ML20234E025
| ML20234E025 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Bodega Bay |
| Issue date: | 07/26/1963 |
| From: | Beck C US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20234A767 | List:
|
| References | |
| FOIA-85-665 NUDOCS 8709220263 | |
| Download: ML20234E025 (77) | |
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D R A F T #2 CKBecksv1 l 7/26/63
. SAFETY ISSUES IN BODECA BAY REACTOR
- 1. The Bodega Bay reactor proposed by PG&E is similar in all important design concepts to several other successfully operating reactors.
- 2. There are no innovations, new principles or concepts of any substance in this application. For example, most of the systems follow the design concepts and principles already in successful use in other plants, i including such systems as:
l the boiling water concept and general reactor features l l the pressure vessel { the primary system i the core design ! the control system the instrumentation system i the poison injection system emergency cooling systems ; the auxiliary power system l radiation monitoring i vaste treatment, storage, disposal. i the general plan of vapor suppression and confinement building containment.
- 3. Matters on which some thought will be given as the detailed design is finalized include:
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- a. The thickness of fuel element cladding, now tentatively proposed
{ i as 11 mils. The applicant may decide later to use thicker cladding. , If thin cladding is used, the occurrence of failures before intended end of life, which would release fission products into the primary coolant, might force uneconomical shutdowns. There is no assurance ; that similar leaks would not occur with thicker cladding, though the probability would undoubtedly be reduced. In any case, there I f t 8709220263 851217 PDR FDIA PDR F I REST DOS-66,5 a
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( ( would be no substantial hazards, and the eventual outcome , I would depend on economic balances to the operator. 1 l
- b. Design changes of control blades, now under study, may ]
result in increased assurance that inadvertent blade drop-out j would not occur. The present system would be acceptable, but improvements are always welcome.
- c. The design details on the over-sil containment volumes and on j individual penetrations of containment must be developed to l accommodate initial and subsequent periodic pressure and l
l leak test measurements to an extent which has not been l required in previously authorized plants. This does not arise from extra hazards in this p?. ant nor from less favorable site, but from realization that procedures in previous plants l l ' have not been adequate. l l h~"we d. The number and location of isolation valves in the primary system pipes leaving the suppression pool containment have not been 1 finally settled. Satisfactory locations can be chosen; the matter I l can be resolved at a later stage. 1
- e. The applicant proposes to operate with a higher percentage of l steam volume in the core than has been customary. Further analytical work and some limited and controlled initial operations of the plant will eventually determine whether or not this is feasible. If not, it can feasibly operate with lower void volumes and would do so initially anyway.
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- 4. There are no systems on which research and development must be done -
to settle concepts, principles and general arrangements.
- 5. On some matters, further work and collection of infonnation must be done in finalizing design details,
- s. Where and to what extent internal baffles are. needed in the i
suppression pool will be determined by experknents now in '
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1 1 progress at Moss Landing. J l b. An environmental monitoring program, including surveys of ocean currents, marine life, etc. as well as radioactivity levels, is being conducted to provide a reference level for determination. ) whether any changes occur later as a result of plant operation.
- c. Detailed meteorological observations are being made to provide j
information needed in determining operating procedures and license limitations, w;ssmes, 6. Accideg ,c calculations and site characteristics lead to determination of containment pressute and leakage specifications and to containment l l volumes, as follows: l l l l l The plant dry well and suppression pool are designed to these values respectively, and, we believe, are adequate.
- 7. The site for this reactor is an excellent one in all respects but one, namely, the proximity to the well-known St. Andreas earthquake, which
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runs 1000-1500 feet to the northeast of the proposed-reactor l l location. The population distribution is favorable, the isolation distance factors are well wichin acceptable ranges, the meteorology, though not good for certain portions of the time, is as good or better than that in California generally, and, with the isolation l l distances, the containment characteristics already described, and ) l generally known and successfully proven plant characteristics, are j considered acceptable. g.r.,
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The only real issue yet to be settled, either in plant location or in the extra and additional design safeguards on otherwise adequate plant systems, is the proximity of the plant to the St. Andreas fault and the implications this has for the plant. l l Except for the earthquake question, the site and design features would 1 permit granting of a construction permit forthwith. Because of this question, however, on which further data remain to be assembled, the acceptability of the site is not settled, and if acceptabic, whether and to what extent extra safeguards and structural reinforcements on components and systems may be required also remain to be determined. h l I 1 i
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i '( ( DRAFT CKBeck:v1 7/26/63-I l HAZARDS ANALYSIS by'the .j l DIVISION OF LICENSING AND REGUIATION in the matter of l PACIFIC CAS AND ELECTRIC COMPANY \ 1. BODEGA BAY ATOMIC PARK UNIT NUMBER 1 CONSTRUCTION PERMIT DOCKET 50-205 l i
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I. Introduction '4'" The Pacific Gas & Electric Company (PG&E) has proposed to construct and operate a ' nuclear' power plant on Bodega Head in Sonoma County, California. PG&E will design and supervise con-I struction of the unit, and the General' Electric Company (GE) will furnish the nuclear steam supply- system and the turbine generator.
, The proposed plant, designated by PG&E as Bodega Bay Atomic Park Unit Number 1, will produce nuclear energk at the rate of 1,008 gegawatts (Mw). The gross electrical generating' capacity 4s will be approximately 325 Mw.
The Bodega Plant is similar [in many respeete7, in all its .- I essential concepts to boiling water power reactors now in operation'. 1 Steam will be generated in a direct cycle, forced circulation boiling
,p/ water reactor. fite detailed design will be based en operating expertenee from the Valleestes Boiling Water Reseter and the Dresden, ,;nsss Genowners, and Humboldt Bay remeters. 'These features of the plant-wheek require receareh er developmental effort in~ erder to provide ..
engineering informatten neeeseary for their detailed design er evaluation will be dieeuesed in Section-IV of this analysis. Eneept for a dieeussion of seismological and geologieel feeters this report sentains a dieeussion of first, the-significant features ef the eine and envarennent whiek have a bearing on safety and seeendly, the significant featuree-of design whieh affeet the probability of er eensequenees of. seendente er eseuvrenees of safety-signifiennee to the general publie. The' redfelegieel impaat of neraal routine operation of the planty ineluding. thedieekargeofradieaetivematerialsareeensidered,y I l - . . , , . . . . , . . , , , , . , , , - , _ _ , . .. L_______.___.___.___
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ps ;g The vapor-suppression concept of containment with a super-posed low- q 3 leakage repelling building, ' first used on the Humboldt Bay reactor, will .j
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. be used at Bodega. The Bodega site proposed is' near the St. Andressi earthauake fault. This raises cuestions as to the suitability of the -
site, and, if suitable, whether and to what extent extra safeguards and structural specifications may be required because of the possibility' -j of future earthquakes. - l 3
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( . l l g-II. Background fj On December 28, 1962, PG&E submitted an application to the AEC for a construction permit and operating license pursuant to Title 10, Chapter 1, Code of Federal Regulations, Part 50 (10 CFR 50). ] l The application, which includes a " Preliminary llazards Sumary Report", j dated December 28, 1962, and Amendments 1, 2 and 3 to the application I dated March 4, April 5, and June 13, 1963, respectively, has been j
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reviewed by the staff of the Division of Licensing and Regulation _d_ ~ 'l Technical eensultante in specialized areaa-ales advised the regulateey staff. The application has alee been eensidered by the AE6 1s Advisery
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i Gemmittee en Reaeter Safeguarde (A6RS)-as required by the Atemte Energy i Act and the regulations of the AEG. The weeemmendatione of the AGRS, ) as expressed in ses report of April 18, 1963,-(a-eepy-ef-whteh-is -
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attaehed herete se Appendix-6) were aise eensidered in the-regulatory staffle evaluation ] , with the assistance of technical consultants in specialized areas of technology. The application has also been' considered by the AEC's Advisory Committee on Rr. actor Safeguards (ACRS). The final recommendations of the ACRS are expressed in its report of April 18,1963, (a copy of which is attached hereto as' Appendix C). fIe is evetemarily the ease-in reaetor facilities prior to the essenensement of construstica, there-are a number of features of plant design and operaties wktek have met, as-yet, been-definitely resolved. The Ceaunissionis we8vlations-paevide for-the-issuanee-of-a-senstruetten permit en a provisional basis la cases-sveh-as this, in which some espects of design have not been-completed. A provisional-senstruction parmi,e gypy be teamed, aceerding to section-60,35 GFR-en-the-beed.e-ef 5 Tindinge,-among ehav_% .t;hR;; (1) the upplicant has described the.penpnW j
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design of the facility; including, but not limited to, the principal architectural and engineering eviteria for the design, and-has identified the major features er sempenente en which further teeknical information se requiredt (2) the emitted teeknical information will be suppliedt (3) the applicant has proposed,' and there will be eendueted, a research and development program reasonably designed to reselve the safety ;guestiene, if anyy with respeet to these features er components whiek require research j and development and that' (4) en-the basis-of-the-foregeingt-there-is-reasonable
? l assuracee that-(1)-such-eafety questiene-will-be-eatisfeeterfly-resolved-et i
er before the latest date stated in the application-for sempletten of eenstruetion of the propesed faeility-and (ii) taking inte eensideration the oike-eriteria-eentained in Part 199 1 the propened facility een be eenstructed and operated at the proposed leeation without undue risk to the health and safety-of the publie.
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The preposed eenstrueksen permit is granted would autherise eenstruetion only, The Gemmissien would require timely reports from PG&E with respeet-te results of research and development and final design of the mere significant design features. The AEG etaff would eenkinue-ite evaluation of the safety of the plant in light of this information. An operating lieanse will not be sessed until the final-design has been-eempleted and evaluated by the AEG etaff and the AGRS, In addition, the definite plans for operations would be evaluated by these two groups f' Pursuant to a Notice of Hearing published , the issuance of a f provisional construction permit to PG&E will be considered at a public hearing to be held in the hearing room of the Board of Supervisors of l Sonoma County, Santa Rose, California, at 10 a.m. PDT, on 1963 before an Atomic Licensing and Safety Board appointed by the AEC. The issues to be considered at the hearing are: q
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- 1. Whether the applicant has submitted sufficient information to provide reasonable assurance that a facility of the general type proposed in the application can be constructed and operated at the proposed location witho t undue risk to tho'
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health and safety of the public; (h viauals. 4,f course, the implications of the proposed location in close proximity to the St. Andreas 8 earthquake fault line)
- 2. Whether there is reasonable assurance that the technical information omitted from and required to complete the application will be supplied; l
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- 3. Whether the applicant is technically qualified to design and construct the proposed facility;
- 4. Whether the applicant is financially qualified to design and construct the proposed facility; and
- 5. Whether the issuance of an authorization for the construction
> of the facility will be inical to the connon defense and security or to the health and safety of the public.
/_Yhe propeeed plant if eenstructed en Bedega Head would be subjeet to severe sheek from earthquakes, h ere is aise a possibility that earthquakes l
l in the vietnity might eause faulting beneath the plant which would eause ! severe damage to the facility. ne possible effeets of sueh seismie activity en the preposed plant are still under study by the Geenaissionis-regulatory staff. De staff has not yet determined whether-er not a plant aan be , eenstreeted and operated safely et thie leention, At this time further l information which.must be. beeed on exploration of the site, among other l l tMnge, must ee ebtedneds Further consideration mus.t.also be g.iysa 10 criterte which must be eppiled in the design of systems er eempenents of-the feet.11ty which are of importanee to safety, especiaMy these eyetene i _._a_-___.. _ _ _ _ _ _. _ _ - _ _ . . _ . _ _ _ _ _
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1 which must be relied upon in an emergency, such as a severe earthquake, l l to prevent personal injury, i This analysis and the conclusions contained therein are made without- ! regard to the special safety considerations which must also be taken into account in view of the seismicity of the proposed site. 1 The staff's evaluation of the proposed Bodega nuclear power plant I described in subsequent sections of this report and its position on the 1 issues at the forthcoming hearing are based on /J1J.7 the technical
* ' information submitted as part of the applicant's request for a construction l -l I permit and the report from the ACRS. /All of this7 This information is available for inspection and review at the Comission's Public Document Room in Washington, D.C., and at the Comission's San Francisco Operations I Office, 2111 Bancroft Way, Berkeley, California.- This evaluation and proposed recommendation is subject to modification in the light of any . %. further information which may become available, including /Jhe_7 any evidence that may be introduced at the hearing. /jndertheGemissienle l
1 regulations, any pereen whese interest may be affected may appear at the hearing er may f61e a petition to antervene and, 6E granted, may partiespate in the preeeeding, The decision of the-Gemission wall be based upon the entire reeerd in the preeeedingy7 III. deser6ption and Safety Analysis 7 Identification of Safety Issues:. The Bodega Reactor is a direct cycle, foy;ced circulation boiling
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water reactor with provision for internal / separation of water entrained with the steam leaving the reactor vessel. Nuclear energy released in i
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. the reactor at the rate of 1,008 megawatts will be transferred to the
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circulating water coolant /wh6eh is e6reulated throph the remeter2 [
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- which is partially converted into steam W the electricity generating turbines. Steam generated [a the remeter atJ at a pressure of' 1,075. psia-
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flows to a turbine generator Iwit[ which has a gross electrical generating 3 1 capacity of about 325 megawatts. Reactor [eelantwh6ekhasbeea]' water separated from the existing steam is recirculated through four external pipe loops each containing - a ptanp rated 'at 29,000 gpm. After passing through the ~ turbine the steam is condensed, and the condensate after demineralization 1 is returned t'o the reactor vessel. 8his watery wh&e[ The circulating water (and steam condensate), which will contain some radioactive materials, will be circulated within a closed system from which the only normal effluent will be a continuous discharge of noncondensible gases. This gaseous material, I also containing radioactive components, will be monitored continuously and i y, released from the reactor stack if the contained radioactivity is below perm- ' issible limits. [Asineenventionalpower,plante,thesendenserwillbe ' eeeled-by water drawn from a nearby seures. Inthiseasygatertocool the condenser will be taken from Bodega Bay and discharged into the Pacific Ocean. Fromtimetotime,[eguinted]controlledandmeasured
-t quantities (( radioactive _, liquids wastes acetanulated from plant ' operation, i i
will be mixed with the condenser coolant water and discharged to the ocean.
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Thus, from the routine op'eration of the plant three cuestions of potential l huced, arise: 1 (
- 1. Are the quantities and identity of the routinelv generated waste liquids, the storage tanks, monitoring instruments and mechanisms for l
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- 11 controlled release to the atmosphere such that safety limits will be complied I
with on level of radioactivity released to the environment? and !
- 2. Are the cueantities and identity of the routinely generated waste liquids, the storage tanks, monitoring instruments and mechanisms for controlled release to the ocean such that safety;11mits ujil be complied with on the level of radioactivity relecsed to the ocean?
- 3. Are there adequate plans and procedures for handling and disposal l
of the spent demineralized resins (highly contaminated with collected radioactivity) and other radioactive solids accumulated during plant operation? ) Within the 592 fuel assemblies of the Bodega reactor, about 1 gram of nuclear fuel vill be burned each day. About 1 gram per day of fission products l per day will be left in the fuel as a residue. It is this accumulated inventory , I of highly radioactive fission products which constitutes the real potential i threat of this reactor or of any reactor to the health and safety of the public
- m. in surrounding areas. Normally, the inventory remains distributed within the dense fuel matrix, which is clad, for added assurance of radioactivity retention, with a thin, impervious layer of stainless steel. It is oniv if some substantial fraction of this material should be released to the environment, that any j threat to health and safety would occur.
The clad fuel elements in turn reside within the closed, high pressure ( primary cooling steam generating system, and this in turn is contained sithin a heavy concrete, high pressure capacity, low-leakage suppres:sion system which would be capable of retaining the steam, water and radiocetivity which i ! would be released from any break of the primary systen, piping within the vapor suppression container. Fast acting stop valves are provided in all
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primary system ifnes going beyond the container to prevent release if a leak should occur _in pipe lines outsidd the container. Several possible accidental combir,a, tions of circumstances can be visualized by which release 1 of some of the fission product inventory could occur. .In any reactor r a safeguards are provided to prevent the occurrE.co of any foreseeable r ( circumstances which would release dangerous materiy to the envirornnent. l
. . 3 Therefore, the crucial objective of any reactor safeg evaluation is to ;
i I l determine the adecuacy of safeguards to prevent release of hazardous amounts of she fission product inventory of the plant to the environment. In the followinn sections the summary of this extensive analysis,,,,,whicN , ,, has been carried out by the Regulatory Staff will include (in addition to l 1 and 2 above) the following: i i
- 1. A stranary of the principles, design concepts and systems in the l Bodega Reactor which have been well ey_t,ablished t through their use in other
.wn,.. reactors.
- 2. Particular questions on the containment structures:
- a. Design specifications
- b. Leak testing l
! c. ' Isolation valves in primary lines.
- 3. Thicknesu of cladJing on the fuel elements
- 4. Percentage of steam voids in the core '
i h'. Improved design of control rod blades / r
- 6. Pumping capacity in emergency system (37a)
- 7. Implications of the proximity to the St. Andreas fault.
l 8. Strmnary on suitability on site. 1 l l
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, [~9a evere11 judgment eescerning the safety of remeter operatione l
on.the meeeptability of potential hasards must be based vpen a
,) hanber of individual safety ensiderations. Our judgmente at this 4
time are based wpen an evaluation of design details, design eviterien . l OMIT ALSO THE ATTACKitD 3 PAGES_/ f n n
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.m' III.- Safety Evaluation:
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; 3 7 1. Solids. Ifauid and naseous wastes from normal operation.
- a. Radioactive wolid wastes. which will include spent j s
' demineralized resins. filters, scran eauipment and miscellaneous laboratory trash. will be anoroorf ately packaged and shioned to I
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, an authorized disposal faci 11tv.
- p. Leakane of crimary water from valves. orimary water accumulated from maintenance operations. radioactive liauids from decontamination 1
"d procedures and liould laboratory wastes es the extent of. from a few tens of thousands to perhaos a few hundred thousands of gallons per month. constitute the radioactive licuid wastes exweeted from this olant. Most of the liauid would be oniv slightiv contaminated. little 9
I i f any of it would be in the "high level" radioactivity range.
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All liould wastes will be collected by a system of drains into
,m.,,,, special trucks for storare and monitorina. Facilities will be provided for converting high level wastes into solid form for shipment to a disposal site. The ordinary low level-wastes will be iniected-at a monitored and carefully controlled rate into the effluent condenser cooling wasten which flows into the ocean. The temperature of the e outlet water will be to above that of the ocean at that point.
Ocean water already has a natural level of radioactive content. Regulations of the Atomic Energy Conunission, which is based on the recommendations of the Federal Radiation Council, permit additional limited cuantities s .'; of radioactivity to be released into ocean water, provided the resulting
- n concentrations do not exceed those stipulated in published regulations l
(10 CFR 20). These permissible levels aie set sufficiently low to meet y de ourity requirements for continued use as drinking water. ! i u ,, . m .m. , , ,,,; . , . . . . , , , . , . . ,
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I The radioactive liquids from the plant in mixing with the l condenser coolant ( gpm) are greatly diluted. In general, best I advance estimations and experience with other sbnilar reactors indicate j that the upper limits of radioactivity concentration in the effluent would seldomly need to be reached except for short periods of tLme and otherwise the levels would be well below the permissible 1Laits. In any case, adequate monitoring procedures would be observed to insure that l l l l the permitted levels were not exceeded. l . The Commission believes that the levels of radioactivity permitted to be discharged into ocean waters are far below that which would cause hazard to any individual or ecologic disturbance to the plant and animal life in the area. These levels have been set only after extensive study of all information available and the advice of the most knowledge- ! I able experts in this field. a nx.a Extensive radiation and detailed ecological surveys of the area i are in oronress and will be comnieted prior to any authorization of plant oneration. to established backcround reference information on the present situation at Bodenn and in surrounding ocean and land areas. A check list of marine farms and flora is being prepared. Detailed measurements of the ocean currents, rate of mixing of water near the proposed condenser l cooling water outlet, and dispersal patterns are being made using drift poles, uramic dye, temperature gradients and salinity data, i Thus, background information on the existing situation will be i fully and quantitatively known. A subsequent phase of environmental monitoring observations will be extended into any operating periods F h-k____.___._
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which may be authorized. This post-start-up surveillance program will . p.&tV se be adeouste to discoversany significant changes in the environment thi64% l gym, ($rp)Nu5 lWff""? Yht'"'4WrdM$ l Qf From aj $ Crud $e.*Wdffkehjetbo JW y0M<'J cJ)jVacQA'hh our knowledge of the op rational experience with other reactors 4 il and from the effects predicted by anticipated operating conditions and by ] calculations, we believe that any effects on the natural conditions of , biota in the area resulting from releasing the degree warmer I condensers, cooling water or the permissible levels of radioactivity
# into the ocean will be so small as to be negligible.
- c. Some radioactive waste gases will be dispersed into the atmosphere in regular operations c' the plant. Such gases will be diluted by the efm. exhaust ventilation air stream from the top of the ft. stack l l
and released at a carefully monitored and controlled rate. The radio-active cases will consist mostiv of non-condensible gases routinely ) I w pi e removed from the primary coolant system by the air eiector. The ma_ior radioactive components of these gases will normally be N-13, with some N-16 N-17 and 0-19, and, inlease of minor defects in fuel elements, noble ges fission products Argon and Krypton.au/pu/ cl/dE6
- 1 The permissible amounts of radioactivity which may be released to the atmosphere in normal operations are determined by the limits established by the Commission on concentrations of radioactive components p u' in the air which would flow beyond/there members of the public may I'
be located. From the point of stack release to the site boundary dispersion and dilution will occur, depending on distance and meteorological parameters.
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1 I l In calcuintinn the stack effluents which may be permitted, l _so that r,armissible concentrations are not exceeded beyond the site k 1 l boundaries, averagina over periods of time. rather then limitation odf l any-instantaneous transitory peak concentrationjmay be permitted,Mence, ' specified limits are based on time averanes of meteorological diffusion parameters. A 250 ft, meteorological tower is being constructed on the 1 site to furnish data on temperature, temperature gradient and speed, directions. l etc., for the eventual determination by the Commission of air dilution factors ! I which may be assured and hence, of permissible levels of radioactivity , in the stack effluents. Monitors in ef fluent lines. alarms and automatic division of effluent streams to a system of storane tanks in case 3rf
'l excessive levels are reached, will be provided to insure that permissible j levels are not exceeded.
1 l The environmental surveillance pronrams mentioned above will establish l l . background information on existinn atmospheric radioactivity levels and provide a means of establishing any later channes which may occur. We i believe the plant can operate within the prescribed permissible concentration limits, that such levels of radioactivity will not constitute an ' undue hazard to health and safety and that there will not be a cumulative environ-mental contamination of any consequence from the routine effluents.
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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, California, approxi-mately 50 miles northwest of San Francisco. Bodega Head is bounded 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
- towards Bodega Head and forms a natural breakwater for .
l l Bodega Harbor. In the orientation of Bodega Head and' the 1 4 proposed plant site with respect to the main California . ) coastline as shown in figure 1. l The environmental factors which are considered to be l l significant with respect to the safety for this site and which
,_. ,g, have been examined in detail include: (1) the location with respect to the nearby population, . (2) the meteorological factors,and(3)themarineenvironmentalfactors.(Asnoted previously, geological and seismological factors,are signi-ficant to safety and will be the subject of further considera-i tion by the Regulatory Staff These environmental factors are important insofar as they might affect the normal operation of the plant or cause damage to essential equipment and with respect to the bearings, the population and other environmental factors have on the course of accidents.
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c c 1 n i p. 6 Plent'Iocation . The proposed location for the reactor plant is on ! 1 a 225 acre tract'of land at Campbell Cove near the Southern" .l end of Bodega Head. The property owned by the applicant includes the entire southern end of Bodega Head.' - The adjacent property'to the north 'is under' acquisition , by the University of' California for a research' f acility. - l 3 ,,. The reactor would be located on the east side of Bedega Head near Campbell Cove and across the~cntrance' channel to Bo'd ega Harbor from Doran Park. The nearest' , k edge of Doran Park, which is owned by Sonoma County and q contains no residences, lies approximately 1,300. feet east of the proposed reactor.. The traffic through the entrance channel to Bodega 6ff?ti lMW#fP' Harbor consists primarily of commental and sports ' fishing boats. Usage of both the Channel and Doran . Park could be controlled under emergency conditions if l this should become necessary for protect:1o'n of the kf 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 approxi-mately 2,700 feet (0.5 miles). The population data submitted by the applicant based on the 1960 census shows population lar@ than about u j; 200 within 10 miles of the' site, and bone large han -
.t about 3,000 within 20 miles of the s te. The nearest.
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C C r, ,, cities of more than 10,000 are Santa Rosa (31.037) and Petaluma (14,035) which are 21.and.24 miles respectively from the - site. The population density within 25 miles is as follows: Persons / Square Mile Total Population'- Distance of Land Area In Area 0-1 0- 0 1 -5 21' 500 5-10 16 1600 10-15 81 15,700 15-20 97 30,000' 20-25 180 66,700-1 I The above . tabulation shows that the proposed reactor site' is favorably located vith' respect to population distribution and density. The location on a peninsula provides natural barriers against ' the ! future development of housing within at least two miles of the reactor plant. The tabulation shows that the i population density is low essentially out to the i distance of Petaluma and Santa Rosa and these cities' account for about 75 percent of the population in the annular area that lies between 20 and 25 miles from the ( reactor site. ' Based on the population' data the staff believes that the low population distance and the population center distance for the Bodega site can be considered to be 24 miles, which is the distance to Santa Rosa. i s --
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- 2. Site Meteorolocv
, In general, climatology of the coastal area at Bodega Bay is typical of the central to northern coastal area of California and is characterized !
by a wet season extending from about November through March and a dry season from about April through October. The topography of the area inland from the site is characterized by a series of' hills r.nd valleys, with the hills rising to elevations varying from approximately 400 feet to approximately 1000 feet. The roughness of this terrain would be expected to enhance the atmospheric turbulence and result in , .. ,4 rapid dilution of gaseous effluents that might be-transported from the site to inland locations. On the other hand under strong inversion conditions, the range of hills along the coast, which is approxi-mately three miles from the proposed site, would - tend to restrict the transport of airborne materials
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to the inland areas. Detailed meteorological information for the 7 proposed reactor location is not presently available. However, observationsoof vind directions and velocities of the coastal locations at Point Reyes, approximately 22 m11cs south of Bodega Head and t
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;. . . Jenner approximately~12 miles north of the Head,:
shows the wind direction towards the coastline approximately 60 percent of the time with the prevailing wind direction from the northwest. The information 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 has proposed diffusion parameters for use in the Sutton equation for estimating the amount of atmospheric dilution under various meteorological conditions, as follows. Meteoroloniest Condition. Strong Moderate. Lapse Parameter Inversion Inversion n .5 .3 .22 l t Cy .2 .21 .6 j Cz .02 .07 .2-tr (miles per hr.) 5 5 10- ! The staff believes the parameters provide adequate conservatism for estimation of the possible exposures under accident conditions except with i respect to the wind speeds proposed for the inversion l conditions. In our opinion, a wind speed of one meter per second, which corresponds to about 2.2 miles pw
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u, , l per hour, would be more representative of a ! coastal location during periods of slow diffusion. Table I, Appendix III of the applicant's Pre- . liminary Hazards Summary Report, which shows a , relatively high frequency of the' lower wind speeds 0 in the zero to three miles per hour range confirms our judgement in this respect. _ j X The meteorological f actors discussed here have been used in the analysis of the maximum credible. accident evaluated by the regulatory staff and described in Section VI. PGiE has constructed a meteorology station with a tower approximately 250 feet in height to collect l meteorological data, such as the frequency of wind 7-l
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speeds and directions of various atmospheric and stabilizing conditions which would be appropriate for use at the proposed site. It is expected that sufficient
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l i data vill be available for estimating the capacity of the atmosphere to safely dilute radioactive gases that might be released from the site. These d'ata should be available before the reactor plant would be ready to operate.
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- 3. The Marine Environment Approximately 250,000 gpm of cooling water for l
the 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 so, therefore, does not become i
- h. radioactive from exposure to neutron irradiation.
L-This cooling water, however, will contain liquid 1 i
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effluents that are periodically released afj ter ' monitoringj from the rad waste facility. The con- 2
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centration of radioactivity in the condenser cooling ) 1
, water before discharge to the ocean would bel controlled to meet the requirements of 10 CFR Part 20 of the Commissien's Regulations.
At the request of the AEC, the U. S. Bureau of Commercial Fisheries of the Department of the Interior' has reviewed the effects of reactor operations on the marine environment of Bodega Head. The Bureau's report, which is attached as Appendix II, indicates potentially significant effects from the discharge of this effluent to the ocean ast l l
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(1) Possible concentration of radioactivity in seafood, (2) Possible concentration of radioactivity along beaches used for public recreation. - The Bureau of Fisheries states that it is well established that certain levels of radioactive wastes can be discharged into the oceans without adverse effects on fish and wildlife, since circulation insures mixing of ' radioactivity with large volumes of water, quickly diluting and dispersing the radio-activity. Their report further states that the permissible levels and rates of discharge of l radioactivity are difficult to determine cdvs5iEe- f t. l
.msmmi for any specific area, a ecommends an extensive monitoring program to insure that concentrations of radioactivity in the marine life do not exceed pre- 2 detemined levels. '
, j The experience at Windscale, England, where
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radioactive wastes have been discharged to coastal ! waters for several years was cited by the Bureau of Fisheries as an exa=ple that shows no adverse effects on fish or shellfish from reactor operations. The English have determined on basis of a monitoring program carried out over several years that 3,000 curies
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, , i per month of strontium 90 could be safely released to the coastal waters at Windscale.
PG&E has specified that the concentration of radioactivity in the condenser cooling water discharge will not exceed that' permitted by Part 20 of'the I Commission's Regulations. Based on a discharge rate of 250,000 gallons per minute the regulatory' staff 4 b has estimated that approximately 4.1 curies of
. strontium 90 per month could be discharged to the v Pacific Ocean, if all of the radioactivity in the
( condenser cooling water consisted of this nuclide. This value appears extremely conservative in comparison with permissible release value determined m'
~ 4.,um by the English for Windscale. L
'The applicant has initiated extensive studies of l
oceanography and marine biology to evaluate the 7 1 marine biological aspects of the proposed reactor operation. These studies are as followst - h 1
- 1. An oceanographic survey will be carried out to determine the circulation ^ pattern of the ocean in the vicinity of the outf all, and the capacity of the oceen to dilute the condenser cooling water discharge.
1
- 2. An ecological survey which will include an inventory of the marine organisms in the ;
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17 vicinity of the outfall will be carried' out.
- 3. An environmental survey of the marine environment will be conducted to determine .
existing levels of radioactivity before the reactor'becomes operational. This program would continue af ter the reactor s is in operation to_ determine any tendencies for reconcentration by the marine environment of radioactivity w-released to the ocean from operation of the plant. The staff has considered the information submitted-4,o-w by the applicant and the' comments of the U.'S. Sureau- i of Connercial Fisheries and concludes that ' the liquid effluents produced by operat' ion of the. proposed .I reactor can be disposed to the Pacific Ocean without s i
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exceedingPart20 limits,[a rthermor$ in view of this it is extremely unlikely that adverse effects 'i [4 on the marine environment itself will be observed.
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The staff further believes that the programs proposed by the applicant for maintaining vigilance over the marine environment will be adequate to discover any I anomalies that may occur so that'these could be corrected.
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., 4 The containment system serves as a principal safeguard against the .
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release of radioactive materials to the environment in the e f a najor
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rupture ix of the primary system inside the containment structure. Other _ w emergency systems are generally provided, as in this case, to either prevent - 3 or retard the release of fission products under various accidenti ' conditions . 3
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These zrmx various emergency systems operate independently to prevent the release of radioactive materials beyond the boundaries of the plant. Con- (
.." ament and other emergency systens are discussed in this analysis in the l
hi hb context that they are required to safeguard operating personnel and the publ e t
,~ in ih m n*v the event that the best effort to design the plant to the '
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highest standards and to construct and maintain and operate the plant 1 i m++4=++y competently should fail. The proposed design as designed .n this section + - + has been used in the analysis of the maximum credible accident, which is discussed in Section 6. That analysis is the basis for j
- establishing 6 important design parameters such as dryvell. and suppression chamber design pressures. -
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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
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~ Bay reactor facility. Significant features of the Bodega Bay Plant containment design include the following: j (1) Plans for the Bodega Bay Plant call for a dry well having a 60 ft, diameter spherical lover section and
{ a 26 ft, diameter cylindrical upper section. The over- l all height of the dry well is approximately 100 feet. (2) The reactor vessel and four reactor recirculation loops, each with a pump, vill be located within the dry well. (3) The dry well vill have an airlock entrance. Personnel entry is not pjanned during reactor cperation, but is contemplated with the reactor hot and pressurized, my w,, (4) The suppression chamber will be in the form of a torus i and vill have a major diameter of 93 ft. and a cross-section diameter of 26 ft. It vill contain about h65,000 gallons of water and have an air space above the water of about 80,000 cubic feet. 2 Both the dry well and the suppression chamber vill be designed and constructed in accordance with the ASME boiler and Pressure Vessel Code, Section VIII. Piping restraints vill {
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be provided at containment penetrations to assure that failure j of the pipe vill not cause containment rupture. A concrete k I
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,g, 43 i refueling ' building vill contain the dry well and suppression chamber. Pressure and leak rate specifica* ions for these containment system components are as follows:
Leak Rate Component Design Pressure (% of volume in 2h hours) Dry well 62 psig 0.5 (at design pressure) i Suppression chamber 35 psig 0.5 (at design pressure) Refueling building 12 in. H 2O
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100 (at 1/4 in. H 2O) W In order to oof tesf the Bodega Bay pressure suppression design, Pacific Gas and Electric is conducting a test program at its Moss Landing Power Plant. As in the Humboldt Bay project, the I EL N ~N m m m1nn1 - applicant has constructed a full scale segment of the suppression system. In the test of Bodega Bay design a single 24-inch diameter vent pipe from the dry well to the suppression chamber was used. Since the full size plant is to have 112 of these vent pipes, the I testequipmentrepresentsa1/112segmentofthecontainment. Tests vere conducted with this mock-up to simulate various accident conditions. A flow comparable to 1/112th of the flow resulting from a complete circumferential break of one of the 28-in, reactor coolant recirculation lines (with flow out both sides of the break) was taken as the " maximum credible operating accident" (MCOA). Highest containment pressures observed in these tests were 52 psig in the dry well and 30 psig in the suppression chamber. These pressures were observed when the mock-up dry well was pre-heated to 2550 F and when the mock-up reactor vessel vater was l subcooled 350F. Tests at higher and lover dry well temperatures and at higher and lover reactor vater subcooling yielded lower dry well and suppression chamber pressures. m p me g9 e4 9 9 1 b- . 'I __.--A
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.10-In another test a break area 2.5 times that of the MCOA was simulated. In this test the peak dry well pressure observed was 63 psig. Further Moss Landing tests are being conducted to determine whe' t her baffles are needed in the suppression chamber. i As another significant containment design featur
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Gas and Electric proposes that in a number of instances a single s e i isolation valve vill be installed at the containment - vall in f pipes or ducts penetrating the containment; however, each such line vill have two isolation valves, one
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of which may be a remotely operable process valve located else-where. (:;n ? '
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Two isolation valves located at the dry well vall in each \
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< l q main steam line are to close on a manual signal or automatically
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on the occurrence of any of the following: (1) Lov condenser vacuum (2), Main steam line leak (in the pipe tunnel) , (3) Lov reactor water level, of PG&E is giving consideration to providing ihn protection mx the main steam line isolation valve:against foreign catter, which might interfere with proper valve action. l l 6
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3 . l 1 The Bodega Bay design is such that during refuelind, the spent fuel stora6e pool vill connect directly to the sh' eld i water above -\ the reactor, thus permitting direct underwater transfer.of fuel without the need for a transfer cask. In our opinion this fea'ture provides in a aimple and reliable way for both continuous shieldin and cooling of spent fuel during transfer and storage. ' i i k i l i 1 1 i+,.. + l I l 1 i k 5 4 g
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h ' 20 - The refueling' building in which the' dry well'and suppression chamber system are located is provided with a controlled release ventilation system which discharges.to the plant stack, The. i building and ventilation system design 'is such that the refueling l building can be maintained at a negative pressure. Discharge. r from the building passes through halogen and radioactive particulate f cleanup equipment prior to discharge to the stack.. PG&E has in- k , dicated that,(in accordance with the' recommendations of M'CQ n the system vill be designed _to permit frequent testing of the - ability to filter particulate and to remove iodine at specified efficiencies. f , (The following 2 paragraphs will be revised to reflect the 8/ ( expected Amendment 3. )
-k' The staff believes that the general containment. scheme proposed by PG&E is adequate for the proposed reactor. It is our opinion,-
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however, that some important criteria for the' design of the con-tainment features have not at present been specifically proposed. ) , Such additional criteria, including those mentioned explicit,1y by the A CRS report to the Commission, are necessary to' issure 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 consequences of a maximum credible
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accident.
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The regulatory staff, in revieving the details of design;vhen they are I developed, intends to assure that the following ' criteria on containment
/. testing, penetration design, and isolation valving are me.h 3 6 i
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- 1. The design should permit initial integral leak rate i
j testing af the dry-vell and suppression chamber at their respective design pressure after the installation of all i penetrations (including piping' conduits, electrical. con- ! duct 6rs, and gasketing closures) and subsequent periodic testing at suppression pool design pressure. 2 the G initial testing, the leakage rate of the containment system 1 y 1 should be determined as a function of pccr pressure up to ; i full design pressure.
- 2. The design of penetrations should take into account, in h addition to the pressure load, the loads or deformations l
imposed by themal expansion, impact'of missiles, l l
, reactions of ruptured pipes, and disturbances incident to installation, maintenance or repair. Penetrations should f ..
be shielded from missiles to the extent practicable. 111 penetrations should be designed so as to allow frequent l
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periodic leakage rate tests of the penetrations only (including points of attachment to the containment shell), at full design pressure. 3 All pipes and conduits which communicate with interior of r the primary system or the contminent system, and other piping (such as instrumentation and control piping) which
~\ cannot be adequately protected against accidental rupture, should contain double isolation valves. All valves per-forming the function of isolation valves should be provided k
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22.. } vith protection against materials in the system which proper might prevent / closing and should be provided with reliable automatic and manual actuation features. Isolation valvin g ) I should be designed so as to permit periodic leakage rate I thsts. Appropriate closing times for isolation valves should I be determined on the basis of analyses of system ruptures ; which would release fission or activation products outside the dry well while the valves are not fully closed. ; 1 The Staff believes that PG&E shoulds 'ubmit for 1 1 Cnmmission review the results of further developmental i tests of the suppression pool concept and final design plans for the containment as soc as they can be made %) l available. (Hote : /.dditional ec= cats wi' L aude on Tine effect5 ' i
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1 l ( 22a C-l p, { 2ea c h o- A(acl ear- Syslc~s As indicated previously the design of a reactorshould be expected to r4 Y I inherently provide a large measure of assurance in the safety of operation. ? , That is,the features of the plant which are necessitated by considerations 7 of convenience and economics do, incidentally, provide safety and reliability. *
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With additional consideration for safety values which are involved, b the features of the reactor core, reactor controls, and primary systems fil which are necessary for normal operction can be designed to provide the ' k principal protection against the development of radiological hazards resulting from reactor accidents. _ To safeguard a reactor core against damage that would release fuel l and fission products from the confinement profided by fuel elements, one must adequately provide fcr (1) control of the chain reaction, (2) removal j of the heat generated in the fission process and the radioactive decay of fission products, and (3) protection of fuel elements against mechanical damage. Control of the rluelear chain reactor is necessary to limit the ' rate at which energy is released within the fuel element in the form of 4 heat. ! Excessive heating or inadequate heat removal vould cause overheating i 1 of the elements to the extent that cladding is weakened or subjected to 1 an excessive internal pressure. Mechanical forces, such as reactor pressure 1 and hydraulic forces, which impose loads on the fuel element structure, I must also be taken into account in fueltelement design. 9 Generally speaking the mechanical design of the reactor fuel elements and the design of controls and the heat removal system should be
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, such that, in normal operations and under many conceivable
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" e accident situations, radioactive ' fission' products.would
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be confined to the fuel elements themselves. If this objective is met then the primary coolant system serves as a m secondary containment system. k
. Primary System The functional integrity of the primary system ~ is.necessary to the 0
integrity of the fuel elements themselves. Its integrity is also necessary in normal operations to properly confine the relatively minor quantity of radioactive fission products which are expected to leak from fuel elements. 4 As p,-opous Gy FE/Ef the reactor core will be located in a reactor, pressure vessel designed, built.and tested in 4 accordance with Section VIII of the Boiler and 6 Pressure Vessel Code of the American Society of Mechanical Engineers. The 50 ft, high by 15 ft. I diameter vessel will be constructed of carbon steel - , , approximately 6 inches thick. The interior of the - 1 ! vessel will be clad with stainless steel applied by welds overlay methods. The design pressure will be 1235 psig at 575'F. Steam generated in the reactor is passed through 1 axial flow steam separators and. driers located and then inside the reactor vessel. through two 20-inch steam lines 4
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{ I to' the turbine, which is located in a separate , 1 structure. . Water, after separation far. from 7 4 { steam in the' reactor vessel, passes from the s-
-reactor vessel through four 28-inch pipes. Four I
.J recirculation pumps, one in'each loop, provide the.
. driving force for circulating water through the ' 1 reactor core. - Feed water from the condenser is ,
i l injected into the reactor vessel by a pump driven by the main turbine. Pressure vessels and piping located within the dry well will be designed, tested and constructed in.- ' p accordance with applicable requirements of the ASME
- s Boiler and Pressure Vessel Code. . Piping outside
\( the dry well will conform to . the requirements of the American Standards Association Code for Pressure Piping. Twelve safety valves, arranged to discharge i
into the suppression chamber, are provided to protect the reactor and primary system from over-pressure. The general. concepts and criteria of primary system design proposed are, in important aspectsy similar to those used in several other nuclear power plants. The staff believes that a system designed according to hese plans ;
. can be expected to fulfill its safety functions.
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- 2. Core Design <
Since the nuclear characteristics of a reactor.pape dependenpupon
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thefuelconfiguration,[amongotherthings] structural stabEi] of J rces of iechanical fp o 1 core components is important to reactor safety. < or thermal origin Vaich exert stresses upon civiliane components of the I core must be taken into consideration in the, design. Dmtm Neutron absorbing control blades which are necessary for the adjustment of reactivity and
/
control of the nettron chain reaction must be provided with channels in / which they nay move freely. The power output of a reactor, which is of prime importance to the 1' j l operator depends upon many factors involving nuclearj hydraulic, and thermo-dynamic phenomena. The objective of thesafety evaluation is to determine 1
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the effects on coy and fuel elecent integrity of operation under the j d proposed modes as they affect power distribution and stability, flow rate / and themo] dynamic properties of the coolant, and fuel temperatures. I l
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( 2Sa 1 ( i f - j The reactor core 411 be' composed of 592 fuel assemb ies each of which provides a vertical channel through which
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the mixture of steam and wat9r passes. The core will ) have approximately the form of u, right circular cylinder 140 inches in diameter and 125 inches high. One hundrcd
/
and forty-five control rods will enter the core from 4
, . below the fuel assemblies through conttel rgd' guide tubes'.
1 ' s The fuel assemblies are held in proper l position by upper a. d j i i s lower grid plates which are attached to the cylindrical ! l core shroud. The weight of the fuel assemblics is borne b by the control rod guide tubes which extend to the lower head of the reactor pressure vessel. 1 ! I
.. Each fuel assembly will be composed of 49 fuel rods in a square array. Fuel pellets of UO2 enriched to 2.5% U-235 will be contained within stainless steel tubing. PG6E has tentatively proposed that this tubing would have a nominal
/
l thickness of 0.011. inches and would be abic to withstand an exposure of 15,000 Mh'D/ TON. On the basis of present information and operating experience, one cannot bt assured that fuel with cladding of this type and thickness
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e can be irradiated for the exposures stated without experiencing excessive rupturing of the cladding.
!!owever, the design of fuel for the Bodega reactor will
[ ( f not be completed until further data from a General [ 9 - Electric research and development program are availabic. In our opinion, this program is reasonably designed to provide an engineering basis for a suitable fuel element k design. In any event, extensive experience with other M fpowerreactorsprovidesreasonableassurancethata l design suitable from a safety standpoint can be developed , Thermal and hydraulic factors, which ultimately determine the permissible power level of the reactor, have only been briefly described by PGGE. While there is not at present a firm basis for establishing thermal limits i as high as is suggested by some of the parameters specified
,...s n the applicatio PG6E has estab ed a criterion for
[ determining the proper detailed thermal and hydraulic design factors for the Bodega reactor, namely, that the fuel willM l l t operate without loss of cladding integrity over the design wme exposure period at the maximum heat fluxes possible within burnout limitations. Operating experience at 'other boiling water reactors has indicated that this cri'erion can be met and that it is acceptable from a safety standpoint. The power distribution which is expected in the Bodega Reactor core has been estimated for the purpose of making a preliminary determination of the thermal margins
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\ power distribution in the course of design of the reactor.
After operation commences the power distribution will be g 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 g the thermal margins that are experienced in operation.
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The stability of reactor power is affected by operating co ons which in turn have transitory effects on the neutron econog of the chain
/h 8 reaction. The volume of steam within the r'eactor coolant is one.such op.erating condition which has an important bearing on reactor stability.
e Preliminary calculations indicate that at rated I operating conditions the steam volume fractions would t , be as follows: Average Core Voids-37% ' l l Average Exit Voids-58% i
- Although the staff is not aware of any substantial I operating experience that would confirm the acceptability I of operating at void fractions this high, PGGE believes I
' that analog computer studies being made will show that the '
i plant can be designed to exhibit satisfactory dynamic- ; performance with such high void content. In any event, ' l g however, we believe that with appropriate limitations high Q void conditions can be safely approached in reactor tests l devised so as to determine the proper range of void content for nazzan nc .1 reactor operation. a ; i n 4 .w. -s .w. er ,e . y me i*g. c .w - ea ..w, e- 4 +e< .4-- . . - .m +.
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- 3. R_eactor Controls Nuclear safety safety requires-that there be , {
reliable means for controlling the reactivity of a i M nuclear reacto g Reactivity can be related directly to I the rapidity with which the neutron chain reaction t changes. When reactivity is positive the chain reaction gro : that is, the rate at which nuclear energy is released by fission 1
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is increased. Conversely, when reactivity is negative the
- chain diminishes and power falls. The operating condition of the reactor, its temperature, pressure, power level, j 1
void content, and fuel exposure all affect reactivity. l In our opinion the general way in which each of these factor s
; );
, affects reactivity is quite well known, and the theoretical ;
and experimental methods for investigating reac tivity effect s I are sufficiently developed to perma design v.' the Bodega l I e*w geactor control system to proceed with confidence. 1 It is expected that within the range of operating variables contemplated for this reactor the reactor shouldbe) table. That is, an increase in reactor power I causes changes in operating conditions which have a strong na+unti Atendancy to decrease reactivity and limit the power increase. The main purposes of the Bodega reactor control system, therefore, are to provide a means of precisely and reUahly
~
'j adjusting reactivity - '
p(.1 Necessary safety objectives for the control system are to provide 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, and (3) maintaining stable level of pover safely within the capabilities of the heat removal system. i 4s%B8 e%_ S M=see s e q $M-F
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. ptoemed The control systema f or this reactor is an array of 145 movable control blades, which have sufficient react ,vity worth to keep the reactor safety shut down, even though one l
, of these blades (or rods) might be stuck out of the reactor CorC.
The k-effective of the n actor with all control blades in the reactor core is calculate'd to be 0.97, 1 The material in the movable blachs vill be boron carbide contained within 0.175-in. O.D. stainless tubes similar to those presently in use at Ih esden. Additional control devices, removable only through a loading procedure, are provided by fast control curtains of 0.11, boron stainless 1 steel,which will be semi-permanentlylocated between selected fuel elements. The fixed control devices provide a flexibility in The reactor design so , that reactivity of the core can be easily adjusted to attain the shutdown I 1 margin required fromsafety considerations. 'r'-:-- =- ;- - - 21 r- J - h m 1 l v-i na - . ; , -111,61 l l k l l m . . - . . . . . = . . . - .%- _,u%, e e. .a.
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( ! a f The reactor design also incorporates a liquid poison system that can be used to inject xxxx sodium pentaborate 1 into the core in the event complete shutdovn cannot be ) d roA. I'b 'Y" *#
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hydraulic control rod drives to be used in the Bodega Bay plant are to be designed using the same basic concepts as have been employed in drives in use. in boiling water l i reactor plants at Dresden, Big Rock Point, Hu=boldt Bay I - i and the SENN plant in Italy. Water used as the hydraulic '
,. fluid can be applied to either side of a piston which is !
4 mer' 'nically coupled to the control rod, thus providing for
- i
{ either upward or downward rod motion. Only one rod can be moved outvard (increasing reactivity) at a time,and it may be moved either continuously or in 6 inch steps. Rod speed is controlled by orifices which regulate the flow of water I m.m - % i away from the low pressure side of the piston. ! at - Therodspeedfornormalwithdrawalvillbecontrolledsothat/themaximum ra ;e of withdrawal the reactor power would not increase at a fxxt rate fact enough to lead to serious consequences. can be inserted simultaneously, shutting the reactor down. Rods are scrammed upward by applyin8 pressurized water from fN'7 ' either the reactor or from accumulators to the bottcm side j of the drive pistons and simultaneously relieving the kx volume above the top ixd side of the pistons to the scram du=p tank. The drive is locked in fixed positions by collet fingers which enga6e Brooves spread at 6-inch intervals along the movable index tube. The collet fingers support.the weight of the rod and the downward forces due to reactor pressure.
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.-3r-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. This includes Dresden-PoM experience as well as initial Big Rock 4operations.
At Big Rock Point, there have been two isolated occurrences of rod " drift-out". In one of these, the cause was attributed 1 to an inadvertent release of de=ineralizer resins resulting in ' t l the collet fingers being jammed in the open position so that i j the rod was free to drift as influenced by the forces due to ' gravity and hydraulic pressure. In the second case it was .j reported that a hard particle became trapped between the collet q piston and a sleeve which is located between the collet and the I j index tube. This again is believed to have caused the collet
. ~m em /
fingers to be jammed in the open position, thus permitting rod drift. The hard particle was never found. It should be noted, I however, that in neither of these cases nor in any other case - has there been any apparent stea^-:-_ impaiment of scram l capability. Detailed desi6n of drives for the Bodega reactor has not been made. General Electric is considering modifications of earlier designs that vill minimize the possibility of ! us a p htm opeah.Mr. foreign material accumulating in the rod driveg. The applicant has also inddcated that functional and endurance tests will be made on the prototype h%;zz Bodega mechanisms, but the detailed procedures for these tests and the acceptability criteria have not been proPO*ed
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- Control. systems which are designed to react rapidly .to demands for shutting a reactor down generally have some potential for accidentally increasing reactivity as well. This aspect- {
of the PGSE control system design is discussed later in this report (Section V), where consequences of a rod drop ^out are v considered. PG6E has indicated that devices for limiting individual rod worth and for impeding the fall of a rod are under development. Such devices could enhance.the-safety of operation and simplify the procedures that are pxs presently used with similar drives to provide the necessary j i assurance that a rod dropout-accident cannot cause a~ serious
] l public hazard. i l
I
/
In view of the importance to safety of the detailed design of the control rod drive system, the staff believes that PG6E should submit timely reports to the Commission nsAn.. during facility construction on development, design, and.
/ testing of the Bodega control system. ' ,
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c C 4 4 Control and Safety Instrumentation The instrumentation necessary for safety in a nuclear power plant generally involves a large number of i 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 collected by the measuring instruments is used to guide the operating staff in g controlling the plant and to actuate automatic control f devices. The instruments, circuits, and control devices which are ) of most importance to public health and safety are: (a) f those necessary for and contributing to stable reactor operation,
'r (b) those used in control of radioactive fluids and effluents, j
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1 e43_,, and (c) those used for control of emergency equipment. j
; -- kneral adequate description of posed instrumentation
, & fa a%sce.pfon is adqwfc 4
- is presented w in the application.3 _ this time / there is uvi sufficient information available to determine whether 1
instrumentation provisions have been made for all essential functions or to determine the degree of reliability that should be attributed to the reactor protection system, which is described by PG6E as " fail-safe". 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 provide an acceptable design. The staff intends O
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( ( to eval e the reactor control an safety instrumentation in detail pr r to reactor operation order to assure that
. proper attention as been given to the eed for automatic functions and the re ' ability of safety instrumentation.
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e C C E. Energency and Safety Systems i Energency systems provide means either for safely continuing operation in the event of some equipment failure or operator error or for limiting the extent of damage and resultant hazard. In many instances, design features of the 1 l facility which have been provided for the primary purpose of i making plant operation more convenient, reliable or economic 3 l .are, in effect, emergency systems. Other features are designed primarily for the purpose of providing emergency functions. The principal emergency systems and components proposed l for this reactor are: (1) Alternate power supplies for critical electrical loads - { e%~= t (2) Reactor control safety devices and circuitry; l (3) Liquid poison injection systems l (h) Energency cooling system l (5) Bleed and feed system (6) Core spray system . (7) Containment system Some of these systems have already been discussed Q is. Principal features of other emergency and safety systems are discussed in this section. In all such systems a high degree of reliability must be provided so that the system vill perform properly in adverse circumstances. This requires not only careful design
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l 36 - of the principal features but also attention to such related i equipment as signal and control circuits, power supplies, and instrumentation. Maintenance and frequent testing of emergency systems provides the final assurances of readiness of emergency systems to response to the demands placed upon them. These factors must, therefore, be taken int [infinaldesign. _ ' The applicant in his analysis of.. accidents has taken credit for the effectively action of the core spray system to/ reduce the calculated doses to In rsons offsite. As is indicated in the discussion of accidents in Section 6 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 to limit the consequences of
=-
serious reactor accidents to an acceptable level. ) +no 4 >
- 1. Power Supply
'. Protection of power supplies is provided on several levels, as described in PG&E's application. The plant is tied into the PG&E distribution system by two 220 kv circuits i
to Ignacio Substation. All plant auxiliary power require-ments can be met by either a transformer tied to the station generator or by a transformer tied to the 220 Ev-lines. An additional external transmission line and transfomer of limited capacity and an engine-driven generator provide emergency power to equipment necessary i for safe plant shut-down. Station batteries vill supply l the electrical energy for the more critical loads. l
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- 2. Emergency Cooling Systems 8
1 A number of different means vill be provided for removing after-heat' generated in the reactor core as.a { consequence of radioactive. decay of fission products. Such provisions are necessary to remove decay heat after reactor shutdown to prevent melting or rupture of fuel. elements, which would lead to the release and dispersal of fission products. These provisions vill include: .j
- i (1) The normal condensate-feed-water system (2) An emergency condenser which can be put into operation in event the reactor must beIisolated-r from the main condenser (3) A low pressure shutdown cooling system -
(4) A bleed-and-feed system which ieleases steam at a 4, 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) vill ( be available to restore water lost through accidental ruptures or through. bleed-and-feed operations. Both high head and lov some . head pumps vill be provided with back-up 3 pumping arrangements. , In the event of a ma,jor rupture of the primary system, emergency action should be capable of reducing to a great. l extent the amount of fuel damage and fission.prcduct release" from the reactor. 4
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- PG&E has indicated in Amendment 3 to the Application that the 1
emergency cooling system vill be provided riyh with pump backup !
. j beyond that.provided by the auxiliary Um,-
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] J feedwater pumps. S e staff believes'that the plant should be provided 6both high pressure and low pressure reserve pumping capacity beyond
. s that described specifically in the application. These features, in addition to the final design factors already discussed must be carefully reviewed in detail. There is .no reason to believe,' however, that the suitable. final ,
l design for the emergency systems cannot be made 'so as to provide for the > L teessary functions usually encountered in nuclear power plants. [ MY m 4 l l i 1
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R <T. Radiation Monitoring Radiation monitoring equipment be provided for two olvingyi.gfety of operating personnel and the general public: (1) for monitoring of radioactive effluents and (2) for dete2 mining levels of radiation in work areas in j the plant.
- The following measures, which are typical of other facilities, vill be employed at Bodega: .
(1) Batches of liquid vastes vill be analyzed radiochemicallypriortodischargetodetermine quantities and types of activity present; (2) Primary coolant and fluids in various auxiliary systems will be monitored by radiation detectors and sampled for determination of quantities and types of activity present; (3) continuously discharged gases vill be continuously He monitored to determine g quantity of activity discharged; (4) A program of radiological monitoring of the environ-ment v111 be conducted; and { (5) Fixed and portable equipment vill be used to measure radiation levels in occupied regionso[44e plani. The type of plant proposed and the environmental conditions do not 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.
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( G. Waste Treatment. Storane and Discharne The applicant has de' scribed in general terms the radioactive vastes that would be produced during operation
, oftheproposedreactorfaciiityandhasproposedgeneral methods for management of' these wastes in order to 41,a la'oL4bne of
,, meet3 10 CFR Part 20 of the Commission's Regulations.
The sources and general citaracter of these vastes and . the general methods proposed for meeting the health and 4 - safety requirements are s_ummarized briefly in the following paragraphs. '
- 1. Radioactive Liould Wastes The principal sources of radioactive liquid wastes from a plant of the proposed type consist of small amounts of leakage of primary coolant from valves and equipment when maintenance is performed, and wastes from decontamination procedures. Other sources include laundlering operations for contaminated clothing and laboratory operations which are carried out as a part of the reactor and power plant operating control procedures.
The amount of radioactivity in these liquids p er % elly is variable and will depend3upon the concentration
, present in the primary coolant water system. The radioactivity in the primary coolant consists of "
et fission proddcts that may be released from the fuel f r. I through minor imperfections in the fuel cladding, f 4 0
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and of irradiate'd impurities that may be . present in the cooling water. Such impurities
'wouldincludecobosionproductsfromthe' coolant i system and fissi$'n products from traces of j uranium impurities that may exist in the fuel l
cladding surface.
, Present experience with this general type ;
of reactor at Dresden indicates that the range of radioactivity concentration ^in these liquid wastes may vary f' rom values which would be low enough to meet the drinking water requirements l ,
' of 10 CFR Part 20 for the public without treatment I
I to as much as several microcuries per cubic centimeter. l i weer The volume of the'se wastes may vary from a few l tens of thousands of gallons per month to a few hundreds of thousands of gallons per month. . , The applicant is proposing to construct a special system of drains and tanks for collection - of the radioactive liquids from all potential n sources, and to provide a Radwaste Facility for monitoring and decontamination of these wates. Radioactive liquid waste will generally be disposed 4 of by injecting it into the condenser cooling ' watereffluentstheamaftermonitoringfor - ir radioactive content., This disposal.will be so 6 P
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. controlled that. concentrations in the effluent streamatthepointofdischargetothePacific
, Ocean will not eSceed those specified'in 10 CFR 20.
The staff has reUiewed the general concepts o-proposed by the applicant for the design of thisrystemandbelievesthattheconceptisadequate to meet the requirements of the Commission's Regula-
~
ti ons. _
- 2. Radioactive Solil Wastes e u The principal sources of solid wastes inc1Gde n ppent demineralize,er resins, filters, scrap
, . equipment and' miscellaneous trash such as hand tools, laboratory vare, etc.
The applicanN proposes to collect the radio-activedeminerallzerresinsinstoragetanks which would be located in the Radwaste Facility 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 nd ICC standards for shipment of radioactive materials. All other i solid waste would be collected and stored in a
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vault constructed for this purpose pending 11-packaging and shipment to the licensed waste 1 .' i disposal facility. l a-D e * ** wt * == i .-ase- w>p,, ..p ,,.,-e.no= m p g. .,,e , , ,,w,,. , , ,
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The staff bd11 eves that the concept proposed by the applicant for handling these i il solid wastes is satisfactory.
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- 3. Radioactive Case,,ous Waste
{ Gaseous was$$s from a reactor of the type I proposed consist principally.of non-condensible i I radioactivegase'Athatareremovedfromthemain '
}W.
condenser by th air ejector, and from the turbine
' gland seal system. The process areas, laundey.
1 i and laboratory will contain trace amounts of i 1 l contaminated dusEs, mists and vapors. Further,
, , the dry well will contain radioactive argon asaresultofdeutronirradiationoftheairwithin this cavity, alEhough this air would only be i I '
released to the#$ tack on an intermittent basis whenever personnel access to the dry well would l ,,, berequireddfo{maintenancepurposes. The air ejector off-gases, which comprises i the bulk of the gase6us radioactivity to be released to the atmosphere through the stack,. 1 consist of various isotopes of nitrogen and i f oxygen. The off-gases also may contain non- ..
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d f. ! condensible fission product noble gases, mostly e xenon and krypton, that may leak through minor c , i
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If I s r. imperfections in the fuel cladding into ! the primary coolant system or result from , irradiation of trace quantitles of residual uranium contamination on the fuel cladding surface. The applicant has listed the isotopes of nitrogen
- and oxygen which would be present in i
l the air ejector off-gases as follows: N-13,
, , 1 1
N-16, N-17 and 0-19. Based on the experience j i at the Dresden Plant the staff would expect that Nitrogen 13 would be the major contributor to 1 the total gaseous radioactivity release to the atmosphere during normal operations. In this regard, experience at Dresden has shown 1 that concentrations of radioactive gases released l l are well within 'the limits established by the
- Cocntision's Regulations.
{ i
, , The average annual rate of gaseous radio- i activity release from the stack will be limited to a specified quantity at the time of granting of an operating license to assure that the i
requirements of the Consnission's Regulations are met. The meteorological and topographical condi-
,t tions at the site as well as the engineering
- l. design of the, plant will be taken into account 1r when this limit is established. The applicant
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l proposes to measure the amount of radioactivity
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released from the air ejector and instantaneous 1 radiation level to provide for an alarm and for closure of the off gas vent system valve if the radiation level would exceed limits pre - determined by the Commission to be acceptable.
,f The concept proposed by the applicant for a bh,
" f measurement and control of the radioactive 1
~
gases appears to'be an application of conceptions that have been pireviously. approved by the Commission A 1 l L3nr m*L-r:;;ter y1.ats. Accordingly, the staff . 1
/ !
j believes that the methods proposed will be ade- I
./
/ quate to satisfy'the health and safety require-Nm 81 ~
ments for operation of the' type of reactor
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/ facility proposed.
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- 17. Research and Development Program j While the design of this nuclear power plant is similar in many l respects to other plants now in opgation, certain features of t,he design requi.re research and development work to establish complete adequacy of intended 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) Meterology. A meteorological facility is beirg installed at the site to provide necessary data for atmospheric diffusion studies. l Instruments will be mounted at three levels on a 250 ft. tower and ; will measure temperature and wind speed and direction. All readings
, will be digitized and recorded on paper tape.
(b) Oceanography. The capacity of the ocean to diffuse the conden'ser l i cooling water and minimize the effects of temperature and radio- ! I , .m activity on the marine biota is being investigated in a series of test 3 l conducted at the site. These tests include use of drift poles and uranine dye as well as measurements of temperature and salinity. They will continue through at~least one annual cycle of oceanographic, and meteorological conditions.' (c) Marine Biology Survey. An ecological survey is being conducted to i prepare check lists of the marine fauna and flora of Bodega Head and Harbor. 1 M (d) Radiological Survey. A preoperational monitoring survey of the site and its enviro $2s will be initiated two years before commencement of operation of the reactor. The= details of this program have not been completed for Bodega Bay. .However, it is anticipated that it will be similar to that conducted for the Company's Humboldt Bay nuclear unit. _ _ _ _ _ .1
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.i (e) Pressure Suppression Tests As described in the applicant's hazards .!
summary report, Appendix I, extensive tests of the pressure supp{essioi concept have been conducted. Additional tests will be conducted at . the Company's Moss 1,anding Power Plant to determine whether or not i baffles between vent pipes are required in the suppression pool. In addition to the research and development work being carried out by l the Pacific Gas and Electric Company .the General Electric Company is. carrying out a number of research and development programs of safety significance that will influence the design of the Bodega Bay plant. These 1 are: - (a) Fuel Development Results from fuel element development tests and h experience with fuel designs now employed in existing reactors will form the basis for the selection of the Bodega fuel. (b) Instrumentation Development In-core startup range neutron detectors
. ~ . . ,
are being developed as a possible substitute for the previously planned out-of-core detectors. (c) Control System Development A, prototype Bodega control rod drive-is currently being manufactured. - It will be subjected to extensive developmental testing before lhe final drive design is released for
- ' manufacture. Several devices which could reduce the likelihood or
,# magnitude of a control rod dropput . accident are being developed for
\
possible use in the Bodega control system.
- V ts (d) Nuclear Excursion Analysis Development Analytical models are being_
dxt { developed for the more accurate prediction of the. physical consequences A' i of nuclear excursion. - l i It is the opinion of the Staff Nwweat these research and development programs
.ro are properly oriented. As provided for by paragraph 50.35, Title 10
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l of the Code of Federal Regulations the Staff should be informed of the results of these research and development programs and will considor these ! results in its evaluation of the propriety of granting an operating license to this facility. 9 *t}
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_ g_ G V. Safety Analyses j
, The design features of the plant have been described in the previous sections, and in many ~
cases the safeguards provided by } I a particular design feature or the operational limits imposed by cafety considerations for a particular feature were discussed. 1 In general the criteria for plant design should include: (1) . means to control radiation hazards (including discharge of radio-activity) during normal operation; (2) design features to minimize op#is. . , . w the probability of having an accident; and (3) design features for mitigating the consequences of an accident should one occur. ' The means for controlling radiation hazards during normal. *
-4 operation vill be provided by suitable shielding and' radiation.
monitoring in the case of direct radiation emitted from the-reactor and by proper monitoring of radioactive vastes which are discharged from the plant site. For vastes discharged from the w.vasnm plant, the release rates shall be such that they do not result in personnel exposures in excess of 10 CFR 20 limits. ! The adequacy of the design features that are incorporated to
.g mitigate the consequence of an accident in the unlikely event that one should occur are evaluated in the following section on the maximum credible accident. The consequence of this accident'to the health and safety of the public is presented ta1Hng into
' ' ~
9' consideration the safety features afforded by the containment pnd olher emage *cy
,I Asystemsand the environmental character of the site. J A i's
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- .i To evaluate the design features that are incorporated into the plant design to minimize the probability of having an accident l a number of. representative abnormal conditions,l equipment mal- 1 functions and operator ' errors were. postulated and ev'aluated by .
]
the applicant. Those which vere' presented in the Preliminary L q Hazanis Summag Report included: s
- a. Changing pressure regulator handvheel setting b.
Continuous control rod' withdrawal or. insertion I
- c. Loss of' elec't rical load '
- d. Control rod drive malfunction
- e. Recirculation pump failureo !
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- f. , Main steam valve closures h
' g. Failure of reactor safety valve to reseat j
/
- h. Failure of reactor safety system wwm
- 1. Fuel cladding failure J. Loss of feedvater l k. ' Loss of condenser vacuum i 1 Loss of auxiliary power
- m. Instrument air failure
- n. Pressure regulator failure
- o. Energency condenser tube failure
- p. Reactor system ruptures inside the dry vell- .
- q. Failure to replenish cooling water in emergency condenser' rj Startup accident i'
- s. Fuel loading and handling accidents i
- t. Cold water accident.
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'l In addition to those conditions listed above, three equipment failures termed." Major Accidents" were evaluated by the applicant.-
Ttiese accidents included: f
.m-
- a. Control rod drop accident- -.
- b. Main steam line rupture outside the dry well <
)
4
- c. Reactor system rupture in the dry well.
In some of the malfunctions and failures presented, the l evaluation is not yet completed; however, the applicant has stated that when the analysis is complete, the results vill be usedascriteriainthedetailedplantdesign(forexample, to size the pressure relief valves and to set the isolation valve . '
, closure specifications). .
In our opinion the evaluatierothat M*been completed at this time have um formed a satisfactory basis for determining the nature and consequences-a of the maximum credible accident and for establishingproper perspective of the necessity for emergency systems, except for further considerations-
- vhich chould be given to the control rod drop accident. l 1
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- calculations by the applicant indicate that the.most reactive 1
control blade could have a reactivity vorth as high as .036. I Additional calculations show that if this blade were to drop
.i I .E 1 1 free of the core a minimtm period of 3 milliseconds'could' result,
)
and the average fuel temperature vould reach 5500oF in the.un . O " ll o 4 7 controlled fuel zone. The consequences t,og meetQ' - ( in the event of this accident are not datirely clear. .The j ]
/ .
1 applicant has indicated they are developing analytical models- '
- I for more accurate prediction oE the cbasequences of such a 1
nuclear excursion and that the forthcoming SPERT destructive .j s test vill be used to check the modem'that is developed.- ' I l k In addition to the analytical work, a rod' worth minimizer computer and a rod dropout velocity limiter are being developed
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The. rod worth computer would continually monitor control rod patterns to reinforce procedural controls provided to sure that patterns causing . individual rods to assume undesirsbly high reactivity worth are i not used. Conceptual designs for flow restricting devices that vould limit potential control rod dropout velocities to safe p . f,;
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I m ever that by use of these alternatives one :an obtain adequate '
}! assurance that the control rod drop accident would not have consequences as serious as the nav4=m credible accident discussed. An the next section.
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Sb bl WJ/bn Of kg o UM f~ [40t$t thrn.hg VI. Wxtrmrn Credible Acelday,g Evaluation - s for the purpose of evaluating the adequacy of the proposed containment concept, the applicant has hypothesized a major I accident involving a substantial release of radioactive fission products from the reactor fuel, and has estimated theLconsequences of th*is accident in terms of potential radiation exposure to the public, taking into consi-deration moderating effects on such exposures of the containment system and the environmental characteristics of the site. The maximnn credible l tst. accident chosen by the applicant results from an instantaneous complete l rupture of one primary coolant line inside of the dry well af ter reactor l I operation at rated power for an extended period of time when the fission product inventory is at a maximum. The pipe rupture would release i a the pressure in the reactor system (assumed to be at 1250 N,ig), resulting in i r all of the reactor coolant system water flashing to steam; An immediate buildup of steam pressure in the drywell to about 62 psig' would en. sue and pressure would increase in the suppression chamber to about 35 psig. The !
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due to steam condensation in the water contained within the suppression pool. Other assumptions cormerning the magnitude of the accident and the effectiveness of the engineered safegterds systems for alleviating the ' l severity of the consequences are as fcL!ows: l a (1) The loss of coolant from the reactor would result in melting of one-half of the reactor fuel and damage to the cladding in the remainder of the fuel. (The applicant will provide an emergency core spray system f for preventing such significant damage to the core under j these severe conditions, however, for the purposg o.T this j
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' Noble Gases 20 100 '60.
Halogens 20 100 60 Other Solids 0 1 0.5 I u.# ,, (3) One-half of the halogens would be removed in the dry well and suppression pool by11ste out and scrubbing action of the { vater. The remainder of the ha!ogens and all of the noble gases would be "available" for leakage to the reactor building . at a rate of 0.5 percent per day at design pressure. As the l pressure in the dry well and suppression chamber decrease, ~ there would be a corresponding decrease in leak rate. As noted
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I previously, the applicant will be required to demonstrate q the assumed integrity of the dry well and suppression chamber prior to reactor operation by suitable leak rate tests. (4) The reactor building air which is maintained at a slight negative pressure (k" of water) would be exhausted to the atmosphere through f11ters for removal of particulate and halogens, at a volume flow rate equivalent to 100 percent of the reactor building air volume in 24 hours. The particulate and iodine removal filters are assumed to be 95 percent effective. j As has been stated, provisions will be made in the design for i verification of the effectiveness of these filters on a periodic basis. . 4 _
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i The applicant's evaluation of the consequences _of this accident assumed release to the atmosphere through a stack 300 feet high, although'at the present time no specifications have been proposed for the stack design. Assuming that the wind direction and velocity were constant during the course of the accident, the applicant calculated exposures as follows: C.W . 1. For bood meteorological diffusion (lapse) conditions and a wind speed of 10 miles per hour the maximum exposure rate at ground level would occur at a distance of approximately 0.6 miles I from the stack.
- a. The maximum potential dose rate to the thyroid was calculated-by the applicant to be approximately a,g,g.au, 8 millirems per hours (0.008 rems per hour) and the
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- b. 'Ihe maximum potential whole body dose rate due to 1 noble gases was calculated to be~ approximately 2 s a
m1111 rem per hour with a ' total potential dose 'for. the duration of the accident approximately 0.024 rem.
- 2. For moderate inversion conditions with a vind' speed of Q.y 5 miles per hour the applicant estimated the maximum
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3 miles from the site. ' Under these conditions the maximum ' potential dose rate to the thyroid and total dose for the , duration of release was estimated to be less than 40 millirema per hour (0.04 rems per. hour) and 7 rems, 1
, respectively. The whole body potential riose rate and-integrated dose were estimated to be less than 10 1 1 millirems per hour (0.01 rems per' hour) and one rem, respectively.
As previously stated in the section of this report describing the meteorology of the site the staff believes that wind speeds
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during inversion conditions may be somewhat lover hn assumed by .j the applicant in this evaluation. In this regard,'the staff has made calculations based upon the above assumptions which take 0
- ,e into account the possibility for the accident to occur at a vind-speed of one meter per second- (2.2 miles per hour) under either .
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height of 300 feet and the one meter per second. wind speed' i the staff estimates that the maximum potential whole body and thyroid dosages for the duration of the accident are 0 94 rem
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and 53 rem,' respectively. ,i ' h e staff has considered both the applicant'arass'umptions concerning the postulated maximum credible' accident:and the
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the consequences'of such an accident.- As' indicated in the-above analysis, the amount 'of fission product released from i the fuel depends on the extent of ' core ' damage, which in e' . i
' turn, depends on the effectiveness of the emergency core spra.y, i d
The staff believes that a suitable core spray design which votad provide for an adequate supply of emergency cooling water would I as m .. j substantially reduce the extent of damage. to the fuel, even! { to the point of preventing any melting. On the other hand, if. l the emergency core spray failed to function at all the inventory
-Ah e 1 of fission products released from the core would be increased by '
i approximately a factor of two. In this case the dosage values vould be increased to approximately 2 rem whole' body and-to 100 rem to the thyroid. From the aboves analyses the staff has concluded that.the: I! engineered safety features proposed for this facility should be capable of limiting the degree of' harm that could result if an' accident-M such as postulated should occur. herefore, since it is believed
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that the occurrence of such an accident is highly unlikely, the staff concludes that operation of the proposed reactor at the Bodega site would not represent an undue risk to the public.
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VII. Technical Qualifications of Applicant , l h e technical qualifications of PG&E are described in the application for a construction permit. PG&E has constructed and is now operating a boiling water nuclear power plant at Humboldt Bay near Eureka, California. heir principal contractor for the Bodega construction, the General El.ectric Companyjdesigned and furnished the major components of the Humboldt nuclear steam supply i system, including the reactor with its controls and instruments- I l n., %. o.. - tion. GE has also designed and furnished similar equipment for ) several other boiling vater reactors in this country and abroad. l l 1
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