ML20151X436
| ML20151X436 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon, 05000000 |
| Issue date: | 11/23/1967 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20151W779 | List:
|
| References | |
| FOIA-88-156 NUDOCS 8808250356 | |
| Download: ML20151X436 (23) | |
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U. S. ATGfIC ENERGY COMISSION DIVISION OF REACTOR LICENSING REPORT TO THE ADVISORY C04MITTEE ON REACTOR SAFEGUARDS IN THE W5TER OF PACIFIC GAS AND ELECTRIC CCMPANY DIABLO CANYON NJCLEAR PIANI DOCKETI NO. 50-275 REPCRT NO. 2 I
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Note by the Director, Division of Reactor Licensirg The attached report has been prepared by the Division of Reactor Licensing for use by the Advisory Co=mittee on Reactor Safeguards at its Dece:nber l
1967 ceeting.
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w.meTV A n U T R T /Th M it T r,jcg w arsu w ue va m n 1.0 Introduction Pacific Gas and Electric Company submitted an application dated January 16, 1967, for a construction permit for its proposed Diablo Canyon Nuclear Power Plant. A previous report to the ACRS dated September 20, 1967 vas prepared which included our preliminary evaluation of the site, seismic design, core physics, and themal-hydraulic design. This report presents the results of our evaluation of the proposed facility design in those areas where reserva-tions were previously expressed as well as items not included in the previous report.
In certain areas the staff has not accepted the applicant's proposed design. We have infor=ed the applicant of these areas and they are discussed in the followin6 sections. It is our understanding that the applicant proposes to file an ameniment prior to the December ACRS meeting date to fo mally document oral co=mitments.
2.0 Site Characteristics In our first report to the Committee the only sitin6 matter that was not resolved vas the problem of suitable plant protection against potential tsunamis. The applicant has proposed that the use of a 20 foot tsunami (including peak storm and high tide) for protection design purposes was sufficiently conservative for this site and presented information in support of its view. This information was reviewed by our consultants in the USC&GS and ESSA. Based upon this review and a discussion with the applicant on November 21, 1967, our consultants have not chan6ed their opinions and we believe with them that protection a6ainst floodin6 from a tsunami should be provided to an elevation of 30 feet above mean low low water. At the conclusion of this meeting, .the applicant orally a6 reed to protect all AUpppp A n n fem nutn nz.
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~anw ww Amff ITY@R m u O Mit V Class I structures to this elevation. As originally proposed, all Class I structures and equipment except the intake structure are located 80 or more feet above MSL. The top of the intake structure as designed vould be 20 feet above MLLW (Mean Iov Lov Water) and to accomodate the added tsunami height a 10 foot vall vill be built on top of the intake structure around the fire and auxiliary sea water pump motors (the pumps needed to maintain the nuclear facility in a safe shutdown condition) protecting them to a 30 foot level.
In the judgment of our consultants the maximum drav-down due to the tsunami could result in a lowerin6 of the sea water level of approximately 25 feet belov mean lov lov vater. They further stated that the duration of the dravdown condition vould be short, taking less than one hour for a complete cycle vith only a few minutes at the maximum dravdown.
The applicant stated that the intake stracture vill be designed to provide a "vet vell" of adequate capacity for assurin8 at all times a sufficient volume of water for operation of the auxiliary sea water pumps.
This design concept provides for a veir type arrangement to trap vater in the intake structure to a depth of about 12 feet. Under conditions of extreme dravdown, sufficient water would be trapped in the intake structure to permit operation of the auxiliary water pumps for approximately 30 minutes. This design vill require shutdovn of the me.'.n cooling vater pumps when the dravdown exceeds a given elevation because these pumps also drav from the same source. To assure that the main cooling water pumps vould be i
i shut devn the applicant has stated that they vill receive warning of potential tsuna=1 conditions throu6h the ESSA alerting system. Upon receipt of the alert, the applicant stated that an observer, who vill be in contact with AcWUenAU gpa u uvunu 4MR U Yde "&""LH
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the control roor., vill be posted and when the water at the intake structure reaches a pre-set level the plant vill be shut down. Ve and our consultants feel that with these design provisions the Diablo Canyon facility vi]l be adequately protected against tsunamis. Selection of the level for shutdown and whether or not automatic protection is required are beig deferred to the operating license review stage.
3.C Seismic Deaign.
OJr previous report to the ACRS included a sectior. on the seismic desi6n criteria proposed for the Diablo Canyon facility. At that time our review of the containc.ent design was co=;1ete except for a fev outstanding items The design criteria where clarifiestion was requested from the applicant.
for other Class I stn.ctures vas still under review at the time of our last report.
Additional irforme. tion, presented in A:endrer.ts 5 and 6, has been revieved by toth the staff and our cor.sultants. Our positiorc for the containment structure and other Class I structures and components are discussed separately below. Ve expect that our consul, ants report vill be available prior to the December ACRS meeting.
3 . *_ Containment Desigq Factored loads ior the design of the containnect structure have been proposed which cor.J:ine dead loads, pressure loads, terperature leads and The earthquake loads (or vind load if greater than the earthquake load).
formlae for the three loading conditions are presented on page 510 of the PSAR. The containment vill be designed such that the most restrictive loading cofination for each pe,rticular region of the containment results in average stresses not Breater than the yield point. The staff acd our coneuharts concur in the design approach proposed by the applicast.
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Tne reactor contain ent structure, consictic6 of a steel-lined, reinforced concrete, straight circular cylinder, with a hemispherical doce and a flat totto=, presents two new features; a helical reinforcin6 pattern Tne in the concrete shell and a hinge at the base of the cylindrical vall.
concrete cylinder is reinforced with helical bars, inclined at an an61 e of 300 from the vertical. The vall reinforcing bars are continuous with the do=e reinforcin6 Additional hoop reinforcing is provided in the cylindrical vall. The continuity of the vall and dose reinforcing does not require termination and anchorage of any bar in the dome, and is an attractive feature of this reinforcing arraL6esent. Ar.other advanta6e is the direct transmission of shears throu6hout the structure. Tne applicant presents a prelinicary arrangement of the reinforcin6 pattern which vill require farther attention as outlined below. Detailed arrangement of the reinforcing tars including the location of the splices, the possible interferences between the bars, the erection sequence of the reinforcing, the arrangenent of the reinforcing at special points such as openin6s, zones of discontinuities, We do not foresee any Broups of penetrations, have still to be worked out.
insurmountable problems in the prelicinary design and reco6n17e that alternate possibilities may be used if unexpected difficulties should arise during the final design stage.
The design at the base of the vall incorporates a system of vertical The benes are hinged at their steel teams, spaced four feet on centers.
base and are 20 feet long. The base of the vall is divided into three concentric layers. The inner layer, approximately twelve inches thick, supports the liner. The intermediate layer, approximately 16 inches thick, contains the vertical steel beams anchored into concrete ad,jecent to the=.
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The The exterior layer contains the helical and the hoop reinforcing bars.
two surfaces of contact between the three layers, and the steel bea=s vill be treated with a bond-breaking substance, to ensure independent action of all elements. The purpose of this arran6ement is to ensure transmission of the radial shears from the vall into the base. This is a new design and vill require more studies and tests to clarify its behavior under all possible load combinations.
It is not clear hov the stresses vill be transmitted from the beans into the adjacent concrete slabs and vice versa. It is also not clear how the hinge action vill be ensured across three layers of concrete. Finally, the rotation at the hin6e may influence the behavior of the liner at this location in an unfavorable =tr.euver. However, if further studies disclose unexpected difficulties, alternate arrangements may be used.
The design of penetrations, descrited in general terms, is acceptable to us. Additional studies vill be required, however, to clarify all the details of the arrangement of reinforcing bars at the openings, of the liner, and of the ancht.rs.
32 Class I Structures, The applicant presentet in Asendment 5 a document entitled "Ultimate Strength Criteria to Ensure No Loss of Punction of Piping and Vessels under Earthquake Loading," WACP-5890, Redsion 1. This document contains stress loading criteria which Westinghouse proposes as their basis for designing vessels and piping.
Our present position is that all Class I structures, systems, and components should be designed to vithstand:
(a) Load combinations including nor=al design loads and design WEECH/4 U5E ONLY"
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earthquake loads within nomal working stress or deflection li its.
(b) Load combinations including maximu= earthquake loads and applicable design basis accident loads, without loss of function of the specific structure, system, or component.
The Class I items can be broadly subdivided into three categories:
Buildings and Structures, Mechanical Systems, end Instrumentation and Control.
Since Class I items are intended to perform different functione, they will require, in Seneral, different acceptance limits under type (b) load combinations.
The seismic design criteria for Class I cechanical systems, some of which are listed below have been specifically reviewed is discussed in subsequent sections: .
(a) Reactor vessel, its supports and vessel inte a. - includin6 fuel assemblies and control rod drives.
(b) Beactor coolant system, including piping, valves, steam gene:?ators, pressurizer, pumps and component supports.
(c) Emergency core coolin6 system, including piping, valves, water tanks, accu =ulators and pumps.
(d) Containment safe 6uards systems including pipin6, tanks, valves, ducts, fans, coolers and spray headers.
In response to our request for a definition of the proposed load combinations and stress or defor=ation limits, the applicant supplied information for reactor internals, vessels, piping, and supports in the Fifth Supplement, pages 19 through 52. The stress limits for type (b) leading (caximum earthquake plus pipe rupture loads) were supplied with the Fourth Supplement in the report WCAF-5890-1.
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We have reviewed these submittals and ve consider the loadin6 combicatiocs assumed by the applicant (Table 10-1) both realistic and satisfactory. Tha proposed stress or deformation limits for the specific components are discussed in more detail below:
32.1 Reactor _ vessel Internals To be able to perform their function, i.e. allow core shutdown and cooling, the reactor vessel internals must satisfy deformation limits that are more restrictive than the stress limits for other compocents. The applicacit stated that the internals vill be designed to withstand normal design loads plus earthquake loads within Section III limits, with exception of materials not covered by the Code, such as fuel rod cladding. Seismic stresses vill te cor. tined in the most conservative way and vill be considered as primary stresses. We consider these criteria satisfactory.
For the type (b) loading, including maximum earthquake loads and tlovdevn We effects due to a pipe break, the deflections are listed in Table 10-3 consider these deflections to be reasonable. We intend to review the applicant's calculations for selected internals at the operating license stage of our reviev.
32.2 vessels, P3 ping and &apports We have reviewed the stress li=its for these components, proposed by the We applicant in Table 10-1 (Fifth Supplement) and the report WCAP-3890-1.
find the Section III or B31.1 Cede limits,'for vessels and piping respectively, satisfactory for type (a) load conbination (normal design loads plus deaign earthquake loads).
We agree also that for type (b) load combination, (corresponding to load combination 4 in Table 10-1) the allovable extent of plastic determa*, ion can AUi7U_(f'U.A_sU._.
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be larger than that associated with the Section III stress limits. We hC leva, de : ,
hovev?r, that it vould be prudent to assure that the primary stresse exceed the "collapse stresses" as defined in the "Criteria of Sectin III of the AfME Boiler and Pressure Vessel Code for Nuclear Vessels," pages 5 and !.
These primary stress limits based on plastic collapse are discussed alec in CBNL-NSIC-21 "Technology of Steel Pressure Vessels for Water-Cooled Paclear Reactors," pages 341 throush 346.
The "collapse" stresses for combined primary loading have been ettlined on the basis of limit design theory and perfect plasticity with co stradn-hardening. The actual strain-hardening properties of spt itfic caterials, talsaced to a certain extent by imperfections in the caterials, vill provide larger or enller carsics of safety.
Our position is also in a6:eement vith that expressed ty the "Ientative Regulatory Supple entry Criteria for ASME Code - Constructed INeleer Pressure Vessels," which on page 29 states that where limit analysis is used the combined loadings shall be limited tc 90 percent of the yield compse load.
Since the stress limits, proposed by the applicant in WCAF-5890-1 for
.ype (b) loading, exceed those described above, ve conclude that they do not provide an adequate cargin of safety. We intend to have the appli unt identify specific components for which streeses under type (b) loading vnld exceed the "collapse" stresses used as a tasis for Section III stresa limits.
We intend also to find out what design modifications are necessary to reet these limits.
In conclusion, it is our finding the.t the desi6n method is acceptatie, however the stress limits proposed for type (b) loadings should be modified
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to provide an adequase margin of safety.
h.C Core Thermal, Hydraulic, a_nd Physics Design The ther-al-hydraulic and physics aspects of the Diablo Canyon f acility were presented in our previous report to the ACRS. Since the time of that report additional infomation has been received on programs for fuel developner.t.
use of fixed poison rods and additional information en the use of partial length control rods. A table summarizing the important oore parameters of the Indian Point II an'. Diablo Canyon designs is presented in Table I.
k.1 Physics Aspelts The Diablo Canyon facility physics design tesis has been modified to f.cclude fixed burnable poison in the first fuel cycle. Borosilicate 6 ass 1 encapsulsted in stainless steel rods vill be distributed throughout the core in unused control rod guide tubes. It is proposed that about 1144 of these rods be installed in vacant control rod guide tubes, held in place by a spider assem':ly compressed beneath the upper co e plate to ensure flov forces vill not cause motion. These rods vould have a combined worth of 7 2% delta k/k, and as a Tr.e consequence the dissolved boron concentration during operation is reduced.
reduced dissolved boron concentration results in negative moderator temperst'.:rt coefficients itich vill reduce the potential severity of loss. of coolant accidents and rod ejection accidents and, according to the applicant, vill ds p induced xenon occillations.
The react.ivity vorth of the borosilicate glass rods is being evaluated at the Westinghouse Reactor Evaluation Center by comparing calculated and measured worths from critical experiments. Based on preliminary evaluation, Westingtone has confidence in predictir.g the reactivity worth of the poison rods. Long term performance of these rods in a power reactor environment will emnm e o e vett e AMW UF T MeMS.e UtJe %vaNLW u
&JECsidL UsE or L Table I Comparison of Diablo Canyon and Indian Point II Diablo Canyon Indian Point II 3250 2758 Total Heat Generation, Hv(t) 2 207,000 175,600 Average Heat Flux, BIU/hr ft 583,000 570,800 PeakHeatFlux,PTU/hrfte Average Linear Heat Generation, 57 6.7 kv/ft 18 9 18.5 Peak Linear Heat C' :'eration, hv/ft 2 6 2 56 x 10 Core Mass velocity )~ nr ft 2 56 x 10 539 543 Core Inlet Temperature, *F.
Feakin6 Factors 2.82 3 25 Fq 1.70 1.88 F4H 1.81 1.81 DND retio (W-3)
Boron Concentration for Keff = .99 1600 3400 all rods ?ut, cold, ppm Moderator Temperature coefficient,Ak/k*F .5 to -3 0 x 10 -fl.0 to -3 0 x 10~4 Fuel Enrichments 2.2 2.23 Region 1 27 2 38 Region 2 33 2.68 Region 3 V
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The applicant states in kr.eudsent 7 (page II-1) that inclusion of burcat;e poisons vill damp xenon oscillations in the X-Y plane since the moderater coefficient is negative by a sufficient mar 6 1n. The threshold for X-Y instability due to feedback from the moderator te:nperature coefficient is calculated to be ".CTT x 10 deltak/kOF. The applicant analyzed uncertain-ties in the variables used in the prediction of stability and has related these variables to the ma6nitude of moderator coefficient. The applicant believes the design moderator temperature coefficient is sufficiently negative to ensure stability.
Insofar as axial stability is concerned the applicant vill install The partial length rods to te moved as a bank to damp induced oscillations.
pa-tial rods vill also be us2d to provide flattening in the axial direction and hence the peaking factor for heat flux has been re$uced from previous Westinghouse desi6cs. Additional comments on this aspect are presented in the thermal-bydraulics section.
4.? Tcemal-Hydraulics
~fhe core design for the Diablo Canyon reactor takes advantage of reduced peaking factors which are made possible by the use of partial length control rods. This chan6e makes it possible to increase the avera6e power of the core 18% compared to previous designs, yet maintain peak specific fuel powers in line with past designs. In effect, although the minimum DNB ratio in the een remains constant, the number of ruel rods which are oper?ing close to the minimum DIGR is increased in Diablo. To illustrate this point, discussions with Westin6house personnel have indicated the following comparisons t* ween Diablo Canyon and Indian Point II:
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~w .ilPliouam use ONM Ihmber of rods with DNB less than indicated Diablo Canyon Indian Point II 100% power, normal flow, design inlet temperature 0 0 DNBR of 1.8 <.10 19 150 550 110 2.0 125% pover, normal flow, design inlet temperature 750 105 DNER of 13 1000 15 2500 100% power,90%ofnorme.1 flov, design inlet temp-erature 0 0 DNBR 13 0 15 0 .
250 15 17 500 19 1500 100% power, 80% of normal flow, design inlet temp-erature 0 0 DNBR 13 50 15 550 17 2300 700 The design basis fer analyzins; transients in this core is that the minimum DNBR shall not be less than 13, and we have concluded that even though a greater nunber of fuel rods vould be ir*olved which approached D!E (e.g., more rods could have a calculated DNBR between 13 and 1.l+), statistis-ally there is ample margin of safety.
Ve do not agree; however, that sufficient instrumentation is being pro;esed to ensure that the axial flattening (peaking factors) vill in practice be achieved. The applicant has proposed that relis.nce be placed entirely on the 11"hTT11/f"aft w avaruw A 1Twwa UIG!LTTwOvF.J
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The in-core conitors for Diablo Canyon, as presently proposed, are six traveling flux probes which may be positioned in any of 58 thimble locstions in the core. These in-core channels are not designed to operate at full power for more than a fev months. The applicant's position on in-core moniters is that test programs (primarily at SENA) vill adequately demonstrate the capability of the external long ion chambers to predict power patterns within the core.
Cur position in this regard is that intelligence fro: in-core monitors .ast be provided to an operator to position the partial rods in order to assure proper axial power flattenin6 If, at sece later date, experience shows that the 9xternal conitors vill detect in-core anomalies with adequate sensitivity ve vould change our position.
One other aspect of our review for Diablo Canyon is that of fuel perforce.c:e at proposed peak powers correspondin6 to expected burnup. The applican provided a sumary bar chart shoving both the present and proposed irradiation test programs to demonstrate acceptable fuel perfomance for this reactor.
We have plotted the expected peak rod operating :haracteristics on this tar cl. art. As is evident, at the present time there is no satisfactory operating experience at the linear power generation levels contemplated for the Diablo Canyon reactor. We believe, however, the test prograss for Saxton ant Zorita vill provide a basis for predicting operation of the Diablo Canyon facility.
50 Instrumentation and Control (This section under preparation and vill be completed and transmitted to co=nittee as soon as poesible. )
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waluated in cenfer .ar.ce with the guidelines of Fart 100.
Although our assumptions differ sotevhst frc these used by 4.he applicant, all of the resultinE deses, vitn the excepti n cf the TID 113;4 type accident, are well telow the 10 C5100 Euideline dese levele at the evcilaile exclusion rene radius (0.5 mile) and the low populatien one trd iur (~.5 miler) without any thyroid dere reduction factore needed.
Fcr 'he lcrs of ecclent accident which results in -he TID lL6LL fissien prcduct relence fractiens (100fc noble gas, 259 icdine, end 15 solids) avail-e have calculated the folleving n ie fcr leckage, with nt, iodine reductien, scre leve:s:
2 Ecur Icse (Fet' 30 ley Icte (Femi 7 0.5 mile 7 ~.5 miler "hyroid 370 15L
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- 1. '<eteerclogy - Ground release, cer.terline, Fasquill Type F, 1 =/eec. , and vake of the tuildinE (Volt".etric source and e = 1/2) fer the first 8 hcurs of the accident; from 8 to 2L hours ground release, Fes quill Type F,1 m/see. , uniform dis-persien into a 22-1/2' sector; and 1 day to 30 days - the stet $lity, vind speed, and directicn vere varied.
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', 2. Leak Pete - 0.15/ day for 2k ' hours, and 0.0L5%/ day for the duration of the accident.
It is apparent from the above table that an todine reduction facter of about 3 is needed to meet the 2-hour thyroid dose l!=it of 300 rum at the site boundary. No reduction factors are needed to meet 30 day doue limits at the available low population zene distance.
Although the design basis for sizing the emergency core cooling system is to limit fission product release from the fuc1, it has been our position that the containment and its associated enE ineered safety features be capable of limiting potential doses in conformance to Fart 100 criteria. The appli-cant initially proposed a containment spray system usinE sodium thiosulfate to provide the needed iodine removal. In Amendment No. 2 (pages 128--i30) l a test,proEram for this system was described. Eesults cf these tests and a research and development pregram vere further defined in Amendment No. 6 (pg. 6-7). We have discussed the propcsed research and development program with the applicant, and PG&E has stated that space is _ being reserved near the air recirculation units so that charcoal filter units can be added in the event the research and development does not provide cenclusive evidence to l
l support needed iodine removal rates.
In addition to making independent dose calculations in conformance ,
vith Part 100 guidelines, we have reviewed the design bases for the emergency core coolinE system and tne containment heat removal systems.
Our evaluation of these systems follows:
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its nominal heat transfer geometry and zircaloy-vater reactions vill be limited to an insignificant amount. The emergency core cooling system (accumulator tanks and Safety Injection System) will be designed to provide sufficient injection of torated vater to meet this crittrion f or all reactor coolant pipe break sizes and locations up to and including a double-ended rupture of the reactor coolant pipe." .
We have held meetings with the applicant with regard to ;he degree of redundancy required to meet the design objective given above. Our stated position to the applicant is that redundant systems shculd be provided ruch that an active component failure for toth short and ler.g term conditiens and passive failure for long term cooling requirements can tv tolerated without jeopardizing the ability of providing core cooling. In effect what this means is that co=on headers as originally proposed for safety injection Eis criterien, in our and long term recirculation were not acceptable.
view, also applies to the component cooling vater system and the auxiliary salt water system since single failures in these systems could also negate long term core cooling. In response to our interpretation of Criterion kk, the applicant modified the salt water system in Appendi.x A of Amendment No. 2, modified the auxiliary coolant vater system in Appendix B of i o
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h.endment No. 3, and modified the safety Injection system in h.er.dment No. 7. We have reviewed these revised drawings and find certain exceptions to the desired design goal. The applicant still retains single valves which l
Join otherwise redundant independent systems. While it may be desirable or even necessary to have the capability of transferring flow around specific components in one cooling loop snd utilize components in the opposite loop, use of single valves to accomplish this objectivi can place both systems in jeopardy because of a single failure. If, for examp e, this single valve i
should teSin to leak excessively, toth recirculation systems vould have to te secured to isolate the failure. Instellatien of dual valves in these locations vould eliminate this cbjection. One other location ve have identified where a sinel e failure cannot te tolerated is in either of the .
isolation valves on the containment sump lines. We have stated our objections to the applicant and have indicated that medification durinE design vill need to be made. :
We have reviewed the performance of the Safety Injection System in ;
being capable of meetinE the design objectives. Specific ansvers to questions by the staff with reEard to ECCS capatility vere made in Amendment No. 3 (pp 114 248). We have reviewed the information submitted l
and believe the system as proposed is Eenerally adequate (the specific area of thermal shock is still under review). The performance of the system with 3 of 4 accumulators and 1 of 2 s.I.s. pumps can te su=marized as 4
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- 2. 3 0 ft 1615 2 1795 o 3 0.5 ft Study of the problem of , thermal shock during core cooling system actuation by Babcock & Wilcox, Combustion Engineering, General Electric and Westinghouse continues. Two modes of potential failure are being considered: ductile yielding and brittle fracture. 2.e latter is teing treated using both the Pellini-Puzak diagram approach and fracture mechanics.
We are presently waiting for the results of calculations, promised by Westin6 h ouse in a topical report, to establish thermal stress distributien patterns near the crack tip as the crack progresses through the thickness of the vessel. Since the information submitted by Westinghouse, sc far, in connection with the Diablo Canycn application is insufficient, ve consider the them.al shock problem unresolved at' this time.
6.2 Enginee;ed Safety Teatures for Heat Removal from the Containment Tne Diablo Canyon Containment vessel is designed for an accident pressure of 17 psig. The applicant was asked to perform calculations to show the capability of the containment to withstand various assumed energy releases during the course of an accident. The ansvers to these questions appear in Amendment No. 3 (pp 181-219). Engineered safety natures for the containment structure are redundant and the applicant's analysis shows that pyggen A tr wse e A ntu m
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. 20 eperation of 3 cf 5 ccntainment air coolers and 1 of 2 containment spray systems is adequate to maintain the calculated pressure telow design pressure.
We have reviewed the accident medel tnd have concluded that the containment and its heat removal systems are adequately siced.
One aspect which ve telieve needs further attention during detailed The design is that of leak detection en external recirculation Jystems.
recirculation features are closely associated with the ECCS (for long term heat removal) and our concern is that of detecting cnd being capable of isolating leaks in either of the tue systete. If the leakage frc: valves and packings are within design limits, the dose centribution can be tolerated within Part 100 guidelinee. If major leaks should develop during the re-circulation phase, activity leakage to the er.virennent (there is no pro.
visien for iodine removal in the auxiliary building ventilation system) ceuld teccie excessive unless an cperater has previsien to detect and isclate the source.
6.*; Control Room Shielding The accident dose criteria for this centrol room (including ingress end egress) is 2.5 re vtele tedy and 330 re te the thyroid for the course of an accident. ! our epinien, an iodine removal system should te in.
corporated into the control room ventilation system or other censures should te taken to limit potential thyroid doses in the centrol room to values more l'.. line with Criterien 11 censideraticns.
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7 Cenclusiens Ascuning catisfactory resciutien, as the final decign evelves, of specific problems enumerated in the foregoir.g sections, ve have concluded that there is reasonable assurance the Diablo Canyon facility can te built and operated at the proposed location without undue risk to the health and safety of the public.
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%9CCb% Uuu Me As u 22 APPENDIX A LIST OF AMENDMENTS - DIABLO CANYON FACILITY i
- 1. Amendment No. 1 dated July 10, 1967 which contained answers to questions, design methods based on ultimate strength criteria, and described part length absorber rods.
- 2. Amendment No. 2 dated July 24, 1967 which contained answers to questions.
- 3. Amendment No. 3 dated July 31, 1967 which contained answers to questions t
and additional information on site geology, 4 Amendment No. 4 dated September 8, 1967 which provided a cross reference i-to pages in the PSAR which dealt with each of the proposed General P Design Criteria.
- 5. Amendment No. 5 dated October 18, 1967 which contained financial data, s additional tsunami information and revised information on the ultimate strength design criteria.
- 6. Amendment No. 6 dated November 6, 1967 which contained answers to questions, outlined research and development programs, and presented topical reports on the use of burnable poison rods and experimental results on DNB studies in rod bundles. .
- 7. Amendment No. 7 dated November 9, 1967 which contained answers to
- questions.
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