ML19329E134

Chapter 1 of AR Nuclear 1 PSAR, Introduction & Summary. Includes Revisions 1-18

ML19329E134
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/24/1967
From:	ARKANSAS POWER & LIGHT CO.
To:
References
NUDOCS 8005300711
Download: ML19329E134 (54)

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TABLE OF CONTENTS SECTION TITLE PAGE 1 INTRODUCTION AND

SUMMARY

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1.1 INTRODUCTION

1-1 1.2 DESIGN HIGHLIGHTS 1-2 1.2.1 SITE CHARACTERISTICS 1-2 1.2.2 POWER LEVEL l-2 1.2 3 PEAK SPECIFIC PCWF2 LEVEL l-2 1.2.4 REACTOR BUILDING SYSTEM l-2 1.2 5 ENGINEERED SAFEGUARDS 1-2

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1.2.6 ELECTRICAL SYSTEMS AND EMERGENCY POWER l-3 1.2 7 ONCE-THROUGH STEAM GENERATORS 14 13 TABUIAR CHARACTERISTICS 1-4 1.4 PRINCIPAL DESIGN CRITERIA 1-6 1.4.1 CRITERION 1 1-7 1.4.2 - CRITERION 2 1-8 1.4.3 CRITERION 3 1-9 1.4.4 CRITERION 4 1-9 1.4.5 CRITERION 5 1-10 1.4.6 CRITERION 6 1-10 1.4.7 CRITERION 7 1-11 1.4.8 CRITERION 8 1-12 1.4.9 CRITERION 9 -

1-12 t-1.4.10 CRITERION 10 1-13 '

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g00 53007// lo-31-69 Supplement No. 13

TABLE OF CONTENTS (CONT'D.)

SECTION TITLE PAGE 1.4.11 CRITERION 11 1-14 O

1.4.E CRITERION 12 1-15 1.4.13 CRITERION 13 1-15 1.4.14 CRITERION 14. 1-16 1.4.15 CRITERION 15 1-16 1.4.16 CRITERION 16 1-17 1.4.17 CRITERION 17 1-18 1.4.18 CRITERION 18 1-20 1.4.19 CRITERION 19 1-20 1.4.20 CRITERION 20 1-21 1.4.21 CRITERION 21 1-21 1.4.22 CRITERION 22 1-22 1.4.23 CRITERION 23 1-23 1.4.24 CRITERION 24 1-24 >

1.4.25 CRITERION 25 1-24 1.4.26 CRITERION 26 1-24 1.4.27 CRITERION 27 1-25 15 RESEARCH AND DEVELOR4ENT REQJJIRH4ENTS 1-25 151 XENON OSCILLATIONS TEST 1-26 152 THERMAL AND HYDRAULICS FROGRAMS 1-26 153 IUEL ROD CLAD FAILURE TESTS 1-26 1 5.4 HIGH BURNUP FUEL TESTS 1-26a 155 INTERNALS VENT VALVES TESTS 1-26b 1 5.6 CONTROL ROD DRIVE-LINE TESTS

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TABLE OF CONTENTS (C0fff"d.)

SECTION TITLE PAGE 157 ONCE-THROUGH STEAM GENERATOR TESTS 1-264 1 5.8 SELF-POWERED DETECTOR TESTS 1-26d 159 BLOWDOWN FORCES ON INTERNALS 1-26e 13 1 5 10 CHEMICAL SPRAY SYSTEM FOR IODINE REh0 VAL 1-26e 1 5 11 SMALL xsREAK ANALYSIS 1-26f 1.6 IDENTIFICATION OF CONTRACTORS 1-27 1.7 CONCLLGIONS 1.27 I

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r LIST OF TABLES TABLE NO. TITLE PAGE NO.

1-1 INGINEERED SAFEGUARDS 1-28  ;

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l-2 ' C(EPARISON OF DESIGN PARAMETERS 1-29 4

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LIST OF FIGURES FIGURE NO.- TITLE 1-1 THE MIDDLE SOUTH UTILITIES SYSTEM 1-2 GENERAL ARRANGEMENT UPPER FLOOR PIAN 1-3 GENERAL ARRANGEMENT LOWER FLOOR PLAN 1-4 GENERAL ARRANGEMENT SECTION A-A 1-5 GENERAL ARRANGEMENT SECTION B-B 1-6 GENERAL ARRANGEMENT SECTION A-A 1-7 GENERAL ARRANGEMENT SECTION B-B 17 1 GENERAL ARRANGEMENT MISCELLANEOUS PIANS AND SECTIONS 1-9 EQUIPMENT AND AREA INDEK i

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5-4-70 Supplement No. 17 0014 W

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Answer:

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Emergency injection coolant is provided 'to maintain the core sufficiently covered to prevent core melting for the complete range of postulated reactor coolant system rupture sizes up to the maximum size of a 36-inch ID pipe. In the process of cooling the core, the metal-water reaction is limited to an in-significant amount. (Section 14)

Makeup pumps with a capacity of approximately 900 gpm will inject water into the reactor coolant system. This system is pr4 - ily effective early in the accident while the reactor coolant system pressure is above 100 psig. (Sec-tion 6)

A core flooding system is supplied to cool the core at intermediate to low pressures. This system consists of two independent core flooding tanks, each of which is connected to a different reactar vessel injection nozzle.

Decay heat pumps provide 9,000 gpm of water to cool the core when it is parti- ,

ally or totally uncovered, and the reactor coolant pressure has dropped below 100 psig. (Section6) 1.4.3 CRITERION 3 Protection must be provided against pcssibilities for damage of the safeguard-ing features of the facility by missiles generated through equipment failures inside the containment.

Answer:

- Protective walls and slabs, local missile shielding, or restraining devices will be provided to_ protect the Reactor Building liner plate and engineered safeguards within the Reactor Building against damage from missiles generated by equipment failures. The concrete enclosing the reactor coolant system serres as radiation shielding and as an effective barrier against missiles.

Local missile barriers will be provided for control rod drives and where re-quired because-of openin6s in the concrete shield enclosing the reactor cool-ar'. system. (Sections 4, 5 1.4.10 and 6) An evaluation of the fire protection syv ,' as it affects Class I structures or equipment protective functions is 17 cont . ad in Section 9.8 1.4.4 CRITERION 4

'Ihe reactor must be designed to acccmmodate, without fuel failure or primary system damage, deviations from steady state norm that might be occasioned by abnormal yet anticipated transient events such as tripping of the turbine-generator and loss of power to the reactor recirculation system pumps.

Answer:

The reactor is designed with a margin above normal operating conditions to accommodate anticipated abnormal deviations from the steady state operation.

This margin allows for deviation.s of temperature, pressure, flow, reactor

-power, and reactor-turbine power mismatch. The reactor is operated at a constant average coolant temperature above 15 per cent power and has a negative b power coefficient to dampen the effects of powert 'ransients. The reactor control system will maintain the reactor operating parameters within preset 0(31 5 limits, and the reactor protectica system will shut down the reactor if normal loperating limits are exceeded by preset amounts. (Sections ggnd14) 1-9 Supplement No. 17

The Unit- is shut down automatically by the reactor protection system if a com- '

plete loss of. electrical power occurs. Upon loss of external system electrical load, a reactor pcwer reduction occurs, and the reactor continues .to generate Unit power needs at reduced load. Se resultant reactor coolant system tempera-ture and volume increases for both of the above are held within design limits by relieving steam through the bypass to the condenser and/or secondary system re-lief valves to the atmosphere, thereby preventing excessive reactor coolant system pressures. Accordingly, these transients will not produce fuel or re-actor coolant system damage. (Sections 7 1 and 14.1.2.8)

The reactor coolant pumps are provided with sufficient inertia to maintain adequate ficw to prevent fuel damage if power to all pumps is lost. 'Ihe cri-terion for core prot ?ction following loss-of-coolant flow is to maintain a De-parture from Nucleate Boiling Ratio.(DNBR) equal to, or greater, than that at the design overpower level for initial power conditions up to and including the maximum operating power level of 107 5 percent power. Natural circulation coolant flow will provide adequate core cooling after the pump energy has been dissipated. (Section 14.1.2.6 and Figure 9-8) 1.4.5 CRITERION 5

'Ihe reactor must be designed so that pcwer or process variable oscillations or transients that could cause fuel failure or primary system damage are not ,

possible or can be readily suppressed.

Answer:

The ability of the reactor cor. trol and protection system to control the oscil-lations resulting from variatioa of coolant temperature within the control

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system dead band and from spatial xenon oscillations has been analyzed. Varia-tions in average coolant temperature provide nagative feedback and enhance re-actor stability during that portion of core life in which the moderator tempera-ture coefficient ^is negative. When the coefficient la positibe, rod motion will compensate for the positive feedback. The maximum power change rate result-ing from temperature oscillations within the control system dead band has been calculated to be less than 1%/ minute. Since the unit has been designed to follow ra=p load changes of 10%/ minute, this is well within the capability of the con-trol system.

Control flexibility with respect to xenon transients is provided by the ccmbina-tion of control rods, in-core instrumentation, and out-of-core instrumentation.

Within control rod limits, transient xenon related to load changes is controlled by the automatic control system. Axial, radial or aximuthal neutron flux changes will be detected by the nuclear instrumentation. Individual or groups of control rods can be positioned to suppress and/or correct flux changes. (Section 3.2.2.2 3) 1.4.6 CRITERION 6 Clad fuel must be designed to accommodate throughout its design lifetime all normal and abnormal modes of anticipated reactor operation, including the design overpower condition, without experiencing significant cladding failures. Unclad or vented fuels must be i.esigned with similar objectives of providing control over fission products. For unclad and vented solid fuels, normal and abnormal

! modes of anticipated reactor operation must be achieved without exceeding design -

, release rates of fission products from the fuel over core lifetime.

1-10 5-4-70 00.16 h .

Supplement No. 17 O

Answer:

Fuel clad integrity is insured under all normal and abnormal modes of anticipated operation by avoiding clad overstressing and overheating..

The evaluation of clad stresses includes the effects of internal and

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external pressures, temperature gradients and changes, clad-fuel inter-actions,' vibrations, and earthquake effects. The free-standing clad design prevents collapse at the end volume region of the fuel rod and provides sufficient radial and end void volume to accommodate clad-fuel

' intersections and internal gas pressures. (Section 3.2.4.2)

CLsd-overheating is prevented by satisfying the following core thermal and hydraulic criteria: (Section 3.2 3 1.1)

(a) At the design overpower no fuel melting will occur.

(b)- A 99 per cent confidence exists that at least 99 5 per cent of the fuel rods in the core will be in no jeopardy of experi-encing a DNB during continuous operation at the design over-power of 114 per cent.

1.4.7 CRITERION 7 The maximum ractivity worth of control rods or elements and the rates with which reactivity can be inserted must be held to values such that no single credible mechanical or electrical or electrical control system malfunction could cause a reactivity transient capable of damaging the primary system or causing significant fuel failure.

Answer:

Reactivity control will be accomplished by movement of control rod assemblies and by_ changes in soluble boron concentration in the reactor coolant. Each control rod assembly consists of a cluster of 16 control rods. The rod drives and their controls will have an inherent feature to limit overspe.cd in the event of malfunctions.

Approach to criticality and low power operation will be by manual rod with-drawal. The remaining rod assemblies (or rod assembly groups) will be inter-locked to permit withdrawal.on automatic control only after the rod groups used for approrch to criticality and low power operation have been fully withdrawn. - Rod assemblies used for automatic control will be arranged in four groups and interlocked to prevent simultaneous withdrawal of more than two groups. That is, simultaneous withdrawal of two automatic groups will be permitted over approximately the first 25 per cent of the second rod

^ group stroke and the last 25 per cent of the first rod group stroke.

The maximum reactivity insertion rate associated with simultaneous withdrawal ofaregulating,12-rodgroupis5.8x10~5(4k/k)/sec. Assuming a single electrical failure occurs that invalidates the interlock and permits the 25 control rod assemblies 'on automatic control to be withdrawn simultaneously,

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a maximum reactivity-insertion rate of 2 3 x 10 - (ok/k)/sec.couldresult.

Reactivity transients of this magnitude have been analyzed, and the resultant poker transients will not produce reactor coolant system or fuel failure.

(S ectio'n '14.1.2. 3) y- 0017

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A reduction in the reactor coolant soluble boron concentration will require ~

operator . initiation, and will be prohibited by interlocks until the control rod assemblies are in an acceptable pattern for dilution. A second safety feature will physical]ar limit the maxbum rate at which dilution water can be added to the system. A third safety measure will consist of a relay which will limit the tcc'al time of dilution. The gximum reactivity insert-ion rates from moderator dilution will be 7.0 x 10 rates are not sufficient to produce damage to either (A thek/k)/

fuelsec. These or reactor coolant system.

1.4.8 CRITERION 8 Reactivity shutdown capability must be provided to make and hold the core suberitical from any credible operating condition with any one control element at its position of highest reactivity.

'Ansuer:

The reactor is designed with the capability of providing a shutdown margin of at least 1 % d k/k with the single most reactive control rod fully with-drawn dition.

at any point in core life with the reactor at a hot zero power con-The minimum hot shutdown margin of 2.1%d k/k occurs at the end of life.

Reactor suberitical margin is maintained during cooldown by changes in soluble poison concentration. The rate of reactivity compensation from boron addition is greater than the reactivity change associated with *he maximum allowable N reactor cooldown rate of 100 F per hour. Thus, suberiticality is assured ~)

during cooldown with the most reactive control rod totally unavailable.

1.4.9 CRITERION 9

- Backup reactivity shutdown capability must be provided that is independent of normal reactivity control provisions. This system must have the capability to shut down the reactor from any operating condition.

Answer:

Soluble boron addition will provide an independent backup to the control rod assemblies for reactivity shutdown. Boron addition will be accomplished using the makeup and purification system. There are three makeup pumps to insure flow availability under au credible operating conditions. These pumps take' suction from the borated water storage tank, which contains water with 2.270 ppm boron, or from the makeup tank. In the latter case, a solution containing 8,750 ppm horon is supplied to the makeup tank from a mixing tank.

Two transfer pumps are provided. (Section 9.1)

The makeup pumps and the two sources of concentrated boron solution insure the capability of being able to shut down the reactor without any control rods from any operating condition. The following table demonstrates the capability of shutdown with control rods for two modes of makeup and puri-fication system operation.

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Soluble Poison Shutdown Canability Negative Reactivity Time to Shut Down from 100%

Feed Insertion Rate, Rated Power to Hot Zero Pcwer Concen- Feed 1,4 k/k/ Minute Condition (1), Minutes tration, Flow Rate, pum boron gem BoL EoL BOL EoL 8,750 20 0.0179 0.0217 67 106 4,121 70 2 0.0224 0.0353 54 65 3,196 140f3 0.0278 0.0543 43 42 1.4.10 CRITERION lo Heat removal systems must be provided which are capable of acco=modating core decay heat under all anticipated abncr=al and credible accident conditions, such as isolation from the main condenser and complete or partial loss of primary coolant from the reactor.

Answer:

R: actor decay heat will be removed through the steam gererators until the reactor 3 coolant system is cooled to 250 F. Steam generated by decay heat will supply the main and emergency feedwater pump turbines and can also be vented to atmosphere and/orbypassedtothecondenser. The steam generators are supplied feedwater by (1) the main steam-driven feedwater pumps, which can be operated at a reduced flow rate for decay heat removal, (2) by a steam-driven emergency feed pump, or (3) by a motor-driven emergency feed m energized from the emergency diesel-generator system.

The main feedwater pumps supply water contained in the feedwater train and the condensate storage tank to the steam generators. The emergency feed pump takes suction from the condenser hotwell and the condensate storage tank. These sources provide at least 250,000 gal. of water storage which is sufficient for decay heat removal for about one day after reactor shutdown with the condenser isolated.

The condenser is normally available so that water inventory is not depleted.

(Section 10).

Without use of reactor coolant pu=ps, decay heat will be removed by natural circulation through the reactor coolant system. (Section 14.1.2.8)

(1) Reactivity balance on Dopper and moderator equal to 1.2% dk/k for BoL and23%4k/kforEoL.

(2) Makeup to makeup tank at 20 gpm of 8,750 ppm boron from boric acid mix

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tank plus 50 gpm at 2,270 ppm boron from storage tanks.

(3) Makeup to makeup tank at 20 gpm of 8,750 ppm boron from boric acid mix tank plus 120 gpm at 2,270 ppm boron from storage tanks. ~

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'_' Supplement No. 3

..Under conditions of complete or partial loss-of-coolant from the reactor, decay heat vill be removed from the core by coolant supplied by the emergency injection coolant system. The source of injection water vill be the borated water storage tank. When this source is exhausted, the low pressure injection pumps will take suction from the Reactor Building sump. The return flov is cooled and pumped to the reactor vessel to continue core cooling. This system contains redundancy of equipment to insure availability of flow when required.

If complete loss of external electric power occurs, on-site sources supply sufficient electric power for all engineered safeguards and cooling water systems. (Section 14.2.2 3)

-1.4.11 -CRITERION 11 Components of the primary coolant and containment system must be designed and operated so that no substantial pressure or thermal stress vill be imposed on the structural materials unless the temperatures are ull above the nil-ductility temperatures. For ferritic materials of the coolant envelope and the contain-cent, minimum temperatures are NDT + 60 F and NDT + 30 F, respectively.

Answer:

The reactor vessel plate material opposite the core is purchased to a specified NUfT of 10 F or less and is tested to verify conformity to specified require-ments. (Section 4.2 5)

The end-of-Unit-life-NUIT value of the reactor vessel opposite the core vill s be not more than 260 F. Unit operating procedures will be established to limit j the operating pressure to 20 per cent of the design pressure when the reactor coolant system temperature is below NDTT plus 60 F throughout Unit life.

Surveillance specimens of the reactor vessel shell section material vill be installed between the core and inside vall of the vessel shell to monitor the NDTT of the vessel material during operating lifetice. (Section4.1.4)

The reactor vessel material is protected from excessive radiation damage by coolant water annuli between the core and the reactor vessel. The thickness of these annuli limits the total fast flux grggter than 1 Mev incident on the reactor vessel vall to an nyt value of 3 x 10e in 40 years at an 80 per cent l unit capacity factor. The thermal shield contributes to a further reduction l in vessel material radiation damage. (Section4.1.4) l The Reactor Building liner is enclosed within the Reactor Building and thus will not be exposed to the temperature extremes of the environs. The Reactor Building ambient temperature during unit operation vill be between 100 and 110 F which is expected to be vell above the NDT ' temperature +30 F for the liner plate.. The liner plate is completely enclosed by the thick concrete valls, slab, and roof of the Reactor Building, and thus will not be subject to sudden variations due to changes in external temperatures. In addition, the bottom liner plate is protected by a minimum thickness of 12 in, of l concrete. The penetrations vill be made of material which exhibits, by test, a transition temperature at least 30 F below the minimum service metal temperature.  !

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F 1.4.12 CRITERION 12 Capability for control rod insertion under abnormal conditions must be provided.

Answer:

Control rod assemblies will provide the normal means for changing reactivity to shut down to a hot suberitical condition. They may be inserted independ-ently of manual means. Both modes of insertion override reactor control system signals by interrupting power to the rod drives. Without power the control rod assemblies insert into the core by gravity. Soluble boron is added to maintain suberiticality from a hot to a cold zero power condition.

The principal safety criteria for the control rod drives are:

~(a) Ho single failure in the drive shall result in the loss of safety function.

(b) Trip action shall not require power, and no single failure or chain of failures shall prevent trip action to more than one mechanism.

(c) The trip ecmmand shall override all other commands. Trip action shall be nonreversible.

The reactor vessel, reactor vessel supports, reactor vessel internals, fuel assemblies, control rod assemblies, and the control rod drives are all designed to resist, without loss of function, the effects of seismic loadings established by the seismological analysis of the site.

The control rod assembly is never withdrawn completely from the fuel assembly.

The guide structure is oriented with respect to the fuel assembly by a common grid structure which maintains full stroke control rod guidance into the fuel assembly. The drive line is designed and will be tested to be fully operable under conditions of the maximum misalignment specified. (Section 3 3 3.4.1) 1.4.13 CRITERIOU 13 The reactor facility must be provided with a control room frcm which all actions can be controlled or monitored as necessary to maintain safe operational status of the plant at all times. The control room must be provided with adequate prctection to permit occupancy under the conditions described in Criterion 17 below, and with the means to shut down the plant and maintain it in a safe condition if such accident were to be experienced.

Answer:

The facility will be provided with a control room (Section 7.4). It is a design goal that entrance to this control room is maintained at all times.

As is discussed in Section 11.2.1.1, radiation exposures to plant personnel I during the 30 day period following a Maximum Hypothetical Accide.nt will not exceed 25 ren to the whole body. However, the accident doses may exceed the

- l limits specified in 10 CFR 20 for routine exposure to the plant staff but will not' exceed 10 CFR 100 limits.

1-15 0021 O

Jihe reactor protection ~ system.is designed to be essentially fail-safe without '

operator control. -Thus, safe shutdown can be achieved and maintained.

All instrumentation land controls necessary to operate the reactcr will be located l within the control room. This instrumentation includes the power and process variables, power load indication, temperature, pressure, flow,

-flow-mode, and control rod positions.

The engineered safeguards _will be controlled and maintained from the control room. The status of all ES' equipment will be indicated as will the variables pertinent to the ES system (Section 6.3). The various local radiation monitors will- have readouts displayed and always located in the control room (Section 11.1).

The control room ventilation system (Section 9.7, Figure 9-9) will limit-ingestion of. air borne containments to the control room to limit exposure as set forth above..

- 1. 4.14 - CRITERION 14 Means must be included in the control room to show the relative reactivity status of the reactor such as position indication of mechanical rods 'or concentrations of chemical poisons.

Answer:

The position of each control rod assembly will be displayed in the control room. The reactivity status of soluble boron will be indicated by the position of the control rod assemblies. The soluble boron concentration will be. adjusted to be _ consistent with specified red patterns and control rod -

assembly group position. . Accordingly, continuous indication of soluble boron concentration will not-be required. . The operator will receive results of laboratory analyses of the soluble boron concentration. (Section 7)-

1.4.15 CRITERION 15 JA reliable reactor protection system must be provided to automatically initiate appropriate action to prevent safety limits from being exceeded. Capability must be provided for testing functional operability of the system and for determining that no component or circuit failure has occurred. For instru-

.ments and control--systems in vital areas where the potential consequences of failure require redundancy, the redundant channels must be independent and

- must be capable of being tested to determine that they remain independent.

Sufficient redundancy must be provided that failure or removal from service of a single component -or channel will not inhibit necessary safety action when required. These criteria should, where applicable, be satisfied by the instrumentation associated with containment closure and isolation systems',

' decay _ heat removal and core cooling systems, systems _ to ' prevent cold-slug i accident's, and other vitalfsystems, as well as the reactor nuclear and process safety l system.

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Answer:

The reactor protection system is designed to provide the features specified in this criterion. A minimen of four sensors is provided for each trip variable except start-up rate. Two sensors are provided for start-up rate monitoring. Reactor trip is provided when the following parameters exceed preset values:

(a) Reactor power.

(b) Reactor outlet temperature.

(c) Reactor pressure.

(d) Reactor start-up rate.

If a portion of an instrumentation channel is removed frcm service, the channel assumes a tripped condition. One channel in a tripped condition places the protection system in a half-tripped mode such that a trip of any one of the remaining channels causes a reactor trip.

Reactor Building isolation and engineered safeguards are initiated from a 3-channel system described in Section 7 The pcwer supply for each individual channel will be from one of the two redundant battery-backsd vital busses. (Section 8) The channels are normally energized, and loss of power to one bus causes a reactor trip.

(Section 7)

Provisions will be included for testing the protection systems and/or components under administrative control on a periodic basis. Normal testing will include the insertion of a simulated signal to dynamically check response and performance of each channel's components except de-t ectors. Tests of each protection system channel will insure a high confidence level of system operability. (Section 7.1 3.5) 1.4.16 CRITERION 16 The vital instranentation systems of Criterion 15 must be designed so that no credible combination of circu= stances can interfere with the performance of a safety function when it is needed. In particular, the effect of in-fluences common to redundant channels, which are intended to be independent; must not negate the operability of a safety system. The effects of gross disconnection of the system, loss of energy (electric power, instrument air), and adverse environment (heat from loss of instrument cooling, extreme cold, fire, steam, water, etc.) must cause the system to go into its safest state (fail-safe) or be demonstrably tolerable on some other basis.

Answer:

Protection systems instrumentation is designed to operate in Reactor Building ambient conditions ranging from 40 F to 140 F without adverse effects in accuracy. ' Reactor Building temperature will be normally controlled in the L

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innge :of 60 F to 110 F. The protection system instrumentation, exclusive

. of the neutron detectors in the Reactor Building, will withstan:1 the external pressure and temperature for the. duration.of a loss-of-coolant accident and still-be operable (but subject .to - several per cent inaccuracy). The out-of-

-core neutron detectors are designed for continuous operation in,a temperature of 175 F and a pressure of 150 psig.

L . Redundant instrument channels are provided for all reactor protection and-engineered safeguard systems. -Ioss of power to'each individual reactor.

protection channel will trip that individual channel. Loss of all instru-ment. power will trip the reactor protection system, thereby releasing

[ the control rods. Engineered safeguards are normally de-energized controls.

They will be activated- through redundant controls and power systems.

(Section 7 1) l Manual reactor trip is' designed so that failure of the automatic reactor trip circuitry will-not prohibit or negate the manual trip. The same is true with respect to manual operation of the engineered safeguards equipment.

(Section 7.1) 1.4.17 CRITERION-17 The containment ' structure, including access openings and penetrations, must be designed and fabricated to accommodate or dissipate without failure the pressures and temperatures associated with the largest credible energy release including the . effects of credible metal-water or other chemical reactions uninhibited by active quenching systems. If part of the primary -

l- coolant -system is outside the primary reactor contaimnent, appropriate ^ ]1 - L.

j safeguards must be provided for that part if necessary, to protect the l health and safety of the public, in case of an accidental rupture in that

_p art of the system. The appropriateness of safeguards such as isolation-valves,' additional containment, etc., will depend on environmental and popu-lation conditions surrounding the site.- ~

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Answer:

The reactor containment, a' continuous, reinforced concrete structure, with a welded steel _ liner'to provide leak tightness, will completely enclose the entire reactor and. reactor coolant system, to ensure with certain engineered safeguards that an acceptable upper limit for leakage of radioactive materials to thef environment will not be exceeded, even if gross failure of the reactor coolant system were to occur (Sections 5 and 7). . It is a prime design goal to mainta4.n the integrity of the containment under both normal and accident conditions.

. The Reactor' Building, its openings and penetration, has a design pressure of 59?psig at 284 F. The greatest transient peak pressure, associated with a Tseverance of a reactor. coolant pipe and the resultant energy release from

. metal-water reactions and hydrogen burn up, will not exceed these values.

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The Reactor Building and engineered safeguards systems have been evaluated for various ccmbinations of credible energy releases. The emergency I cooling systems are sufficient to prevent overpressurisation of the structure and return the Reactor Building to near atmospheric conditions within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (Section 14.2.2 3.4)

The use of injection systems for core flooding will lbnit the Reactor Building pressure to less than the design pressure. If a metal-water reaction is un-inhibited by the active quenching systems the resultant peak Reactor Building pressure is less than the design pressure.

l No lines which contain high temperature, high pressure reactor coolant pene- l trate tne Reactor Building except the sampling lines. These small sampling lines are normally isolated by two valves in series. Therefore, it is only during a sampling operation that a line failure would require operator action to prevent escape of coolant external to the Reactor Building. This is a procedure that the operator would normally perform. I The makeup and purification system diverts a small amount of reactor coolant outside of the Reactor Building. This high pressure and high temperature coolant is cooled before it leaves the Reactor Building. Lines serving this function contain isolation valves that can be closed to prevent uncontrolled release of reactor coolant in the event a line fails external to the Reactor Building. The letdown coolers are supplied with water from the intermediate cooling system. Any leakage of reactor coolant through the letdown coolers will be into this system rather than to the environment.

The intermediate cooling system is monitored to detect leakage of reactor coolant.

Leakage of contaminated coolant from engineered safeguards equipment located external to the Reactor Building has been evaluated, and the resultant envir-onmental consequences are well below 10 CFR 100 limits at the site boundary, and have been included in the total accidental dose calculations.

The high pressure injection and low pressure injection systems have redundancy of equipment to insure availability of capacity. (Section 6.1)

Some engineered safeguards systems have both a normal and an emergency function, thereby providing nearly continuous testing of operability. For example, the makeup pumps are in continuous use for seal injection and makeup- ;he low pressure injection pumps are in use for decay heat removal during each shut-down; and two service water pumps are in continuous use.

During normal operation the standby and operating units will be rotated into service on a scheduled basis. In cases where separate equipment is used solely for emergency conditions, such as the Reactor Building spray pumps, recirculating lines are provided, and instranentation is installed to provide means for conducting tests. The equipment is located to facilitate inspection during operation. (Sections 6 and 9)

Electric motors, valves, and damper operators, which must function within the 1

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- Reactor Building during accident conditions, will operate in a stemn-air atmosphere at 284 F and 59 psig.

1-19 0025

1.4.18 CRITERI0II 18  %

Provisions must be made for the removal of heat from within the containment structure as necessary to maintain the integrity of the structure under conditions described in Criterion 17 above. If engineered safeguards are

.needed to prevent containment-vessel failure due-to heat released under such conditions,'at least two independent systems must be provided, pre-

'ferably of different principles. Backup equipment (e.g., water and power systems) to such engineered safeguards must also be redundant.

Answer:

m Reactor Building cooling following the loss-of-coolant accident is provided

' by two independent systems: (1) the Reactor Building spray, and (2) the Reactor Building emergency coolers. The capability of either of these cooling systems, or both at partia1' capacity, is sufficient to prevent

' excessive Reactor Building pressure during loss-of-coolant accident conditions.

The Reactor Building spray system supplies 3,000 gpm from the borated water storage tank'into the Reactor Building. After the borated water storage tank is emptied, recirculation from the Reactor Building sump begins. This recir-culated water.is cooled in heat exchangers by the service water system. The service water system.is always in operation, and therefore has continuously indicated availability.

Two sets of nossles, located in the upper portion of the Reactor Building .m structure, are arranged to provide a uniform spray pattern. Redundancy in j both pumping and heat exchanger capacity exists. (Section 6.2) -

To prevent excessive temperature rise following an accident, the Reactor Building cooling system has three emergency cooling units which reject heat to the service water system. Pumps and heat exchangers are redundant to insure availability. (Section 6.2)

Upon loca of external sources of electric power, one of the two diesel generators will permit operation of all required engineered safeguards equip-ment. (Section 8.2.3) 1.4.19 CRITERION 19 The maximum integrated leakage from the containment structure under the condi-tions described in Criterion 17 above must meet the site exposure criteria set forth in 10 CFR 100. The containment structure must be designed so that the

. containment can be leak tested at least to design pressure conditions after completion and installation of al1~ penetrations, and the leakage rate measured over a suitable period to ' verify its conformance with required performance.

The plant ~must be designed for later tests at suitable pressures.

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@ 002s

Answer:

It is a major containment design goal to provide a biological shield for as long as necessary under nor=al and accidait conditions (Section 51). The 3 design leakage at 59 psig of 0.20 percent by volume per day will not be exceeded under the accident conditions detailed in Section 14.2.2.4. The design pressure of 59 psig and design temperature of 284 F under accident conditions include a conservative margin for subsequent chemical reactions which might occur.

The environmental hazards resulting from an NHA, assuming the above specified leak rate from the Reactor Building, will be well below the limits of 10 CFR 100 (Section 14.2.2 3.5)

The contaiment and penetrations are designed to a 0.20 percent by volume per day leak rate to conform with the requirements of 10 CFR 100. The basis 3 of the pre-operational leak rate test is the reference volume method (Section 5 9).

The containment shall be designed so that the leakage rate can be periodically checked to design pressure of 59 psig, if necessary.

1.4.20 CRITERION 20 All containment structure penetrations subject to failure such as resilient seals and expansion bellows must be designed and constructed so that leak-tightness can be demonstrated at design pressure at any time throughout operating life of the reactor.

Answer:

All electrical penetrations have provisions for pressurizing between the double seals to 59 psig at any time during operation or shutdown to allow for leak checking by observing the pressure decay. There are no pipe penetrations to the containment which require a bellows seal between the pipe and the containment.

1.4.21 CRITERION 21 Sufficient normal and emergency sources of electrical power must be provided to assure a capability for prompt shutdown and continued maintenance of the reactor facility in a safe condition under all credible circumstances.

Answer:

In the event of loss of all off-site power, a drop in load to auxiliary load on the nuclear power system is accomplished by the step load reduction de-  ;

tailed in Section 7.2.1.1. In addition, emergency power sources provide a '

dependable supply of power for the critical services in the unlikely event of simulta.neous loss of normal and standby power (Section 8.2.3). Two diesel generators supply the 4160-volt buses and station batteries, and will provide  !

the power required for the vital auxiliaries, instrumentation, control equip- l ment and emergency lighting to enable safe plant shutdown. '

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5-3-68 Supplement No. 3

The failure of a single active component in the offsite power supply will result in loss of the offsite power supply. Sufficient redundancy :.s provided by redundant busses, two diesel generators, and station batteries to enable

'the engineered safety features to operate assuming a failure in a s:.ngle active component in the onsite system (c.f. Figure 8-1). A complete failure of the offsite transmission system will mean a drop in load to 5 per cent full load on the nuclear power supply as noted in Section 7.2.1.1.

1.4.22 CRITERION 22 Valves and their associated apparatus that are essential to the containment function must be redundant and so arranged that no credible combination of circumstances can interfere with their necessary functioning. Such redundant valves and associated apparatus must be independent of each other. Capability must be provided for testing functional operability of these valves and assoc-iated equipment to detemine that no failure has occurred and that leakage is within acceptable limits. Bedundant valves and auxiliaries must be in-dependent. Containment closure valves must be actuated by instrumentation, control circuits, and energy sources which satisfy Criteria 15 and 16 above.

Answer:

All penetrations vital in the containment functions are part of the isolation system (Section 5.4). inere are four types of isolation valving de}endent upon the individual system requirements under accident conditions. There is sufficient redundancy in the instrumentation circuits of the safeguards ,

actuation system to minimize the possibility of inadvertent tripping of the ,

isolation system. #

The isolation system closes all fluid lines (except those associated with engineered safeguards systems) penetrating the Reactor Building in the event of a loss-of-coolant accident. Reactor Building isolation occurs on a signal of approximately 4 psig or by manual actuation from the control room.

The criterion for isolation valve requirements is:

Leakage through all fluid penetrations not serving accident-consequence-limiting systems is to be minimized by a double barrier so that no s1ngle credible failure or malfunction of an active component can re-sult in a loss of isolation or intolerable leakage. The double barriers take the form of closed piping systems both inside and outside the Reactor Building and various arrangements of isolation valves.

(Section5.2)

. Fluid penetrations serving engineered safeguards systems also meet this criterion, but the actuators are manually operated from the control room.

The control circuitry that initiates Reactor Building isolation is part of the engineered safeguards protective system and is. designed to meet Criteria 15 and 16. (Section 7.1.3.2) 0028 O 1-22 ..

9

Tasting of the isolation valves is proviced for in the cperation sequence of the esscciatei process systems. The associated apparatus essential to the cantainment function is redundant and can be tested in conjunction with the engineered safeguards system. Valve leakage can be detected through the monitoring instrumentation provided for in Section 7.3.

1.4.23 CRITERION 23 In determining the suitability of a facility for a proposed site the accept-ance of the inherent and engineered safety afforded by the systems, materials and components, and the associated engineered safeguards built into the j facility, will depend on their demonstrated performance capability and re-liability and the extent to which the operability of such systems, materials, components, and engineered safeguards can be tested and inspected during the life of the plant.

Answer:

All engineered safeguards systems are designed so that a single failure of an active component will not prevent operation of that system or reduce the capacity below that required to maintain a safe condition. Two independent Reactor Building cooling systems, each having full heat removal capacity, are used to prevent overpressurization. (Sections 6.1 and 6.2)

The high pressure injection and low pressure injection systems have redundancy of equipment to insure availability of capacity. (Section 6.1)

Some engineered safeguards systems have both a normal and an emergency function, thereby providing nearly continuous testing of operability. During normal operation, the standby and operating units will be rotated into service on a scheduled basis. The answer to Criterion 17 (Section 1.h.17) gives more detail regarding redundancy, testing, and normal and emergency operation of9 engineered safeguards.

Engineered safeguards equipment piping, which is not fully protected against missile damage, utilizes dual lines to preclude loss of the protective function as a result of the secondary failure. (Section 6)

All components of the containment pressure-reducing systems can be inspected thrcugh testing. Pumps, valves, no: les, and blowers can be flow inspected, and sumps, lines and tanks can be visually inspected.

1.4.24 CRITERION 24 (

All fuel storage and waste handling systems must be contained if necessary to prevent the accidental release of radioactivity in enounts which could affect the health and safety of the public.

k

.f b.

y 1-23 S

I I esver:

. w This decay heat retcval is accomplished by the spent fuel ecoling system. In thecaseamaximumof1-1/3coresarestoredduetocompleteunloadingofthe reactorvesselwith1/3coreinresidence,twopumpsandtwocoolerswill maintain the spent, fuel pool at 132 F. One pump and cooler will maintain the pool at 169 F. If all cooling is lost, it will take 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> for the pool to reach 212 F. In this time repairs could be made to restart at least one spent fuel cooling pump. The most serious failure would be complete loss of water in the storage pool. For protection, cooling connections enter near or above the water level so that the pool cannot be gravity-drained. These precautions together with the shielding specified in Section 11.2, will prevent radioactive release to the plant environs.

Accidents assuming rupture of a waste gas tank have been evaluated and the consequences of the release show to be well below the guideline values of 10 CFR 100. (Section 11.1.h)

Damage to a fuel assembly in the spent fuel pit releasing radioactive g3ses to the auxiliary building was evaluated. With filtration and exhaust of these gases through the plant vent the off-site dose is a factor of more than 10 below the 10 CFR 100 guidelines. (Section 14.2.2.1.2) 1.L.25 CRITERION 25 The fuel handling and storage facilities must be designed to prevent criti-cality an'd to maintain adequate shielding and ccoling for spent fuel under all anticipated normal and abnormal conditions, and credible accident con- '}

ditions. Variables upon which health and safety of the public depend must be monitored.

Anawer:

Criticality in the new and spent fuel storage systems is prevented through storage in a geometrically safe condition. The new and spent fuel assemblies are stored in racks in parallel rows having a center-to-center distance of 21 inches in both directions. This spacing is sufficient to maintain a keff of less than 0.9 for the new fuel when wet (Section 9.6;. The new and spent fuel storage racks are designed so that it is impossible to insert fuel assemblies. in other than the prescribed locations, thereby insuring the necessary spacing between assemblies. The 21 in. x 21 in. spacing insures an eversafe geometric array in unborated water. Under these conditions, a criticality accident during refueling or storage is not considered credible.

The shielding provided in the spent fuel and waste storage area is the -

shielding for Zone II (Section 11.2) enabling the plant to meet the require-ments of 10 CFR 20.

~1.4.26 CRrfERION 26 Where unfavorable-environmental conditions can be expected to require limit-

~

ations upon the release of operational radioactive effluents to the environ- gc-ment, appropriate holdup capacity must be provided for retention of gascous, (g(()

liquid, or. solid effluents.

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0030

!hk%D .1-24 gg -- es---- M

Answer:

The liquid radioactive waste control system collects, treats, stores, and disposes of all radioactive liquid wastes. These are collected in sumps and drain tanks and then transferred to the appropriate tanks in the auxi-liary building for treatment, storage and disposal. Wastes to be dis-charged from the system are processed on a batch basis with each batch being processed by such methods which are appropriate for the materials present.

After analysis, processed liquid wastes are discharged through the circulating water discharge canal.

Solid radioactive wastes are collected, processed, packaged, and stored temporarily on the site in a shielded structure to permit decay or accumulation prior to shipment from the plant for permanent storage.

Equipment is provided to compress and holdup radioactive gases for decay before release through the station vent at a controlled rate.

(Section11.1.32) 1.4.27 CRITERION 27 The plant must be provided with systems capable of monitoring the release of radioactivity under accident conditions.

Answer:

Monitoring of all station solid, liquid, and gaseous releases is accomplished with the appropriate instrumentation (Section 7 3 and 11.1.3.4). Releases from the containment ventilation are monitored systematically, prior to release (Section 5 5). The plant ventilation (Section 9.7) is similarly monitored and dilution controlled to achieve 10 CFR 20 concentrations. All li uid effluents are monitored both prior to and after treatment with appropriate hold up available for decay purposes (Section 11.1). The solid effluents are baled and checked for radioactivity before shipnent off-site. Hence, monitoring of the releases within the facility environs is controlled so that the releases are never more than allowed by 10 CFR 20. Beta-gama detectors located in selected areas of the station used with operating procedures will, assure that personnel exposure does not exceed 10 CFR 20 limits (Section 11.2.2).

Monitoring and alarm instrumentation is provided sensitive to the operation of the spent fuel and decay heat cooling systems (Section 7 3 2). Continuity in safe operation is dependent upon the spent fuel and decay heat systems and the appropriate shielding (Section 11.2) is provided to enable access in the event of decay heat removal system trouble.

1.5 RESEARCH AND DEVEIDPMENT REQUIREMENTS The research and development programs that have been initiated to establish final design or to demonstrate the capability of the design for future oper-ation at a higher power level are summarized as follows:

A m.'

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1-25

1 5.1 XENON OSCILLATIONS An analysis to evaluate the possibility of xenon oscillations throughout core life is underway. A modal analysis to determine critical parameters has h been completed, and the detailed spatial calculations are in progress. If it is determined that such oscillations may occur, appropriate design changes to eliminate or control the oscillations vill be incorporated.

See also 3.2.2.2.3. and Questions 1.31, 1 5 (Item 6), 3.1, 3.2, h.2, and 6.6 of Supplement 3 to this PSAR.

1 5.2 THERMAL AND HYDRAULIC PROGRAMS B&W is conducting a continuous research and development program for heat trans-fer and fluid flow investigations applicable to the design of the Russellville Nuclear Unit. Two important aspects of this program are: 5

a. Reactor Vessel Flow Distribution and Pressure Drco Tests A 1/6-scale model of the vessel and internals is under test to measure the flow distribution to the core, fluid mixing in the vessel and core, and the distribution of pressure drop within the reactor vessel.

b. Fuel Assembly Heat Transfer and Fluid Flow Test Critical heat flux data have been obtained on single-channel tubular and annular test sections with uniform and non-uniform heat fluxes, and on the multiple rod fuel assemblies with uni-form heat fluxes. These data have been obtained for a range of pressure, te=perature, and mass velocities encompassing the reactor design conditions. This work is being extended to include multiple rod fuel assemblies with non-uniform axial heat generation. Additional mixing, flow distribution, and pressure drop data will be taken on models of various reactor flow cells and on partial full-scale fuel assemblies.

See also 3.3.2, and Question 1.3d of Supplement 3 to this PSAR.

1 5.3 FUEL ROD CLAD FAILURE A study of clad failure mechanisms associated with a loss-of-coolant accident is presently underway. This study has included identification of the potential 4 failure mechanisms, a search of the literature to obtain applicable data, evaluation and application of existing data, and scoping tests to obtain data on potential failure mechanisms. The initial results of this study include the identification of the failure mechanisms, an evaluation of the information avail-able in the literature concerning these mechanisms, and an evaluation of the effects of these mechanisms on the reactor system design.

0032 "

9 T-3-68 Supplement No. 5 m

The objective of the study is to ensure that there are no potential failure mechanisms that might interfere with the ability of the emergency core ecoling systems to terminate the core temperature transient and remove decay heat in h the event of a loss-of-coolant accident. These potential failure mechanisms include clad melting, zirconium-vater reaction, eutectic formation between the Zircaloy-clad and the stainless-steel spacer grids, the possibility of clad embrittlement as a result of the quenching during core flooding, and clad per-foration or deformation accompanying its failure. In the case of clad melting and zirconium-water reaction, our present design limit for peak clad temperature precludes these as possible failure modes. Information available in the liter-ature, along with experimental evidence from tests conducted by B&W, show that brittle fracture of the cladding vill not occur as a result of quenching fol-loving a loss-of-coolant accident, and that eutectic formation between dissimilar core materials will not interfere with the flow of emergency core coolant after the accident.

B&W has undertaken a program to evaluate the effects of perforation and deforma-tion'of fuel rods during the temperature transient following the loss-of-coolant accident. Preliminary tests have been run on nine samples of Zirealoy h cladding filled with ceramic pellets, and additional experiments are planned to gain a clearer understanding of the effects of temperature excursions on Zircaloy-clad fuel elements. Current plans include performance of a three-phase program. In the first two phases which are experimental, single-rod excursions will be performed to better establish temperature-presnure relation-ships at the time of clad perforation. The single-rod tests of the first phase vill also investigate the extent of deformation to be expected under the vary-ing conditions associated with simulated in-reactor temperature excursions.

These vill include the effects of hydrogen concentration and oxide films. The second phase of the program vill consist principally of multirod tests to ex-plore the effect of the restraining action of spacer grids and adjacent fuel rods and.to determine the randomization of the localized deformation in an assembly of fuel rods. In the third phase of the program, the data obtained from the two experimental phases vill be applied to the analysis of the effects in a loss-of-coolant accident.

See also Questions 1.3h and 1.5 (Item 5) of Supplement 3 to this PSAR.

1.5.h HIGH BURNUP FUEL TESTS B&W is conducting a program to obtain a better understanding of fuel growth rates and irradiation effects on cladding, the influence of hydrogen on clad-ding, and fission-gas release at high burnup for the specific design burnup projected for peak-power regions in the reactor.

The fuels irradiation program vill test fuel specimens at design temperatures and at exposures in excess of thoseobtained in the fuel rod. The specimens irradiated to the design burnup are scheduled to be completed in mid-1969 The program vill provide information on the swelling rate of UO as a function 2

of burnup, density, heat rate, and cladding restraint. Fuel specimens will be operated at heat rates up to 21.5 kv/ foot, which is in excess of the peak  ;

specific power in the core. The burnup will range up to 75,000 MWD /MTU. The l fuel rods will operate with a cladding surface temperature of 650F. '

i 0033 I

6-5-68 l-26a Supplement No. 4 i

A detailed report of sources of information for the irradiation of clad and fuel has been presented in the PSAR, 3 2.4.2 plus references. In addition to the PSAR references, irradiation of fuel assemblies or partial fuel assemblies with Zircaloy-clad UO2 is in pr gress in the Saxton and Big Rock Point reactors. These data will demonstrate the behavior of fuel assemblies under the combined effects of irradiation, pressure cycles, .

thermal gradients, reactor coolant environment, and fuel-clad restraints.

A program has been carried out to determine the effects of irradiation on the mechanical properties of Zircaloy-4. Tests were conducted to temperatures as high as 775 F.

This testing will be used to provide design information for advanced cores.

B&W has established a schedule to complete testing and hot cell examination 13 late in 1971 and to write a report early in 19'72.

See also Question 1 3f of Supplement 3 to this PSAR.

155 INTERNALS VFliT VALVES The internals vent valves will be designed to relieve the pressure gener-ated by steaming in the core following the LOCA so that the core will re-main sufficiently covered. The valves will also be designed to withstand the forces resulting from rupture of either a reactor coolcnt inlet or outlet pipe.

Testing of the valves will consist of the following:

a. A full-sized valve assembly (seat, locking mechanism, and socket) will be hydrostatically pressure-tested at static conditions to the maximum pressure expected to result during the blordown.

b. Sufficient tests will be conducted at zero pressure to determine the frictional loads in the hinge assembly, the inertia of the valve cover, and the cover rebound resulting from impact of the cover on the seat so that the valve response to cyclic blowdcwn forces may be determined analyti:: ally.

c. The valve assembly will be pressurized to determine what pressure differential is required to cause the valve to begin to open. A determination of the pres -

sure differential required to open the valve to its maximum open position will be simulated by mechanical means.

1-26b 10-31-69 Supplement No. 13 i

0034

d. A valve assembly will be installed and removed remotely in a test stand to judge the adequacy of handling equipment.

e. A valve assembly will be prototype tested ever an appropriate range of vibration frequencies and amplitudes to verify the analytical results showing that the valve will not unseat because of vibration duric& normal operation.

Since the temperature differential existing ucross the valve assembly during normal operation in the reactor is only approximately 55F, and since the same material is used for the valve seat, socket, and cover, there is no need to conduct tests at elevated temperatures.

See also 3.3.4, and questions 1 3e and 1 5 (Item 8) of Supplement 3 to this PSAR.

b 1-26c 10-31-69 Supplement No. 13 ,

1 s

.0035

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i l

1.5.6 CONTROL ROD DRIVE LINE TEST The test assembly for this program is a full-sized fuel assembly with associ-ated control rod and control rod guide, adjacent internals, and control rod drive. The unit is bein6 tested under conditions of temperature, pressure, flow, and water chemistry specified for the full-sized reactor installation.

This program vill embrace a prototype phase in which the unit vill be subjected to misali6nment, varying flow, and temperature. The second phase of this program is one of life-testing where the unit vill be continuously cycled to cover the number of feet of travel and the number of trips anticipated for its life in the reactor. Both phases of the program vill confirm the operability of the drive line in normal and misaligned conditions, confirm .the rod drop times and load-carrying characteristics of the actuator, indicate vibration and fretting wear characteristics of the control rods and fuel assemblies, and determine the wear characteristics of all the drive line components. Also, a component test program is being conducted using autoclave testing of selected components at reactor pressure and temperature. The purpose of this program is to seek out potential material and/or design problems prior to production unit testing.

See also 3.3.3.1 and 3.3.3.h, and questions 1.3b of Supplement 3 to the PSAR.

157 ONCE-THROUGH STEAM GENERATOR TEST Testing necessary to prove the adequacy of the once-through steam generator design for service at the initial power level and to confirm the size and confi6uration of the units has been completed. Steady-state and load-changing operations using once-through steam generator models were performed to demonstrate the ability of the unit to follow the transients and to demonstrate the interaction of the 5

control system with the water level, steam pressure, and flows. The test

, equipment consisted of one 37-tube full-length unit, one 19-tube full-length unit, and a full-length 7-tube unit. The tubes were fabricated in accordance with the production techniques anticipated for the full-sized unit.

The latter portion of the program included tests to determine the natural frequency of the tubes in the steam generator by subjecting them to artificially induced vibrations from an external source. The buckling and vibration characteristics verify the structural integrity of the tube design.

Primary and secondary blevdown tests on the models have demonstrated the integrity of the unit under conditions of rapid depressurization and large tube-to-shell temperature differentials. The results of these tests are being used in the development and verification of flowdown analyses.

See also Question 1.3a of Supplement 3 to this PSAR.

1 5.8 SELF-POWERED DETECTOR TESTS The test units for this program are the self-powered detectors described in 7.3.3. These units have been tested in the B&W Test Reactor at conditions of temperature and neutron flux anticipated in a central station reactor. These -

units are currently being tested in the Big Rock Point, Nuclear Power Plant where they are exposed to temperature, neutron flux p and flow for conditions

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1-26d 10-31-69 SUPPLEMENT No. 13

+

v

(pproximating tho:2 in th2 Russ211v111o Nuclear Limit. Th2 racults of thero programs will prcvid2 a dattetor syctem with pradictable characteristics of performance and longevity under incore conditions.

See also Question 1.3c of Supplement 3 to this PSAR.

159 BLOWDOWN FORCES ON INTERNALS

~

B&W has developed an analog computer model to obtain detailed information on the forces imposed on the reactor vessel internals during the subcooled por- '

tion of blowdown following a reactor coolant system rupture. , The model can be used to simulate leaks at any location in the hot and cold leg piping. Leak sizes up to a complete shear of a main coolant pipe can be simulated. Resis-tance to flow between all regions is simulated, as well as the inertia of the fluid in the connecting flow paths. Pressure in each region is calculated by using the equations for conservation of mass and momentum and assuming an isentropic expansion of water in each region. The mimimum pressure in each region is restricted to the saturation pressure corresponding to the temperature in that region.

Test results have been obtained by the Phillips Petroleum Company for the blow-down of a vessel with and without simulated reactor vessel internals. Addi-tional blowdown testing has been conducted and is still underway using the 1/h-scale LOFT vessel. Tests have been conducted with and without internals in the vessel.

The tests that have been completed, together with those that are underway,.Will provide an adequate amount of test data to verify the B&W analytical model.

N See also Question 1.3J and 1.5 (Item 7) of Supplement 3 to this PSAR.

• i 1510 CHEMICAL SPRAY SYSTEM FOR IODINE REMOVAL *

.)

B&W, ORNL, and o'thers have been and are now generating significant q.uantities of data which aid in the evaluation of the chemical spray system. Spray tests at ORNL have demonstrated that the chemical spray system vill function as an effective engineered safety system for iodine removal following an MHA.

Some additional R&D is desirable to provide more detail coverify th'at the chemical spray solution system vill continue to function effectively under post-accident conditions.

The objectives of these additional programs are to verify that the chemical solution:

a) Does not exhi' bit chemical or physical changes that are

[ detrimental to the solution's effectiveness for iodine l

retention or produce excessive amounts of undesirable decomposition products when it is exposed to temperature, pressure and irradiation doses associated with the post-accident conditions.

b) Dces not result in significant chemf' cal attack on or

[ corrosion of the' primary construction materials .

(Zircaloy,-Inconel, stainless steel, carbon steel,

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l .

paint, and concrete), during the post-accident J.

period. CHD'317 cR5

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10-31-69' i

@ l-26e SUPPLEMENT NO. 13

c) Is compatible with soluble boron compounds..

1 5.11 SMALL BRFAK ANALYSIS This analysis has been complete 6 .v Babcock & Wilcox and the results are included in the-DUKE FSAR. v J FSAR will also report this work.

The following Babcock & Wilcov Topical Reports will be incorporated in our FSAR to the extent * ; .eable:

SUBMITTAL NUMBER TITLE OR DESCRIPTION DATE BAW-10001 In-Core Instrumentation Test Program September 1969 BAW-10002 Once Through S-G R&D August 1969 BAW-10005 Internals Vent Valve Evaluation July 1%9 BAW-10006 Reactor Vessel Material Surveillance Program June 1969 BAW-10007 Control Rod Drive System Test Program July 1969 BAW-10008 Reactcr Internals Stress and Deflections Part I Due to LOCA and Maximum Hypothetical Earthquake July 1969 BAW-10008 Supplement (to deal with core) October 1969 BAW-10009 Fuel Rod Failures During LOCA Early 1970 BAW-10010 Stability Margin for Xenon Oscillation August 1969 Part I Madal Analysis BAW-10010 Supplement (to deal with multi-dimensional analysis) Early 1970 BAW-10012 Vessel Model Flow Tests October 1969 13 BAW-10014 Analysis of Sustained DNB Operation August 1969 BAW-10017 Chemical Sprays R&D October 1969 BAW-10018 Reactor Vessel IDCA Thermal Shock Analysis May 1969 The above reports, which are not proprietary, will be incorporated in our application by placing a formal letter addressed to the Atomic Energy Commission in the public record. Proprietary reports will be handled in the same manner as they were for the DUKE FSAR.

Programs underway at B&W will provide the data needed to satisfy the foregoing objectives. B&W will have this data in time to incorporate it into the system design as required to meet constructica schedules.

• For details of the chemical spray system R&D Program, refer to Met Ed PSAR for Three Mile Island Nuclear Station, Docket No. 50-289, Amend-ment No. 2 (Supplement No. 1) dated October 2, 1967, question 5 13, and Amendment No. 4 (Supplement No. 3) dated December 8,1967, question 17.4.

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4$ 0038

-1.6 IDENTIFICATION OF CONTRACTORS Arkansas Power & Light Campany, as owners, have arranged for the purchase of equipr'.nt and consulting, engineering, and construction services for the in-

- stal_ation of the Russellville Nuclear Unit. As sole owners, Arkansas Power &

Light Company is responsible for the design, construction and operation of the unit.

The Bechtel Corporation has been retained for engineering, procurement and construction services. They will also provide assistance in obtaining li-censes and permits, in employee training, in acceptance testing, in quality control, and in initial start-up of the project.

The Babcock & Wilcox Company has been contracted with to design, manufacture and deliver to the site a complete nuclear steam supply ystem and fuel. In addition, the Babcock & WIlcox Company vill supply competent technical and professional consultation for erection, initial fuel loading, testing and initial start-up of the complete nuclear steam supply system, Babcock & Wilcox will participate in initial plant personnel training.

The Westinghouse Electric Company will supply the turbine generator and its auxiliaries.

1.7 CONCLUSION

S It is concluded from this report that the proposed Russellville Nuclear Unit can be designed, constructed and operated in a safe manner; that the proposed design vill provide adequate protection to the public fran any sequence of events resulting in disablement of equipment from causes, natural or mechanical; and. that Arkansas Power & Light Company is qualified to start, operate,and maintain the Russellville nuclear unit in accordance with all applicable laws and regulations and in a manner satisfactory to the Atomic Energy Commission and to the public interest.

0038 b . .

6-5-68 1-27 Supplement No. 4

Table 1-1 Engineered Safeguards Function Total Equi 1xnent Installed High Pressure Injection 3 Pumps (makeup) ]

1 Storage Tank >

Core Flooding System 2 Tanks Iow Pressure Injection 2 Pumps (Decay Heat Removal) 2 Heat Exchangers Reactor Building Spray 2 Pumps System 1 Sodium Thiosulfate Tank 3 Reactor Building Cooling 2 Pumps System 3 Emergency Cooling Units F

1-28 0040 W>d' '

5-3-68

.,7 m - .

,_s m.._ _ _ . _ . . _ - _ _ _ ~ _ . . . - _ ._.

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Ts.ble 1-2 ~

Comparison of Desta,n Parameters

, 1(per station unit basis unless noted)

Oconee Russellville Nuclear Crystal River Plant Nuclear Station' . Turkey Point

-- Item Unit Unit 3 or 4 Unit 1; 2, or 3 Unit 3 or 4

'.1 Hydraulic ' and Themal Design Parameters

. Rated Heat Output (core), W t 2,452 2,452 2,452 6 2,0 71 ~6-Rated Heat Output (core), Btu /br 8,369 x lab 8,369 x 10 8,369 x 10 '7,157 x 10

. Maximum overpower, $ 14 14 14 12 -

System Pressure (ncainal), ~ psia . 2,200 2,200 2,200 2,250 Systemt Pressure (minimum steady' state), psia 2,150. 2,150 2,150 2.220 PJuer Distribution Factors Heat Generated in Fuel and Cladding, $ 97 3 VT 3 97 3 97.4

,(nuclear) 1.85 1.85 1.85 1 75- #

F nuclear) 3 15 3 15 3 15 3.12 Hot annel Factors <

Fq (nue, and mech.) 3.24 3 24 3 24 3.25 .

DNB Ratio at Rated Conditions 2.27(W-3) 2.27 (W-3) 2.27(W-3) 1.85(W-3).

1.60 (BAW-168) 1.60(BM-168) 1.60 (B M-168)

Minimum DNB Ratio at Design Overpower 173 (W-3) 173 (W-3) 1.73 (W-3) 1.30(W-3) 1 38 (BM-168) 1 38 (B M-168) 1.38 (Bm-168)

Coolant Flow 6 6 Total Flow Rate, lb/hr 1313x109 131 3 x 1 131 3 x 106 1co.6xlg Effective Flow Rate for Heat Transfer, Ib/hr 120.9 x 10P 120 9 x 1 120 9 x 10 93,5 , yn Effective Flow Area for Heat Transfer, ft'- 47 75 47 75 47.75 39.0 +

l 15 70 15 70 15 70 Average Velocity Along Fuel Rodg, ft/sec . 6 6 6 13 9 6 Average Mass Velocity, lb/hr-ft 2 53 x 10 2 53 x 10 2 53 x 10 2 33 x 10 Coolant Temperature, F '

Nominal Inlet 555 555 555 546.5 Maximum Inlet due to Instrumentation Error and Deadband 557 557 557 550 5-Average Rise in Vessel 47.8 47.8 47.8 54 .

Average Rise in Core 49.3 43 9 49 3 59 Average in Core 579 7 579 7 579 7 577 Average in Vessel 578.9 578.9 578.9 574 1 Nominal Outlet of Hot Channel 644.4 644.4 644.4 647 t

.AverageFilmCoefficient, Btu /hr-ft2 F 5,000 5,000 5,000 5,500

. Average Film Temperature Diffarence, F 31 31 31 30 -

Heat Transfer at. ICef, Power Active Heat Transfer Surface2Area, ft 2 48,578 48,578 48,578 42,460

'O AverageHeatFlux, Btu /hr-ft 167,620 167,620 167,620 164,200 Maximum Heat Flux, Btu /hr-ft2 1

Q Average Thermal Output, kw/ft 543 5.4,000 543,000 %3,000 533,600 i

@ Maximum Thermal output, kw/ft 5.4 5.4 53 .i 17 5 17 5 17 5

- *4 Maximum Clad Surface Temperature at Nominal.

17 3

! Pressure, F 654 654 6% 657-Fuel Central Temperature, F Maximum at 100% Power 4,160 4,160 4,160 4,Cr/0 Maximum at 114% overpower 4,400 4,400 4,400 4,270 Thermal Output, kw/ft at Maximum Overpower 19 9 19 9 19 9 19.4 2 . Core Mechanical Design Parameters

- Fuel Assemblies Design CRA can CRA can CRA can SCC canless Rod Pitch in. 0 558 0 558 0.558 0 563 m _ ,. _ r .__ _ _ . , _ _ . __m - , __

(

\ Tabli 1-2 c. ' d)

Oconee Russellville Nuclear Crystal River Plant Duclear Ste. tion Turkey Point Item Unit 'Jnit 3 or 4 Unit 1, 2, or 3 Unit 3 or 4 Overall Dimensions, in. 8.522 x 8.522 8.522 x 8.522 6.522 x 8.522 8.f26 x 8.426 4

Fuel Weight (as U0 2 ), lb 201,520 201,520 201,520 179,000 Total Weight, lb 283,200 283,200 283,200 226,200 Number of Grida per Assembly 8 8 8 8 Fuel Rods Number 36,816 36,816 36,816 32,028 Outside Diameter, in. 0.420 0.420 0.420 0.422 Diametral Cap, in. 0.006 0.006 0.006 0.0065 Clad Thickness, in. 0.026 0.026 0.026 0.0243 Clad Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy Fuel Pellets Material UO2 sintered UO2 sintered UO2 sintered UO2 sintered Density, $ of theroretical 95 95 95 94-93 Diameter, in. 0.362 0 362 0 362 0. 3ff>9 Length, in. 0.8 0.8 0.8 0.600 Control Rod Assemblies (CRA)

Net.* van Absorber 5% CJ-15% In-80% Ag 5% CJ-15% In-80% Ag $$ Cd-15% In-80% Ag 5% Cd-15% In-80% Ag r Clr.s lir.g Material 304 SG-cold worked 304 Es-cold worked 304 SS-cold worked 304 SS-cold worked Clad Thickness, in. 0.018 0.018 0.018 0.019 Number of Assemblies 69 69 69 41 Number of Control Rods per Assembly 16 16 16 20 Core Structure Core Ba:Tel ID/0D, in. 147/150 147/150 147/150 133.5/137 25 Thermal Shield ID/0D, in. 155/159 155/159 155/159 141.0/147.5 3 Preliminary Nuclear DesiFn Data Structural Characteristics Fuel Weight (as UO 2 ), lb 201,520 201,520 201,520 179,000 Clad 6".ght, lb 43,000 43,000 43,000 35,600 Core Diameter, in. (equivalent) 128.9 128 9 123.9 119 5 Core Ileight, in. (activ- fuel) 144 144 144 144 Reflector Thickness and Composition Top (water plus steel), in. 12 12 12 10 Bottom (water plus steel), in. 12 12 12 10 Side (water plus steel), In. 18 18 18 15 ll20/U (unit cell-cold) 2.'J7 2JJ7 2 97 3.48 Number of Fuel Assemblics 177 1,7 177 157 FuelRods/FuelAssembly 208 208 208 204 O Performance Characteristics C Loading Technique 3 region 3 region 3 reeton 3 region M Fuel Discharge Burnup, WD/MTU y' Average First Cycle 12,460 12,460 8,260 14,000 Equilibrium Core Average 28,200 28,200 28,200 27,000 FeedEnrichments,w/oU-235 nos. I and 3 Region 1 2.29 2.29 2.24 2.28 Region 2 2.64 2.64 2.47 2.43 Region 3 2 90 2 90 2.77 2 73 Equilibrium 2 94 2 94 3 09 --

Contaol Characteristics Effective Multiplication (beginning of life) Nos. 3 and 4 Hos . It3 No.2 Cold, No Power, Clean 1.302 1 302 1 31a 1.255 1.275 liot, No Power, Clean 1.247 1.247 1.258 1.201 1.225 liot, Rated Power, Xe and Gm Equilibrium 1.158 1.158 1.167 1.119 1.170

Tabla 1-2 (Cont's, Oconee Russellville Nuclear Crystal River Plant Nuclear Station Turkey Point Item Unit Unit 3 or 4 Unit 1, 2, or 3 Unit 3 or 4 Control Rod Assamblies Material 55 Cd-15% In-80% Ag 55 Cd-15% In-80% Ag 5% Cd-15% In-80% Ag 55 Cd-15% In-80% Ag Number of Assemblies 69 69 69 41 Number of Absorber Rods per CRA 16 16 16 20 Ak Total Rod Worth ( k),$ 10.0 10.0 10.0 70 Boron Concentrations To Shut Reactor Down With Rods Inserted (clean), cold / hot ppa 1290/1080 1290/1080 1290/1150 2300/2500 Ak Boron Worth (hot), $( k)/pps

• 1/100 1/100 1/100 1/130

• Ak Boron Worth (cold), $( k)/pIm 1/75 1/75 1/75 1/100

. Kinetic Characteristics Ak +1.0 x 10-k to +1.0 x 10 to +1.0 x 10 to +1.0 x 10 to Moderator Temperature Coefficient, ( k)/F k -3 0 x 10' -3.0 x 10'

-Ak -1.0 x IO~

-3.0x10fto -1.0 x 10 to -1.0 x 13 to -3.0x10'fto

-1.0 x 10' Moderator Pressure Coefficient, (T)/ psi +3.0x10:g +3.0 x 10'g +3 0 x 10'q Ak +1.0 x 10 to +1.0 x 10 to +1.0 x 10' to +3 0 x 10'3 to ModeratorVoidCoefficient,(k)/% void -3.0 x 10-3 3 0 x to-3 -3.0 x 10~3 +0.5

-2.0 xx10~

10~3 Ak -1.1 x 10-5 to 1,1 x 1o-5 1,1 x lo-5 to 1,o x 1o-5 to DopplerCoefficient,(1)/F -1.7 x 10~5 1 7 x 10'S -1.7 x 10'S -2.0 x 10'5 4 Principal Design Parameters of the Reactor Coolant System System Heat Output, W t 2,468 2,468 2,468 2,0CJT System Heat Output, Btu /hr 8,423 x 106 8,423 x 10 6 8,423 x 106 7.156 x 10 6

Operating Pressure, psig 2,185 2,185 2,185 2,235 Reactor Inlet Temperature, F 555 555 555 546.5 Reactor. Outlet Temperature, F 603 603 603 600.6 Number of Loops 2 2 2 3 Design Pressure, psig 2,500 2,500 2,500 2,485 Design Temperature, F 650 650 650 650 Hydrostatic Test Pressure (cold), psig 3,125 3.125 3.125 3,110 Coolant Volume, including pressurizer, ft3 11,800 11,800 11,800 9,800 Total Reactor Flow, spa 352,000 352,000 352,000 266,400 5 Reactor Coolant System Code Requiren.ents Reactor Vessel ASME III, Class A EME III, Class A ASME .II, Class A EME III, Class A Steam Generator Tube Side ASME III, Class A ASME III, Class A EME III, Class A ASME III, Class A C Shell Side EME III, Class A EME III, Class A ACME III, Class A A34E III, Class C Q Pressurizer ASME III, Class A ASME III, Class A EME III, Class A ASME III, Class A g Pressurizer Relief Tank ASME III, Class C EME III, Class C ASME III, Class C EME III, Class C Pressurizer Safety Valves ACME III EME III ASME III EME III

'M, Reactor Coolant Piping USASI B31.1 USMI B31.1 USASI B31.1 USASI B31.1 Reactor Coolant Pump Cesing ASME III, Class A ASME III, Class A A3tE III, Class A 6 Principal Design Parameters of the Neactor Vessel Material SA-533, Grade SA-533, Grade SA-533, Grade SA-302, Grade B g clad with B, clad with B, clad with B, clad with 10-8 Stain- 18-8 Stain- 18-8 Stain- Type 30 +! 4 less Steel less Steel less Steel austenitic SS

s T'bla 1-2 (Cont'd)

Oconee Russellville Nuclear Crystal River Plant Nuclear Station Turkey Point Item Unit 3 Unit Unit 1, 2, or 3 Unit 3 or ')

Design pressure, psig 2,500 2,500 2,500 2,485 Design Temperature, F ~ 650 650 650 650 Operating Pressure, psig 2,185 2,185 2,185 2,235 Inside Diameter of Shell, in. 171 171 171 155.5~

Outside Diameter Across Nozzles, in. 249 249 249 Overall Height of Vessel and Closure 240/235-3/8 Head (over CRD nozzles), ft.-in.

Minimum Clad Thickness, In.

39-0 1/8 39-0 1/8 39-0 1/8 41-0 5/32 lt 7 Principal Design Parameters of the Steam Generators Ntanber of Unita 2 2 2 3 Type Vertical, once- Vertical, once- Vertical, once- Verticsl, U-tube,.

through with in- through with in- through with in- with integral tegral superheater tegral super- tegral super- moisture separator heater heater

[ Tube Material Inconel Inconel Inconel Inconel Shell Material Carbon Steel Carbon Steel Carbon Steel Carbon Steel Tube Side Design Pressure, psig 2,500 2,500 2,500 2,485 Tube Side Design Temperature, F 650 650 650 650 65.66 x 106 Tube Side Design Flow, lb/hr 65.66 x 106 65.66 x 10" 33.53 x 106 Shell Side Design Pressure, psig 1,050 1,050 1,050 1,085 Shell Side Design Temperature, F 600 600 600 600 Operating Pressure, Tube Side, Nominal, psig 2,185 2,185 2,185 2,235 Operating Pressure, Shell Side, Maximum, psig 910 910 910 1,005 Maximum Moisture at Outlet at Rated Ioad, % 35 F superheat 35 F sulwrheat 35 F superheat Hydrostatic Test Pressure (tube side-cold), psig 1/4 3.125 3,125 3,125 3,110 8 Principal D3 sign Parameters of the Reactor Coolant Pumps Number of Units 4 4 4 3 Type Vertical, single Vertical, single Vertical, single Vertical, single stage stage stage stage Radial flow with C bottom suction and C Design Pressure, psig horizontal discharge 2,500 2,500 2,500 2,485 M Design Temperature, F 650 650 650 M Operating Pressure, Nominal, psig 650 2,185 2,185 2,185 2,235 Suction Temperature, F 555 555 555 546.5 Deaign Capoeity, gpm 88,000 88,000 88,000 88,800 Design Total Developed Head, ft 370 370 370 256 Hydrostatic Test Pressure (cold), psig 3,125 2,125 3.125 3,110 Motor Type A-C Induction, sin- A-C Induction, A-C Induction, A-C Induction, gle speed single speed single speed single speed MotorRating(nominal),hp 9,000 9,000 9,000 5,500

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

\ _'

T:ble 1-2 '(cont'd) ocones Russellville ' Nuclear Crystal River Plant . Nuclear Station Item. - Turkry Ibint Unit Unit 3 Unit- 1, 2, or 3 Unit 3 or 4

.9 . Principal Design Parameters of ttie Reactor Coolant Piping Material Carbon Steel clad Carbon Steel clad Carbon Steel clad Austenitic SS witt. SS with SS with SS Hot Iag (ID), in. 36 36 A Cold Leg (ID), in. .36 29 28 - 28 28 Between Pump and Steam Generator (ID), in. 28 27-1/2

. 28 28 31 lo - Reactor Building System Parameters

-Type Steel-lined., pre- Steel-lined, pre - Steel-lined, pre. Steel-lined, pre-stressed, post- . stressed, post- stressed, post- stressed, post-tensioned con- tensioned con- tensioned con- tensioned con-crete, vertical crete, vertical crete, vertical crete, vertical' cylinder with cylinder with cylinder with . cylinder with flat bottom and flat bottom and^ flat bottom and shallow domed flat bot. tom and shallow domed shallow domed shallow domed roof roof roof roor Design Parameters Inside Diameter, ft 116

' 130 116 116

'lieight, ft 1 206

" 187 206 Free Volume, ft3. 1,900,000 177 2,000,000 1,900,000- 1,550,000 Reference Incident Pressure, psig 59 55 59 56 Reference Incident Energy (El), Btu 306,700,000 306,700,000 306,700,000 Energy Required to Produce Incident Pressure 272,000,000

~(E2 ), Btu 341,806,000 335,200,000 341,806,000 Ratio: El/E2 300,000,000 o.897 0.915 0.897 0.907 Ratio: (E2 - El)/El O.115 0.093 0.115 0.103 Concrete _ Thickness, ft Vertical Wall 3-3/4 3-1/2 Dome 3-3/4 3-1/2 Reactor Building Leak Prevention 3-1/4 3 3-1/4 3 and Mitigation Leak-tight penetra- Leak-tight pene- Icak-tight penetra- Leak-tight penetra-tions and cor. tin- trations and tions and contin- tions and contin-uous steel liner. continuous uous steel liner. uous steel liner.

Automatic isola- steel liner. Automatie isola- Automatic isola-tion where re- Automatic iso- tion where re- tion where re-quired lation where quired quired required O

C 4

UI Gaseous Effluent Purge Discharge vent Discharge vent Discharge vent Through particulate above top of above top of above top of filters and moni ~

Reactor Building Reactor Building Rcactor Building tors. Part of the

(-200 ft above (-200 ft above (-200 ft above main exhaust .

grade) grade) grade) system

r T-ble 1-2 (Cont'd)

Oconee Item-Russellville Nuclear Crystal River Plant Nuclear Station Turkey Point Unit Unit 3 or 2+ Unit 1, 2, or 3 Unit 3 or 4-11 Engineered Safeguards Safety Injection System No. of High Head Ptaapa 3 3 3 3 No. of Iow Head Pumps g Reactor Building Air Coolers g 3 2 No. of IA2its 3 4 Air Flow Cap'y. Each, at Accident 3 3 Condition, cfin 54,000 54,000 54,000 80,000 Core Flooding System No. of Tanka 2 2 2 Total Volume, ft3 2,820 2,820 3

Postaccident Filters e 2,820 3,600 No. of Units One None Hone inside Reactor None Air Flow Cap'y. Each, at Accident Building. Leak-Condition, etia Type age from penetra-tions collected, filtered and dis-charged through station vent

, Filtration Reduction RateggR/V, (nf - 0.9 per pass), hr Reactor Building Spray No. of Pumps 2 2 2 2 Including Sodium Thiosulfate Injection YeS res No Emergency Power No 3 Generator Units, No. 2 3 2 2 for both Units Type Diesel Engineered Safeguards Operable from Diesel Not comparable

• Diesel Emergency Power Source (minimum) All engineered All engineered All engineered 1 liigh Head Safety safeguards e- safeguards e-quigment is safeguards e- Injection (SI) quipment is quipment is Pump capable of being capable of cal.oble of being 1 Iow Head SI Ptanp operated from being opera- operated from onsite e;.nergen- 3 Containment Air ted from on- onsite emergency recirculation units cy power site emergency power 1 Containment spray power pump 1 Service water pump

• Two 70 KW hydro-O electric units with C one overhead and

$% one underground O feeder. Also, one of three 44 M7A gas turbine units loca-ted 30 miles distant dedicsted solely for backup emergency

. Power

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