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1 DIVISION OF REACTOR LICENSING REPORT TO THE ADVISORY C04MITTEE ON REACTOR SAFEGUARDS IN THE MATTER OF ;
1 PACIFIC GAS AND ELECTRIC C04PANY DIABLO CANYON NUCLEAR PLANT DOCKET NO. 50-275 pt .
REPORT NO. 2 l, N[
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Note by the Director, Division of Reactor Licensing i The attached report has been prepared by the Division of Reactor Licensing for use by the Advisory Committee on Reactor Safeguards at its December 1967 meeting.
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l 1.0 Introduction Pacific Gas and Electric Company sule.itted an application dated January 16, 1967, for a construction permit for its proposed Diablo Canyon Nuclear Power l Plant. A previous report to the ACRS dated September 20, 1967 was prepared
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which included our preliminary evaluation of the site, seismic design, core l physics, and thermal-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 l - report.
i l In certain areas the etaff has not accepted the applicant's proposed l
l design. We have informed the applicant of these areas and they are discussed in the following sections. It is our understanding that the applicant proposes to file an amendment prior to the December ACRS meeting date to fomally document oral commitments.
2.0 Site Characteristics In our first report to the Committee the only siting matter that was not resolved was the problem of suitable plant protection against potential tsunanis. The applicant has proposed that the use of a 20 foot tsimmi (including peak storm and h16h tide) for protection des 1 n6 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 against flooding from a tsunami should be provided to an elevation of 30 feet above mean lov low vater. At the conclusion of this meeting, bthe applicant orally a6 reed to protect all
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I feet above MLIN (Mean Iov Lev Water) and to accomodate the added tsunami height a 10 foot van vill be built on top of the intale structure around l the fire and anviliary sea water pump motors (the pumps needed to maintain j the nuclear facility in a safe shutdown condition) protectin6 them to a 1
30 foot level.
In the jud ment 6 of our consultants the maximum drav-down due to the tsunami could result in a lowering of the sea water level of approximately 25 feet below mean lov lov vater. They further stated that the duration of the dravdown condition vould be short, taking less than one hour for a i
complete cycle with only a few minutes at the maximum iravdevn.
The applicant stated that the intake structure vin be designed to provide a " vet well" of adequate capacity for assuring 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 water in the intake structure to a depth of about 12 feet. Under conditions of extreme dravdovn, suffic$ent water would be trapped in the intake structure to permit operation of the auxiliary water pumps for approximately 30 minutes. This desi 6n vill require shutdown of the mais cooling vater pumps when the dravdown exceeds a given elevation because these pumps also draw from the same source. To assure that the main coolin6 vater pumps vould be shut down the applicant has stated that they will receive warning of potential tsunami conditions through the ESSA alerting system. Upon receipt of the alert, the applicant stated that an observer, who will be in contact with
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. . . STOIY the control room, 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 vill be adequately protected aEainst tsunamis. Selection of the level for shutdown and whether or not automatic protection is required are being deferred to the operating license review sta6e.
30 Seismic Design Our previous report to the ACRS included a section on the seismic design criteria proposed for the Diablo Canyon facility. At t! tat time iur review of the containment design was com;1ete except for a few outstanding items vbere clarifiestion was requested from the applicant. The design criteria for other Class I structures was still under review at the time of our last report. Additional information, presented in Amendments 5 and.6, has been reviewed by toth the staff and our consultants. Our positions for the j containment structure and other Class I structures and components are >
discussed se;arately below. We expect that our conr:ltants report. Will be available prier to the December ACRS meeting.
31 Containment Design Factored loads for the design of the containmer. structure have been proposed which combire dead loads, pressure loads, temperature leads and ea-thquake loads (or vind load if greater than the earthquake load). The formulae for the three leading conditions are presented on paEe 5-10 of the PSAR. The containment vill be designed such that the most restrictive loading combination for each particular region of the containment results in avera6e stresses not greater than the yield point. The staff and our consultants concur in the design approach proposed by the applicant.
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The reactor containment structure, consisting of a steel-lined, reinforced concrete, straight circular cylinder, with a hemispherical dome and a flat bottom, presents two new features; a helical reinforcing pattern in the concrete shell and a hinge at the base of the cylindrical vall. The concrete cylinder is reinforced with helical bars, inclined at an an61t of 30 from the vertical. The vall reinforcing bars are continuous with the ;
dome reinforcing. Additional hoop reinforcing is provided in the cylindrical vall. The continuity of the vall and dome reinforcing does not require j termination and anchorage of any bar in the dome, and is an attractive feature of this reinforcing arrangement. Another advantage is the direct i
transmission of shears throughout the structure. The applicant presents a j l preliminary arrangement of the reinforcing pattern which vill require further i
i attention as outlined below. Detailed arrangement of the reinforcing tars includir.g the location of the splices, the possible interferences between 1 \
the bars, the erection sequence of the reinforcing, the arrangement of the reinforcing at specirl points such as openings, zones of discontinuities, groups of penetrations, have still to be worked out. "a'e do not foreste any insurmountable problems in the preliminary design and reco6nize that alternate possibilities may be used if unexpected difficulties should arise du*ing the final design stage.
The desi6n at the base of the vall incorporates a system of vertical steel beams, spaced four feet on centers. The beams are hinged at their base and are 20 feet long. The base of the vall is divided into three concentric layers. The inner layer, approximately twelve inches thich, supports the liner. The intermediate layer, approximately 16 inches thick, contains the vertical steel beams anchored into concrete adjacent to them.
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The exterior layer contains the helical and the hoop reinforcing bars. The two surfaces of contact between the three layers, and the steel besas vill be treated with a bond-bresking substance, to ensure independent action of I.
all elements. The purpose of this arrangement is to ensure transmission of l the radial shears from the vall into the base. This is a new design and vill require more studies and tests to clarify its behavie under all possible
( load combinations.
It is not clear how the stresses vill be transni;ted from the beams into 1
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 I rotation at the hinge may influence the behavior of the liner at this l
location in an unfavorable maneuver. However, if fur.her studies disclose l .
l unexpected difficulties, alternate arrangements may be used. l l
The design of penetrations, described in general terms, is acceptable to l
us. Additional studies vill be reouired, however, to clarify all the details of the arrangement of reinforcing bars at the openings, of the liner, and of the anchors.
l 32 Class I Structures The applicant presented in Anendment 5 a documen entitled " Ultimate Strength Criteria to Ensure No Loss of Function of Piping and Vessels under Earthquake Loading," WACP-5890, Revision 1. This doernent contains stress loading criteria which Westin6h ouse 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 withstand:
(a) Lead combinations including norma] deeign leads and design
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(b) Lead combinations including maximum earthquake leads and applicable design basis accident loads, without loss of function of the specific structure, system, or component.
f The Class I items can be broadly subdivided into three cate6ories:
Buildings and Structures, Mechanical Systems, and Instrumentation and Centrol. i l
Since Class I items are intended to perform differest functions, they will ,
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require, in general, different acceptance limits under type (1) lead combinations.
The seismic design criteria for Class I mechanical systems, some of which are listed belov have been specifically reviend as discussed in subsequent sections:
1 (a) Reactor vessel, its supports and vessel 1:ternals including fuel 1 assemblies and control rod drives.
(b) Reactor coolant system, including piping, valves, steam generators, l pressurizer, pumps and component supports.
(c) Emergency core cooling system, including ;1 pin 6, valves, vater I l
tanks, accumulators and pumps.
f (d) Containment safeguards systems including piping, tanks, valves, ducts, fans, coolers and spray headers.
In response to our request for a definition of the ;roposed load combinations and stress or deformation limits, the applicant supplied information for reactor internals, vessels, pipiry!;, and supports in the Fifth Supplement, pages 19 through 52. The stress limits for type (b) loading (mav1==
earthquake plus pipe rupture loads) were supplied with the Fourth Supplement l l
in the report WCAP-5890-1.
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l We have reviewed these submittals and we consider the loamg combinations i assumed by the applicant (Table 10-1) both realistic and satisfactory. The !
proposed stress or deformation limits for the specific components are discussed in more detail below:
3 2.1 Reactor Vessel Internals To be able to perform their function, i.e. allw core shutdown and cooling, i the reactor vessel internals must satisfy deformation limits that are more restrictive than the stress limits for other components. The applicant stated '
I that the internals vill be designed to withstand no: mal 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 be combined in the most conservative way and vill be considered as primary stresses. We consider these criteria satisfactory.
For the type (b) loading, including maximum ea-thquake loads and blovdovn effects due to a pipe break, the deflections are listed in Table 10-3 We consider these deflections to be reasonable. We in end to review the applica:t's l
l calculations for selected internals at the operating license stage of our review.
3 2.2 Vessels, Piping and Supports We have reviewed the stress limits for these co=ponents, proposed by the applicant in Table 10-1 (Fifth Supplement) and the report WCAP-5890-1. We i find the Section III or B31.1 Code limits, for vessels and piping respectively, satisfactory for type (a) load combination (normal design loads plus design earthquake leads).
We agree also that for type (b) load combination, (corresponding to load combination !& in Table 10-1) the allovable extent of plastic deformation can 10)FMGML USE NLY l#/W 7 pp e l
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f however, that it would be prudent to assure that the primary stressec de net exceed the " collapse stresses" as defined in the " Criteria of Section III of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels," pa6es 5 and 6.
These primary stress limits based on plastic collapse are discussed airc in ORRL-NSIC-21 " Technology of Steel Pressure Vessels for Water-Cooled Nuclear Beactors," pa6es 341 throush 346.
The " collapse" stresses for combined primary loaiing have been obtained on the basis of limit design theory and perfect plasticity with no strain-hardeninE. The actual strain-hardening properties of specific materials,.
balanced to a certain extent by imperfections in the materials, will provide lar6er or emeller margins of safety.
Our position is also in agreement with that expressed ty the " Tentative Regulatory Supplementary Criteria for ASME Code - Constructed Nuclear Pressure Vessels," which on page 29 states that where limit analysis is used the combined loadin6s shall be limited to 90 percent of the yield c:C.19.pse I load.
l Since the stress limits, proposed by the applicant in WCAP-5c90-1 for j type (b) loading, exceed those described above, ve cenclude that they do not j provide an adequate margin of safety. We intend to have the applicant ;
identify specific compenents for which stresses under type (b) loading veuld i
l- exceed the " collapse" stresses used as a basis for Se: tion III stress 1:Ir.ite.
We intend also to find out what design modifications are necessan to meet these limits.
In conclusion, it is our finding that the desi5n method is acceptable, l
however the stress limits proposed for type (b) loadings should be modified -
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e.0 Core Themal, Hydraulic, and Physics Design The themal-hydraulic and physics aspects of the Diablo Canyon facility were presented in our previous report to the ACRS. Since the time of that report additional information has been received on pro 6 rams for fuel development. l use of fixed poison rods and additional information en the use of partial j
length control rods. A table summarizing the important oore parameters of the !
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Indian Point II and Diablo Canyon designs is presented in Table I.
l L.1 Physics Aspects The Diablo Canyon facility physics design basis has been modified to include l fixed burnable poison in the first fuel cycle. Borosilicate glass encapsulated l
in stainless steel rods vill be distributed throughout the core in unused v
l control rod guide tubes. It is proposed that about 11 % of these. rods be installed in vacant control rod guide tubes, held in place by a spider assembly ;
compressed beneath the upper core plate to ensure flow forces vill not cause ]
motion. These rods vould have a combined worth of 7 2% delta k/k, and as a t
consequence the dissolved boron concentration during operation is reduced. The reduced dissolved boron concentration results in negative moderator temperatre coefficients which vill reduce the potential severity of loss of coolant accidents and rod ejection accidents and, according to the applicant, vill damp induced xenon oscillations.
The reactivity worth of the boros111cate glass rods is being evaluated at the Westin6house Reactor Evaluation Center by comparing calculate'd and measured worths from critical experiments. Based on preliminary evaluation, Westinghouse has confidence in predicting the reactivity vorth of the poison rods. long term performance of these rods in a power reactor environment will gTUenAU~ ~~ -- HIQit?
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_t( FIGALvUSE~DREY Table I j Comparison oi' Diablo Canyon and Indian Point II Diablo Canyon Indian Point II Total Heat Generation, Mw(t) 3250 2758 175,600 Avera6e Heat Flux, M /hr ft2 207,00C Peak Heat Flux, M/hr ft2 583,000 570,800 AveraBe Linear Heat Generation, 6.7 57 kv/ft Peak Linear Heat Generation, kv/ft 18 9 18 5 0
CoreMassvelocitylb/hrft2 2 56 x 10 2 56 x 10' Core Inlet Temperature, 'F. 539 543 Peaking Factors Fq . 2.82 3 25 1 70 1.88 FAH 1.81 1.81 DND ratio (W-3)
Boron Concentration for Keff = .99 )
all rods out, cold, ppm 1600 3400 l
l Moderator Temperature coefficient,Ak/hOF .5 to -3 0 x 10 fl.0 to -3 0 x 10 4 Fuel Enrichments Region 1 2.2 2.23 i Region 2 2.? 2 38 Region 3 33 2.68 l
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CELCIAL USE'OfiCif 11-be evaluated from in-pile testing of two rods in the Saxton reactor.
The applicant states in Amendment 7 (pa6e II-1) that inclusion of burnatie poisons will damp xenon oscillations in the X-Y plane since the moderator coefficient is negative by a sufficient margin. The threshold for X-Y instability due to- feedback from the moderator temperature coefficient is calculated to be ~.CT[ x 10 deltak/k*F. The applicant analyzed uncertain-ties in the variables used in the prediction of stability and has related these variables to the magnitude 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 will install partial len6th rods to be moved as a bank to damp induced oscillations. The pa-tial rods vill also be used to provide flattening in the axial direction i and bence the peaking factor for heat flux has been reduced from previous 1
Westinghouse designs. Additional comments on this aspect are presented in i 1
the thermal-bydraulics section. l l
4.2 Thernal-Hydraulics The core design for the Diablo Canyon reactor takes advantage of reduced peaking factors which are made possible by the use of partial len6th control
! rods. This change makes it possible to increase the avera6e power of t.he core 18% compared to previous designs, yet maintain peak specific fuel powers in line with past desi6ns. In effect, although the minimum DNB ratio in the core remains constant, the number of fuel rods which are operating close to the minimum DNBR is increased in Diablo. To illustrate this point, discussions with Westinghouse personnel have indicated the following comparisons between Diablo Canyon and Indian Point II:
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. s f u L Number of rods with DKB less 4 than indicated Diablo Canyon Indian Point II_ l 100% power,normalflev, design inlet temperature DNBR of 1.8 0 0 19 150 ( 10 2.0 550 110 l
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l design inlet temperature DNBR of 13 750 105 15 2500 1000 100% power,90%ofnormal l 1
flov, design inlet temp-erature DNBR 13 0 0-15 0 0 17 250 15 19 1500 500 l
i 100% power,80%ofnormal flov, design inlet temp- j erature DNBR 13 0 0 15 550 50 17 2300 700 i
The design basis for antlyzing transients in this core is that the l minimum DNBR shall not be less than 13, and we have concluded that even
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thou6h a greater number of fuel rods vould be involved which approached DIE I
(e.g., more rods could have a calculated DNBR between 13 and 1.4), statistic.
ally there is ample margin of safety.
We do not agree; however, that sufficient instrumentation is being proposed to ensure that the axial flattening (peaking factors) vill in practice be achieved. The applicant has proposed that reliance be placed entirely on the
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1 four external flux monitors to detect and correct abne:ml power patterns. !
The' in-core monitors for Diablo Canyon, as presently ;roposed, are six traveling flux probes.which may be positioned in any of 58 thimble locations in the core. These in'-core channels are not designed to operate at full power j for more than a few months. The applicant's position on in-core monitors is.
that test programa.(primarily at SENA) vill adequately. demonstrate th< capability-of the external long ion' chambers to; predict power patterns .within the core, j Our position in this regard is that intelligence from in-core monitors mast be provided to sn'eperator to position the partial rois in order to' assure proper axial power flattening. If, at some later date, experience shows that j the external monitors vill detect in-core anomalies with adequate sensitivity ve, vould change our position. ,
Or.e other aspect of our review for Diablo Canyon is 'that of fuel performan:e -
at proposed peak powers corresponding to expected bursp. The applicann ,
i provided a summary bar chart showing both .the present and proposed irradiation test pro 5 rams to demonstrate acceptable fuel performa::e for .this reactor.
l We have plotted the expected peak rod operating chart:1 eristics on this ba-f
( chart. As is evident, at the present time there is no satisfactory operating "3 .. ;
experience at the linear power generation levels contemplated for the Diablo Canyon reactor. We believe, however, the test progra s for Saxton and Zorita- ;
o vill provide a basis for predicting operation of the .ablo Canyon facility.
50 Instrumentation and Control (This section under preparation and will be. completed and transmitted to committee as soon as possible. )
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T.* Aceident Evaluation Accidents for the Diablo Canyon facility !a < e leer evaluated in ecnfornance with the guidelines of Part 100.
Although our assunptiens differ somewhat fren 4:re used by the applicant, all of the resulting doses, with the excep-ion of the TID 148ht type accident, are well below the 10 CFE 100 Suideline dose levels at the :
- available exclusion sone radius (0.5 mile) and the lov population zone t
radius (7 5 miles) witPout any thyroid dose reduction factors needed. !
l For the loss of coolant accident which results in the TID lL8kk fission i product release fractions (100% netle gas, 25% iodine, and 1% solids) avail-able for leakage, with no iodine reduction, we have calculated the following dose levels:
2 Hour rose (Fen' 30 Day Lose (Fem)
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~he following additionel assu ptions were made 1: calcula ing these deses:
- 1. Meteerclegy - Grcund release, centerline. Pa ;uill hTe F, 1 m 'sec. , and wake of the tuilding (volu .etri: source and e = 1/E for the firct E retrs cf the eerider : frc 9 to Eh hours ground release, Pas quill "ype F,1 - 'si : , unifcrm dis-persien into a 22-1/20 sector; and 1 day to 22 days - the stability, vind speed, and directicn were varied.
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It is apparent' from the above table that an iodine reduction factor of about 3 is needed to meet the 2-hour thyroid dose limit' of 300 rem at ,
the' site boundary. No reduction factors are needed te meet 30 day dose. i i
limits at the available low population zone distance.
Althou6h the design basis for sizing the emer6ency core cooling system is to limit fission product release from the fuel, it has been our position that the containment and its associated engineered safety features be capable of 11.mitirg potential doses in conformance to Part .100 criteria. The'appli-cant initially proposed a containment spray system using sodium thiosulfate to provide the needed iodine removal, In Amendment E:. 2 (pages 128-130) i a test program for this system was described. Results of these tests and a research and development program vere further defined in Amendment No. 6 1
(pg. 6-7). 'de have discussed the propcsed research and development program l l
vith the applicant, and PG&E has stated that space is being reserved near the air recirculation units so that charcoal filter units :an be added in the i event the research and development does not provide c::clusive evidence to f
l support needed iodine removal rates. ,
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! In addition to makin6 independent dose calculati::s in conformance !
with Part 100 guidelines, we have reviewed the desi n6 bases for the i
emergency core cooling system and the containment Fea removal systems.
Our evaluation of these systems follows:
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The criteria for these systems as given in Amend:ent 2 is "that the l
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maximum calculated zircaloy clad temperature vill not et any point in the core exceed the melting temperature of zircaloy. The : ore vill remain in its nominal heat transfer Beometry and zircaloy-vater reactions vill be )
limited to an insignificant amount. Ihe emergency core cooling system l I
(accumulat'or tanks and Safety Injection System) vill te des'igned' to provide sufficient injection of borated water to meet this criterion for 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 the degree of redundancy required to meet the design objective give above. Our stated position to the applicant is that redundant systems sh:uld be provided such-that an active component failure for both short and 1:rg term conditions and passive failure for long term cooling requirements can be tolerated without jeopardizing the ability of providing core cc.: ling. In effect what this means is that co= mon headers as originally prop: sed for safety injection and long term recirculation were not acceptable. This criterion, in our view, also applies to the component cooling vater system and the auxiliary salt water system since single failures in these systers could also neEate.
long term core cooling. In response to our interpretation of Criterion kh, the applicant modifi~ed the salt water system in Appendix A of Amendment No. 2, modified the auxiliary coolant water system in Appendix B of
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/cendment No. 3, and modified the' Safety Injection System in Amendrent No. 7. We have reviewed these revised drawings .and find certain exceptions to the desired design goal. The applicant still retains single valves which join otherwise redundant' independent systems. While i; may be desirable'cr even necessary to have the capability of transferring flow around specific-components in one cooling loop and utilize components in the opposite' loop, use of single valves to accomplish this objective can place both systems in jeopardy because of a single failure. If, for exemple, this single valve should begin to leak excessively, both recirculation systems would have to be secured to isolate the f?ilure. Installation of d:a1 valves in these locations vould. eliminate this objection. One other location we have identified where a single failure cannot be 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 ncdification during design vill need to be made.
We have reviewed the performance of the Safety ~ fection System in being capable of meeting the design objectives. Specific answers to questiens by the staff with regard to ECOS capabili:p were made in Amendment No. 3 (pp 114-248). We have reviewed the information submittee and believe the system as proposed is generally adequa:e (the specific area of thermal chock is still under review). The perfor Ance of the system with 3 of 4 accumulators and 1 of 2 S.I.S. pumps car ':e su=marized as follows:
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- 1. Double-ended coolant 1 pipe break 2120 <1
- 2. 3 0 ft e 1615 0 2 1795 o 3 0.5 ft .l Study of the problem of thermal shock during core cooling system actuation by Babcock & Wilcox. Combustion Engineering.-General Electric. l and Westinghouse continues. Two modes of potential failure are being considered: ductile yielding and brittl'e fracture. Se latter is being -
treated using both the Pelli:.1-Puzak diagram approach and fracture ~
.J mechanics.
We are presently waiting for the results of cale'.:lations, promised by Westinghouse in a topical report, to establish the al stress distribution patterns near the crack tip as the crack progresses through the thickness of the vessel. Since the information submitted by Wes-inghouse, so far, in l
'I connection with the Diablo Canycn application is insu'ficient, we consider l the thermal shock problem unresolved at this time.
6.2 Engineered Safety Features for Heat Removal from the :entainment l l 1 The Diablo Canycn Containment vessel is designed for an accident i pressure of LT psig. The applicant was asked to perfera calculations to-l' show the capability of the containment to withstand vs.rious assumed energt releases during the course of an accident. The answers to these questions i l
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appear in Amendment No. 3 (pp 181-219). Engineered Safety Features for the containment structure cre redundant 'and the applicant's analysis shows that 1
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operation of 3 of 5 containment air coolers and 1 of 2 containment spray systems is adequate to maintain the calculated pressure belov design pressure.
We have reviewed the accident model and have concluded that the containment and its heat removal systems are adequately sized.
One aspect which we believe needs further attention during detailed design is that of leak detection on external recirculation systems. The l recirculation features are closely associated with the ECCS (for long term l l
heat removal) and our concern is that of detecting and being capable of isolating leaks in either of the tvc systems. If the leakage from valves and packings are within design limits, the dose contribution can be tolerated within Part 100 guidelines. If major leaks should develop during the re-circulation phase, activity leakage to the environment (there is no pro-vision for iodine removal in the auxiliary building ventilation system) could become excessive unless an operator has provision to detect and isolate i
\
j I the source.
6.3 Centrol Room Shie,1 ding The accident dose criteria for-this control room (including ingress and egress) is 2.5 rem whole tcdy and 300 rem tc the thyroid for the course of an accident. In our opinion, an iodine removal system should be in-corporated into the control room ventilation system er other measures should be taken to limit potential thyroid doses in the control rcom to values more in line with Criterion '1 considerations.
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- 7. Conclusions Assuming satisfactory resolution, as the final design evolves, of l
specific problems enumerated in the foregoing sections, we have concluded l
that there is reasenable assurance the Diablo Canyon facility can te built I
and operated at the proposed location withcut undue risk to the health and safety of the public.
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' "EHith 4#cJf.a %//AMM W APPENDIX A i 1.IST OF AMENDMENTS - DIABLO CANYON FACILITY l l
- 1. Amendment No. 1 dated July 10, 1967 which contained answers to questions, l l
design methods based on ultimate strength criteria, and described part l
l i length absorber rods.
- 2. Amendment No. 2 dated July 24, 1967 which contained answers to questions.
- 1 l 3. Amendment No. 3 dated July 31, 1967 which contained answers to questions I l
and additional information on site geology.
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- 4. Amendment No. 4 dated September 8, 1967 which provided a cross reference l to pages in the PSAR which dealt with each of the proposed General i
Design Criteria.
- 5. Amendment No. 5 dated October 18, 1967 which contained financial data, 1
additional tsunami information and revised information on the ultimate i
- 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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