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.I DRAFT 2 e Date August 5, 1963 I ^ '). . .
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HAZARDS ANALYSIS hM. Ag f a by .th._ e ..
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M, < . -DIVISION OF LICENSING AND REGULATION hy% ,.
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, j PACIFIC GAS AND ELECTRIC COMPANY a
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'.j BOIEGA BAY ATOMIC PARK
. J. m IEtIT NUMBER 1
-' CONSTRUCTION PERMIT Jt DOCKET'50-205
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! TABLE OF CONTENTS
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, j W I. Introduction d'ipj :
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II. Background
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% s.y A. Site and Environmental Factors ~i
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- : 1. Plant Location
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- 3. Marine Environment ,
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f B. Containment Design f %
?"i C. Nuclear Design Features L ,
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ty;% 1. Reactor Core and Fuel Elements W}
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- 2. Reactor Control ( ,
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- W 3. Control and Safety Instamentation
- k. Primary Coolant System
- 3. - D. Emergency and Safety System Design s . ,.
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'- 1. Power Supply q
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E. Radiatica Monitori5k Systems Design
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? ' F. Waste , Treatment. Storage and Discharge Design Features -
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,3jj 1. Radioactive Liquid Wastes -
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-j -3. Radioactive Gaseous Wastes
".19 IV. Seismic Considerations
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Epj; Y. Research and Development Program
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% VI. Analyses of Potential Accidents
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, .VII. Maximum Credible Accident Evaluation
. VIII. Technical' Qualificatidas
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,[ IX. Reports of Advisory Committee en. Reactor Safeguards !
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ct - I. Summary and Conclusions 'j
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., se Introduction 3- I. 'I io
<,s Me Pacific Cas 4 Electric Company (PG4E) has proposed to construct and
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- operate a nuclear power plant on Bodega Head in' Sonoma County' , . California.
fhk PG4E will design and supervise construction of the unit, and the General e
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Electric Company (GE) will furnish the nuclear steam supply system and the sd
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'f .W The proposed plant, designated by PG4E as Bodega Bay Atomic Park Unit Number 1, will produce nuclear energy at the rate of 1,008 megawatts (4t).
De gross electrical generating capacity will be approximately 325 Mr.
he Bodega plant is similar in many respects to boiling water power j- reactors now in operation. In general, its detailed design will be based x, a .
M: ,ij en operational experience with the Vallecitos ' Boiling Water Reactor and the WM
$$ Dresden, Consumers, and Humboldt Bay reactors. H ere are some features of J.W.
the plant which require research or developmental effort in order to provide engineering information necessary for their detailed design. nose features
) are discussed in Section IV of this report.
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c Except for a discussion of the effects of seismological and geological factors and plant safety, this report contains an evaluation of the significant hy features of the. site and environment which have a bearing on public health and safety, and the significant features of the proposed facility design which j
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to the general public. He effects on public safety of normal routine opera- .,
x k+Q.iMZ i tion of the plant, including the discharge of radioactive materials are also considered in this report.
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3 x II. Background -;
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hn M.T On December 28,1962, PG4E submitted an application ~ to the AEC for a
,$:; construction pemit and operating license pursuant to Title 10, Chapter 1, 6
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Code of Federal Regulations, Part 50 (10 CFR 50). We application, which 1 includes a " Preliminary Hazards Summary Report", dated December 28, 1962, and Amenchments 1, 2 and 3 to the application dated March b. April 5, and Juno 131
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d : respectively, .has been reviewed by the staff of .the Division of Licensing and 15W Mn Regulation. Technical consultants in ' specialized '
areas also advised the AEC
Q.. regulatory staff. He application has also been considered by the AEC's yk '
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- ,]p Advisory Committee on Reactor Safeguards (ACRS), as required by the Atomic ja Energy Act and the regulations of the AEC. De recommendations of the ACRS, 1 . as expressed in its report of April 18,1963 (# copy of which is attached "1 .
. I j hereto as - Appendix C) were also considered in the regulatory staff's evaluation. {
d As is customarily the case in reactor facilities prior to the commencement I
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of construction', there are a number of features of plant design and operation e,-
17 which have not, as yet, been definitely resolved. De Commission's regulations j w; ~ #j provide for the . issuance of a construction permit, on a provisional basis in cases such as this, in which aspects of the detailed design have not been I
completed. A provisional construction permit may be issued, according to 1
Section 50.35, 10 CFR,on the basis of findings, among others, that (1) the J
- i applicant has described the proposed design of the facility, including, but j
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not limited to, the principal architectural and engineering criteria for f y,
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,Q the design, and has identified the major features or components on which ;
q further technical information is required; (2) the omitted technical info-dE mation will be supplied; (3) the applicant has proposed, and there will be ej '
'M i?g conducted, a research and development program reasonably designed to resolve
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the safety questions, if any, with respect to those features or components ffh rh .
which' require research and development; and that (4) on the basis of the 4i foregoing, there is reasonable assurance that (i) such safety questions will be satisfactorily resolved at or before the latest date stated in the
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application for completion of constwetion of the proposed facility and (ii)
,j , taking into consideration the site criteria contained in Part 100, the proposed j
1-facility can be' constructed and operated at the proposed location without undue
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risk to the health and safety of the public.
4 1, Be proposed constmetion permit, if granted, would authorize constme-
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@J tion only. He Commaission would require timely reports from PG4E with respect
[ to results of research and development and final design of the more significant 1 , design . features. he AEC staff would continue its evaluation of the safety f
} 4 of the plant in light of' this information. An operating license would not l- be issued until the final design had been completed and evaluated by the AEC
- staff and the ACRS. In addition, the plans and proceduns for operations 4-would be evaluated by these two groups.
,[ ;, Pursuant to a Notice of Hearing published , the
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[7" issuance of a provisiona1' construction permit to PG4E will be cansidered at l
a public hearing to be held in the hearing room of the Board of Supervisors
, of Sonoma County, Santa Rosa, California, at 10:00 a.m., PDT, on l
1963 before an Atomic Licensing and Safety Board' appointed by the AEC. He
, l issues to be considered at the hearing are:
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- .J 1. Whether the applicant has submitted sufficient information to provide
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y reasonable assurance that a facility of the general type proposedTi
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iy the application can be constructed and operated at the proposed Iccativ. wikat undue risk to the health and safety of the public; a,
.2. Whether there is reasonable assurance that the technical information W,1 '
y@m omitted from and required to complete the application will be supplied; g
- 3. ' . Whether the applicant is technically qualified to design and construct
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- the proposed facility; and
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, y, 4. Whether the issuance of an authorization for the construction of the m facility will be inimical to the common defense and security or to the
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. . health and safety of the public.
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- I ne proposed plant to be constructed on Bodega Head would be subject to 1 a.
potentially severe shocks from earthquakes'. There is also a possibility that 1
earthquakes in the vicinity might cause faulting beneath the plant which I
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could cause severe damage to the facility. The possible effects of such 4'O
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. seismic activity on the proposed plant are still under study by the Commission's s.u n regulatory staff. He staff has not yet detemined whether or not a plant can M
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be constructed and operated safely at this location due to these considerations.
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At this time further information which must be based on exploration of the site, among other things, must be obtained. Further consideration must also be given to criteria which must be applied in the design of systems or com- '
a ponents of the facility which are of importance to safety, especially those dQ systems which must be relied upon in an emergency, such as a severe earthquake, 3y[u
.h ye to prevent undue hazard to public health and safety.
5,4 .-
This analysis and the conclusions contained in this report, therefore,
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'., are made without regard to the special safety considerations which must also 1
l be taken into account in view of the seismicity of the proposed site, u,(
- 4 The staff's evaluation of the proposed Bodega nuclear power plant described in subsequent sections of this report and its position on the issues
- >g at the forthcoming hearing are based on all the technical information submitted
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j$Ik , as part of the applicant's request for a construction permit and the report 9.b ~ 4
'.Y from the ACRS. All of this information is available for inspection and review f
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{ at the Commission's Public Document Room in Washington, D. C., and at the Sck - Commission's San Francisco Operations Office, 2111 Bancroft Way, Berkeley, ,
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?/@] California. His evaluation and proposed recommendation is subject to j e, ,'. . .
1 Q , f, modification in the light of 'any further information which may become avail-3) able, including the evidence introduced at the hearing. The decision of M/ the Commission will be based upon the entire record in the proceeding.
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, lhe Bodega reactor is a direct cycle, forced cird:Istion boiling water reactor with internal steam separation. Nuclear ~ energy generated in the reactor
, - at the rate of 1,008 megawatts will be transferred to the water coolant which '
.'[. is circ'ulated through the reactor. Steam generated in the reactor at 1,075
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pais flows to a turbine generator with a gross electrical generating capacity of about 325 megawatts. Reactor coolant which has been separated from the steem is recirculated through four loops each containing a pump rated at 29,000 gym. After passing through the turbine the steam is condensed, and the condensate after demineralization is returned to the reactor vessel. 1his j
water, which will contain some radioactive materials, will be circulated within a closed system from which the only normal effluent will be a continuous y
y discharge of noncondensible gases. This. gaseous material will be monitored 3
Y continuously and released from the reactor stack,1f the contained radio-
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activity is below permissible limits. As would be the case in a conventional t -
- power plant, the condenser will be cooled by wster drawn from Bodega Bay and discharged into the Pacific Ocean. From time to time, regulated and measured quantities of radioactive liquids will be mixed with the condenser coolant
! water from this facility and discharged to the ocean.
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y;q An overall judgment concerning the safety of operation of the Bodega
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,l reactor or the acceptability of potential hazards must be based upon a i-0 musser of individual safety considerations. Our judgments at this time are
.Ml' based upon an evaluation of the design details, design criteria and design
,w 7,h concepts described in the PG4E application, the known facts concerning the
- proposed site and its environment, and an analysis of the effects on public
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p health and safety of normal operations and of potential accidents during these operations.
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. As noted previously, detailed design of a number of features of the facility is not yet completed. For those features, the present evaluation by the Com-
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mission staff is based upon the principles of design rather than upon details hh
,:n of design themselves. . Ir. the case of features in.this category which are 'of particular importance to safety, the staff proposes to require and expects to v' m'
~jf. teceive information on final design of these features before PG4E has expended
-[ any substantial amount of effort in the construction of those features.
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The general olijective of nuclear safety is to prevent or limit to an
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acceptable degree the exposure of persons to ionizing radiatien. It is proposed 1 that the following general features be provided to achieve this objective:
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i (1) Re first and principal safeguard is found in the design features of the plant which contributes to the confinement of all radio-
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active materials to their proper place. Confinement of fission '
f#~ ' products and the products of' neutron activation is attained by the 3
, three-fold containment afforded by the fuel elements, the primary
- coolant system, and the containment building. The isolation of i
the plant site provides a fouirth" measure of protection of the public !
against the radiation from materials routinely or accidentally J released from the plant. I '
.: The design of the fuel elements and provisions for reactor
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control and cooling have the purpose, from a safety standpoint, J of retaining fission products within the fuel elements, in which e
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location the shielding of their radiation is most easily effective.
$, ne primary coolant system,' consisting of various vessels
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and types, serves to retain any fission products which might be
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released from the fuel elements and any other radioactive materials formed in the course of operation. Condenser cooling water which is discharged continuously to the ocean is completely separated from f .
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i The reactor vessel, tho' recirculation pumps, piping, and other portions of the primary coolant system are located within a structure
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in.y which is capabis of safely containing the radioactive contents of the Dh .. . .
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., . 1 (2) Radioactive fluids and solid waste are contro11ed' and processed in a manner which confines these potentially hazardous materials to
, systems which are designed to permit their safe preparation for disposal. Paths through which radioactive materials are or might be discharged are . continuously monitored by radiation detectors.
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. es E e'A.:g Storage, processing, samplin'g , and monitoring are employed to 2M '
E.7 assure that quantities of radioactive materials released to the environment will be within limitations established by AEC regulations.
l (3) Shielding and area radiation monitoring are used to continually
. assure the safety of workers from radiological hazards. Such monitors also serve as an additional safeguard against the release i of radioactive materials to the public.
u;A g (4) Emergency systems are designed to prevent or reduce the hazards to
..n a.f.1 plant personnel and the general public should accidents occur.
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fhe following sections of this report discuss in greater detail the
$,N more important safety considerations which have led to the staff's conclusions
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- with respect to the . safety of operation of the proposed plant.
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- .W J A.
Site and Environmental Factors
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he site proposed for the Bodega Bay reactor is located on Bodega Head, a small peninsula along the Pacific Coast in Sonoma County, California,
.~j approximately 50 miles northwest of San Francisco. Bodega Head is bounded
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. on the east by Bode,ga Harbor, by Bode 2s. Bay on the south and the Pacific s ocean on the west. A sand spit known' as Doran Bea$h or Doran Park extends 1
j') from the mainland towards Bodega Head and. forms a natural breakwater for y Bodega Harbor. ,
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i s .The environmental factors which are significant with respect to the t kh.. .
"!l/ safety hr this site and which have been exaEned in detail includei (1) l l' the location with respect to the nearby population, (2) the meteorolog4bal I i
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. . factors, and (3) the marine environmental factors. As noted previously, I
i geological and seismological factors are significant to safety and will be
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', j4 the subject of further consideration by the regulatory staff.
- 1. Plant Location t
a he proposed location for the reactor plant is on a 225 acre tract u
V of land at Campbell Cove near the southern end of Bodega Head. The property j
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owned by. the applicant includes the entire southem end of Bodega Head. De adjacent property to the north is under acquisition by the University of l California for a research facility.
l i ne reactor would be located on the east side of Bodega Head near i Campbell Coveland across the entrance channel to Bodega Harbor from Doran Park. He nearest edge of Doran Park, whf ch is owned by Sonoma County and 1
contains no residences, lies approximately 1,300 feet east of the proposed
[,] reactor.
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. he traffic through~ the entrance channel to Bodega Harbor consists
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<gl5 primarily of commercial and sports fishing boats. Usage of both the Channel n - -
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- and Doran Park could be cc.atrolled under emergency conditions if this should
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- become necessary for protection of the public. Accordingly, the exclusion m
'?( distance for this site can be considered to be the distance to the northern 3
, site boundary, which is a minimum of approximately 2,700 feet (0.5 miles).
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d' Me population data for this area based on the 1960 census shows no popu-
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U lation conter larger than about 200 within '10 miles of the site, and none larger
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?- than about 3,000 within 20 miles of the site. He nearest cities of more than
"! U 10,000 are Santa Rosa (31,037) and Petaluma (14,035) which are 21 and 24 miles, Ah 7d respectively,from the site. De population density within 25 miles is as
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j (Gm v Persons / Square Mile Total Population i q'[.D
- Distance of Land Area in Area
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, 1)fr - 5 21 soo 5 -lo 16 1,600 {
1 to - 15 81 15.700
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16 0 , 15 - 20 97 3o,ooo ] i 2o - 25 180 .
66,700 he above tabulation shows that the proposed reactor site is favorably l l
located from a safety standpoint with respect to population distribution and density. Not only is the immediate area sparsely populat4d at the present time, but the location on a peninsula provides natural barriers against the future development of housing within at least two miles of the reactor plant.
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- . out to the distance of Petaluma and Santa Rosa, and these cities account
( for about 75 percent of the population in the annular area that lies between
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20 and 25 miles from the reactor site. Accordingly, the actual " low popu-g+.:; -
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- P -] lation distance" and the " population center distance" as defined in the l
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3; ) Commission's site criteria regulation for the Bodega site can be considered M il
$,:. to be 24 miles, which is the distance to Santa Rosa. I't is shown in a later )
i section of this report that the maximum required " low population" and j 5!
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, . y, *
.f'" " population center" distances are miles and miles, respectively.
- 2. Site Meteorology-' o 34 ff In general, climatology of the coastal area at Bodega Bay is typical q
- -a
.'r ] of the central to northern coastal area of California and is characterized by k . .
3 a wet season extending from about November,through March and a dry season
. .1
.g g frun about April.through October. ' he topography of the area inland from the- {
l qN ,y ' -
site is characterized by a series of hills and valleys, with the hills rising i
~;W a -\
i
,g(
to elevations varying from approximately 400 feet' to approximately.1000 feet.
,. ~
9 he roughness of this terrain would be expected to enhance the charac-(
".'li teristic atmospheric turbulence of this area of California and result in i 4 j
- rapid dilution of gaseous effluents that might be transported from the site fj to inland locations. On the other hand under strong inversion conditions, 7$j{ '
0M the range of hills along the coast, which is approximately three miles l(il from the proposed site, would tend to restrict the transport of airborne q
'4/ materials to the inland areas.
Detailed meteorological information for the proposed reactor location 3 is not presently available. However, observations of wind directions and l velocities of coastal locations at Point Reyes, approximately 22 miles south of Bodega Head, and Jenner, approximately 12 miles north of the Head, shows
.an.y
%f that the wind blows towards the coastline approximately 60 percent of the M
9 time, with the prevailing wind direction from the' northwest. His infomation
' id.4
,em l;
also indicates that the remainder of the time the wind blows either offshore e.C is4D
, up*4 or generally parallel to the coastline.
W.9 Based on the meteorological information presently available the applicant
[% 1
-r ,j has proposed the following diffusion parameters for use in Sutton equations Nj gg for estimating the amount of atmospheric dilution at the site under various
- 9. p T meteorological conditions:
, s' tW
~
3
>* I 9' .f f
- T
- .c
, i %. . '*,,. i4.. { }*
- j r ie p l , QWE3YEM?.M&WWhWEMi&QT2@M@*nsGM~M&MiTC'G3?NQ?M3?kiffN$55CW w
g,,,.;
<dA ,- .,;
a q}5, 4 ph" m -
s' : .
nui Meteorological 13Cendition
,m6 Strong e Moderate i"
e, Parameter , '
. - h. ,.' se -
.+- Inversie!fejj & Inversion' - tspse ;
PR v:l$','*y n .5 ;3 4 .22-k; Cy .2 j'.21 ,
.6
' 3 3 Cs
~
, 1 .02 '
~ ;07 .2 cx ,
[.jj$ A (miles per hr.) 5 5 10 h, '
n i1j ,
Our evaluation has disc 1'osed no unusual meteorological conditions at this aj
>j: i:
d*
- site and we believe that the parameters proposed by PG4E provide adequate .
v .
.. . conservatism for the estimaticra of atmospheric dilution at Bodega Head, except; with respect to the wind speeds' proposed for the inversion conditions. In
- g our opinion, a. wind speed of one meter per second, which corresponds .to about ,
. 2.2 miles per hour, would bhmow representative of this coastal location .
tun .
.~.~. .
g.g 'during periods of slow diffusion. Table I, Appendix III of the applicant's
[ Palininary Hazards Summary Report, which shows's relatively high frequency l of the lower wind speeds in the zero to three miles per hour range, tends to N corifirm our judgmen't in this respect. -
lhe meteorological factors, as discusse'd here and modified above, have t 3
.y . been.used in .our evaluation. of the potential consequences of the maximum, 4h ,.
Hjf credible accident, which is. discussed in Section VI of this report.
..g. . .e ~ . . s. . .
PG4E has constructed a meteorology station with a tower approximately
]t.. - l
, c}.h.- w 250 feet in height which will be used to collect meteorological data during
'e tt?'
-t y; reactor construction, such as the frequency of wind speeds and directions of :
~
various atmospheric and stabilizing conditions which' would'be appropriate for 4i
?. use at the proposed site. We believe that thess facilities should provide *l MJ
$$, sufficient data for determining prior to reactor operation the capacity cif p: 1 E, the atmosphere to safely dilute radioactive gases that might be released i
\
nf w from the facility.
vh _bb__. M9 (h&NN'h*hN *[ 'N /.V +. N#,bDie NC+ v +- v:. OW Ib
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1 a 3. The Marine nyfronment 4
>]j Approximately 250,000 gpm of. co'oling water for the plant condenser will be withdrawn from the Pacific Ocean' at Campbell Cove on the east side of
. e tr '
N'% Bodega Head and discharged to the ocean _ on the west side of the Head with an
- k h;,y &
,[
estinated maximum temperature rise of> 1A*F. The conderser cooling water does 7
jl
%b .,f not pass through the reactor and will nue become radioactive from exposure to
.y $g
,7,$ neutron irradiation. , This cooling water, however, will contain liquid efflu-ents to de periodically released after monitoring from the rad waste facility.
- The rAtsu.tration of radioactivity in the condenser cooling water before a
discharge to the ocean would be controlled to meet the requirements of 10 CFR
\
Part 20 of the Commission's Regulationar At the request of the AEC, the U. S. Bureau. of Commercial Fisheries of
~
A
, a; l the Department of the Interior. has reviewed the effects of reactor operations
.e ,
7%: on the marine environment c'f sodpa Head. '!he Burosufs report, which is
~~
attached as Appendix II, inflicated that there could be potentially signifi-
'4 cant effects from the discharge of this effluent to the ocean, such as:
e ?;
- ) (1) Possible concentration of radioactivity in seafood, l
~
(2) Possible concentration of radioactivity along beaches used for G public recreation.
9
- .9@,y
[, However, the Bureau of Fisheries stated that it is well established p?j
?] that certain levels of radioactive wastes can be discharged into the oceans an without adverse effects on fish and wildlife, since $:irculation insures ly
,x'
[ ,
' mixing of radioactivity with large volumes of water, quickly diluting and
,. q
(' dispersing the radioactivity. Their report further stated that since the h -
permissible levels and rates of discharge of radioactivity are difficult to J
N$d qgg ( determine in advance for any specific area, an extensive monitoring program 3 s . 'to insure that concentrations of radioactivity in the marine life do not
- 5
'J:[ ;
i ' exceed predetermined levels should be undertaken.
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- ]'
- The experience at Windscale, England, where radioactive wastes have been Aschirmed to coastal waters for severui. years was cited by the Bureau of j 1 .
l x., Fisheries _ as an example that shows no adverse effects ori fish or shellfish l 1b . ,
, yj s fun weetor' operations. N Englisti have determined on basis of a monitoring l .-
y,ff 6 ,
c GpW program carried out over several years that 3,000 curies per month of strontium
-,.g,4 co-90 could be safely released to the coastal waters at Windscale.
i by,i
./;fc:-7 .,
6 k A- i.,
Qy
, .g PG4E)easiepecified that the concentration of radioactivity in the condenser i .
coolimg!watAr discharge will'not exceed that permitted by Part 20 of the Com- .
t t
- ? mission's Regulations. Based orda discharge rate of 250f000 gallons por, 3
, i 1 ;
minute we have estimated that,approximately 4.1 curies of strontium 90 peh i
month could be discharged to the Pacific Ocean, if all of the radioactivity ad in the condenser cooling water consisted of this nuclide. This value is
.:2 i costde61y conservative in comparison with permissible nisase values deter-5
" mined by the English for Windscale.
Extensive studies of oceanography and marine biology to evaluate the ;
i i marine biological aspects of the proposed reactor operation similar to those i\
recommended by the Bureau of' Commercial Fisheries have been initiated by the applicant. These studies are .as follois:
. 1. An oceanographic survey will be carr!.ed out to detemine the
.c o 2;h circulation pattern of the ocean in the vicir.ity of the outfall, c a ..
n+d.;.
' and the capacity of the ocean to dilute the condenser cooling water 4 .s y
. .a.:,
discharge.
NW 2. An ecological survey which will incluie an inventory of the marine
&l '
organisms _in the vicinity of the outfall will be carried out.
ng
??.h -
- 3. An environmental survey of the marine environment will be conducted I 39 ,\
%. to determine existing levels of radioactivity before the reactor j
. ' z. becomes operational. T. Is program would contime after the reactor
,e is in operation to determine any tendencies for neoncentration by sj .'
-q . , a
.n ': :v q. . y
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. ._ mi. Q iq . w A 4 p .m a:.qa m;>.:n.p;+, .
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ne experience at Windscale, England,yhere radioactive wastes have been
@ dissharged to coastal waters for several Wars was cited by the Bureau of
- ,J -
- .;j Fisheries as an example that shows no adverse effects on fish or shellfish w
Mjj from reactor operations. The English have detezzined on basis of a monitoring L:Q; Q; ;,3 program carried out over several years that 3,000 curies per month of strontium M
scur 90 could be safely released to the coastal waters at Windscale.
4 9cl PG4E.hasispscified that the concentration of radioactivity in the condenser 2?
7 cooling water discharge will not exceed that permitted by Part 20 of the Com-r j mission's Regulations. Based ceAa discharge rate of 250,000 gallons per
6 , .
, . minute we have estimated that approximately 4.1 curies of strontium 90 per month could be discharged to the Pacific Ocean, if all of the radioactivity p in the condenser cooling water consisted of this nuclide. His value is wk >
Me r est:4m61y conservative in comparison with permissible release values deter-
..;p
'j, mined by the Engli.h for Wirdscale.
- Extensive studies of oceanography and marine biology to evaluate the
-1 marine biological aspects of the proposed reactor operation similar to those recommended by the Bureau of Commercial Fisheries have been initiated by the i applicant. Thess studies are as follows:
, ' j> 1. An oceanographic survey will be carried out to determine the g%:Q circulation pattern of the ocean in the vicinity of the outfall,
- ge and the capacity of the ocean to dilute the condenser cooling water
'Mi discharge.
en
.,j$ 2. An ecological survey which will include an inventory of the marine
.we - . .
f1 b.b organisms in the vicinity of the outfall will be carried out.
%~
W
- 3. An environmental survey of the marine environment will be conducted M.g to determine existing levels of radioactivity before the reactor
- J.
becomes operational, his program would continue after the reactor is in operation to determine any tendencies for reconcentration by
. . , , , y e .* <8' , 4 f 'l ph. ."'E / d ,
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the marine environment of radioactivity released to the ocean from
' ~' ;"'
operation of the plant. ;
'k :. . .
We have considered the information submitted by the applicant and the Tif.: cosments of the-U. S. Bureau of Commercial Fisheries and have concluded that MC jg - the liquid effluents produced by operation of the proposed reactor can be
<N. . disposed as proposed to the Pacific Ocean without exceeding Part 20 limits.
.r .. . , . .,$
I.,w,i In view of this, we believe 'it is extremely unlikely that any adverse effects r f- :
i on the marine environment will result. Further, it is our opinico that the
/
programs proposed by the applicant for maintaining vigilance over the marine l '
i l environment will be adequate to observe any anomalies that may occur despite i
i these precautions so that these could be corrected well before any safety
,. l problem develops.
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B. Cont'ainment Design
["ll.
is The containment system proposed for the Bodega reactor will serve as f a principal safeguard against release of radioactive materials to the environment m .c in the unlikely event of major rupture of the primary system inside the contain-k]T;G w.4 6 mont . structure. As in other nuclear power reactors, the containment system in
. i?cl,t p,@ this facility is not provided to protect against accidents that are expected or
.n y likely to occur; rather it is provided to safeguard operating personnel and
~ ~
j the general public in the event that the best efforts to design the plant to j the highest standards, and to construct, maintain, and operate the plant com-
,,; potently and safely should fail. In addition, other emergency systems, such
+
. as the core spray system, are generally provided, as in this case, to either pavent or utard the release of fission products under various accidental yd1& . -
conditions.j nese various emergency systems operate independently of the q
kb fg.g containment system to prevent the' release of radioactive materials beyond ,
.y the boundaries of the plant. ney are discussed in a later section of this j report. We adequacy of these systems and the containment system in particular e
is an important safety consideration, l
. :.\ .
The proposed design of the containment system as discussed in this
'de ... t section has been used in the analysis of the maximum credible accident, which
'My is discussed in Section 6 of this report. Dat analysis is the basis for
%2m
- 3) establishing important design parameters such as drywell and suppression 5M g 3 .,7 chamber design pressures, which are later described.
fi[E The containment system proposed for this facility is one which utilizes -
pw,3 iM
. ::p the pressure suppression concept. Its design is similar in many respects to 1M wp that used at Humboldt Bay reactor facility. Significant features of the
, "t Bodega Bay plant containment design include the following:
1
't
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i-t; (1) Plans 'for the Bodega Bay plani:' call for a drywell having a 60 ft.
. diameter spherical lower section anda ' 26 ft. diameter cylindrical y
1: upper section. The overall height of the drywell is approximately
,:.y .
de 100 ft.
hk' ,
T '-
(2) The reactor vessel and four reactor recirculation loops, each with
- s. .
~
..g1 ' a pump, will be 1ocated'within the drywell.
d; fl,th.< (3) The drywell will have an airlock entrance. Personnel entry is' not
- sm planned during reactor operation, but is contemplated with the
,"f reactor coolant system hot and pressurized.
t
] (4) '!he suppression chamber will be in the form of a torus and will j have a major diameter of 93 ft. and a cross-section diameter of
. q.
MJ
~::.;
26 ft. It will contain about 465,000 gallons of water and have an
- @ , air space above the water of about 80,000 cubic ft.
%~ :S Both the drywell and the suppression chamber will be designed and con-structed in accordance with the ASE Boiler and Pressure Vessel' Code, Section i VIII. Piping restraints will be provided at containment penetrations to assun that failure of the pipe will not cause containment mpture. A concrete
] ,
refueling building will be constmeted above the drywell and suppression chamber. Pressure and leak rate specifications for these containment system
'. Q
!'yd components are as follows:
t :(.??
l , e W.M',p.
Yq Leak Rate
%. 3 . Component Design Pressure (% of volume in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
- s3
's pp$.
Drywell 62 psig 0.5 (at design pressure) gd. ~
q Suppression chamber 35 psig 0.5 (at design pressure)
INf.,d Refue' ling building 12 in. H 2O 100 (at 1/4 in. H 2O) i
,..c -
- s.9.' n. ..
In order to proof test the Bodega Bay pressure suppression design,
% Pacific Gas 4 Electric Company is conducting a test program at its Moss
, . .nn.
N
,[ h . s D V , [j: N O d A. *.S e 0 L. .;- 'q \
. e, "
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l rdi'"* ~riu mu%2m'm?"vwc'~ms;'hyRi<dmMPE:Ts@MQQ@fg.'r.h;&Wy52
. i l
- i. _- y y, y,
- 3 ,,, v m. < -- . , . .
..--.. y .- . . _ . . .
17 -
l Landing Power Plant. As in the Humboldt-Bay project, the applicant has con-
, j stmeted a full-scale segment of the, suppression system. In the test of Bodega E' Bay design a single 24-inch di'ameter vent pipe from the drywell to the suppression 11
~.
i hs. , chamber was used. Since the full size plant is to have 112 of these vent pipes,
),di the test equipment represents a 1/112 segment of the containment.
s.g .
$$ '~
Tests were conducted with this mock-up to simulate various accident con-d ditions. A flow comparable to 1/112th of the flow resulting from a complete
$.mf4 circumferential break of one of the 28 in. reactor coolant recirculation
.]
lines (with flow out both sides of the break) was taken as the " maximum
. l credible operating accident" (MCOA). Highest containment pressures observed l
! in these tests were 52 psig in the drywell and 30 psig in the suppression j
\ .
l' r . chamber. These pressuns were observed when the mock-up drywell was pre-
- $,M ;L heated to 255'F and when the sock-up reactor vessel water was subcooled
'[5 f
~ 35'F. Tests at 1}igher and lower drywell temperatures and at higher and lower reactor water subcooling yielded lower drywell and suppression chamber
'3 pressures. ,
In another test a break area 2.5 times that of the MCOA was simulated.
. .) In this test the peak drywell pressure observed was 63 psig. Further Moss l
Landing tests are being conducted to determine whether baffles are needed in
. ;lW
?,g the suppression chamber. We believe that Humboldt Bay experience and the
- VD '
y..* Moss Landing tests will provide an adequate basis for designing the Bodega il.sp containment system.
As another significant containment design feature, Pacific Gas 6
@M l'c Electric proposes that in a number of instances a single isolation valve
% 1p
- will be installed at the containment well in pipes or ducts penetrating the Gg '
n .s containment. However, each such line will also be provided with a second j
'f isolation valve which mey be a remotely operable process valve located
- 'h .
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. ?wt&ueiPremmheenkmekd:skinxn.-~+%@ykgepue'.mp"cdQf:,3&mg:yqy.-
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~
elsewhen. The two isolation valves located at the drywell wall in each main L -[
..', steam line are to close on a manual signal or automatically on the occurrence of any of the following:
!.d (1) Low condenser vacuum
' FEM
,;. ~ , (2) Main steam line leak (in the pipe tunnel)
$ (3) Low reactor water level n$
. [g..n fyj PG4E is also giving consideration to providing protection of the main steam
,[ line isolation valve against foreign matter, which might interfere with proper valve action.
~I no Bodega Bay design is such that during refueling, the spent fuel storage pool will connect dinctly to the shield water above the reactor, m .4 thus permitting direct underwater transfer of fuel without the need for a
.m transfer cask. In our opinion this feature provides in a sir;ple and
.:..m m -
niiable way for both continuous shielding and cooling of spent fuel during g
,; transfer and storage.
[i ne refueling building in which the drywell and suppression chamber
, system are located is provided with a controlled release ventilation system
., which discharges to the plant stack, ne building and ventilation system i
- design is such that the refueling building can be maintained at a negative AC Qq
- fi pn ssure. Discharge from the building passes through halogen and radioactive JM particulate cleanup equipment prior to discharge to the stack. PG6E has L. ~. ?
nm
% indicated that, in accordance with the recommendations of the ACRS, the yh :D system will be designed to permit frequent testing of the ability to filter pA 4:
.3
?? particulate and to remove iodine at specified efficiencies.
,A
'4M -
g -
(De following 2 paragraphs will be revised to reflect the expected -
'iN
,Q Amendment 3.)
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l
,) The staff believe's that the general'coritainment scheme proposed by PG6E !
- is adequate for the proposed reactor fscility. However, it is our opinion gB #4 7'.. *f/ that the detailed design should reflect the specific considerations discussed W
U.N in the following paragraph which we believe are necessary to assure that the W
m-containment as proposed can be reasonably expected to provide the high degree
.Q .;
$.y of integrity, proposed at any time that it might be called upon to contain the jq! consequences of a maximum credible accident. These include those recommended
}.] containment features mentioned explicitly'iri the ACRS report on this project to the Commission. The regulatory staff, in reviewing the details of contain-ment design as they are developed, intendt to assure that these considerations i
with respect to containment testing, penetration design, and isolation valving have been satisfied: 7 Y 1. The design should permit initial integral leak rate testing of the
- a. ,"
drywell and ' suppression chamber ~at their respective design pressure after I
the installation of all penetrations (including piping conduits, electrical
,j conductors, and gasketing closures) and subsequent periodic testing at suppression pools design pressure. In the initial testing, the leakage rate of the containment system should be determined as a function of pressure
. up to full design pressure.
M -m
- 2. The design of penetrations should take into account, in addition to
.;Q the pressure load, the loads or deformations imposed by thermal expansion,
'. :.2 i impact of missiles, reactions of ruptured pipes, and disturbances incident
[5 5 to installation, maintenance or repair. Penetrations should be shielded J'l V:3
, from missiles to the extent practicable. Penetrations should be designed
/., :
P -
so as to allow frequent periodic leakage rate tests of the penetrations n.n
',;,y only (including points of attachment to the containment shellg, at full
.(e
i design pressure. The required frequency of integral high pressure tests I
.hk +
.4 f.% . . . . . w. . ,s a - _
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+
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lp.
of the entire containment system after. operation commences will be dependent
~
y upon the frequency that individual penetrations are tested at design pressures.
M 3. All pipes and condsits which communicate'with inteiior;of the_ primary
. ~ .
}{}
w ;
system or the containment system, and oth6r piping (such as instrumentation l f4 and control piping) which cannot be adequately protected against accidental ,
g nptures, should contain at least two isolation *.alves. A11' valves per-forming the function of isolation valves should be provided with proteci: ion i i$7N
{7i against materials in the system which might prevent proper closing and should
^
, 4 be provided with reliable automatic and manual actuation features. -Isolation ,
valving should be designed so as to permit periodic leakage rate tests.
i 1
'J' Appropriate closing times for isolation valves should'be determined on the j WQ
- 3 basis of analyses of system ruptures which would nicase fission or activation
. 6{ '
]
- $ products outside the drywell while the valves are not fully closed. j
.[q.tp: In order to assure that these considerations have been satisfied, we -
believe that PG4E should submit for Commission review the. results of further l 1
] developmental tests of the suppression pool concept and final design plans
. 1 j for the containment as soon as they can be made available.
l
! C. Nuclear Design Features
- i
. , As noted previously, every reasonable effort is made to design '
w d nuclear power plants such as the Bodega facility such that they can be
%}i;g operated safely even without such extra safeguards as the containment system. .
WYi
@) 1his degree of safety can be assured if the fission products which are produced g in the nuclear processes are confined within the fuel in which they are
,4j.
+
j generated. Since the first leak tight barrier which normally contains these 2J g,,
. fission products is the cladding surrounding the fuel elements, special Uni
]g precautions are taken to insure that the cladding remains intact during both 'j w
, ]. normal and abnormal operation.
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e unm-mawannnne.m
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. 21 -
he integrity of 'the fuel cladding barrier can be maintained if it is not subject to a combination of undue thermal or mechanical forces or stresses.
E 3ernal stresses can be produced by an excessive nuclear reaction rate within v ..
3% the fuel which would cause the' fuel to generate heat at a rate exceeding the ph capacity of the cooling system which ordinarily maintains the temperatures 54 of the fuel cladding within predetermined safe limits. Mechanical stresses die can be caused by forces such as reactor coolant pressure and hydraulic flow.
To safeguard a reactor core against damage' that would release fission f products from the confinement provided by the fuel element cladding, one must:
I (1) Provide a cladding of suitable material and thickness which will
{
2'?
retain its integrity when subjected to the thermal, mechanical, g
and radiation condition of its environment; 9: . -
- (2) Provide for ' control of the chain reaction; and
?(,.l .
(3) P* ovide for removal of the heat generated in the fission process
~
and the radioactive decay of the fission products.
Generally speaking, .the design of the reactor core, including the fuel !
elements, the design of control and instrumentation systems, and the design of the heat removal system should be such that, in normal operation and
,:C
$aw under many conceivable accidental conditions, rsdioact1ve fission products jh would be confined to the fuel elements themselves. ,
MM
- $E 1. Reactor Core and Fuel Elements
'l'M .
Qi De proposed reactor core will be composed of 592 fuel assemblies f each of which provides a vertical channel through which the mixture of steam A ' ', 9
.h and water passes. De core will have approximately the fom of a right i ilN k':
circular cylinder 140 inches in diameter and 125 inches high. One hundred and forty-five control rods will enter the core from below the fuel 5
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4;.4 g gg, *. y assemblies through control rod guide tubes. The fuel assemblies are held in s.
f proper position by upper and lower grid plates which are attached to the cylindrical core shroud. The weight of the fuel assemblies is borne by the
~
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control rod guide tubes which ' extend to .the lower head of the reactor pressure hpp vessel. s p .. ,, .y , i
. .e
.{$ ff Each fuel assembly will be composed of 49 fuel elements in a square
~
array. A fuel element'is made up of a number of fuel pellets of UO2 enriched yuQ. to 2.5% U-235. Rose fuel pellets will be contained within stainless steel tubing which provides the fuel cladding boundary. PG4E has tentatively b proposed that this cladding would have a nominal thickness of 0.011 inches ~l y and would be able to withstand an exposure of 15,000 WD/ TON. l v
-)
g
, Determination of the maximum nuclear reactor rate or power level of a l
M reactor core of this configuration which is consistent with maintaining the JW -
l pji
' integrity of the fuel' $1ement cladding requires a complex. analysis of the s;3 . ,
{
h4 nuclear characteristics of the core, the thermal dynamic properties of 3- the fuel, cladding, and c~oolant, and the hydraulic properties of the coolant
- 4) and coolant system as well as the mechanical properties of the fuel and the cladding both initial and when ' subjected to irradiation.
- 1 The power distribution caused by nuclear fissions which is expected
%q g in the Bodega reactor core has been estimated by the applicant for the
~ -
g@ purpose of making i preliminary determination of the thermal margins which ,
1 Ib M;:J would occur.in the hottest fuel assembly. Rose estimates will be refined y:% by detailed calculations of power distribution in the course of design -l
,l
$n n 9 of the reactor. After operation commences the power distribution will be t
fhff -monitored continuously by a system of in-core flux monitors. Such methods
~w f' have been successfully used in other reactors and should provide a reliable l means of establishing the thermal margins that are experienced in operation.
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N. ' Thermodynamic and hydraulic factors, which ultimately detemine the )
l 1
permissible power level of the reactor, hue only been briefly described by
- PG4E. While at present there does not appear to be a firm basis for establish-
- [ ing.themal limits as high as is suggested by some of the parameters specified
!!N) .
W5sk in the application, PG4E has established a criterion for detemining the proper
)*(( detailed thermal and hydraulic design factors for the Bodega reactor core; 08 .
W namely, that the fuel will operate without loss of cladding integrity over
.ff?
g the design exposure period at the maximum heat, fluxes possible within burn-
.r j out limitations. Operating ex'perience at other boiling water reactors has i
indicated that this criterion can be met and that it is acceptable from a '
~
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safety standpoint. ,
I g The thickness of fuel element cladding tentatively proposed is 11 mils, 4gw4 Cladding thickness for the bulk of the Humboldt Bay fuel elements is I l
'Q'y!
~
~
, mils. On 'the basis of present information and operai:ing experience, one 1
'q cannot be assured that fuel with cladding of this type and the smaller i thickness can be irradiated for the exposures stated without experiencing
.)j excessive pin-hole leaks in the cladding. However, the design of fuel 1
for the Bodega reactor will not be completed until further data from a a
1 General Electric research and development program am available. In our I Q) ify opinion, this program is reasonably designed to provide an engineering y -
kl.N basis for a suitable fuel element design. In any event, extensive experience
- p
.MO5 with other power reactors provides reasonable assurance that a fuel element q%
IN cladding design suitable from a safety standpoint can be developed for the
^Q$
g Bodega reactor, h
M'
- 2. Reactor Control
~.
'$.a Nuclear safety requins that there be reliable means for controlling '
," the nuclear reaction rate or reactivity of a nuclear reactor. Reactivity
,;. can be related directly to the rapidity with which the neutron chain reaction ,
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'When reactivity id positiire',' the, chain reaction grows; that is, the j
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, - rate at which nuclear energy is released by .; . fission is increased. Conversely, j hyl when reactivity is negative,the chain diminishes and power falls. He opera- l
,y I g :d ting condition of the reactor, its temperature, pressure, power level, void
-. l g.g content, and fuel exposure all affect reactivity. In.our opinion, the )
u4 , . . .
i generaliay in which each of these factors affects reactivity in a reactor 1.l$$
C,T of this type is quite well known, and the theoretical and experimental b,b.L lg methods for investigating reactivity effects are sufficiently developed to
- n. c N
3 permit design of the Bodega reactor control system to proceed with confidence.
7 w
The stability of a reactor is affected by operating conditions which l
. have transitory effects on the neutron economy of the chain reaction.
. > e.
.y.~. .
It is expected that, with the possible exception of the volume fraction of 3 i
' 4lg, steam within the reactor coolant, this reactor should be stable within l4.5 ..
.};4 ~ the range of operating conditions contemplated. D at is, an increase in
,' o '
- reactor power will cause changes in operating conditions which will have a strong natural tendency to decrease reactivity and limit the power increases, a' With respect to steam volume fractions, preliminary calculations indicate j 1
that the following would result at rated operating conditions at Bodega: I l
4 average core voids - 37%
average exit voids - 58%
ze "Although we are not aware of any substantial operating experience that ')
de (gg would confirm the stability of ' reactor operation at steam volume fractions 4W this high, PG6E has stated in the application that analog computer studies being made as part of the research and development program will show that dw
' the plant can be designed to exhibit satisfactory dynamic performance with
'W y
^
2 such high void content. In any event, we believe that with appropriate
,g% y E limitations the high void conditions proposed can be safely approached in 1.1 my
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reactor tests devised so as to determine the proper range of void content for
. O, -
,. ,q stable reactor operation.. He main' purposes"of, the Bodega reactor control system, therefore, will be to provide a means of precisely and reliably 3-l .,,.
iQ adjusting reactivity. .
4:<
- FQ Necessary safety objectives for the Bodegsi control system are to
.1 ,
M provide means for (1) shutting the reactior down by a safe margin under any KSi ;
g:g circumstance, (2) starting the reactor and increasing power at a safe rate,
.S
, $s7 and (3) maintaining the reactor power level within the capabilities of the heat removal system.
3g The control system proposied for this reactor is an array of 145 moveable control blades, which have'suffielent reactivity worth to keep the
', ii
'Im reactor safely shut down, even though one of these blades (or rods) might be W, ,. stuck out of the reactor core. ne k-effective of the reactor with all W .
D, control blades in tho' nactor core is calculated to be 0.97. .In our opinion, ,
?
'the strength of the controls is suitable for safely shutting down the reactor under any credible circumstance.
He material in the moveable blades will be boron carbide contained t '
i within 0.175-in. O.D. stainless tubes similar to those presently in use at I
Dresden. Additional control devices, removable only through a loading )
,gJ procedure, are provided by fast control curtains of 0.1% boron stainless el ~
steel, which will be semi permanently located between selected fuel l
@q - a i
Dh elements. He fixed control devices provide a flexibility in the reactor !
Md lC design so that reactivity'of the core can be easily adjusted to attain the g&d M .ggy shutdown margin required from safety considerations.
..g w-no.zwactor design also ince rporates a liquid poison system that can fv be used to inject sodium pentaborate into the core in the unlikely event 2;* complete shutdown cannot be achieved by use of the control rods. Similar
, g, ,
systems are provided in boiling water power reactors now in operation. ]
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- w. .
,_ . 's The hydraulic control rod drives to'be.used in the Bodega Bay plant ;
e ,: < . e: n (
}, are to be designed using the same bs. sic';c:acepts as have been employed in drives )
? ni .
..{ .
in use in boiling water reactor plants at'Dresden, Big Rock Point, and Humboldt j
- j. Bay. Water used. as the hydraulic fluid can be applied to either side of a
.' i . .
iy f piston which is mechanically coupled to the control rod, thus providing for
- 't .
sgd either upward or downward rod motion. Only one rod can be moved outward ;
(increasing reactivity) at a time, and it may be moved either continuously 1 .c
% ' .. I or in 6 inch steps. Rod speed is controlled by orifices which regulate the l
ij 1 flow of water away from the low pressure side of the ;iston. The rod speed for normal withdrawal will be controlled so that at the' maximum rate of withdrawal the reactor power would not increase at a rate fast enough to 1 "
i M lead to serious consequence.s.
)
. y. W ' All rods can be inserted simultaneously, shutting the reactor down.
~
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yd i
,l y,Q . Rods are scrammed upward by applying pressurized water from either the
- . : I reactor or from accumulators to the bottom side of the drive pistons and 4
h simultaneously relieving the volume above the top side of the pistons to the t
l scram dump tank. The drive is locked in fixed positions by collet fingers which engage grooves spread at 6-inch intervals along the moveable index
] tube. The collet fingers support the weight of the rod and the downward forces due to reactor pressure.
'. Since drives similar to these have been used at other plants, an .;
- ]h;g
{, important part of our evaluation of these drives is based on previous experience with these drives. This includes Dresden experience as well
- Q@j g.{fg as initial Big Rock Point operations.
,j Detailed design of drives for the Bodega reactor has not been made. !.
$$.- General Electric is considering modifications of designs already in opera-
- l I tion in other reactors that will minimize the possibility of foreign !
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y- material accumulating in the rod drives and impairing operability. The.
4 J
!.3- applicant.has also indicated that functinest and endurance tests will be
- i. made on the prototype Bodega mechanisms, but the detailed procedures for I-
] these tests and the acceptability criteria have not as yet been proposed.
",, Control systems which an designed to react rapidly to demands for
.g3 .
g 4.:
. , shutting a reactor down generally have some potential for accidentally increasing reactivity as well. 1his aspect of the PG4E control system 4
c.ft design is discussed later in this report (Section V), when consequences A s
.l . of a rod dropout are considered. PG4E has indicated that devices for
., limiting individual rod worth and for impeding the fall of a rod are under b
3 development. Such devices could enhance the safety. of operation and simplify i
l the procedures that are presently used with similar drives to provide the q' necessary assurance that a rod dropout accident cannot cause a serious
}9:
q.
- . public hazard. .
~ . , . ,, .
i c.g,
. jf. We believe that the design criteria for the reactor control system is
? a ac. cept able. In view of the importance to safety of the detailed d'asign of l the control rod drive system, the staff believes ..that PG4E should submit timely toports for review by the Commission during facility construction on i
development, design, and testing of this system.
1
- 3. Control and Safety Instru;nentation
..Y.a ,
[,
.The instrumentation necessary for operating the control system
. a.
'1 -'
and the reactor safely in a nuclear power plant generally involves a large 1
3 number of sensors throughout the various process systems. These sensors i D.3 j; measure a variety of variables, such as neutron flux and gamma radiation na .
'!ic levels, and temperatures and pressures of various fluids. Information col- l
.inf; .
,- . w. .: , .c .. ,
i 91
, v,t lected by the measuring instruments is used to guide the operating staff in !
<- controlling the plant and to actuate automatic control devices.
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The instzw....., circuits,. and control devices which are of most importance
., to public health and safety are: (a)- th1e necessary for and contributing to
- 1 d stable reactor operation, (b) those used in control of radioactive fluids
.]
l and effluents, and (c) those used for control of emergency equipment.
e V A generally adequate description' of proposed instrumentation is presented M
in the application. Sufficient details are not yet available to determine 3M whether instrumentaticu provisions have been made for all essential functions
E} Ii or to determine the degree of reliability'that should be attributed to the b reactor protection system, which is described by PG6E as " fail-safel. ~i These, however, are design problems which appear to be recognized by the applicant and which require only the application of well-known engineering methods to i
provide an acceptable design. The staff intends to evaluate the reactor
-i f.g control and safety instrumentation in. Setail prior to reactor operation ng p% in order to assure that proper attention has been given to the need for e w. -
automatic functions and the reliability of safety instrumentation.
- 4. Primpry Coolant System -
k Water is circulated through the reactor core by this system in Jj order to remove the heat generated by the nuclear fuel. The heat changes 1,
the coolant to steam which is converted to mechanical energy in the turbine j
! i j in order to produce electricity. l
}ly The functional integrity of the primary system is necessary to the y
d integrity of the fuel element cladding since heat must continually be removed
,q:
,, y during reactor operations in order to preserve the integrity of the cladding barrier. If this objective is met, then the primary coolant system also
.x <
serves as an essentin11v leek t1&* , ladependent imr*ier, in addition
,C l
]&'{ to the fuel element cladding to retain radioactive fission products within ,
s .
); the facility. The principal components of this system are the reactor i
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pressure vessel, the primary coolant piping, pumps and isolation valves.
P 2p1 - As proposed by PG4E, the reactor core will be located in-a reactor i
- ,j pnssun vessel des 5ed, built anditested in accordance witir Sectiour VILL r
[.
b ?.:
of the Boiler and Pressure Vessel Code of the American Society of Nechanical f$$ .
Engineers. Tho'50 ft. high by45 ft. diameter vessel will be constructed
.,u
'.T J of carbon steel approximately 6 inches thick. The interior of the vessel -
i will be clad w'ith stainless skeel applied by welds overlay methods. The
~
p I g% ,-
)
,g I design pressure will fd 1235 psig at 575'F. I Steam generated 'in the reactor will be passed through axial flow
]
. 6? i I
. I 9< -
steam separators and driers located and then inside the reactor vessel Ul . through two 20'-inch steam lines to the turbine, which is located in a separate building. Water, after separation from stets in the' reactor vessel,
.dth passes from the reactor vessel through four 28-inch pipes. Four recircu-
^y
' g@ . lation pumps, one in each loop, provide the driving force for circulating 1.va: s n .. w . .
rP water through the reactor core. Feed water from the condenser is injected into the reactor vessel by a pump driven by the main turbine.
Pressure vessels and piping located within the drywell will be designed, tested and. constructed in accordance with applicable . requirements of the ASME
~.
1 q Boiler and Pressure Vessel Code. Piping outside the drywell will conform .
L to the requirements of the American Standards Association Code for Pressure
-gu w
W Piping. Twelve safety .c valves, arranged to discharge into the suppression y -- -
. si g
,u.
chamber, are provided to protect the reactor and primary system from ,
271 .. _ ,
+
over-pressure. .
{&m N 7.M@,
The general concepts and crite'ia r of primary system design proposed are, in important aspects, similar to those used in several other nuclear j g&;. a;. , . , . r, ,
W power plants. We believe that a system designed according to these plans t
u
.Q. $. can be expected to fulfill its s' afety functions.
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D. Emergency and Safety System Design
. Emergency. systems provide means.either for safely continuing reactor ,.
'j' operation in the event of some' equipr.ent failure or operator error, or for
,. ?); . limiting the extent' of' damage and resultant hazard. 'In many instances, design n .
NQ features of the facility which have been provided for.the primary purpose ~ of 4$p.v.4, making plant operation more convenient or "reliablef are, in effect, emergency -
1 m c
.3 j systems. Other features are designed primarily for the purpose of providing co .
. :.1 yd emergency functions.
.1he principal emergency systems and. components' propose'd for this
%, reactor 'are:
, (1) Alternate power supplies for critical electrical loads j
. i f;,w; (2) Reactor control safety devices and circuitry
.}?' 4 (3) , Liquid poison inpection s,ystem
.y (4) Emergency cooling system 4 -. .
sj -(5) B,leed and feed system (6) Core spray system j (7) Containment system I j
Some of these systems have already been discussed in this analysis.
4 Principal features of other emergency and safety systems are discussed
' .'. ) .
u r.
-[ in this section.
As is indicated in the discussion of the consequences of potential M.'4*. v .
,~
,', I' accidents in a 1 ster section of this report, a number of these emergency
=. y p systems, mainly containment and emergency cooling systems with their
.g f.h 2c.
associated water supplies, power supplies and controls, sust be relied on
, ;;.gsv g/jp to limit the ' consequences of serious reactor accidents to an acceptable a y "
<4 f. level. In all such systems, therefore, a high degree of reliability must M
e ,. be provided so that the system will perform properly in adverse circumstances.
c:w.
This requires not only careful design of the principal features but also W2
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attention ':o such related equipment as signal and control circuits,. power supplies, and instrumentation. - Maintenar.ce 'and frequent testing of emergency
+
,j
.n systems provides the final assurances of readiness of emergency' systems to response to the demands. placed upon them. These factors must, therefore, be taken into account in final design.
m
?$ ,
, , . 1. Power Supply 1
j
?
- 9 $ Protection of power supplies is provided on several' levels, as j
$$ 1 hc: described in PG4E's application. De plant is tied into the PG6E f.($ distribution system by two 220 ky circuits to Ignacio Substation. All Uk plant auxiliary power requirements can be met by either a transformer tied
'. to the station generator or by a transformer tied to the 220 kv lines. An
[..
additional external transmission line and transforiner of limited capacity E}g9 and an engine-driven generator provide emergency power to equipment necessary u
I .
for safe plant shut down. Station batteries will supply the electrical 4 . . . . . .
q '
{j energy for the more critical loads. In our opinion, these provisions are
'[ suitable from a safety standpoint.
i.
% 2. Emergency Cooling Systems
- w j:'
A number of different means will be provided for removing' after- ,
l heat generated in the reactor core as a consequence of radioactive decay of fission products. Such provisions are necessary to remove decay heat e.v g after reactor shutdown to prevent melting or rupture of fuel elements, y[f}j which would lead to the release and dispersal of fission products. Rese GAh
' 'i(' provisions will include:
DM (1) he normal condensate-feed-water system $
1
?[W (2) An emergency condenser which can be put into operation in event ~!
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y the reactor must be isolated from the main condenser t N:; e
.)
y (3) A low pressure shutdown cooling system O 'l
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(4). A bleed ~and-feed system which releases steam at a controlled rate w
M- to the suppassion pool' l
, [ (5) A high pressure core spray. system.
,. (6) A low pressure core spray system y
3g A number of sources of. watoir (and pumping capacity) will be available ;
I 4!j' . -- .
j to restore water,,lost x through accidental ruptures or through bleed-and-feed
.y ,
, ' operations. ~ Both high head and' low head pumps will be prwided with some J fh -
back-up pumping arrangements. In our opinion, these systems in combination
\
'pll-
- 9 -
with emergency operation action should be capable of reducing to an a
acceptable degne the amount of fuel damage and fission product release c..,
cO , s from the facility which would usult in the event of a major. rupture of
- g. : ' the primary system. ;
i .
(# PG4E has indicated in Amendment,3 to the Application that the emergency L .- i' y ,]
. cooling system will be provided with pump backup beyond that provided by p . < . . ,. .
g f, the auxiliary (startup) feedwater pumps. We believe that this pump in conjunction with those previously described should be designed so that
,j the plant can be provided with both high pressure and low pressure reserve
, pumping capacity beyond that described specifically in the application. ;
- %ese featuns, in addition to the final design factors already discussed t
.3 must be carefully reviewed in detail prior to reactor operation. H ere is
.g
.: f no reason to believe, however, that the final design for the emergency 4"'
systems cannot suitably *piNwide for the necessary safety functions for g- '
Ed
- , +
this facility.
4.Xq 73 E. Radiation Monitoring Systems Design
-A
% A Radiation monitoring equipment is provided for two purposes involving !
KA skfety of operating personne1 and the general public: (1)
~
for monitoring 4.y , >
3,I of radioactive effluents and (2) for determining levels of radiation in m.
work anas in the plant.
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. ..hy s
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- Ai ne following measures, which are typical of other facilities, will j a.( '
4 49 be employed at Bodega: l
.p.,,
(1) Batches of liquid wastes will be " analyzed radiochemically prior ,
j
+., .
4 Qlj to discharge to determine quantities and types of activity present; '
M i6 (2) Primary coolant and fluids in various auxiliary systems will be h/M ij monitored by radiation detectors and samples for determination 5/d of quantities and types of activity pusent;
- ffli i
($<p (3) Continuously discharged gases will be continuously monitored '
9(.i:
pp.i to determine the quantity of activity discharged; SO i; (4) A program of radiological monitoring of the environment will 9a be conducted; and s -t f (5) Fixed and portable equipment will be used to measure radiation
- q. , .
g
.&,h levels in occupied regions of the plant.
i;M,.e ne type of plant proposed and the environmental conditions do not wrii
!e pose any unusual' requirements for monitoring methods and equipment. We l believe that proper monitoring equipment and methods are available to
, ': .t
. fulfill the requirements of safety at Bodega.
'M ]
~ ~ '?( F. Waste Treatment, Storage and Discharge Design Features 3 he applicant has described in general terms the radioactive wastes
- . t w
jg that would be produced during operation of the proposed reactor facility
%: and has proposed general methodstfor management of these wastes in order T93 N, ,u ,D to meet the limitations of 10 CFR Part 20 of the Commission's Regulations.
~;.c2 \
,Qg'3 he sources and general character of these wastes and the general methods my l l
],('pf proposed for meeting the health and safety nquirements are summarized I v.c3 '
.)M briefly in the following paragraphs.
y~M ,
' =
- 1. Radioactive Liquid Wastes
' atj%g{ ,c The principal sources of radioactive liquid wastes from a plant l w.R of the proposed type consist of small amounts of leakage of primary Ob'? *
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- 4 j '
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coolant from valv" and equipment when maintenance is performed, and wastes 1
.. .r. e Q
~
from decontamination procedures. Other' sources include 1aundering opera-i
, Nj tions for contaminated clothing and laboratory operations which are carried ]
.w
., + i*y
,Gy out as a part of the reactor'and power plant operating control procedures. ;
1 '<*p. '
y L The amount of radioactivity in' these liquids is variable and will depend
]
& 1
, primarily,upon the concentration present in tho' primary coolant water j l
@ system. The radioactivity in the primary coolant consists of fission- J products that may'be released from the fuel through minor imperfections M
xj.)
in the fuel cladding and of irradiated impurities that may be present in
- g s ._
the cooling water. Such impurities would' include corrosion products from
,-j
% the coolant system and fission products from traces of uranium impurities f
- gi that may exist in the fuel cladding surface.
ue-
. Present experience with this general type of reactor at Dresden aMJ ,
indicates 'that the range of radioactivity concentration in these liquid
~ -
igi wastes may' vary from values which would be low enough to meet the drinking ;
.5 i
-Q water requirements of 10 CFR Part 20 for the public without treatment to j y ;
M as much as several microcuries per cubic centimeter. The volume of these
!82 wastes may vary from a few tens of thousands of gallons per month to a few 4,@:
?7 1,, hundreds of thousands of gallons per month. j e,. )'
1he applicant is proposing to construct a special system of drains
,, T n
? and tanks for collection of the radioactive liquids from all potential 7 A ,.
yv 1 sources, and to provide a Radwaste Facility for monitoring and decontainination
%.a u -
s.c j9[ of these wastes. Radioactive liquid waste will generally be disposed of by M3 injecting it into the condenser cooling water effluent stream after ff R ~n .
.2. 9 monitoring for radioactive content. This disposal will be so controlled ,
-: 1 ,
W +
. , . . )
- ," ' that concentrations in the effluent stream at the point of discharge to the Pacific Ocean will not exceed those specified in 10 CFR 20. We have
_(/*
der
. l,
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. . g. ,. m a
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y .
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a.c ' '
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~
reviewed the general concepts described in the application for the design
~
of this system and believe that they are adequate to meet the safety require-monts of the Commission's Regulations.
' ' .[.m$
- 2. Radioactive Solid Wastes
}f n,n.
'f.!Ed he principal sources' of solid wastes include spent domineralizer '
. ' ith
. usins, filters, scrap equipment and miscellaneous trash such as' hand tools,
.q@s4 laboratory ware, etc.
J
\
gj he applicant proposes' to' collect the radioactive domineralizer
, k: iy resins in storage tanks which would be located in the Radwaste Facility
..g '
/[.jy o ,,
and to dispose of those materials by' shipment to an AEC-licensed waste
.I disposal facility. De shipments, of course, would be required to meet
. 7y ,
j.; the appropriate AEC and ICC standards for shipment of radioactive materials.
.m .
% All other solid waste would be collected and stored in 'a vault constructed i$
g$ for this purpose pending packaging. and shipment to the licensed waste W
. u, v. .c QM, disposal facility.
.7
. 4
% ;3 We believe that the facilities proposed by the applicant for handling l ip,y these solid wastes are satisfacto'.'y.
I 3. Radioactive Gaseous Wastes
- p.
, Gaseous wastes from a reactor of the type proposed consist principally of non-condensible radioactive gases that are removed from the
.uhw
-(p main condenser by the air ejector, and from the turbine gland seal system.
N
&,j0.
he process areas, laundry and laboratory will contain trace amounts of g
contaminated dusts, mists and vapors. In addition, the drywell will contain g'*t:r., radioactive argon as a result of neutron irradiation of the air within m aj M this cavity, although this air would only be released to the stack on an
- intermittent basis whenever personnel access to the drywell would be required d for maintenance purposes.
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., A 1he. air ejector off-gases, which comprises the bulk of the gaseous-
- j. rad'ioactivity to be released to the ntriobhere through the stack, will i
L a .
J consist of various isotopes of nitrogen and oxygen. The off-gases also may
,dg e 4!d contain non-condensible fission product noble gases, mostly xenon and 4 h
-pyy
.~
krypton, that may leak through minor imperfections in the fuel cladding .
,}- into the primary coolant system or result from irradiation of trace quan-e < . .
tities of residual uranium . contamination on the fuel cladding surface.'
^
o o 7
Se isotopes of nitrogen and oxygen expected to be present in the air ,
.}
. . .< .+ :
ejector off-gases as follows: N-13, N-16, N-17 and 0-19. Based on the
.s. A;a.. ~/ .if m'
50 experience at the Dresden Plant we believe that Nitrogen 15 will be' the .'
g ,,1 mjor contributor to the total gaseous radioactivity release to the atmos- j H;y phere during normal operations. In this regard, experience at Dresden has
)
et ^
.Y.m; shown that concentrations of radioactive gases released an well within the r 1 s: .x ,- .
limits est ablished by the Commission's Regulations. I f.Q ,f 4"p ~
he average annual rate of gaseous radioactivity release from the stack j at this facility will' be limited to a specified quantity which will assure
'q Y that the requinments of the Commission's Regulations are met. The
,: .g meteorological and topographical conditions at the site as well as the
', engineering design of the plant will be taken into account in order to
,m.
,& establish this limit. The applicant proposes to measure the amount of
- M radioactivity nieased from the air ejector and instantaneous radiation a op; l
N6 level to provide for an alarm and for closure of the off-gas vent system ,i Qh valve if the radiation' level would exceed limits pre-determined by the !
$g.
Ma4 Commission to be acceptable.
The system proposed by the applicant for measurement and control of the o2 '
_ 1!
radioactive gases'is an application of concepts that have been previously (
@Jj wM approved by the Commission for other reactor plants of this type. Accordingly, g we believe that the methods proposed will be adequate to satisfy the health and
- %g g -
safety nquirements for operation of the type of reactor facility proposed.
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~, '
V. Research and Development Program
'h 3 While the design of this nuclear power plant is similar in many
- l
.. , : respects to other plants now in operacion, certain components and features J
^~
of this facility will require research and development work in order to com-
,u,
.].yf. t t',p plete their detailed design and to establish the adequacy of the intended design. and operational parameters. In recognition of this need, Pacific
~
$j Gas and Electric Company is conducting research and development programs p-n
$1j as outlined below:
(a) Noteorolog. A meteorological facility is being installed at the site to provide necessary data for atmospheric diffusion studi,es.
j Instruments will be mounted at three levels on a 250 ft. tower and will
.j measure temperature and wind speed and direction. All readings will be
].),p digitized and recorded on paper tape.
m.y
- @ (b) Oceanography. De capacity of the ocean to diffuse the condenser
$P .
cooling water and minimize the effects of temperature and radioactivity on the marine biota is being investigated in a series of tests conducted at the site. Rose tests include use of drift poles and uranine dye as well
/ as measurements of temperature and salinity. Wey will continue through at l 1 east one annual cycle of oceanographic and meteorological conditions.
4 .
gg j (c) Marine Biology Survey. An ecological survey is being conducted n
[
Mm to papan check lists of the marine fauna and flora of Bodega Head and wjy.PM Harbor. .
(d) Radiological Survey. A preoperational moniterring survey of the
. jf, p']
site and its environs will be initiated two years before commencement of 1.5-;
NM operation of the reactor. The details of this program have not been com- ,'
4;gg
'j . "pleted for Bodega Bay. However, it is anticipated that it will be similar
, N,[U 37 : L*;
to that conducted for the Company's Humboldt Bay nuclear unit.
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(e) Pressure Suppression Tests. As -described in the applicant's hazards ' summary report, Appendix'I, extensive tests of the pressure sup-J 3 pression concept have been conducted. Additional'. tests will be conducted c{
.at the Company's Moss 1.anding Power Plant to determine whether or not baffles h
. ch,u.
between vent pipes are required in the suppression pool.
i g
' .In addition to the research .aiH development work being carried out by
- mk- the Pacific Gas and Electric Company,' the teneral Electric Company is carry-nyp.!
y'rp ing out a number of research and development programs of safety significance that will influence the. design of the Bodega Bay plant. 'Ihese are: q P
(a) Fuel; Development. Results from fuel element development tests ;
.'L -l- and experience with fuel designs now employed in existing reactors will F
form the basis for the selection of the design of the Bodega fuel elements.
.j
. q{:.a (b) Instrumentation Development. In-core startup range neutron detectors
~ fj,y are being developed as a possible substitute for the previously planned
- I.+"m4f**.J . ,
2R out-of-core detectors.
t (c) Control System Development. A prototype Bodega control rod drive is currently being manufactured. It will be subjected to extensive
'h ,
developmental testing before the final drive design is released for manu.
)
facture. Several devices which could reduce the likelihood or magnitude y[
$ of a control rod dropout accident are being developed for pssible use in the Bodega control system.
l l
l
'l
- h NWl ' ~
(d) Nuclear Excursion Analysis Development. Analytical models are
}K E!M '
being developed for the more accurate prediction of the physical conse-quences of nuclear excursion y .)
g j s
2[.y
- ... It is our vinion that tlie*;fueEFoE~snd
- development programs proposed to - '{)
$?,d ~
'4
, .W;;n lg ,
becconducted arecreasonably designed"to resolve the'.sefetyrgiettions with respect '
l'6) ms M thdee features; or components'.*of' theJBodegar remetor"whleh reqdire research and
- "'M '
9-p@
. development 'in. order to complete their detailed' design or to establish the ade--
,..~,
- 9. quacy of that design ~ int light: of the proposed operational patameters. As i
. ,! iTAM94specifically noted in various sections of this report, we believe that the ,
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[ L applib: ant shculd submit the results of@ticular research and.develognent programs p
t, d as they are completed in order that the yugulatory staff may determine the ade-
., .j. 4
.kfP] 'y,q quacy of the proposed detailed design during facility constructive.
-b 9 VI. '.* Analyses of Potential Accidents 7/
9..AM gj
,. ~ -
+
The design features of the plant have been described in the.pravious
,. - _,... # ,~ . ,
^(9 ~ sections, and in many cases the safeguarda provided by a particular design _
N [.. q f.@s featum or the operational limits' imposed by safety considerations for a Jt' bu ,
, . particular feature have been discussed), In geieral the criteria fp[ plant 9
WO . design should includes
~
(1) means to control radiation ~ hasards (including
~
,j.,, .' ..
discharge If radioactivity) during normal operation; [2) design features to
- R. .
}' minimise the probability of having an accident; and (3) design features for Kf ' mitigating tho' couseguences of an accident should one occur.
n:M - -, .
.&' The means for controlling radiation hasards during normal oporr'. ion p 'x. .. ..
- T? ' ' will be provided by suitable shielding and radiation monitoring in the case
~.y. *
'# - of direct radiation emitted from the reactor and. by proyer monitorf,ng of radioactive wastes which are discharged from the plant site. For wastes
, discharged from the plant, the release rates shall be such that they do not result in personnel exposures in excess of 10 CFR 20 limits.
2 The adequacy of the design features that are incorporated to mitignte V[4.%
the consequence of accidents in the unlikely event that one should occur 9-Jte' are evaluated in the following parngraphs. The consequence of these tcm Nh
- accidents to tne beslth and safet* of the public is presented taking into K.:,
'y '
consideration the safety featv.ree afforded by the containment and other Q .-
' g' ;
emergency systems at:d the environmental character of the site.
- To evaluate th5 adequacy of design features to be incorporated into Ghl the plant design Eo minimise the probability of accidents, a number of
%.,'l, jg M.; representative abnormal conditions, equipment malfunctions and operator Mh g,Mtr( errors were postulated and evaluated by the applicant. Those which were
. A,
?'
N &, W,,m,,. presented %w,. . ,.psygsu in the Pnliminary up. ~ 1.w Hazards Summary. ....nReport included:
...w- ..w. ;, M
~i= >- - . .
T: 5 ? W Y k Y , f M W h!'.4d W f T h h, 3 *l* , **:Yh.j. *N,f.Qg}t ?f.??Yh.Q'j$,9f b
[*; '
< , r . ro ,; .
- 2+g :
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' . . . .'--i . y -
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. 41 ..
t
}
- a. Changing pressure regulator handwhed! setting
, b. Continuous control rod withdriwal or' insertion ,
i, l i . c. Loss of electrical load
,..a m d. Control rod drive malfunction Yd.!
jyg e. Recirculation ptsap failures !
DIEr', -
~ 'i ; f. Main steam valve closures
[a, .
- 4. . Q.. , g. Failure of reactor safety valve to resent
- h. Failure of reactor safety system
- i. Fuel cladding failure l 1
.: j. 14ss if feedwater
.l
.l
, k. Loss of condanser vacuum
.t 7 ', . 1. Loss of auxiliary power -
ur; .
Instrtunent air failure g;1 4
m.
. . . .. :.,a
. .n, Pressure regulator failure
- o. ~ Emergency condenser tube failure
- p. Reactor system ruptures inside the drywell
,. q. Failure to replenish cooling water in emergency condenser
! r. Startup accident
- t 1-f . s. Fuel loading and handling accidents
~ MfM 't. Cold water'secident yeg ,
~. -
'e fj In addition to these conditions listed above, three equipment failures 9,. . 3 termed " Major Accidents" were evaluated by the applicant. 1hese accidents
..f
+g..j. .
4 included:
1.h. w .- .
j).y[ . , , ,
- a. Control rod drop accident
,up,4 .
nf
- b. Main steam line rupture outside the drywell >
. W, '
- c. Reactor system rupture in the drywell
.w.
- a ;
, .;
- w
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4 s - 42 -
l In some of the usifunciions and failures presented, the evaluation is J. not yet completed; however,.the applicant has stated that when the analysis
- i
. is complete, the results will be used as criteria in the detailed plant
- e. ,
design (for example, to size the pressure relief valves and to set the py;j yt isolation valve. closure specifications). j l:,97 a.: ..a ,,. , . . a, ;. . ,
4
- In our opinion the evaluations that have been completed at this time 7l.l']..e have fomed a satisfactory basis for determining the nature and consequences j l
l- of the hazards of reactor operations, . including the consequences of the maxi. j mun credible accident. In addition, these analyses have established a proper i
.t
, ; ! basia for the detailed design of emergency systems, except for further 1
consider.Ltions which should.be given to the control rod drop accident.
. In this accident, calculations.by the applicant indican that the most 0;g ,
l 1
3
.MJ reactive control blade could have 'a reactivity worth as high as .036.
T... Additional alcula,tions 'show that if this blade were to drop free of the core, a minimum period of 3 milliseconds could result, and the average fuel l
temperature would reach 5500*F in the uncontrolled fuel zone. The conse-3 quences to the reactor vessel in the event of this accident are not entirely
, clear. As part of the research and development program, the applicant has
! indicated that they are developing analytical models for more accurate
.g prediction of the c, consequences of such a nuclear excursion, and that the W forthcoming SPERT destructive test will be used to check the model that is 1
- ) developed.
.1
- ' d In adducion to the analytical work, a rod worth minimizer computer and a rod dropout velocity limiter are being developed for possible use
.~ , -in the Bodega Plant. The rod worth computer would continually monitor con- -
A trol rod patterns to reinforce procedural controls provided to insure that
, . ~ - ,
patterns causing individual rods to assume undesirable high reactivity ty
?.4
- W ed,;N &&!bki&55.,4- 5k.5t,.: -*<- y. 4 WL ' '& a: ?.e . . dnAc#Ni.1$h.
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^ l [,' [ .s#. . h3 1 i yorth are not used. Conceptual designs for flow restricting devices that
.i ,
would limit potential centrol rod dropout velocities to safe values are
}
o being developed. Should there be unsatisfactory results of the research and
,. i gh. . development work to develop experimental verification of the applicant's 3N position that a rod dropout accident of this type vill not endanger the n?s '. - t ~
kj mactor vessel, h believe that other design fratures, such as' the rod
, x.y- .
Q ~-
vorth minimiser computer or the rod dropout velocity limiter, should be incorporated into the plant design. However, we believe that by use of
,fj these alternative approaches 'one can obtain adequate assurance that the
,i control rod drop accident would not have consequences as serious as the ,
maximum credible accident discussed in the next sectico of this report.
, VII. Maximum Credible Accident Evaluation h
n.
- cs For the purpose of determining the upper limit of public hazard incident
.f . to operation of this facility, the applicant has hypothesized a major nuclear accident involving a substantial release of radioactive fission products
'l from the reactor fuel, and has estimated the consequences of this acciden.t in terms of potential radiation exposure to the public. This analyses takes
}.f
-i
- into consideration the moderating effects on such exposures of the containment 7; , and other emergency systems { and the environmental characteristics of the y
pg site. The maximum credible occident chosen by the applicant results from
- n . ..
', f '
an instantaneous complete rupture of one primary coolent line inside of the t
M
- 3:1 drywell after reactor operation at rated power for an extended period of jj.h time with the fission product inventory at a maximum. The pipe rupture Md
, would release the pressure in the reactor system (assumed to be at 1250 <
.[qa -
e M
- n peig), resulting in all of the reactor coolant system water flashing to
$.en$r steam. An ismediate buildup of steam pressure in the dryvell to 'about 62 3
.. psig would ensue and pnssure would increase in the suppression chamber to 4
c.. .
'[ is
)yd$h . W # w.; A,9 4Nm94.%-,.%.1 ts.; . . . . a..cy ,g. g m d
- y~v~.y=nwarmttw
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.. i SW* Y :, , :.:.h.. ~. : .. .: ~ n . z. .
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about 35 pois. The pressure in the suppres's' ion chamber would be reduced within a few minutes due to steme condensation in the water contained within a z, the suppression pool.
r:
.h~h;i Other assumpticas concerning the magnitude of the accident and the
. 6J% s car > .
effectiveness of the engineered safeguards systems for alleviating the Wo.'e , , ,
. gf. . i tg severity of the consequences are as follows: '
- wy,'
(1) The loss of coolant from the reactor is assumed to result in melt-
/
' ~
ing of cue-half of the reactor fuel and damage to the cladding in the remainder of the fuel. (The facility will be provided with an
- . emergency core spray system for preventing significant damage to the core under these severe conditions; however, for the purpose
'I d / of this malysis, it is assumed to be only 50% effective.)
M. 4 .d..
-(2) The fission product nicase frost the fuel is assumed to be the aW . , . . , .
following:
.. s: '
3.,
From Damaged From Melted Total Release
's Fuel Cladding Portion From Fuel i Noble Cases 20% 100% .# 60%
ilj
. 1 Halogens 20% 100% 60$a I
Other Solids 0 .0 5%
- y;y -l 1%
,'il41 '
if'waNd .(3) One-half of the halogens are assumed to be removed in the dryvell
- ;;'., a .<l gj and suppassico pool by plate out and scrubbing action of the
-9..s,a tie 4 vater. The remainder of the halogens and all of the noble gases
%.d9 -
are "available" for leakage to the reactor building at a rate of 2,4f31 uby .
1 g
- 0 5 percent per day at design pressure. As the pressure in the .
dryvell and suppression chamber decreases, the leak rate :
1 U5:c.
%p correspondingly decreases.
y:jf
,w,e j
'Q.
_ ]
~
- . J
'I 4'd 'e - I m.. y '+ , s .... phare through filters for removal of particulate and halogens at
@ a volume flow rate equivalent to 100 percent af the reactor build-
,j$b -
q .') ing air volume in 2h hours. The particulate and iodine removal i ra; ic, -
J 4 ,, filters are assumed to be 95 percent effective. As has been 6 4
- ,J,p
- stated, provisions vill be made in the design for verification"
.s of the effectiveness of these filters on a periodic basis.
(5) The applicant's evaluation of the consequences of this accident l l
1 q assumed release to the atmosphere through aastack 300 feet high, l although at the present time no specifications have been proposed t \
, for the stack design. .
I K@ .
C Aesuming that the vind direction and velocity were constant during
[e[i.Ci the course of' the accident, the applicant calculated exposures as follows:
e 1. For good meteorological diffusion (lapse) conditions and a vind t
) -
speed of 10 miles per hour, the maximum exposure rate at ground level would occur at a distance of ,approximately 0,6 miles from the stack. The
. maximum potential dose rate to the thyroid is approximately 8 millirems per hour (0.008 rems per hour), and the total potential dose for the duration
.,$ of the release to the atmosphere is approximately 1.5 rems. The maximum l d
,g potential whole body dose rate due to noble gases is approximately 2 millirem
?!
ap
, f;h
,, per hour, with a total potential dose for .the duration of the accident of "dC
,G( approximately 0.02h rem.
d5 2.
J.g{g For moderate inversion conditions with a vind speed of 5 miles
'9 E,ya
- per hour, the applicant estimated the maximum exposure rates at ground i;p kIf level vould occur approximately 3 miles from the site. Under these
, ytb
^
conditions the maximum potential dose rate to the thyorid and total dose
.6 I ,, q %'. I9 - ie
% e 8m 4 -h ,
,*J h h - 4 \ WR ' / , a p
u'lDgA'*iE@ ,7NJ M.@QW'M"!t9M7mt/A"M/MMMETM.K3MN1RT1"rm* MEN 1
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1 --
.- . . .- ; .e ...--.w. . . . - .. . . , , i gn
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- 7
- h6 - ,
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.4 for the duration.of release is lesasthan kJ millirems per hour (0.0k rems per- !
ri:
- hour) and 7 rems, respectively. T1.e whole body potential dose rate and M
QM integrated dose is less than 10 millirems per hour (0.01 rems per hour) and W.3 L s one rom, nspectivedy. ,
?
..* As previously stated in the section of this report describing the I meteorology of the site, .we believe that vind speeds during inversica con-
-Q:(.j ..
hx1 -
ditions may be somewhat lower than assumed by the applicant in this evaluation.
WtM .
Accordingly, we have made calculations based upon the above assumptions whh:h /
> y. -
jg take into account the possibility for the accident to occur at a wind speed
^j.q of one meter per second (2.2 miles per hour), under either lapse or inversion conditions. Using an effective stack height of 300 feet and the one meter q
d/ per second wind speed, we estimate..that the maximum potential whole body
' &y '
Q xWn and thyroid dosages for the duration of the accident would be 0 9 mm N$
y.4 and 53 rem, respectively. '
- s-4 The staff has considered both the applicant's assumptions concerning 5
the postulated maximum credible accident and the general concept of the I safety features proposed for mitigating the consequences of such an accident.
As indicated in the above analysis, the amount of fission product released
- .g from the fuel depends on the extent of core damage, which in tum, depends
.96%
jfj, cn the effectiveness of the emergency core spray. In our opinion, a t rw ,
f:{
suitable core spray design which would provide for an adequate supply of M
D emergency cooling water would substantially reduce the extent of damage to
!/6d l
.- the fuel, even to the point of preventing any melting. On the other hand,
,- p , .t
]
y3 < even if the emergency core spray failed to function at all, the inventory ,
4 ,
j of fission products released from the core would be increased by approximately
,f 3, .
a factor ef two. In this case the dosage values given above would be increased y to approximately 2 mm whole body, and to 100 rem to the thyroid.
&b
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Ad From the above analyses, we have, concluded that the engineered safety J*~ features proposed for this facility sh>uldt be capable of significantly limit-
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of public hazard incident to operatien of the Bodega reactor is an acceptable J .
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?%. VIII. Technical Qualifications
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, ~9 31 The technical qualifications of PG&E an described in the application for a construction permit. PG&E has constructed and is now operating a
., boiling water nuclear powe'r plant at Humboldt Bay near Eureka, California.
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. .ge; Their principal cetractor for the Bodega construction, the General Electric
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Company, designed sad furnished the major compments of the Humboldt nuclear f}qlg, ..
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..-pm.a 33 mentations. GE has also designed and furnished similar equipment for _
.,-), several.other boiling water reactors in this country and abroad. On the basis of their demmstrated ability in similar endeavora, we have concluded
~m that PG&E and their principal contractor are technically qualified to con-J. . .. ! struct the proposed facility.
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