ML20235A693
Text
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_Y 1.'
AUG 7 1963 l
t Files sl.
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J. Bowell, Chief, Site.-
Environmental Branch, BIAR L
j USGS REPORT ON BODEGA R&Y 50** S On this date I received the fe11 suing information from 3
Mr. Al Clebach, USGS.
1.
k. Baker has indicated to Mr. Cimbseh that the
'j Department af Interior Beientific Adviser agreed to forsard the USGS Geologie nport on Bodega Bay I
to Seen tary Udall en assarames that a report on seismology of Dodega Esad will be prepared by Wes.
2.
Br. ML111am Bluest, who is a geophysicist in the Geologie i
BLvision of USGS (theoretical Geophysical Braneh) has i
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rossived the easignment for preparation of this report.
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Dr. DLeest moeived his doctorate in geophysics from I
servard. Prior to employment with the USGS be worked j
as an azploration geophysicist for an oil eospany, and his experience with the USGS has largely been in this j
general field, although in addition he was associated age with a program involving a study of esismic signals y
resulting from madergm und analaar m ' M ons, boot R. Iowenstein Sqs E. G. Case
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R. Bryan G. Hadimek l
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DRAFT 2 Date: August 5, 1963 HAZARD 6 ANALYSIS by the DIVISION OF LICENSING AND REGULATION in the matter of r*
PACIFIC GAS AND ELECTRIC COMPANY 2
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BODEGA BAY ATOMIC PARK 1
~f UNIT NUMBER 1 A
- CONSTRUCTION PERMIT c.1
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DOCKET 50-205
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TABLE OF CONTENTS
.. 3; I.
Introduction 1
II. Background III. Important Safety Consideraticos A.
Site and Environmental Factors j
i 1.
Plant Location
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2.
Site Meteorology
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3.
Marine Environment
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B.
Containment Design g:)
C.
Nuclear Design Features
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Reactor Core and Fuel Elements
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'a 2.
Reactor Centrol a
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Centrol and Safety Instrumentation j
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Primary Coolant System j
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D.
Emergency and Safety System Design O
4 1.
Power Supply iq e
2.
Emergency Cooling System l7' E.
Radiatim Monitorink Systems Design j
4' F.
Waste Treatment, Storage and Discharge Design Features l.
Radioactive Liquid Wastes u _._;
2.
Radioactive Solid Wastes 3.
Radioactive Gaseous Wastes IV.
Seismic..C msiderations V.
Research sad Development Program VI.
Analyses of Potential Accidents
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-i VII. Maximum Credible Accident Evaluation VIII. Technical Qualificati me IX.
Reports of Advisory Conunittee m Reactor Safeguards -
1 X.
Susanary and Cmelusims l
E.
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9 I.
Introduction he Pacific Gas & Electric Company (PG6E) has proposed to constmet and operate a nuclear power plant on Bodega Head in Sonoma County, California.
PG4E will design and supervise construction of the unit, and the General Electric Company (CE) will furnish the nuclear steam supply system and the turbine generator.
He proposed plant, designated by PG6E as Bodega Bay Atomic Park Unit Number 1, will produce nuclear energy at the rate of 1,008 megawatts (Me).
H The gross electrical generating capacity will be approximately 325 4t.
he Bodega plant is similar in many respects to boiling water power reactors now in operation.
In general, its detailed design will be based
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21 on operational experience with the Vallecitos Boiling Water Reactor and the Dresden, Consumers, and Humboldt Bay reactors. Dere are some features of
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the plant which require research or developmental effort in order to provide j
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engineering information necessary for their detailed design. These featu ns 4
are discussed in Section IV of this report.
Except for a discussion of the effects of seismological and geological factors and plant safety, this report contains an evaluation of the significant 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 affect the probability of or consequences of accidents of safety significance j
to the general public. he effects on public safety of nomal routine opera-tion of the plant, including the discharge of radioactive materials are also considered in this report.
II.
Background
On December 28,1962, PG6E submitted an application to the AEC for a l
construction permit and operating license pursuant to Title 10, Chapter 1, 1
Code of Federal Regulations, Part 50 (10 CFR 50), no application, which includes a " Preliminary Hazards Summary' Report", dated December 28, 1962,.
and Amendments 1, 2 and 3 to the application dated March 16,. April 5, and June 131953, respectively, has been reviewed by the staff of the Division of Licensing and Regulation. Technical consultants in specialized areas also advised the AEC regulatory staff. The application has also been considered by the AEC's Advisory Committee on Reactor Safeguards (ACRS), as required by the Atomic r
Energy Act and the regulations of the AEC. De recommendations of the ACRS, 4
as expressed in its report of April 18,1963 (a copy of which is attached hereto as Appendix C) were also considered in the regulatory staff's evaluation.
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's As is customarily the case in reactor facilities prior to the commencement
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of construction, there are a number of features of plant design and operation-a which have not, as yet, been definitely reso'1ved. We Commiss' ion's regulations provide for the issuance of a construction permit on a provisional basis in i
cases such as this, in which aspects of the detailed design have not been completed. A provisional constmetion permit may be issued, according to j
Section 50.35, 10 CFR,on the basis of findings, among others, that '(1) the applicant has described the proposed design of the facility, including,'but not limited to, the principal architectural and engineering criteria for the design, and has identified the major features or components on which further technical information is required; (2) the omitted technical info-mation will be supplied; (3) the applicant has proposed, and there will be conducted, a research and development program reasonably designed to resolve the safety questions, if any, with respect to those features or components
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l which require research and development; and that (4) on the basis of the foregoing, there is reasonable assurance that (i) such safety questions wi1L-be satisfactorily resolved at or before the latest date stated in the I
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3 application for completion of constructim of the proposed facility and (ii)
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taking into consideration the site criteria contained in Part 100, the proposed i
i facility can be constructed and operated at the proposed location without undue l
risk to the health and safety of the public.
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ne proposed constmetion permit, if granted, would authorize construc-
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tion only. De Commission would require timely reports from PG6E with respect to results of research and development and final design of the more significant fI i
design features. De AEC staff would continue its evaluation of the safety of the plant in light of this information. An operating license would not l l J
be issued until the final design had been completed and evaluated by the AEC j;j staff and the ACRS.
In addition, the plans and procedures for operations would be evaluated by these two groups.
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Pursuant to a Notice of Hearing published
, the h
issuance of a provisional construction permit to PG6E will be considered at
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a public hearing to be held in the hearing room of the Board of Supervisors N1 l of Sonoma County, Santa Rosa, California, 'at 10:00 a.m., PDT, on
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1963 before an Atomic 1.icensing and Safety Board appointed by the AEC. The
[e issues to be considered at the hearing are:
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1.
Whether the applicant has submitted sufficient information to provide i
reasonable assurance that a facility of the general type proposed Ts the application can be constmeted and operated at the proposed
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w andue risk to the health and safety of the public; 2.
Whether there is reasonable assurance that the technical information omitted from and required to complete the application will be supplied; 3.
Whether the applicant is technically qualified to design and construct the proposed facility; and l
4 Whether the issuance of an authorization for the construction of the facility will be inimical to the common defense and security or to the health and safety of the public.
The proposed plant to be constructed on Bodega Head would be subject to potentially severe shocks from earthquakes. There is also a possibility that earthquakes in the vicinity might cause faulting beneath the plant which could cause severe damage to the facility.
The possible effects of such seismic activity on the proposed plant are still under study by the Commission's regulatory staff. The staff has not yet determined whether or not a plant can C
be constructed and operated safely at this location due to these considerations.
4 At this time further information which must be based on exploration of the i
a site, among other things, must be obtained.
Further consideration must also e
i d be given to criteria which must be applied in the design of systems or com-y ponents of the facility which are of importance to safety, especially those 4
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systems which nust be relied upon in an emergency, such as a severe earthquake, to prevent undue hazard to public health and safety.
This analysis and the conclusions contained in this report, therefore, I
are made without regard to the special safety considerations which must also i
f be taken into account in view of the seismicity of the proposed site.
The staff's evaluation of the proposed Bodega nuclear power plant i
l described in subsequent sections of this report and its position on the issues t
at the fortheeming hearing are based on all the technical information submitted as part of the applicant's request for a construction permit and the report l
from the ACRS. All of this information is available for inspection and review at the Commission's Public Document Room in Washington, D. C., and at the Coninission's San Francisco Operations Office, 2111 Bancroft Way, Berkeley, Califomia. This evaluation and proposed recommendation is subject to j
modification in the light of any further information which may become avail-4 able, including the evidence introduced at the hearing. The decision of I
the Commission will be based upon the entire record in the proceeding.
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III. Tmoorthot SafhyiConsiderattymgp3d ne' Bodega reactor is a direct cycle, forced cirdblation' boiling water -
l 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 circulated through the' reactor. Steam generated in the. reactor at 1,075 ~
psis flows to a turbine generator with a gross electrical generating capacity I
of about 325 megawatts.
Reactor coolant which has been separated from the-s steam is recirculated through four loops each containing a' pump rated at 4
29,000 gpe. After passing through the turbine the steam is condensed, and the.
condensate after demineralization is returned to the reactor vessel. his a
water, which will contain some radioactive materials, will be circulated
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i5 within a closed system from which the only normal effluent will be a continuous
.3 discharge of noncondensible gases. This gaseous material will be monitored.
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continuously and released from the reactor stack if the contained radio.
activity is below permissible limits. As would be the case in a conventional
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power plant, the condenser will be cooled by water drawn from Bodega Bay and
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discharged into the Pacific Ocean.
From time to time, regulated and measured
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1 quantities of radioactive liquids will be mixed with the condenser coolant
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water from this facility and discharged to the ocean.
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An overall judgment concerning the safety of operation of the Bodega y
reactor or the acceptability of potential hazards must be based upon a 1
number of individual safety considerations. Our judgments at this time are La based upon an evaluation of the design details, design criteria and design concepts described in the PG4E application, the known facts conceming the
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proposed site and its environment, and an analysis of the effects on public health and safety of nomal 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-mission staff is based upon the principles of design rather than upon details of design themselves.
In the case of features in this category which are of particular importance to safety, the staff proposes to require and expects to receive information on final design of these features before PG6E has expended any substantial amount of effort in the construction of those features.
De general objective of nuclear safety is to prevent or limit to an acceptable degree the exposure of persons to ionizing radiation.
It is proposed that the following general features be provided to achieve this objective:
(1) ne first and principal safeguard is found in the design features of the plant which contributes to the confinement of all radio-active materials to their proper place. Confinement of fission f$
products and the products of neutron activation is attained by the three-fold containment afforded by the fuel elements, the primary coolant system, and the containment building. The isolation of
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the plant site provides a fourth measure of protection of the public g
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against the radiation from materials routinely or accidentally
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The design of the fuel elements and provisions for reactor control and cooling have the purpose, from a safety standpoint, of retaining fission products within the fuel elements, in which location the shielding of their radiation is most easily effective.
The primary coolant system, consisting of various vessels and types, serves to retain any fissior, products which might be 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
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7 the primary coolant.
The reactor vessel, the recirculation pumps, piping, and other portions of the primary coolant system are located within a structure which is capable of safely containing the radioactive contents of the plant in the event of an unlikely major accident which might cause a rapid and uncontrolled release of fission products from the fuel
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(2)
Radioactive fluids and solid waste are contro11ed' and processed in
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l a manner which confines these potentially hazardous materials to g
systems which are designed to permit their safe preparation for disposal.
Paths through which radioactive materials are or might L,
4 be discharged are continuously monitored by radiation detectors.
Storage, processing, sampling, and monitoring are employed to assure that quantities of radioactive materials released to the q
environment will be within limitations established by AEC regulations.
(3) Shielding and area radiation monitoring are used to continually assure the safety of workers from radiological hazards. Such f
monitors also serve as an additional safeguard against the release of radioactive materials to the public.
i (4) Emergency systems are designed to prevent or reduce the hazards to plant personnel and the general public should accidents occur.
The following sections of this report discuss in greater detail the L
more important safety considerations which have led to the staff's conclusions with respect to the safety of operation of the proposed plant.
1 A.
Site 'and Environmental Factors The site proposed for the Bodega Bay reactor is located on Bodega Head, a small peninsula along the Pacific Coast in Sonoma County, Califomia, approximately 50 miles northwest of San Francisco. Bodega Head is bounded k__.-______
on the east by Bodega Harbor, by Bodega Bay on the south and the Pacific Ocean on the west. A sand spit known as Doran Beach or Doran Park extends from the mainland towards Bodega Head and forms a natural breakwater for Bodege Harbor.
The environmental factors which are significant with respect to the safety (or this site and which have been examined in detail includet (1) the location with respect to the nearby population, (2) the meteorological l
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factors, and (3) the marine environmental factors. As noted previously, L
geological and seismological factors are significant to safety and will be N
the subject of further consideration by the regulatory staff.
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Plant Location
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ne proposed location for the reactor plant is on a 225 acre tract
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of land at Campbell Cove near the southern end of Bodega Head. The property owned by the applicant includes the entire southern end of Bodega Head. He I
adjacent property to the north is under acquisition by the University of California for a research facility, j
ne reactor would be located on the east side of Bodega Head near I
Campbell Covekand across the entrance channel to Bodega Harbor from Doran f -
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Park. The nearest edge of Doran Park, which is owned by Sonoma County and l
contains no residences, lies approximately 1,300 feet east of the proposed
- reactor, y
he traffic through the entrance channel to Bodega Harbor consists primarily of commercial and sports fishing boats. Usage of both the Channel and Doran Pare ~ could be controlled under emergency conditions if this should become necessary for protection of the public.
Accordingly, the exclusion distance for this site can be considered to be the distance to the northern site boundary, which is a minimum of approximately 2,700 feet (0.5 miles).
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1 ne population data for this area based on the 1960 census shows no popu-4 lation conter larger than about 200 within 10 miles of the site, and none larger than about 3,000 within 20 miles of the site. The nearest cities of more than 10,000 are Santa Rosa (31,037) and Petaluna (14,035) which are 21 and 24 miles, respectively,from the site. The population density withnt 25 miles is as follows:
Persons / Square Mile Total Population Distance of Land Area in Area 4
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14 - 5 21 500 q
W 5 - 10 16 1,600 f
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10 - 15 81 15,700 3
.. J l is - 20 97 30,000 y
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20 - 25 180 66,700 4 ' ',
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The above tabulation shows that the proposed reactor site is favorably 4*
located from a safety standpoint with respect to population distribution a
and density. Not only is the immediate area sparsely populated at the present i
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time, but the location on a peninsula provides natural barriers against the 3
future development of housing within at least two miles of the reactor plant.
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Re data also shows that the population density is quite low essentially out to the distance of Petaluma and Santa Rosa, and these cities account M
for about 75 percent of the population in the annular area that lies between 20 and 25 miles from the reactor site. Accordingly, the actual " low popu-lation distance" and the " population center distance" as defined in the Consnission's site criteria regulation for the Bodega site can be considered to be 24 miles, which is the distance to Santa Rosa.
It is shown in a later section of this report that the maximum required " low population" and
" population center" distances are miles and miles, respectively.
2.
Site Meteorology In general, climatology of the coastal area at Bodega Bay is typical l
of the central to northern coastal area of Califomia and is characterized by a wet season extending from about November through March and a dry season i
from about April through October. The topography of the area inland from the site is characterized by a series of hills and valleys, with the hills rising h
to elevations varying from approximately 400 feet to approximately 1000 feet.
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.1 The roughness of this terrain would be expected to enhance the charac-
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- 4 teristic atmospheric turbulence of this area of Califomia and result in g
rapid dilution of gaseous effluents that might be transported from the site d
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to inland locations. On the other hand under strong, inversion conditions, j
I the range of hills along the coast, which is approximately three miles
'l from the proposed site, would tend to restrict the transport of airbome '
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e materials to the inland areas.
Detailed meteorological information for the proposed reactor location 7,
is not presently available. However, observations of wind directions and i'
velocities of coastal locations at Point Reyes, approximately 22 miles south Q
I of Bodega Head, and Jenner, approximately 12 miles north of the Head, shows that the wind blows towards the coastline approximately 60 percent of the time, with the prevailing wind direction from the northwest. This information 4.--
also indicates that the remainder of the time the wind blows either offshore or generally parallel to the coastline.
Based on..the meteorological information presently available the applicant l
l has proposed the following diffusion parameters for use in Sutton equations for estimating the amount of atmospheric dilution at the site under various meteorological conditions:
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Meteorological Condition Strong Moderate Inversion Lapse Parameter
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.5
.3
.22 C
.2
.21 6
y Cg 02
.07
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A (miles per hr.)
5 5
10 l
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Our evaluation has disc 1'osed no unusual meteorological conditions at this site and we believe that the parameters proposed by PG6E provide adequate I_
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conservatism for the estimation of atmospheric dilution at Bodega Head, except f
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with respect to the wind speeds proposed for the inversion conditions.
In
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our opinion, a wind speed of one meter per second, which corresponds to about 2.2 miles per hour, would be more representative of this coastal location
- M during periods of slow diffusion. Table I, Appendix III of the applicant's 1
Preliminary Hazards Summary Report, which shows a relatively high frequency
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of the lower wind speeds in the zero to three miles per hour range, tends to
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confirm our judgment in this respect.
3 The meteorological factors, as discussed here and modified above, have been used in our evaluation of the potential consequences of the maximum,
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credible accident which is discussed in Section VI of this report.
PG6E has constructed a meteorology station with a tower approximately LJ 250 feet in height which will be used to collect meteorological data during reactor construction, such as the frequency of wind speeds and directions of various atmospheric and stabilizing conditions which would be appropriate for l
use at the proposed site. We believe that these facilities should provide sufficient data for determining prior to reactor operation the capacity of the atmosphere to safely dilute radioactive gases that might be released from the facility.
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3.
The Marine Environment i
Approximately 250,000 gpm of cooling water for the plant condenser will be withdrawn from the Pacific Ocean at Campbell Cove on the east side of Bodega Head and discharged to the ocean on the west side of the Head with an estimated maximum temperature rise of 18'F.
The condenser cooling water does not pass through the reactor and will not become radioactive from exposure to Tl neutron irradiation. This cooling water, however, will contain liquid efflu-j i
l ents to be periodically released after monitoring from the rad waste facility.
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9 The concentration of radioactivity in the condenser cooling water before q
disch,arge to the ocean would be controlled to meet the requirements of 10 CFR Q
Part 20 of the Commission's Regulations; l
At the request of the AEC, the U. S. Bureau of Connercial Fisheries of e
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l the Department of the Interior has reviewed the effects of reactor operations 1
on the marine environment of Bodega Head.
The Bureau's report, which is
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- i l attached as Appendix II, indicated that there could be potentially signifi-j cant effects from the discharge of this effluent to the ocean, such as:
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Possible concentration of radioactivity in seafood, g
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Possible concentration of radioactivity along beaches used for t
s public recreation.
However, the Bureau of Fisheries stated that it is well established that certain levels of radioactive wastes can be discharged into the oceans L,
without adverse effects on fish and wildlife, since circulation insures mixing of radioactivity with large volumes of water, quickly diluting and dispersing thE radioactivity. Their report further stated that since the permissible levels and rates of discharge of radioactivity are difficult to determine in advance for any specific area, an extensive monitoring program to insure that concentrations cf radioactivity in the marine life do not exceed predetermined levels should be undertaken.
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The experience at Windscale, England, where radioactive wastes have been discharged to coastal waters for seven) years was cited by the Bureau of Fisheries as an example that shows no adverse effects on fish or shellfish from rea:: tor operations. The English have determined on basis of a monitoring program carried cut over several years that 3,000 curies per month of strontium b
90 could be safely released to the coastal waters at Windscale.
I PG6E has ispecified that the concentration of radioactivity in the condenser cooling water discharge will not exceed that permitted by Part 20 of the Com-
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mission's Regulations. Based onia discharge rate of 250,000 gallons per minute we have estimated that approximately 4.1 curies of strontium 90 per
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month could be discharged to the Pacific Ocean, if all of the radioactivity J
in the condenser cooling water consisted of this nuclide. This value is extzteely conservative in comparison with permissible release values deter-h mined by the English for Windscale, f~
Extensive studies of oceanography and marine biology to evaluate the
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marine biological aspects of the proposed reactor operation similar to those l
8 recommended by the Bureau of Commercial Fisheries have been initiated by the I'
applicant. These studies are as follows:
1.
An oceanographic survey will be carried out to determine the circulation pattem of the ocean in the vicinity of the outfall, l
and the capacity of the ocean to dilute the condenser cooling water l
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discharge.
2.
An ecological survey which will include an inventory of the marine organisms in the vicinity of the outfall will be carried out.
3.
An environmental survey of the marine en-Lronment will be conducted to determine existing levels of radioactivity before the reactor becomes operational. This program would continue after the reactor is in operation to determine any tendencies for reconcentration by 1
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the marine environment of radioactivity released to the ocean from operation of the plant.
1 We have considered the information submitted by the applicant and the l
conments of the U. S. Bureau of Commercial Fisheries and have concluded that j
the liquid effluents produced by operation of the proposed reactor can be l
disposed as proposed to the Pacific Ocean without exceeding Part 20 limits.
In view of this, we believe it is extremely unlikely that any adverse effects pIi l d
cn the marine environment will result. Further, it is our opinion that the 9
programs proposed by the applicant for maintaining vigilance over the marine r
environment will be adequate to observe any anomalies that may occur despite
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- g these precautions so that these could be corrected well before any safety g
l problem develops.
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4 B.
Containment Design ne containment system proposed for the Bodega reactor will serve as a principal safeguard against release of radioactive materials to the environment in the unlikely event of major rupture of the primary system inside the contain-l ment structure. As in other nuclear power reactors,' the containment system.in this facility is not provided to protect against accidents that are expected or j'
likely to occur; rather it is provided to safeguard operating personnel and p
the general public in the event that the best efforts to design the plant to the highest standards, and to construct, maintain, and operate the plant com-
.1 petently and safely should fail.
In. addition, other emergency systems, such as the core spray system, are generally provided, as in this case, to either prevent or retard the release of fission products under various accidental j
3 conditions.3 Rese various emergency systems operate independently of the M
containment system to prevent the release of radioactive materials beyond the boundaries of the plant. They are discussed in a later section of this
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report. Re adequacy of these systems and the containment system in particular a
is an important safety consideration.
Re proposed design of the containment system as discussed in this
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section has been used in the analysis of the maximum credible accident, which is discussed in Section 6 of this report.
That analysis is the basis for establishing important design parameters such as drywell and suppression L
chamber design pressures, which are later described.
The containment system proposed for this facility is one which utilizes the pressure suppression concept.
Its design is similar in many respects to that used at Humboldt Bay reactor facility.
Significant features of the Bodega Bay plant containment design include the following:
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(1) Plans for the Bodega Bay plant call for a drywell having a 60 ft..
I diameter spherical lower section and a 26 ft. diameter cylindrical upper. section. The overall height of the ~ drywell.is approximately:
100 ft.
1 (2) The reactor vessel and four reactor recirculation loops,- each with a pump, will be located within the drywell.
i (3) n e drywell will have an airlock entrance.
Personnel' entry is not t.
planned during nactor operation, but is contemplated with the
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reactor coolant system hot and pressurized.
J (4) ne suppression chamber will be in the form cf a torus and will H..
3 have a major diameter of 93 ft. and a cross-section diameter of
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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.
Both the drywell and the suppression chamber will.be designed and con-oy structed in accordance with the ASE Boiler and Pressure Vessel Code,' Sectien
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Piping restraints will be provided at containment penetrations to b -
1l assure that failure of the' pipe will not cause containment rupture. A concrete
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refueling building will be constructed above the drywell and suppression chamber.
Pressure and leak rate specifications for these containment system-
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components are as follows:
A L3 Leak Rate 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 />)-
Drywell 62 psig 0.5 (at, design pressure)
Suppression chamber 35 psig 0.5 (at design pressure)
Refueling building 12 in. H O 100 (at 1/4 in. H O) 2 2
In order to proof test the Bodega Bay pressure suppression desip, Pacific Gas 6 Electric Company is conducting a test program at its Moss 3
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C Landing Power Plant. As in the Humboldt Bay project, the applicant has con-structed a full-scale segment of the suppression system.
In the test of Bodega Bay design a single 24-inch diameter vent pipe frce the drywell to the suppression i
chamber was used. Since the full size' plant is to have 112 of these vent pipes, the test equipment represents a 1/112 segment of the containment.
Tests wen conducteri with this mock-up to simulate various Ixcident con-(
ditions. A flow comparable to 1/112th of the flow resulting. froin a complete circumfenntial break of one of the 28 in, reactor coolant recirculation 4'
lines (with flow out both sides of the break) was taken as the " maximum credible operating accident" (MCOA). Highest containment pressuns observed in these tests were 52 psig in the drywell and 30 psig in the suppression chambe r.
These pressures were observed when the mock-up drywell was pre-heated to 255'F and when the mock-up reactor vessel water was subcooled Pf 35'F.
Tests at higher and lower drywell temperatures and at higher and y
lower reactor water subcooling yielded lower drywell and suppression chamber pres sures.
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In another test a break area 2.5 times that of the MCOA was simulated.
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In this test the peak drywell pressure observed was 63 psig. Further Moss Landing tests ere being conducted to determine whether baffles are needed in the suppression chamber. We believe that Humboldt Bay experience and the Moss Landing tests will provide an adequate basis for designing the Bodega containment system.
L As another significant containment design feature, Pacific Gas 6 Electric proposes that in a number of instances a single isolation valve will be installed at the containment well in pipes or ducts penetrating the containment. However, each such line will also be provided with a second isolation valve which may be a remotely operable process valve located
tol j f' l,
F' (1
j
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^ i I
elsewhere. The two isolation valves located at the &cywell wall in each main steam line are to close on a manual signal or automatically on the occurrence of any of the following:
i (1)
Low condenser vacuum 7
(2) Main steam line leak (in the pipe tuntiel)
(3)
Low reactor water level 9f PG6E is also givtag consideration to providinc.;rotection of the main steam 8
line isolation valve against foreign matter, which might interfere with 1
,proper valve action.
He Bodega Bay design is such that during refueling, the spent fuel storage pool will connect directly to the shid1d water above the reactor, w
thus permitting direct underwater transfer c!,/ fuel without the need for a transfer cask.
In our opinion this feature proiides in a simple and reliable way for both continuous shielding and cooling of spent fuel during dransfer and storage.
he refueling building in which the drywell and suppression chamber syst'em are located is 1,rovided with a controlled release ventilation system which' discharges to the plant stack. The building and ventilation system
{
i design is *f,uch that the refueling building can be maintained at a negative pressure. Discharge from the building passes through halogen and radioactive L
g particulate cleanup equipment prior to discharge to the stack. PGSE has
}
indicated that, in accordance with the recommendations of the ACRS, the system will be designed to permit frequent testing of the ability to filter
~
particulate and to remove iodine at specified efficiencies.
(The following 2 paragraphs will be revised to reflect the expected
(
Amendment 3.,)
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Se staff believes that the general containment scheme proposed by PG6E is adequate for the proposed reactor facility. However, it is our opinion that the detailed design should reflect the specific considerations discussed i
in the following paragraph which we believe are necessary to assure that the containment as proposed can be reasonably expected to provide the high degree of integrity proposed at any time that it might be called upon to contain the q
l 8,
consequences of a maximum credible accident. These include those recommended j
containment features mentioned explicitly in the ACRS report on this project to the Connaission. The regulatory staff, in reviewing the details of contain-ment design as they are developed, intendd to assure that these considerations
,3 with respect to containment testing, penetration design, and isolation valving
.2 have been satisfied:
1.
The design should permit initial integral leak rate testing of the j
drywell and ' suppression chamber at their respective design pressure after i
the installation of all penetrations (including piping conduits, electrical
.3 a
conductors, and gasketing closures) and subsequent periodic testing at f
l suppression pools design pressure.
In the initial testing, the leakage b_:
rate of the containment system should be determined as a function of pressure l"
l up to full design pressure.
2.
ne design of penetrations should take into account, in addition to the pressure load, the loads or deformations imposed by thermal expansion, impact of missiles, reactions of ruptured pipes, and disturbances incident to installation,' maintenance or repair.
Penetrations should be shielded from missiles to the extent practicable.
Penetrations should be designed so as to allow frequent periodic leakage rate tests of the penetrations only (including points of attachment to the containment shell), at full design pressure. The required frequency of integral high pressure tests l
l
(
of the entire containment system after operation commences will be dependent upon the frequency that individual penetrations are tested at design pressures.
3.
All pipes and conddits which communicate with interior of the primary system or the containment system, and other piping (such as instrumentation and control piping) which cannot be adequately protected against accidental l
ruptures, should contain at least two isolation valves. All valves per-T" forming the function of isolation valves should be provided with protection J.
I against materials in the system which might prevent proper closing and should be provided with reliable automatic and manual actuation features.
Isolation l
' M valving should be designed so as to permit periodic leakage rate tests.
q Appropriate closing times for isolation valves should be determined on the u
basis of analyses of system ruptures which would release fission or activation products outside the drywell while the valves are not fully closed.
U t
In order to assure that these considerations have been satisfied, we 3
3 I<
believe that PG6E should submit for Commission review the results of further developmental tests of the suppression pool concept and final design plans
(.
3 for the containment as soon as they can be made available.
?-
C.
Nuclear Design Features 1
As noted previously, every reasonable effort is made to design l
nuclear power plants such as the Bodega facility such that they can be
~
operated safely even without such extra safeguards as the containment system.
This degree of safety can be assured if the fission products which are produced in the nuclear processes are confined within the fuel in which they are generated.
Since the first leak-tight barrier which normally contains these fission products is the cladding surrounding the fuel elements, special precautions are taken to insure that the cladding remains intact during both normal and abnormal operation.
(
21 -
ne integrity of the fuel cladding barrier can be maintained if it is not subject to a combination of undue thernal or mechanical forces or stresses.
Dermal stresses can be produced by an excessive nuclear reaction rate within the fuel which would cause the fuel to generate heat at a rate exceeding the capacity of the cooling system which ordinarily maintains the temperatures of the fuel cladding within predetermined safe limits. Mechanical stresses J
4 can be caused by forces such as reactor coolant pressure and hydraulic flow.
l o'
To safeguarti a reactor core against damage that would release fission
.] i
-. 3 products from the confinement provided by the fuel element cladding, one y,
E_
must:
-=e (1)
Provide a cladding of suitable material and thickness which will
- [=j U
retain its integrity when subjected to the thermal, mechanical, i
and radiation condition of its environment; (2)
Provide for control of the chain reaction; and j
5 (3)
Provide for removal of the heat generated in the fission process and the radioactive decay of the fission products.
[
8-Generally speaking, the design of the reactor core, including the fuel j
elements, the design of control and instrumentation systems, and the design 1
e of the heat removal system should be such that, in normal operation and under many conceivable accidental conditions, radioactive fission products would be confined to the fuel elements themselves.
1.
Reactor Core and Fuel Elements ne proposed reactor core will be composed of 592 fuel assemblies each of which provides a vertical channel through which the mixture of steam and water passes. The core will have approximately the form of a right 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
(
l assemblics through control rod guide tubes. The fuel assemblies are held in 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 control rod guide tubes which extend to the lower head of the reactor pressure vessel.
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
{
to 2.5% U-235.
Rese fuel pellets will be contained within stainless steel tubing which provides the fuel cladding boundary.
PG6E has tentatively "M,
proposed that this cladding would have a nominal thickness of 0.011 inches and would be able to withstand an exposure of 15,000 WD/ TON.
]
Determination of the maximum nuclear reactor rate or power level of a reactor core of this configuration which is consistent with maintaining the integrity of the fuel element cladding requires a complex analysis of the nuclear characteristics of the core, the thermal dynamic properties of I
the fuel, cladding, and coolant, and the hydraulic properties of the coolant l'
+w and coolant system as well as the mechanical properties of the fuel and the cladding both initial and when subjected to irradiation.-
The power distribution caused by nuclear fissions which is expected f
in the Bodega reactor core has been estimated by the applicant for the purpose of making a preliminary determination of the thermal margins which L
would occur in the hottest fuel assembly.
Rese estimates will be refined by detailed calculations of power distribution in the course of design of the reactor. After operation commences the power distribution will be monitored continuously by a system of in-core flux monitors.
Such methods have been successfully used in other reactors and should provide a reliable means of establishing the thermal margins that are experienced in operation.
l
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l
[
< Thermodynamic and hydraulic facters, which ultimately determine the i
permissible power level of the reactor, have only been briefly described by PG6E. While at present there does not appear to be a firm basis for establish-ing thermal limits as high as is suggested by some of the parameters specified in the application, PGSE has established a criterion for determining the proper detailed thermal and hydraulic design factors for the Bodega reactor core; T7 namely, that the fuel will operate without loss of cladding integrity over f
the design exposure period at the maximum heat fluxes possible within burn-l 1
out limitations.
Operating experience at other boiling water reactors has 1
indicated that this criterion can be met and that it is acceptable from a "1
22g safety standpoint.
2
'A The thickness of fuel element cladding tentatively proposed is 11 mils.
.s Cladding thickness for the bulk of the Humboldt Bay fuel elements is gj mils.
On the basis of present information and operating experience, one k
cannot be assured that fuel with cladding of this type and the smaller
~
I thickness can be irradiated for the exposures stated without experiencing I ]~
...Vn excessive pin-hole leaks in the cladding. However, the design of fuel I,
p.,
for the Bodega reactor will not be completed until further data from a
[_;
i General Electric research and development program are available.
In our i
opinion, this program is reasonably designed to provide an engineering basis for a suitable fuel element design.
In any event, extensive experience L
with other power reactors provides reasonable assurance that a fuel element cladding design suitable from a safety standpoint can be developed for the Bodega reactor.
l 2.
Reactor Control 1
Nuclear safety requires 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
('
changes. When reactivity is positive, the ch'ain reaction grows; that is,' the rate at which nuclear energy is relbased bk fission is increased. Conversely, when reactivity is negative,the chain diminishes and power falls. The opera-ting condition of the reactor, its temperature, pressure, power level, void content, and fuel exposure all affect reactivity.
In our opinion, the general way in which each of these factors affects reactivity in a reactor i
of this type is quite well known, and the theoretical and experimental methods for investigating reactivity effects are sufficiently developed to "1
permit design of the Bodega reactor control system to proceed with confidence.
The stability of a reactor is affected by operating conditions which have transitory effects on the neutron economy of the chain reaction.
h]
It is expected that, with the possible exception of the volume fraction of steam within the reactor coolant, this reactor should be stable within p
p' the range of operating conditions contemplated. That is, an increase in
['
reactor power will cause changes in operating conditions which will have a strong natural tendency to decrease reactivity and limit the power increases.
(
With respect to steam volume fractions, preliminary calculations indicate I
that the following would result at rated operating conditions at Bodega; i
I average core voids - 37%
f average exit voids - 58%
Although we are not aware of any substantial operating experience that
.i would confirm the stability of reactor operation at steam volume fractions this high, PG6E has stated in the application that analog computer studies being made as jiart of the research and development program will show that the plant can be designed to exhibit satisfactory dynamic performance with such high void content.
In any event, we believe that with appropriate limitations the high void conditions proposed can be safely approached in t
i
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m_._
nactor tests devised so as to determine the proper. range of void content' for stable nactor operation. De main purposes of the Bodega reactor control system, thenfore, will be to provide 'a means of precisely and reliably adjusting reactivity.
Necessary safety objectives for the Bodega control system are to poride means for (1) shutting the reactor down by a safe margin under any circumstance, (2)' starting the reactor and increasing power at a safe rate, f
and (3) maintaining the reactor power level within the capabilities of the i
heat removal system.
The control system proposed for this. reactor is an ' array of 145 moveable control blades, which have suffiele'nt" reactivity worth to keep the reactor safely shut down, even though one of these blades (or rods) might be 4
stuck out of the reactor co n.
The k-effective of the reactor with all.
control blades in the react 6r core is calculated to be 0.97.
In our opinion, j
the strength of the controls is suitable for safely shutting down the y
reactor under any credible circumstance.
1 Re material in the moveable blades will be boron carbide contained within 0.175-in. O.D. stainless tubes similar to those presently in use at jp.
Dresden. Additional control devices, removable only through.a loading j
I precedure, are provided by fast control curtains of 0.1% boron stainless steel, which will be semi-permanently located between selected fuel
);
elements. De fixed control devices provide a flexibility in the reactor..
design so that reactivity of the core can be easily adjusted to attain the shutdown margin required from safety considerations.
The reactor design also incorporates a liquid poison system that can be used to inject sodium pentaborate 'into the core in the unlikely event complete shutdown cannot be achieved by use of the control rods._. Similar systems are provided in boiling water power reactors now in operation.
--w-.
(.
ne hydraulic control rod drives to be used in the Bodega Bay plant are to be designed using the same b'asic concepts as have been employed in drives in use in boiling water reactor plants at Dresden, Big Rock Point, and Humboldt Bay. Water used as the hydraulic fluid can be applied to either side of a piston which is mechanically coupled to the control rod, thus providing for 4
sitter upward or downward rod motion. Only one rod can be moved outward 4
(increasing reactivity) at a time, and it may be moved either continuously p];
or in 6 inch steps.
Rod speed is controlled by orifices which regulate the flow of water away from the low pressure side of the piston. De rod speed 3
for normal withdrawal will be controlled so that at the maximum rate of 4
4 withdrawal the reactor power would not increase at a rate fast enough to T
i i
lead to serious consequences.
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All rods can be inserted simultaneously, shutting the reactor down.
N k
Rods are scrammed upward by applying pressurized water from either the j,_.
,(
reactor or from accumulators to the bottom side of the drive pistons and pi 2
simultaneously relieving the volume above the top side of the pistons to the k
scram dump tank. De drive is locked in fixed positions by collet fingers 8
y i ?
which engage grooves spread at 6-inch intervals along the moveable index yJ'
~
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 important part of our evaluation of these drives is based on previous experience with these drives, nis includes Dresden experience as well as initial Big Rock Point operations.
Detailed design of drives for the Bodega reactor has not been made.
General Electric is considering modifications of designs already in opera-tion in other reactors that will minimize the possibility of foreign
(.
material accumulating in the rod drives and impairing operability. The i
applicant has also indicated that functional'and endurance tests will be made on the prototype Bodega mechanisms, but the detailed procedures for these tests and the acceptability criteria have not as yet been proposed.
Control systems which are designed to react rapidly to demands for shutting a reactor down generally have some potential for accidentally I
increasing reactivity as well. This aspect of the PG6E control system design is discussed later in this report (Section V), where consequences of a rod dropout are considered. PG6E has indicated that devices for
.s limiting individual rod worth and for impeding the fall of a rod are under 3
development. Such devices could enhance the safety of operation and simplify the procedures that are presently used with similar drives to provide the necessary assurance that a rod dropout accident cannot cause a serious public hazard.
f We believe that the design criteria for the reactor control system is acceptable.
In view of the importance to safety of the detailed design of the control rod drive system, the staff believes that pC4E should submit r
timely reports for review by the Commission during facility construction on l-1 1
development, design, and testing of this system.
1i 3.
Control and Safety Instrumentation
+
The instrumentation necessary for operating the control system L--
and the reactor safely in a nuclear power plant generally involves a large number of sensors throughout the various process systems. These sensors measure a variety of variables, such as neutron flux and gamma radiation levels, and temperatures and pressures of various fluids.
Information col-1ected by the ceasuring instruments is used to guide the operating staff in l
controlling the plant and to actuate automatic control devices.
(,
circuits, and control devices which are of most importance The instrue".
o, to public health and safety are:
(a) thcr t necessary for and contributing to stable reactor operation, (b) those used in control of radioactive fluids and effluents, and (c) those used for control of emergency equipment.
A generally adequate description of proposed instrumentation is presented in the application. Sufficient details are not yet available to determine T~
whether instrumentation provisions have been made for all essential functions or to determine the degree of reliability that should be attributed to the
~i reactor protection system, which is described by PG6E as " fail-safel. '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 f$^
provide an acceptable design. The staff intends to evaluate the reactor control and safety instrumentation in detail prior to reactor operation h
1 in order to assure that proper' attention has been given to the need for gj s
automatic functions and the reliability of safety instrumentation.
u 4.
Prinney Coolan* System x
g Water is circulated through the reactor core by this system in order to remove the heat generated by the nuclear fuel. The heat changen fi the coolant to steam which is converted to mechanical energy in, the turbine i
in order to produce electricity.
The functional integrity of the primary system is necessary to the
..~
integrity of the fuel element cladding since heat must continually be removed during reactor operations in order to preserve the integrity of the cladding barrier.
If this objective is met, then the nrimary coolnnt systen also serves as an essenti,11-1;r a r i s'e nen de a t.o r-i e r, in addition to the fuel element cladding to retain radioactive fission products within the facility. The principal components of this system are the reactor
( pressure vessel, the primary coolant piping, pumps and isolation valves.
As proposed by PG6E, the reactor core vill be located in a reactor 4
pressure vessel designed, built and tested in accordance witir Feetion VIII of the Boiler and Pressure Vessel Code of the American Society of Nechanical Engineers. The 50 ft, high by 15 ft. diameter versel will be constructed of carbon steel approximately 6 inches thick. The interior of the vessel will be clad with stainless steel applied by welds overlay methods. The design pressure will be 1235 psig at 575'F.
l Steam generated in the reactor will be passed through axial flow j
a steam separators and driers located and then inside the reactor vessel j
r through two 20-inch steam lines to the turbine, which is located in a g
separate building. Water, after separation from steam in the reactor vessel,
]
passes from the reactor vessel through four 28-inch pipes.
Four recircu-4 1ation pumps, one in each loop, provide the driving force for circulating l
,1 water through the reactor core.
Feed water from the condenser is injected into the reactor vessel by a pump driven by the main turbine.
y
.1 Pressure vessels and piping located within the drywell will be designed, 19 tested and constructed in accordance with applicable, requirements of the ASME Boiler and Pressure Vessel Code.
Piping outside the drywell will conform i
f to the requirements of the American Standards Association Code for Pressure i
Piping. Twelve safety valves, arranged to discharge into the suppression chamber, are provided to protect the reactor and primary system from ove r-pres sure.
The genera 1 concepts and criteria of primary system design proposed are, in important aspects, similar to those used in several other nuclear power plants. We believe that a system designed according to these plans can be expected to fulfill its safety functions.
(.
D.
Emergency and Safety System Design Emergency systems provide means either for safely continuing reactor operation in the event of some equipent failure or operator error, or for limiting the extent of damage and resultant hazard.
In many instances, design features of the facility which have been provided for the primary purpose of making plant operation more convenient or ' reliable. are, in effect, emergency sys tems. Other features are designed primarily for the purpose of providing p
i1 i
emergency functions.
t The principal emergency systems and components proposed for this g
reactor are:
5 i
(1) Alternate power supplies for critical electrical loads i:]9 (2)
Reactor control safety devices and circuitry 3
(3) Liquid poison injection system e
l 3.1 (4) Emergency cooling system
~
5 (5) Bleed and feed system q
-q (6) Core spray system 1
! ?
l l
(7)
Containment system
[g Some of these systems have already been discussed in this analysis.
!i
- p. g Principal features of other emergency and safety systems are discussed
}
in this section.
)
As is indicated in the discussion of the consequences of potential I
b accidents in a later section of this report, a number of these emergency systems, mainly containment and emergency cooling systems with their associated water supplies, power supplies and controls, must be relied on
- i to limit the consequences of serious reactor accidents to an acceptable level.
In all such systems, therefore, a high degree of reliability must be provided so that the system will perform properly in adverse circumstances, i
This requires not only careful design of the principal features but also i
)
__-_________________a
~ -..
(
31 attention to such related equipment as signal and control circuits, power supplies, and instrumentation. Mainter.ance and frequent testing of emergency systems provides the final assurances of readiness of emergency systems to response to the demands placed upon them.
These f actors must, therefore, be taken into account in final design.
1.
Power Supply Protection of power supplies is provided on several levels, as F
described in PGGE's application. He plant is tied into the PG6E i
distribution system by two 220 kw circuits to Ignacio Substation. All 1
~j plant auxiliary power requirements can be met by either a transformer tied 4y to the station generator or by a transformer tied to the 220 kv lines. An additional external transmission line and transformer of limited capacity
~
and an engine-driven generator provide emergency power to equipment necessary for safe plant shut down. Station batteries will supply the electrical
]
energy for the more critical loads.
In our opinion, these provisions are j
l suitable from a safety standpoint.
j 2.
Emergency Cooling Systems g-I' A number of different means will be provided for removing after-i heat generated in the reactor core as a consequence of radioactive decay j
1 of fission products. Such provisions are necessary to remove decay heat
[
aft-r reactor shutdown to prevent melting or rupture of fuel elements,
~
which would lead to the release and dispersal of fission products. These l
provisions will include:
1 (1) ne normal condensate-feed-water system (2)
An emergency condenser which can be put into operation in event the reactor must be isolated from the main condenser (3) A low pressure shutdown cooling system
____.m._____.._________.__.__
- si.
(
(4) A bleed-and-feed system which releases steam at a controlled rate to the suppression pool (5) A high pressure core spray system (6) A low pressure core spray system A number of sources of water (and pumping capacity) will be available to restore water lost through accidental ruptures or through bleed-and-feed operations. Both high head and low head pumps will be provided with some m
back-up pumping arrangements.
In our opinion, these systems in combination L
with emergency operation action should be capable of reducing to an i
acceptable degree the amount of fuel damage and fission product release q
from the facility which would result in the event of a major rupture of
.H
. ~,
the primary system.
PG6E has indicated in Amendment 3 to the Application that the emergency
- q lM cooling system will be provided with pump backup beyond that provided by 5..
the auxiliary (startup) feedwater pumps. We believe that this pump in conjunction with those previously described should be designed so that j
2 the plant can be provided with both high pressure and low pressure reserve 9
J. '
t j
pumping capacity beyond that described specifically in the application, t
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?
These features, in addition to the final design factors already discussed f
must be carefully reviewed in detail prior to reactor operation. There is no reason to believe, however, that the final design for the emergency d
systems cannot suitably firovide for the necessary safety functions for this facility.
E.
Radiation Monitoring Systems Design Radiation monitoring equipment is provided for two purposes involving safety of operating personnel and the general public:
(1) for monitoring of radioactive effluents and (2) for determining levels of radiation in work areas in the plant.
~
(,
The following measures, which are typical of other facilities, will be employed at Bodega:
(1)
Batches of liquid wastes will be analyzed radiochemically prior to discharge to determine quantities and types of activity present; (2)
Primary coolant and fluids in various auxiliary systems will be monitored by radiation detectors and samples for determination of quantities and types of activity present; p
(3) Continuously discharged gases will be continuously monitored to determine the quantity of activity discharged; (4) A program of radiological monitoring of the environment will j
be conducted; and (5)
Fixed and portable equipment will be used to measure radiation levels in occupied regions of the plant.
The type of plant proposed' and the environmental conditions do not I
pose any unusual requirements for monitoring methods and equipment. We believe that proper monitoring equipment and methods are available to fulfill the requirements of safety at Bodega.
[ ;
F.
Waste Treatment, Storage and Discharge Design Features
(
The applicant has described in general terms the radioactive wastes l
that would be produced during operation of the proposed reactor facility f
and has proposed general methods, for management of these wastes in order to meet the limitations of 10 CFR Part 20 of the Commission's Regulations.
The sources and general character of these wastes and the general methods proposed for meeting the health and safety requirements are summarized briefly in the following paragraphs.
1.
Radioactive Liquid Wastes lhe principal sources of radioactive liquid wastes from a plant
. of the proposed type consist of small amounts of leakage of primary e
(.
coolant from valves and equipment when maintenance is performed, and wastes from decontamination procedures. Other swrces include laundering opera-tions for contaminated clothing and laboratory operations which are carried out as a part of the reactor and power plant operating centrol procedures.
The amount of radioactivity in these liquids is variable and will depend primarily upon the concentration present in the primary coolant water system. The radioactivity in the primary coolant consists of fission products that may be released from the fuel through minor imperfections j '
l in the fuel cladding and of irradiated impurities that may be present in
-3 I
the cooling water. Such impurities would include corrosion products from I
)
P the coolant system and fission products from traces of uranium impurities C*
l that may exist in the fuel cladding surface.
'd i
Present experience with this general type of reactor at Dresden g
indicates that the range of radioactivity concentration in these liquid
.]
i wastes may vary from values which would be low enough to meet the drinking water requirements of 10 CFR Part 20 for the public without treatment to as much as several microcuries per cubic centimeter. The volume of these
,i i
wastes may vary from a few tens of thousands of gallons per month to a few a.1 5 -.s 3
hundreds of thousands of gallons per month.
The applicant is proposing to construct a special system of drains I'
and tanks for collection of the radioact&ve liquids from all potential sources, and to provide a Radwaste Facility for monitoring and decontamination of these wastes.
Radioactive liquid waste will generally be disposed of by injecting it into the condenser cooling water effluent stream after monitoring for radioactive content. This disposal will be so controlled 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
35 -
j 4
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 Regulatior.s.
2.
Radioactive Solid Wastes The principal sources of solid wastes include spent domineralizer resins, filters, scrap equipment and miscellaneous. trash such as hand' tools, 1aboratory ware, etc.
The applicant proposes to collect' the radioactive domineralizer I
resins in storage tanks which would be located in the Radwaste Facility
.)
1 and to dispose of these materials by shipment 'to an AEC-licensed waste
.?
disposal f acility. The shipments, of course, would be required to. meet the appropriate AEC and ICC standards for shipment of radioactive materials.
All other solid waste would be collected and stored in a vault constructed k
for this purpose pending packaging and shipment to the licensed waste I
disposal facility.
We believe that the facilities proposed by the applicant for handling e) these solid wastes are satisfactory.
3.
Radioactive Gaseous Wastes f,
5 Gaseous wastes from a reactor of the type proposed consist I
principally of non-condensible radioactive gases that are removed from the main condenser by the air ejector, and from the turbine gland seal system.
L 1he process areas, laundry and laboratory will contain trace amounts of contaminated dusts, mists and vapors.
In addition, the drywell will contain radioactive argon as a result of neutron irradiation ~ of the air within this cavity, ilthough this air would only be released to the' stack on an intermittent basis whenever personnel access to the drywell would be required for maintenance purposes.
l
sa I
36 -
The air ejector off-gases, which comprises the bulk of the gaseous radioactivity to be released to the atmosphere through the stack, will consist of various isotopes of nitrogen and oxygen. The off-gases also may contain non-condensible fission product noble gases, mostly xenon and t
krypton, that may leak through minor imperfections in the fuel cladding into the primary coolant system or result from irradiation of trace quan-T~
tities of residual uranium contamination on the fuel cladding surfsce.
1he 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 experience at the Dresden Plant we believe that Nitrogen 13 will be'the O
mjor contributor to the total gaseous radioactivity release to the atmos-phere during normal operations.
In this regard, experience at Dresden has shown that concentrations of radioactive gases released are well within the g{
9 limits established by the Commission's Regulations.
The average annual rate of gaseous radioactivity release from the stack i
at this facility will be limited to a specified quantity which will assure that the requirements of the Commission's Regulations are met.
f;a.4 The.
meteorological and topographical conditions at the site as well as the 3
j engineering design of the plant will be taken into account in order to 4
establish this limit. The applicant proposes to tacasure the amount of radioactivity released from the air ejector and instantaneous radiation L-level to provide for an alarm and for closure of the off-gas vent system valve if the radiation level would exceed limits pre-detemined by the Commission to be acceptable.
The system proposed by the applicant for measurement and control of the radioactive. gases is an application of concepts that have been previously approved by the Commission for other reactor plants of this type. Accordingly, l
we believe that the methods proposed will be adequate to satisfy the health and l
1 safety requirements for operation of the type of reactor facility proposed.
l 1
( -
l' V.
Research and Development Program While the design of this nuclear power plant is similar in many respects to other plants now in operation, certain components and features l
of this facility will require research and development work in order to com-plete their detailed design and to establish the adequacy of the intended l
design and operational parameters.
In recognition of this need, Pacific Gas and Electric Company is conducting research and development programs as outlined below:
1 (a)
Meteorology.
A meteorological facility is being installed at i
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the site to provide necessary data for atmospheric diffusion studies.
Instruments will be mounted at three levels on a 250 ft. tower and will
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measure temperature and wind speed and direction. All readings will be 7l
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digitized and recorded on paper tape.
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(b)
Oceanography. The capacity of the ocean to diffuse the condenser 7 i cooling water and minimize the effects of temperature and radioactivity
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. l on the marine biota is being investigated in a series of tests conducted at jl k
the site. These tests include use of drift poles and uranine dye as well
- Ii, as measurements of temperature and salinity. They will continue through at ls least one annual cycle of oceanographic and meteorological conditions.
(c)
Marine Biology Survey. An ecological survey is being conducted f
to prepare check lists of the marine fauna and flora of Bodega Head and I
b_.
Harbor.
(d)
Radiological Survey. A preoperational monitoring survey of the site and its environs will be initiated two years before commencement of operation of tee reactor. The details of this program have not been com-pleted for Bodega Bay. However, it is anticipated that it will be similar 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-d pression concept have been conducted.
Additional tests will be conducted
]
at the Company's Moss Landing Power Plant to determine whether or not baffles between vent pipes are required in the suppression pool.
In addition to the research and development work being carried out by the Pacific Gas and Electric Company, the General Electric Company is carry-1 ing out a number of research and development programs of safety significance a
that will influence the design of the Bodega Bay plant. These are:
I (a) Fuel Development.
Results from fuel element development tests f
l and experience with fuel designs now employed in existing reactors will form the basis for the selection of the design of the Bodega fuel elements.
(b)
Instrumentation Development.
In-core startup range neutron detectors l
l are being developed as a possible substitute for the previously planned l
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out-of-core detectors.
1 1
(c) Control System Development.
A prototype Bodega control rod drive l )"
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is currently being manufactured.
It will be subjected to extensive lj developmental testing before the final drive design is released for manu-1]
[,
facture. Several devices which could reduce the likelihood or magnitude l'
l of a control rod dropout accident are being developed for possible use in j
the Bodega control system.
(d) Nuclear Excursion Analysis Development.
Analytical models are W
being developed for the more accurate prediction of the physical conse-quences of nuclear excursion.
It is ou3 opinion that the titieiFefahd development programs propcsed to -
~
be conducted are reasonably designed to resolve the safety,quettions with respect t6 those features or components 'of the ' Bodega remetor'vhich reqdire research and r
development in order to comolete their detailed' design or to establish the ade--
quacy of that design in light of the proposed operational parameters.
As specifically noted in various sections of this report, we believe that the
l I
l h0 applicant should submit the results of,particular research and development programs
-i as they are completed in order that the reculatory staff may determine the ade-
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quacy of the proposed detailed design during facility construction.
l VI. ' Analyses of Potential Accidents _
i The design features of the plant have been described in the previous
.i sections, and in many cases the safeguards provided by a particular design feature or the operational limits imposed by safety considerations for a particular feature have been discussed. 'In general the criteria for plant j
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~i design should include:
(1) means to ' control radiatica hazards (including g
discharge of radioactivity) during normal operaticn; (2) design features to
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minimize the probability of having an accident; and (3) design features for
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'.l mitigating the consequences of an accident should one occur.
4 The means for controlling radiation bazards during normal operation 9'l vill be provided by suitable shielding and radiation monitoring in the case
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4 of direct radiation emitted from the reactor and by proper monitoring of I
radioactive vastes which are discharged from the plant site.
For vastes
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discharged from the plant, the release rates shall be such that they j
do not result in personnel exposures $n excess of 10 CFR 20 limits.
j i
a The adequacy of the design features that are incorporated to mitigate j_
the consequence of accidents in the unlikely event that one should occur
,Ll are evaluated in the following paragraphs. The consequence of these j
accidents to the health and safety of the public is presented taking into consideration the safety features afforded by the containment and other i
emergency systems and the environmental character of the site.
To evaluate the adequacy of design features to be incorporated into the plant design to minimize the probability of accidents, a number of representative abnormal ccaditions, equipment malfunctions and operator errors were postulated and evaluated by the applicant. Those which were j
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presented in the Preliminary Hazards Summary Report included:
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- 41 a.
Changing pressure regulator handwheel setting d
i b.
Continuous control rod withdrawal or' insertion c.
Loss of electrical load d.
Control rod drive malfunction e.
Recirculation pump failures f.
Main steam valve closures 1
-t g.
Failure of reactor safety valve to rescat 1
h.
Failure of reactor safety system 1
W 1.
Fuel cladding failure Jj
- j. Loss tf feedwater k.
Loss of auxiliary power E
m.
Instrument air failure n.
Pressure regulator failure i
o.
Emergency condenser tube failure
,d p.
Reactor system ruptures inside the drywell t ;
q.
Failure to replenish cooling water in emergency condenser j ;
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r.
Startup accident
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4 s.
Fuel loading and handling accidents I4
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Cold water accident i
In addition to those conditions listed above, three equipment failures Q
termed " Major Accidents" were evaluated by the applicant. These accidents included:
a.
Control rod drop accident b.
Main steam line rupture outside the drywell c.
Reactor system rupture in the drywell
(
i In some of the malfunctions and failures presented, the evaluation is not yet completed; however, the applicant has stated that when the analysis is complete, the results will be used as criteria in the detailed plant design (for example, to size the pressure relief valves and to set' the isolation valve closure specificatens).
In our opinion the evaluations that have been completed at this time I~*
have formed a satisfactory basis for detsemining the nature and consequences
(
i of the hazards of reactor operations, including the consequences of the maxiy A1 mun credible accident.
In addition, these analyses have established a proper qj 9
basis for the detailed design of emergency systems, except for further
_j Ti considerations which should be given to the control rod drop accident, q
In this accident, calculations by the applicant indicate that the most b.
reactive control blade could have a reactivity worth as high as.036.
.- g Additional calculations show that if this blade were to drop free of the 4j j -, ;
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-qw quences to the reactor vessel in the event of this accident are not entirely y
c.
clear. As part of the research and development program, the applicant has I
indicated that they are developing analytical models for more accurate j.
prediction of the c. consequences of such a nuclear excursion, and that the j
forthcoming SPERT destructive. test will be used to check the model that is developed.
In addition 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-trol rod patterns to reinforce procedural controls provided to insure that patterns causing individual rods to assume undesir61y Ngh reactivity
(
h3 a worth are not used.
Conceptual designt, fc.y flow natricting devices that would limit potentia 1' control rod dropout velocities to safer values are-being developed. Should there.be unsatisfactory results of. the research and
' development work to develop experimental verification 'of the applicant's position that a rod dropout accident of this type will not endanger the reactor vessel, we believe that other design features, such as the rod T~
vorth minimizer computer or the rod dropout velocity limiter, should be incorporated into the plant design. However, we believe that by use-of i9 these alternative approaches one can obtain adequste assurance that the j
control rod drop accident would not have consequences as serious as the maximum credible accident discussed in the next section of this report.
I VII. Maximum Credible Accident Evaluation y
For the purpose of determining the upper limit of public hazard incident to operation of this facility, the applicant has hypothesized a major nuclear accident involving a substantial release of radioactive fission products from the reactor fuel, and has estia.ated the consequences of this accident bi in terms of potential radiation exposure to the public. This analyses takes s
into consideration the moderating effects on such exposures of the containment 1
and other emergency systems ( and the environmental characteristics of the site. The maximum credible accident chosen by the applicant results from an instantaneous complete rupture. of one primary coolant line inside of the y
dryvell after reactor operation at rated power for an extended period of time with the fission product inventory at a maximum. The pipe rupture would release the pressure in the reactor system (assumed to be at 1250 psig), resulting in all of the reactor coolant system water flashing to steam. An immediate buildup of steam pressure in the dryvell to about 62 psig would ensue and pressure would increase in the suppression chamber to
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about 35 psig. The pressure in the suppression chamber would be reduced within a few minutes due to steam conder.sation in the water contained within the suppressica pool.
Other assumptions concerning the magnitude of the accident and the effectiveness of the engineered safeguards systems for alleviating the severity of the consequences are as follows:
(1) The loss of coolant from the reactor is assumed to result in melt-ing of one-half of the reactor fuel and damage to the cladding in
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the remainder of the fuel.
(The facility will be provided vith an C
emergency core spray system for preventing significant damage to i
~;j the core under these severe conditions; however, for the purpose c
of this analysis, it is assumed to be only 50% effective.)
15k (2) The fission product release from the. fuel is assumed to be the t
following:
From Damaged From Melted-Total Release Fuel Cladding Portion From Fuel
,1 Ji L M Noble Gases 20%
100%
,* 60%
Halogens 20%
100%
60$
Other Solids 0
15 0 5%
(3) One-half of the halogens are assumed to be removed in the dryvell sad suppression pool ty plate out and. scrubbing action of the L
v ate r.
Ite remainder of the halogens and all of the noble gases are "available" for leakage to the reactor building at a rate of 0.5 ' percent per day at design pressure.
As the pressure in the dryvell and suppression chamber decreases, the leak rate correspondingly decreases.
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(h) The reactor building air, which.is maintained at a slight negative pressure (1/4" of water), is assumed to be exhausted to the atmos-l phere through filters for removal of particulate and halogens at.
1 a volume flow rate equivalent to 100 percent of the reactor build-
.f 1
ing air volume in 2h hours. The particulate and iodine removal l
filters are assumed to be 95 percent effective. As has been r.___
stateu, provisions vill be made in the design for verification i
of the effectiveness of these filters en a periodic basis.
1 (5) The applicant's evaluation of the consequences of this accident assumed release to the atmosphere through a stack 300 feet high, d
R although at the present time no specifications have been proposed
'O for the stack design.
Assuming that the vind direction and velocity were constant during b;
the course of the accident, the applicant calculated exposures as follows:
j, u
For good meteorological diffusion (lakee) conditicms and a vind 1.
i speed of 10 miles per hour, the maximum exposure rate at ground level j
y 1
would occur at a distance of approximately 0.6 miles from the stack. The j
q maximum potential dose rate to the thyroid is approximately 8 millirems per I
hour (0.008 rems per hour), and the total potential dose for the duration f.. '
of the release to the atmosphere is approximat$1y 15 rems. The maximum 4
i potential whole body dose rate due to noble gases is approximately 2 millirem W
per hour, with a total potential dose for the duration of the e.ccident of approximately 0.02h rem.
2.
For moderate inversion conditions with a vind speed of 3 miles per hour, the. applicant estimated the maximum exposure rates at ground level would occur approximately 3 miles from the site. Under these i
conditions the maximum potential dose rate to the thyorid and total dose
A-.__
U; h6 for the duration of release is less tr.ta 40. millirems per hour (0.0k rems per hour) and T rems, respectively. Tne whcle ' body potential dose rate and integrated dose is less than 10 milliremn per hour (0.01 rems per hour) and one rem, nspectively.
- As' previously stated in the section of this report describing the meteorology of the site, we telieve that vind speeds during inversion con-m ditions may be somewhat lower than assmned by the applicant in this evaluation.
I Accord,tugly, we have made calculations based upou the above assumptions wMch
.i
^1 take /< nto account the possibility for the accident to occur at a vind speed f
i j
n of one meter per second (2.2 miles per hour), under either lapse or it/ version i
cr:nditions. Using an effective stack height of 300 feet and the one mter i
l
?j per second vind speed, we estimate the.t the maximum potential cole body t 9
l d
and thyroid dos.ges for the durt.tica of the accident would be 0 9h rem and 53 rem, nspectively.
.),
The staff has considered both the applicant's assumptions concerning 9,<
s the postu1Nted maximum credible accident and the general concept of the i
safety features proposed for mitigating the consequences of such an accident.
s As indicated in the above analysis, the amount of fission product released
['
from the fuel depends on the extent of core damage, which in turn, depends en the effectiveness of the emergency core spray.
In our opinion, a suitable core spray design which would provide for an adequate supply of La emergency cooling water would substantially reduce the extent of damage to t,he fuel, even to the point of preventing any melting. On the other hand, even if the emWrgency core spray faile'd to function at all, the inventory of fission products released from the core would be increased by appro imatt/ly
'I a factor of two.
In thin case the dosage values given above vould be increai,ed to approximedely 2 rem whole body, and to 100 rem to the thyroid.
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.,Q From the above analyses, we have ccr. eluded that the engineered safety
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d features proposed for this facility should be capable of significantly limit-
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ing the degree of harm that eculd result if a severe accident such as postu-lated should occur. Therefore, since it is believed that the occurrence of I
such an accident is highly unlikely, we have concluded that the upper limit
- s cf public hazard incident to operation of the Bodega reactor is an acceptable i
R risk to public b.alth and safety.
1 VIII. Technical Qualifications 9
-]
The technical qualifications of PG&E are described in the application
[_.5 for a construction permit. PG&E has constructed and is now operating a
.~,d boiling water nuclear power plant at Humboldt Bay near Eureka, California.
Ysn Their principal contractor for the Bodega construction, the Generti Electric yf Conpany, designed and furnished the major compcments of the Humboldt nuclear
! 4 steam supply system, including tne reactor with ito ^cor.trols and in'stru-j',
l l
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ment ations. GE has also designed and furnished similur equipment for s
q several other boiling vnter reactors in this countsy and abroad. On the 3
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basis of their dencestrated ability in similar endeavors, we have concluded 1
that PGiE and their principal contractor are. technically qualified to con-
- l' l
i j
struct the,; proposed facility.
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