ML20235C892

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Div of Reactor Licensing Rept to ACRS in Matter of Comm Ed CP Application for Quad Cities Units 1 & 2
ML20235C892
Person / Time
Site: Quad Cities, 05000000
Issue date: 11/21/1966
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US ATOMIC ENERGY COMMISSION (AEC)
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NOV 21 1966 s

U. S. ATOMIC ENERGY COMMISSION D_IVISION OF REACTOR LICENSING _

REPORT TO ADVISORY COMMITTEE AFECUARDS_

ON REACTOR S IN THE MATT 5R OF COMMONWEALTH EDISON COMPANY CONSTRUCTION PERMIT APPLICATION -

-CITIES UNITS 1 AND 2 FOR QUAD  !

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h 'hh ldi Note by the Director of the Divisioneactor of R Licensing The attached report has been prepared b consideration by the ACRS at its December meeting.

1966y the Division of React

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TABLE OF CONTENTS Page r!

I. Introduction......................................................... 1 II. New. Plant Features and Features Unique to a Dual Reactor System .. . .. 2 A. Spent Fuel Poo1s.............................................. 2 B. Physics Calculations.......................................... 2 C. Powdex System................................................. 3 D. Reactor Shutdown Cooling System............................... 3 E. Reactor Core Isolation Cooling (RCIC) System.................. 4 F. Interconnections of Units 1 and 2............................. 7

1. , containment Interconnections........................... 8
2. Electrical System...................................... 8 III. Site....................... .......................................... 10 A. Description................................................... 10 B. Meteorology......... ... ..................................... 11 C. Geology and Fydro1ogy......................................... 12 D. Seismology.................................................... 13 E. Environmental Monitoring...................................... 13 F. Off-site Hazards.............................................. 14 IV. En g i n e e re d S a f e gu a rd s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 A. Emergency Core Cooling Concepts............................... 14 B. Analysis of Individual Core Cooling Sys tems . . . . . . . . . . . . . . . . . . . 17
1. High Pressure Coolant Injection (HPCI) System.......... 18
2. Core Spray..................................... ....... 18 0FFECHAL USE6tY ,

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3. Low Pressure Coolant Injection (LPCI) Sys tem. . . . . . . . . . . . 18
4. Automatic Depressurization............................. 20
5. Unlimited Feedwater Supp1y............................. 20
6. Alternate Water Sources for Engineered Safeguards. . . . . . 21 C. Containment Cooling........................................... 21 D. Prevention of Rod Ejection.................................... 22 E.. Conformance of- the Engineered Safeguard Systems to the Consnis sion 's P ropos ed Crite ria. . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 V. Containment.......................................................... 27 A. Steam Line Isolation Valve Testing............................. 27 B. S e c on d a ry Co n t ainme n t Te s t in g . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 27 VI. Instrumentation...................................................... 28 VII. Primary System Surveillance.................'......................... 31 A. S umma ry o f P r o p o s a l s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
3. Hydrostatic P roof Tes ts During Ves sel Life. . . . . . . . . . . . . . . . . . . . 32 VIII.. Turbine Orientation.............. ..................................... 32 IX. Accident Analysis.................................................... 34 X. Review of Previous ACRS Concerns..................................... 36 XI. Items Requiring Continuing Review..................................... 38 XII. ' Conclusion........................................................... 38 i

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4FFECHAbUSE ONL I. Introduction The Commonwealth Edison Company has submitted an application dated May- 31, 1966, for a construction permit and f acility license for a single cycle boiling water reactor of 2255 MWt, supplied by General Electric, to be Unit I at its' Quad-Cities site near Cordova, Illinois. The application was amended to include an identical Unit 2 by Amendment No. 1, dated August 18, 1966. This report is an evaluation of both proposed units.

Additional information, in response to s taf f questions, was provided by the applicant in Amendment Nos. 2 and 3, dated September 9,1966, and October 18, 1966, respectively. Both amendments dealt with turbine f ailure, engineered safeguards, liquid waste disposal, and primary system integrity. The topics covered in Amend-ment No. 3 included elaboration of certain areas in which the ACRS subcommittee expressed further interest at their meeting held Septenber 16, 1966.

The subcommittee meeting of September 16, 1966, included a visit to the pro-posed site. An additicnal subcommittee meeting was held November 17, 19M , at _j which further documentation was requested in the areas of turbine rotor failure, i suppression pool corrosion, and Jeak detection criteria and ir.strumentation.

The proposed reactors are substantially similar to the Dresden Units 2 and 3 for which provisional construction permits have been issued to the applicant and, except for size, to the Millstone Point reactor.

(Dockets 50-237/249 and 50-245.)

Except for the description and evaluation of the proposed site, this report is essentially an extension of previous analyses performed on the Dresden and Millstone reactors. Systems which have been previously evaluated by the staff and ACRS are not discussed except where .nore information has been developed as the result of staf f or ACRS interest. New systems, such as the non-regenerative (Powdex) l GFFHCHAtUSE ONLY m.

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demineralized in the waste disposal system, and systems for which the staf f had not previously submitted a written evaluation, such as the Low Pressure Coolant Injec-tion (LPCI) system, are also discussed in this report.

Similarly, the Commission's proposed criteria, dated November 22, 1965, have been discussed only in those areas in which new information was available. At the end of each discussion topic, a brief statement of how the applicable criteria have been satisfied is included.

II. New Plant Features and Features Unique to a Dual Reactor System A. Spent Fuel Pools The spent fuel pools of Quad-Cities reactors Units 1 and 2 will be adjacent and will be connected by a transfer canal, a design feature not included in Dresden Units 2 and 3. The interconnected fuel pools will permit the transfer of refueling tools between the storage pools and will facilitate improved fuel cycles involving the alternate irradiation of fuel elements in the two corce. It is our understand-irg, he. cver, that fresh fuel will be used for the initial startups for b th reactors. We believe that conformance of the fuel storage facilities to No. 25 of the Commission's proposed criteria which requires that adequate shielding and cool-ing for spent fuel will be supplied is not affected by the interconnection.

(Section X-1.0; Amend. 3, p. 1-3.)

B. Physics Calculations Revised reactor physics calculations have ind'.cated smaller values for excess reactivity of the clean core as opposed to Dresden Unit 3 (0.22 vs. 0.26 d k).

The control rod and borated control curtain worths also were reduced (0.17 vs. 0.18 d k and 0.09 vs. 0.12 d k, respectively. Since there has been no physical change in the fuel or control components, no significant change in b k/k values results f rom the change in reactivity. The shutdown margin with the strongest rod out OFMCHAMSE-ONLY- -

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$FRGAt-USE~0 INLY remains ' unchanged at 1% A k/k, and we believe that the change in the reactivity balance presents no decrease in plant safety. (Amend. 2, p G-4-1.)

C. Powdex System The Quad-Cities liquid radioactive waste disposal system will utilize Powdex non-regenerative demineralized units which, af ter approximately 17 days of operation, will be back-flushed and the resia disposed as solid waste. The Powdex units will increase the cleanup capability of the system and discharge into.the river will be i I

reduced by about a f actor of 20 over the regenerative system. The resin regenera- l tion process of previously proposed f acilities resulted in liquid chemical wastes

= of a high specific activity which increased the comp 1exity and hazard of waste  ;

disposal. Owing to the reduced liquid waste activity, the holdup tanks will now meet the Class II earthquake standards instead of Class I. l The units are deemed to satisfy No. 24 of the Comission's proposed criteria by significantly decreasing the amount of radioactivity which might be accidentally released even though the probability of such a release by seismic activity is increased. Furthermore, the applicant has stated, and we agree, that even in the event of inadvertent release of liquid waste, the concentration of radionuclides in the discharge ca.nal would not exceed 10-7 microcuries/cm . (Section VII-3.0; Amend. 2, p. H-2-1.)

D. Reactor Shutdown Cooling System The shutdown cooling system circulates water from the primary system, under low pressure conditions, through a heat exchanger and back to the primary system to remove decay heat. The heat exchanger used is either of the LPCI-containment spray heat exchangers. The shutdown cooling system has only one intake f rom the recircu- )

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lation lines, but primary water can be cooled and returned through both of the LPCI 1

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heat exchanger loops. The rating of each heat exchanger is about 0.8% of full power and either loop is thus sufficient in capacity to provide shutdown cooling about two hours af ter the reactor has been scrammed from full power. Sharing of

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system components with an engineered safeguard system requires that the normal and.

emergency functions be isolable. We believe that with proper attention to the sub-

' ject 'of single valve f ailure, the isolation of function can be assured and sharing of components of this system with an engineered safeguard system is acceptable.

This system, in conjunction with other cooling systems' described in this report, satisfies No. 10 of the Commission's proposed criteria which states that heat removal systems shall be provided to cover all anticipated abnormal and accident conditions. (Section'X-3.0; Amend. 2, pp D-5-1, G-3-1.)

E. Reactor Core Isolation Cooling (RCIC) System For the Quad-Cities reactors, the Isolation Condenser system has been replaced by a Reactor Core Isolation Cooling -(RCIC) system utilizing a turbine-pump unit.

This system is similar to, but of a smaller capacity than, the High Pressure Coolant 1 Injection (HPCI) systen preposed for beth 0:ad-Cities and Dresden reactors and described in Section IV of this report. The primary purpose of the RCIC and Isols-tien Condenser systems is to supply cooling in the event of a turbine trip ,

accompanied by loss of a.c. power which would result in a loss of feedwater to the primary system. (Section X-4.0; Amend. 2, pp. F-1-1 to 7-6-1.) I The RCIC system consiets of a steam turbine requiring only enough steam to drive a centrifugal pump to supply about 350 gpm of makelp water to the reactor vessel. Excess steam generated by decay heat is released through the primary system

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EFHCHAL4JSE-ONHA-Fo11owing' isolation of the reactor from the normal heat sinks, the primary.

system pressure increas'es, and a reactor scram is initiated at' approximately 1050

. psi. ..At approximately 1080 psi, the pressure relief valves open and relieve to the suppression pool.- Normally, coolant makeup will be supplied. from' the condensate storage tank by the feedwater pumps.

-In the event that feedwater is unavailable, either through depletion of the condensate supply or loss of a.c. power, the steam supply line to the RCIC system will be automatically opened on low reactor water level (by a d.c. operated valve) permitting. the passage of steam through the RCIC turbine- and into the suppression pool. Pumps driven by. the RCIC turbine will supply makeup water from the condensate storage tank. Sufficient-condensate will be available at a minimum to permit con-tinuous operation of the RCIC system for more than eight hours. In the event that the condensate source is unavailable, water frem the suppression pool can be used as an alternate scurce.

Opera:icn of the FCIC cpten .311*. cause a slow heating of the suppression pool.

In response to a question by the staff, GE stated that the suppression pool tempera-ture will not be allowed to exceed 140*F to provide assurance that the contain-ment capability to suppress steam released in the maximum credible accident is not l- reduced. One unit of the suppression pool cooling system will be capable of raain-l taining the pool temperature below 140*F. The cooling system is operable from the station diesels.

In addition to the core cooling functions that were formerly performed by the isolation condenser, the RCIC system will be capable of supplying makeup for small leaks in the event that feedwater is unavailable.

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Testing of the RCIC system will be possible at any time the primary system is pressurized and steam is available. Makeup water under test conditions will be lq i

directed through a bypass line into the main condenser. Formerly, the isolation j 1

condenser could only be tested when the plant was shut down.

Although the RCIC system is the substitution of an active shutdown system for a passive one, se seliew that based on our review, on the demonstrated operability of this system at the time of installation and on a comprehensive pro-gram of regularly scheduled tests when the facility is operational, the RCIC system j as proposed will provide emergency cooling equivalent in reliability to those systems proposed for previous f acilities.

With reference to fission product leakage resulting from RCIC operation, 1' General Electric supplied the following information by letter dated April 15, 1966, in response to staff questions which were asked in conjunction with review of a General Electric " company private" document on the RCIC system submitted in Oc:ober 1965.

(1) During a two-hour pressure reduction to 150 psi, the amount of primary sys tem water that would be released to the containment would be about 50,000 gallons. If no heat exchangers functioned to cool the suppression pool, the pool and air temperature would be about 170'F and the containment pressure about 7 psig. The applicant calculates, and the staff agrees, that even if a primary containment leakage rate of 0.4% per day is assumed, the environmental doses would be negligible in the case of high gross fission product activity (50 oc/cc) in the reactor water.

(2) In the case of eight hours of use at full pressure, the expected amount of activity released to the containment system with a gross fission product of 50 uc/cc in the reactor water would be approximately 16 curies to the suppression pool. Of 4FRGAL4JSE-ONLY -+&,b x .

,J M GAL 4JSE4 Nb"t the 16 curies gross fission product in the pool water, not more than 0.5 curie could be expected to be released to the containment air. Operation of the suppression pool heat exchangers will insure no increase in pool temperature or containment pressure.

From the above, and on the basis of independent calculations, the staff con-cludes that negligible amounts of fission product leakage would be expected during operation of this system. The potential exists for accumulation of fission products in the suppression pool water, however, and high radioactivity levels in the torus could prevent personnel entry for maintenance and inspection. The applicant has agreed to provide means (requiring manual hookup) to process suppression pool water.

We believe that the proposed RCIC system (which can be backed up by the higher capacity High Pressure Coolant Injection system, if necessary) will provide pro-tection at least equal to that of the two isolation condensers found on previous plants. This system, in conjunction with other core cooling systems described in this report, satisfies No. 10 cf the Conmission's proposed criteria which states that systems must be provided which are capable of remcving core decay heat under all anticipated abnormal and accident conditions.

F. Interconnections of Units 1 and 2.

As discussed in the answer to question No. 1 of Amendment No. 3 to the appli-cation, a number of systems or f acilities will be shared by Units 1 and 2. Of these features, only the interconnection of the spent fuel storage pools is unique to the Quad-Cities station. The most important interconnections are (1) the secondary con-tainment, (2) the standby gas treatment system, and (3) the electrical system.

(Section I-8.0; Amend. 3, pp 1-1.)

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1. Containment Interconnections The secondary containment and standby gas treatment system combine to form the engineered safeguard which would provide the necessary airflow to maintain the  !

1 combined secondary containments of Units 2 and 3 at a slight negative pressure and which would filter and release the secondary containment atmosphere through a 310-foot s tack. Thus, a major fraction of fission products which might be released to the reactor building in the event of an accident would be trapped in the filters, and that fraction which would pass through the filters would be released to the environment from the stack. The off-site doses as a result of the design basis accident are based on release of the secondary containment atmosphere through the gas treatment system and stack at a rate of 100% of the building's free volume per day. The difference from a single unit is thet the standby gas treatment system must be sized to evacuate a larger volume. Test procedure for the system will con-sist of inserting a sin.ulated high radiation signal and measuring a previously calibratad pressure drop in the exhaust duct. This tes: procedure is further discussed in Section V of this report.

2. Electrical System The outputs of the Quad-Cities Units 1 and 2 are delivered to the 345 KV switchyard which feeds four 345 KV transmission lines and the plant auxiliary power transformers. One of these lines runs in a westerly direction to the Hills Substation. Two lines extend to the east on separate rights-of-way, and connect to a substation near Rock Falls. The fourth line runs to Barstow and shares a common set of towers with the line to Rock Falls to a point about one and one-half miles east of the station.

Part of the normal auxiliary power supply for each of the Quad-Cities units is provided through a transformer connected to its generator leads. Startup power and OFRGAL USE-ONLY V4 - i

.JFEHGArUSE ONiul the rest of the normal auxiliary power supply are through transformers supplied from the 345 KV switchyard. The applicant has stated as a criterion: "The trans-formers are connected to the 345 bus so that outage of any line will not cause an interruption to the transformers. Modern high speed relaying is provided on the 345 KV lines."

At each unit, the auxiliary busses are connected by appropriate switching to the standby emergency diesel generators. There are three diesel generators sized such that any two can supply those engineered safeguards which would be required after a coolant-loss accident in one unit and vital normal shutdown systems for the second unit.

A general design requirement is that duplicate services will be supplied from different auxiliary busses. Thus, failure (e.g., short to ground) of any one bus will not preclude the operation of adequate safeguards devices from the remaining bus.

All r rotective circui t breshers are sized according to standard electrical industry practice where maximum current interrupting capabilities of the circuit breakers exceed the available line to short circuit taking into account the impedances of the generator, transformers and other electrical system components, l Circuit breakers are electrically operated from the 125 volt d.c. system (station bat t e ry) .

i The staff believes that adequate redundancy and isolaticu exists between the i

345 KV lines which, respectively, run in the easterly and westerly directions. The staff also believes that the applicant's criteria regarding the redundancy of emergency diesel generators, the distribution of duplicate safeguacds loads on separate busses, and the capability for circuit isolation in the event of f aults are acceptable.

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JFRECHAL-USE-ONLT III. Site A. Description The site of the proposed Quad Cities plant consists of 488 acres in Rock Island County, Illinois, on the east bank of the Mississippi River opposite the mouth of the Wapsipinicon River, approximately 3 miles nor-h of Cordove,

' Illinois. It is about 20 miles northeast of the Quad Cities emas (Davenport, Iowe; Rock Island, Moline end East Moline, Illinois).

From the reactor building and from the plant steck the distance to the nearest site boundary is approximately 1190 feet to the north. The distence to the nearest residence is 1550 feet,to the north. There are approximately 15 river front houses within one mile of the reactorss Ihe nearest town is Cordova (1960 population 502), approximately 3 miles south of -he site. The f nearest city is Clinton, Iowe (1960 population 33,589), to which dister.ces are approximately 5 miles to the ed6t and lo miles to the center. 2e total popule- ,

tion within 5 miles in 1965 was 5369, end this is expected te :ncreece to sp- I l

proximately 10,500 by 1980. We feel that the popule cion center distence may be teken as 5 miles and the low population distence es about h miles. The table below gives the total population et various distences:

Distence Miles 1965 Populcticn 0-l @

1-2 75 2-3 226 3h 1,oh9 h-5 3,959 1960 Population 1950 Population 0-5 5,089 10,500 ,

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B.' . Meteorology The:epplicant has presented data on the climatology of.the site from Moline, Illinois, .with supplementary data from a number of other' cities in the -

state. From this information, it is concluded that: typical temperate con-tinental climatology exists at this site, and that no unusual espects of diffusion will be found.when' detailed date is taken at the site and analyzed.

For the present, the applicent has assumed the standard GE diffusion. model, based on' en assumption of sigma-theta times wind speed of 0.l' radian-meter /second ]

for stable conditions, and . conventional diffusion coefficients for the Sutton equation during unstable conditions. We feel that the amount of diffusion predicted in this menner is reasonable, since it is the seme as aesumed for most other continental sites. ' The Weather Bureau has also stated that they' expect the diffusion at this site to be similer to that at most others in the i 4

k; central portion of the country.

l The maximu:n sustsined wind speed thet 'hes been etserved is 67 mph, with un-officie3 estimates of gusts of 110 mph. Major structures vill be designed for l

susteined vinds of 100 mph. In the past 52 years, there have been 8 tornadoes f i

in Rock Island County, and 10 in the adjacent Henry County. ]

With reference to tornado protection, components which are required for safe shutdown of the plant are located either under the protection of re-inforced concrete or are located underground. Seismic design considerations i bring the strength of the concrete reactor building ve11s to a level such that

~-they can. resist winds of h00 mph within normal alloveble stresses, and would not feil until e speed of approximately 600 mph is reached. In addition, we 1

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believe-that sufficient-independence of.v' ital systems has been provided to-insure that s' safe shutdown of the plant can be accomplished even in the' event of's tornado-generated missile penetrating the secondary containment. (Amend 2,-

p B-1-1; Amend 3, P 4-1).

C. Geology and Hydrology All major structures vill be supported on dolomite bedrock, which in most places is no deeper than the foundations themselves. The information submitted bylthe applicant indicates'that the geologic and hydrologic conditions at the site present no unusual' design or construction problems for this facility. There i

is no evidence of major faulting in the area, and major tectonic deformation'has n'ot occurred in some 60 million years.

The ground water grsdient in the vicinity of the site is relatively flat, end slepes generally.toward the. Mississippi River. The finished grade elevation at the site vill be 59,.5 1 foet, while the surrounding land elevetien is ap-proximately 605 feet. , Tr.e normal poci elevation of the river, when flow is being controlled by the dams, is 572 feet, . and' the maximum flood elevation during 92 years of record was 586 feet. The 1000 year flood level is estimated to be 589.2 feet, which is still some 5 feet below site grade. It is concluded that flooding of the site is not a problem..

At the request of the staff, the USGS has reviewed the geological and hydrological aspects of the site. We have not received the USGS corrents. How-ever, based on discussions with them we believe their conclusions support the

- views of the staff.

All of the Quad Cities take their water supply fro- the Mississippi, beginning at a point approximately 15 miles below the site. The applicant has considered GFM6 hat-USE-ONLY u..n,.

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'. dilution in .the river, flow times, flow rates, intake rete, and storage capacity in an effort to assess the consequences of en accidental release of liquid effluent. It was found that the storage capacity. is great enough to permit a citda torn off its water intake for the. time during which the' concentration is more than half the peak,.even for'the most adverse conditions of flow and consumption. During the remaining time, the maximum potential concentration would be at about Part:20 limits assuming vaste inventories in the storage tanks associated with failed fuel conditions in.the-reactor. (Amend 3, pp 2-1 5-3 , .

6-1).

D. ' Seismology The. site is located in an eres of low seismicity. The applicant has proposed e design earthquake resulting in a maximum ground acceleration at the site of 0.12 g. In addition, for a ground acceleration of 0.24 g the plant will be designed such that there vill be no impairment of function of critical structures end components, and a safe end -orderly shutdown vill te assured. Eased.cn our discussion of the seismicity aspects with the USC&GS, vhich has reviewed the seismicity of the site, ve believe these criteria are adequate.

E. Environmental Monitoring It is stated that a study of environmental radiation levels will begin approximately 2 years before the plant is started up. The plans have not been formulated, but the program vill include air samples on site, veter samples from the river and nearby wells, and samples of soil, vegetation, and milk from the surrounding eree. It is anticipated that the Fish and Wildlife Service i

vill recommend monitoring of fish, and this vill be required by the staff at the 1

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appropriate time. With this addition, we feel the scope of the program is adequate.

F. Off-site Hazards

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The applicant has stated, as a result of a study of adjacent chemical l

plants, that any accident, such as an explosion or fire, at the adjacent plants would not affect the functional capability of the primary containment, reactor and turbine buildings or radioactive vaste control facility. In addition, the accidental dumping of a large quantity of hcl was postulated to occur upstream of the station and was stated to have a negligible detrimental effect on the system since the diluted acid would not be present over an extended timo period.

Based on the above, the staff believes that no credible mechanism exists by which (1) en accident could be caused or (2) the contal ment or shutdown systems could be impaired as a result of accidents et adjacent industrial plants.

(Amend 2, p A 4-1).

IV. Engineered Safeguards A. Ihergency Core Cooling Concepts In case of inventory loss from the primary system, e minimum of two heat removal systems of independent and different principles are provided to dissipate core decay heat in order to prevent fuel clad melting. The abnormal conditions considered ere the loss of auxiliary AC power and/or the loss of coolant due to a primary system rupture up to and including a double-ended severance of the largest primary system pipe resulting in the automatic scram of the reactor and closure of the primary system isolation valves. (Amend 2, p G-5-1).

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. Failure 'of auxiliary AC power negates.the'use.of the' normal heat removall by-the main steam condenser in the primary coolant-system.';Under'this' con-

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'dition, core 'ooling.is c accomplished by the RCIC-(Reactor Core Isolation Cooling j l . system). , ' The operation of this system.was discussed ' in Section II.E.

During .the loss of both AC power and ' coolant, four different systems

operate in various combinations to provide' sufficient coolant makeup
to 'ac-commodate the entire spectrum of. primary system rupture sizes. Decay' heat from 1

the reactor core"is removed by the steam or water flowing from the reactor, 1

- through the rupture, 'into the containment vessel and finally entering the. pressure-suppression pool.  !

If the size ~of the rupture is not large enough to provide a rapid depressuri-

':l zation of the reactor vessel, the primary system relief valves can be intermittently..

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operated with the High Pressure Coolant Injection (HPCI) systeni providing make-up a

coolant. Like the RCIC system,.the HPCI system consists of a steam turbine driving s l centrifugal pmp, associated N operated velves, and piping, t~ne HPCI-turbine is driven by steam extracted- from the' steam line of the isolated reactor, and it exhausts this steam to the pressure suppression pool where it is con--

densed. In order to maintain the proper veter-level in the reactor vessel, the pump supplies make-up water initially from the condensate storage' tank and later, as-required, from the pressure suppression peol.

~ Failure of the HPCI system to' function properly causes its overlapping backup system to actuate. This independent backup system is an automatic depressurizing system that senses loss of flow from the HPCI, low water level and high drywell pressure, and actuates the primary system relief valves to relieve the primary system pressure. The released steam is condensed in the l I

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' pressure suppression pool. Following.this depressuriza-ion, either of the

" low-pressure, high-capacity" systems can prevent clad melting. The two in-dependent low pressure systems are' the Core Spray systen and the Lov Pressure.

Coolant' Injection-(LPCI) system..

The HPCI previously described . is a "high-pressure, low-capacity" system

'that provides water for small primary system breaks. If the rupture is too i

large for. the make-up capacity, of the- HPCI, then the rupture is also large enough to depressurize the primary system so that the " low-pressure, high-capacity" systems can then provide sufficient make-up coolant.

The Core Spray system consists of two loops with each loop having its own 100% capacity pump which takes water from the suppression pool and delivers it to a spray 'sparger within the core shroud. The coolant is distributed on the top of the core. This system does not' depend on maintaining a minimum veter level

' in the reactor core, but is sized to distribute enough veter to cool the core even if it is entirely uncovered by the loss-of-coolan: condition / Supportir.6 experimental evidence and core heating calculations were provided in Amendment No. 5 tb the Dresden Unit 3 application. This system tas an overlapping tackup system in the form of the LPCI system.

The LPCI system works on a different principle. It vill reflood the core to the top of the/ jet pumps (2/3 of the core height). he applicant has stated that experiments'show that flooding a single fuel bt.ndle to 1/2 its length will provide sufficient cooling to prevent clad melting. (Iresden Unit 3, Amendment 5).

The LPCI system consirts of two loops with each loop haring two pumps. Three of the four pumps are required for full system flow. This system also draws veter from the suppression pool. The pumps in these two "lov-pressure, high-capacity" cOFRGHAL USE-ONLY7 .

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.JEBOAEUSE-ONU systems, es well es the service-veter pump used for the removal of heat from the LPCI heet exchangers can be povered by the station standby diesel generators.

As discussed in Section IV.C. the pumps are shared with the containment spray system.

'If the loss-of-ecolant condition occurs without loss of AC power, any of the systems used for the previous condition can be used to provide core coolinE and, in addition, the feedvater supply system and control rod drive coolant supply can provide make-up veter. The control rod drive coolant supply system has little capacity and therefore can only accommodate very small leaks. The feedwater supply system does not have enough cepecity to u: eke up veter for large primary system ruptures that are in the veter portion of the primary system; however, it can accommodate all steam line ruptures. Se feedvater pumps require AC power to operate and drev their veter from the main steam condenser hotvell.

Service veter can be pumped into the main steam condenser to provide a continuous supply of teke-up va er.

B. Analysis of Individual Core Cooling Systems The ability of the Core Spray and HPCI systems to cover the range of primary break sizes was discussed in Peport No. 2 to the Committee on the Dresden Unit 3 application. The ability of the engineered safeguards complex to cover the entire spectrum of primary system breaks is illustrated in a tar graph in Amendment No. 5 to the Dresden 3 application (p 14 of the subsection dated August lo, 1966).

Experimental evidence for the effectiveness of the core spray and also for the flooding concept ves discussed in the above submittal. Staff reservations and further analyses of the systems are presented below:

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1. : HPCI System (Section VI-9 0; Amend. 3, p 13-35) a .

Accidental' initiation'of the High Pressure. Coolant Injec. tion system could: inject colducondensett into the reactor.. This incident has been snelyzed

. -ry the applicent and foun'd to. result in additional'aubcooling and would possibly .

result in a reactor scram on overpower, but thermal limits vould not be exceeded.

The staff believes that this is a reasonable essessment of this accident.

. l The staff has' the same reservations on this system as stated in th eDresden Unit 3 Report No.,2; namely, (1) redundancy in critical valves and instruments-

. tion should be provided,. and (2) the condensate storage tank should be made a l

Class 1 structure from a seismic design standpoint. We understand that the c o-i densate storege' tank ' vill.be made a Class 1 structure. We believe that these questions should be resolved during the development of the final design.

2. Core Sprcy (Section VI 6.0; Amend ' 3, p 13-7) i

, A core sprey complex has been proposed for this syst/,n that has less 'J redundancy than that proposed to the Committee on Dresden Unit 3 Subsequent to the initial Dresden Unit 3 proposel, (2 loops, each with 200% capacity), a core flooding system ves proposed and General Electric stated that the need for the extra capacity of the core spray systems ves being r'eviewed. As a con-sequence the Quad-Cities proposal is for two loops, each with 100% capacity.

Ve believe that this is adequate when combined with the backup capability of the Low Pressure Coolant Injection (flooding) system.

3 LPCI System (Section VI-7.0; Amend 3, pp 13-1,13-32)

The lov Pressure Coolant Injection system is a hi8h cepecity flooding system which enters the reactor vessel through the recirculation lines and jet

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QFMGHAL41SE'OfE pumps, It requires the operation of several valves including the closure of one of two isolation valves to prevent backflow throu6h the recirculation lines.

The design of this system has not been completed but the method of entry throu6h the recirculation lines has been found preferable by the applicent to J

entering through the top or bottom vessel heads. The chief disadvantage of the bottom entry ves the mechanical forces exerted on the internels. The chief dis-advantage of entry through the top head was said to be the piping which would have to be removed each time the vessel head was removed.

The staff has several reservations on this system: ,

(1) A cross-tie between the two sides of the system essures that the i required three of four pumps can deliver veter to eitler recirculation loop. We believe that isolation velves should be provided in this cross-tie to assure that any component failure would te isoleble end would not prevent et least partial functioning of the system. We understend that this modification vill be made.

(2) If a break of a recirculation line were to occur, the single fir.el valve on the unaffected side must open. We belie ~e trat an analysis should be performed before the desi6n is completed to essess the desirability of pro-viding redundant volving in place of each normally clesed velve on each line.

(3) Detailed bases end analyses for minimum core flooding requirements (including make-up requirements for leakege and boil-eff) should be provided.

This information is now being developed. We believe that these questions should be resolved before the final design is fixed.

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k.. Aut'omatic'Depressurization (Section IV-3 0)

Automatic depressurization was proposed.by General Electric as a back--

up to the High Pressure Coolant Injection. System. It vould be actuated on simultaneous signals of high drywell pressure, low reactor water level and lack

of HPCI flow. As presently envisioned, e delay would te built in to allow the operator to negate a spurious actuation. After the lO-ninute depressurization, the core spray or flooding systems vould be relied on to keep the core cooled.

Because of the requirement for a high drywell pressure signal, we believe that the system may not be adequate to cover all break sizes. Owing to the suppression characteristics 'of the containment, the possibility exists that the steam from a small break could be condensed without actuating the pressure trip. It may therefore be necessary to eliminate the drywell pressure signal as a requirement for depressurization. It is possible that the RCIC sys em might provide a back-up for-this critical break size.

Although the design basis for the vessel includes c e such depressurizatien transient duri,ng the vessel lifetime, we have ~ been infermed that the vestiel vould withstand many of these transients without sustai.ing damage. We believe that an accidental initiation of the system would requi e an analysis of the vessel integrity based on the recorded pressure-tempere ure history of the

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transient.

We understand that the applicent is continuing to a:alyze these problems and final resolution can be accomplished at the operating license stage.

5 Unlimited Feedwater Supply (Section VI-10.0) l Connections have been provided such that one of three large service water

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pumps could provide en unlimited supply of untreated veter to the main condenser  ;

l I j hotvell. The veter could then be transferred to the primary system by the feed-  !

veter pumps. This would depend on outside AC power to cperate the feedwater end f

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the service water pumps. We believe that this capability;is an important safety feature since it allows eventual flooding of the containment which would be re-quired for' final accident recovery.

6. . Alternate Water Sources for Engineered Safeguards (Amend. 2, p G-7-1;

. Amend 3, p'17-1)

The LPCI and core spray systems are fed from a single header connected to the suppression pool in the concrete compartment surrounding the suppress' ion ' torus.

The applicant has. stated that this header is of equivalent > integrity to the con-tainment. Its. loss, however, could be worse than the loss of containment integrity.

since with no core or containment cooling, the containment eventually would be breached and much larger quantities of fission products released. The applicant has provided an alternate water source, the condensate storage tank, for the LPCI and Core Spray Systems. The treated water will also be used for system test purposes. The HPCI system uses the condensate storage tank as the primary water source and uses the suppression pool as a secondary source. We believe that the 1

alternate sources provide the requisite backup.

C. Containment Cooling 1 The containment spray system is fed from the LPCI system pumps, and the i

applicant has stated that the containment spray would be used only after the LPCI l

l system has completed its flooding function. The basis for this sequential mode of operation is that the core is flooded to the 2/3 level in about three minutes, and the containment spray would not be immediately required even in the event that no core cooling was accomplished.

The flooding system, however, continues to require makeup for leakage through the core shroud and for boiloff due to decay heat. The staff therefore believes that the containment spray should be automatically operabic in parallel with the

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. makeup requirement of the LPCI system. GE is currently looking at orificing and additional valving to resolve this . problem but has indicated that initiation of Lthe spray would be manual. (Only one 'of the four pumps available 'to the LPCI ~

system is necessary to supply the full flow requirement'of'the containment spray.):

General Electric has indicated that the LPCI systen. would override the .

containment spray whenever the water level in the core dropped below the 2/3 -

mark. We believe that this is not acceptable since the time that the contain-ment spray would be ' required, if at all, vould be when the core had been. uncovered, metal-veter reaction had resulted and cooling of non-cendensible gases would be.

required. This is not to say that the core should not be flooded with large quantities of water when the water level in the core is _ low; rather.ve believe-that adequate pumping capacity should be provided to operate both the LPCI and the containment spray simultaneously. Because' of the increasing complexity of the' LPCI-containment sprey heat removal system, we believe it may be preferable to provide sept. rate simpler syster.s.

We have not completed our review of the interaction between the protective functions provided by these two systems since analyses by the applicant are not yet available. We believe the details of this problem can be handled during the final desi 6n of the facility.

D. Prevention of Bod Ejection (Amend 2, p A-3-1)

Thimble supports are provided in the design to prevent ejection of a control rod during operation. In case of a thimble failure, these supports would limit downward movement of the thimbles and internals to a few inches. To insure that no further control rod movement is possible, in case the rod is unlatched at the 1,

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.4FMGHAh-USE-ONL u time of the accident, the collet assembly must have the ability to engage and lock the index tube at the maximum velocity expected during failure conditions.

This maximum irelocity would be limited to about' 10 ft/sec due to ' hydraulic restrictions within the. drive assembly (General! Electric report GE CR-50891 and errata).

Tests have been performed in which the collet.vas allowed to engage a weighted section of index tube at velocities up to 15 ft 'sec. The index tube travelled less than one inch efter finger engagement although the fingers were bent et the maximum test speed. We believe'that the ability of the collet fingers to engage and lock the index tube has been adequately de onstrated. A detailed presentation of the test results are being prepared in GICR-5268.

E. Conformance of the Engineered Safeguards Systems to the Commission's Proposed Criteria Besed .on our enalysis of the engineered safeguards systems and with the -

reservations stated above, we believe that the final design of the systems can satisfy those:of. the Commission's proposed criteria which relate to engineered safeguards. The applicable criteria are listed below. Cnly those subjects on which new enelyses are available are discussed in detail.

Specifically, those criteria which we believe can be satisfied by the final design are:

I (1) No 1(a), which states that quality standards that reflect the importance of the function to be performed be used in the design of those features essential to the prevention o' accidents or to the mitigation of accident consequences. )

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(2) No. 2, which states that provisions must be included .to limit the extent of credible chemical reactions.

'(3) No.17, which states that the containment must te designed to take.

the largest credible energy release iricluding the effects of metal-veter reactions ' uninhibited by active quenching systems.

Although the applicant has stated that he believes that the design l basis should be " consistent with the performance objectives of the core cooling systems" (which results in less than 1% metal-water reaction) calculations have been presented to indicate the capability of the containment to withstand a large metal-veter reaction. The staff believes that this capability is required to account for the possibilities of-delayed initiation, partial ineffectiveness of the core cooling systems, or other uncertainties in this complex problem. ,

' An exemple of the necessity for designing for some metal-veter reaction is the case of cn eight-minute delay in actuation of the core spray. This delayed initiation leads to a calculated 12% clad melt 4

and a 10% metal-water reaction. Clearly, if this were to occur, some capability for accommodating metal-water reaction would be necessary.

An even higher amount of metal-water reaction could be associated with a similar amount of clad melt if the core cooling were only partially effective over a long time period.

The proposed containment is designed to acco=modate the time-rate of reaction associated with no core cooling. We believe that this is a conservative design and that Criterion No. 17 is therefore satisfied.

(Amend 3, p 10-1).

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(4) No.18, which states that two independent systems must be provided

' t'o remove decay and metal-water reaction heat from the. containment.

Containment spray is provided through either of two headers and from one of four pumps to-remove heat from non-condensible gases in the dryvell atmosphere to the suppression pool. Any steam generated.by boiloff in the reactor vill be condensed in the suppression pool.- The suppression pool is cooled by heet exchangers in the LPCI-containment sprey loops. (Amend 3, p 11-1).

(5) No. 21, which states that normal and emergency electrical power -

sources must assure continued maintenance of the facility in a safe .

.I shutdown condition under all credible circumstances. Three diesels are provided for emergency power, two of which vill provide minimum engineered safeguards .for one plant and also sustain systems essential to the shutdown of the second plant.

(o) No. 23, which states that acceptance of safeguards vill depend on the reliability and testability of the systems. Reliability has been provided by requiring the.t each type of primary system break be

" covered" by et least two independent systems, each of which has redundant components.

Testability of the systems varies. The RCIC system can be tested at any time during the operation of the plant. The Lov Pressure Coolant Injection, the High Pressure Coolant Injection and the Core Spray systems are testable up to the last isoletion valve l .

during operation by returning the water through test lines. Pro-visions vill be made to fully test the LPCI and Core Spray during OFFHCHAL USE9"iWhY-ltd,~ m .

OEESAL USE-Olha shutdown. Final test procedures are not available at this time but vill be prepared as the final designs of the systems progress.

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UMC4AL-USETML V. Containment A. Steam Line Isolation Valve Testing In response to staff questions, the applicant has described static tests to be perfctmed by the valve manuf acturer to determine valve closure times and leakage characteristics. A prototype valve.will be tested under pressure and temperature but with no flow and each valve will have a cold hydro and leakage test. Each valve will also have a closing test in. dry nitrogen at 1000 psig and seal leakage i

tests with air at 50.psig. In addition, closure tests under normal operating con- I dicions can be performed at any time during the life of the valve by reducing power to 75% and closing one of the four steam lines.

No prototype tests are proposed under the flow conditions which might be encountered as a result of a steam line break. Since the operation of at least c.ne of two isolation valves on each affected line is essential to the containment of the primary system water af ter a steam line break in this ' direct-cycle BWR, the l staff believes that the adequacy of this barrier should be damenstrated in prete-type tests. In addition to demonstrating closure ability within the specifications, j leakage properties of the valve should be determined for the case in which the valve is seated while hot and then cools. Cold leakage after a hot seating would be the probable sequence for a loss-of-coolant accident within the drywell.

(Amend. 2, p. D-2-1; Amend. 3, p. 8-1.)

B. Secondary Containment Testing l

The design specification of the secondary containment structure is 100% per day in-leakage at a negative pressure of 1/4 inch of water and that this pressure differential shall be maintained by the standby gas treatment system which exhausts to the facility stack. The proposed method of determining that the secondary L GEEHCHAL-USE ONL"i g.7 3 m.. 3

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containment meets 'this design ' specification is to measure the flow rate through the standby gas treatment system. Since the flow rate is fixed for a-given secondary

' containment pressure differential, any deviation from previous leakage conditions should be noted by a change in flow.

In our opinion, however, determination of in-leakage in this manner does not give adequate assurance that there is no. ground level out-leakage at any ' point on the building walls. We therefore believe that provisions should be made to measure pressure differentials at various points on each external wall to. illustrate that the pressure differential and thus the in-leakage specification has been met.

Details of this program will be worked out on similar reactors now approaching the operating license stage. (Amend. 2, p. D-6-1; Amend. 3, p. 9-1.)

VI. Instrumentation The reactor protection (scram) system is based on the dual logic (1 of 2, times 2) channel concept, utilizing in-core instrumentation. This coincident-redur. dant system has been' studied during our reviews of Dresden tinits 2 and 3, and

.. found to be inherently f ail-safe, testable for all credible faults, and immune to -

single faults. The dual logic system will also be used to initiate containment isolation and engineered safeguards. All containment isolation trips require manual reset, thus precluding the possibility of automatic de-isolation. Manual scram circuits are independent of the instrument channels. Instrument channels I used for control are independent of those used for safety.

Since our review of the Dresden plants, General Electric has made several changes to the design of their overall protection system. These changes will be incorporated within the Quad-Cities design and are described and analyzed herein:

(1) The rod-block system will be redundant and will consist of two relay logic 4FFHCHAL USE-ONL-Y ma i Ufw \

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chains which', when de-energized,'respectively interrupt' power to the rod select and' rod drive systems. . Those channels which use the rod block function to protect against' local' excursions due to accidental rod withdrawal (e.g., power-to-flow, APRM) .will be connected into this system in such manner as to take advantage of this redundancy (1/2 logic) .

(2) Containment isolation systems involving redundant valves will' use identical type valves inside and outside of containment. Each valve will have two solenoids, one a.c. and one d.c. , and will be air-held (open), spring icaded, and will f ail closed on loss of power. De-energizing both solenoids at each valve is required to initiate closure of a valve. A staff analysis has shown that, with proper attention to system logic and subsequent wiring installation, it is possible to design such-an isolation system within the requirements of Nos.15 and 16 of the Commission's proposed criteria. ';

(3) Discussions have been continuing with General Electric regarding a change in the design criteria for their refueling mode interlock system. Iriefly, it is the staff's desire to see the system designed to be immune to any conceivable single (equipment) failure which could allow the potential for a refueling excursion ,

accident to develop. This matter is under continuing study by both GE and DRL. We see no obstacle to its eventual satisfactory resolution.

(4) Two independent core spray systema are provided. Core spray is initiated by the low-low water level sensors connected in the dual logic (1 of 2, times 2) array.

Manual actuation capability is also provided. One pump (100% capacity) can be carried in the emergency load of the diesel generator units. Two pumps (200%

' capacity) can be carried by the nognal auxiliary power sources. Because of the inherent redundancy of the systems and their associated electrical supplies, and  ;

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the f ail-safe and redundant nature of the level-sensing instrumentation,'we believe the core spray system instrumentation can be designed to conform to our criteria.

(5). The Low Pressure Coolant Injection (LPCI) system will also be initiated by the low-low water level instrumentation. Further, the applicant has stated that instru-mentation.will be provided to ensure that full flow will be available to the reactor

. in the event of a broken recirculation line by preventing flow through the broken line._ This instrumentation is'not further described. However, since it appears to be. essential' to plant safety, the final design will be judged against criterion 15 which' states that testability and immunity to single f ailures must be provided. j (Amend. 2, p. H-8-1.)

(6) The High Pressure Coolant Injection System requires only d.c. power for its operation. It is initiated by low water level signals and stopped by high water level signals. Necessary valves are operated from the station battery.

The staff believes that the water level sensing instrumentation and associated cir:uits can be designed such that no f ailure therein will disable the HPCI system.

However, the system depends upon d.c. power for its operation. The reactor ,

depressurization system which, in combination with the low pressure injection systems,

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is intended to provide backup to the HPCI system for a certain range of break sizes, 1

also requires d.c. for its operation. The staff believes that the requirement for '

additional redundancy in instrumentation should be evaluated for these systems to 1

preclude failure of the system due to a local or complete failure of the d.c.

supply. This matter is under continuing study and will be resolved with the applicant prior to issuance of the operating license.

(7) The RCIC system also depends solely on d.c. power for operation of the vital j instrumentation, and we believe that the desirability of additional redundancy should also be evaluated for this system.

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, . . 2 VII. : ' Primary Sys tem Surveillance A. Summary of Proposals-

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Samples of the vessel material will be placed within the vessel as' irradiation

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surveillance. samples, and records of tests on these samples will be maintained for

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continuing surveillance.

Portions of the reactor vessel, recirculating water, feedwater,' and steam systems will be equipped with removable insulation to permit visual inspection and

, non-destructive testing of ex't'ernal surf aces. Types of tests mentioned by the applicant as performable were ultrasonic, dye penetrant, magnetic particle, and gamma radiographic. All primary system welds were listed as accessible. Only representative welds in the' steam feedwater, engineered safeguard, and other auxiliary high pressure system lines are to be accessible for non-destructive test-ing. The entire inner surface of the reactor vessel will be available, if required, for visual' inspection either directly or remotely.

A schedule for inspection of the primary sys tem was outlined in response to question 15 of. Amendment No. 3. Representative internal and external portions would be inspected over a five-year period. The total cycle would include all surfaces of the upper head, upper flange studs, nuts, internal and external surf aces of the upper shell course (12 feet) of the vessel including the lower closure flange and the exterior of all bimetallic welds in the primary system.

Similar inspections will be performed on lower parts of the vessel but at less frequent intervals, and the center course of the vessel would be inspected only once during the vessel lifetime (selected welds and surfaces during the 15th year).

Not all parts of the vessel would be inspected during the service life.

( (Amend. 2, pp. G-1-1, C-2-1; Amend. 3, p.15-1 and Appendix.)

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The Commonwealth Edison Company has stated a general policy of "backfitting" items related to safety to plants previously reviewed at the construction stage. {

Thus, any improvements in inspection frequency or technique decided on for Quad-Cities are at the same time determined for Dresden Units 2 and 3 now under construction and, to the extent possible, for Dresden Unit 1. The in-service inspection plans and associated design provisions for the reactors under construc-tion are being formulated concurrently.

B. Hydrostatic Proof Tests During Vessel Life The applicant proposes that tests of the integrity of the vessel be performed (at 125% of the design pressure) only as an initial test and af ter major repairs on the primary system. (Amend. 3, p. 15-3.)

Tightness tests will be performed each time the vessel head is replaced at a pressure below the lowest relief valve setting. The applicant considers it "not relevaet" to test at 125% of design at regular intervals since normal operation is at ET. cf design, and the l west relief valve setting is 86% of the design pressur2.

The design basis for the primary system does not include periodic testing above design pressure, but the applicant has stated that testing cycles could be accommodated within the present design. There appear to be inherent undesirable features in periodic hydro testing, such as imposing additional f atigue cycles on the vessel. The staff believes that further investigations of this problem must be carried out.

VIII. Turbine Orientation In response to questions by the staff, an analysis of potential turbine fa$1ure modes was presented. The analysis included discussions of the improved i design of recent turbines, automatic controls which prevent turbine overspeed, and

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a tabulation of the penetrating potential of theoretically possible missiles in the event that the overspeed controls failed. (Amend. 3, p. 3-1.)

The probability of turbine failure at operating speed has been reduced by j improved fabrication techniques which include additional heat treating cycles, more careful control of metal chemistry and composition, new metals with lower brittle-to-ductile transition temperatures, and improved mill practices. General Electric concludes that based on their improved turbine design, the probability of failure  !

at rated speed is virtually eliminated.

To protect against failure at overspeed, the turbine will be equipped with l control valves which are designed to trip shut at overspeed and stop valves or a trip-throttle valve. These will be controlled by the main governor, or emergency governor, or both. The speed governors are designed to limit steam flow to no more than 120% rated speed. A vibration pickup system will also be providec to trip the steam supply valves shut on signals caused by turbine distortion which would exist at above 120% of rated srced. Uncon: rolled steam flow could cause the turbine to accelerate to 200% of rated speed in the event that the overspeed steam control failed, the vibration sensor failed and partial turbine or generator failure occurring at lower speeds did not impede rotation. The failure at 200%

rated speed was taken as a limiting case by General Electric for the purpose of analysis to determine the maximum penetrating potential of various missiles.

Missiles from the high pressure turbine sections and from the generators are calculated to not escape their housings. There is no direct path by which a missile from the low pressure turbines could reach either drywell. Low trajectory missiles travelin8 between 19.5* and 13.5' from the horizontal plane and deflected from tie l vertical plane at the proper angle by some collision within the turbine housing or within the machinery room could strike either drywell but at reduced energies.

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b The applicant's analysis states that penetration of the drywell is not possible by credible missiles. Although the secondary containment could be breached, the safe shutdown of the reactor could still be accomplished by several independent means.

High trajectory missiles from the low pressure turbine could theoretically penetrate the secondary containment roof and damage, but not penetrate the shield plug. Since only a small deflection from the vertical would be required to reach

, the target, there would be little difference in the probability of striking the j l

shield plug for a differently oriented turbine. We understand that the applicant intends to consider missiles which might be generated from a rotor f ailure.

We believe that if proper consideration is given to protection of the primary containment and reactor shutdown systems from low trajectory missiles, that no advantage would be gained by re-orienting the turbines on an axis directed 'away from the containments. We believe the shield plug should be adequate to withstand a high trajectory missile.

IX. Accident Analysis The staff's evaluation of major accidents in the Quad-Cities Units 1 and 2 indicates that the consequences of a significant accident in one unit would not be propagated to the unaffected unit to cause secondary accidents. The worst influence on the operation of the second unit would be the restriction of access for routine maintenance in the common secondary containment. Accordingly, the following dis-cussion is limited to considerations of fission product release from a single typical unit.

The four major accidents postulated by the applicant (those involving the potential release of significant amounts of fission products) explore four possibic routes by which fission products might escape from the containment. The rod-drop accident releases fission products from the fuel within the confines of the primary OFFHCHAL-tJSFrONhY y,.

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coolant system. The refueling (fuel drop) accident releases fission products from the fuel directly to the secondary containment. The steam line break accident )

releases fission products entrained in the primary coolant directly to the atmosphere.

The coolan't-loss accident releases fission products from the fuel to the pressure-  !

suppression (primary) containment.

The postulated coolant-loss accident, involving rupture of a large line and a core heatup .and a time-rate of metal-water reaction associated with no active quenching systems is the worst accident, in terms of pressures, temperatures, and fission product release that the pressure suppression containment must withstand. i The design basis for the containment is further discussed above in Section IV.E.(3) of this report.

The staff's calculations of off-site consequences of these accidents are similar to those performed for Dresden Unit 3.

The lower exclusion distance.

0.225 miles, as opposed to 0.5 miles in the Dresden case led to an increase of about 20*

in the po te .tial t.co-hour coses calet. lated by the staff at the exclusion i

boundary except for the assumed steam line break (ground release) which increased by a factor of 4.

The 4-mile low population distance at the Quad-Cities site, as opposed to the 10-mile low population distance at the Dresden site leads to calcu-lated potential dos,es for the course of the accident which are about 40% higher for either unit of the Quad-Cities reactors. All doses calculated remain well within Part 100 guidelines.

A discussion of the detailed staff and applicant assumptions for each major accident is included in Report No. I to the Committee on Dresden Unit 3 and will not be repeated here except for the refueling accident, j

j A different mechanism has been postulated by the applicant for purposes of (

1 illustrating a fission product release directly to the secondary containment.

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In previous applications, a fuel bundle drop into an array made just critical by the withdrawa'l of two adjacent control rods was postulated. to occur through a I

series of operator errors and procedural interlocks. In the Quad-Cities applica-tion, the mode of fission product release is physical damage.(releasing gap fission products) occasioned by the drop of a fuel element onto the core.

. For this case, the staff calculations indicate that if 5% of 'the halogens and 10% of the noble gases were released from damaged fuel rods, release through the standby gas treatment system and the stack at 100% of the building volume per day would allow 100% of the core to be damaged without exceeding Part 100 guidelines.

As discussed in section VI of this report, we believe that the interlock system can be designed such that no single component failure, even when combined with operator errors, could lead to the drop of a fuel bundle into a critical fuel array. We have, nevertheless, in this case asked the applicant to consider the potential physical consequences associated with the postulated excursion to determine the p tential r.sgnitude of the consequences. The applicant has calculated that although a water slug would be expelled from -the pool, no physical damage to the secondary containment would be sustained and fission products released from the refuelin8 Pool would pass through the standby gas treatment system and exit from the stack. The staff calculations, with conservative source and plateout assumptions, indicate that potential doses would remain well within Part 100 guidelines.

X. Review of Previous ACRS Concerns This section is restricted to those items which have been identified in the Committee's letters on Dresden Units 2 and 3 and Millstone Point as requiring further l analysis or more detailed information.

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A. Inspectability of_the Primary System As discussed in Section VII of this report, the applicant has formulated a proposed inspection ' schedule (which will be common to the Dresden Units 2.and 3

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as well as Quad-Cities Units 1 and 2). We believe that continued discussions with the applicant on this subject will lead to mutually acceptable inspection procedures and techniques by the operating license stage.

B. Jet Pump Development The applicant has stated that a jet pump test program is currently in progress which will investigate jet pump stability and also steam carryunder in the pumps.

These tests, together with a carefully conducted startup program, will be reviewed to insure that all operational problems of safety significance can be recognized.

(Amendment 3, p. 14-1.)

C. Reactivity Transients Further information on design margins and the relationship of large excursions to the integrity of the princry system and internals were presented in the applica-tion. The staff plans to maintain an interest in large reactivity excursions in future reviews and in particular the damage threshholds of the primary system and internals will be further studied.

D. As discussed in section IV of this report, the applicant is continuing to develop and analyze those systems required for emergency core cooling. The staff's current analysis of the available information is also presented in section IV.

E. The staff also intends to continue to review other items such as missile pro-tection, pipe whipping, blowdown forces on vessel internals, and pressure vessel f abrication quality control. 4

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I dFHQALJLISE6LV XI. Items Requiring Continuing Review _

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. In this analysis, we have identified the several items listed below which will )

require further evaluation as the design progresses. Each item is common to other General Electric boiling water reactors, and we intend to continue our review of these items during subsequent licensing reviews.

(1) Design and analysis of the engineered safeguards complex including:

a. The redundancy of critical valves and instrumentation. i
b. The relationship between the LPCI system and the containment spray system operation,
c. The bases and analysis of minimum core flooding requirements.
d. The adequacy of the automatic depresst:.rization system to meet the design requirements.

(2) The seismic design of the water source for the HPCI system.

(3) The testing of steam line isolation valves for closure and leakage under simulated accident conditions.

(4) The desirability of " proof" testing during the vessel lifetime at pressures {

above the design pressure.

XII. Conclusion on the basis of our safety evaluation of the proposed design of the Guad-Cities Units 1 and 2, we believe that there is reasonable assurance that this facility can be built and operated at the proposed location without undue risk to the health and safety of the public.

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