any unusual requirements for monitoring u.V + . '. {. methods and equipment. he foll. sing measures, which are typical of other 1 [ facilities, vill be e= ployed at Bodega: t (1) Batches of liquid vastes vill be analyzed .. s s* .!.[.[ [~ D c radiochemically. prior to discharge to determine . ':q; e yl_,., p. j%m -C !.7, quantities and types of activity present; ~ av m;;.. ,"C ' g a.. ,. l (2) Primary coolant and fluids in various aux 111A7 I, ,c. w

• r g.

q systems vill be monitored by radiation deteer.o1 . p gj and as= pled'for deter =ination of quantitles and 4n ;T, 0 ..1 -{} types of activity present; ,.s. ,. t ;..T (3) continuously. discharged gases vill be continuously .t'

c i

.. ; :. 0 \\ Og { ?,J'4* monitored to determine quantity of activity discharged; )

n:

A .%'.oi;.;A D ~(4)' . w> g,9 program of radiological monitoring of the environ-i '.w.1 2 ment vill be conducted; and e N,.,akfi l (5) Fixed and portable equipment vill be used to measure ~ O.. SW nt. .. J..-

a..n -. rmiiation levels in occupied regions.

l e .a ,/n" 4l4 -'< 7 1 ,1. 1' . :.' J.. v., s.. i./f.4p;p) t ,4 gxz -. 'i ;+3 v.- .~ . r x.,

, v. -.; *..

...,...o n' . +. _..

p

-fi; _ f q.;T

p*.

.w.._p. $ed. 3 n gm4 v .e,u ,,s%. ,.',c., y' 1 gy 'Q* 1% 8f . -e ,s 7* &$ &%i%%idbSN N$$Yh%O$? Uk * ' >;S'd 1'WO*"NY', ^

4,w;ra.v Wm+.% t.'dO.a.k*d M$dh'hbOhh~FMwl:v.tsW?d'MeNI$..r73M'M.m$'5ClftMT ! E my Wh wwu - -c. ww ~w-p+-%gew.mm k%% } Nf L::4 r ],w .,..~ ~ ~- .c. 4

, C.

Weste Treatment. Storsee and Disehnree i The applicant has described in general terms the w.Ea q v ,g radioactive vastes that would be produced during operation Ff52 i 3470 , of the proposed reactor facility and has proposed general g.$..q; v.; n r .r.... 1 7.K?f methods for management of these vastes in order to 3.: \\ t '; meet 10'CFR Part 20 of the Commission's Regulations. $j./b %' g. The sources and general character of these vastes and I the general methods proposed for meeting the health and p q safety requirements are su::ntarized briefly in the following paragraphs. 1. Radionettve Lieuid Wastes .c

• r51 W.,

'p. ,s g >, h q 77 ;./.lf, ,The principal sources of radioactive liquid v .F. a ;?. i.,: 2 a e .e A g ' +. .u . -. - '..r.,4... ",y., e ..vastes from a. plant of the proposed type consist of .4

g.;q

.g.. . e,., j 'small amounts of leakage of primary coolant from e valves and equipment when maintenance is performed, 4 and vastes from decontamination procedures. Other

.1

'f'i sources include laundering operations for contaminated '. t. Jh[@d /L clothing and laboratory operations which are carried jp . " s.. w '., ' out as a part of the reactor and power plant operating 3,d.9 r .".(id'T.f control procedures.

.yW

:

/fG . 0. 4. :.r.+. The amount of radioactivity in these liquids + dl.Ir.y'b gj.'.'[j ~' E. is vartsble and will depend upon the concentration - g. 4 W;;;j

p.

..f present in the primary coolant water system The p w i . ;.w., ;: y q..; 43

/

g:,.:7 radioactivity in the primary coolant consists of ~' ' ' * '

,'+ 3 i.b y

. ~. ' fission products that may be released from the fuel . W. s.

c,Q S'.fM through minor imperfections in the fuel cladding, m

' e. ; :* O r; ^ r, ~m % ;j f*,u,' * - ~.. a ' ? '*.

.-. "d.',

4 : - d. O -c'.$4&,Sdg*,s.- @..~~."' (>

• \\

y"g'Tf] a. m.:w ung:-i,.p?!c%tp'z% W

• ' 4 - 9. @it %Wh., M;N 1 u.

s@p % u g.:MP %f W e t

+.. ;.+ n l.e v.. ..,..,m s =.:.4 ._,,...y,;... .,v.

y. Q ^?'?i W'N## CfM W.GS W W W SM W.gif4'P 9s?.yG W Q7%i%'?,' DIM $$CEj@ A G G.l, Q';;;b,t.. y

,. - 9 '.~.;. a M9 m. - A. .x. y fry p,W ' p 9-ffg ce s - a 4 ,N i - 40 t i w.e 7 f. and of irradiated impurities that may be ~- . 3;.. f., 6.! y4j { present in the cooling water. Such impurities 'Qc L ~ d) - .r...7..q n o.~.. p..,. c would include corrosion products from the coolant 4 9;.p.2 ,.v ~

s. Q

),

p.- + 1+.& system and fission products from traces of ~

• A.-

u <' m:4 .5M.i.~ ; ; ~ uranium impurities that may exist in the fuel Ty [,:w*,.- - l cladding surfa,ce. !j. I Present experience with this general type ( of reactor at Dresden indicates tihat the range .ae ' i of radioactivity concentration in these liquid .j 1,, 3i, (j wastes may vary from values which wou.ld be low ^ y;ay,.:.;, "".s I/Dj.,, [, ~ ' p ;o', , (.. ;enough to meet the drinking water requirements 4%,.., ,,., : = zg # 2 i.. , '.:z.j of 10 CFR Part 20 for the public without treatment e .. Q, 4 4 to as much as several microcuries per cubic centimeter. c j The volume of[ these wastes may vary from a few a ' 4 ~c ;Q':} tens of thousands of gallons per month to a few X hundreds of thousands of gallons per month. , c. .Ml::..,. ' The applicant is proposing to construct a

,.n

..c. u..; 'yy g?; special system of drains and tanks for collection r che . g.u ,i z .~g ~. ~ - 3.1i

• 1*,,6,l

[ .9,* mQ; 'g of the radioactive liquids from all potential l ..r,~

,1 n:

a %$i sources, and to provide a Radvaste Facility for 4%$ c)' Par @ i p' i.. g . monitoring and decontamination of these wates. . qk, qg i ,,..,,s MtM*$j w..., .r "T Radioactive liquid waste will generally be disposed gp. e

s. ' e 1

?0 St r of by injecting it into the condenser cooling i.x. W y s.y ,N..$.: -R) water effluent stream after monitoring for -./,m, eg - n it- ' h:3$ I 9 . This disposal will be so radioact!ve content. .,Q.fi x. n .s... 4. k * >nk. j;.h0?hl$,bQbN$k;f WM ~ h k Y N N ?' @ : '. L' *bY SY 4 NY

I 1 '.,,M %'.a'g %.:. 4, w' + w-un.w-~ v.w-,., r ~~t%#m&ww.n~m^ G 98.m*tvwmTm"".tyrT_T'.M kW, x

Y wom c

w-w - @. c. t - 1,. m JJ m:,a ta - -r s fto.f,[th b h.hj% V 6c.,C p m 'w '.. t3 +- -+ 41 - i i 4 N controlled that concentrations in the effluent .v 1 pjg

1j_

3m ,. stream at the point of discharge to the Pacific qr i ,.,.,'.4 .y p.c ,; g.u,.,_;...t,.p y :+9j,. t ocean will not. exceed those specified in 10 CFR 20. .. m 3 n .r. w, c ,<me n f Md,a y ~ M)& The staff has reviewed the general concepts ' a u ,1 - y fj.; : .I g 7.;:.2p ;.v.. a.. ~ R8 proposed by the applicant for the design of ] ^ this system and believes that the concept is adequate .j to meet the requirements of the Commission's Regula- ,I i I tions. .j. 2,, Radionettve Solid % stes , i$..- ,; -...:.vp:si:.;ic h-7.: T The principal sources of solid wastes inclnde: y L ll -

• Wn, v 5

A....%. , f..*? A L4

• ' v. pow}.b ' r. 4 4Q, 2

p,.,e: p.:n.,. ,r

, w h;y ng.,5 4,_ mm'a ppent7 demineralized resins, filters, scrap g,g s, -

,). o - c;. + .,t. j equipment and miscellaneous trash such as hand o. g., .I i tools, laboratory vare, etc. t. s ,.c The applicant proposes to collect the radio-4,.< ;. 2 ' : '.fo.t. active demineralized resins in storage tanks .,w.,. .a. bid which would be located in the Radwaste Facility .:n ~.N...gjh.) -s p

J

'and to dispose of these materials by shipnent f. <a. 3,3- .gh.;,sd; _.....v,j. ~. G " to an AEC licensed waste disposal facility. .:.ee.g

c,...

s 4,;; a.y ., w w -.. J$g ,m., 7 - 'The shipments, of course, would be required to ,b gg. r. . f. W.2 :;.,t meet the appropriate AEC and ICC standards for .c,,,vg

e. %

.. u

y, 4.,;

p g .... u.g, ,, 3 + g+- + shipment esf radioactive materials. All other g.f.g.y .prrm a r k,g c a... ~ - 'yt I hin %,y.? :.5 . solid waste would be collected and stored in a 1 p ,w

• i mv - ~~

1 9)[.M.y Q. ?...j-s, vault constructed for this purpose pending u: de... ., m.. $<.d .R packaging and shipment to the licensed waste 1 .,.a s ,.:..,; )pa. l ,G YUF disposal facility. m

v

';.tk 4, h,

• r

$., y &.,;. 3. + ' ', .M / s.M..%va$m mr. gh.Od %:.' W* > "3 ume p Q j$ v "? w " ;l5b&ir .a. ~ u'@ ?@:I.hr~) S ' ?* ' 2 " W^ C': ^ ?.

• d
• 'L

+( - _$ % i< wSQ3.D~f3% # C14 8LQ W 61% % W h M d M W W i% W 6 W & E M N

..'N. mRm Mr& Wit:2QMf6.Rygg:gygg&.p}prM&hgp@g q v I h.

m..., \\.......c

' y,. W %:p';, C < f. :& r i.1. _ :.,,.. - ,g . ?,y g 4 .j ..e s-y5 42 - 4

r.. p.

d[g.: 'M$f T.' The staff believes that the concept shi yg.' Proposed by the applicant f.or handling these ." -m .t., '3

g..

. solid wastes is satisfactory. ~ . ~.7. 4, ~,,, m j e., p. e.

n.

.s, c m o ' S,.s.,4. ,3. Radioactive Caseous Waste r.. m ..m t (iA,,, 5*' Gaseous wastes from a reactor of the dy,pe. [ h.. proposed. consist principally of non-condensible radioactive gases that are removed froin the main

• 9 condenser by the air ejector, and from the' turbine I.)'

gland seal system. The process areas, launday ?j Qy ' 'fc l V.,,.

and laboratory will contain trace amounts of e'

h s m %.. u. 4.1 %:m 1...::,.... &J.. Mk;J f f.d W S f. O+w, + .a.j & q Q contaminated dusts, mists and vapors. Further, u ., Qyr e .4 y:W a s.3._ -d;M+ y.,. P " the dry well will contain radioactive argon r E.1 as a result of neutron irradiation of the air within ,. = - this cavity,.although this air would only be . r:,d.A.c i ]'h' :) .. '.t ,e released to the stack on an intermittent basis g s. '.,.. f t whenever personnel access to the dry well would M. 4. I;f'; NP be requiredd for maintenance purposes. -WJj,;. klh I?. xp. TThe air ejector off-gases, which comprises ..c. : c ... j. w . i. s- ,1 i. M the bulk of the gase6ns radioactivity to be c. < ;;; .,,k,. s j., . y= 4 I h5 y released to the atmosphere through the stack,. Jy.%a.n .~. y,i j..- 9,. consist of various isotopes of nitrogen and c i i 4 3 r,. J * "$ a ..._, f.r.. !. '.',,.,R, oxygen..The off-gases also may contain non-e m y, y, ~ ,y.,,W condensible fission product noble gases, mostly ~. ..e.v Q] j.*- zenon and krypton, that may leak through minor ,W, ,:l \\h i F

(', } f.S

e. [,

.L n,4 s: 'k Q.. s -, m w..:. s. [S ' [.' v'* '8 4 4

• 4 4

' @ A*),4 '[.' e gi e .h'f. p f . p. }* ' 4.* 'r g L h 4

• [ NYb $.d**r i

5 E.' M *3 ' 5 O.

5.S h 5 & !h-$ h $ $ $ h hhh h

h. b'?

- - W-W2' $rS$ 55; uw %E q!q. N'.'.".l ~. '. .+iciY:lU.'Ti%+' t,. e. w 8., A!G 't ,a ?y.j}.,,,:,* cg. 'm <' , V2.t -.i g. .'l y 43 - n.; g .y imperfections in the fuel cladding into y'Oh':.M.)' y the primary coolant system or result from p

4..

J'y ;. _4 .~,.&..

• 4., ;, p.

w. .:.. m r." irradiation of trace quantities of residual g , me r s.~ .4. ', 1) ,' uranium contamination on the fuel cladding surface.

s...

@!%..i The applicant has listed the isotopes of j 1 ~ ,,. l, nitrogen and oxygen which would be present in j S '. -A t the air ejector off-gases as follows: N-13, N-16, N-17 and 0-19. Based on the experience .K ).i st the Dresden Picnt the staff would expect that { .p... ..e. . - a , Nitrogen 13 would be the major contributor to f s d,... l, 2 s y;';;... \\ z .,up t, 7.4s. -.,. w... V.' - m.. ,,p. E'Cthe total gaseous radioactivity release to the ,,g,g.; - A.. ggn r.p.4; 4, 7. 7a atmosphere during normal operations. In l~1 this regard, experience at Dresden has shown a f 9j that concentrations of radioactive gases released

n.a

) y;2.1 are well within the limits established by the L*; $ :.:l. Cocuiston's Regulations. M'[a'i v 'E t The average annual rate of gaseous radio- ..y y,,o. q .%y.;/, . y.m s .. + e -., ..~-- o@e e activity release from the stack will be limited . l .y ;., a. n.

4. -

-m KW 2 g 2

ny....

to.a specified quantity at the time of granting M;p.;y p., . _,e f$ 1 ?W3

41,:;

of an operating license to assure that the .y, y '. gip .a_. 7..y.wp/_-.F. requirements of the Commission's Regulations are '\\ 4 m ky 9 ... pr,n+ a.- - .hh ' 2 d h.$ h ymet. The meteorol'ogical and topographical condi-d -j. ..., ~ ~ u 4 .i n.: n: gg -} SV 'tions at the site as well as the engineering y e m.: :.,; : :. v;.Dj ~~ 7,<.'J n$ design of the plant will be taken into acccunt m -s.. %;g-lJq

'[] % ' when th'ts limit is established. The applicant y

p.- i . g g..- . w~n.; w

a -

e a { ' '

bh h$$ h k WYf 'YYO N OI*M Wf~Gh$d9Mesndhr i;fR'l..l~.tfMW : g.Mg.gl., w,

• g.-

s.

m. i.p&'k p:wh,.gy y.:. ?y...: ;y,.: .v n. . m.. . p.,. .n, R g2a- ~ ~ 'l ff-t. 4.

.a M.b

- proposes to measure the amount of radioactivity .c D.

fgg

.:QL'

"~ ' A., re eased from the air ejector and instantaneous l .i s, 2 ~ . s, y.. ,,w .; s ' - .. + y Q, q,, J.y.:., /a ".. s.,, ,:f 1 - t,.,.. :.,.w_,gu g, r,,;4. 7[ Q ~ / radiation level to provide for an alarm and for c p.g , s ,.. W ; '. 4 M. ~*=. ./,.,,. mLt ' i l ',i a. ,3'. j'l :*['p',.,,, ;ey 4 ..,~

7. -..-

c. .g.. v ' closure of the off gas vent system valve '_3 ? 7;,, r, .r,.

1. ..w.. O..,..~. ; :., '..

m.. ~. r,.. ... ;..,E. g..;<v c e,';Hif the radiation level would exceed limits pre- .c e 333.. .1 It - A M; i - determined by the Commission to be acceptable. .i 1 Q

t The concept proposed by the applicant for

'~ i -F. ,j measurement and control of the radioactive 3 . i. . gases appears to be an application of conceptions @,pjl y, ..a-?. Z.J 1 -- ;C...g a.P.. ch..n.g?.;.<:.:;that have been previously approved by the Commission w, ...t -e,,. u..,,,.~ Accordingly, the staff ..;,....,. n.,.. n -. .s ,. r 4

q+;p.- s c

.

,y s.u b. - y-,..py32. ], R..g.,r.M,.,au; g.,for other reactor plants.- , %qa; y 12 i?i . e. 3.,,w..a,,: 1...., g. c p%... M ' believes that the methods proposed will be ade- .j w. g f'] quate to satisfy the health and safety require-o a.- 3 9., ments for op,eration of the type of reactor .p ' q..e 19di facility proposed. 4 I ".. 2.,..,.y ~ s - %d.) .~ ;~4-v;,*iQ ^ ~ .{15'q* D c-f,, t,

y!.

• **'N I. a.

Q '*,,

.? *b a }4$ g ; 9 ,.-.g..g. 3,y, w., '.. .5 1;p?r alW

• C (::'.y :.., '; Q I T v."' h :.\\.

?? r,.. . l gf'{ i ),,'K j Tr'*J.$ 4T.g:.J Me M)"*. b ryer g,'k* ;de. @ o ~l ",P, c S ".? '; A ew p. 4 M.b [.,[' /s b [. aw;+ . hq * : %.s. <Ms. ; e W .g< -Q .a y " ek ,K'I a. D h /e wa,p 4 ., } uf s_.'.1pf *.3,

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n + n .M-g[,Q;J( 4,s u .e. ,c r *s. e 'l+. ld... t i .,-.,.t(J, c e.-

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- h,,, h. '2. Q h,.: ;? C.., f* 4 ~[ w s . ?.w; y .e + . t N 'i ~ n, 4 [, 3-~\\ h4 [i, * ;.'" ~ ',,k f~ k[,, ~ 1 s

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g. .)

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~ 4 s , (,_ 'j - 4 p. s ..t'. s ' kg, u hk ,, qa ~;r.3.; us.. Q. S ~,4C.rt~ y Q.- : ,.s q, ,' & 0 y, ~ :., A ~. - m, MMWMEWWMMWG3#@wsWtM@mSFMM&M@M

$$ ihh 5eb$5Y WW NA'M WMIbNY %hh !,k 6 y g q..f )y*C. " Qv r W W L -:::;-3.;=.a2 W n.e ~ ~.4s

s. - -g c.. g 3 6 b?

1 .t 3..g. p j t s. q;t.- I .t m 2 4 7 s I;;3 a pa. : . Research and Development Program ... ' d 17. y'd While the design of this nuclear power plant is similar in many We ~ !;g respects.toother plants now in operation,' certain features of the design W h t" ,,,. req 61re research and development work to establish complete adequacy of j .. w. 3,;..:y. 4 s,.,..... ?L ]2~ >fintendeddesignandoperationalparameters. In recognition of this need, W,- 3 Pacific Gas. and ZLectric Company is conducting research and development M.$ Zl c.W .c programs as outlined belows. .p 4 . g.. + , 2E (a) Meterology. A meteorological facility is beirg installed at the p. site to provide necessary data for atmospheric diffusion studies., d .j Instruments will be nounted at three levels on a 250 ft.' tower and -4 c.e 1 [.~.f 4 7 'q . will measure temperature and wind speed and direction. All readings i "*r. [, 4 -'.i.. s . _ ; j.0 y.. ,.a.e W<-h.. J - 7,:ic will be digit'ized'and,r'ecorded on paper, tape. d. m. p% .,. ;w~, ;% . y., n::.,,... 2 . qa e. w ..a .ys.l.c 74

-@(b)G0 oceanography. <The '; capacity of the ocean to diffuse the conden, er -

s u.,.,, 1 cooling water and minimize the effects of temperature and radio-j ;d. activity on the marine biota is being investigated in a series of tects 1, ;)al conducted at the site. The'se tests include use of drift poles and .ptM . p1 uranine dye as well as measurements of temperature and ralinity. ,v, 1 5d They will continue through at least one annual cycle of oceanographic [?$:'hp'h -. W. ,m.. i and meteorological conditions.

u. +y n,p a,...

w, o - > ~,m(. i :i N. c). Marine' Biology Survey. . ecological survey is being conducted 'to n;l@%y,y g w -x- ' my,! ("3l

r-

., 3 m prepare check lists of the marine fauna and flora of Bodega Head and n .X.. g.. ',.- [ T k;.Jhu - -m. p p.,.@ ,., m. . Harbor. s ,,kl3r w $n ..~ g_ .m . 'u s. , M;h..lh.1k, .[' ) RadioldafSuney. A preoperational monitoring survey of the site " s. . s. .~ 2: . V:~g tj 4.,,.. ggy w and its environs will be initiated two years before commencement of i b.c. a; }.6 ( y + . r?.U . operation of the reactor. The details of this program have not been ( ~ I]... i;ij completed for Bodega Bay. However, it is anticipated that it will be ~ s .,}-g ' simib to that conducted for the Company's Humboldt Bay nuclear unit.

. A:

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Pressure Suppression Tests As described in the applicant's hazards l ~ .. ).t x E

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summary report, Appendix I, extensive tests of the pressure suppression 9 Qjl[ concept have been conducted. Additional tests will be con' ducted at & :b.a ,.M. the Company's Moss Landing Power Plant to determine whether or not d.gn: - u ,... ~, _ x, a....,., .s.., gT. j ' * ' " haffles between' vent pipes are, required in the suppression pool. n% i d.k j,, In addition 'to the research and development work being carried out by g'. a.s -3:] l the Pacific Gas and Electric Company the General Electric Company is 1 t, carrying out a number of research and development programs of safety q g significance that will influence the design of the Bodega Bay plant. These,

• 7. :.

are: .,3q J@.;:<cfo - 'M',s:.$,es-7 % (a) Fuel Development.Results from fuel element development tests and i D.;' ,h 4Q$:(( ....'c-th'e' selection of;the Bodega fuel. .w ';7;..%j. experience with fuel des'igns now employed in existing reactors will .,..;. c, nsu.. 3 -. n. Q,,4 %R ~ ' M ;y g , x .A; "". e

• (*.: e sform the basis 'for.

(,., ,p ,e .g : ;, O! g ,(b) Instrumentation Development In-core startup range neutron detectors m, 4,q arb being developed as a possible substitute for the previously A di planned out-of-core detect rs.

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. E.;j (c) Control System Development A prototype Bodega control rod drive is

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~. Q@%? "f currently being manufactured. It will be subjected to extensive , j. ,Qfh. developmental testing before the final drive design is released for a m .n -,- s.+., o.-... D$ [ manufacture. Several devices which could reduce the likelihdod or

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, N'-X$ $5SM ~ ^ magnitude of a control rod dropout accident are being developed for o M .m . possible use.in the Bodega control system. mw y .. > u..w a.g ,_m. '..Q) Nuclear Excursion ' Analysis ' Development Analytical models are being W 1 ^ =3 , g',, ;., Q,,. ? + .\\ 4- ^

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p. .,+. e u develope'd for the more accurate prediction of the physical consequences gy .u rep 3DiW [ h.:h of nuclear excursion. .W MC It is the opinion of the Staff that these research and development progra'ms . ;qg jy M>" m, - are properly oriented. As provided for by paragraph 50.35, Title '10

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h.4, g, et,4.,- of the Code of Federal Regulations the Staff [ f-fb b' e informed of the. .w. 4,* 4 results of these research and development programs and will consider these j p j .}:i results in its evaluation of the propriety of granting an operating license

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k%j previous sections, and in many cases the safeguards provided by

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a particular design feature or the operational limits imposed by fy.j& 3,; 9. s w p. 7.s.7,,.w 7 safety' con:1derations for.a particular feature were discussed. gi. c r . i. ? -m*'., i .e . %.o Q. In general the criteria.for plant design should include: (1) %y"a q j W"4 '6 means to control radiation y= e (i.ncluding discharge of radio-

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uy, ,......a ',; k,9 g. activity) during normal operation; (2) design features to *4m4ze .y l l I l the probability of having an accidenS; and (3) design features for d ', mitigating the consequences of an accident should one, occur. 5 Se means for controlling radiation during normal .4.*A Aa. -_ s .-1x-..__f . p;..a.., ,m . - ;;m -r-ty . SfMoperation, vill be provided by suitable shielding and radiation -- @W Q

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,,,.se, Je,.j plant, the release rates shall be such that they do not result in .q m PJ personnel exposures in excess of 10 CFR 20 limits. >-1 ,);. W.h . 's] he adequacy of the design features that are incorporate'd to .cc..: mitig,ste the ' onsequence of an accident in the smlikely event that .s ...qq c s

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' j #'n 4 G,M.% marinnnn credible accident. he consequence of this accident at'a Mdh .,,-\\ w #1 - "# -.*(pw= 4u 'dd.<. A' M-j?d .~. ,,,l. 7, +k-k-- ' +k

t,- -* the public is presented taking into M,Q W

3 D yg $- considerat' ion the safety features afforded by the con +a4-nt '1 . n.. ,. 4. - ~.g, s. M,'.% y l

• 4.p u.J, o i N :Ayptem and the environmental ccaracter of the site.

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47 l i j 1 \\ r.i 1 4. 4 To evaluate the design features that are incorporated into j j. l the plant design to minim 4ze the probability of having an accident 4,, .gf p q, _a number of representative abnormal conditions, equipment mal-j j . r... sg,p,; a a u c. j n6v.,. ~.,vfunctions and operator errors were postulated and evaluated by ,y yn.,.; 1 g fG[ a ,,,the applicant. Those which were presented in the Pre 14m4"a v y 'h ! ' Hazards Bummary Report included: Ji' d',,I Changing pressure regulator handwheel setting 1 a. '. 0 0 1 j d. 1

b.. Continuous con't;rol rod withdrawal or insertion e

. I 'j q c. Loss of elec'rical load t o 31 J l . d. Control r,od drive malfunction l +.. y g, ~ g.; '.~ :q e,. Recirculation pump failures ~ 2'. ,T d' T.. 4:/:Vllf 'h. z ,~. L. o .... ) .5ke.'d f.W. 4.' @ T 9 % 3.f 7/ Main steam valve closures .~ J '.p '. M l ' [q.% a ~ g. ' Failure of reactor' safety valve to reseat r d >I- , j h. Failure of reactor safety system - ;4 * ~ l 1. Puel eladding failure 3 ~ i .il 4 \\ .J;;)

j.. Loss of feedvater

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?,1 '{'d Fuel loading and handling accidents .s-l "O ' t. Cold water accident. ...i., s .s J c.,, .m.

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i' 48 - l . s' g.j - o i ~, M .) ', m In addition to those conditions. listed above, three' equipment 5, ' a; L failuns termed " Major Accidents" were evaluated by the applicant.

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,1 These accidents. included:. 4 w ..a... e .i. 'h.Y.s,[.*h . %. ~,.,+. 1 < + MX ',_ r C+4 .~ 7.; . & ~,. n, e,. h ~,': g ..r.i.: .~ a m,- .,,, ) ',r f. Wer,y#;. i c ... v,,-.s-,t.s. ' Control rod drop accident ..~.; k.e, ;. c. ~.... 1

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. b.' Main steam line rupture outside the dry well 1 . 2. W. ,,r.. t4 : Reactor system rupture in the dry well. y,,. c. y-i ,, 'a In some of the malfunctions and failures presented, the . l' i 4 ] evaluation is not yet completed; however, the applicant has ') Y ~ r stated that when the analysis is complete, the reculta vill be .,b, -.. ~ ,jt i ] Q used'as criteria in the detailed plant design (for example, to a;g,

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,.;w. s. .... :. w, i... - . pu.... .,...,myy... size the pr. essure.'..re.. lief v.alves and to set the isolation valve 4 .--w a ~. .-,4 ~. ~:. a,,,;:.s rm n.~u n... w... e. y c ~ ~ x ...,.~,. 7 a,..

+ A g. y 4 $ ' closure specifications). In our opinion, the evaluation results

..c d " J R.1 which are complete and the stated design objectives for plant J ) v ,.p l. systems and components where evaluation is inec=plete are satis- ,. q a y y ' ?.5 - fact N.1

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. ;.? Additional calculations show that if this blade were to drop 1 a t,g,,,g ....m. q. e.g f ~ L 61., free of the core a minimn= period of 3 mnis eccnds could result, s ~ q[: h. .j ) .y,:.)." '" 3.c.. and the average fuel temperature vould reach 55000F in the un-s . t. . controlled fuel zone. Se consequences to the reactor vessel 1 in the event of this accident are.not entirely clear. Se 1 i 1 - ]'l:- applicant has indicated they are developing a:falytical models a for more accurate prediction of the consequences of such a a, g l nuclear excursion and that the forthcoming SPERT destructive .;d.. A: n.

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A M,.: e w',..%~itest will be used to check the model that is developed. 2"

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.g, o In addition to the analytical work,*a rod worth min 454zer e.

.j computer and a rod dropout velocity limiter are being developed 1 j ) .i for possible use in.the Bodega Plant. Se rod vorth computer i t 't .ni..,a, a vould continually monitor control rod patterns to reinforce 4 3 \\ (w procedural controls provided to insure that patterns causing z., I 9 insb

-s.,p.4p indivM ual rods to assume undesirable high reactivity worth are

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.L <r. @m u.not tset. Conceptual designs for flow restricting devices that .i .. r - i .u. ; h ?.*:! 'Q.U.. vould litit potential control rod dropout velocities to safe ~.- s y;.3 e . ;,p,y .e [QT ' values art also being developed. In the absence of experimental pd M.a.,g %g.i[1 .-j' ' verification of the applicants position that a rod dropout 9 d.ff , 5 ' ' accident of this type vill not. endanger the reactor vessel, we 3& ll ..., yu ~.q.q.7,,..: a .G %.. m a I [A';% ,.. ~ 'j,"i,believe that other design, features,.such as the roi vorth .f n ys i n E.d ? ^ c minimizer computer or the rod dropout velocity limiter, should n.g ~,3 4 6 ' h>f ^ be incorporated into the plant desi n. 6 u-. l 'We Y '/

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,.N [ s.,;s c 9?p. VI. Maximum Credible Accident Evaluation g.. { p.g For the purpose of evaluating the adequacy of the proposed ' 1 .Rg . containment concept, the applicant has hypothesized a major' 7 i lJ ' accident involving a substantial release of radioactive fission g o.< y,. s . n..., x m. productaifrom the reactor fuel, and has estimated the consequences ,? , F;"

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{ MG y '.of this accident in' terms of potential radiation exposure to the 7 [[# -[ M L public, taking into consideration moderating effects.on such e.m... lg'i exposures of.the containment system and the environmental

• 2.

U],' characteristics of the site. In our opinion the maximum credible n ] accident postulated by the applicant represents the upper limit [54 0f Potentia 1' accidents that could occur incident-to operation-3 M .h # "$h of this reactor. Themabaumcredibleaccidentchosenbythe t d(.;,x.u ;s. W w'. q,;.:. w s re.vt S. v. u - ~y .q) M sm,,.@.;. i gnyW ti wcp;3.tapplicant results from an instantaneous complete rupture of. mw , i.4 s l <.: 9 (~,', ~ ~.,.. one primary coolant line inside of the dry well after reactor

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y operation at rated power for an extended period of time when j

+ .j the fission product inventroy is at a maximum. The pipe rupture i k.' would release the pressure in the reactor system (assumed to G' i,T be at 1250 psig), resulting in all of the reactor coolant system -gt n, j D water flashing'to steam. An immediate buildup of steam pressure R. h,.a 3 .N. . c. r.,..... + i .ype f.ij ,m ' (, t in the drywell to about 62 psig would ensue and press' re would 1 u .:..nw

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, ;,,vm i m*. - e increase in the suppression chamber to about 35 psig. The

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e a .4.. e, W ii, y 7 pressure in the suppression chamber would be reduced within ) 3 v.5.m .s J VW,. 4/ 1 4 ,,7 , E a few minutes due to steam condensation in the water contained + 3.pe., 3., xa ,. m 4 1 w . t. s .4 <?B;k_.....v y within the suppression pool. i%* ~~ l '%'e.. a.y*:y,. Ji ~ Other assumptions concerning the magnitude of'the accident and ,92 7, < m.u A. f). the effectiveness of the engineered safeguards systems for cc: 3's %- - sa .t i s gff v. .+. . &i k a'"

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' g j.,, ., y ., [. 4 *. f. 4 e. ,{ f.,,.g v.:.*', ga ', >1, N.e... alleviating the severity of the consequences are as followst up h,. 1. The loss of coolant from the reactor would result in L Jw a, m: M-..g'i .. melting of one-half of the reactor fuel and damage ~ j p. . g ;g, < m...,. '3. - z:.x,+;J nto the cladding in the remainder of the fuel. (The u Udv;; -o W. 1,1, c 1.. ,f , applicant will provide an emergency core spray system //% .W4 s.u ,,.i_l.. for preventing such significant damse to the core under' ,1 .,1 :)h, these severe conditions, however, for the purpose of this , ".}e analyses, it is assumed to be only 507. effective). s i .t 2. The fission product release from the core to the dry well t. Yd U. .' ?. ..' and suppression shamber was assumed as followst - 1 n:

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+ w...- d'*j , f~. _.,. /.%,.2 FISSION PRODUCT RELEASE FROM REACTOR CORE (Percent) tiht n 1,b O..~.. .- q m-~ c.;;.. c4 w@c -;y%g m.j.e.w.;3p:6; From Damaged From Helted . Total Core \\ l r .m y s.}*lp/ . Fuel Cladding Portion Inventory Release! m "i Noble Gases 20 100 60 .l Halogens 20 '100 60 ) Volatile Solids / 0 50 25 .c Other Solids 0 1 0.5 t i.:p 3. One-half of the halogens would be removed in the dry well

s. p.1 x

s. 4% M and suppression pool by plate out and scrubbing action of the wp re:.~,: . y r-s.. ,.,. a water. The remainder of the halogens and all of the noble ,, i WMS.. m.e.3,.i, y -..;(~ gases would be "available" for leakage to the reactor building g,m, . :e.i m. f. f., s h %,u %x.y,. -h:.; . bap #gM5 at a rate of 0.5 percent per day at design pressure. As .y e.: ..n m' -w,,. ip$ j'.t,j @.t. 4noted previously, the applicant will be required to e .py ,.tP+-

M;yp N 4.* -

' demonstrate the' assumed integrity of the dry well and iI %.t },1 a n.v p +u:,. y. a;., f{ ek . J, ,e ,s .4.s,- $)J -.0.e suppression chamber prior to reactor operation by suitable Wy: 'Q '~ &;M4 9 c-leak rate tests. ,. 4. -.+b. % t 9'Ir '

• )

s j"' \\; Q 7* s e 'g. Q 3. v .. h [ .waM .. 9'i .

• W N.. N

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a., -

j s. x y :.~.m. u .Q.p wq -wwmen~- cr -- -m; 4mw.n.m;, ; ;g, n.y.....;,r,g.. _. -,,n,$ g,q g..pg4,..,49g,w _, 4, y.ny.. +.u.m;;,.;.y. e w m r %p .'.g.w : %.;z -' ;q, y, .q wq~~ f ' y .:q y.:, - p y'qQ'jpppp9 ~7" s, ,, 3; !.,.gr,

,., j 4,.g.

3 ,y,. , n..~ } ' m.,-

a..

.Jg ~l 52 - J di l' 4. The reactor building air which is maintained at a slight L;?p. j if 1 i .. g. negative-pressure (k" of water) would be exhausted to the i v.- e. 1 7 g e -Wjm.* atmosphere through filters for removal of particulate s.e;; ,.m. e-v , :., o;,.c re,,y; %%..:, d halogens, at a volume flow rate equivalent to 100 percent ..e ., p,4v 1,; '. an . p. ge.. of the reactor building air volume in 24 hours. The parti- .$9 . /@,. 'l 3 u w . culate and iodine removal filters are assumed to be 95 percent ,"qw 9. g,,,. ' effective. As has been stated, provisions will be made in 1 ll 1 .. the design for verification of the effectiveness of these 1 1 .j filters on a periodic basis. [ The applicant's evaluation of the consequences of this accident ce ' ~ 1 fl assumed release to the atmosphere through a stack 300 feet high, .m$..% . m:., n- ..n .4.; 4. ~.a$p$although t the present time no specifications have been proposed E. t m. ..f c f N the stack design, i

t

..g , Assuming that the wind direction and velocity were constant 1 during the course of the accident, the applicant calculated ,c, ,f, exposures as follows ..,a ,r y ,]o 1. For good meteorological diffusion (lapse) conditions and q. m.: . a wind speedof 10 miles per. hour the maximum concentration V 't' Ma.4 m e.W %;~u -;mp ,a 1.e,..,2 s c ' M.1 'l a4 .of radioactivity in'the air at ground level and the maximum ,.i %.3-q exposure rate would occur at a distance of approximately T. '3. ' y.) - 9., v. e,1. ~ 0.6 miles from t.he stack. ~ .9. -. e .v ~--' dyi n.j 2 .a. .s..,.. m .%@w i.1

- D.gega

.. a. The maximum potential dose rate to the thyroid was g ...,h,4 /. 8 .T-l, _ 1 d,. N... *.% ' N, 'G2'p,".%. U%jfhalculated by the applicant to be approx;j

• %.44 tq, y

3.B ;{gj .,,....%,,g,, m. v.e~y, e,y . r;: v -;. m' .' J - 8 millirems per hour (0.008 rems per hour) and the'

g. ffsj a

L;, h 5].,' _ total potential dose for the duration of the release r ({ j{f. .....'k. . y., ' y.,,,

• t%.

_. c. 1 9'v .,. 3:) %'9 7.xs;: j !%%&M9t#@m g$Ms.4s%%.wwm.wa.-a..:;,nwsind ..:..n. s.u.

f('*~rninam~x=.m*-r? wemm7~;ygm,ing..u.;,.y..e.q.. gag ppgg g4. ~ Q

. m 'p

gpMQL.,

m' y. ,~

. ;.z.

j -6 J p.M p.g M $~@., p .h y~ e-w y..z. A. g.e -y. 53 - Oi .,~ ,ar lbG y'w to the atmosphere was estimated to be approxi-w.w w.gh [, mately 1.5 rems. g

• 3 x..

! /.'

gd, . g*p.F~ . m 4.i,..._ b. The mar.imum potential whole body dose rate due to ' .,, : v.x u': a arv 'w us ~ ~.. w s 2 k ,.3, 3 .w -T millirem per hour with a total potential dose for the 'ip '.vy a9 duration of the accfdent approximately 0.024 rem. f w.3... For moderate inversion conditions with a wind speed of i 2. 5 miles per hour the applicant estimated the' maximum 4' concentration in the air at ground level would occur ..s.. t .Q,. ~~ .., ; q ?e'l l? f

s. :approximately 3 miles / rom the site. Under these conditions f

, x. b ;;x. 9 :. M.s t ~' .. e.. - C y,. ,,. ': 4 y 7', c \\ -es 7.- w.. %. pc.,bw.the maximum potential' dose rate to the thyroid and total p m. g,t c.o .m-yc..;,, g. ,p.. ..., m.;. s, m .rL .es ',3c . w e '. dose for the duration of release was estimated to be m 1ess than 40 millirems per hour (0.04 rems per hour) and J f.: .[ 7 rems, respectively. The whole body potential dose rate e ~ s)s 4 M-and integrated dose were estimated to be less than 10

j millirems per hour (0.01 rems per hour) and one rem,

.,n.

n? j l

3

.C L

respectively. (.)m

'5 d1

~. 4 .,.._.1.c.,.,W;,,.nJAs previously stated in the section of this report describing kp'W.nm'gi;c. W r.w:- ~ ~ ' s ,,, ~ x.M .-ap. kd..w[M.s ,,. h,, 4,..the meteorology of the site the staff feels that wind speeds q u gif f:t sps:E.during inversion conditions may be somewhat lower than assumed by , w g;,... , m. w,.- e, aj g ...a.. $.E. M.5 . ' 7. ns@ the applicant in this evaluation. In this regard, the staff has l t. O i "MO ' S' 4 .,9 J Q iO';V.%.. 4.$0 }Q2 M':'fGj... made calculations based 'up'on the above assumptions which tak q -.;q .I 'f~'~=' r n.v.a d /., _ . - n..., w, m:m. a..;p g.f '. w - l i ' q@#.i%..t.4

.. g 4

< N';'into account the possibility for the accident to ' occur at a wind..,...' a .m - ... %.s x :.v. e*.

c. ;

• i

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" speed of one meter per second (2.2 miles per hour) under either

.i d. s. p . ~. Wu,e k 4 A, W.n .s : .-~

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• .k'dC '
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- h4-a j '. ! MI,. Using an effective stack '("T', -M lapse or inversion conditions. ,M;fh .. j height of 300' feet and 'ttie' one meter per second wind speed e$ 3 .,. # e l f

.;i..., -..,..,..,-pa.{gthe staff estimates that the maximum potential whole body and 2$.,. ;-

sy. 3. thyroid dosages for. the duration of the accident are 0.94 rem 1 ' f;., ~ and 53 rem, respectively. 1 .l$M "r7 The staff has considered both the applicant's assumptions { i Y. concerning the postulated maximum credible accident and the _{ ~ i general concept of the safety features proposed for reducing g a M '..c ; to a minimum the consequences of such an accident. As indicated W[!.,i; .G l> ~ ~ - 3 ( J ,~: .;[ ([f P . fin the above' analysis, the amount of fission product released ll:s , 'j ^ y.f 4O. WMfrom the fuel depends'en the ~ extent of core damage, which in ,. t, 6m ~. n +?4.V y a+ W X ;.:7 .. w '~ ~.., [ 3,,.Qg, +.y g. , +.

2. L

- turn, depends on the effectiveness of the emergency core spray. .,.4 The staff believes that a suitable core spray design which would i provideforanadequatesupp[yofemergencycoolingwaterwould substantially reduce the extent of damage to the fuel, even x,J to the point of preventing any melting. On the other hand, if My y'. A ~ g @pg - -,g a. ,9 , ;.the emergency core spray failed to function at all the inventory

h @

, i l of fission products released'from the core would be increased by ,.,,a... y, ,r..... ..@ w. ',) .., r3 'C ~.) ' approximately a factor of two. In this case the dosage values c .?N k ~~ wp Q .would be increased to approximately 2 rem wholc body and to 100 9 f.yj g. y.. p$ R

, A'. . s rem to the thyroid.

cy y

• 9' '

m..g,.q.r.o q. From th'e above analyses the staff has concluded that the engineered t

.y; Idi j safety features proposed for this facility should be espable of Q.y -lr

=. ' 'h / ' Ij limiting the degree of harm,,that could result if an accident ! H.a M <(y ij such as postulated should occur. Therefore, since it is believed O. 7 gQ's. ;.*

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l - g cca,.m m ~swww.cfdwr d&Ms.WMe-@han@*w2W/AN&S*W -bn M, w, :%.y $:&eJa%;.y;. M.@4 - y.r. Y,a&yQ*; .s.g.>4.. o. A.. w. nn -. ml*;.: m .y;n W %.2, M. w.a 4.V t +,g,. Ay.. >.; m x~ n L 4 s. w s a e-n & s a.. u-

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e-a ~.w t..p..f.. 4y 54a - J m,. s ....s.. t e. c... e c. Ry I-that the occurrence of such an accident is highly unlikely, the

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,-staff concludes'that operation of the proposed reactor at %yj j %.. qu ,,,s.,,,the Bodega site yould not represent an undue risk to the public. n 1 . l I ^ M.~. g(JR,, .ii [.;i g,s &:: % k.pjp ~ ~n P %. .w i w~. 9.y.n J,l. k 'f '{ < g' 1., +s i* +

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$5 " ~Nsy&.$t*dMp % m 'W M fa.'1:'T =' " A' ' gg ,o ss - ,m .dX gy.g: i .3 VII. Technical Qualifications of Atulicant '. 6,"- ) l . 2..\\$[ he technical qualifications of PG&E are described in the .J + p. an"/y M' 6.y-application for a construction permit. PG&E has constructed and gg m l L z is now operating a boiling water nuclear power plant'at Humboldt

j, g4

. s..a. 2., ; ,.L Q g,..3 .4%.--.gpW.: Bay near Eureka, California, heir principal contractor for the g;, ..e n a ry. .--. - -- ~ , _ Mega construction, the General El.ectric Company) designed and -h:&.? d

• ^

furnished the major components of the Humboldt nuclear steam supply m'.. ,-2. system, incimmg the reactor with its controls and instrunenta- ...wx j ;Ep ' tion.. GE has also designed and furnished similar equipment for several other boiling water reactors in this country and abroad, t. c. (Special remarks on gm141 cations necessa:7 for seismic %@,yn . J; .s,. ~? 'ami coa,&sh%s ' [- A . ~. x.~, e mo. a.c a...

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..',. 1 t. VIII.' Report of ACRS .e -q73 he Comission's Advisory Comittee on Beactor Safeguards -ws h. '(ACES) has reviewed'the site and pre 14m4m design of the pro- 'Vi$$h, 7;$,i .c

posed Bodega plant and reported thereon in.a letter dated

.q_ g,y.f.. y g v. s s. %g p.3 ' ? April 18,1963 to the Chairnan of the Comission. A copy ;of f4g; U.1 .m ~ .m i " this report is attached hereto. 4 N:l7 'Fi.p. ~At its forth-seventh meetihg, April n-13, 1963, the ACRS ut!,,.. ' c a , %."G C., discussed the 3'acility, with representatives of the Pacific Gas and Electric Company, the General Electric Company, the U. S. ,y l Bureau of Commercial Fisheries, the U. S. Geological Survey, ..,2. '.?.,. . M a.nd the AEC regulatory, staff. Se Comittee concluded, as ~ , '.N.. - - n.,. 3, ;'%. : .? qt m1-: 4. M; I;g :.. ~lb $ [; indicated in its letter that: S/ m me, j n - .r + .,a . L.~ v; <..,,q ggw..,.;. ., 3: ug7 v: G. f.j g y. he ACRS believes that, sub$ect to the above conditions, ~ .v.n d j M' the proposed reactor can be designed and built at the r.3; ~ r, ~ ~ proposed location,on Bodega Head with reasonable assurance w j .~ i n" that it can be operated without undue hazard to the health y.:j j}:

Q and safety of the public."

La M k:h 39 .e... Se staff agrees with and has taken into account the w;..r.g. .u. . r. 3,. 1'8c .$p# t:97 ..,h S /:H.4 comments contained in the letter suggesting that further con-M J 4.. e,% w e y y%a;. s 1 9 %s, ?J i "? F I

w. sideration be given to several points involving exploration, ri a

~

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+. ;

Wd e,8 ;L ~..,. ~..L lvd,esign, or testing. Each of these matters has been discussed A.. m y i 5. g gy,, j... $ 3 ,M;p.: T,in some detail in the appropriate section of this Hazards , :,,... :ds.,.h c. mc yn y., w :., .m .. < p .s*ps e 4.g'a!- ,[,. ", '{h;t.*j d.h

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he# ,/ ...A. ' ' l. ll t 9 n: Introduction

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In accordance with;the. Atomic Energy Act of 1954, as amended, c i it is unlawful for anyone to acquire or use a reactor without . i.&..' , obtaining.a license'fromithe Atomic Energy Commission. The~ applicable! R ~

p...../

j;. u 4 j,.] procedures,and regulations pertinent to the issuancie of licenses $1 ./ i - @$s to construct or opera'te nuclear reactors N e contained in 10 CFR,. [. M .,f =y 2 . +.. .g ,]@ Parts 2g 20, 50'and;100 The first time..that.the Division'of j t3 ge. ..7 Reactor, Licensing (published findings that a site was not suitab / (... .__ _ g y h' for loca$ing a proposed nuclear. power plant was Pacific Gas & l', T g-y, Electric Company a proposed Bodega Bay. Nuclear Power Plant site. 9 ,,} Oil Discussion My . '.. ~ : ]M),- On December 28, 1962, the Pacific Gas & Electric Company..,c. ;j 7 ~ q,,, ',j filed an application tonconstruct and operate a nuclear pow'er C- .m.. c.- f g u

plant at Bodega Bay, California.. The proposed reactor was to 4..

be a 1008 MW(t) boiling water reactor. In the course of the E , g.. Division of Reactor Licensing review nine amendments to the t . h application were f'iled. In addition to the technical evaluation i J y,! ~ within.the Division of Reactor Ilicensing, various consultants \\i ..y.. ~ $W Mg aissisted the AEC staff in specialized areas. The application was WM! :. s. e '.S '. gj 'also considered by._the Advisory Committee on Reactor. Safeguards ,x..:;, .(ACRS)- aind its views o'n., the proposed Bodega Bay Nuclear Power t L.Q.i N Q ...'.... Plant were expressed in letters to the Chairman of the AEC M . dated April 18, 1963, and Ocitober 20, 1964, i y.:N }.h In the evaluation ofthe proposed nuclear power plant'for .yjg h e.+r 'e g; Bodega Bay, a number of site and design features 1. d 9 m .f:y ' yM, .k s ,a e 9 'D 3 r, ..fc e.+ .A t'

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i ; r.yn-mg.~. iscra w w w c [ ..s p y. y i .r, 2-important safety implications were considered. Among these d J. ;. .were the suitability of the nuclear reacto'- design, fuel handling i.f facilities, radioactive waste disposal systems, emergency safety y) features,, containment design,.and the site related aspects. ., l. a t~ '; Early in the review it was recognized that the proximity of the o %d f iMj proposed Bodega Bay nuclear power plant to the San Andreas fault %,, v., h r.

(y, zone would cause serious reservations, particularly if fauli:ing

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were found at the site. In discussions between the ACRS-and the ~ .a. w. Division of Reactor Licensing staff in April of 1963, the staff recognized that acceptance of the site would imply agreement that

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.g4: the likelihood of the occurrence of an earthquake which would ..e M3 cause faulting under the structures was acceptably small. The $[Q applicant, Pacific Gas & Electric Company, claimed that, slippage < G.L = .v

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is more likely to occur in existing faults and that the develop-j ,]' jg ment of new faults would not-be expected to occur. The AEC ,.g AM staff did not have a basis for disagreeing with the applicant's i \\1 assertions as to the existence or likelihood of development of s K s 3 faults; however, the staff believed that further information l',

a;,3 WM gig should be developed during excavation on the possibility that Lyy M

faults existed under the site. ~~ 4.+$g Q;j The ACRS in its letter of April 18, 1963, also expressed ,f g / concern on locating reactors on faults. The ACRS letter in @g s7@. part read: j! ,1 s? h$? '/l'. s .A7 %.h N /Nf /- t

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"h "The: requirements that are imposed on plant design- .l because of location in an active seismic area have been. considered by the applicant, and the referenced documents contain the recommendations of seismologists y. J. who have been consulted on this question. Tentative Q M: . exploration indicates that the reactor and turbine buildings will.not be-located on an active fault.line..

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The Committee believes that if this. point is established, .? i the design criteria for, the plant are. adequate-from the 6h standpoint of hazards associated with earthquakes. M Careful examination of' the quartz-diorite rock below M MfE '. should be made during building excavation, to confirm 'this point. Furthermore, the Committee suggests that,. during design, careful-attention shoul,d be.given to W the' ability of emergency shutdown syst' ems to operate 1l. .. properly during and subs'equent to violent earth. shocks, and to the stress effects that might be introduced be-cause the reactor building and the turbine building 1 2 are to be anchored in different geological formations. The need for earthquake-induced shutdown and isolation of the primary system can be considered at a later time." ii The'AEC staff continued its evaluation.of the seismological ey an g".: and geological characteristics particularly as related to the \\ "3 c y existence of a shaft fault in the. underlying rock. Assisting l g [ the AEC staff in the evaluation were the U. S. Geological Survey and the U. S. Coast & Geodetic Survey with regard to the .w

• b and geology / seismology'of the Bodega Bay Site.

Dr. Nathan Newmark 'of the University of Illinois, and Er. Robert Williamson of 0 i 4# Holmes & Narver, Inc. assisted in the review of the structural xA a wh .t aspects. The advice-received from our geologic and seismology M $,f4 consultants indicated that the reactor and its containment Ag structures be designed to withstand a ground response spectrum j$ 1of2/3g,wkthpeakaccelerationsup'to1.0g. Further, the y$E possibility of differential shear ground motion up to 3 feet h existed in the event of a severe earthquake in the area. . b. 19.e jV' f?d a -.~. ~.:....:,= ~. . ij

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... +.. .s. < '-y ewugn.mm %.na m ., +. ' &,en s '/. ,9e y g>" ~f* ? o-I 4-4 o The applicant, while disagreeing.with the credibility of such an extreme earthquake, nevertheless proposed a design which it 4 j

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believed would. safely ride through a 2/3 g earthquake with* peak i cfm l[hf;i l acceleratibas up to 1.0 g accompanied by differential shear dis-Q(; i . / 49 placements up to 3 feet either horizontal or vertical. The AC35 .MG i' .]'jg reached the conclusion.that the engineering principles and general I;y@.. j design proposed were sound,. The AEC sta$f considered all of the .i ^ information available and the views of the consulthLnts and ACRS i on this; matter and.did not lightly reach an opposite conclusion. The ABC. staff recognized that in this type of technical Judgement, reasonable men may differ,. and it..was our judgement-c,. At that accepting the conservative ~ recommendat' ions raised Substantial' W ' f* safety questions'o'n the reliance of unproven design measured. 7, The AEC staff, therefore, concluded that Bodega Bay was not a suitable location for-the proposed nuclear power plant. a., \\\\\\ j

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,.ia.,..a..,..+ W.c.ma @;a..,,,.<. y' - w. w u #.4 *f h =s h, e' n g : e W iw w- ...,..,. : :,... n i..u. H.,..... a ;..:,.: A @ ' g.,. g; /T A.4,,", .s, cn ~,. n t ., 4 r :: n .n.%h. j ..,w,. .1 e. 9.y. 1 I-n, c 'e,; ~. ./ <~ CHRONOLOGY 1. ' December 28, 1962 Application filed for Bodega Bay.- Nuclear' Power Plant. n 9@

2. - February 26, 1963 Letter with list of questions trans-cftlp

.mitted to PGEE. .a. S} 3.. March.4, 1963 Amendment No. 1 filed by PG&E. . h.J $ad 4. April 5, 1963 ,'- Amendment No. 2 filed by PGhE. 5. -May'24, 1963 Letter requesting information from PG&E.1 6. June 13,1963 Amendment No.'S fileM by PGhE. ') a \\ ~~ 7. August 9,:1963 ~ ' Amendment No. 4 filed by PG&E. ' 8. January 22,.1964 Amendment No. 5 filed by PG&E. ? 9. March 16, 1964 Amendment No. 6 filed by PG&B. %, i y - 10. March 31,'1964 . Amendment No. 7 filed by PG&E. ,w ~ h -11. July 8, 1964 ,- ' Letter with list of. question's trans-n. ~ mitted to PGER. 12. July 20, 1964 ' Amendannt No. 8 filed by PG&E. 4f 13. August 26, 1964 ~ mitted to PGhE. 2 Letter with list of questions trans-Q; - ^ 14. September 16, 1964 - Amendment No. 9 filed by PG&E. 15. ~ October 26, 1964 Summary Analysis by the Division of f. Reactor Licensing released. m ~ ~.m. ' 16 _ Letter from PGh3 requested withdrawal h November 4, 1964. - f@4 -___._ of application. 0 d4 [ i t .L

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,-.,h j [.:.' ' : c,.; p' 3 G A = W " * ~ .u.n_=g..p.~--,, ta 4 a d h prs.asiped, b ards to the proposed plant from such a esissia p a twofoldt. (1) shaking of the greuet ano to seisate unse propuesties, ama (R) possible pamenneet at=ftr - --E et the ibammastian 1 rook due to faniting. The hasard hi to abaking is being tapestigates ty others, insladting the 9!W_w Divisten, & E. Coast and Goodetie Survey. Pmplicties of possible permanent at=h _1 most be benes largely esa the distributica sad characteristics of the surance f=el+ tag preenced by the 1905 sarthquake and to a lesser ecsent en the distribution ( i of faults in the. sacavatica for the rwacter and ta the entire baiega j 1 I!ehd tree,, 'The evidecce is cat adquate to h A more than a general statcuent of probabilities. i b site tu approximately 1,000 foot west at the west edge of the eetive Een Andross felt some, whiah is aggeesimately 11/2 miles wide 1 s f hora,, N semia surface rupture in this wisimity Aartas the 1906 marthquaks tock place ar. the east side of the ecr.e and bed a horizontti dispt- -t = - -. of 10-00 rw t. !?hrt;m6 out Bo&rga rienA, tectonic fadits sad,foints are h i comen in the granitio rock; h most prominent enes tned morthwert, northeast, and east. At the site, a principal structure is the Shaft 3.. Imelt, ammed frees its exposurse in the shaft exeersted far the roaster. This faalt, one et many %eetente taalte in the'granitie rests, is the only one that has been tremot douusera from the surtees throesh Pleistocene y se4 mmats into the wJeelying granitic reeks. It strikas E. kO* & em4 bas boom traced on the surface a total of about 230 fast. 1 i } e. r l l l e e 3 '. q. , i '~~

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