ML20151X719

AEC Safety Evaluation in Matter of Licensee Plant

ML20151X719
Person / Time
Site:	Diablo Canyon, 05000000
Issue date:	01/23/1968
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML20151W779	List: ML20151W777 ML20151W807 ML20151W834 ML20151W945 ML20151W976 ML20151X022 ML20151X037 ML20151X057 ML20151X067 ML20151X076 ML20151X087 ML20151X097 ML20151X119 ML20151X125 ML20151X306 ML20151X322 ML20151X358 ML20151X381 ML20151X405 ML20151X436 ML20151X461 ML20151X466 ML20151X585 ML20151X612 ML20151X719 ML20151X755 ML20151X798 ML20151X821 ML20151X882
References
FOIA-88-156 NUDOCS 8808260086
Download: ML20151X719 (55)

Text

4 I l

, ( I January 23, 1968 .

l I

SAFETt EVALUATION BY THE DIVISIOR & PEACTCR LICENSING

- UNITED STATES AfCHIC ENERGY C(3 MISSION IN THE MATTER OF PACIFIC GAS and EIECTRIC CCMPANY DIABLO CANf0N NUCIZAR POWER PLANT DOCKET No. 50-275 8000260006 000721 PDR FOIA PDR MCMILL A88-156

s TABLE & CONTENIS Pa6e

... . .. ........... 1 1.0 INIROUJCTION . . . . . . .

2.0 SITE . ....... .. . . .. ...... ..... ... 3

.. .............. 9 30 NUCLEAR STEAM SYSTD( DESIGN 4.0 CONIAI MENT ...... . ... .............. 32 41 30 EMINEERED SAFETY FEAIURES . . . . . . . . . . . . . . . . .

6.0 ESIGN & CLASS I STRUCIURES FOR SEISMIC AND ACCIDENT LOADING . . . . . . . . . . . . . . . . . . . 47

. ... ... .............. 31 70 QUALITY ASSURANCE 8.0 ELECTRICAL SYSTEM ... . . .. ... ....... .... 31

... ... ........... 33 90 RADI0 ACTIVE WASTE CONTROL 10.0 ACCIENT AhALYSIS .. . .... ...... .... .... 33 11.0 RESEARCH AND EVELOIMENT ... ... ........... 38 12.0 REPORT & THE ADVISORY C004ITIEE ON REACIOR SAFEGUARDS ..... .... .............. 60 61 13 0 TECHNICAL QUALIFICATIONS . . . . ... ...........

61 14.0 C0h70RMANCE TO THE GERERAL IESIGN CRITERIA . . . . . . . . .

.. .............. 62 15 0 C04 MON DEFENSE AND SECURITY 16.0 CONCW SIONS ... .. .. ... .............. 63 APPENDICES A Chronology of Diablo Canyon Review ............ 63 B Report of Advisory Co=mittee on Reacter Licensing ..... 67 C-1 Report of the U. S. Veather Bureau ............ 73 I

s 1.0 INTh00]CTICH The Pacific Gas and Electric Conpany (applicant,P.G.&E. ), by application dated January 16, 1967, and subsequent amendments, requested a license to constmet and operate a pressurized water reactor at its Diablo Canyon site which is located in San Luis Obispo County, California.

The proposed reactor is designed to operate at 3250 W(t) with an expected ultimate capability of producing 3391 W(t). The applicant has designed the major components including tha containment structure and energency cooling system for a power leve) cf 3391 W(t), and has used this power level in analyzing postua .ted accidents in conformance to the guide-lines of 10 CFR Part 100. E have evaluated the containment and emergency cooling systems for 3391 W(t); however, the thermal and hydraulic character-istica vere evaluated at 3250 W(t). Before operation at any power level above 3250 W(t) is authorized by the Consnission, the Commission must perfom a safety evaluation to assure that the facility can be operated safely at the higher power level.

Our technical safety review of the proposed plant has been based on the applicant's Preliminary Safety Analysis Report (PSAR) and the nine The subsequent amendments, all of which are contained in the application.

technical evaluation of the preliminary design of the proposed plant was accomplished by the Division of Reactor Licensing with assistance from tb4 Divinion of Reactor Standards and various AEC consultants, as requested.

l Within Reactor Licensing, the Reactor Projects group vas responsible for the review, for coordinating parts of the review involving personnel within l

o s, .

The issues to be considered, and on which the findings cast be cade by an Atomic Safety and Licensing Smard before the requested license may be issued, are set forth in the Notice of Hearing issued by the Co:mnissio's l

and published in the Federal Register on January 13,1968, 33 F.R. 516. j l

2.0 SITE 2.1 Description _

The site for the proposed nuclear ple'.t is located adjacent to the Pacific 0:ean and Diablo Canyon Creek in San Luis Obispo County, California.

The site consists of 585 acres south of Diablo Canyon Creek which are leased to the company for a term of 99 years with an option to renev for an additional 99 years and an additional 100 to 203 acres north of the creek which P.G.&E. vill acquire. The exclusion area distance from the reactor to the nearest site boundary on land vill be one-half mile.

The 1960 populatipn distribution as a function of distance from the site, and the projected population for the year 1980 is presented in the following table.

Table 2.1 Population Distribution as a Function of Distance from Diablo Canyon Site 1960 1980 0-5 miles 12 80 5-10 miles 1,560 6,850 10-20 miles 47,630 111,460 20-30 miles 37,980 90,500

9 l -

netcorological measurements on a.250 foot tover near the plant location and on a 100 foot tover at the top of the 914 foot bill on the site.

Tracer diffusion studies using fluorwscent particles and smoke vill also be performed. We believe the proposed meteorological program is adequate to provide a basis for the development of a gaseous radioactive release limit and to confirm the conservati'sa of diffusion parameters used in the analysis of potential accidental releases.

23 Geology and Hydrology The geologic features of the plant site vere presented in the Preliminary Safety Analysis Report and in Amendment No. 3 As described, the plant structures vould be founded on bedrock, which is predominately sandstone.

Tne site has been invectigated by cutting four trenches down to the bedrock.

Tnis investigation revealed no evidence of a major fault in the area. There is evidence of surface disturbances, scue of which are faults, in the plant site. None of the breaks offsets the interface between the bedrock and the terrace deposits and none extends upward into the surficial cover. The a6e of the breaks at the site has been established to be at least 100,000 years, indicating that the possibility of fault-induced displacements at the site is sufficiently remote to be disregarded. The U. S. Geological Surysy (USGS) has reviewed the application ani other available literature, and has examined the exploratory trenches at the site. It has concluded that the applicant's analysis appears to be carefully derived and to present an adequate appraisal of those aspects of the geology which would be pertinent i

Earthquake A: A ca6nitude 8-1/2 carthquake (Gutenberg-Richter scale),

along the San Andreas Fault 48 miles from the site, resultin6 in a ground acceleration of 0.106 at the site.

Earthquake B: A magnitude 7-1/4 earthquake alon6 the Nacimiento Fault 20 miles from the site, resulting in a ground acceleration of 0.12g at the site.

Earthquake C: A magnitude 7-1/2 earthquake along the off-shore extension of the Santa Ynez Fault 50 miles from the site, resulting in a ground acceleration of 0.05g at the site.

Earthquake D: After-shock, with an earthquake ma6nitude 6-3/4 earthquake at the site, which results is a ground acceleration of 0.20g at the site.

The applicant has cesstructed response spectra for earthquake B normalized to a ground acceleration of 0.15g, and earthquake D noma 11 zed to a ground acceleration of 0.20s. For long periods the response spectra from earthquake B result in higher spectral accelerations %ah those from earthquake D. Analysis of the frequency distributions from the earthquakes shoved that the response spectra from earthquakes B & D give larger spectral accelerations than earthquakes A & C.

The U. S. Coast and Geodetic Survey has reviewed the applicant's analysis and agrees with the applicant's choice of predicted maximum ground accelerations as normalized. The U. S. Coast and Geodetic Savey report is attached as Appendix E to thic evaluation. On the basis of its report, ve have accepted the ground acceleration values selected by the l

.. e s , -f j l

years prior to plant operation. They propose to monitor airborne gn=a l

activity, air particulate activity, bovine thyroid, milk, leafy vegetables, and aquatic flora and fauna. The program proposed by the applicant vill provide a fim basis upon which the post-operationci environmental radio-activity monitoring program can be developed. The Fish and Wildlife Service evaluated the enviromental monitoring program and found it to be generally satisfactory, with the exception that additional vater and sediment analysis should be included. The applicant has stated that it will implement the recommendation in this regard. Comments of the Fish and Wildlife Service, Department of the Interior, are attached as Appendix 0.

2.0 Conclusions On the basis of the discussion in Section 2.'O ve conclude the site is acceptable for the proposed nuclear facility.

30 NJCIZAR SI1AM SYSTEN IESIGN 31 Bamma n Description The nuclear steam supply system consists of a light water moderated pressurized nter reactor (PWR) vtdch transfers reactor heat to four steam generators. Steam generated in the secondary side of the steam generators passes to a turbine-generator unit. The fuel for the reacter is lov enrichment UO2 Pellets enclosed within Zircaloy tubes 0.422 inches in outside diameter and about 12 feet in length. The fuel rods are arranged in a 15 x 15 array and are supported axially by nine spring grid assemblies.

Nozzles at the top and bottom, as well as the grids, are velded to 20 control rod guide thimbles and provide the structured support for the fuel.

l

s beneash the upper core plate to ensure they se not roved by hydraulic forces. Dissolved boric acid is used to compensate for fuel burnup and is controlleu by a chemical addition and control system.

The nucleac flux level in the core can be monitored by external neutron detectors and internal detectors. Indications Nom neutron flux as well as measured coolant pressure, temparature, and slow can initiate a reactor scram through the reactor protection system which causes the full length control rod assemblies to fall by gravity into the core.

i .mtor coolant at 2235 psig is circulated through the core by four centrifugal purpc. Dring rated operation the coolant enten the reactor After exitir.g the reactor, the at 539 F and exits from tne core at 602 F.

coolant passes throu4h .four steam generators which u of vertical U tube design. Secondary coolant in the steam generator receives energy throg$

the U-tubes from the primary coolant c<-using boiling at a secondary pressure of nbout ),000 psig. Saturated stesa is passed through the contaiment structure and enters the turbine. Exhaust from the turbine is condensed and pumped back through feedvater lines to the stesa gsinerators.

The nuclear steam supply system, as designed'oy Westinghouse, is sinilar to that of other pressurized water reactor systems that have been licensed. A new feature of this design is the us- of part length t. eol rod ersemblies to control axial power shape, and this feature is discussed in follovA -

ecctions of this evaluation.

32 Naelear D< 2 The 7 : ten heet " =g ed to operate at 3250 MW(t) to an average fuel burr.r e> ..

- t* iays per metric % n of uranium. The i

k f

.t

I

- 3-vorth of abo at .072dk resulting in a reduction of the initial operatir4 boron concentration to about 1200 p-m. The barnable poison rods vere included to eliminate a positive moderator temperature coefficient.

Calculations for this core have indicated that potential diametrical xenon oscillations vould be self-dampir4 vith a threshold moderator temperature coefficient of reactivity of -0.07 x 10 4 k/ T, vib the stability increased with more negative coefficients. Using barnable poison results in s calculated moderator temperature coefficient of reactivity at pcver from -0 5 x 10' d k[F to -3 0 x 10~dk/F. The effect of the negative moderator temperature coefficient of reactivity also rett.ces the potential serverity of a loss of coolant accident and rod ejection accident. Use of the burnable poison restits $n the same end-of-life negative moderator temperature coefficient, hence the potential consequences of a steam line break accident (evaluated at end-of life conditions) are unaffe %ed.

We have evaluated the analysis of the applicant and have concluded that the nuc h design is acceptable, with the reservation that adequat; nuclear instrumentation must be provided to determine the axial power distribution. The ACRS also made this point in its letter. If the part length rols were positioned to cause the axial power to be shifted to the top of the core, for example, unlepirable departure from nucleate be111n6 '"KB) sargins could occur. Nr conclusions on the instru=entation design to mon.itor povcr distribooo are discussed in Sec.'.on 3 4 of this report.

.. 15 The proposed mechanical design of internals, fuel assemblies, and control elements is adequate.

34 Thermal kLA Hydraulic Design The core design for the Diablo Canyon reactor takes advanta68 of reduced peaking factors made possible by the use of part length control rods. With reduced peakin6 factors, it is possible to increase the avers 6e Pover of the core 18% compared to previous designs, yet maintain the peak specific fuel power in line with past desi6ns.

The desi6n basis for the reactor is that the minimum Dh3 is not less than 13 during design basis transients. By some optimization of the inlet coolant enthalpy, but primarily by reducing the hot chancel factors, the mini =um DNBR for this core compared to earlier designs remains essentially the same. To illustrate this pcint a comparison of thermal. hydraulic parameters of the Proposed Diablo Canyon reactor and the Indian Point II reactor are presented in Table 1.

The expected fuel performance at the higher linear heat generation rate coupled with fuel barnup has been snelyzed by the applicant. A sumary bar chart shoving both the present and proposed irradiation test programs to demonstrate acceptable fuel perfomance for this reentor was p- ided in Amendment No. 6. We have evalaated the expected peak fuel rod operating characteristics far this reactor in relation to the test program. Thus far, there has been little operating experience at the linear power Eeneration icvels contemplated for the Diablo Canyon reactor.

It is telieved, however, that test programs to be performed at the Saxtop

4 s

and Zorita reactors vin confirti the expected fuel perfora.ance predicted for this facility prior to its operation. In the event such predictions are not borne out, fuel protection can be provided by 112nitations such as shorter fuel exposure.

One consequence of increased power flattening is that during operation or a design transient a larger number of fuel rods have DNB ratios in a specified interval. We have evaluated this nopect and have concluded that the miMmm DNBR of 13 for design transients as used in previously designed plants provides an adequate ma gin of safety.

One aspect of the design which the staff vill follow is that of instrumentation to assure that the axial power distribution is adeq'mtely controlled. The applicant has proposed that the four external flux monitors vill detect abnormal power patterns. The in< ore monitors for Diablo Canyon, as presently propose ' are six traveling flux probes which ray traverse any of 58 thimble locations in the core. These in-core channels are not designed to operats in the core at full power for more l

f than a few months. The applicant believes that the test programa (primarily at SENA) vill adequately demonstrate the capability of the external long ion etv.bers to predict power patterns within the core.

oar position in this rea,ard is that information from in-core me:dtora must te provided to an operator to position de partial rods in order to assure proper axial power shaping. We vould change our position if, at some lator date, e.cperience shove that the extsrnal monitors vill detect i

in-core ownalies with adequate sennitivity.

l l

19 potentially unsafe condition is sensed, the reactor protection cystem trips the reactor. The Diablo Canyon reactor protection system vill differ from that described in recent PSAR's for Westinghouse designed plants.

Changes were made to the systen in accordance with the provisions of the Proposed IEEE Standard on Nuclear Power Plant Protection Systems. Other changes were made as a resd.t of the higher power density and the use of partial length rods. Iecause of the known design changes in the instru-nentation, the staff specificany asked for additional infomation for the Diablo Canyon facility staff review. This additional information was presented in Amendments No. 7 and d.

The reactor protection system vi n be designed on a channelized basis to provide for isolation between redundant protection channels.

Isolation of redundant analog channels viu originate at the sensors and continue back through the field viring and containment penetrations to the analog protection racks. Isolation of field viring vin be achieved using separate vire vays, cable trays, conduit runs, and containment penetrations for each redundant channel. Redundant analog equipment vin be isolated by locating the equipment in four separate protection racks. The four racks of equipment win be ener$ized from separate alternatin6 current (a.c.)powersources.

j Each reacter protection system instrument channel vill teminate in The I

a reactor trip bisteble mounted in one of the four protection racks.

trip bistable ic the final operational component tu the analog channel.

7 The transition from reactor protection instrument channel identity to

O s

21-d.c. buses are igrtant in assuring that the first detectable failure does not fail the system. There is a sin 61e three phase a.c. bus between the trip breakers and the d.c. power supplies. 22e applicant has stated that this bus vill consist of totally enclosed bus bass. We believe that the voltage and current requirement (about 400 KVA at 260 volta three-phase) and the enclosed bus arrangement provides adequate assurance of meeting the single failure criterion.

The individual reactor protection channels feeding rasctor trip signals into the logic channels are as follows:

Coincidence logic Trip Parameter HiSh helear Flux (source range) One out of two

- high level HighNuclearFlux(intermediate One out of two ,

range) - hi6h level Hi6h helear Flux (pover range) Tve out of four lov power trip High helear Flux (pover range) Tve out of four high power trip Lev pressuriser pressure Tve out of four H16h pressurizer pressure Two out of four High pressurir.er water level Two out of three Turbine trip Two out of three Two out of three in any one loop Lov reactor coolant flev above75% power .

Two out of three in any two loops above10% power Reactor coolant pu=p breaker One out of one in any one loop opening above75% Power One out of one in a- vo loops above10% power

overpoverL T and over temper ature d T reactor trip settings if the power is unequally distributed.

Each instrumentation channel, both nuclear and process, which supplies a signal for reactor protection is read out in the control room. The re A-out allove the operator to detect failures in the analog portion of protection channels by cross-comparing channels monitoring the same variable and those monitoring variables having a known relation to each other.

The applicant has stated that the reactor protection system vill be designed, built and tested in accordance with the Froposed IEEE Stanclard on Nuclear Power Plant Protection Systems. The staff has examined the applicant's preliminary desi6n to evaluate its compliance with the Proposed IEEE Startiard.

Based upon our reviev ve have concluded that the applicant's proposed design meets the Proposed IEEE Standards with the possible exception of the section en Control and Protection Interaction. Section 4 7 of the Proposed Standards is addressed to th'e condition where a plant transient which requires protective action can be brought on by a failure or malfunction of a control system and the same event prevents proper action of a protection system channel designed to protect against the resultant unsafe condition. The proposed standard requires that after such a malfunction the rnatning portion of the protection system independently meet the sir 41t failure criterion. Plant designs in which the protection system and control syste=s are: not interconnected comply with Section 4 7 without further design provisions.

O

0 23 signals vill also be used te control feed flov and steam generator level. The low.lov steam generator level reactor trip uses a 2 of 3 lo6 c1 from any steam generator. One of the three level channels used for the reactor trip can be selected to control the level of the steam generator. The applicant in this case has stated that these trip functions are provided for equipment protection and are not necessary for reacter safety. If the analyses of the final design shows that these trips are not essential to safety, they vould not be within the scope of the Proposed IEEE Standard. If en the other hand, it is determined that these trips should be a part of the safety system, the applicant has agreed to confern it to the Proposed IEEE Standard.

In its review, the ACRS recommended that control and protection instrumentation be separated to the fullest extent practical. The applicant has stated that this rece smendatien vill be folleved.

On the basis of the foregaing, we have concluded that the applicant We vill review the design vill provide an adequate protection system.

of the system and refer the design of this system for further ACRS I consideration in accordance with the ACRS request.

3 6.2 Engineered Safety Features l

The applicant stated that the circuits vbich actuate engineered safety features vill also be designed to the Proposed IEEE Standard.

l It is proposed that the same channelized approach vill be used for these I

circuits as is proposed for the reacter protection system.

i y -

0 g the three channels cich actuate contr.in=ent i solation. We believe that this circuit, which vill be designed to the Proposed IEEE Stardard, is satisfactory.

We have discussed with the applicant the ability of the engineered safety feature electrical equipment to parfona its function in an accident environment. The applicant has stated that available data and additional testing as indicated vin be presented prior to operatign which vin prove the capability of this equipment to function in the conbi',1 temper.

ature, pressure, hunidity environment associated with the design basis accident. This equipmer.t includes cables, motors, detectors, and valve operatora located inside the containment which are associated with the en61 neered safety features.

3.6.3 Naclear Irstrumentation An originany proposed, the nuclear instrumentation and control room dispi,ay was co= parable to other WR designs except that no period or startup rate instrumentation was provided. As proposed only a log level recorder vould 'e provided to assist the operator in maintaining a controllable period.

We believe that period or startup rate indication over the range of neucron flux from telow criticality to plant bestup should be provided, and that the log level recorder can supplement the period information, but not replace it. Period indication reduces the possibility of initiatin6 trancients requiring the operation of the reactor protection system by providinE infonnation to the operator to avoid uncontrollable periods.

. 29 reac tor. Each centrifu6al pump is a vertical, ains,le stage pump vith botte 6uction and horizontal discharge havic6 a rated capacity of 88,500 gpn at 272 feet of head. Each steam generator is of vertical U-tube design with primary coolant circulated throu6h the U tubes. A pressurizer is connected via a 14 inch I.D. line to one of the 29 inch I.D. primary coolant pipes. Safety valves are included on the pressurizer which discharge to a relief tank within the containment. The componer.ts of the primary system are designed as indicated in Table 2.

Table 2 M ="7 Coolant System Design Component Code Design Design Requirements Pressure Temperature psig eF Primary vessel ASG III-A 2485 650 Steam Generator a Tube side ASE III-A 2485 650 b Shell side ASE III-C lo85 600 ASME III-A 2485 680 Pressurizer Coolant Piping ASA B31.1 2485 650 Safety Valves ASE III ----

The reactor vessel is fabricated from SA302 ors.ie B lov alloy steel with all internal surfaces clad with Type 304 austenitic stainless ste-l.

19 2 For design purposes a time integrated fast neutron flux of 3 7 x 10 n/cm is used, which results in a calculated NDT ahift of 2/5 F. The actual calculated fast neutron exposure based on 0.8 load factor for 50 years at

31-

!. 5 Secondary System Steam generated in each steam generator passes throu6h steam lines (one for each steam generator) via an isolation valve, step valve and control valve into the high pressure turbine. On each line upst.eam of the isolation valves are pressure relief valvec, atmospheric dump valves, Down.

and a line leading to the turbine-driven emer6ency feedvater pump.

strean fro:n the isolation 'talves there is a header which allovs bypassing steam directly to the condenser. This bypass line in conjunction with the atmospheric dump valves is designed to permit a cocplete lead rejection vithout a reactor trip. This system in conjunction v1th the resctor control system vill provide the capability of reducing reactor power and maintaining the unit at aux.iliary load after a net load rejection.

Steam exiting the high pressure turbine passes throu6h reheaters where it is heated by condensin6 steam from the bypass line, and then it is delivered to the lov pressure turbine elements and the turbine-driven feedvater pu=ps. After passing throud;h the low pressure turbine elements, the steam is condensed in the main condenser. Condensatt is then pumped throu6h generater hydrogen coolers, stator coolers, gland steam condensers and the steam jet air ejector by condensate pumps to the suction of the condensate booster pumps. Three hal#-size condensate pumps and three hs,1f-site condensate booster pumps s,re provided. The condensate booster pumps deliver the condensate throu6h five sta6es of feedvEter heaters to the feedvater pu=ps. Two half-size feedvater pumps deliver condensate through hi6h pressure heaters to individual lines leading to each stea:s

. l l

l l

\

and it vin be 11+2 feet fro:n the base slab to the springline of the doze.

l The structure vill be 38 feet thick in the cylindrical portion, and the da:ne vin be 2) feet thick. The steel liner vill have a mini e thickness of 3/8 inch in the cylinder van and dome, and vill have a minimum thick-nessoffinchinthebaseslab. The liner vin be anchored to the reirforced concrete with L shaped anchers fastened to the reinforcing and velded to the liner plate on a spacin6 of 20 inches vertically and hori-r.entally. The free volume of the containment vessel vill be about 2.6 x 10 cubic feet.

Internal atmetures consisting of equipment supports, shielding and floors are supported by the concrete base mat which in turn rests on bedrock. The internal concrete structures win be designed to surround .

the pristry coolant systems to provide for dad 1* 'prc,tection in the event of component failure. The dewign vill also include provision to withstand forces associated with a double-ended rupture of a main coolant pipe.

Missile protection vin include protection for failures of valves including valve stems and bonnets, instmment thimbles, closure bolts, and essplete control red drive mechanisms.

Further detail en the pre 14mim design is presented in fenoving sections.

k.2 Containment M ading Factored loads for the design of the containment stmeture have been proposed which combine dead leads, pressure loads, temperature loads and The earthquake loads (or vind loads if greater than the earthquake losis).

usin6 the Vestin6 house FLAS11 code, and calculations of the conts,inment energy camputed throu6h use of the Westin6 house COCO code. The design basis accident 2

(3ft break) results in a peak containment pressure of 39 6 psig.

Analyses vere nise performed by the applicant to demonstrate the siequacy of the containment design pressure by including additional energy from internals and reactor de%y heat, and by including metal-vater reactions.

Inclusion of a metal-water reaction which involves about 32 percent of the Zirealoy within 1000 seconds causes a peak containment pressure of 45 7 peig, which is below the design pressure.

Earthquake loadings vill be computed on the basis that the vertical acceleration values are two-thirds or the horizontal Breund acceleration values with the effects of the horizontal and vertical loadings combined on the assu:sption that they act simultaneously. The design basis earthquake and tvice the design basis earthquake are defined in tetts of acceleration in Section 2 of this report. Our cenaultants have reviewed the seismic aspects of the containment design and their conclusions, with which we adopt, are included in Appendix F to this report.

4.3 Structural Design Details fbe preliming design of the reacter containment structure contains two new features which we consider particularly significant, a helical reinforcing pattern in the concrete shell, and a hingo at the base et the cylindrical vall. The concreta cylinder is reinforced with helical bars, inclined at an an61e of 30' from the vertical. The vall reinforcing bars are continuous with the done reinforcing. Additional heep rein *ercing is w

. 37 be subjected to the folleving tests in addition to these required by the manufacturer:

ASN C 114, Standard Methods for Cnemical Analysis of Hydraulic Cement.

ASm C US or ASM C 204, Standard Method of Test for Fineness of Portland Cement by the Tv.rbidimetet, or Standard Method of Test for Fineness of Portland Cement by Air Permeability Apparatus.

ASM C 191, StaMard Method of Test for Time of Setting of Hydraulic Cement by Vicat Needle.

Construction tests of the concrete vill be peri'ormed by a group other than the contractor, taking trial mixes from the contractor for testin6 The testing vill include tests as follows:

ASN C 231, Standard Method of Test for Air Content of Freshly Mixed Concrete by the Pressure Nethod.

ASm C 143, Standard Method of Test for Slump of Portland Cement Concrete.

AS M C 232, 8tandard Method of Test for Bleeding of Concrete.

ASW C 192, Standard Meths,1 of Making and Caring Concrete Compression and Flexure Test Specimens in the Imboratory.

ASM C 39, Standard Method of Test For Compressive Strength of Molded Cenerete Cylinders.

In addition to tne tests en trial mixes, after construction has started field tests vill be performed using appw:riate methods listed above plus the felleving:

AS N C 172, Stsndard Method of Eampling Fresh Concrete.

b-

. 37 ASIM A 20, General Requirements for Delivery of Bolled Steel Plates of Flange An.1 Firebox Qualities.

Qus.lity assurance during conetruction is discussed in Section 7 of this report.

45 structural Testing A program of acceptance testin6 has been indicated which we believe vill provide a high degree of assurance that anomalous structural behavior vin be detected.

Detailed attention is being given to liner inspection during construction.

Use of vacuum texes has been proposed to identify and allev correction of potential source 5 of containment leakage as construction proceeds. Applicable portions of the ASME Boiler and Pressure Vessel Code,Section VIII, vin be foneved during construction. Tne amount of radiography and other tests proposed by the applicant is sdequate to assure elding quality.

In addition to the leak tests perfomed e.2 the individual velds, a proof test of the containment vin be made at 54 psis and an integrated leak rate test of the containment structure at 47 psi 6 vin be performed.

4.6 Centainment leaisse An integrated leak rate test of the completed containment structure vin be performed at 47 peig vith an douule penetration and veM channel zones open to toe atmosphere. This test vin be performed in contwaance to ANS 7 60 of the American Fuelear Society. This integrated leak rate test vill be performed to demonstrate that the total leakage rate is less than 0.1% of the containment volume per day, t _

. .bl.

5.O ENOIEERED SAFETY FTA'NRES 51 meign _of Ihergency core ceeling systems The applicant's design basis for the emerr,,ency core coolin6 syste:n is to adequately cool the core and to limit potential radioactive releases from the fuel folleving less of coolant from the primary system. To provide adequ ste cooling of the core requires that core geometry be maintained and that clad tempers 6ture and metal-vater reactions be limited daring the accident.

The applicant's criterion for maintenance of mechanical integrity during the blevdevn is that deformation of reactor internals shall be limited to ensure the capability to scra:n control rods and also to cool the core. The applicant has proposed deformation limits to be met in the analysis of blowdown forces on reacter internals.- As discussed in Section 6 of this report, we believe that the applicant's basis for design to maintain mechanical integrity during blevdevn is acceptable. As recommended by the ACRS, we vill review the results of the detailed blevdown calcula-tions more fully when these become available.

Core cooling for any location and site of primary coolant pipe break up to the double-ended rupture of a recirculation pipe vill be provided by high pressure injection pumps, low pressure injection pumps ant accu:r.1-lators. We have discussed with the applicant the degree of redundancy It is our required to meet the design basis for tia system given above.

position that redundant systems abould be provided to the extent that an active co=ponent failure during betb short and long tem cooling, or a w .

1

/ -43 l

l i

I The safety injection system as proposed includes four accumtlators,  !

tve- hi6h-head safety injection pumps and two lev-head residual heat ,

)

removal pumps. Each accumulater is connected to a cold leg on the reactor coolant piping. Each accu:sulator contains about 770 cubic feet of berated water with compressed nitrogen gas occupyirg about 330 cubic feet, puring normal operation the accumulators are valved inte the primary coolant system with only check valves being closed. In the event of a loss-of-coolant accident, when the pressure decays to the n!trogen gas pressure (about 600 psig) flow to the coolant system is established auto =atically.

Each of the two safety injection pumps are rated to deliver about 800 gpn at 2500 ft. of head. The discharge of these pumps is into two hot legs and two cold legs of the reacter coolant piping. The residual heat removal pumps take suctf or. from the containment sump after the injection system has exhausted its source of vater from the refueling vater tank. These pumps deliver coolant through the cold legs of the four reacter coolant pipes. Each residual test removal pump is rated to deliver about 3000 gpm at 350 feet of head.

We have revieved the capability of the Safety Injection System to meet the design objectives. The performance of the system with 3 of 4 accumulators and 1 of 2 safety injection system pumps can be nummaritei as follevs:

l l

1

containment air coclers and 1 of 2 cone 'n- st spray systems is adequate to maintain the calculated pressure belev design pressure. We have reviewed the accident model and have concluded ths; the containment and its heat removal systems are adequately sized.

The containment spray system is also used to remove radio-iodine from the containment atmosphere in the event of an accident. Initially sodium thiesulfate was the chemical additive to be used in the borated water for iodice remeval, however the use of other additives is beitg studied with the final choice dependent upon a research and development program. The applicant has also reserved space within the containment to install charcoal absorbers in the event this reses,rch and development program does not result in development of a suitable chemical. additive for the sprays.

One aspect which the staff believes should receive further attention during detailed design is that of prevision for leak detection and for isolation of a leak on the external recirculation systems. The recir-culation features are also associated with the ECCO for long tenn heat I

removal and ear concern is with being able to detect, and having the capability of iselating, leaks in either of the two systems, i .

5 3 Auxiliary Electric Fever _

The en61ceered safety feature electrical leads are connected to f

three 14160 volt vital buses. When off-site power is not available, I

cach bus is povered by a separate iiesel gecerator. The redundancy of f

nafety feature leads e.nd the arrargement of leads en the buses is such C

47 6.0 DESIGN T CLASS I SirRUCWRES Fm SEISMIC AND ACCIDENT LGADINGS 6.1 General _

These structures, 3ystems and components of the nuclear plant which are important to nuclear safety, i.e., failure of which might cause or increase the severity of a loss-of-coelant accident er result in the release of excessive acnunts, of radioactivity, are ter-ned Class 1.

These components which are not essentis1 to the safe shutdown of the reactor and failure of which would not result in the release substantial amounts of radioactivity, are considered Class II. Those stnactures and components not related to reactor operation e.re Class III. The subsequent discussion on design criteria contained in this section relates to the Class I etructures and components.

6.2 Earthquake Response Spectra and Da= ping The magnitude of ground acceleration for earthquakes at this site, including the review by the U. S. Coast and Geodstic Survey was discussed previously in Section 2. The response spectra for the 0.15 g ground acceleration earthquake ("far avay") and that for the 0.20 g ground acceleration er.rthquake ("close-by") were presented in she applica'. ion.

Todether these response spectra constitute the design basis earthquake.

A larger 0.40 g ground accel'eration earthquake response spectnxm is aise contained in the applicatioh which represents what can be considered the maximum potential earthquake.

Darity detailed design, vibrational response characteristics vill be calculated for the assumed earthquakes. We Agre with the ACRS that s

. 49 (b) Load combinations including exn earthquake loads and applicable design basis accident loads, without loss of function of the specific structure, eyeus, or co=wnent.

The applicant presented in Amendment No. 5 a document entitled "Ultimate Strength criteria to Ensure No Ioss of Function of Piping and Vessels under Earthquake Leading," WCAP-5890, Revision 1. This docunent

ontains design criteria proposed as the basis for detailed des 16L of Class I vessels and piping.

As a result of discussions with the applicant, the stress-strain criteria for load combination (b) have been redefined in Amendment No. 9 to limit strains to more conservative values. We have revieved there subnittals and we consider the leading combinations and limits now The proposed by the applicant to be both realistic and satisfactory.

proposed stress or deformation limits for specific components sre discussed in more detail belev.

Reacter Vessel Internals To be able to perform their function, i.e. to allev core shutdevn and cooling, the reac%er vessel internals must satisfy defemation limits that are *. sere restrictive than the stress limits. The applicant stated that the internals v111 oe designed to withstand nomal design leads plus earthquake loads within Section III limits, with the exception of materiale not esvered by the Code, such as fuel red cladding. Seismic stresses vill be combined in a conservative way and will be considered as primary stresses.

-5L 70 QUALITY ASSURANCE The applicant has described its program for quality assurance in the application. The staff has also discussed this area in additional detail with the applicant. The staff considers it essential that in addition to designing structures and equipment to adequete codes and specifications, and requiring contractors to build and inspect in conformance to these codes, that a group independent of the construction force itself also inspect the quality of vorkmanship and have authority to stop construction if any deviations are noted.

The applicant proposes that materials and componeLte furnished by all subcontractors vill be revis,ved by Vestinghouse or FG&E engineers.

The constmetion for PCAE v111 be carried out by its Construction Division.

The verk of the Constmetion Division and of Vestinghouse vill be reviewed independently by PGLE'c Engineering Department. Tnis group has independent responsibility to test and inspect caterials and has authority to stop construction if discrepancias are noted from the specifications or plans.

We believe that the quality assurance program as proposed is adequate.

8.0 EIECTRICAL SYSTEM The preposed electrical transmission system for the facility includes two 500 kv lines and tvo 230 kv lines. Electricity generated at 25 kV 6.

the facility vill feed the main transformer and tvo unit auxiliary trans-formers. The main transformer vill step the voltage up to 500 kv to supply N

. .s 3 complate load rejection without tripping, the primary source of power vould be the main generator. In the event of a turbine generator trip, the auxiliary load would be auto =atically transferred to the Stand-by startup transformers. In the event power to the vital buses cannot be successfully restored, a pro 6racced svitchiD4 vill occur and diesel generators vill be automatically connected to these buses and the essential loads restored in a proper sequence.

We believe the external power sources fot use at the plant provide a hi6h degree of assurance that power vill be available when required.

As with previous applications, however, ve consider that adequate on-site power should be available to meet the sin 61e failure criterion.

90 RADI0 ACTIVE WASTE CONIT01 2 The sizing of the vaste handling and storage equipment has been per-formed on the basis of continued reactor operation with clad defects in 1% of the fue4 rods. The primary system water chemistry is maintained by the chemical and volume control system. Coolant is taken from a cold leg of t'ne reactor coolant system, reduced in pressure, passed throu6h a heat exchanger, and passed throu6h the demineralizers as necessary and then routed to the volume control tank. Water level control in the primary system is provided by pumping the vater in the volume control tant through the seal water lines or via charging lines into the primary system. Addition to er dilution of the boric acid content of the vater is also accomplished by this sy. item by making up vater to the volu:.e control tank from the chemical addition system.

\

35 10.0 AOCIDENT ANALYSIS 10.1 Incidents _

A nunber of plant operating transients vere considered by the applicant including rod withdrawal during startup and from power, moderator dilution, loss of. coolant flow, loss of electrical load, and loss of off-site AC pover, in order to assess the safety margins of the plant design. The criterion for detailed design of the reactor control and protection systen is to be able to automatically take corrective action to cope with any of these transients. Preliminary analyses as presented in the application vill be recalculated during detai'.ed plant design to verify that trantients are within the capabilities of the reactor control and protection systems. Previous staff evaluations of other IVR designs at the operating license stage have denonstrated that anticipated transients have been terminated with adequate margia to a minimum DNB ratie of 13, and we believe that this limit can be met in this facility.

The ACRS in its letter reconsnended that consideration be given to the developa.ent and utilization of instraentation for prompt detection of grosc failure of a fuel element. The applicant has a6 reed to give this question further study to determine its feasibility.

10.2- Accident Evaluation Potential accidents which could result in radioactive releases to the environment have been analyzed by the applicant. We have evaluated these occidents and the engineered safety features provided to limit the potential exposures. Accidents which have been considered are:

.' available exclusion zone radius (0 5 mile) and the lov population zone rs.iius (6 miles) vithout any thyroid dose reduction facters needed.

For accidents involving less of coolant from the primary system the emergency core cooling systems are designed to limit fuel cladding tempera-tures as described in Section 5 te vel.1 delev melting temperature. Althegh the design basis for sizing the emergency core cooling system is to limit fission product release from the fuel, ve continue to tske the position that the containment and its associated engineering safety features shall be capable of limiting potential deses in confermance to 10 CTR Part 100 guidelines by assuring releases of fission products fron the fuel. For the less of coolant accident which results in the TID 14844 fission product releasefractions(100%noblegas,'25%iciine,and1% solids)available for leakage, with no further iodine reduction, the staff has calculated potential deses by aise using the felleving conservative assu=ptions:

1. Ateorelegy - Oreund release, centerline, Pasquill Type F, 1m/sec.,andvakeofthebuilding(volumetricsourceand e e 1/2) for the first 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the accident; fren 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ground release, Pasquill Type F,1 m/sec., unifers dispersion into a 22-1/2* sector; and 1 day to 30 days - the stability, vind speed, and direction vere varied.

2. Imak Rate - 0.1%/ day for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 0.045%/ day for the duration of the accident.

The potential deses at the exclusion area boundary a:e 6 rem whole t+dy and 870 rem to the thyroid. Potential doses for the coure- of an

l l

l o

'important aspect of emergency core cooling is the. prediction of the mode of fuel failure of the fuel rods as related to the l ability to provide core cooling. Studies have indicated that localir.ed cladding damage, rather than gross dama6e, vould be the mode of failure and this mode of failure should not impede cooling of the cladding. In Amendment No. 6 verk which is being conducted by Westin6h ouse to confirm the expected fuel behavier during the postulated less of coolant accident is described. As stated in Item No. 3 of Amendnent No. 6, deferv.ation and burst characteristics of the fuel cladding vill depend upon inter-related parameters of the cladding strength and ductility, internal pressure within the fuel reds, and the fuel red heating rate.

The experimental pro 6 ram tw ^<.st these inter-related parameters is described in Amendment No. 6. We a6ree with the A38 that further evidence should be obtained to show that fuel red failures in less of coolant accidents vill not significantly effect the ability of the BCCS to prevent clad melting. ile intend to evaluate the test data ha it becomes available.

2) Development of Tinal Core Thermal-Hydraulic, Nuclear and l

i Mechanical Design Parameters- Included in this area art the studies en use of part length control reds and burnable poison reds, additional analytical refinements in reactivity transients, analytical studies on predicted DNB raties during operating and transient conditions, and evaluation of the effects of blevdevn

, - --n..- ,

, ./' .

The report of the ACRS concluded ". . .The Ce::::nittee believes that with due concideration to the foregoin6 items, and in view of the isolated site, the nuclest plant proposed for the Diable Canyon site can be constructed with reasonable assurance that it can be operated without undue risk to the nealth and safety of the public."

13 0 TECHNICAL QUAIITICATIONS The applicant, Pacific Gas and Electric Ce:pany, has exteasive experience in the design, construction and operation of electric generating plants. Pacific Gas and Electric Ce:pany personnel have been involved with nuclear power generation for a number of years having operated the Hu:nboldt Bay Power Plant in co=:ercial operation since 1963 The nuclear steam system supplier, Westic6 house Electric Corporation has designed and constructed a number of pressurized water reactors vbich have been licensed by the Ce= ission.

On the basis of the above considerations, and based upon our evaluation of the responsible personnel, ve believe that the applicant is qualified to design and construct the proposed facility.

14.0 CONFCfMANCE TO THE GENERAL DESIGN CRITERIA In November 1965, the Cat =nission published its General Design Criteria for Nuclear Power Plant Construction Permits, and on July 11, 1967, published in the Federal Register its revised General Design Crf teria taking into account ce:n=ents received on the initial criteria and further developc:ent of the criteria by the regulatory staff. The applicast in A. endment No. 4 cross-referenced the infonnation as presented in the application with the

16.0 CONCWSIONS Based on the proposed design of the Pacific Gas and Electric Company's Diable Canyon Facility, on the criteria, principles and design arrangements for systems and components thus far described, vbich include all of the important safety items and on the calculated potential consequences of routine and accidental release of radioactive materials to the environs, on the scope of the development program which vill be conducted, and on the technical competence of the applicant and the principal contractors, ve have concluded that, in accordance with the previsions of para 6raph 50 35(a),10 CFR Part 50 and paragraph 2.lO4(b) 10 CTR 2:

1. The applicant has described the proposed desi6n of the facility, includin6, the principal architectural and engineering criteria for the design and has identified the Wer features or cosponents for the protection of the health and safety of the public;

2. Such further technical er design information as may be required to complete the safety analysis ami which can reasonably be left for later considerations, vill be supplied in the final safety analysis reports; 3 Safety features er components, which require research and develop- j zent have been described by the applicant and the applicant has j identified, and there vill be conducted, a research and develop-ment program reasonably designed to resolve any safety questions associated with such features or components; l

l l

I

• l i

16.0 CONCWSIONS  :

l Based on the proposed design of the Paci*1c Gas and Electric Company's Diable Canyon Facility, on the criteria, principles and design arrawngements for systems and conponents thus far described, which include all of the important safety items and on the calculated potential consequences of routine and accidents 1 release of radioactive materials to the environs, on the scope of the development program which vill be conducted, and on the technical conpetence of the applicant and the principal contractors, we have concluded that, in accordance with the provisions of paragraph 50 35(a),10 CFR Part 50 and paragraph 2.104(b) 10 CFR 2:

1. The applicant has described the proposed design of the facility, includin6, the principal architectural and engineerin6 criteria for the design and has identified the Wer features or cosponents for the protection of the health and safety of the public;

2. Such further technical or design information as may be required to complete the safety analysis and which can reasonably be left for later considerations, vill be supplied in the final safety analysis reports; 3

Safety features or components, which requin research and develop-ment have been described by the applicant and the applicant has identified, and there vill be conducted, a research and develop-ment program reasonably designed to resolve any safety questions associated with such features or components;

APPENDIX 'A Chronology of Diablo Canyon Reviev Event Date_

1) January 16, 1967 Date of Application for Construction Fer.:dt.

2) March 21, 1967 Representatives of Applicant and Regulatory Staff met to discuss seismic design aspects.

3) April 20-21,1967 Representatives of Applicant and Regulatory Staff met to discuss general design.

4) May 5,1967 Questions on cite, plant layout and design of major structures transmitted to the applicant.

5) May 16,1967 Representatives of the Applicant and Regulator Staff met to discuss seismic design.

Questions concerning instrumentation and

6) May 18, 1967 control systems transmitted to the applicant.

7) June 30, 1967 Questions concerning reacter design, engi-neered safety features and accident analyses trans:aitted to the applicant.

Amendment No.1 containing ansvers to

8) July 10, 1967 questions, design methods based on ultimate strength criteria ani description of part length absorber rods filed.

9) July 19, 1967 AcRs Subcommittee visit to the site.

Amendment No. 2 containing ansvers to

10) July 24,1967 questions filed.

Amendment No. 3 containing anr.rers to

11) July 31, 1967 questions and information en geology filed.

12) August 15, 1967 Representatives of the Applicant and Regu-lator/ Staff met to discuss scismic design.

Questions concerning research and develop-( 13) August 31, 1967 rent prograas transmitted to the applicant.

l l

[

l

o /J7FiTIX B ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

' UNITED STATES ATOMIC ENERGY COMMISSloN WAS HIN GToN. O.C. 1056 DEC 2 0 B57 3 .,

Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission #

Washington, D. C. .

Subject:

REPORT ON PACIFIC GAS AND ELECTRIC COMPAhY NUCLEAR UNIT --

DIABLO CAln0N SITE

Dear Dr. Seaborg:

At its ninety-second meeting, on December 7-9, 1967, the Advisory Committee on Reactor Safeguards completed a review of the application by the Pacific Gas and Electric Company for authorization to construct a nuclear unit at its Diablo Canyon Site, in San Luis Obispo County, California. This project had previously been considered at the Commit ,

tee's ninetieth meeting, on October 5-7, 1967, and at Subcommittee meetings on October 4,1967 and December 1,1967. ' Seme members of the Committee visited the site on July 19, 1967. During its review, the Committee had the benefit of discussions with representatives of the Pacific Gas. and Electric Company, the Westinghouse Electric Corporation, and the AEC Regulatory Staff and their consultants. The Committee also had the benefit of the documents listed below.

The Diablo Canyon site comprises approximately 800 acres adjacent to the Pacific Ocean on an isolated promontory between Morro Bay and Pismo Beach.

Eighteen people live within six miles of the site; the nearest boundary of the City of San Luis Obispo (1965 population of 25.750) is 10 miles distant; and only three cities of more than 10,000 population are located within 60 miles.

The contavinment structure, which encloses the reactor and steam generators, will consist of a steel-lined concrete shell in the form of a reinforced-concrete vertical cylinder with a flat base and a hemispherical dome. This and all other Class I structures and components have been designed not to exceed normal working stress or deflection limits during a design earthquake of 0.2 g acceleration and to assure no loss of function at twice this ground acceleration. In addition, protection will be afforded against seismic sea

{

- - . , ~ .

i.

i lionorabic Glenn T. Scaborg ggg 9 ; 6 07

~

The higher power and increased power density of this reactor, compared to similar pWR's previously approved, place increased dependence on correct manipulation of control rods. The information available to the operator (rod positions, neutron flux and temperature profiles) must be sufficiently reliabic, complete, and comprehensible that the propo' sed procedural control can be effective in achieving the predicted flux peaking factors. When information becomes available from large PWR's scheduled for operation earlier than Diablu Canyon, the applicant and the Regulatory Staff should review carefully whether the adequacy of the proposed system for high-power-density operation is justified by the data.

The applicant indicated that a system of fixed in-core neutron detectors and continuously operating readouts could be added, if later shown to be ,

necessary, with protection functions if needed. Additionally, the Commit-tee believes that the operator should have available to him readouts of the positions of all the control rods, without the necessity of switching a single indicator to each of 61 rods.

~

'The applicant has proposed using signals from protection instruments for control purposes. The Committee believes that control and protection instrumentation should be separated to the fullest extent practicabic.*

The Committoo believes that the present design is unsatisfactory in this respect but that a satisf actory protection system can be designed during

. the construction of this reactor. The Committcc wishes to review an

• improved design prior'to installation of the protection system-Consideration should also bc given to the development and utilization of instrumentation for prompt detection of gross failure of a fuci element.*

f During the course of final design and construction, studies will be made to determine the vibration characteristics of the major reactor components and the response of safety instrumentation to seismic loadings. Considera-tion should be given to obtaining experimental verification, to the extent practical, of the anticipated behavior in earthquakes of important components and instruments. It is also desirable that, prior to reactor operation, meant be developed and provided to guide or implement decisions concerning reactor

operation in the event of a large earthquake in the region of the site.

The Committee continues to emphasize the importance of quality assurance in fabrication of the primary system and of inspection during service life.*

Because of the higher power level and advanced thermal conditions in the Diablo Canyon reactor, these matters assume even greater importance. The Committee recommends that the applicant implement those improvements in-primary system quality which are practical with current technology.

l s 69 .

i l

l

_ _ _ _ _ _ _ _ _ _ . . ~ __ __. . _ _ _ _ _ . - _ _ _

_ - - - - - _ . - . - . . - . - . ~.- -- - - . - -

Honorabic Glenn T. Scaborg DEC '* O E67 References - Diablo _ Canyon

~

1. Letter fro >n Pacific Gas and Electric Company dated September 30, 1966; Preliminary Site Report, Diablo Canyon, dated September 30, 1966.

. 2. Letter from Pacific Gas and E1cetric Company dated January 16, 1967; i' Nuclear Power Planta Diablo Canyon Site, License Application, dated January 16, 1967; Prel.iminary Safety Analysis Report, Volumes 1 and 2.  ;

3. Letter from Pacific Gas and Electric Company dated July 10, 1967; Amendment No. I to License Application, Pirst Supplement to Prelicinary Safety Analysis Report.

4. Letter from Pacific Gas and Electric Company dated July 24, 1967; Amendment No. 2 to License Application, Second Supplement to Preliminary Safety Analysis Report.

i 5. Pacific Gas and Electric Company letter dated July 31, 1967; Amendment

! No. 3 to License Application Third Supplement to Preiteinary Safety

, Analysis Report.

. 6. Pacific Gas and Electric Company letter dated October 18, 1967; Amend- .*

ment No. 4 to License Application.

7. Pacific Gas and Electric Ccmpany letter dated October 18,1967; Amend-mont Nc. 5 to License Application, Pourth Supplement to Preliminary Safety Analysis Report.

8. Pacific Gas and Elcetric Company letter dated November 6,1967; Amend-ment. No. 6 to Licer.sc Application, Pifth Supplement to Preliminary i Safety Analysis Report.

9. Pacific Cas and Electric Company letter dated November 9,1967; Amend-ment No. 7 to License Application, Sixth Supplement to Preliminary

- Safety Analysis Report. ,

10. Pacific Gas and Electric Company letter dated November 30,1967; Amend-ment No. 8 to License Application, Seventh Supplement to Preliminary Safety Analysis Report. ~

11. Pacific Gas and Electric Company latter dated December 6,1967; Aliiend-ment No. 9 to License Application, Eighth Supplement to Preliminary ,

Safety Analysis Report. ,

. -71 =

• APPENDIX C-1 Comment s on Diablo Canyon Site Pacific Gas and Electric Company Preliminary Safety Analysis Report Volumes 1 and 11 dated January 18, 1967 Prepared by Environmental Meteorology Branch Institute for Atmospheric Sciences March 20, 1967 As discussed in our comments on October 19, 1966, the important features of the transport and diffusion climate of the Diablo Canyon site are the high frequency of onshore winds f rom the west and northwest, the persistent occurrence of a marine inversion at about a height of 1000 to 2000 feet above sea level and the rough mountainous terrain rising to a height of over 1000 feet within a distance of 3000 feet from the shoreline. While at first glance the semi-permanent existence of a marine inversion would suggest poor dilution conditions, it is necessary to consider that good dilution usually exists under the inversion lid during the daytime onshore flow. Thus, an ef fluent released at the site near the ,

ground would probably undergo considerable mixing in the first mile '

or so before being restricted in the vertical, by the inversion alof t.

The stabilizing effect of air trajectories over a smooth, cold water surface before reaching the site is rapidi1y changed en an unstable ef fect within a few hundred feet of travel over rough and heated inland terrain.-

The dif fusion parameters used to calculate the site dispersion f:ctors (Table 2-4) are conservative. A ground source is assumed as well as inversion conditions (Pasquill F) with a wind speed of 1 m/see for periods up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. No credit is taken for the meandering of the wind direction over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. ' The adjust-ment to take into account building. induced turbulence amounts to a f actor of 4 at the site boundary of 800m, which is reasonable.

It is noted that a comprehensive on-site meteorological program is planned including a meteorological tower at the reactor site and on the 914-f t on-site hill as well as surf ace measursments at four other locations and a series of smoke and fluorescent particle tracer tests.

In sammary, no unusual meteorological aspects are anticipated with regard to the safety analysis of the Diablo Canyon site. The assumption of a ground level source, low wind speed, inversion conditions and a constant mean wind direction over a 24-hour period is quite conservative.

73 l

-- - _____-___ D

AFFENDIX C-2 Comment s on Diablo Canyon Site Pacific Gas and Electric Company Preliminary Safety Analysis Report Third supplement dated July 31, 1967 .

Prepared by Environmental Meteorology Branch Institute for Atmospheric Sciences Environmental Science Services Administration August 28, 1967 In regard to question VII.G.5, it is obvious f rom the comparison made by the applicant that there is very little difference beyond the first 300 meters between using a virtual point source or a volumetric source technique to account f .r added building-induced diffusion. It should be pointed out that both techniques are an empirical means to account for a phenomena which is not clearly understood but which L1 observed. Por comparative purposes it is interesting to note the ratio between concentrations without accounting for building turbulence and the volumetric approach. The table below shows the Diablo Canyon calculations compared to recent fiwid tests at the National Reactor Testing Station conducted under inversion conditions and moderate wind speeds (G = 6.3 m/sec).

Di stance (meters) Diablo Canyon EBR-II Tests 100 3.5 4;2 200 300 9.8 400 6.2 3.9 600 3.2 700 3.2 800 2.8 1000 2.3 Extrapolating the results, one would conclude that an .added dilution factor of 2.8 is reasonable at the alte boundary of 800 m.

e f

APTENDIX D C 3,7 7 %; i !TED ' ~"S

• Maj DEPARTMENT OF THE INTERIOR '

"*** GEOLOG: CAL SURVEY WASHINGTON. o C. 20242 SEP 211967 Mr. Harold L. Price .

Director of Regulation U. S. Atomic Energy Consnission

.. 4915 St. Elmo Avenue Bethesda, Maryland 20545

Dear Mr. Price:

Transrsitted herewith in response to the request of Mr. Edson G. Casa dated February 6,1967, is a review of the geologic and hydrologic aspects of the license application of the Pacific Gas and Electric Co=pany, Diablo Canyon Nucicar Plant Site.

This review prepared by Henry W. Coulter and Eric L. Meyer of the Geological Survey has been discussed with ce=bers of your staff and we hc.ve no objections to your t:aking it a part of the public record. .

Sincerely yours, Actin 8 Director Enclosun s .

4 I

4

-y

Diablo Canyon Site AEC Docket 50-275 Hydrology The site is on the shore of the Pacific Ocean in San Luis Obispo County, near the mouth of Diablo Canyon Crcok. The reactor is to be located on a terrace on the left bank of the Creek at a grade altitude I of 85 feat above mean sea level. Cooling water is to 'be obtained from an intake at the ocean south of the reactor and discharged about 1,200 feet north of the inta'tc. A spit of land extends about 1,000 feet into the ocean between the intake and discharge points.

The reactor location would not be affected by floods of Diablo Canyon Creek, the only developed drainage nearby. The switchyard, however, is shown as occupying a parc of the canyon where it could be affeeted by flooding.

There are no reports of ground water develop:ents in the vicinity of the site. It does not appear that the reactor vould affect t'resh-water resources of the area.

Geology The analysis of the geology of the Diablo Canyon Nuclear Plant Site presented in A.E.C. Docket No. 50-275 and supplements was reviewed and compared with the available literature, and the exploratory trenches at the site vea examined on August 1L-15,1965. The analysis appears to be carefully derived and to present e.n adequate appraiial of those aspects of the geology which vculd be pertinent to an engineering evaluation cf the site. ..

e APPENDIX E

,y ,

ri

• S.

u.s. r"oARTMF.NT oF COMMERCE

. ENVIRONMENT AL SCIENCC SERVICES ADMINISTRATION COAST AND GEODCTIC SURVEY -

• • '"#[ ROCKVILLE. MO. 20052 .

=e September el,1967 ina n yau 1o,C23 Mr. Harold L. Price Director of Regulations U. S. Atomic Energy Commission Washington, D. C. 20545

Dear Mr. Price:

In accordance with your request, we are forwarding 20 copies of our report on the seismicity of Diablo Car.-

yon, San Luis Obispo County, Califori.la. The Coast i and Geodetic Survey has reviewed and evaluated the information on the seismicity of the area presented ,

by the applicant in their "Preliminary Safety Anal-ysis Report," and find that it is satisfactory with respect to both distant and nearby earthquakes. We have also included a statement about the tsunami run-up at the site.

If we may be of further assistance tc you, please do not hesitate to contact us.

Sincerely yours, ,

l .

C 'fison,[. '

ear Admiral, USE SA Director 1

Enclosure e

I e

REPORT ON THE SEISXICITY OF THE NUCLEAR PIANT AT THE DIABLO CANYON S7.TE, CALIFORNIA At the request of the Division of Reactor LicensinS of l the Atomic Energy Commission, the Seismology Division of the Coast and Geodetic Survey has evaluated the seismicity of the area around the proposed reactor site in Diablo Can-yon, San Luis Obispo County, California, and has revjewed  ;

i the similar analysis made by the applicant in the Prelimi-nary Safety Analysis Report of the Nuclear Plant at the l Diablo Canyon Site, Pacific Gas and Electric Company. The applicant's seismicity report is complete for both nearby ,

and distant earthquakes which may have affected the proposed i

site. The study of sources of the potential maximum earth- i quakes along the San Andreas Fault, the Nacimiento Fault, [

the Santa Ynez Fault and the San Andreas af tershocks west of ,

this fault includes not only a review of the historical [

earthquakes but a discussion of those parameters related to  :

the carthquake frequency spectrum as fault length, depth of  !

, (

focus, slip and duration of strong shaking. l Based upon the review of the seismic history and re -  ;

i lated earthquake frequency spectrum data and the related  :

, geologic considerations, the Coast and Geodetic Survey agrees with the applicent's statement of 0.20 g at the site nd en r

l

APPENDIX F REPCRI TO AEC REGULATORY STAFF AIAUACY CF DIE SEUERAL CRITERIA FOR THE DIABLO CANYCE SITE NUCIRAR FIMI Pacific Gas and Electric C o w (Docket 50-275) by N. M. Few. ark ani V. J. Hall 9

December, 1967

'l

e AIEAUACY OF THE STRUCWRAL CRITERIA FOR THE DIABLO CAhTON SITE h'JCLEAR PLANT by N. M. Nev:: ark and W. J. Hall Ih'IECIUCTION This report concerns the adequacy of the contaiment structures and cosponents, reactor piping ani reactor internals, for the Diablo Canyon Site Nuclear Plant, for which application for a construction permit ani operating license has been r.ade to the U. S. Atomic Energy Ccc=ission (McKet No. 50-275) by the Pacific Gas and Electric Company. The facility is to be located in San Luis Obispo County, California,12 miles vest southwest of the city of San The Luis Obispo, asd adjacent to the Pacific Ocean ani Diablo Canyon Creek.

site is about 190 miles south of San Francisco and 150 miles northwest of Los Angeles.

Specifically this report is concerned with the evaluation of the design criteria that determine the ability of the contalment system, piping and reactor internals to withstand .a design earthquake acting simultaneously with other applicable loads foming the basis of the design. The facility also is to be designed to withstani a r.aximum earthquake siza11taneously with other This applicable loads to the extent of insuring safe shutdown ami containment.

report is based on inforcation and criteria set forth in the preliminary safety analysis report (PSAR) ani supplements thereto as listed at the eni of this report. We have participated in discussions with the AEC Regulatory Staff and the applicant and its consultants, in which many of the design criteria were discussed in detail.

o

.q.

created by the discontinuity at the base and to provide a gradual transiti:n cf load carryir4 elements tetween the base and the cylindor van. These boa =s do not participate in resisting either uplift due to pressure or shear and tenaien due to earthquake loading; these forces are to be resisted by the diagonal steel

< reinforcing just described. The concrete van in this lover zone is divided into three zones. The inner zone, about 1 ft. thick, conoists of reinforced concrete and is the element Lc which the liner is attached. The middle zone contains the vertical steel I-beans which in turn act as supports for the 16 in.

thick reinforced concrete slab spanning the space between the beams. The ou*Ar zone consists of about 14 in, of concrete in which the dia6onal and hoop reinforcement are embedded. The three zones are provided with bond-breaking material to insurc that the aler.cate vill act separately. The reinforcing steel for the acc, cylindrical vans and boso mat vin te h1 h 6 strength reinforcing conforming to the AS24 A432 specification. The A432 reinforcing tars of size larger than No n are to te spliced v'th Cadveld splices except in cases where accessibility makes velding maniatory.

The liner, as described in Supplement 2, vin bb a minimum of 3/6 in.

thick for the doze and cylinirical vnus and 1/4 in, thick for the base slab.

The anchor studs are to be L shaped and win be fusion velded to the liner plate.

The studs vill be spaced at the corners of a 20 in. square grid, and the design is intended to preclude major affects arising from buckling of the liner.

Personnel and equipment access hatches are provided for access to the containment vessel. In addition there are other penetrations for piping and electrical conduits.

Re facility includes a sea water intake structure located at sea level at the base of the cliff with circulating water conduits and auxiliary ralt vater amiuits leadina to the nuclear plant.

e 5

twice the maxi =um acceleration noted above, nascly 0.kog and 0 30s, but with f

the latter earthquako having a maximum ground velocity corresponding rouchly to a value of 0.40s ground acceleration. The U. S. Coast and Geodetic Survey j report (Bef. 3) concurs in 0.20s and 0.40g values of maximum ground acceleration for design and maxisram conditions, r

4

' Class I piping and equipment, as discussed in Sdpplements 2 and 5, vill be designed for normal loads, (internal pressure, dead load, thermal expansion, etc.) combined with pipe nipture loads and earthquake loading.

The reactor internals are to be designed to resist earthquake '

co=binei vith blev-down loadings and other applicable loadings.

t CCMGhTS ON ADEQUACY OF DESIGN Seistic Design 4

For this facility the contairment design is to be made for two earthquakes corresponding to maximum horisontal ground acchlorations of 0.20g (Earthquake D) and 0.15g (Earthquake 3). For the maEimum earthquake loading the two earthquakes are characterised by horisontal grouni accelerations of twice the values just cited, namely 0.40s and 0 30s. Spectra corresponding to these earthquakes are presented as Figs. 2-11 through 2-14 of the PSAR and-again in Supplement No. 3 beginning on page 22, along with an envelope of the spectra for the no-loss-of function condition (Fig. III. A.12-5, supplement 3).

We concur with the response spectra for the earthquakes when they are used in the following manner.

Since the response spectrum values for Egrthquake D give values that control for high frequencies, and for Earthquake 3, values that control for i

[

t

• 6 that the dynamic analysis to be folloved for the Class I cocponents and structures is the modal participation factor method. Further the modal analysis zny be carried out either through the uso directly of the encothcd tpactra, or employing a time history of 6round motion, employing eerthquake records with emplitude values scaled which lead to essentially the same smoothed spectra. Discursion of this point is presented by the appliccnt in answer to question III. A.13 in Supplement 3 We concur in the use of the modal participation n2thod in the analysis and desi6 n, as well as the use of 3 either the smoothed spectra or the time history input method, provided that the time history input yields the same response spectra as given in the report without any mahr d(vtations telov those cmoothed response spectrum values presented in the PSAR for the envelope of the two earthquakes conoidered. The applicant has advised that the time history input used in its analysis yields ,

substantially the sa:e response spectra as the envelope spectra of the two earthquakes considered.

Vertical acceleration values in all cases vill be tak as two-thirds the corresponding -4~" horizontal ground acceleration, and the effects of horizontal and vertical earthquake loniings vill be ccabined, and considered to act simultaneously. In addition in the elastic analysis, for the containment structure the usual fractional increase in stress for short tem loading vill not be used. We concur in these criteria.

The dm= ping values to be used in the desi6n are given on page 2-29 (revised 7-31-67) of the PSAR and ve concur with the values given therein.

General D sf a B ,.-it , ic fc Cort: ..en We have reviewed the design stress criteria presented on page 5-9 of the PSA'.t and the load factor expressions to be e= ployed in the design and find these reasonable. Parther, ve Lote on page 5-12 of the PsAR that no steel

e o

.g.

Oahold proccco and that icss than 1 percent of the splices vill be inaccccsib'e for Cadveld oplico units, and vill therefore require velding. The proposed approach is acceptable to us.

D e design of the intake structure located at sea level is described in detail in the PSAR and the various supplements. h is vill be designed.as a Class I structure, with due regard for expected tsunami vater heights. Althogh it appears that some protection has been provided against the possibility of rock merses from the cliff falling onto, or into, the pump house, ve recommend that consideration be given to impair =ent of the controle or the pumping system through any possible rock falls or slides.

Crane s_

ne containment crane is listed on page 2-27 (revised 7-31-67) of the PSAR as a Class I structure. We call attention to the design of the cranes to insure that these cranes cannot be displaced from the rails during the design er caximum earthquake, or otherwiso to have damage result from the move::.ent of items supported by them which could cause impairment of the containment or the ability for safe shutdown.

Fenetrations A dircussion of the design of the containment penetrations is given in answer to question III.A.2 of Supplement 1. It is noted there that for the large penetrations the diagonal rebars vill be velded directly to a heavy structural steel ring through use of Ca,1 veld sleeves. D is approach appears satisfactcry to us.

Re applicant further notes in the same section that the strers concen-tration in the vicinity of the opening vill be considered in the analysis. Although this approach may vell be satisfactory, ve believe that the penetration design should take account of any secondary effects arising from local bending, thermal effects, and so on, to insure that the penetration door detail behavea 95-

e o

ne decign critoria and design approach as descrited above are

, acceptable to us.

ne isolation valve desigr. is discussed in several places but particularly in answer to Question II. A.14 of Supplement 1. D e approach outliced there is acceptable to us, ne design of the reactor internals has been reviewed in sone detail with the applicant. De internals are to be designed to withstand the combiced

"*v4-m earthquake spectrum corcerrent vith blev down in such a manner that moderato yielding would not ir. pair the capability of safe shutdown. On the basis of our discussion with the applicant, and the material presented in Supplement 5, the design criteria and design approach proposed for the internals are acceptable to us.

CCNCW SIONS In line vith the design goal of providir4 serviceable structures and components with a reserve in strength and ductility, and on the basis of 4 the information presented, we believe the design criteria outliped for the containment and other Class I co=ponents incimiing the reactor internals, piping, vessels, and supports can provide an adequate margin of safety for seismic resistance.

^

R &ERENCES

1. "Preliminary Safety Analysis Report, Volumes 1 and 2," Naclear Plant, Diablo ,

Canyon Site, Pacific Gas and Electric Company,1967

2. "Pre 11rAney Safety Analysis Report, Suppler 4nts 1, 2, 3, 4, 5, and 6j" '

Racluir Plant, Diablo Canyon om, Pacific Gas and .:lectric Com;&ny, l967 3 "Report on the Ceismicity of the Diablo Canyon Site," U. S. Coast and GeMetic Sevey, Rockville, Maryland, September 21, 1967 e

' < APPENDIX G-1 m am.y aumo. '

igMQN..

/Es., i t UNITED STATES

[ij' hM(h [i DEPARTMENT OF THE INTERIOR 4

.gy FISH AND WILDLIFE SERVICE

" WASHINGTON 25, D. C.

h Mr. Harold L. Price JUN 2 3 B67 Director of Regr.lations U. S. Atcmic Energy Consnission Washington, D. C. 20545 i

Dear Mr. Price:

his is in reply to Mr. Edson O. Case's letter of February 6, .

1967, requesting our cccments on the application of the Pacific Gas and Electric Ccepany for license for its proposed Diablo Canycn Nuclear Pover Plant, San Luis Obispo County, California, Docket No. 50-275 he project would bc located adjccent to the Pacific Ocean on the California coast about 12 miles vest southwest of San Inis Obispo, California. A pressurized water reactor, designed for an initial output of 3,250 thezval meEavatts, and an approximate net electrical output of 1,060 megavatts vculd be used as a pover source.

4 A radioactive vaste disposal system and other facilities required for a cceplete and operable nuclear power plant would be provided. Gasecus vastes vould be vented to the atmosphere, and liquid vastes vould be discharged to the Pacific Ocean via the condenser cooling water. Condenser cooling water would be 4 punped trea the Pacific Oc'ean through traveling vater screens and returned to the ocean at the water's edge in Diablo Cove..

Detailed design criteria fcr the cooling water system have not been finalized at this time.

he applicant plans to initiate a radiclogical survey of the ,

arca about two years before plant operation. Details of this program have not been ccepleted, but spceial emphasis vill be placed on the edible marine species which are present in Diablo Cove.

We applicant has contracted for ecological studies of the marine environment, and the Resources Agency of California has made similar irvastigations. he results indicate that existing flora and fauna ase a highly diversified mixture of varuvatar and cold-vater forms. R ese studies vill be ccatinued in order to adequately describe seasonal changes. Special emphasis will be placed on seaveed, which appears to be the chief source of organic activity.

3

-99 $

o o

^  ;

1. Cooperate with the Fich and Wildlife Service, the ,

Federal Water Pol.lution Control Adrinistration, The Desourus AGen./ of Caliiocnia, and other '

interested state a6encies in developing plans for radiological surveys.

2. Conduct or arranEe for the conduct of preoperational

• , radiological surveys of selected organisns that concentrate and store radioactive isotopea, and of the environnent including water and sediaent sanples. These surveys should be conducted by scientists knowledgeable in the fish and vildlife field.

3 Prepare a report of the pre-operational radiological survey and provide five copies to the Secretary of the Interior for evaluat:lon prior to project operation.

4. Conduct radiolecical curveys, alsilar to those * -

specified in reccenendation 2 c.bove, analyze the '

data, and prepare end submit reports every three  ;

nonths during the first year of reactor operation  !

and every six ncnths thereafter or until it has been

' conclusively deconstrated that no significant tdverse conditions exist. Submit, tive copies of these reports to the Secretary of the Interior for distribution to the appropriate State asi Federal agencies for evaluatics.

5 Beduce the discharge of radioactive vastes to acceptable levels, if the post-operational surveys ,

establish that %e release of radioactive effluent at levels permitted under Title 10, Part 20, Ccde of Federal Regulations, results in harmful concentra-tions of radioactivity in fish and v11dlife. ..

We understand that it is the Cosnission's opinion that its regul,atory .

authority over nuclear power plants involves only those hazards associated with radioactive materials. .We have reccomended in past rpplicaticrc that before the permit 16 issued, ther al pollution and other detr,",n:. ital effects to fish and vildlife VM.:b may result from ,

plant cons',ruction and operation be called to the attention o applicant.

compar * .s sisaw t,n acreement with the these proble.- n " wo cc: ..ct sL les which would identify I

Resources Age.,cy oi Califo m hamful effects resultimg free other than radiological cases, and to However, we reccesend that the mitigate losses if they occur.following observations be brought to the a t

I

-101-  !

e

• ' ,, *. N O -

1 .$.

1. Cooperate with the Fish and Wildlife Service and the Eccources Acency of California and other interested state agencies in developin6 plans for ecological surveys; initiate these surveys at least two years before reactor operation; sad continue them on a regular basis during operation or until it has been conclusively demonstrated that no significant adverse conditions exist.

2. F.eet with personnel of the Fish and Wildlife Service and the Resources Agency of California at frequent intervals to discuss new plans and to evaluate results of cxisting surveys.

3 F. eke such modifications in project structure and operation as ray be deternined ricessary as a result of the surveys.

4. Provide a screening device at the cooling water intake, the specifications of which would be deter-cined in cooperaticn with the Fish and Wildlife Service and the Resources Acency of California.

5 Provide cea.pensatica for any losses of fish and vildlife that may occur as a result of construction crcperation of the project.

the opportunity for presenting our vim on this project is appreciated.

Sincere,'y yours, G. ./o / . .

C QQ OM Aethg omiss o er 4

-103-N

0 4

, o . , . ,

APPEliDIX 0-2 m amy ,oca to,

. ? , . - *. #

UNITED STATES  ;

f DEPARTMENT OF THE INTERIOR W g>y r

-.ey? j edf, FISH AND WILDLIFE SERVICE ,

y . ' WASHINGTON, D.C. 20240

'l Mr. Harold L. Price N 3 * }963 Director of' Regulations 1 U. S. Atee.ic Energy Co mission )

Washington, D. C. 20$5

Dear Mr. Price:

' This is in response to Mr. Boyd's letters of July 31, and August 11, 1967, transmitting Anend=ents lios.1, 2, and 3 to the Preliminary Safety Analysis Report of the application by the Pacific Gas and Electric >

Company for a construction per: nit for the proposed Diablo Canyon fluelear l

Pover plant, San Luis Obispo County, California, Docket No. 50-275.

The Service has reviewed k.endments Nos.1, 2, and 3 and has the following '

co=ents on the proposed envirorcental monitoring program contained' in

/cendment lio.1 and on the cooling vater intoke structure contained in Anend ent No. 3:

' The proposed envirorcental monitoring program generally conforms

• to the recesendations in our June 23 letter to the Cocaission; however, the applicant's proposal does not include collection and examination of water and sedinent samples which we reco= ended and consider a necessary part of the conitoring program. Fish and wildlife resources vould be insured of r. ore adequate protec-tion from radiological hazards with implementation of the-nonitoring program along with other proposed safeguards, provided that water and sedinent sa ples are included in the prograss and that initial liquid vaste discharge limits are adjusted if

' co.sidered necessary.

i Specifications for the cooling water intake structure contained

' in A endment No. 3 do not include adequate information for an analysis of possible effects on fish and vildlife. The bar racks and traveling screens designed primarily to keep debris out of the system could serve the additional purpose of minimi- '

sing harm to fish and wildlife provided that intake velocities vere not limiting. Being cognizant of the Coesission's opinion that its regulatory authority does not apply to other than i radiological hazards, we to tomend that the Comission urge ,

' the Pacific Gas and Electric Company to consult with the Fish and Wildlife Service and the Resources Agency of California

' in estabidshing final design criteria for the intake structure.

l 105-a

~ 33