ML20114E678
| ML20114E678 | |
| Person / Time | |
|---|---|
| Site: | 05000214 |
| Issue date: | 07/01/1964 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML093631134 | List:
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| References | |
| NUDOCS 9210120238 | |
| Download: ML20114E678 (21) | |
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U. S. ATOiIC STEPGY CCtetISSIai DIVIbWi CF RFACTOR LICDISD10 REP 0hT TO ADVISORY CCMMITTEE Gi REACTOR SAFEGUARDS
_ON LEPAR'IMENT OF WATER AND PCWER OF THE CITY OF LO3 ANGELES MALIEU NUCLEAR PLANT - LWIT NO. 2 APPLICATIGi F0R CQlSTRUCTIGi PERMIT t
i Note by the Directcr. Division of Reactor Licensing The attached repcrt has been prepared by D. F. Enuth, J. F. Newell, and cther members cf the staff of the Division of Reacter Licensing for-consideration by the Advisory Co:::nittee on Reactor Safeguards at its July 1964 meeting, i
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9210120238 920520 PDR ORG NRCHIST l
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L?RODUCTIO!l..*****..................,,
.,... wa 2
A.
Site...........................
b 3.
Meteorology.......................
6 C.
Geology and Hydralecy..................
7 D.
Seismology d Th?L f _u 8
E.
Effects of Liqui:1 Eff.!c a -
s F.
Reactor..............
P G.
Engineered Saferned 11 (1) Containment...........
13 (2) Air Recircule.1:r Eptm..............
Ih-(3) Contairar.ent 5; ray Sys*en..............
Ih (h) Safety Injection S/stm (5) Emerger.07 Ele;;rica l Paer Cupply.........,
15 15 H.
Seismic Design Criteria.................
16 I.
Accident Evaluation......
18 CONCLUSIONS * * * * * * * * *
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1 ft11bu Unit No. 1
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Introducticn
"%e City of Los Angeles Iepartment cf Water and Power has met with the J
Advisory Cc=mittee on Eeactor Eafeguards at various times in the pastLetters dated Ju to discuss propcsed sitec fer nuclear reactors.
1960; June 27,1960; January 14, 1961; September 1],- 1961; April h, j
1962; October 12, 1962; and Nobe=ter 1k, 1962 have been written on the suitability of various reactors at variove proposed sites.
The ACRS letter of October 12, 1962, is addressed s the suita'ility of a l
double containment ccncept for a pressurized vnter reactor at the At the reactor power level presently proposed Corral Canyon site.
c:ntemplated, use of a conventicnal single containment structure did not meet the criteria suggested in 10 CFR 100 pertaining to off-site radiolegical hazards; therefcre, a double centainment concept was Ite ACRS letter of October 12, 1962 con-preposed for censideration.
cluded that the Corral Canyon cite would be suitable vith this con-1 I
tainment system,
~he Department of Water and Power of the City of Los Angeles has prcposed that a 1h73 Mw(t) pressurized water reactor be located at Corral Canyon which is about 29 miles vest of the center of Los Angeles. An application for a ecnstruction permdt for this facility vas filed with the Commission on November 18, 1963 A report en-J titled " Preliminary Hazards Summary Report, Malibu Nuclear-Plant l
Amendment No.1", was submitted in support of this application.
Nc.1 which contained annual operating costs, requests for specini nuclear material allocation, and legal information was submitted on February 13,196h.. In response to a request for additional technical information on this proposed plant by the Commission, Amendments Numbered 2, 3, and 4 vere submitted on May 7,1964; May 21,1964; and l
June 3, 1964 respectively.
The technical review of this facility is based cn these documents.
The Department of Water and Tower it the evner-general contractor and vill design and construct the convantional portica of the plant.
A project manager has been designated by the Department of Water and Fever who is responsible for-the overall project including review of 4
centractors work.
The nuclear steam supply system vill be designed and constructed by Westinghouse Electric Corporation and Stone and Webster Engineering Corporation.
The most novel feature of this facility is the reinforced concrete containment structure containing two steel membranes which separate the containment volume from the environment. The spac6 between the tVo steel membranes Vill be maintained at a negative pressure with a pump back system during norual operatien and all accident conditions, and hence the applicant believes that direct leakage from the containment volume vill be essentially zero. Also proposed as engineered safeguards for this an air recirculation system, a safety centainment and reactor are:
injection system, and a containment spray system.
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Our approach in this evaluation has been to determine how much reliance would have to be placed on the engineered saf eguards in order for this f acility to meet the siting criterist and secondly, to determine if the criteria for designing of the engineered saf eguards provides suf ficient assurance that the necessary reliance on engineered saf eguards can be realized in order that a severe reactor accident will not result in unacceptable of f-site radiological hazards.
In the subsequent sections the site, the reactor, and engineered safeguards will be briefly described with our appraisal of particular features with which we have had reserva-tions. The last section on Accident Evaluation summarizes the calcula-4 tions that the staff has made of the of f-site consequences of the " maximum hypothetical accident" (primary system rupture followed by 100% core meltdevn).
Discussion A.
Site The proposed reactor site is at the mouth of Corral Canyor. about three miles from the comunity of Malibu Beach, tt:n miles from Santa Monica and about thirty miles vest of downtown Los Angeles.
U. S. Highway Alternate 101 runs alon8 the south edge of the site between the Pacific Ocean coastline and the reactor location. The Santa Monica mountains, which are parallel to the coastline, lie immediately north of the reactor site and rise to a he1Cht of about 2500 f eet separating the site f rom the San Fernando Valley.
The Department of Water and Power is presently in the process of acquiring approximately 305 acres of land in order to provide an exclusion distance around the reactor f acility. The area under acquisition is roughly semi-circular in shape with U. S. Highway Alternate 101 along the diameter (about one mile). The reactor site at the mouth of Corrat Canyon is about equidistant from the semi-circular boundary and apprcrimately 350 f eet f rom the edge of the highway.
The nearest population to the reactor will be in a triangular section of land which lies between the highway and the Pacific Ocean. Approxi -
mately 200 people live on this parcel of land and the saturation population of this secticn is said to be approximately 400 people.
The distance to the nearest residence from the reactor containment building is approximately 1100 f eet although the shortest distance to this criangularly shaped parcel is approximately 850 feet.
The present population distribution (1960 Census) within thirty miles of the reactor is listed in the following table.
3-TABLE _I_ - POPULATION WITM1H E IRTY M112S OF TRE REACTOR BUILDI Popolation D,1etance 6.200 5
11,900 10 100,800 15 700,000 20 25 1,805,000 3.100.000 30 According to information provided by Los Angeles Department of Water and Power when this site was under consideration by the Committee in 1962 approximately 507. of the population within five miles was in the vicinity of Dume Point three to five miles vest of the site; 251 in the Malibu Beach cocaunity; and 25% along the beach and in the mountains surrounding the site.
The population projection for 1980 within five siles of the reactor site provided by Los Angeles Department of Water and Power in 1962 and based on the los Angeles County Regional Planning Commission estimates is as follows Radius (Miles) 1980 Estimated Popuistion 0-1 5,000 1-2 9,000 2-3 11,000 3-4 21,000 4-5 25,000 The Malibu Citizens for Conservation in a letter to the AEC dated June 4, 1964, estimates the 1980 population to be 3,200 and 11,800 at one and two miles, respectively of the Corral Canyon site.
t Based on the configuration of the property under acquisition by LADWP, the exclusion distance for the proposed reactor site is considered to be 850
_ S f eet (0.16 mile), which is the t41 stance to the nearest boundary of pri-vately owned property. When cempared with the exclusion distance of approximately.88 mile based et. the assumptions of TLD 14844 for the size of reactor proposed, it is apparent that extensive engineered saf eguards vill be required for this reactor in order to meet the requirements of Part 100. In this regard, calculations indicate that a reduction in containment leakage by a f actor of approxinetely thirteen would be necessary. As discussed in another section of this report we believe this can be achieved by the types of safeguards proposed.
On the basis of the existing population, the lew population distance presently available at the proposed site is approximately ten miles, although on the basis of future population estimates this distance could be as low as about four to five miles by 1980 (within about ten to twelve years f rom the tice the plant would be in operation).
Since TLD 14844 assumptions would result in a low population distance of approximately 13.3 miles, a reduction in potential fir.sion product release to the environment by a f actor of approximately six would be necessary to satisfy a 16v population distance of four miles.
It is apparent that enginaered safeguards which would be adequate for meeting the Part 100 dosage criterion at the exclusion distance of 0.16 miles available at this site would also satisfy the dosage criterion for a low population zone distance of four miles.
B.
METEOROLOGY Southern California, which is noted for its mild climate, is not subjecc to the severe storms such as tornadoes, hurricanes, and severe f rontal passages that occur in other parts of the United States. How-ever, because of the site location along the Pacific Coast close to the mouth of a steep canyon, the air circulation pattern at the proposed reactor site at Corral Canyon is complicated by the mountain valley air circulation and channeling of air flows in the canyon valleys.
Wind rose data is available from two wind measuring stations, located near the site at Zuca Beach five miles west of the site, and at Malibu Beach three miles east of the site. Although these data are not adequate for predicting the precise flow patterns of air transport f rom the site, they indicate the general influence of the sea breeze and the rountain-valley winds, with prevailing air flow towards the Santa Monica area and possible recirculation of air over the Malibu coastal area.
In order to obtain more precise data, the applicant has proposed to conduct a meteorological sttdy to determine these f actors as well as the atmospheric diffusion charseterisites pending completion of arrangements for acquiring the site, rd particular interest in regard to the meteorology of this site is the well knmm subsidence inversion which serves as a lid over the entire Los Angeles coastal area as f ar as air pollution climatology is concerned.
Information provided by the U. S. Weather Bureau indicates J
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thet the setsidence in+ev vien may rettrict the vertical dif fusion significant percent of the time when the inver-of an aisborne ef'. vent 4
sion base is below the ilevation of the mountain range immediately inland f rom the sit e.
Since the subsidence inversion lid is usually a f ew hundred f eet above the ground surf ace, this restraint would not generally affect dilution et riose-in distance when the air flow is Hevever, vnen the air flow is inland parallel to the coast.
towards the mountain range, the dilution of an airborne effluent would be restricted.
The minimum dilution to be expected at the site, i
however, would generally be ersociated with an intense ground level inversion such as occur a at night anywhere in the United States.
l The atuospheric dif fusion equations and parameters normally employed for estimating the evneentration of materials released to the atmosphere can be applied at this site for the close-in distances, and the dif fusion paramet tre proposed by the applicant in this regard are considered to c;nservatively represent the amount of dilution to be expected under these conditions. However, because of the oscillatory nature of the land sea-breeze phenomena which results in a relatively strong on-shore breeze during the day and a weak off-shore breeze at night, and the presence of the subsidence inversion, the stardard dif fusion equations cannot be relied upon for estimating the expected concentration of atmospheric releases at this site for s
travel distances greater than a mile or so from the plant.
I The applicant proposes that gaseous effluents produced during routine operations wm:1d be retained for decay of short lived nuclides and then released under taeteorological control so that exposure of off-site areas would not exceed fif ty millirem per year (a factor of ten lower than Part 20 requir ements).
With respect to the maaicum hypothetical accident the applicant pro-poses to retain all fission products that may be released to the containment until the short lived nuclides have decayed to insignificant levels. Af ter this decay period the applicant then proposes that controlled purging of the remaining gaseous activity could be accomplished within Part 20 requirements under the most f avorable atmospheric dilution conditions.
apparent
- hat the curtrelled releases of ef fluents from routine It 1:
operations as well aa from the hypothetical accident vould depend on a detailed knowledge of the atmospheric dif fusion conditions such as air flow patterns, vind spenda, and stability considerations. _In this regard the meteorologiesi program that is proposed by the applicant would be important in establishing the t:6ais for controlling these releases.
Based upon comments provid(2 to us tiy the U. S. Weather Bureau we belie-6 that a program such as proposed by the applicant would be adequate'for this purpose.
. CE01.0GY AND HYDROLOGY The proposed site is underlain with a bedrock known as the Monterey formation, which consists chiefly of interbedded sandstone, siltstone and mudstone, overlain by about ten to sixteen feet of quarternary terrace deposits of si.:y-clay and silty-sand containing gravel, cobbles and boulders. The reactor plant would be founded in the Monterey bedrock formation which is considered to be an adequate foundation material for this plant, although the bedrock formation has been tightly folded and moderately to intensely sheared due to the seismic history of the area.
The features of geology and hydrology that are most important from a saf ety stt' point relate to the nature and extent of landslides and the relation of the Halibu whether they might endanger the plant; Coast f ault to the reactor location and tha liklihood of differential ground motion through the plant due to earthquakest and the possibility of flooding of the plant due to a cloudburst type storm occurring on the Corral Canyon watershed.
At the request of the staff the li. S. Geological Survey has examinpd the extent of these problems and has advised the staff as follows:
Landslides cov?r about one-third of the site area and are most 1.
subject to initiation when the ground is saturated, and during periods of ground vibration such as might be caused by earth-quakes. Bovever, f ailure of steep slopes may occur at anytime.
The bluf f s that flank the mouth of Corral Canyon on either side 2.
The of the proposed reactor site are both active landslides.
direction of movement of these slides, is towards the coastline and theref ore would not af f ect the reactor plant. However, it will be necessary for the applicant to provide safely designed slopes for these canyon valls if they are cut during grading activities to avoid the possibility for initiating conditions which could result in land lide damage to the plant. The applicant has indicated that he is aware of this problem and will take whatever action is necessary in this regard.
3.
There is no likelihood f or landslide movement of the bedrock formation in the bed of Corral Canyon.
During an intense rainstorm the possibility exists for a landslide 4
to form a dam in Corral Canyon up-canyon f rom the reactor site.
This could result in a flood with an estimated flow of 10,000 cf s when the landslide dam vashed out.
Flood flows in the range of 1500 cf s to 3000 cf s can be expected in Corral Canyon once in fifty years. It is our understanding that the applicant is deve-loping a plan f or drainage of flood waters in Corral Cr.nyon although the criteria for this design has not been furnished at 4
this time. However, we see no reason why a suitable design could not be developed which would be adequate for prevesqng hazardous operating conditions for the reactor due to flood flows in the Canyon.
5.
The reactor site is situated in the Malibu fault zone which is a vide structursi zone that has been cut by several f aults, trending east-west for at least twenty miles along the Malibu coastline. The surf ace trace of the Malibu Coast f ault lies -
approxinetely 500 f eet north of the proposed site. According to the USGS, all of the known ground displacement on the NW11bu coast fault zone occurred in prehistoric time--that is, sometime between about 200 and 400,000 years ago. Based on the frequency and severity of earthquakes along the Malibu coast and related f aults and the lack of evidence of surf ace ruptures within historic time, the USGS has concluded that the probability of ground dis-placement at Corral Canyon within the next fif ty years is very low. On this basis, the staf f has concluded that the probability of potential hazard to the public from differential ground move-ment due to en earthquake at this site is low enough to be disregarded.
D.
SEISMOLOGY As indicated in the geology section of this report the proposed site is in one of the active seismic areas in California. It is located close to the trace of the Malibu Coast f ault, about forty miles from the San Andreas f ault, and about twenty to thirty miles from the Santa Ynez and Inglewood faults.
I urvey has At the request of the staff, t he U. S. coast and Geodetic c
examined the seismicity and ttunami potential of the site'und has advised as follows:
1.
Based on the past seismic history the site vill be subjected periodically to shaking from earthquakes occurring along the f ault systems in the area. In this reggrd, however, the magnitude of earthquakes that occur along the minor fault systems near the site are in the range of about 4 to about 6.5.
Earthquakes of Magnitude 7 and above are not expected to occur along these minor fault systems, but do occur along the San Andreas fault. The caximum ground acceleration likely to be experienced at the site would be due to a large earthquake along the San Andreas f ault instead of the smaller earthquakes along the minor faults near the site. Based on these considerations, the U. S. Coast and Geodetic Survey has estimated that the site would be likely to ex-perience a maximum ground accelerations of 0.23 g in the period range of 0.3 to 0.6 seconds, but that an acceleration of 0.3 g in the same period range is possible and should be taken into consideration in the f acility design.
For the purpose of earth-
i 8-1 i-quake design the applicant has assumed a maximum credible ground motion corresponding to an acceleration of 0.3 g., which compares f avorably with the value predicted by the U. S. Coast and Geodette Survey.
With regard to the potential for a tsunsai striking the Malibu 2.
coastline the U. S. Coast and Geodetic Survey believes that the potential exists for water encroachment to a height of fif ty f eet above meta sea level.. In this respect the U. S. Coast and Geodetic Survey does not consider that a seismic es wave will-strike the site with " destructive force" but instead that the-tidal effects could result in a build-up in water elevation of this amount, with i
the period for buildup. and recession being anywhere from a few minutes to approximately forty-five minutes. Since the proposed l
grade elevation at the site is thirty-five f eet above mean tea level a rise in water elevation as predicted would result in the site being inundated to a depth of fif teen feet. The applicant has been requested to describe the saf ety considerations involved if a tsunami of this proportion were to occur, and the criteria they, would apply for minimizing hazards to the public from inundation of the plant.
E.
Ef f ects of Liquid Ef fluents on Fish and Shellfish.
l The Department of Water and Power has proposed to control the concen-tration of radioactivity in liquid wastes released to the ocean environment in the condenser cooling water discharge line to one-tenth of 4
the amount permitted by Part 20 of the Cennission's Regulations.
We have requested comments from the U. S. Fish and Wildlif e Service on s
these aspects of the reactor operations and have been advised that adverse effects to the fisheries would not be expected due to the radioactivity releases proposed, but that the greatest effect the proposed plant might have on aquatic lif e would be due to the tempera-ture the condenser cooling water released to the bay. In this' respect, however, there is no reason to expect that the temperature effects would be any different from this plant than from a fossil fueled plant of similar capacity.
4 The Fish and Wildlife Service has reconnended a preoperational monitoring program to begin two years in advance of the plant operation to include ecological surveys, radiological monitoring and ' oceanographic studies.
The applicant has stated that he will cooperate with local regulatory agencies in order to astabitsh parameters for a cuitable radiological monitoring program, and that a consultant has been retained for the oceanographic aspects, especially related to prob 1rms concerning the sea water intake and the condenser cooling water discharge structures.
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Reactor i,
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The proposed Mslibu reector vill in many respects te similar to -
4 that propcced for the Connecticut Yankee plant,.which in turn is~
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similar to that proposed fcr the Southern California Edison plant, j
The reacter vill be c Ik?3 N(t) precourized water -esctor utilizing i
j dissolved boric acid in the-mcderator, in addition :: mechanical centrol reds, for reactor cortrol. Four parallel' recirculating pri-mary coolant loops-ree cennected to the reactor; ehch loop contains j-a recirculation pump and a ctesn generator.
N prc;osed loop.
i design does not include tic 6k valves nor check valves in the primary loop piping; hence it is not possible to isclate a loop -
from the reactor.
One desirable attribute of this feature is that
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i e severe cold vator accidem is not possible with this design.
Without valves, however. n led in the primary systen due-to 'a f
piping failure er steam generator failure cannot-.be isolated.
In some cases, this.could make the consequences of such a failure
.L more severe.
Although the etaff believes it vould-te more prudent to have loop isolation valves, we have not found any serious i
pctential safety problems which vould require their inclusion, f
l The proposed fuel for the first core is sintered ~ UC: contained l-within 16 mil thick stainlect steel cladding. A to al of 156 i
fuel assemblics are planned ror this core with each assembly containing 217 fuel rods. Eclear coefficients are expected-to-3
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be negative except for the coclant void and te=pera:ure co-
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efficients which may te nero. It la anticipated ths; three. separate i
enrichmnts vill be used in three concentric fuel regions. Each region would contain 5 ' nezenblies with enrichmente from the l
center outvard of 3.k%, 3.6% and k.2%.
Refueling c' the core i
vould take place with an invard loading schedule; i.e. at. each refueling the 52 assemblies in the inner moat fuel region vould i
te removed and the second tc.d third regions vculd te moved-invard j-with new fuel-loaded in the cutermost region. The'geal exposure is stated.to be an average Mrnup. of 10.000 MdD/T. Although stainless ateel cladding is planned for the'firnt cycle, it.is=
l anticipated that zirconium alleya may be used in fu'ure cores.
i The applicant hac atated. that one criteria for desig.ing the j '.
. containment structure ic the-possible energy contritution' from a metal-vater reaction. Che ataff has asked for -additional in-I formation on the rate of potential metal-water reactions, and the energy that vould be liberated as -a consequence; hevever, the
-ansver in' Amendment No. 2 indicated that the applicant's analysis is not ecmplete. 'lhe answer further states that an amendment-vill he supplied at some later date. In a meeting with ;ie appa.icant on June 16,196k, we were advised that the metal-vater reaction studies with circonium would not be completed in. the near future,
. and that an amendment would not be filed until after any public L
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~ hearing. Accordingly, we telieve that a,$udgement cf the adequacy L
cf the containment ehculd be limited to the considentions of a stainlees steel core.
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t Central ef this preposed reactor vill be with a sol.ile poison in i
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the moderator and with silver-indium-endmium alloy c:ntrol rods.
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The centrol rods will te-clad with stainless steel : minimize the
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contamination of the system with activated silver.
3elow the..
absorber section cf each centrol rod is a follover fabricated of l
Investigatien in in progrees to replace this follover j
zirealley.
Each centrol rod will be positioned with a fuel bearing material.
ty a magnetic Jack drive mechanism which is mounted on the reactor j
The reference core sheva a total of 32 control rod i
vessel head.
The con-positiens but this may be changed in the final design.
l trol rods vill have sufficient worth to control the short-term core reactivity changes and to render the~ core suberitical under l
operating-conditions by a safe margin in the event cf a scram.
j The soluble neutron poison vill be used to adjust f:r long-term re-i E e maximum single control rod setivity changes due tc burnup.
vorth has not been repc2ted.
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The staff asked for an evaluation of an accident-due to ejection cf a control rod, and an evaluation of an accident where cold, i
The results:of non-berated water is injected into the core.:
vas stated these evaluations have not yet been Teported, but i l
by the applicant that in the case of tue control rec ejection l
accident, the consequencea vtuld be less severe than the " maximum hypothetical accident". ~ Mearc of limiting the consequences of i
both of these type accidents were enumerated by the applicant to i
l include administrative rad worth limits and mechanical locking devices in the case of the control rods,and possible means of
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e of the unbersted water injecting additional beren in disagree with.the applicant's i
injection.The staff has no 2eaean s conclusion as to the upper litit of consequences cf these accidents,
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however, we plan to continue our evaluation in these areas.
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~~esent plans for mcnitoring the neutron power level and flux shape within -the reacter are with 6 external monit:rs; h around l
the periphery of the core,1 at the tcp cf the core,Jand 1 at We sre not in a pcsition at this time the bottom of the cere.
to agree with the applicant that sufficient reliance can be
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i placed on calculated flux chapea to. reliably predict important parametern such su burnout margins with ohly these instruments.
We believe that conditions are present (large core, high flux, and means cf perturbing the flux) which could induce xenon spatial escillations within the core. - With cnly 6 external monit:rs, and ne in-core monitors,. detection of the existence or absence of such oscillations may be difficult. The applicant has stated that at approximately 18% cverpower, the margin to departure from nucleate beiling _ is calculated to te 1.25 (H.DIGR).
Frcm this, it would then appear that flux oscillatiens causing local power increases cf this magnitude could cause potential fuel tube turnout problems.
We 'are continuing our evaluation in this area, but believe that this,; articular pr:blem need not j
be recolved at the construction permit stage, s
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Encineered Safegardr.
5 (1) Centainment The containment etructure prcroced for this res: tor is in l
many recpecta similcr to that pr:pc.'ed for the.Mvensvcod j
The contain-ent building vill te a cy* indrical resctor.
reinforced concrete structure witn a hemispheri:al concrete I
'"he inside diameter vill te 135 feet and the inside d ome.
There.till be tvc steel membranes e
height vill be 190 feet.
l of at least 0.25 inch thickness separated ty 2*i" of pop-d corn concrete, hternal tc the :vo ':.embranes :: the cylin-t drical portion vill be k'2" cf neraal reinfereed concrete to provide etructural strength. The hemupheri:al dome vill 2
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te constructed of.3'2" of r,crmal reinferced cen: rete.
reinforced cencrete vill te designed to meet the require-tents cf ACI-318 of the bilding Code Recuirements for Peinforced Ccncreto.
'Ihe reinferaed concrete stncture vill be designed to vithstand an internal pressure of ho psig.
In sddition, the structure vill be designed so that in the event-o.,
i the " maximum hypothetical accident", the stresses due to internal pressure and themi strerse,a from the accident I
plus the seismic stresser from an equivalent c.;0 E accel-eration result in otrenes not to exceed the yield point.
Additionalco=eatsentheseismicdesigncritegiaare After presented in Section D " Seismic :esign Criteris..
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construction is complete the s,pplicant. propeses to proof i
test the building at 11% cf design pressure.
'Ihe leak tight inte5rity of the tuilding vill te provided by the double steel membrance. After each merirane is con-structed, prior to pouring the reinforced ccncrete, the space between the nectrunes will be pressurized with approx--
imately 2 psig of a heon mixtre.
Each memb:sne vill then te checked with a halide 1%k detector fcr the presence of Icaks. After the reinforced concrete structmre has been completed an integrated leak' test vill be perf:rmed on each
- .embrane.
Ihe specificatien fcr leakage is th=_t each membrane j
shall leak no more then 0.1 percent cf the certained volume in one day at a differential pressure of 15 prig.
I During operation the popec:r. cencrete zere ber.een the steel membranes vill be maintained at a negative pressure of approximately 4 inches of water. Three ecmpressors each rated at 10 CDI will b( med te maintain the regative pressure zone by taking cucticn frem the pervi:us concrete zone and discharging to the inside of the enntainment structure.
During normal operation partial intermittent l_
to handle operation of one comprecscr vill be cufficien:
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leakage through both rembranes (as-predicted fr:: 0.lyday at 15 psiE) and under accident precsure conditi:ns the applicant states that the leakage should not ex:eed the capacity of 2 compressors. _ A gas meter is installed on the pumpback system in Order to permit _a continuouz cumulative -
assessment of the total leakage into the pervi:us concrete
- ene.
By either monitoring the pressure rise in the '.nner zone or by recording the volume of gas purged to the stack to maintain a uniform pressure in' this zone, it is theoretically-possible to determine the leakage through the : uter membrane.
l the total inleakage to the pervious sc:e and the Kncvin6 outer membrane leakage, the inner leakage can te obtained Se leakage limits, as deter ined by this by subtraction.
method, ct which reactor shutdown vill be necessary vill be set such that the extrapolated leakage through both mem-branes at accident ecnditions will not exceed the capacity An example of_ the methods c:ntemplated of 2 compressors.
for performin6 leelage extrapolation was presented in answer No. 2 of Amendment No. 3 More conservative extrapolation factors used by the Staff -(resulting in higher leakage rates at accident conditienc) indicate that the leaksge could exceed the applicants criteria.
This particu' tr problem could be solved by either increasing the capacity of the pump-back compressors or changing the allovable no-al leakage rates;however, we de not believe this has to te resolved at the construction Ier:1t stage.
Under accident pressures with the compressors :perating, any leakage through the membranes vill te retrned' to the inner containment volume'.
Leakage through the outer mem-brane at the design leak rate would add air t: the inner volume at a slow rate; over-pressurization of the contain-i i
ment is calculated to require hundreds of days.
The conse-quences of simultaneonc pumpback fa_ lure with the " maximum hypothetical accident" are discussed in Secti:n E " Accident Evaluation".
Penetrations which pa:s through the. containment structure l
l are to be provided with sleeves which are vented to the All piping penetrati:ns that pervious concrete zone.
(1) are not normally vented to the pervious ::ne (e.g. ventila-l' tion ducts), (2) which may cc:=unicate with -f:e containment atmosphere in the event of an accident, and, ;3) are not used in an accident '(e.6. cooling vater lines. circulating cooling water) are to be protected'vith a block valve seal water injection system. Eis syctem utilizes isole.; ion valves on the pipin6 on each side of the containment penetration, cnd -
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i a cyetem to inject vater between these valves at a pressure greater than the internal containment accident pressure. With this system unmenitored leakage which could bypass the negative We believe that this pressure zone is greatly minimized.
system, which was not proposed for Ravenswood, greatly: reinforces the applicant's argument that the containment outleakage vill te essentially zero.
One system, the pump tack ecepressor system, has been identified i
as a system from which significant amounts of direct leakageThis from the inner containment to the environment is possible.
system is located in the vaste disposal building which is maintained The at a slight negative pressure and vented to the stack.
consequences of a leak from this system corresponding to 0.1% per day of containment volume referenced to 15 psig, directly from the i
inner containment to the stack are evaluated in the " Accident Evaluation" Section. Althou6h this amount of leakage may be considered-tolerable, we are continuing our evaluation c: the feasibility of requiring provisions for periodic leak tests cf this system during normal o;.eration, (2) _ Air Eecirculation This system has a dual function.
It is used during normal operation to control temperature within containment, and would be used under The post-accident conditiens to remove heat from he centainment.
thick reinforced concrete containment struct; e.is a very effective thermal barrier and were heat removal systems not provided to -
transfer energy out of the structure, continued after-heat genera-j tion vould cause additional vaporization of spilled. coolant and result in over-pressurization of the centainment vessel.
It has been calculated that with no heat removal frec, nor water injection into the containment volume; within about one hour, the containment pressure in the event of the " maximum hypothe.ical accident" would be about 150% of design pressure. - It is the-efore necessary that this engineered safeguard (or the. containment spray system which.is also capable of removing heat from the building) centinue to function properly e.s long as necessary after an accident.
The air recirculation system vill consist of four identical parallel Power for units, each of which contains a fan and heat exchanger.
these fan units can be obtained from the 230 kv transmission lines, 3h.5 kv auxiliary cables, or from the-on-site emergency generator.
Heat transferred to the heat exchangers is nc= ally transferred to the component cooling water system or to an s.lternate supply of water. These unito will be functioning during normal reactor operation, hence their usability under accident conditions is better assured.
Staff calculations indicate that a: least three of these units must continue to operate for approxima:ely 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the
" maximum hypothetical accider}t"; thereafter, two units must continue -
to operate for approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />; and after about.20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> one unit should limit the building pressures to less than the design pressure.
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1 (3) Containment Spray System This system is capable of performing the same function as the J
air recirculation units of transferring heat from the containment in the event of an accident. This system when used in conjunction 4
with the air recirculation units would reduce the containment pressure more rapidly. The system is. actuated manually as desired by the operator. The system as proposed 'is an internally located 4~
system which consists of two 1600 gpm pumps taking suction from the containment vessel sump and pump through a residual heat i
exchanger to a number of spray nozzles spaced through the contain-
-- ment vessel in order to provide for cooling and _vashing of internal surfaces. As with the air recirculation system, power is available Heat to operate this system from the on-site emergency generator.
transferred to -the internal heat exchanger is also via the componenti cooling water system.
Flow testing of _this system is not possible during reactor operation i
because the sumps are normally cry. Maintenance of any component of this_ system is not possible in the event of an accident (or-during normal reactor ' operation) because all components are located inside the containment structure and hence inaccessible. The staff believes that location of these components outside the containment vessel with provision for connection to the refueling storage tank, an is the case in the Connecticut Tankee design offers a much more reliable design; however, location of these components inside the containment vessel may be 0,astified provided design considerations include adequate protection for the system, ano menas are incorporated for testing for system operability to meet the reliability _ required uncer accident conditions. The applicant has stated that the external location of components would give rise to possible unmonitored leakage paths outside. of containment due to learing equipment and as such precludes external location. We are still evaluating the adequacy of the design of this system.
l (h) Safety Injection System The applicant has included an additional engineered safeguard for which credit was not taken in the analysis of the " maximum hypothetical accident". This feature is the safety injection 4
system which will pump borated water from the refueling storage tank into the reactor. In the case of the maximum credible accident, as defined by the applicant (breax of_ the largest pipe connected to the primary system), this system is designed to be adequate to prevent any core meltdown. Power for this' system comes only from off-site supplios.
As proposed, this system consists of two safety injection pumps which tske suction from the refueling water storage tank (which is borated) and supply _ water through suitable valving to the feedwater pumps. By closing the normal valtes in the feedwater pumps and opening other valves, two of these pumps will pump directly
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into the four prieru7 recirculation loops..It is not clear what the censequences vould be if the normal supply of vater to the feed i
i pumps remained open (demineralized, unborated vater in the secondary The boron system) and this water were injected _into the reactor.
liquid poison is worth on the order of 20%
6%/k in the cold, clean condition (excess reactivity cf 25% vith about 5% held in cont ~rol' j
rods); however, we do not know the maximum rate of reactivity addition Additional i
or total reactivity insertion possible with this system.
information on this potential reactivity transient has been requested of.the applicant.
(5) Eoergency Electrical Power Supply As described above some of the emergency systens can be connected to the on-site, h80 volt emergency generator. Recent discussions were held with the applicant on the size of this unit and the 4
expected load in the event of loss of off-site power under (a) nor:::al operationi and (b) accident conditions. Oral infomation indicated that the capacity of on-site emerSe:Cy Generator was bein6 increased to supply approximately 1000 horsepower.
This amount of power should be adequate for decay heat removal folleving Under accident conditions, the power requirements are a scram.
increased (h air recirculation fans alone recuire about 800 horse-i power) and these conditions vill determine the required on-site generator capacity. We believe that sufficient emergency on-site power should be available to remove decay heat under both normal a'
and accident conditions and are continuin6 our evaluation of the power requirements to assure a sufficient degree of reliability.
3 D.
Seismic Design Criteria For the purpose of seismic design, plant structures have been categorized as follows:
j y
Class 1 Those structures and components or parts thereof whose
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failure might cause or increase the severity of a loss of coolant accident or result in an uncontrolled release of excessive amounts of radioactivity.
Class 2 Those structures and couponents or parts. thereof whose f s11ure could not cause nor increase the severity of a loss of coolant accident nor result in an uncontrolled release of excessive amounts of radioactivity, but could interrupt power generation or result. in a controlled plant shutdown or scram.
Those structures and components or parts thereof whose Class 3 f ailure could inconvenience operaticn, but which are not essential to power generation.
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. For Class I structures, operating stresses pulse seismic-stresses will be maintained within allowatie working stresses for a 0.3g eartnquake response spectrm. Purther,- operating stresses when combined with seismic stresses equivalent to a 0.6g earthquake response spectrum will-be such that the function of the structure or component will not be impaired.
For Class 2 structures, the operating stresses and seismic stresses for a 0.2g earthquake response spectrum will be maintained -
within allowable worring stresses. Class 3 structures will-be de-signed in conformance to code requirements.
In consideration of the comnents of our consultants, the desigr, approach i
for these structures appears reasonable with one exception. Although i
the containment structure is categorized as a Class 1 structure, for design purposes special crit.eria are applied. For the containment structure, it is propwd that operating stresses, including _ the in-ternal pressure anc ;,hermal stresres from the "mmr%u hypothetical accident" combina with seismic stresses will not exceed the yield stress for t. 0 3g eWquake response spectrum. Since we believe_that the internal m essure from a " maximum hypothetical accident" should be cons 2m ed as an operating-stress, there is not-the same safety factor preunt in the design of this structure as fcr other Class 1 structures.
urther clarification on the matter has been requested.
E.
Accident Evaluation In the Preliminary Hazards Summary Report and subsequent amendments, tne applicant has evaluated the off-site hazards _ associated with potential radioactive efflumt releases. - Although this section of this report is-concerned primarily-with possible accidental releases of radioactivity, it is appropriate here to mention that the applicant's criteria for routine racioactive gaseous disposai from the stack is to limit possible exposures to less than 50 mrem per year.
The criteria for normal liquid waste disposal in_ the concensa effluent is_ stated to be 10% 'of the values permitted in 10 CFR 20.
The accidents which have tieen considered by the applicant include:
(1) Steam generator tube rupture (2) Waste disposal component failures (3)
Fission gas released from fuel stored in. the spent fuel storage pool (h) Accidents with the reactor, with the most serious considered the_"maximu-hypotnetical accident" The analysis of the first three accidents listed above was presented in answer number 15 of Amendment Number 3 The most severe of these accidents in terms of off-site-radioactive hasards was not substantially more severe than is permitted in terms of 10 CFR 20 criteria (0 5 rem as averaged over a year). The staff has checked the consequences of
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licant.
these accidents and eencurs $n the analyses performed by the app A' number of potential accidents of greater consequence were evaluated t
such as those initiated by a steam line break, loss-by the applicant, As was previously de-j of coolant flow, and reactivity insertions.
scribed, we have asned for additional _information in some of these 1
areas (e.g. rate of reactivity insertion possible due to unborated -
i water); however, the applicant has stated,. and we agree i;
be greater than any or these accidents.
i as is the case for recent For purposes of clarity, this application, Westinghouse reactors, terms the maximum credible accident as the For severence of the largest pipe connected to the reactor system.
j breaks of this size, the safety injection system is designed to have For the Malibu reactor i
sufficient capacity to prevent any clad melting.
j the " maximum hypothetical acci lent" is defined as a rapia primary _
Although not coolant release with near instantaneous meltdown.
specifically stated, the pressure rise in co..tainment corre j
rupture of a primary coolant line, and is the accident usually considered by the staff as the maximum credible accidert for a pressurized water i
For the "r.aximum hypothetical accident" as presented by the applicant, off-site doses dere non-existent bec,ause of the zero reactor.
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containment leakage and the reinforced concrete structure attenuation The staff has asxed for of the potential direct gimma exposures.
additional infonnation on potential dose calculations from such an j
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accident for (a) ground level Aeaxage as a consequence of pump back comprassor failure, and (b) elevated leakage to the stacx as could We have also asked for an occur from pump back compressor leakage.
estimate on the magnitude of _ leakage' from the pump bacx compressors that coule remain undetected during normal operation.
Because we have not fully accepted the principle of absolutely zero i
outleakage from containment in the event of a serious accident, we have performed calculations of the potential off-site doses to de -
termine_ how close one must approach the zero outleakage concept for -
Preliminary calculational results are i
this reactor at this site.
presented in Table I which includes an explanation of the calcu-lational models. These calculationq show that for the " maximum hypothetical accident", if a single containment structure were proposed for the Malibu reactor, the doses woula be higher by. a factor of 11 to 50 (dependtng on tne point of leakage from tne One conserv-structure)' than that usually considered acceptable.
If one ative factor is the assumption of instantaneous meltdown.
the doses would be reduced computes meltoown as a' function of time, by a factor of-approximately 2, thereby reducing these factors by a like amount.
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L The dual me.erane containment concept actually proposed for-this reactor could provice the ceeded reduction factors. As.previously ciscussed,-if grout.d-leazage occurred an ' additional reduction factor 2
of about 25 over that proviced by' single containment would be re-With the present seal water injection: system pmvided quired.
for the ' pipe penetrations, there axe no penetrations which -have i
been identified vnich would cause significant leakage at ground A severe test of the containment would be to consider a level.
simitaneous pump back compressor system failure with the-a phm hypothetical _ accident". Under these conditions the rcombined leakage would result in a reduction factor of approximately I
75 for the first two hours. We therefore believe that this contain-ment is adequate from the standpoint of potential ground level i
leagage.
For an elevated leax to the stack considering the same accident conditions as discussed previously, :the. dual membrane contain-ment would have to have approximately 6 times the leak tightness l
of that for a conventional containment l structure. One system (the pump back compressor system) could leak to the stack; however,
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component testing, as orally proposed:by the applicant,_ could be expected to detect leakage before it would.- approach these j
In addition, there are filters in the _stacx which, with levels.
reasonable efficiencies, could provide the required reduction factor.
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One feature -that the proposed Malibu containment design provides over other conventional containment structures is the fact' that dual membranes, separated by approximately. 2'6" of concrete, are proviced. If an accident could occur (which.has traditional 1Z,_ -
l been considered incredible) where missiles are, formed that would
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penetrate the containment shell, this reactor design provices an additional though uncalculable factor-of safety. -
s Conclusions
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j As described previously, we have some reservations concerning specific design features -of' this facility and further information will be pre-sented by the applicant in these areas. Assuming that the problems i-
-discussed are resolv9d by information presented in the expected j
amendments, we believe that there is reasonable assurance that the j-facility can be-built ano operated at the proposed Corral Canyon-site without undue risk to the health and safety of the public.
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1e' Containment Barrier Table I --Dose Calculations Based on a Sint:
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i 2 hr_ thyroid-
'2 hr whole Maxianum reduction l
dose, rem body, dose, factor necessary rem-to meet site i
criteria f
(a) Instantaneous meltdown with.
3,900 30 13 l
0.1% per day leakage at f
ground level I
(b) Instantaneous meltdown with 15,200 125 50 containment. pressure as a j
function of time and ground level leakage at 0.1% per j!
day at 15 psig.
l (c)* Instantaneous meltdown with 3,200 25 11 containment pressure as a l
function of time and a a
150 foot elevated stack release at 0.1% per day i
at 15 psig-l
- This dose occurs at approximately 2h00 feet vnere terrain ~ is at same j
elevation as the stack discharge point.
l-Case (a)-represents instantaneous meltdown with 0.1% per day leakage _at ground level _ This calculation is a reference calculation made for a single l
containment ba rier with a fission product release corresponding _ to that made in TID-lhBhlt (100% noble gases, 50% halogens, and 15 particulates) j j
with a 50% plate out factor in the case of halogens.
It dois not assume any credit for meltdown as a function of time or pressuze decay as a j_
function of time (hence leakage as a function of time).
The leakage j
rate assumed is, for this case, independent of. pressure.
Case (b). was calculated by computing instantaneous meitdown, pressure Jecay as presented in the hazards report for 3 air recirculation fans, i
ano a leakage rate of 0.1% per day referenced to 15 psig.-.This case is 2
essentially that for a single containment' barrier with only engineered safeguards to reduce pressure.
Case (c) was -computed as for case. (b)' above except for an elevated re-This could correspond to components such as the pump back com-lease.-
pressor leaking at a combined rate of 0.1% per day out the stack.
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