ML20030A516

Proposed Tech Specs 1.0-7.0

ML20030A516
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	01/15/1962
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8101090828
Download: ML20030A516 (50)

Text

,

PROPOSED TECHNICAL SPECIFICATIONS Big R ock P oint Nu clear Plant t

7 January 15, 1962 l

6 Consumers Power Company FAVo9683/

i TABLE OF CONTENTS PAGE 1.0 Introduction 1

1.1 Scope 1

1.2 Definitions 1

2.0 Site 2

2.1 Location 2

2.2 Boundaries 2

23 Principal Activities 2

30 Eeactor Containment 2

31 Containment vessel Design Parameters 2

32 Containment Vessel Dimensions 3

33 Construction 3

3.4 Penetrations 3

35 Post-Incident Spray System 5

36 Containment Eequirements 5

37 Containment Leakage Testing 6

h.0 Eeactor and Power Systems Equipment 6

4.1 Eeactor System Equipment 6

4.2 Power System Equipment and Associated Facilities 10 50 Eeactor Core and Controls 13 51 Design Features 13 52 Principal Calculated Thermal, Hydraulic and Nuclear Characteristics 16 53 Characteristics of Research and Development Puel 20 5.h Principal Core Operating Limitations 21 6.0 Plant Safety and Monitoring Systems 25 6.1 Peactor Safety System 25 6.2 Control Rod Withdrawal Permissive System 30 63 Refueling Operation Interlock System 30 6.4 Plant Monitoring Systems 31 70 OIerating Procedures 35 71 Basic Orerating Principles 35 72 Procedural Safeguards 36 73 Initial Core Loading and Critical Tests 37 74 Power Operation Test Program 39 75 Normal Operation 41 76 Research and Development Program 43 f

77 Refueling Operation 44 7.8 Maintenance 45 79 Operational Testing of Nuclear Safeguard Systems 45 1

t f

1.0 INTRODUCTION

1.1 SCOPE 1.1.1 These technical specifications set forth the principal design features and operating limits and requirements which have an effect on the safety of operation of the Big Rock Point Nuclear Plant (the plant).

1.1.d In the order of presentation, Section 2.0 concerns the site, Sections 3 0 through 6.0 cover the plant and its systems, and Section 7 0 presents procedures for plant start-up, procedures for normal and emergency operation of the plant, and procedures for Phase I of the research and development program. This section also includes administrative and procedural safeguards to the ex-tent that these have a potential effect on nuclear safety.

1.1 3 The dimensions given in these specifications of the plant design features are subject to normal manufacturing tolerances. The operating limits and requirements which are significant from the standpoint of nuclear safety are specified with each of the plant's systems.

1.2 DEFINITIONS 1.2.1 Power Operation - is any operation other than cold shutdown with the reactor vessel closures bolted in place and when reactor criticality is possible.

1.2.2 Fefueling OTeration - is any oyeration with any of the reactor vessel closures open during which either core alterations are being made, or other operations which might increase the reactivity of the core are in progress.

1.2 3 Cold Shutdown - is a reactor condition involving either no fuel in the reactor or a condition meeting the following requirements:

(a) The control rods nre fully inserted in the core, and their withdrawal circuit is locked by neans of the mode selector svitch in the shutdown position to prevent withdrawal. The key to the mode selector switch must be in the possession 6

of the Shift Supervisor.

I (b) The reactor coolant system is at atmospheric pressure.

I l

o i

'.I'l

'I

2 (c) The core shutdown reactivity control margin requirement has been demonstrated in accordance with Section 5.4.2 (b).

2.0 SITE

}

2.1 LOCATION - is in Charlevoix County, Michigan, about 4 miles northeast of Charlevoix, Michigan, and about 11 miles vest of Petoskey, Michigan.

2.2 BOUNDARIES - surround sbout 600 acres, with the nearest landside property line about 2c80 feet, and the nearest shoreline proIerty line about 200 feet from the plant's containment vessel.

23 PRINCIPAL ACTIVITIES - are those associated with the operation of the Big Rock Point nuclear Plant and may include the other activities directly connected with the generation, transmission and distribution of electric energy.

30 REACTOR CONTAIIRCTP Peactor containment consists of a spherical steel vessel, herein-after referred to as containment vessel, sphere or enclosure. The reactor, recirculation piping, pumps, steam drum, fuel pool, and equipment for removal of shutdown heat are among the equiprent in-cluded in the containment vessel.

31 C0!TTAU2ETT VESSEL DESIGN PAPM ETERS Design pressure, internal, psia 41 7 Design temIerature rise, F 190 (coincident with design internal pressure)

Design maximum temperature, F 235 Wind load ASA Std A58.1 Without snow load (Basic vind pressure =

30 psf)

With snow load 60 mph Snow load ASA Std A58.1 (Max = 40 psf at top)

Lateral seismic acceleration, percent 5

of gravity (coincident with dead load and snow load only)

(

l i

Permissible air leakage rate at 417 psia 0.5 at ambient temperature, percent Ier day of free volume (including all penetra-tions) 32 CONTAII27DT VES3FL DIMENSIONS Diameter, feet 130 Heicht above crade, feet 103 Approximate free volute, cubic feet 9.4 x 105 33 coNSTRUCTIO" - The principal taterial of construction is SA-201 Grade B, firebox steel produced to SA-303 specifications.

Design and constructiun is in accordance with ASME Boiler and Pressure Vessel Code, Sections II, VIII and IX, as modified by the appli-cable nuclear code cases.

34 PEUETFATIONS 3.h.1 Design Features (a) Total nunber 100 (b) Total nudber of access air locks 3

(c) Size of nozzle Ienetrations:

Maximum diameter, inches 24 Minimum diameter, inches 3/h (d) Types of penetrations:

Fire size same as nozzle size Direct weld Pipe size smaller than nozzle Weld to nozzle size cap hole Honrigid renetration connections Pellows seal All electrical conductor pene-Hermetically sealed trations connector or com-pound filled nipple 3.4.2 Methods of closure g

(a) p l

Access air locks have two in-series gasketed doors, which are interlocked to insure that at least one door is locked closed at all times when containment integrity is required.

k

-b-I (b)

Lines open to the free volume of the containment vessel have two valves in series, at least one of which closes automatically whenever necessary to prevent outward flow in the event of an accident. Except for check valves, both valves can be closed by tanual initiation from either the control room or from other stations that would be tenable after an accident.

(c) Lines open to the reactor vessel or any portior of the reactor recirculating loop are treated in the manner described in the previous paragraph, with the added feature that the two valves are on opposite sides of the containment shell.

(d) Lines normally closed have only a single valve.

A lock, interlock, or operating rules protect this valve from being oIened during operating conditions which require containment integrity.

(e)

Certain lines enter and leave the sphere without any oren-ings to the containment free volume.

Others leave and return to the sphere without any orenings to the atmos-phere.

Such lines do not require isolation valves.

(f)

The two 24-inch ventilation orenings, one for supply and one for exhaust, are designed to close within six seconds after any scram signal.

In order to prevent the possi-bility of excessive external pressure on the sphere due to atmospheric changes, the two valv<:s in the ventilation supply line are automatically oIened whenever the differ-ential pressure exceeds 1 psi, overriding all other signals.

The valves reclose when the internal pressure is still slightly below atmospheric.

3.43 Operatinc Fequirements (a)

Normally open lines carrying fluids out of the sphere are closed automatically upon a signal indicating high sp:,ere pressure or low water level in the reactor vessel.

These automatic isolation valves also close upon instrument air or power failure, and upon manual trip from the control room.

(b)

)

Normally open lines, carrying fluida into the sphere, are each equipped with a check valve to prevent backflow upon loss of inward propellent force.

In addition, these lines can be secured by manually-operated gate valves or by k

air-operated control valves. The latter close upon air or i

power failure, with exception of the supply line to the con-trol rod drive hydraulic system. Valves in this control ro1 hydraulic system line fail open to insure continuous water supply, and backup isolation is provided by integral valves in the control rod drive pumps.

35 POST-INCIDENT SPPr! GYSTBI Containment effectiveness is supplemented by the enclosure post-incident splay system in the event of an accident involving loss

+

of coolant from a primary syctem rupture.

3 5.1 resign Features (a) Number of sets of spray nozzles 2

(b) Capacity of sprays, spm Ier set h00 (c) I;ozzle pressure, psia 100 (d) System actuation 1 set automatic, 1 backup set manual (e) Signal used to actuate High sphere pressure (f) Signal trip setting 2 psi above atmospheric 3 5.2 overatinc Pequirements (a) Automatic operation of this spray system involves a 15 minute time delay after cystem actuation, to allow the oIerator to override a possible spurious actuation.

This time delay feature may be manually overridden to actuate the spray system prior to the expiration of the 15 minute re riod.

e (b) 4 Water addition to the containment vessel must be manuclly stopIed before the accumulated water level reaches an ele-vation of 596 feet.

l (c)

The automatic controls of the spray system will be functionally tested at least every 15 months.

3.6 CorAIN72:TP HEQUIRE!E!:TS p

}

Containment integrity shall be maintained during power operation, refueling operation and cold shutdown conditions except as specified by a system of pre:edures and controls to be established for occa-sions when containment must be breached during cold shutdown con-dition.

E

37 CcIcAIICE?C LEAKAGE TESTIIIG To assure that the permissible leakege rate for the containment i

vessel and its penetrations is not exceeded during the life of the plant, a procedure covering future inspection and rnethod j

of testing vill be submitted for Commission approval prior to the elapse of 18 months followir.g issuance of the provisional operating license.

4.0 FFACTOR AI!D POWER SYSTEMS EQUIP!EIC 4.1 PEACTOR SYSTEM ECUIPIEIR

!~

)

The reactor system equipment j

the steam drum, the safety reJ ^f' consists of the reactor vessel, valves, the reactor recir-culating pumps, the shutdown cooling r.ystem, the emergency j

cooling systems and the interconnecting piping and valves.

h.l.1 Feactor Vetsel l

(a) rusign Features i

Length, over-all, feet 30 Inside diameter, inches 106 Wall thickness, excluding cladding, inches 5-1/h Cladding thickne 2 minimum, in" i

5/32 l

Design prest

.e, psia 1715 Design temperature, F

650 Approximate initial nil ductility transition temperature, F 10 bl

-T-(b)

Principal Materials of Construction Component Material Vessel shell and heads Specification Steel Flanges and nozzles ASTM SA-302 Grade B Steel l

Cladding ASTM SA-336 Stainless steel Head studs Types 308 and 309 Steel Head nuts ASTM-A193, AISI 4340 Steel ASTM SA-194, AISI 4340 (c)

Reactor Vessel Penetrations Nozzles Number Diameter, Inches

_ Location Coolant water inlets 2

Steam-vater mixture outlets Bottom 20 6

14 Shutdown heat exchanger outlet Shell 1

Access ports in top head 6

Shell 3

Control rod drive penetrations 10 Head 32 Liquid poison inlet 4

Bottom 1

Emergency core spray inlet 3

Botto=

1 Vessel vents 3

Shell 2

In-core flux monitor penetrations 3

Head 8

Instrument nozzles 2

Bottom 4

Seal leak monitor 3

Shell 1

1/2 Shell flange (d) _ Operating Eequirements shall be limited to approximately 100The rate of cha r vessel vall vessel pressurization in excess of 100 psig shall F per hour. Reactor allowed to occur at temperatures below the calculat d not be ductility transition temperature.

e nil temperature calculations shall be made at lea tUil ductility transition 4.1.2 s

once each year.

Main Coolant Eecirculation System vessel, the steam drum, the reactor recirculatin 1

e reactor interconnecting piping and valves, and the safety mps, the i

relief valves.

k PDDR BRBlHR

F

-8.

(a)

Design Features Number Number of recirculating loo Approximate intern lof recirculating pu ps mps per loop isolation v lvexcluding reactor 2

a v lume o

core of system, 1

Steam Drum:

es, cu ftand internals, a

1 to 3830 Length, ov Inside diameterer-all, feet Wall thickness, inches 40 Cladding thicknes, excluding cladding 78 Design pressure s, minimum, inches, inches

, psia 4-3/8 Design temperatur 5/32 Eecirculating Pump e,

F Type s:

1700 650 Rating Single Seals

'7,000 stage, centrifugal gpm @ 78 Pump suction Mechanical, 3 in-vefeet of head op:

ries Pump dischargev lve, size, inche a

a s

v lv, size, Pump discharge butt e

inches 2h Pump discharge erfly "alv, size inches e

v lv 20 a

Pump discharge e bypass yalve

, inches 20

, aize, per minute valve Safety Relief Valv ope ing rate, ir)ches n

5 Number es:

5 Maximum rupture disc, psiasettir6 of firrt v l a

Sequentio' pressu ve including 6

rerE uing valvesre increment settin Minimum

, psi 1700 ca g of setting),pacity per v lve a

Valv pounds per hour (1202 psia 10 e orifice area, sq in.

2 36 x 10 3 97o, m

t

$.I Reactor Power Operatio i

Coolant material n Cooling:

Type of cooling system Demineralized water g

l Syctem pressurization Forced recirculation Minimum loops operating c Boiling water currently (or equivalent) on-Number of passes throu h Flov direction through g core 1

Eeactor Shutdown Coolin core 1

Upward Design pressure, psia g:

Design temperature,UF Number pumps 315 Number heat exchangers 425 Heat removal cap <1 city pe 2

loop, Btu /hr r

2 Reactor Emergency Cooli Emergency Condenser:

20 x 10 ng:

Type Number of tube bundles Minimum capacity per tub Tank with tube bundles bundle, Btu /hr e

2 Minitum cooling time av from water ailable 16 x 106 Number storage, hours syctems providing makeup supply 4

Design external pressur psia 2

e of tank, Design pressure of tube bundles, 41,7 psia Minimum time to put syst operation following signalem in full 1700 Core Spray System:

, seconds Type 30

\\

\\

Sparger Capacity of sprays nozzlering with spray

, gpm 400 L

r j!.

t, Nozzle pressure, psia 115 Core Spray System Recirculation:

Number pumps 2

Nudber heat exchanger 1

Heat removal capacity, Btu /hr @ 28.4 F 6

l log mean temperature difference 8 x 10 (b) Operating Requirements t

E:

A minimum of one reactor recirculating loop or its equivalent shall be used during all reactor power operations when reactor y

power level is above 1.0 Mwt. The safety relief velves shall p

be set appropriately for all planned reactor operating pres-sures so that the alloveble pressure of 1870 psia (1700 plus 10%) in the nuclear steam supply system is not exceeded. The emergency condenser and the core spray system shall be operable and ready for service at all times during power operation. The core spray system and shutdown cooling system shall be operable and ready for service during refueling operations.

4.2 POWER SYSTEM EQUIPMENT AND ASSOCIATED FACILITIES 4.2.1 Electrical System (a) Auxiliary Power i

The auxiliary power system is the normal source of power to the plant under operating and shutdown conditions. Auxiliary power is obtained from either the main generator or from the trans-r

[

mission system through the station service transformer connected to the 2400 volt switchgear bus. Each of two 480 volt systems obtains power from a separate transformer connected to the 2400 volt bus.

The reactor safety system and related circuits are fed from J

three 120 volt AC buses. Each of two buses is supplied from a different h80 volt system through its own motor-generator 1

set.

Each motor-generator is equipped with a flywheel to 7

sustain operation during momentary power system disturbances.

The third bus is supplied from the 125 volt DC system through a static inverter. The 125 volt DC cattery system also fur-nishes power for other critical services including:

i Liquid poison system controls Motor-operated automatic enclosure isolation valves Enclosure ventilation system isolation valves j

Emergency condenser drain valves i

?

5

i l

i '

(b) Emergency Power Emergency power is provided by' a diesel-generator which is automatically started on loss of auxiliary power. As soon as the diesel-generator reaches its rated potential of 480 volts, power is available automatically to service the following equipment:

Containment vessel access air locks Fire protection system electrically driven pump Emergency lighting Instrumentation including monitoring systems Other electrical services may be energized manually as re-qui red.

h.2.2 Main Condenser (a) Design Features Type Radial flow surface condenser with deaerating hot vell Condensing surface area, square feet 27,500 Design condensing pressure, inches Hg absolute 15

. l Condensing capacity, pounds per hour @ 15 inches Hg absolute 460,000 Condensing capacity during full f

load rejection, pounds per hour 948,000 F

Air ejector capacity - 10 cubic feet per minute of air plus

[

1.1 pounds per hour of hydrogen plus i

8 3 pounds per hour of oxygen (b) Operating Requirements The following condenser pressure trips vill be orerative during reactor power operations when system pressure is f

i 350 psia or higher:

i f

f 5

-f

. t-

I I

~ Annunciate, inches Hg absolute 50I05 Heactor scram,. inches Hg absolute

[8.0't05

-Turbine trip and bypass valve closure, inches Hg absolute 10.0 t 0 5 4.2 3 Turbine Bypass Control System (a) ~ Design Features-l

-Flow capacity at 1015 psia, pounds per hour

-739,000 Flow capacity at 1465 psia, pounds per hour 963,000 4.2.4 Condensat'e and Feed-Water System The condensate and feed-water system will operate during power operation to perform as follows:

(a) Pressurize and return feed water to the stea drum.

(b)- Provide means for'demineralizer cleanup of the full condensate flow.

-(c) Provide.means for regenerative heating. of feed water by -

j extracting steam from various points off the turbine to j

supply the feed-water heaters.

j l

4.2 5 Rettetor Cooling Water System

]

The reactor cooling water system is a closed cooling loop utilizing demineralized water to remove heat from the ' following.

pieces of equipment:

Reactor shield cooling ~ panels I

Reactor cleanup nonregenerative heat exchanger Reactor shutdown heat exchanger Fuel pit cooling water heat exchanger Miscellaneous sample coolers l

Reactor recirculating pump coolers 4.2.6 Fire Protection System.

In addition to furnishing vater for conventional plant fire-fightirc, equipment, the fire protection system shall furnish water as follows:

ty-

i i Ie i

Core spray cooling system Enclosure post-incident spray system Backup for service water system to the sphere The fire protection system shall be operable and ready for service during power operat' on and during refueling operation.

50 REACTOR CORE AND CONTROLS e

The reactor core and controls consist of all the components of the f

reactor core including the reactor reactivity control system, and o

the liquid poison system.

L 51 DESIGN FEATURFS

(

The dimensions for these desi6n features are for the cold con-dition and are subject to normal manufacturing tolerances.

~

5 1.1 Principal Core Materials y

Fuel UO2 Moderator and reflector Light water Structural components including fuel cladding Stainless steel Flow channels Zircaloy and/or stainless steel Control rodo B C filled stainless steel gtubes in cruciform shaped r

g sheaths c

5 1.2 Control Rods Number of control rods 32 Poison material in rods B C powder in SS tubes g

Pitch (square array), inches 10.466 Active length, inches 68 Shape Cruciform

• \\

Width, inches 11-1/2

{

Blade thickness, inches 5/16 L

Sheath thickness, inches 1/16 Number B C filled, SS tubes g

f per control rod 116 4

e

, i

.i SS tube, B C filled, outer g

diameter, inches 0.175 SS tube, B C filled, vall g

?

thickness, inches 0.020 I

513 control Rod Drives i-Type Hydraulic, locking piston Normal stroke length, inches 68 Length between locking notches, inches 30 Normal drive velocity, inches per second 30 Principal materials of con-Type 304 S3,17 4 Ph (M1100)SS, struction Inconel X (springs)

Graphitar 14 (seals) 5 1.4 Liquid Poison System Material Sodium pentaborate solution Available quantity of solution, gallons 850 Minimum concentration of sodium pentaborate in solution, vt percent 19 Initial injection gas pressure, psia 2000 System actuation Remote manual Type of injection valves Explosive, electrically actuated 515 Initial core Puel composition I

The data presented in this section consist of the design features of the fuel which will make up the physical composition of the

}

initial core. These data do not represent specific limitations l

to design and operation of subsequent core loadings for Phase I of research and development program testing. The specific limita-g-

tions which vill be adhered to in the design and operation of the l_

initial and subsequent core loadings are given in Section 5.4.

I l-i

-1

4 235 l

(a)

Initial Fuel Enrichment, approx vt percent U 32 (b) General Core Data Approximate number of fuel buriles in core 56 Total weight UO f r 56 bundles, pounds 19,300 2

Moderator to fuel volume ratio 27 Equivalent core diameter, inches 62 5 (c) Fuel Bundle Geometry, fuel rod array 12 x 12 Standard fuel rods per bundle 132 Spacers per bundle 3

Special fuel rods per bundle (4 special fuel rods, at bundle corners, are segmented) 12 (d) Fuel Rod Cladding Material Stainless steel Standard rod tube vall, inches 0.019 Special rod tube vall, inches 0.031 (e) Fuel Rod Standard rod diameter, inches 0 388 Standard fuel pellet diameter, inches 0 345 Special rod diameter, inches 0 350 Special fuel pellet diameter, inches 0.283 Standard rod active length, inches 70 Corner rod active length, inches (made up of 4 segments) 59 l

Standard rod gas plenum length, inches 2-3/4 Corner rod gas plenum length, inctes j

(total of 4 segments) 4-5/8 (f) Channels Number of stainless steel and/or zircaloy 88 I

t

> Wall thickness:

Stainless Insidee loy, incheste l, inche Zir e

a c

s width:

Stainless 52 Zir PRINCIPALc loy, inchessteel, inches 0 075 a

o, loo E

CHARACTERISTICALCULATED THEFl'AL, HYDRA 6 57 CS The data pr analys e ted ULIC, AND 6.%

es es 11citatio n

of ns to designthe initialin thisNUCLEAR vill be sectio in Se tion 5 4 adhe cor n are e de ig rd c

e diffe or to in the de oper tio.n, butnominal v lue s

3 re t a

n prov l Any cubs que from.

n do a

a sign The and not r prs bas d affe t for those e

indic tion thoper tionspe ific limit e

e c

c ontinued opec lculated vill c

es a

nt et on tio the n

a ns ab cpe ific ility of as a

c 521 of the ations ratio, pr Principal specified the lic n

c ens ovidedn t require fu at actu l v lu o

which cor in Se tion 5 h e

a are the InitialCalculated a

giv ee to that su h rthe es en co r AEC ure mply with opedeviat'o c

Cor Ther (a) ap-e Lo ding mal and i

a Co ns do Hydr r ting limit r

a e ps not er (b) Pe king fa tat r ted stea aulic Chara t a-s a

a c

Ov s (to he m flow, Mvt eristic c or s of er-all at r t applied local) a pow to heat 15 7 ed erpow r (includes flux):

e Ov (c)

Total (produer (ste dy and gross a

and tr ansient effe t )

\\

He t ct) a c

32 Flux s

and "uel Center Tempe 1.27 Gt He Av r tur 4,1 a

er 2a a6e oft ) t Flux, 157 !& @

e Ma imu t

x 157 m@

smper tur,

110,000 127% ma @

a Wt Maxi e

F Temper tur 352,000 157!&

a t

e, UF 5 70 r

44,000 1,1k 7

0 680 3,020 700 3,650 w

-_17 -

s

.(d) Burnout ratio, minimum at overpower (a) 2.0 (e) Maximum fuel cladding stress, psi 48,200 (f) Average core power density at 157 Mwt, KW/L 45 (g) Stability margin at 157 Mwt,1050 psia, degrees phase margin 68 (h) Total recirculating flow rate:

Normal for 157 Mwt, 1050 psia' 6

pounds per hour 12 x 10 Maximum, cold, atmospheric pressure' 6

pounds per hour 18 x 10 (1) Reactor core flr rate, percent of total recirculating flou rate 93 (j) Feactor core pressure drop:

Normal for 157 Mwt, 1050 psia, psi 11 Maximum, cold conditions, psi 15 (k) Eeactor core maximum exit bulk temperature, F

550 (1) Core avera6e void content at overpower,

-percent 23 (m) Minimum core inlet pressure, psia 1050 (a) Based on correlatio given in " Recommended Curves of Burnout Limit for Design and Operation of Boiling Water Reactors," by E. Janssen and S. Levy, November 14, 1961, General Electric Memo Report.

(To be issued later as part of j

GEAP 3892.) This correlation supersedes the previous correlation used in support of the Hazards Summary Report.

5 2.2 Principal Calculated Nuclear Characteristics of the Initial Core (a) Temperature and Void Coefficients I

he following temperature and void coeffic.ients have been calculated for minimum criticul size loadings consisting i

F k~

_- 1

t p

L}-

- either of all steel channels or all zircaloy channels. Sim-

.1 plifying assumptions applied which tend to make the values larger (more positive) are:

iej '

The negative effect_ caused by the-presence' of control rods was not included in this calculation.

The "end of cycle" data correspond to erid of cycle

)

' reactivit'y conditions in which the neutron leakap is detemined by the geometric buckling of the 56 fuel bundle core. Thus, the indicated positive effect of decreased leakage is larger th'an could be encountered in practice, because enough reactivity must be main--

tained to compensate for voids at rated power. Addition of identical reload fuel would ~ make the coefficients more negative.

The fuel temperature was assumed to be 68 F in all these cases.

Moderator Temperature Coefficient (Ak rf/k pp per F) e e

68 F 320 F 550 F Steel channels:

-Clean initial core

+0.4 x 10

-0 7 x 10

-1 3 x 10-End of cycle

+0.6 x 10

+0.1 x 10-

-0 7 x 10 Zirealoy channels:

Clean initial core

-0.l'x 10

-1.4 x 10

-2 5 x-10

-4 End of cycle

+0 3 x 10-0.0

-0.6 x 10 Void Coefficient (Akepp/kepp per Unit Void Within the Channel) 680 0

F 550 F 1

0% Void 20% to 50% Void Steel channels.

j Clean initial core

-0.18

- -0.17 End of cycle

-0.09

-0.14 3

l Zircaloy channels:

t-Clean initial core

-0.24

-0.23

-i End of~ cycle

-0.05

-0.12 I

The isotopic composition assumed for the above temperature and void coefficient calculations corresponds to zero exposure.

t

'[

(.

I.

t

,,y 2ew-y.c.---

. rem.

-e, c

y_---,..--r

19-(b)

UO 2 (Dop le ) C p

Fu l r

e Te o fficie t (O e

mper tur a

n e

k 68 F

g /k 1

g pe 323 Mode F) r F

r tor 1

a 323 68 F

F, Oy' Voi TFe 550 ds Ak F, Ofo se m /k fu l c

550 of e ficie ts F, 20fu Voids m pe e

te (c) mpe r

n F

r tur a

Re apply to k Voids

-1.4 atc ivi ove 7 x 10 -9 e

ty Bala the latti r

-10 3 x 10 -5 gass nce c.

u ing a

-1 15 x 10 ~3 m

Te e

mperatur giv Voids e

en co n tant s

Xenon and Fu l e

sama iu Total Adepletio r

m n

Ak (d) k ad eff n

r m

Multipl equir d ane e

uv ic tio e ing o,op5 r

a 0 01 n Fa to. C 9

c All r

o,o31 r ds o

c 28 ste l out, n xditio 0,127 e

o 28 cha ste l 0 20 n

n e

ste l rd cha nels, 32 cha 2

e o

nn ls e

ot nn ls, 31 r ds in u

e o

23 ste l e

Mo t n ls, 31 r ds in, c o

o k

rd cha ot n

epp (e) u e

~

e tr l 1 20 n

n a

2 Pr b r ds in, o o,943 o

o ab le Ma im uter x

Col u

m Co tr l 0 96 d

n Hot (550 Bod o

5 Hot (550F) n Worth 0 96 o

3 viod

\\

Ma im F) 20 v i x

p u

wi m Fe od on ef lh u tivity Add f

.ak thdr c

eff ep /k awal p

in hep at the itio 0 039 of of 3

ep sec c

the r

n Eate (ba 0.04 od c

od n

per ma imu x

wo 2

rth sec sd 00 ond)ak withdr0 04 m

e 30 2

ef /k aw l f

a r te ic a

epp per 00052 4

V"

. 20 l

.(g) Calculated Worth of Liquid Poison System

) --

Ak

/

68gfr kerr, normal. vater level in reactor and F, 2000 ppm boron

-0.25 j'i' ok rr/k rr, normal water level in reactor and e

e 550 F,1300 -ppm boron

-0.16 Maximum time to bring reactor 'suberitical, hot, minutes 5

~.

t' 53 CHAP.ACTERISTICS OF RESEARCH AND DEVELOPENT FUEL 531 Design Features The-Phase I research and development fuel vill consist of two groups. One group will be instrumented fuel bundles, whose design features are substantially those of the initial core fuel bundles described in Section 515 The design features-of the other group of development fuel bundles vill be substantially as tabulated below:

3 Enrichment, vt percent U 27 Geometry, fuel rod array 11 x 11 Standard fuel rods per bundle 109 i

Special fuel rods per bundle 12 Standard fuel rod diameter, inches 0.400 Special fuel rod diameter, inches 0 320 i

Moderator to fuel volume ratio 2.8 Standard fuel rod pitch, inches 0 588 Special fuel rod pitch, ir.ches 0 503 Fuel rod length, inches 70 Clad th'.ekness, SS, inches 0.010 Spacers per bundle 4

532-Cjeulated Thermal, Hydraulic, and Nuclear Characteristics The range of tests to be conducted in the reactor vill embrace a variety of thermal hydraulic conditions that in every case vill first be analyzed.to assure compliance with the operating limita-tions specified in Section 5.4.

The design of Phase I development fuel is such as to give nuclear characteristics equivalent to the.

j initial core.

t-I I

l

.l-The following data are for a representative test to be conducted during Fhase I of the development program. The limiting parameter in this case is the 1.5 burnout ratio, with reactor power below the specified operating limit (157 Mat). This test will be conducted only to the extent that the stability criterion of measured peak-to-peak flux amplitude is less than 25 percent. The anticipated worst conditions with regard to stability have been shown by cal-culation to have substantial phase margins indicating that the core operating limitations will not be exceeded. Other Phase I develop-ment tests, though differing in specific variables which may be limiting, will comply with the operating limitations specified in Section 5.4.

Feaking factors (to be applied to heat flux)

Over-all at operating power 32 Overpower 1.27 Total (product) 4.1 Eeactor pressure, psia 1500 Core size (including eight development fuel bundles), number of fuel bundles 56 Burnout ratio 15 Maximumheatflux,atoverpower, Btu /hr-ft 475,000 Maximum fuel cladding stress, psi 47,900 Feactor thermal power, Mwt 144 Average core power density, }N/L 41 Maximum fuel center temperature, F 3,500 54 PRINCIPAL COPI OPERATIU3 LIMITATIONS 1

5.h.1 Peactor Power Level (a) Refueling The reactor power shall be limited to 1.0 Mat, exclusive of core decay heat, during operations with the reactor vessel closures open.

(b) Reactor Operation l

The reactor operation will be so limited as to be consistent I

with the most conservative of the following:

Minimum core burnout ratio at overpower 15 Maximum heat flux at overpower, Btu /hr-ft Sh0,000 Maximum fuel clad stress (at limiting con-j dition: end of life, overpower), percent of yield stress 90 i

I

e

. 'C 22 -

4,

'I Stability' criterion: Maximum measured peaksto-peak flux amplitude, (measured t-opposite core center. line) percent of i.

average operating flux 25

~

Maximum stesdy state power level, Mvt-157 Maximum value of. core power density, total core power divided-by total core volume, KW/L 45 Reactor Operating Pressure:

' Maximum,.psig 1485 Minimum, psig

'800

~

[

Minimum recirculation flow rate, Ib/hr 6 x 10 l

5.4.2 Control Rod system

~

(a)' Control ' Rod Perfomance The following limits apply to any control rod which can be withdrawn. It is permissible to tag and valve out the l

hydraulic drive water to a full'- inserted control rod which is defective or does not meet Lnese limits provided the

. remaining rods de meet the l'aits.

The following limits shall've checked at least every 15 o

months.

(1)

Minimum withdrawal time shall be greater than 23 seconds.

- ( ii).

Maximum scram tine from system trip to 90. percent.

insertion shall not exceed 2 5 seconds.

.(b) Core Shutdown Margin Verification The-k pp of the core shall be maintained at less than unity e

vith the most. valuable reactivity-worth control rod com-pletely withdrawn from_the cold, xenon-free core. The core shutdown margin vill be verified by planned control rod move-

'g-ment to demonstrate the absence of criticality or to provide data for the calculation of shutdown margin prior to start-up

$2.

after any shutdown in which the system has cooled sufficiently to be opened to atmospheric pressure and any of the following

' 'I situations exist:

3, j

.\\.

. +-f i

t n,.

,e v....

~

,,,, ~

- + -,,

.. ~ - -.

J

- 23 ' -

1:

f

-t-

[ UI

'Puel has been added and/dr repositioned in a way

' N which is not-definitely known to. reduce reactivity; or

' Lf Any st' eel ' channels have been replaced by zircaloy j-channels during the shutdown; or L

A control rod has been changed and presence of poison has not. been verified;- or 35,000 MWDt have been generated by the plant'since the

,j' previous margin demonstration.

.1 During power operation, if reactivity and control rod. motion -

data indicate a possible loss of poison from a control rod, the reactor shall be brought to' the cold shutdown condition.

(c). Control Rod Withdrawal Verification Checks

]

The following checks shall be made to demonstrate that the

~

control-rod poison section follows the movement of the con-trol rod drive:

(i)

Coupling integrity check involves fully withdrawing a rod and observing that the drive. vill not go to the overtravel position. Each control rod shall be verified.

using this. check prior to commencement of each refueling operation and/or series of critical tests with the reactor vessel head removed.

-l' (ii)

- Appropriate rod withdrawal verification checks shall be performed prior to or during power operation start-up j.

for each control rod which has not been verified since its last scram. Verification may consist of either a coupling integrity check or observation of niclear in-

[

strumentation response to rod withdrawal. Weekly sa___e i

C analysis of reactor system vnter for presence o." boron vill'be made.

(iii) Appropriate rod withdravel verification checks shall I

be performed during sustained power operation. Proyer rod following shall be verified by nuclear instrumenta-l_

tion response.

- ( iv).

At. any time during power operation that it is not pos-sible to verify the coupling integrity either by means -

of nuclear instrumentation response or overtravel check, the potential reactivity increase that could result from I

simultaneous dropping of all unverified control rods vill be limited to 1%21k pp/k pp.

e e

f^

.h:

[i i.

II

r_

i.

24 -

l' i:

I 5 4 3 Liquid poison system-l The liquid poison system shall be available for operation at-all times during refueling and power operation. The reactor shall i

[

be shut down in any situation where the poison solution tank level drops below an equii lent of 850 gallons-19 veight percent sodium.

[

pengaborate or the poison solution storage temperature drops belov

~110 F, or where the ability of the system to inject poison into l

j the reactor is in_ doubt. Components of the system vill be checked l

at one to two month intervals for proper operation except for actu-l I ation of the injection valves. If liquid poison injection is -

initiated, the reactor core vill be poisoned to assure its sub-criticality' in the cold, xenon-free (less than 0.001 okerr/kerr in xenon) condition before terminating liquid poison injection.

]

The' reactor vill not be operated' after poison has been injected until the boron concentration in the reactor water has been re-duced to 100 ppm or less.

.5.4.4 Feactivity Coefficients During initial loading of the core, the moderator temperature and void coefficients will be measured as indicated in Section 7 0 of these specifications.

The reactivity coefficients shall meet the following requirements:

(a) The effect upon reactivity of increasing voids at constant.

pressure shall always be negative. A measurement which shows a negative effect of voids introduced inside a flow channel under cold conditions vill constitute an adequate demonstration of the negative character of the void coefficient.

t

.I (b) The moderator temperature coefficient (inferred from critical control rod position) during uniform heating of the core shall be limited such that the potential maximum Akeff/keff produced by heating the moderator is always less than one dollar.

-[

(c) The over-all effect of increasing reactor power at constant pressure shall be a loss of reactivity when the system pres-sure is in excess of 800 psig.

545 Eeactivity Additions During Core Alterations l

The limits and requirements which apply to rc ctivity additions are as follows:

l.

f I-l -

1

k i.

- (a) The procedure for core alterations which may increase reactivity includes withdrawal of a control rod of maxi-mum worth in the vicinity of the alteration before and after the alteration to verify that the core is suberitical.

j All rods shall be fully inserted during the reactivity addition. Checks vill be made at frequent intervals during core alterations to assure that the core shutdown margin requirement is being met.

(b) Core alterations which increase reactivity shall be limited between suberiticality checks to fuel loading increments which do not exceed the placement of one fuel bundle or the exchange of two zircaloy channels for two steel channels.

6.0 PLANT SAFETY AND MONITORING SYSTEMS The plant safety and monitoring systems are considered to encompass the reactor safety system including related features, the control rod withdrawal permissive system, the refueling operation interlock system, and the plant monitoring systems.

6.1 REACTOR SAFETY SYSTEM 6.1.1 General Features (a) The reactor safety system consists of sensing devices and associated circuits which Eutomatically initiate a reactor scram or other required action. Certain of the sensing devices also initiate automatic closure of the containment vessel isolation valres, while other sensors initiate emergency cooling of the reactor through operation of the emergency condenser or through operation of the core spray system. Controls are also available in the cortrol room to permit manual initiation of a scran, manual initiation of penetration closure, and manual initiation of emergency cooling of the reactor.

(b) The reactor safety system uses two parallel safety channels.

These channels have separate power supplies and separate chains of sensor trip contacts. The channels are designed on the f ail-safo principle (de-energizing vill cause a scram).

+

The reactor stiety system is designed so that both channels must be de-energized to cause a scram and must be reset manually subsequent to a scram and prior to start-up.

.l 1

b i

'i h

i 1

. t 6.1.2 Feactor Safety System During Power Operation The following tabulation gives the arrangment of the reactor safety sys:cm during power operation:

Trip Scram Contacts Coincidence Setting Sensor and in Each in 'sch and Trip Device Channel Channel Tolerance Special Features High reactor 2

1 out of 2 50 5 psi above pressure (4 pres-reactor opera-sure switches) ting pressure Low reactor water 2

1 out of 2 Elevation 610' Closes enclosure iso-level (h level 6" tl inch lation va'.ves.

If switches) reactor pressure in less than 100 psig actua'es core spray systet.

1 8.0 0 5 inches Low steam drum 2

1 out of 2 water level (4 below operating level switches) level Main steam line 2

1 out of 2 5015 percent backup isolation of full closure valve closure (h position switches) t 8.0 0 5 inches Bypassed by master High condenser 2

1 out of 2 pressure (h pres-of Hg absolute switch as described sure switches) pressure in Section 6.1 3 High enclosure pressure (4 pres-2 1 out of 2 2.010.2 psi Closes enclosure iso-sure switches) above atmos-lation valves. Two pheric independent pressure switches actuate a time mechanism that initiates post-incident spray system in 15 minutes unless the con-trol is manually over-ridden.

i i

I i

27 -

Trip Scram Contacts Coincidence Setting Sensor and in Each in Each and Trip Device Channel Channel Tolerance Special Features High scram dump 2

1 out of 2 5/16t1/2 inches Alarms at level of 10 tank level (4 level below tank cen-inches below tank switches) ter line center line.

Recirculation 12 1 set out Approxinately waterline valves of 2 10 percent of closed (two sets full simultane-of 6 position oua closure of switches in each both discharge of 2 channels) or both suction valves or 5015 percent of full simultaneous closure of both butterfly valves or any combina-tion of one of these valves in each loop Loss of auxiliary 1

1 out of 1 52 20 volts Closes all automati-power supply cally actuated enclosure (voltage relay) isolation valves and actuates emergency condenser.

t 120 5 percent Interlocks prevent H1 h neutron flux 3

2 out of 3 6

(each of 3 power of the o to 150 control rod withdrawal range flux moni-percent scale, as described in tors has a trip or 4812 per-Section 6.2.1.

contact in each cent of the O channel) to 60 percent scale (any range)

Short period (each 2

1 out of 2 10t2 second Placement of any 2 out of intermediate period of 3 range switches of range log-N the high level neutron period monitors flux channels in the

}

has a trip contact power range position in each channel) vill bypass this period trip feature.

Manual scram 1

1 out of 1 (1 switch)

L i

'k

~

. ~,.

a

. p i

t.-

- i-The enclosure ventilation valves are designed to close vithin 6 seconds after any trip which initiates reactor scram.

t

- t Any sensor maylbe taken out of_ service Inr maintenance

-since its trip contacts are left in the scram condition.

6.13. Reactor Safety System Bypass

' The following tabulation gives the permissive functional con-ditions during which certain reactor safety system sensors are bypassed by the reactor safety system mode selector switch:

i Mode Selector I-Switch Position Trip Functions Bypassed Run None Start-up High condenser pressure

' Bypass dump tank (a)

Lov steam drum vater level Recirculation waterline valves closed Steam line backup isolation valve closed High condenser pressure High water level in scram dump tank (b)

Refuel Low steam drum vater level Recirculation waterline valves closed Steam line backup isolation valve closed 4

High condenser pressure l

Shutdown None (c) 4 (a) Control rod withdrawal is prevented by interlocks while switch is in this mode position.

f 4

(b) Bypass of this trip function is necessary to enable f'

emptying the dump tank after a scram.

4 s

(c) With the mode switch in the " Shutdown" position, both the scram circuit and the control rod withdrawal cir-

{

cuit are open. Control of the ventilation duct circuit is transferred to a relay which provides penetration closure protection through signals from "high sphere pressure" and

f "lov vater level in reactor vessel." This permits normal I

ventilation in the containment vessel during shutdown when i

the control rods are held in the full-in position.

i

-t I

4 k'

1 J--------

w

,# 3 ;-y

~6.1.4 Related Systems-

-op i

(a) Start-Up Flux Monitors

' Two start-up range flux monitoring channels are provided

,e for use during reactor start-up and low power operation.

~

-(b)

In-Core Flux Monitors i

Twenty-four in-core flux monitoring channels are provided.

These channels vill not be used to initiate reactor scram.

(c) Core Spray System Control

~

The core spray system is automatically actuated by simul-taneous tripping of 'the reactor safety system sensor "lov reactor water level" along-vith the " low reactor pressure" t

trip device. The " low reactor pressure" trip device con-

- sists of a pressure svitch interlock which prevents admis-1 sion valve opening while the reactor pressure is above the i

pressure of water supplied by the fire _ protection system.

[

(d) Emergency Condenser Control 4

A pressure switch initiates automatic operation of the emergency condenser if the reactor pressure reaches 100110 psi sbove the reactor operating pressure setting.

l 3 i'

6.1 5 Operating Eequirements 1

t

-( a) The reactor safety system shall be operable during power i

operation. as indicated in Section.6.1.2. This system shall be j

functionally tested at least every 15 months. All portions of this system which can be made temporarily inoperative -

g without either requiring plant shutdown or jeopardy to-nuclear j~

safety vill be checked at least every 2 months.

(b) The core spray system and emergency cc-denser ccntrol initiation sensors shall be functionally tested at least 1

j-every 15 months.

y 4-l

_(c) At.least one of the two start-up range neutron monitoring 1

channels shall be operable and measuring flux from the core during refueling operations, and during the reactor start-up ' to' power operation.

i l

i

!~

I.

Ll q

' ;{

b

.~~

~-6

-)

r 2.. V i

y, U F

.' j -

--(d) L Any one of the three' power range flux monitors may be taken 6

out of service for maintenance during reactor operation.

[

If one monitor is out of service, a trip on either'of the two remaining monitors will scram the reactor. When main-tenance.is.necessary, no major changes in' power level, flux

[-

' distribution or the control rod pattern vill be made.

~

(e) Sufficient in-core monitors shall be. operable to provide data for comparison with selected calculations pertaining -

to initial approach to power and research and development activities.

6.2 cOuTRoL ROD WITHDRAWAL PERMICSIVE SYSTE4 l

1 6.2.1 Interlocks

{

. Interlocks prevent control rod withdrawal when any of the

-following conditions exist:

(a) With any two of.the tnirty-two scram accumulators at a

~~

pressure below 700 psig.

(b) With two' of the three power range channels reading below 5 percent on their O to 150 percent scales (or below 2 percent on their 0-60 percent scales) when reactor power is above the minimum operating range of these channels.

This interlock is bypassed when all three of the power

[

range channels are set on the minimum operable range.

}

(c). With the scram dump tank on bypass.

(d) With the mode selector switch in shutdown.

1 6.2.2 Operating Requirements The con;rol rod withdrawal permissive interlocks shall always be operable. No further withdrawal of control rods will be.per-i mitted if one of these circuits is found to be inoperable. Per-missive. circuits shall be functionally tested at least every 15 months.

.6.3 REFUELING OPERATION INTERLOCK SYSTB4 i

631 Reactor Refueling System l'

l All of' the trip devices not bypassed by the mode selector switch in the refuel position are operative during all refueling operations.

.]

This-includes the sensors and trip devices of the reactor safety J.

-system as specified for power operation as follows:

.e

'.1 i

[.;

H1 h reactor pressure 6

Lov reactor water level High enclosure pressure High scram dump tank level Loss of auxiliary power supply High neutron flux Short ' period Manual scra:

6.3 2 Rerueling operation controls (a) Interlocks are provided to prevent all motion with any of the refueling cranes (namely, jib crane, transfer cask winch, and monorail crane) which are positioned over the reactor vessel whenever any control rod is not fully inserted in the core and the mode selector switch is in the " Refuel" position.

(b) Key locks or interlocks are provided within the speed controls of the refueling cranes (jib crane, transfer cask winch, and monorail crane) to limit their lowering speeds at elevstions corresponding to the core position, such that their maximum travel rate is no more than 100 inches per minute.

6.3.3 operating Requirements (a) All reactor refueling safety system sensors and trip devices shall be functionally tested at least every 15 months, and shall be maintained in the specified condition during all refueling operations.

(b) The refueling operation controls including position inter-locks and travel speed controls shall be functionally I'

tested at least every 15 months.

6.h PLANT MOUITORING SYSTEMS i

The plant monitoring systems include the process radiation monitoring systems, the radioactive vaste disposal systems, and the area monitoring sys tem.

6.4.1 Process Radiation Monitoring Syctems l

The process radiation monitoring systems consist of the air ejector off-gas monitoring system including the fuel rupture detection system, stack gas monitoring system, the emergency

{

condenser vent monitor, and process liquid monitor system.

?

i f

f (a) Air-Ejector Off-Gas Monitorine System Continuous monitoring of the air-ejector off-gas radio-activity is provided during operation of the reactor to provide a continuous warning of possible future release to the environs. This avenue of release is closed by a valve when the activity level approaches the limit for allowable discharge.

At a release rate calculated to be equivalent to four (4) curies per second, (at time of emission from stack) a trip circuit in the air-ejector off-gas monitor causes an alarm to be sounded in the control room. At a release rate similarly calculated to be equivalent to forty (40) curies per second, the air-ejector off-gas monitor trip circuit initiates action of a time-delay switch, which in turn trips the off-gas shutoff valve closed after a preselected delay adjustable up to 15 minutes.

( Off-gas average holdup time is about 30 minutes.)

(b) Stack Gas Monitoring System An isokinetic probe, permanently fixed in the stack approxi-mately one-third up from the base, collects stack gas and particulate samples that are withdrawn with a gas pump through fltv-metering and regulating equipment. The sample is passed through a replaceable filter which is located up-stream fron the gas sampler. The filter is Ieriodically removed and checked for particulate cortamination. After filtering, the continuous flow gas sample is presented to a continuous monitoring single channel gamma spectrometer.

(c) Emergency Condenser Vent Monitor The emergency condenser vent is monitored to detect a release of radioactive material. This vent monitor con-sists of a gamma sensitive channel employing a scintilla-tion crystal sensing device. This channel has a range of 0.1 to 100 mr/hr, and is provided with an alarm to inform the operator of a release of radioactive material.

(d) Process Liquid Monitor System A process liquid monitor system employing gamma scin-I tillation detector channels is provided to indicate, record, I

and annunciate high radiation level in the process liquid streams as follows:

l l

I t

i t

q -

i Radioactive vaste system effluent to canal Eeactor enclosure cooling water Main condensate demineralizer influent Circulating water discharge Service water return from reactor enclosure The radioactive vaste system effluent to canal channel vill be used in conjunction with radiochemical analysis of batch samples to allow manual control of discharge of liquid radio-active vastes. The circulating water discharge monitor vill normally monitor the main stream of plant effluent at its discharge in the canal.

6.4.2 Radioactive Waste Disposal Systems (a) Airborne Wastes The airborne radioactive vaste disposal system consists of holdup piping, a 240-foot high stack, and sufficient con-trols and instruments to accomplish the following:

Provide short term decay time.

Provide stack flow of approximately 30,000 Crx.

Provide for controlled release and dispersion of the noble gases, xenon and krypton, which may be released due to a fuel element rupture. Provisions include automatic closure features to prevent release of air-borne radioactive materials in excess of the permissible emission rates.

(b) Liquid Wastes The liquid radioactive vaste disposal system consists of collection sumps, receiver tanks, tank mixing eductors, strainers, filter, demineralizer, holdup tanks, concen-I trator, pumps, interconnecting piping and control instru-i mentation.

Release of liquid radioactive vaste is predicated upon con-l siderations as follows:

Those liquid vastes which are likely to contain radio-g nuclides of plant origin are sampled, analyzed as ap-propriate, and released on a batch basis. Other flowing liquid streams are continuously monitored at time of

}

release to the environs.

i l

(

Limits employed for release decisions recognize that certain radioelements cannot be present from process origin with regard to use of " unknown" radionuclide limits. Where appropriate, analysis vill be done to permit use of more realistic per-missible limits.

(c) Solid Wastes Spent demineralizer resins vill be sluiced to a shielded 10,000 gallon storage tank. An underground solid vaste burial vault facilicy ic available for storage of other solid radioactive vastes.

6.4 3 Area Monitoring System (a) Fixed gamm.a monitors employin6 scintillation type detector channels with a three decade range which will be set between 0.01 and 1000 mr/hr are installed at various locations throughout the plant. The output of each channel is indicated and recorded in the control room. The main board annunciator alarms any channel reaching a high radiation trip. The trips vill be set on the basis of experience to provide a warning in the event of abnormal radiation levels.

(b) Two of the area monitoring channels are located in the vicinity of refueling operations to provide gamma monitoring oof the fuel etorage areas.

6.4.4 oreratine Requirements (a) The permissible radio gas release rate from the stack of 4 curies per second as averaged over a period of one year vill not be exceeded. Therefore, the air-ejector off-gas monitors will be set to alarm if the off-gas activity reaches a level that would correspond to a stack release rate of 4 curies per second and to initiate closure of the off-gas isolation valve if the activity reaches a level that would correspond to a stack release rate of 40 curies per second. At least one of the two air-ejector off-gas monitoring systems shall be operable and capable of initiating closure of the off-gas isolation valve during power operation. The calibration ar

• auto-matic function shall be checked at least every two months i

during power operation, f

1

(

4 i e

(b) The stack gas monitoring system vill normally be in service during power operation.

If the normal monitoring system must be taken out of service, other means vill be t

made available for continuous stack gas monitcring during power operation.

(c) There will be an emergency condenser vent monitor in service during power operation.

(d) Liquid radioactive wastes will be released if the gross activity of plant origin in the effluent can be regulated so that it does not exceed 10-7,j4c/cc in the discharge canal or other proIer limits based on analyses of radionuclides present using permissible concentrations listed in 10 CFR Part 20.

l 70 OPERATIID pFOCEEURES This section describes those plant operating procedures and procedural safeguards which have a potential effect on safety.

Operating principles and procedures are presented for initial start-up of the plant, for normal and emergency operation of the plant, for the initial phase of testing within the research and development program, and for operational testing of the nuclear safeguards systems of the plant.

71 BASIC OPERATITO PRItCIFLES 7 1.1 0Feration and control of the reactor and most of the process equipment is centralized in the control room, which is located in the turbine building.

77.2 There vill be at least two operators (one of whom vill be an AEC licensed operator) in the control room for start-up and shutdown of the plant. There vill be at least one AEC licensed operator in the control room at other times during power opera-tion and also during refueling. No licensed operator is re-quired in the control room when reactor is in the cold shutdown condition.

713 operators may perform certain operating functions at control panels and valve racks outside of the control room but only at the direction of or with prior knowledge of the operator in the control room.

I I

7 1.4 start-up, normal shutdown, and all other repetitive operations which may involve nuclear safety will be performed in accordance l

with specific written procedures.

I 7.1 5 Routine maintenance of protective devices and critical operating equipment vill be done in accordance with established written schedules.

7 1.6 Incidents and acts having a potential for detrimental effect on nuclear safety will be investigated to effect procedures to pre-vent recurrence.

72 PROCEIUPAL SAFEGUARDS The following procedural safeguards have been established for the operating safety of the plaut.

7 2.1 Detailed Operating and Emergency Procedures (a) Written procedures for normal and emergency operations which may involve nuclear safety vill be prepared and issued prior to start-up of the plant.

(b)

Instructions for normal operation vill consist of detailed procedures required for the oIeration of plant equipment.

Radiation control procedures vill be compiled to cover asIects of the plant's radiation protection program.

7 2.2 Administrative Procedural Controls The following controls vill be employed to promote safety of the plant:

(a) Training of the oIerating staff so that each employee is acquainted with his sIecific duties and responsibilities and the act!on to be taken in the event of off-standard conditions.

(b) Periodic management review for strict adherence to the operating and emergency procedures and the radiation control procedures.

(c)

Periodic management review for strict adherence to the operating limits and requirements for the plant.

(d)

Periodic management review for strict control of access to the plant.

1 (e)

Periodic management review for strict adherence to the procedure for investigating and reporting unusual or un-exrected incidents.

I.

}

i

_- l 73 INITIAL CORE LGADING AND CPITICAL TESTS 731 Basic Test Conditions 7 31.1 The loading and critical testing program vill begin after the special. initial loading instrumentation and the necessary reactor equipment have been checked and found to be in a safe and operable condition.

7 31.2 At the start of loading, the reactor vessel vill be filled with water.

7 313 neutron sources to provide meaningrul readings on neutron sensi-tive chambers vill be installed before loading starts. Chaiber-source geometric relationships will be ~aintained such that neutron multiplication in the fuel is observed.

7 31.4 During this period, the control rod scram circuit vill be operated by at least four neutron sensitive channels whose chambers see neutrons originating in the fuel. Esch control rod, before and after it is encompassed by fuel, vill be checked for proper func-tioning in all modes. Presence of poison in the control rod vill be verified by the time it's encompassed by fuel.

732 Core Loading and Test Program The initial loading and critical testing program will be divided into three major parts as follows:

7 3 2.1 Loading to Minimum Critical Size Cores Criticality calculations vill be used to guide the experimental work. Fuel vill be installed in an area provided with zircaloy channels selected to be adjacent to or currounding a neutron sou rce. All, or all but one or two, con +rol rods vill be inserted during fuel additions. (a) Loading vill take place in a stepwise fashion; neutron multiplication measurements will be made with all control rods withdrawn between steps. Puel vill be added in steps of one bundle between multiplication measuretents until a minimum critical size core is obtained. When the minimum critical size core for this fuel and channel arrangement has l-been established, the k rf vill be esticated. This core vill e

then be disassembled.

The zircaloy channels vill be replaced with steel channels and a second (larger) minimum critical size core vill be assembled in the same manner and general location. The kcff vill be esti-1 mated and may be increcsed by the addition of fuel bundles to l

provide reactivity for censurements of void coefficients, and temperature coefficicnts.

(a) The refueling operation interlock system as described in 9

j Section 6 3 2 (a) vill be cade ineffective when all control rods are not fully inserted during only this initial period of loading to =inimum critical size core.

f-7 3 2.2 Loading to Intermediate core size Fuel vill be added in increments of one-to-four bundles with all control rods inserted. The loading procedure is as follows:

(a) A control rod adjacent to the area to be loaded will be com-pletely withdrawn to prove suberiticality with one rod out; the rod vill then be inserted.

(b) The control rod in the area to be loaded vill be functionally tested and inserted.

(c)

Fuel vill be loaded. The control rod in the area loaded vill then be withdrawn to prove suberiticality with the rod out 4

and to provide functional testing after fuel surrounds the rod. The rod vill then be inserted, and these three steps will be reIcated.

(d) Loading vill be continued until a precalculated core con-figuration is reached. Tests will be made in this core using any of the following techniques:

(1)

Multiplication measurements (ii)

Positive or negative leriods (iii) Low power level irradiations of detectors During this testing, uniform and nonuniform control rod patterns may be used; comparisons may be made of fuel and controls with standard and/or reference fuel and controls.

7 3 2 3 Loading to " Full core" size The initial " full core" will contain approximately 56 bundles.

Fuel vill be added in increments of one-to-four bundles with all control rods inserted. The procedure is identical with that described in Paragraph 7 3 2.2.

Loading vill be continued until the core contains 56 bundles. If the criterion in 7 3 2.2(a) cannot be met, the loading vill be discontinued and additional i

steel channels vill be installed. When the core is fully loaded, the following tests vill be performed:

(a) Void and temperature coefficients of reactivity are measured.

(b) Critical configurations with uniform and nonuniform control rod patterns are determined, a normal start-up rod-withdrawal pattern is selected and reactivity addition rates near critical are measured.

k 4

(c) The fully shutdown and one-rod-out reactivity margins are estimated.

(d) When full core critical testing is completed, the operating neutron sources are installed. The neutron-sensitive chambers used in the rea: tor vessel for testing are trans-ferred, in a etervice fashion, to their design locations concurrent with measurement of neutron attenuation between in-vessel and design location.

74 POWER OPERATION TEST PROGRAM The power test program, consisting of five phases, vill commence only after the initial loading and critical test program has been completed and the results of this program found to be satisfactory.

7.4.1 Phase A - Heating Power, Approximately 0 to 5 Mut, o to 1035 Psig The tests in this phase are run while the reactor is brought up to rated pressure and temperature for the first time. The normal cold start-up procedures described in the following Paragraph 7 5 2 vill be used.

Reactor power will be maintained in the range of approximately 0 to 5 Mwt. The test program during this phase will include the following:

(a) Control rod performance (the ability to scram rods in the hot pressurized condition vill be demonstrated).

(b) Radiation surveys.

(c) Temperature coefficient observations.

(d) System expansion checks.

(e) Reactor vessel temperature differential measurements.

(f) Emergency condenser performance test.

i (g) Critical control rod configurations in the hot conditions.

5 (h)

Performance testing of reactor auxiliary equipment and cont rols.

i (i)

Radiochemical checks.

f 4

i I

40 -

7 4.2 Phase B - Power, Approximately O to 120 Mvt (C$ to 75% Rated),

1035 Psig The reactor power vill be increased in increments of approximately 40.Mut during this phase of testing. The results at each step shall be found satisfactory with respect to plant safety before proceeding to the next step. The turbine vill be placed in service during this phase. The test program at each step vill include the following:

(a) Determination of reactivity changes resulting from variations of steen flow and transient poison in the core. The effect of minor variations of reactor pressure vill also be determined.

(b) Radiation surveys.

(c) Flux vire irradiations to make preliminary calibration of the in-core system.

(d) hadiochemical tests.

(e) System transient tests.

(f) Observation of reactor stability.

(g) Standard turbine start-up tests.

(h) Gamma probe or gamma scan to determine power distribution after operation at 120 Mut.

7.4.3 Phase C - Final In-Core Calibration at Approximately 120 Mwt (75% Rated), 1035 Psig During this phase the reactor vill be run at a steady power level such that the steam void and reactor power distribution approaches that expected at rated power. Flux vire irradiations will be per-formed and the data obtained will be used for a final calibration of the in-core system before the approach to rated power.

Since this will be the first time the plant will have been at high power for an appreciable length of time, chemical and radiochemical tests vill be run. These test results vill be used as reference values for later comparison.

I t

bl -

i l

7.4.h Phase D - Power, Approximately 120 Mwt to 157 Mwt (75% Rated To Rated), 1035 Psig i

This phase covers the approach to and attainment of full power.

l This vill be done in increments of approximately 20 Mwt with the tests described for Phase B, excepting test (h), carried out at each step. The resul',s at each step shall be found satisfactory before proceeding to the next step.

7.h.5 Phase E - Full Fower Demonstration Test i

A demonstration test consisting of a rated power run to prove that the plant meets design and contractual requirements vill take place during Phase E.

75 NORVAL OPERATION 751 Gene ral Detailed operating procedures for each normal mode of plant operation vill be prepared prior to operation.

The following is an outline of the principal normal operation procedures having a potential effect en the safe operation of the plant.

752 Cold Start-Up After Extended Shutdown A cold start-up will occur each time the reactor is returned to service following an extended sl.utdown. The procedure for a t

normal cold start-up is as follows:

(a) A start-up check list vill be followed prior to beginning the actual start-up so that applicable equipment and systems are in condition for start-up. Containment vessel integrity provisions must be in effec'.

}

(b) Each control rod vill be exercised and scran=ed as a check j

of the control rod hydraulic system and the reactor safety system. A coupling verification check vill be included prior to or during start-up.

(c) The start-up check list vill be reviewed and approved by the Shift Supervisor prior to start-up.

(d) The reactor vill be brought critical by control rod with-drawal following a prescribed withdrawal pattern.

l t

I

i h3 -

1 (e) The power vill be adjusted once criticality is reached to maintain a reactor vessel temperature rise rate not to exceed 100 F per hour.

(f) The turbine shaft sealing system vill be placed in service as soon as sufficient steam pressure is available.

(Approxi-mately 150 psig.)

(g) The condenser will be evacuated with the mechanical vacuum pump and the air-ejector vill be placed in service.

t (h) Turbine heating can be started during this operation sequence.

After turbine heating is completed, and the reactor reaches rated pressure, the turbine is gradually brought up to speed.

(i) The mode of turbine control is next transferred to the initial pressure regulator.

(j) The control rods are adjusted to provide the desired power distribution within the core, 753 Hot Start-Up Whenever the plant has been shut down for a period of time with the reactor vessel and auxiliaries remaining pressurized, a hot start-up procedure may be followed to return the plant to service.

This procedure vill be essentially independent of the cause of shutdown assuming that the cause is recognized and any nonstandard conditions have been corrected. The reactor instrumentation vill be reset and downscaled and a hot start-up check list vill be com-pleted prior to the withdrawal of control rods. A coupling integrity check vill then be made on all control rods which have been scrammed since the last coupling check. An alternate procedure involving control rod withdrawal verification by nuclear instrumentation response during start-up may be substituted for this check. Tne start-up will then proceed in accordance with Paragrapns (d) through (j) of 7 5 2 of the normal cold start-up procedure outlined above.

754 Normal Power Operation During normal power operation, the initial pressure regulator will 4

j maintain the reactor pressure at its rated value by operating the turbine admission valves. The turbine-generator load is established by the control rod positions. The principal function of the opera-l ting personnel during this period vill ce as follows:

(a) The maintenance of a continuous watch in the control roon l

for prompt attention to any annunciated alarms.

i

1 h3 -

l (b) The adjustment of the control rod pattern to acco=modate changes in reactivity and to maintain the desired power distribution.

(c) The evaluation of abnormal conditions and the initiation of corrective action as required.

755 Extended Shutdown An extended shutdown vill be accomplished as follows:

(a) Reactor power vill be reduced by manipulation of the control rods, and the main generator load will be de-creased simultaneously. The turbine-generator vill be separated from the system.

(b) All control rods will be inserted.

(c) The removal of reactor decay heat and the reduction of reactor pressure vill be accomplished by controlling reactor steam flow. The rate of cooling of the reactor vessel vill not be allowed to exceed 100 F per hour.

(d) The reactor shutdown cooling system may be placed in oIeration whenever reactor pressure drops below a pressure sufficient to maintain turbine seals. This system vill corr 'ete the cooling of the reactor water g

to 125 F.

7 5.6 Short Duration Shutdown A shutdown of short duration may be accomplished while main-taining system pressure. The turbine-generator vill be unloaded and separated from the system. Reactor heat will be accom=odated by system losses or bypassing steam to the main condenser.

7.6 RESEARCH AND v2VELOPME!7T PROGRAM i

Written procedures will be prepared and issued prior to the initiation of each test involving the operation of the plant.

These procedures vill provide the information for performing each test, detailed to the extent appropriate, according to the outline as follows:

f (a) The objective of each particular test.

I (b) The step-by-step operational method for the test including l

specific terminal aspects of the test.

-l 1

i

(c) The expected response of the object undergoing test and, as appropriate, the expected response of the system within which the test is being conducted. The course of action to

}-

be taken in the ovent that an abnormal response occurs dur-ing the test will be defined. The responsible individual having authority to direct the continuance or discontinuance of a particular test vill be specified.

(d)

In addition to the details covered above in this outline, there vill be instructive information provided with any special equip-nent that may be employed in the course of testing.

77 REFUELING OPERATION The refueling operation vill be conducted in accordance with the following basic principles:

(a) Written procedures vill be available prior to each rei'ueling outage.

(b) The insertion and removal of fuel bundles and channels vill be done through the top of the reactor vessel after opening reactor vessel head closures as appropriate. Water shielding is provided by flooding the reactor vessel and the refueling extension tank. Fuel bundles and channels vill be handled by means of a grapple, transfer cask, and crane.

(c) The trip devices specified in Section 6.31 vill be in service and connected to the reactor safety system during all refueling operations.

In addition, at least one of the start-up nuclear instrumenta-tion channels vill be in service during all refueling opera-tions.

(d) The procedure which will be used for core alterations which increase reactivity vill involve withdrawal of a control rod of maximum vorth in the vicinity of the alteration before and after the alteration to verify that the core is suberitical.

All rods will be inserted during the actual reactivity addition.

The core vill be suberitical when the control rod in the vicinity 7

of the alteration is withdrawn if the core shutdown requirement is being met. This requirement will be checked at frequent in-tervals during core alterations. Communications between the l

control room and the loading area vill exist during all core alterations.

t I

4 4

t h5 I

i (e) The liquid poison system vill be available and ready for use.

1 (f) Containment integrity provisions will be in effect during refueling operations.

)

(g) Unirradiated fuel vill be stored in air in a new fuel storage area within the reactor containment vessel.

(h) Irradiated fuel and irradiated channels vill be stored in the spent fuel storage pool.

1 7.8 MAINTENANCE The following basic principles will guide the maintenance program at the plant:

7 8.1 Damaged or defective equipment vill be repaired or replaced.

7 8.2 Maintenance check lists vill be used wherever practicable to assure that equipment is included in the systematic preventive maintenance program and to guard against error or damage in carrying out the maintenance effort.

783 A system of equipment history records vill be kept in which Vill be recorded the extent of and type of repair, the regular preventive maintenance actions, as well as any nonroutine main-tenance which is required.

7 8.4 The preventive maintenance program vill include a schedule for exercising of normally idle components.

I 7.8.5 Instrumentation and control systems, especially tne neutron power level instrumentation and the reactor safety system, can g

be tested periodically with the plant in operation, and certain portions of the systems can be replaced with spare units while the plant is in oIeration should it be necessary.

7.8.6 Radiological protection practices vill be observed in maintenance activities.

]

79 OPERATIONAL TESTING OF NUCLEAR SAFEGUARD SYSTEMS l

Procedures for testing of plant components and safety systems which have a potential safeguards function are prescribed in Sections 3 0 through 6.0.

These tests and frequency of testing are tabulated as follows:

1 I

46 -

I Beference Procedure System of Function Interval Between Within These Undergoing Test Routine Tests Specifications Containment vessel 6 months or less access air locks g

leakage rate Functional test of post-At least every 15 months Section 3 5 2 incident spray system i

automatic control I

operation 4

Control rod performance At least every 15 conths Section 5 4.2 I

h Liquid poison system Two months or less during Section 5.4 3 component operability power operation Beactor safety system At least every 15 months Section 6.1 5 scram circuits Enclosure isolation A least every 15 months Section 6.1 5 trip circuits Reactor emergency At least every 15 months Section 6.1 5 cooling systems trip circuits b

Control rod withdrawal At least every 15 months Section 6.2.2 permissive interlocks function Befueling operation At least every 15 months Section 6 3 3 controls function Calibration and functional Two months or less Section 6.4.4 test of air-ejector off-gas and stack gas monitors

[

I l