Text

Tech Specs Re Principal Design Features & Operating Limits & Requirements Which Have Effect on Safety of Operation of Plant
	ML20126E613
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	05/06/1964
From:	; US ATOMIC ENERGY COMMISSION (AEC)
To:	;
Shared Package
ML20126E606	List: ML20126E603; ML20126E608; ML20126E613;
References
	NUDOCS 8102060452
	Download: ML20126E613 (83)
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{{#Wiki_filter:,. _ _ i. O E b i L APPENDIX "A" CONSUMERS POWER COMPANY BIG ROCK POINT NUCLEAR PLANT TECHNICAL SPECIFICATIONS APPENDED TO OPERATING LICENSE NO. DPR-6 k Date: ggj jg64 T/0c206o'f53

i 9 CONTENTS e i Pape No. 1.0 Introduction.............-......... 1-1 1-1 1.1 Scope 1.2 Definitions 1-1 -i 2.0 Site.......................... 2-1 2.1 Location........................ 2-1 I 2.2 Boundaries......................... 2-1 2,3 Principal Activities.................... 2-1 3.0 Reactor Containment 3-1 3.1 Containment Sphere Design Parameters............ 3-1 3.2 Containment Sphere Dimensions 3-2 t 3.3 Construction........................ 32 3.4 Penetrations........................ 3-2 3.4.1 Design Features 3-2 3.4.2 Methods of closure...................... 3-3 3.4.3 Operating Requirements................... 3-4 3.5 Post-Incident Spray System,............... 3-4 i 3-4 3.5.1 Design Features 3.5.2 Operating Requirements............... 3-5 3.6 Containment Requirement s................... 3-5 3.7 Containment Sphere Leskage Testing.... 3-5 4-1 4.0 Reactor and Power. Systems Equipment 4-1 I 4.1 Reactor System Equipment..~................ 4.1.1 Reactor Vessel..'..................... 4-1 4.1.2 Primary-Coolant Recirculation System....... 4-3 r 4.1.3 Primary. System Shielding............. 4-7 4.2 Power System Equipment and Associated Facilities......................... 4-5 4.2.1 Electrical System 4-8 4.2.2 Main Condenser..... 4-9 4.2.3 Turbine Bypass Control System 4-9 4,2.4 Condensate and Feod-Water System......... 4-10 4.2.5 Reactor Cooling Water System... 4-10 4.2.6 Fire Protection System. 4-10 4.2.7 Ventilation System......... 4-11 4.2.8 Reactor Service Water System........... 4-11 4.2.9 Instrument and Service Air System 4-11 4.2.10 Turbine-Generator Unit. 4-11 4,2.11 Fuel Storage............ 4-11 5-1 5.0 Reactor Core and Controls 5-1 5.1 Design Features 5.1.1 Principal Core Materials.................. 5-1 5.1.2 Control Rods..-.................... 5-1 5.1.3 Control Rod Drives..................... 5-4 i 1 +. - wr-e ->.-.#6 e. -y .e,_-.* 4.,.,,ove.4#.e..-.mm,,,,wr-w.-.,,.,ww.,__.e.,..+,,--wy-,,, e.9 y .,w,.,,, .m.,,,.*=,y- -%.y pw p .,yn.,,- ry y,, g-

( ', s 4 CONTENTS (Contd) Pane No. 5,0 Reactor Core and Controls - (Contd) 5,1.4 Liquid Poison System... 5-6 5.1.5 General. Core Composition..... . 5-6. 5.1.6 Sources 58 5.2 Principal Calculated Thermal, Hydraulic and Nuclear Characteristics................. 5-9 5.2.1 Principal Calculated Thermal and Ilydraulic Charseteristics of the Core Loading 5-9 5.2.2 Principal Calculated Nuclear Characteristics of the Core 5-10 5,3 Principal Core Operating Limitations........... 5-12 5.3.1 Reactor Power Level ...'5-12 5.3.2 Control Rod System.. . 5-13 5.3.3 Liquid Poison System.. . 5 16 5.3.4 Reactivity Coefficients 5 ' 5.3.5 Reactivity Additions During Core . 5-17 Alterations. 1.

5. 3 /6 ' Reactivity Additions During Power 5-18 Operation i

6, 0 Plant Safety and bbnitoring Systems............., 6-1 61 6.1 Reactor Safety System 6.1.1 General Features..................... 6-1 6.1.2 Reactor Safety System During. Power operation 6-2 6.1.3 Reactor Safety System Bypass.. 6-6 6.1.4 Related Systems 6-7 6.1.5 Ope rat ing Requi rement s.................. 6-7 62 Control Rod Withdrawal Permissive System. 6-9 6.2.1 Interlocks........................ 6-9 6.2.2 Operating Requirements.. . 6-9 6.3 Refueling Operation Interlock System......... .69 6.3.1 Reactor Refueling System. 6-9 6.3. 2 Re fueling Ope ration Cont rols...........,,.., 6-10 6.3.3 Operating Requirement s,,................ 6 10 6.4 Plant Monitoring Systems...............,, 6-10 6,4.1 Process Radiation Monitoring Systems...... . 6-10 i 6.4.2 Area Monitoring System........ . 6.......... 6.4.3 Operating Requirements.............. 6-13 6.5 Radioactive Waste Disposal Systems...........,. 6-15 6.5.1 Airborne Radioactive Wastes 6-15 6.5.2 Liquid Radioactive Wastes , 6-15 6.5.3 Solid Radioactive Wastes................. 6-16 6.5.4 Operating Requirements.................. 6-16 11

CONTENTS (Contd) l Pan Jo. i .71 70 Operating Procedures......... 1 7.1 Basic Operating Principles 7.2 Procedural Safeguards. 7.2.1 Detailed Operating and Emergent, Procedures 7.2.2 Administrative Procedural Controls . 7-J 7.2.3 Operational Review Procedures.............. 7 3 7-5 7.3 Normal Operation 7.3.1 General......................... 7-3 7.3.2 Cold Start-Up After Extended Shutdown........ . 7-3 7.3.3 Hot Start-Up ......................7-4 7.3.4 Normal Power Operation 7-5 7.3.5 Extended Shutdown,, , 7-5 7.3.6 Short Duration Shutdown.. . 7-6 7.4 Refueling Operation... 7-6 7.5 Maintenance.... . 7-7 7.6 Operational Testing of Nuclear Safeguard . 7-8 Systems........ 8.0 Research and Development Program (Phase II) . 8-1 8-1 Fuel Irradiation Program 8-1 8.1,1 Development Fuel Design Features . 8-1 8.1.2 Instrumented Assembly Design ,65 8.2 Performance Testing. 8-5 8.2.1 Core Performance and Transient Tests 8-5 8.2.2 Sequence of Testing. 86 8.2.3 Analysis of Typical Tests................ 8-8 8,3 Reactor Operating Limits . 8-11 8.4 Operating Procedure. 8-12 8.5 Special Review Procedures,.. . 8-12 lii

1.0 INTRODUCTION

1.1 SCOPE 1.1.1 These technical specifications set forth the principal desiga features and operating limits and requirements which have r2 effect on the safety of operation of the Big Rock Point Nenlear Plant (the plant). 1.1.2 Sections 1 through 7 present the plant operating limits (including original fuel) through the period of the full-term license. In the order of presentation, Section 1.0 concerns the introduction, Section 2.0 concerns the site, Sections 3.0 through 6.0 cover the plant 6 d its systems, and Section 7.0 presents procedures for plant start-up and procedures for normal and emergen:y operation of the plant. This section also includes administrative and procedural safeguards to the extent that these have a potential effect on nuclear safety. Section 8.0 describes Phase II of the research and development program, including parametric variations and operating limits on R&D fuel during Phase II. 1.1.3 The dimensions given in these specifications cf the plant design features are subject to normal manufacturing tele ances. 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 The following terms are defined only for the purpose of clearly indicating the intent of the various provisions given within these Technical Specificaticas. 1.2.1 Power Operation - is any operatien other than cold shutdown with the reactor vessel closures bolted in place and when reactor criticality is possible. 1.2.2 Refueling Operation - is any operation with sny of the reactor vessel closures open during which either core alterations are being made, or other operations which might increase the reac-tivity of the core are in progress. 1.2.3 Cold Shutdown - is a reactor condition involving ei ther no fuel in the reactor or a condition meeting the following requirements: (a) The control rods are fullv inserted in the core, and their withdrawal circuit is locked by menns of the mode selector switch in the shutdown position to prevent withdrawal. The 1-1 i J

l e 3 1.2.3 (Contd) key to the mode selector switch must be in the possession of the Shift Supervisor. (b) The reactor coolant system is at atmospheric pressure. (c) The core shutdown-reactivity control margin requirement has been demonstrated in accordance with Section 5.3.2 (b). 2.0 SITE 2.1 LOCATION - shall be in Charlevoir. County, Michigan, shall be about 4, miles northeast of Charlevoix, Michigan, and shall be about 11 miles west of Petoskey, Michigan. 2.2 BOUNDARIES - shall surround about 600 acres,. and the nearest landside property line shall be about 2680 feet, and the. nearest shoreline property line'shall be about 200 feet from the plant's containment sphere. 2.3 FRINCIPAL ACTIVITIES - shall be those associated with the operation of the Big Rock Point Nuclear Plant and may include other activities directly connected with the generation, transmission and distribution of electric energy. 2-1 1 ) l

s 3.0 REACTOR CONTAINMENT Reactor containment shall_ consist of an externally insulated ] spherical steel vessel, hereinafter referred to as containment i vessel, sphere or enclosure. The reactor, recirculation piping, ] pumps, steam drum, fuel pool, and equipment for removal of shutdown heat shall be among the-items of equipment included in c the containment sphere. 3.1 CONTAINMENT SPHERE DESIGN PARAMETERS SHALL BE AS FOLLOWS* - 1 Design Pressure, Internal, Psia - 41.7 J besign Tcaperature Rise, OF 190 i (Coincident With Design Internal Pressure) j Design 14aximum Temperature, F 235 q l Design Pressure at Minimum Temperature, 41.7 j Maximum Internal, Psia i 1 Wind Load ASA Std A58.1 i l Without Snow Load (Basic Wind Pressure = 30 Psf) With Snow Load 60 Mph Snow Load ASA Std A58.1 i (Max = 40 Psf at i Top) s 3-1 q l l ., _. ~,, ~.... _.... _ ...._.-.~..__.._.-..,_._.._.......,...m..

l 3,1 (Contd) Lateral Seismic Acceleration, Percent 5 of Gravity (Coincident With Dead Load and Snow Load Only) 1 Design External Pressure Psig 0.5 (Not Limiting, Safe External Pressure is 1.22 Psig)- Design Maximum Ambient 130 ) Temperature, *F j Design Minimum Ambient 45 Temperature

F Permissible Air Leakage Rate at 41.7 Psia 0.5 at Ambient Temperature, Percent per Day of Free Volume (Including All Penetra-tions) 3.2 CONTAINMENT _ SPilERE DIMENSIONS Diameter, Feet 130 Height Above Grade, Feet 103 5

. Approximate Free Volume, Cubic Feet 9.4 x 10 3.3 CONSTRUCTION - The principal material of construction shall be SA-201 Grade B, fi'rebox steel produced at SA-300 specifica-tions. Design and construction shall be in accordance with ASFE Boiler and Pressure Vessel Code, Sections II, VIII and IX, as modified by the applicable ' nuclear code cases. The Charpy impact rating of the parent metal shall be appoximately 15 foot-pounds at -50*F. 3.4 PENETRATI ONS_ 3.4.1 Design Features _ Thall Be as Follows: (a) Total Number 100 (b) Total Number of Access Air Locks 3 (c) Size of Nozzle Penetrations: Maximum Diameter Inches 24 Minimum Diameter, Inches 3/4 3-2

l R 3 4.1 (Contd) (d) Types of Penetrations: Pipe Site Same as Nozzle Size Direct Weld Pipe Size fhaller Than Nozzle Weld to Nozzle Size Cap Hole Nonrigid Penetration Connections Bellows Seal All Electrical Conductor Pene-Hermetically Sealed trations Connector or Com-pound Filled Nipple 3 4.2 Methods of Closure (a) Access air locks shall have two in-series gasketed doors, whichshallbeinterlockedtoinsurethatat-4ehstone door is locked closed at all times when containment integrity is required. (b) Lines open to the free volume of the containment sphere shall 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 shall be capable of being closed by manual 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 portion of the reactor recirculating loop are treated in the manner described in the previous paragraph, with the added feature that the two valves shall be on opposite sides of tne shell of the containment sphere. (d) Lines normally closed have only a single valve. A lock, interlock, or operating rules shall insura that this valve is closed whenever containment integrity is re-quired. (e) Certain lines enter and leave the containment sphere without any openings to the containment sphere free k,, volume. Others leave and return to the containment j / '. sphere without any openings to the atmosphere. Buch lines shall not require isolation valves, provided the lines are not in danger of being broken as a result of a reactor system rupture. (f) The two 24-inch ventilation openings, one for supply and une for exhaust,shall be designed to close within six seconds after any scram signal. In order to prevent the 3-3

- =- I 3 4.2 (Contd) possibility of excessive external pressure on the con-tainment sphere due to atmospheric changes, the two valves in the ventilation supply line shall be automatica,lly opened whenever the differential pressure exceeds 1 pai, overriding all other signals. The valves shall reclose when the internal pressure is still slightly below atmos-pheric. 3 4.3 Operating fwquiremente (a) Normally=epen lines,. carrying fluids out er the containment e sphere, shall be closed automatically upaa a signal indi-cating high containment sphere preseure or low water level in the reactor vessel. These automatie isolation valves shall also close upon instrument air or power failure, and upon manual trip from the control room. (b) Normally-open lines, carrying fluids into the containment sphere, shall be equipped with check valves to prevent backflow upon loss of inward pmpellent force. In addition, l these lines shall be capable of being secured by manually operated gate valves or by air-operated contml valves. The latter shall close upon air or power failure, with exception of the supply line to the control rod drive hydraulic system. Valves in this control rod hydraulic system line shall fail open to insure continuous water supply, and backup isolation shall be provided by integral valves in the control rod drive + pumps. (c) Clostra times on actor-operated isolation v'alves shall be j at follows: ] l Description Closists Time (Seconds) Main Steam (M0 'f050) 60 Main Steam Drain (MO 7065) 75 35 Poer-INCIDWFf SPRAY SYSTEM Containment effectiveness shall be supplemented by a containment sphere post-incident spray system in the event of an accident involving loss of coolant from a primary systes rupture. 351 Design Features Shall Be as Follows: t.. (a) Number of Dets of Spray Nossles 2 (b) Capacity of Sprays, Gpm per Set 400 (c) Nossle Pressume, Psia 100 3-4 1 i

3.5.1 (Contd) (d) System Actuation 1 Set Automatic, 1 Backup Set Manual (e) Signal Used to Actuate High Containment Sphere Pressure (f) Signal Trip Setting 2 Psi Above Atmospheric (g) Reserve Water Supply Lake Michigan 3.5.2 Operating Requirements (a) Automatic operation of this spray system involves a 15-minute time delay after system actuation to allow the oper-ator to override a possible spurious actuation. This time delay feature may be manually overrider to actuate the spray system prior to the expiration of the 15-minute period. (b) Water addition to the containment sphere must be manually stopped before the accumulated water level reaches an ele-vation of 596 feet. (c) The proper operation of the automatic valves and associated controls of the spray system shall normally be functionally tested at each major refueling shutdown, but not less fre-quently than orce a year. 3.6 CONTAINMENT REQUIREMENTS Containment sphere integrity shall be maintained during power operation, refueling operation and cold shutdown conditions except as specified by a system of procedures and controls to be established for occasions when containment must be breached during cold shutdown conditions. 3.7 CONTAINMENT SPHERE LEAKAGE TESTING For the purpose of this specification, leakage rate is defined as the percent of the contained atmosphere (weight basis) which escapes per day (24 hrs) under the defined pressure conditions through any letke in the containment boundary and all isolation valves and their associated piping. The maximum allowable integrated leakage rate shall not exceed 0.5%/ day of the containment atmosphere (weight basis) at the design pressure of 27 psig. The procedure for containment sphere leakage testing shall be: (a) At least once every 6 months, the personnel lock, the equipment lock and the sphere supply-and-exhaust ventilation valves shall be pressurized, with air to 20 psig, to test their leak tight-3-5

f i t r ness. The sum of leakage rates from these valves and locks shall' be less than 0.257. per day of the containment atmosphere (weight basis) at 20 psig. (b) At least once every 12 months, the following valves shall be tested for operability from both the manual and automatic modes of operation and, at the same time, shall be tested fer leak tightness by means of a pressure test utilizing air or the cormal working fluid at a pressure not lees than 20 psig. Main bteam Isolation (MO 7050) Main Steam Drain (MO 7065) Cleanup System Resin Sluice (CV 4091) Reactor & Fuel Pit Drain Isolation (CV 4027 - CV 4117) Reactor Enclosure Clean dump Isolation (CV 4031 - CV 4103) Reactor Enclosure Dirty dump Isolation (CV 4035 - CV 4103) All significant leaks revealed by these tests shall require repair of valve seals and retests. Automatic controls and instrumentation associated with these valves shall be tested at approximately quarterly intervals; these tests may be conducted with simulated signals or in such other manner as to obviate plant shutdown. (c) At least once every 12 months the following shall be visually examined for evidence of cerrosion, cracking or deterioraticn All Electrical and Accessible Piping Penetration Nipple Welds All Accessible Piping Welds to Nipples All Expansion Joints and Welds on Expansion Joints Fotting Compound in All Electrical Penetrations t Insulation at piping penetration welds shall be removed to permit _ visual examination. The probable cause of any significant corrosion, cracking or deterioration revealed by such visual examination shall be determined, and evaluated in terms of likelihood of recurrence and probable effect upon other containment sphere penetration, components. An individual component leak detection test shall be performed with air'at 10 psig on the faulty component prior to its repair or moditication. The faulty component, and other components if necessary,'shall be repaired or modified..and an individual component leak detection test performed with air at 10 psig upon each repaired or modified component. All components so repaired or modified shall be visually re-examined at' appro-priate intervals, but not less frequently than once every six months, until the adequacy of annual visual inspection is re- ) established to the operator's satisf action. 3-6 1 i 1 1 a-,- ,r,,g.>-e. w wr p,.-g y- ,,w,,v.,- ,-,-,.ggw-c,ay,,y,- p-c,-s -.-u +*-a-+=+= e-+."-e'='&-*'-ww-d

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\\ 4 After cutting into the sphere or its components, or any dis-assembly of components that would affect sphere integrity, an individual component leakage rate or an integrated leakage rate test, whichever is deemed more appropriate by the operator. shall be performed, with air at a pressure not less than 10 psig. l l It sball be permissible to employ a leak detection test in lieu of the above for insuring containment integrity following dis-assembly of the emergency condenser or the gasketed, bolted closure of the coaxial cable electrical penetratiens. The individual component leakage rate determined from the above tests when combined with the previously measured integrated leakage rate shall yield an overall leakage rate not greater than 0.5%/ day at 27 psig. A containment sphere integrated leakage rate test of at least (d) 24-hours duration shall be performed at a pressure not less than 10 psig. Although routine maintenance m.ay be performed, repairs to items listed in 3.7(a), (t), and (c) shall not be made immediately prior to or during the test. The accuracy of the leakage rate measuring syste.m shall be verified (1) by superimposing a contr:11ed leakege rate equivalent to the allevable leakage rats at the test pressure (measured through a gas flow mete:) upon the existing leakage rate and continuing the test a suf ficient period of time to measure the composite leakage, or (2) by other means of equivalent accuracy. If the leakage is in excess of 0.5%/ day of contained atmosphere (weight basis) at the design p:essere (27 psig) or extrapolated to the design pressure, repaire shall be made and the leakage-rate test repeated until the 0.5% per day of centained atmosphere (weight basis) specification is met. An integrated leakage rate test shall be conducted on the contain-(e) ment sphere within 90 days after the effective date of the full-term operating license, in accordance with the procedures set forth in subsection 3.7(d) abcve. On the basis c,f the rerulte of such test, and other available relevsnt data, the operator etall, prior to 1966, propose a frequency of intes:ated leakage rate testing covering the rentining tern of the full-tenn operating If the proposed frequency te acceptable to the Division 1 license. of Licensing and Regulation, the provisions of this Section 3.7 shall be amended accordingly. Otherwise, such integrated leakage ) rate testing shall be repeated in 1966 and cnce every two years thereafter pending agreement upon a different frequency. 3-7 q ~

i (f) If the results of an integrated leakage rate test show leakage above specification, the required repairs shall be performed. The integrated leakage rate test shall then be repeated until I the 0.5% per day (by weight) specification is met and another integrated leakage rate test shall be performed within 1 year; i except that if the excessive leakage rate is attributed solely to the containment components specified under Section 3.7(a) the integrated leakage rate test need not be repeated within one year, but the frequency of testing required by Section 3.7(a) shall be doubled for that period of one year. (g) All leakage rates determined by tests at pressure less then 27 psig shall be corrected by the following extrapolation factor to establish the leakage rate at design preesure: L P ~l u e g 1,t P2-1 u t a where L, = % leakage rate, at extrapolated pressure L = % measured leakage _ rate, at air test pressure g P = extrapolated pressure, atmospheres, absolute P = test pressure, atmospheres, absolute = test viscosity of air, ar test pressure and u temperature u = viscosity of air-steam mixture at pressure a and temperature of the accident condition. 3-8

4.0 REACTOR AND POWER SYSTEMS EQUIPMHIT k.1 REACTOR SYSTD4 EQUIPMENT The reactor system equipment shall consist of the reactor vessel, the steam drum, the safety relief valves, the reactor recirculating pumps, the shutdown cooling system, the emergency cooling systems and the interconnecting piping and valves, k.1.1 Reactor Vessel (a) Design Features Shall Be as Follows: Length, Overall, Feet 30 Inside Diameter, Inches 106 Wall Thickness, Excluding Cladding, Inches 5d Cladding Thickness, Minimum, Inches 5/32 Design Pressure, Psia 1715 Design Ibmperature, F 650 Approximate Initial Nil Ductility 10 Transition Temperature, OF (b) Principal Materials of Construction Component Material Specification Vessel Shell and Heads Steel ASTM SA-302 Grade B Flanges and Nozzles Steel ASTM SA-336 Cladding Stainless Steel Types 308 and 309 Head Studs Steel ASTM A193, AISI h340 Head Nuts Steel ASTM SA-19h, AISI 4340 (c) Reactor Vessel Penetrations Nozzles Number Diameter, Inches Location Coolant Water Inlets 2 20 Bottom Steam-Water Mixture Outlets 6 14 Shell Shutdown Heat Exchanger Outlet 1 6 Shell Access Ports in Top Head 3 10 Head Control Rod Drive Penetrations 32 h Ecttom Liquid Poison Inlet 1 3 Ecttom draergency Core Sprey Inlet 1 ?' Shell Vessel Vents 2 3 Head In-Core Flux Monitor Penetrations 8 2 Ecttom Instrument Nozzles k 3 Shell Seal Leak Monitor 1 1/2 Shell Flange 4-1 9. .,p

k.1.1 ( Co;;td ) (d) Vessel Closure The top head closure shall be a bolted flange with a 106-inch inside diameter. The seal shall be made by double 0-rings of the self-energising type. The space between rings shall be connected to a leak monitoring systesi. (e) Vessel Support ' Die reactor vessel shall be suspended vertically by 2 5-inch diamster rods attached between vessel support brackets and a series of support bracket segments which shall be anchored in the supporting concrete. The horisontal loads shall be resisted by four stabiliser brackets at the vessel flange and four stabilizer brackets near the bottom of the vessel. Thermal expansion of the vessel and connected piping shall be accoessodated by the above supports and by floribility of the piping support system. (f) Vessel Internals The major internal components of the vessel shall be ar-ranged as follows: The core support plate shall be held by support pads in the bottom vessel head. The support tubes and fuel channels as well as the fuel assemblies shall be supported by this plate. he thermal shield shall be supported on six pads located on the vessel wall. The top guide shall be supported by the thermal shield. h e top guide shall position and hold down the channels and support tubes. The steam baffle and emergency cooling sparger shall be supported from pada located on the vessel wall. All components of the internal support structure of the reactor vessel shall be fabricated of Type 304 stain. less steel. (g) Control Rod Guidance 9he 32 seueifofun-shaped control rods shall be stided and supported by the " fuel channel and support tube" assemblies. + 0.00 heh control rod shall contain 8 rollars of 0 567 - 0.015 inch diameter, the bottom four of which shall move in a mintsum intersupport tube space of 0.658 ineh and the top four of which shall move in a minimum interfuel channel space of 0.628 inch. (h) Irradiation of Reactor Vessel Samples Irradiation of reactor vessel samples vill be undertaken to further verify the expected change in impact values due to irradiation effects on the reactor vessel and to further the knowledge of the irradiation effects on reactor vessels. h;

4.1.1 (Contd) (i) Operating Requirements The controlled rate of change of temperature in the reactor vessel wall shall be limited to approximately 100* F per hour. Reactor vessel pressurization in excess of 20'6 of normal operating pressure shall not be allowed to occur at temperatures below the maximum established nil ductility transition temperature plus 60'F. Nil duct'il'ity transition temperature calcula-tions shall be made at least once each year. 4.1.2 Primary Coolant Recirculation System The primary coolant recirculation system shall consist of the reactor vessel, the steam drum, the reactor recirculating pumps, the interconnecting piping and valves, and the safety relief salves. (a) Design Features Shall Be as Follows: Number of Recirculating Loops 2 Number of Reci rcula. 1 Loop Approximate Internal Volume of 3830 System Excluding Reactor Core and Internals to Isolation Valves, Cubic Feet Approximate Volume of Coolant in 2689 System During 157 Mwt Operation, Cubic Feet i Steam Drum: Length, Overall, Feet 40 Inside Diameter, Inches 78 Wall Thickness, Excluding Cladding. 4-3/8 ) Inches ] Cladding Thickness, Minimum, Inches 5/32 Design Pressure, Psia 1700 Design Temperature, 'F 650 4-3 j y, e

~~ a + h.l.2 (Contd) i Recirculating Pumps: Type Single-Stage, Centrifugal, Motor-Driven Rating 16,000 Gpm @ 76 Feet of Head Seals Mechanical, 3 In-Series Recirculating Valves, Each Loop: Mode of Size Opening Rete Location Type Operation Inches Inches / Minute Pump Suction Gate Motor 2h 12 Pump Discharge Gate Motor 20 5 Pump Discharge Gate Motor 5 12 Bypass Valve Pump Discharge Butterfly Motor 20 1.5* Deg/See Recirculetion System Piping:, Location Size Number Vateriel Risers lk" x 0.712 Wall 6 316 Stainless Steel (SS) Downcomers 17" x 0.858 Wall k 316 SS Pump Suction 2k" x 1.21 Wall 2 316 SS Pump Discharge 20" x 1.009 Wall 2 316 SS Safety Relief Valves: i i 6 Number Type Spring-Loaded Maximum Setting of First Valve 1700 Including Rupture Disc, Psia i Sequential Pressure Increment 10 Setting of Remaining Valves, Psi liinimum Capacity per Valve (1202 Psia 2.36 x 105 Setting), Pounds per Hour Valve Orifice Arce, Square Inch 3.976

The limit of operation of each butterfly valve shall be adjusted to produce a 50% reduction in flow from the full-flow condition with the correspondin6 pump in operation.

kk l

.4 h.1.2 (Ccntd) Reactor Power Operation Cooling: Coolant Material Demineralized Water Type of Cooling System Forced Becirculation System Pressurization Boiling Water Minimum Loops Operating Con-1 eurrently (or Equivalent) Number of Passes Through Core 1 Flow Direction Through Core Upward Reactor Shutdown Cooling: Design Pressure, Psia 315 Design Temperature, P 42 5 thamber Pumpe e ak=har Heat Exchangers 2 Esat Emoval Capacity per 80-a 10 ' loop, Btu /Hr z lleactor bergency Cooling: Dnergency Condenser: Type Tank With Tube j ..hndlea Nwnber of Tube Bundles 2 Minimum Capacity per Tube 16 x 10 Bundle, Btu /Hr binun Cooling Time Available , From Water Storage, Houra " ' ' ' ' ^ Number Systems Providing 2 Makeup Supply Design IDxternal. Pressure of 41 7 Tank, Psia Design Pressure of Tube 1700 Bundles, Psia h-5 w tt-y w 4 + - - w M wwr ye'7

4.1.2 (Cont d) Minimum Time to Put System 30 in Full Operation Following Sign al, Seconds Core Spray System: Type Sparrer Ring With Spray Nozzle Capacity of Sprays, Gpm 400 sozzle Pressure, Psia 115 Core Spray System Recirculation: Number Pumps 2 Number Heat Exchanger 1 6 lleat Removal Capacity, Btu /llr 8 x 10 0 28.4'F Log Mean Temper-ature Difference (b) Operating Requirements A minimum of one reactor recirculating loop or its equivalent shall be used during all reactor power operations when reactor power level is above 1.0 Ekt. The maximum operating pressure and temperature shall be the same as the reactor vessel. The controlled rate of change of temperature in the reactor vessel shall be limited to 100* F per hour. All other components in the system shall be capabic of following this temperature change rate. The safety relief valves shall be set ap-propriately for all planned reactor operating pressures so that the allowable pressure of 1870 psia (1700 plus 10'5) 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. The primary coolant shall be sampled and analyzed daily during periods of power operation. The following are absolute limits which if exceeded shall necessitate reactor shutdown. Corrective action will necessarily be taken at more stringent limits to minimize the pos-sibi,lity of these absolute limits ever being reached. 4-6 l

s 4.1.2 (Contd ) Conductivity (Micromho/cm) Maximum 5 Maximum transient

10 pH (Lower and U~pper Limits) h.0 and 10.0 Chloride Ien (Ppm) 1.0 ltullibrium Mclogen Pedio-25 octivit;- (Uc/ml)

Porcn (Ppa) 100 h.l.' Pricry Syctela Shielding Reactor shielding is ordinary concrete with a density of 3 approximately150lb/ft. Thickness varies in plan and elevation to suit structural requirements. The' shielding thickness directly opposite the core shall be approximately 9 feet, 6 inches. The control rod drive room, which is directly beneath the reactor, has ordinary concrete valls which shall be approximately h feet thick. A removable concrete plug, 3 feet, 6 inches thick, shall close the opening above the top of the reactor. The stc u drum, risers and downcomers are primarily shielded by ordine.cy concrete valls which shall vary in thickness from h feet, 9 inches near the bottom to 3 feet, 3 inches at the tep. A large section,12 feet by 42 feet, of the steam drum enclosure vall serves as a blevout panel and shall contain high density, loose aggregate to'a thickness of approximately'h feet, 9 inches. This provides the shielding equivalent to k feet, 9 inches of ordinary caicrete. The reactor shielding shall be cooled by a water-filled jacket at the in*Lde face. The cooling water system shall be designed to remove 60,000 Btu per hour with the inlet water temperature at 680F. Cooling water shall be supplied from the closed loop reactor cooling water system. The jacket shall be a carbon steel, annular tank divided into eight segments, with water entering the bottom j and leaviug at the top. Provisions are made to convert to air

cooling, i

Conductivity is expected to increase temporarily efter stertups from cold shutdown.

The maximum transient value here state ' io the meximum permicciM ? and applies only to the period subsequent to o cold shutdown betueen criticality and 24 hours after reaching 20% rated pouer. h-7 1 J l i l

i t k2 POVPR SKT. F51Igr*0' AND ASSOCIATED FACILITIEF 4.2.1 Electrical System ) l (a) Auxiliary Power 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 transmission 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 solt bus. .The reactor safety system and related circuits are fed ) from three 120 volt a-c buses. Each of two buses is supplied from a different 480 volt system through its ovn motor-generator set. Each motor-generator is equipped with a flywheel to sustain operation during momentary power system disturbances. The thirti bus is supplied from the 125 volt d-c system through a static inverter whic.h supplies power to the control rod position indicating system i and to neutron monitoring channel No. 3 The 125 volt d-c battery system also furnishes power for other critical services including: Liquid Poison System Controls Motor-operated Automatic Containment sphere Taolation Valves i Containment Sphere Ventilation System Isolax.m Valves Emer8ency Condenser Drain Valves Safety Bystem Annunciators Baergency Lighting Switchgear (b) Emergency Power Deergency power shall be provided by a 200 kv diesel-. generator which shall be automatically started on loss of auxiliary power. As soon as the diesel-generator reaches its rated potential of 480 volta, power shall be available automatically to service the following ega13 ment:. Containment Sphere Access Air Locks Fire Protection System Electrically Driven Ptas:p Emergency Lighting j Instrument and Control Transformer 2B The power requirwoonts of the above items arv. appmxi-mately 100 kv. Other electrical services may be energired mamially as required. k8

h.2.2. thin Condent.cr (a) Design Features Shall Be as Follove: Type Radial Flow Surface Condenser With Deaerating Hot Well Condensing Surface Area, ,27,500 Square Feet Design Condensing Pressure, 15 Inches Hg Absolute Condencing Capacity, Founds ,000 per Hour @ 1 5 Inches Hg Absolute Condensing Capacity During Full 948,000 Load Rejection, Founds per Hour Air Ejector Capacity 10 Cubic Feet per Minute of Air Plus 1.1 Pounds per Hour. of Hydrogen Plus 8.3 Pounds per Hour of Oxygen (b) Operating Requirements The following condenser presrure trips shall be operative during reactor power operations when system pressure is 350 peig or higher: Annunciate, Inches Hg Absolute 50 !05 Reactor Beram, Inches Hg Absolute 8.0 1 0 5 Turbine Trip and Bypass Valve 10.0 I 0 5 Closure, Inches Hg Absolute k.2 3 Turbine Bypass Control System (a) Design Features Riall Be as Follows: Flow Capacity at 1015 Psia, Pounds 739,000 per Hour Flow Capacity at lk65 Psia, Pounds 963,000 per Hour Maximum 8 peed, Full Valve Stroke, Approximately 0.2 Seconds 4-9

4,2.e .tidemnte and Feed-Water System l Two,1000 gpm, half-capacity condensate pumps of conventional design shall be provided to pump condensate from the condenser through the condensate system to the suction of the reactor j feed pumps. The condensate system shall b; designed for 200 psig and 300 F. Three, half-capacity, mixed bed ion exchangere,.. designed to each pass 1000 gpm, shall be provided for removal of reactor solids carry-over and turbine-condenser system corrosion products. Condensate leaving the. demineralizers shall paan through two low-pressure, horizontal, U-tube feed-water beaterm to the suction of the reactor feed pumps. There shall be two,1600 gpm, horizontal, centrifugal,.aotor-driven reactor feed pumps. Feedwater shall then be returned to the steam drum through.t high-pressure, horizontal, U-tube feed-water heater, feed-water control valve and check valvg. The feed-water system shall be designed for 2000 peig, 375 F. 4.2 5 Reactor Cooling Water System The reactor cooling water system shall be a closed cooling loop utilizing inhibited demineralized water to remove heat from the fdlldwing pieces of equipment: ~ Reactor Shield Cooling Panels Beactor Clean-Up Nonregenerative Heat Exchanger Reactor Shutdown Heat Exchanger Fuel Pit Cooling Water Heat Exchanger Miscellaneous Sample Coolers Reactor Recirculating Pump Coolers Two reactor cooling water pumps, each rated at 1500 gpm, shall be provided to pirc 21 ate water through two heat exchangers, each ratedat9x10DBtu/ hour,andthenthroughtheaboveequipment. 4.2.6 Fire Protection System In addition to furnishing water for conventi =3 et fire-fighting equipment, the fire protection system shall furnish water as follows: Core Spray Cooling System Containment Sphere Post-Incident Spray System Backup for Service Water System to the Containment Sphere One electric fire pump and one diesel fire pump, each rated at 1000 gpm, 254-foot head, shall be provided. The fire protection system shall be operable and'ydsedy for service during power operation and during refueling operation. ' bl0 i 4 + -. - - ~,. _. _.. _ _, __j

d.2,7 Ventilation Sys.t.e.m Ventilation shall be provided throurh two fuli-ca; t each rated at 30,000 cfm, located in 1:m ventilt.t Ventilation air to and from the cont ainn.:nt-sphere suli N via equipment located within the ventilating romc. .i. .r is located outside the containment sphere and shali c...ain the isolation vlaves, air heating equipsnt, filt. n~J necessary cont rols. The filters shall be prerided foc elt.n-irg inlet aii. Flow of air shall be fren areas of Ic. m st antar, it : centanination prsebability toward areas of highest probability, anJ then out the stsci,. 4.2.8 Reactor Service Water System ili3 hc d Cooling of the reactor cooling water sr.nen shell L by two full size heat exchangers,. each cated at Dx ,t u /:,c. t.. nd ct supplied with service water fron Lake *:i chir,an vi: w ;ter pt2mps. 1.2.9 Instnimant and Service Air Systc.m nst ra::ent cad service air shril bc supplied by t %. c.on - labricated air compressors, each rm. J :t 70 scf.. it' m:9; air shall also pass through a dry r. J. 2.10 Tu rbinc-Cencretor Unit Tus turfine shall b 3600 rpr. tv i -ccc rour.d. dce' ' - J i m, c(:.de r. sing unit directly conn:.ctv; n. ' roten-en i , c.. ra t u r. The turbine-gener.tcr e n. i:.cluding i,

ts.

s. auxiliaries, shall be desigr.cc. foi e : atio: e.' t e t. ano uttuu. T!,c it.iii n i ratirn of the turbir.e r ' 5 4.500 la - 1% prin, st.turated, and 3.5 inches .4 Ma a h m u *. 4 The gs.neratur initially shall be. r n '..+.. '. 7t,583 kr

t: a 0.85 pouci factor and 30 psit hydro.

. r.; su i s. '11 2 tuib:nc Gell r.orrally be op :. : ' t L.it.tt... r- ~11e t re; ol. 'cr, which pemits the n to m n.it of t!.t reacter. During st: 11 -t .. J u. r om ' r tl o:: et lee lo.a s, operat i c, u ct trcl pc :.nitted. hi.?.?..S._t.o.r a..,%. In g,.noral fuel hanJ1ing shall be et,

)-

l,) guiuance and visual observation ci ai; h, d!i, ?- m. tions. Water shall be ustd as th'. b::. ;c :. m.. i tu. c.. c, for the t ransfer of I rradiated fut 1 in i.c.n th re.ct < 4-11 i .~. -m...

J e L.2.ll (Contd) storage pool. A lead shielded transfer cask, with associated vinch, shall be used for thin operation, in conjunction with the semigantry crane. In addition, a jib crane and two monorail cranes may be used for moving fuel. Storage of fuel shall be restricted to the following areas: (a) New Fuel Storage New fuel shall be stored in the desi6nsted dry-storage racks provided at elevation 632.6 inside the contain-ment sphere. Racks with locking covers shall be pro-vided for storage of 102 fuel bundles. The rack array is 6 by 17 on a square pitch with a spacing of 13 inches between bundle center lines. This provides adequate spacing to prevent accidental criticality even if the area should be flooded. New fuel stored in these racks shall be limited to that described in Sections 5 and 8. Movement of fuel into and.out of the storage racks shall be restricted to one bundle at a time. (b) Spent Fuel Storage Spent fuel shall be stored in the designated storage racks located in the 29-foot deep spent fuel storage pool. Two rack assemblies shall be provided, one with a 6 by 12 array on a square pitch, and the other with a 6 by_8 array on a square pitch, thus giving a storage capacity of 120 fuel bundles. Spacing for all racks shall be 12 inches between bundle center lines, thus providing adequate spacing to prevent accidental criticality. Fuel stored in these racks shall be limited to that described in Sections 5 and 8. Movement of fuel into and out of the storage racks shall be restricted to one bundle at a time. k-12 i l l

i S. O REACIOR CORE AND CONTROLS The reactor core and controls shall consist of all the components of the reactor core including tne reactor reactivity control system and the liquid poison system. The expected configuration, of the core is shown in Figure S.I. 5.1 , DESIGN FEATUR.E.S The dimensions for these design features are for the cold con-dition and are subject to normal manufacturing tolerances. S.I.1 Principal Core _ Materials AFuel (Sintered Pellets or UO 2 Compressed Powder) Moderator and Reflector Light Water St ructural Components (Fuel 304 SS and Incoloy 800 Cladding Will, in addition to 304 SS and Incoloy 800, include Zr-2 and Inconel 600) Flow Channels Zircoloy-II and 304 SS Control Rods B C Filled 304 SS Tubes 4and 304 SS Rods in Cruciform Shaped 304 SS Sheaths S.I.2 Control Rods Number of Control Rods: Type 1 16 Type 2 16 Poison Material B C Powder in 304 4 Tubes and Solid 304 SS Rods ] Pitch (Square Array), Inches 10.466 Active Length, Inches 68 Shape Cruci form Width, Inches 11-1/2 5-1 .-, - - _. -,..,... ~... - - _.. _ ~,

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l l gx (. f 5.1.2 (Contd) 1 Blade Thickness, Inches 5/16 ) Sheath Thickness, Inches 1/16 Number B C Filled, 304 SS Tubes per a Control Rod: Type 1 116 Type 2 88 Nuaber of Solid.304 SS Rods per Control Rod: ( Type 1 0 Type 2 28 304 SS Tube, B C Filled, Outer 4 Diameter, Inches 0.175 304 SS Tube, B C Filled, Wall 4 Thickness, Inches 0,020 Solid 304 SS Rod, Outer Diameter, 0 175 Inches 5.1.3 Control Rod Drives The core shall be controlled by 32 control rods with a pitch of approximately 10.466 inches,. which shall be moved vertically within the core by 32 hydraulically operated drives, The hydraulic pressure for positioning the drives during normal rod movement shall be supplied by one of two positive displace, ment pumps, which shall discharge into the hydraulic system at about 200 psi above reactor pressure (Pr.200) for normal contral movement. The drive mechanism used for both normal operation and scram shall be of the locking piston design. Only one rod at a tiec can be withdrawn from the core. Rods may be inserted into the core singly or all may be scrammed together, To scram a control rod, high-pressure water irca enc t ve independent sources r shall be applied to the und m.a.e e: U.c lu aien and water from the top side of the piston shall flow to the scram dump tank, The two independent sources of scram hydraulic pressure shall be: 5-4 . u

5.1.3 (Contd) (. J Accumulator Pressure i uh of t he ~ accum# aiori,:L..!l wr.

c., ne t., in th: form of water under nitaga p re s uu rt foi sc ramming cont rol rod drive The minimum gas volume in each accumulator, during power operation, shall be 600 cubic inches.

(b) Reactor Pressure - Reactor pressure, if above 4M psig, shall supply energy for scramming the control rod drive if accumulator pressure falls below reactor pressurci The control rod drives shall be mounted on the bottom of the reactor vesel and attached to the bottom of the poison section with a coupling. The half coupling, attached to the control rod drive, shall consist of six finger-like segments which form a hollow spheroid. The other half coupling shall be an integral part of the poison section. The force required to engage the coupling shall be approximately 60 pounds. Once engaged, the coupling can be disengaged only be the deliberate movement of an unlocking mechanism. Control and inst rumentation shall be provided to enable the operator to make a coupling integrity check. Poisition indication at each locking position shall be provided by stationary magnetic switches actuated by a permanent magnet in the movabic drive piston, Drive position indication shall be displayed in the control room by a digitial readout system. The center tube of the drive mechanisais a well containing the position indicator probe. This area is at at mospheric pressure. The probe mounts a series of hermetically scaled, magnetically operated switches, each of which indicates c discrete rod postion. The switches are ope rated by a pe rma - nent magnet carried by the drive piston, The intervening walls are of nonmagnetic material, allowing each svitch to be operated as the piston passes. Extra switches are prtvided at each end of the stroke to indicate limits. A thermocouple is placed at the top of the position indicator probe of each control rod drive and is held in position by spring clips. The temperature readings from these thermo-couples provide an indication of the seal operating conditions and detection of flow from reactor to scram dump tank (via Icaking scram valve). Control rod drive features shall be as follows. r. i.. t 5-5

-i 5.1 s (Lantuj St5p Length Between Locking 3.0 Positions. Inches Normal Withdrawal Velocity, 3.0 In./Sec Normal Insertion Velocity.

3. 0 In./See

Scram Insertion Time, Sec 90% of Stroke 2,5 Maximum r

10% of Stroke 0.6 Maximum Principal Materials of Type 304 55 17-4 Pre-cipitation liardened (11 1100) SS. Inconci X (Spring Material) and Graphitar 14 (Seals) 5.1.4 Liguid Poison System Mate rial Sodium Pentaborate. Solu-tion Available Quantity of Setution, 650-Gallons Sodium Pentaborate in Solution. Wt Percent Minimum Concentration 19 Maximum Concentration 30 Initial Injection Gas Pressure, 2080 s Psia Sy ?.t Ac t u a t i o-. Remote Manual lype of Injection Valves Explosive. Electrically Operated Supplied by Conax Corporation i S 1.5 General Core Comges,it cn Q The data in this section present general design features of the original and research and development fuel that shall make up the physical composition of the core. t

The intenal from the time a scram signal causes the scram solenoid valves to de-energize until the rod has traveled the specified distance of its full stroke length, 5-6

5.1,5 (Contd) (a) Enrichment,,o,f_ Fuel, approximate weight percent U-235 from 2.6 to 3.3!~2 n'c'1u sive, (b) General Core Data Number of Fuel Bundles in Core 84 Total Weight UC in 84 Bundles, Pounds 29,300 2 Moderator to Fuel Volume Ratio 2.65 Equivalent Core Diameter, Inches 76,54 (c) Fuel Bundles The general dimensions and configuration of the two types of fuel bundles shall be as shown in Figures 5.2 and 8.1 of these speci-ficat ions. Principal design features shall be essentially as follows: Research and ,Gencrud Original Fuel hcVilopment Fuel Geometry, Fuel Rod Array 12 x 12 11 x 11 Red Pitch, Inches 0.533 0.580 Standard Fuel Rods per Bundle 132 109 Special Fuel Rods per Bundle 12 (4 Special Fuel 12 Rods at Bundle Corners are Segmented) Spacers per Assembly 3 7 Fuel Rod Cladding Mate rial 304 SS 304 SS, Zr-2, Inconci 600 and/or Incoloy 800 Standard Rod Tube Wall, Inches 0.019

0. 010 t o 0.030, Inclusive Special Rod Tube Wall, Inches 0.031 0,010 to 0.030.

Inclusive [.upl, Rods Standard Diameter, Inches 0.388 0.425 Special Rod Diameter, Inches 0.350 0,320 U02 Density, Percent Theorti-cal 94 + 1 90 to 95, Inclusive ~ Active Fuel Length, Inches Standard 70 68 to 70, Inclusive Corner 59 Fill Gas Helium Helium 5-7

T 5.1.5 (Centd) (d) Channels Number of 304 SS and/or Zircaloy 11 88 Wall Thickness, Inches: 304 SS 0.075 Zircaloy 11 0.100 Inside Width, Inches: 304 SS 6.57 Zircaloy 11 6.54 Length, Inches: 304 SS 79 5/8 Zircaloy 11 79-3/4 (e) Total Weight SupportedgC, ore Support Plate: 84 Fuel Bundles 0 420 Lb/ Bundles. Lb 35,280 88 Support-Tube-and Channels Assemblics 0 100 lb/ Assembly, L5 8,800 86 Orifices 0 10/ Orifice, Lb 860 2 Channel Plugs 0 10 lb/ Plug. Lb 20 1 Flow Distiibutor Assembly, Lb ,2.,500 Total Weight, Lb 47.460 5.1.6 Sources Type Ant imony Be ry lliu.T. Quantity 2 Location

The sources shall be placed in core positions 02 59 and 09 52

) as shown in Figure 5.1. If the required three counts per second count rate with a 3:1 signal-to-noise ratio cannot be obtained with the sources in these locations it shall be permissible to use posi-tions 02-52 and 09-59 bunimum Initial Strength 1660 Curles Total Physical Description The sources shall consist of a steci Jacketed antimony pin. 1 inch diameter by 12 inches long, centrally located on the vertical i

With in vessel low IcVel neutron detectors in service, one operat ine. source R

may be, temporarily relocated as the operator deems appropriate, 5-8

1 ] .r.l.6 (Contd) axis'of a stcol jacketed (Type 304 SS) beryllium cylinder 51/J i UD' by 16 inches long. The entiro assembly,' including support st ructure. is a cylinder 79-7/16 inches long by 6 inches diameter which rests o;. a special ori ficc in a standard _ support tubo and fuel channel. A lif tin; .l bail shall be provided for: handling purposes. The assembly dcsign shall allcw adequate cooling along the surface of the source pin cr.d the oute ' surface of the assembly. 3.2 PRISCIPAL CALCULATED TilERMAL, ilYDFfl,'LIMND NUCLEAR ~~ U. l.XiiXL._ft!T.d. 5.'f.liC.s...~~~ ~ ~ The data presented in this. sectior. are nominal values bared on analyses of the core design. The specific limitations uPon opera-tion of the core are given in Section 5.3. 5.2.1 Principal calculated Thermal and llydraulic Characteristics of tIEi!s4-liUii3Tc' Ccife"!Ea~di'n{ ~ ~~ ' ~ ~ ~ ~ " ^ ' (a) Core Powc.r at Rated Steam Flow, Mwt 240 (b) Peaking. Factors (To be applied to llent Flux) : (i) Overall at Rated Power (includes Gross 2.83 and Local) (ii) Over power- (Steady' State and Transient (Cffect s) 1 22 (iii)Tutal-(Product of (i) and (ii)) 3 15 (c) lleat flux and Fuci Center Temperature: Average Maximum Maximum 4 122*. 9 240 Mwt 0 240 Mwt of 240 St Fuci !acket licat Flux 2 (litu/ilr-ft ) 116,500 329,000 402,000 Cladling 620 720 760 Fuel Center, Temperature, L

l-1,380 3,350 4,050 luc i Rod Power, Kw/Tt 3.8 10.7 13 1 (d)

Bu rnout Ratio, Minimum at Overpower 18 [ 4 (c)

Maximum f uel Cladding Stress, Psi 72,200

. ( f) Average Core Power Density at 240 Mwt,Kw/L 46 (g) Stability Margin at 240 Mwt, 1485 Psig Degrees Phase Margin 28 (h) Total Recirculating Flow Rate, Normal for 240 Mwt, 1235 Psig, Lb/llr 12.5 x 106 (i) Reactor Core flow Rate, Percent of-Total Recirculation Flow Rate 98 (j) Corc-Inlet Conditions: Inlet Velocity i

Phnre 1 Development Powder Fual 5-9 l

1

s 5.2.1 (Contd) Maximum, Ft/ Set 4.5 Minimum, Ft/Sec 2,7 Inlet Subcooling, Btu /Lb 19.5 (L) Reactor Core Pressure Drop, Normal for 240 Mwt, 1235 Psig

5. 7 (1)

Reactor Core Maximum Exit Bulk Temperature, "F 572 (m) Steam Volume Fraction: Average Core Exit Fraction 0.55 Maximum Channel Exit Fraction 0.63 Average Fraction Over Core Length 0,31 5.2.2 P,rin,cip,a1 Calculated Nuclear Characteristics of the 84-Bundle Core (a) Te,mperature and Void Coefficients The following. temperature and void coefficients have been calculated for the 84-bundle core, These values are based on calculated characteristics of the developmental bundles to be inserted together with experimental data obtained relative to the characteristics of the original fuel. Data relating to individual research and development fuel types are included in Section 8,1,1. kbderator Temper,ature Coefficient (21 keff/k gfper

F) e 68*F 550*F Start of Cycle

+0,2 x 10-4 1,5 x 10'4 End of Cycle + 0,6 x 4 - 0,9 x 10-4 yoid Coefficient (tg keff/keff per Unit Void Within the Channel) 68'F 550* F Start of Cycle ~E'2 6*^~" *~ Qi','2'0 ' ~~" End of Cycle -0,10 -0.10 200R.U,. Cog,f,{i,c,,ient ( k /k _per

F)

(D) I Fuel Temperature *F Moderator *F (4 kdQ F), 68-68, 0% Voids -1,47 x 10-5 1323 550, 0*6 Voids -1,03 x 10-5 1323 550, 20% Voids - 1.15 x 10- 5 5-10 -o+- y %3 y-,w,,,y,--,r,*wr,-- y ,,r=-ww we y-t = te 'y* t-'

e

&+M +e---nf- =i*e-e e'==-*--- s =---4+

5, ? 2 (Contd) (c) Re act ivity,Ba1,ance keff 07625 Temperature Voids 0.030 Xenon and Samarium 0,025 Fuel Depletion and Maneuvering 0,.050* Total Ak Required 0 130 heff ~~ilo.glication Factor, Cold Condi-Multi (d) -.n. All Rods In, 18 SS Channels 0.93 All Rods Out, 18 SS Channels 1,13 31 Rods In, 1855 Channels 0.98 (c) Maximum Control Rod Worth Zi keff/keff Cold 0.039 550'F, No Volds 0.042 550*F, 20*6 Voids 0 030 (f) Maximum Reactivity Addition Rate (Based on Withdrawal of the Rod Worth 0.042 4keff/keff at the Maximum Withdrawal Rate of t hree Inches per Second),dkeff/keff 0.0052 per Second (g) Worth of Ligi,d Poi, son 68'eff, Normal Water Level in Reactor and ok 0 30 F, 2000 Ppm Boron 4kgff, Normal Water Level in Reacter and 0 19 550 F,1300 Ppn Boron Maximum Time to Bring Reactor Suberitical, 5 Hot, Minutes Time Required for Poison to Reach Core After Activation of the System, Seconds 35 The effect of removing SS channels is not included in this number. Renaval of 18 SS channels would add 0,050 dkeff. 5-11

~ \\ l 5.2.2 (Contd) (h), Expected Fuel Burnuj Average FNd/ Ton of Contained U ] Original and Phase I Fuel Types 10,000 Phase 11 Fuc1' Types 15,000 5,3' PRINCIPAL CORE OPERATING LIMITATIONS \\s 5.3.1 Reactor Power Level (a) Refueling The reactor power shall be limited to 1.0 Mwt. exclusive of core decay heat, during operations with the reactor vessel closures open. (b) React or_,0pe raQon, The reactor operation, relative to the original fuel. shall be so limited as to be consistent wit h t he nc.st conservative of the following:

Minimum Core Burnout Ratio et Overpower 1,5 Transient Minimum Burnout Ratio in Event of I.5 Loss of Rocirculation Pumps From Rated Power 2

Maximum lleat Flux at Overpower Btu /lly-It 539,999 2 434.000 Maximum Steady State fleat Flux, Btu /llr-f t Maximum Fuel Rod Power at Overpower. Kw/Ft 17 2 1 Maximum Steady State Fuel Rod Power. Kw/lt 14.2 Stability Criterion: Maximum Measured Zero t o-Peak Flux Amplitude, Percent of Average Operating Flux 20 Maximum Steady State Power Level, FW1 240 Maximum Value of Average Core Power Density 60 0 157 Mw(t), Kw/L t

Based on correlation given in " Burnout Limit Curves for Boiling Kater I:cartors."

by E.lanssen and S. Levy, April 14,1962 (APED 3892). 5-12 w ,=,-w-* ~.- mi*--- v a + -s .r em---e v e u-+ -., - - -me -w-v-..- va---- e, i-w~ ~~-

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y 5..L.1 (;;:Ad) Maxamem Value of Average Core Power Density 4e 0240 Hw(t), Es/L Maxin.um Reactor Pressure Daring Power Operation, 1485 Psig Minimum Recirculation Flow Rate, Lb/lir 6 x li? Maximum FNd/ Ton of Contained Uranius for sa Individual Bundic 23,503 Rate of change of reactor power during pcwcr opcra;icr. Control rod wit hdrawal during power operat icr. shall be s'u. t hs t t!e average rate of change of reactor power i< less than 50 fL: }er minutc when power is less than 120 Mut, less than 20 M.t per minute when powc r is between 120 and 200 Mwt, and 10 PM per ninut e whe n power is betwcon 200 and 240 Mwt," 5.3.2 Control Rod System (a) Control Rod Performance The following limits shall apply to any cont rs) red @ ch m be withdrawn. It shall be permissi ble t o t:q, ar,d t ; the hydraulic drive water to a fully inserted cont r oa which is defcetive or does not meet t he se li v.i t s pr o : dt a t c.e remaining rods do mce: the limits The following test s shall be periorracd at c a ch n.aj or r o ut. m shutdown and at least quarterly caring periods of pc.:;; operation: (i) Continuous withdrawal and inser t 2ca of c ach drive c.c: its st rokc with normal hydraulic e st<. pres: m Minicum withdrawal tinte shall be 23 sc;onds (ii) h'ithdrawal of cach drive, stopping at each 1,cLiny position to check latching and uniatching operatiens and the func 'ioning of the posit icn indicat io. >>; r: (iii) Scram of each drive from the full withdrawn positier Maximum scram time from system trip to 90 percent of inscrtion shall not exceed 2 5 seconds. The following test shall be performed at cach cajor refue h of, but not less frequently than once a year: 5-13 v y . y, -e ,,n ,,w,,---- -,,--r- ,,w r-- ~,, -a,,, e,--<, w w w

i s 5.3.2 (Contd) Insertion of each drive over its entire scope with reduced hydraulic system pressure to de-termine that drive friction is normal. (b) Core Shutdown hbr, gin Verification The reactivity of the core loading shall be such that it is always possible to maintain keff at less than 0. 997 with the most valuable reactivity-worth control blade completely withdrawn from the core. The core shutdown margin shall be verified by a demonstration that the reactor is suberitical with the most valuabic reactivity worth control blade fully withdrawn, plus an immediately adiacent blade withdrawr. to a position known to contribute 0.003 x,ff or more to the effective multiplication. In the event that the maximum reactivity condition occurs at a temperature greater than ambient, the demonstration will either be performed at that temperature or a suitable additional margin will be demen-strated at ambient. This verification shall be performed prior to start up af ter any shutdown in which the system has cooled sufficiently to be opened to atmospheric pressure and any of the following situations exist: Fuel has been added and/or repositioned in a way which is not definitely known to reduce reactivity-, or Any steel channels have been repladed by Zircaloy channels during the shutdown; or 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 previous margin demonstration. 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 Drive Temperature Limit - The limiting component of the drive is7stTd To~r~dIiifai~380'F. The internal temperature of the control rod drives as meas red by a thermocouple located in the drive position switch well normally shall be less than 350*F. It shall be permissible to continue operation of the reactor, if drive temperature instrumentation becomes defective, provided it can be established by other means that drive temperature is normal, 5-14 4

i .5.3.2 (Contd) (d) Control Rod Following Checks - Checks shall be made to demon. strate that the ccIntrol roTpoison sections follow the move-ment of the control rod drives. Two methods of verifying rod following are available. The first is the coupling integrity check, which involves fully withdrawing a rod and observing that the drive will not go to the overtravel pcsition., Drise overtravel is possible only if the drive and poison sections have become separated. This method cannot always be con, veniently applied after the reactor is critical. Tac second method is the response of nuclear inst rumentation to rod movement, which has general application after k,ff reaches approximately 0.995. The application of these methods of verifying rod following shall be as follows: Routine Cou ling Integrity Checks Each cont rol rod shall be verl ie uTing the couplih'giittegrity check prior to the commencement of each refueling operation and/or series of low-level critical tests with the reactor closure head removed. Control Rod Following Verification During Reactor Op,eration - During each approach to HII'i'c~aTi'ty7dnt rol^76df with'dEada before keff reaches 0.995 (determined on t he basis of pre-dictior:s or suberitical multiplication measurement s) shall be verified by a coupling integrity check at the time the rod is fully withdrawn. Each control rod which is withdrawn is equal to or greater than 0 995 shall at any time keff be checked to verify proper rod following before the total worth of the withdrawn portion of the rod and any other unverified rods reach 0.016k (as determined froit the ap-proximate Control Rod Worths resulting from the initial calculations and start-up test data). Verification shall consist of either a couplirgintegrity check or the obscr-vation of nuclear instrumentation response to rod withdrawal, (e) Control Rod Exercising During Sustained Power Operation - Each co nt r o'l~BcT~WIIcTi s e i t%'r~ya'r~ tis 11 p' 6T e offil ei'cTy~di t hd rawn. shall be exercised at least once each day. (f) Abnormal Behavior of the Control Rod Sy' stem - An immediate and thorough investigation shT1T b71ndof Ihe occurrence of any e abnormal behavior (including inability to verify any control blade) of the control system to determine the cause and safety significance of the occurrence. The reactor shall be shut down unless: 5-15 l i

t 5. 3. (Cont d) (1) It is determined by the investiration that eny .n)- function which has occurred neither it.ipnirs th. 1.bility to cont ral the reactor nor indier.tes the ivniner.t ir:- pairm.:nt of the perfer:ance of additional ecycncats of the c;ntrol rad syster. (2) The operating h; draulic water to the defective control rod has been tagr.cd and valycd cut to prcycat.itndrawal of the control rod after an nttet.j t has bear r.a!. t o in-se rt th; eentrol roJ. (3) The coec shutdcwn ar"in requirt.nent (descri!:cd ia (b) aba c) caa be met with the rensi:.ing operile c; Trol rods. 1:valt.. tion o f this rec.uiret ent shall le '.: sed cn previous exrcrirental measurent. ate. (.") .M.in._i m. um Accura.c,.y ( f nod Insitio_n_ Indi.c..at_i_ne Sven - The ......g. cont rol rod drive r esaons itsel t u L d:, screte positions ove r it.s t rave l. The position inEcating nes< r is o# n digital readent de.:irn and indicates the drive mition over a range of 1.3 inches for cach position vf tN drive. Control rods shall not be s110'ead to renain in position he re t.':e indict. tion is not operable. If all inJication for a parti-cular control rod in lost, the rod shall not be nond until the inJiention h: been restored, exec;:t for scrar. 5.5.3 Liquid Poison Sy m t The liquid poisc.n syste. shall i.e avai141c for or; ret ie.: at all tiacs during refueling and power operation. Toe repetcr shall be shut down in nny situation where the poison solutic-tant level drops below an equivalent of 850 callons 19 weir.ht percent sodiu: pcutaborato or where tL poison solution sterara ter.;n rat' arc dro, :. to less than 5'F atave saturation tenperature. Thc "xire a l l e.. - able concentration rhsil be 30 weight pare:nt of sufium pents:hernu. The minimum worth of the liould poison syster (bueJ c.i nr rral water level) shall be 23'idieff. Cc: ponents of the ryster shall be checked at one to two month intervals fo: prot.cr o;. ratica except for actuation of the injection valves. The licaid noison systcc shall be used at any time rLen suberiticality cannet be assured by the norcal shutdown rechanisn. Inicction %I1 he continued until a minimu:a shutdown nargir. of 0.01 A..ffM er; is assured in the most reactive core. The reactor shall uit be operated after poison has been injected until the boven con-centration is the reactor water has been reduced to 100 rr, or lus. One squib (from different operatin? Valve grour; and valce position each time, in rotation) shall be ren.oved am' test fired at least every 12 months. 5-16 1

L, 5.3.4 M tivity Co_efficients. During initial' loading of the core, the moderator temperature and void coefficicnts shall be measured as indicated in Section 7.0 of these spec 1fications. The reactivity coefficients shall meet the' following requirements: 4 (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.will, constitute en adequate demonstre.- tion of the negative character of the void coefficient. t (b) The moderator temperature coefficient (inferred from critict.1 control rod position) during uniform heating of the core shall /k f added be limited such that the potential maximum 4 k rf byheatingthemoderatorisalwayslessthanonedoT(ar. e (c) The overall effect.of increasing reactor power at constant pressure shall be a loss of reactivity whenever the reactor is operating so as to produce a net steam flow. The only condition under which a positive effect could be produced would be low temperature, in which case no steam voids would be. formed. The increase in moderator temperaturc~ E could then produce a 'small positive c'ffect. 5.3.5 Reactivity Additions During Core Alterations ~ The limits and requirements which apply to reactivity additions are as follows: i i (a) The procedure for core alterations which cay increase re-l activity shall utilize the procedure outlined in 5.3.2(b) before and after the alteration to verify the core shutdown i margin of 0.3%dkeff/k ff, with the most valuable rod com-pletely withdrawn. Al,l rods shall be fully inserted durin-i the reactivity addition. Checks shall be made at-frequent intervals during core alterations to assure that the core shutdown margin tequirement is being met. j l (b) Core alterations which increase reactivity shall be limited between subcriticallity checks to fuc1 loading increments which do no exceed the placement of one fuel bundle or the exchange of two Zircaloy chann els for two steel channels. At no time will the shutdown margin be knowingly allowed to reactivity worth fully wiIkkrawn. the control rod of highest be less than 0.3%d k,ff/k witt j from the core. 5-17 I i

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,'q l 5.3.6 Rc a c t i v i t y Addi t i ons Duri ng,,Powe r Ope r,at i on Rout ine Lant rol rod wit hdrawal sequences shall be c st ablished ' for use du rinr, normal powe r operation. These wi11 br in a s q'arnec of f. ingle not c h stc ps invoiring nr.m ement of only om-cont rol rud at a t f m. ) The following criteria will apply to the rod withdrawal re, quences Whever the reactor is critical, the withdtrol of t he next not ch in seque. nee shall not cont ribut e t.c ? t thu e.305 j the sc-Ak g/L kk. The worth of any cer:t rol blade n.aved it: i quence sli i be 1imited such t ha-t ht comptet c wit hdrewa1 of Ihe f blade f rom it!. pres (nt posit ion w,11 cent i abut e not morc t han 0.025 4 k gr/keft-t l t i l 5 18

6.1.2 Reactor &fety System During Pcwer Operation The following tabulation gives the arrangement of the reactor safety system that shall be effective during power operation: Trip Scraa Contacts Coincidence Setting Warning Sensor and in Each in Each azzi Instrument Annunciation Trip Device Channel Channel Tolerance Special Features Ranges Trip Set P;1nt 1 1 25 5 psi above 50 5 psi above 100-1700 psig High reactor 2 1 out of 2 pressure (k pres-reactor opera-reactor opers-cure switches) ting pressure ting pressure 1.ow reactor water 2 1 out of 2 Elevation 610' Closes containment Fixed level level (4 level 6"11 inch sphere isolation trip point no evitches) valves. If reactor range pressure is less than 200 psig, actuates core spray system. (Bote: Spray water will not enter reactor vessel until reactor pressure drops below fire header pressure.) 1 8.0 0 5 inches -30" to +30" -k" below 1 ww cteam drust 2 1 out of 2 wet.cr level (k below operating water operating level level switches). level insin steem line 2 1 catt of 2 50!5 percent Position switch - backup isolation of full closure trsins adjust-valve closure (4 able full valve ,pocition switches) travel 8.0 0 5 inches Bypassed by pressure 0 - 30" Hg'vac 1 High condenser 2 1 out of 2 i praesure (4 pzveL of Hg absolute interlock as described - cure switches) pressure in Section 6.1 3-l 4

i '6.0 FLalE s"Tr.. UC MONITenING SYSTEMS The plant safety and monitoring systems shall be considered to encompass the reactor safety system including related features, the control rod withdrawal permissive system, the refueling op-eration interlock system, and the plant monitoring systems. 6.1 REACTOR SAFETY SYSTEM 6.1.1 General Features The reactor safety system shall consist of sensing devices and associated circuits which automatically initiate a reactor scram or other tequired action. Certain of the sensing devices shall also initiate automatic closure of the cont &innent sphere isolation valves, while other sensors shall initiate emergency cooling of the reactor through operation of the emergency condenser or through operation of the core spray system. Controls shall also be available e in the control room to permit manual initiation of penetration closures, and manual initiation of emergency cooling of the reactor. The reactor safety system shall use two parallel safety channels. These channels shall have separate power supplies and separate chains of sensor trip contacts. The channels shall be designed on the fail-safe principle (de-energizing will cause a scrae0 The reactor safety system shall be 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. The safety system response time (the interval from the time a sensor trip contact operates until the sc:am solenoid valves are de-energized) shall be less than 100 milliseconds. The contain-ment sphere ventilation valves shall be designed to close within 6 seconds after any trip which initiates reactor scram. If a sensor is disconnected for maintenance this will automatically cause a trip on the safety channel to which'the sensor is connected. 6-1

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\\ 6.1.2 (conta) Trip JSt:ren i Contacts Coincidence Setting Warning l Sensor an1 in Bach in Dich and Instrument Annunciat m Trip Device Omnwl Channel Tolerance Special Features Ear;tes Trip Set D-g, 2.0 O.2 pai Closes containment 0 - 20 psig High enclosure 2 1 out of 2 i pressure (4 pree-above atmos-sphere isolation sure switches) pheric valves. (Two inde-pendent pressure switches actuate a time mechanism that initiates post-incident spray system in 15 min-utes unless the control is manually overridden.) 5/161 /2 inch Alams at level of Fixed level at -10"I1/2" below High scram dump 2 1 out of 2 1 tank level (4 below tank 10 inches below tank trip point tastt center level switches) center line center line. (no range) 11mm 3bsition switah Recirculation 12 1 set out Approximately ' ' adjustable for line valves of 2 10 percent of ' closed (two esta full simulta-full valve of 6 position neous closure travel switches in ameh of both dis-of 2 channels) charge or both suction valves or rimultaneous closure of the butterfly valves to the positions comparable to a 55 percent decrease in flow from full flow or any con-bination of two of these valves, one in each loop.

4 6.1.2 (CorPA) Trip Sert Contacts Coincidence Setting WarnM; Sensor and in Zach in Each and Instru-ent Annun e '"m Trip revice Channel Channel Tolerance Special Tentures Rang s Trip Set Nint - 5220vo](bs Closes all autcrant-I Lose of auxiliary 1 1 out of 1 power supply } ically actuated con-(voltage relay) l tainment sphere isolation valves. i High neutron flux 3 2 out of 3 12o15 percent Interlocks prevent O to 125% (S (ench of 3 power of the O to 125 control rod with-range flux monitors percent scale, drawal as described has a trip contact or 3812 per-in Section 6.2.1. in each channel) cent of the O to 10 percent i 4 i scale (any range) a Protection against 3 1 downscale 5 percent low 4 picosmmeter cir-1 upscale trip and 1 ad-cuit failure ditional in-strument at 1PO% high trip I lO 2 second Placement of any 2 out -100 see to 15 see inter-Short period 2 1 out of 2 (each of inter-period of 3 range switches of infinity to mediate range neediate range the high level neutron +10 see (There is also Log-N period flux channels in the an annunciator monitors has a power range position operated by trip contact in will bypass this period-the start-up each channel) trip feature. channel period 4 contacts.) Manual ecram 1 1 out-of 1 (1 evitch) n

t l 1 6.1.2.1 Start-Up Channels - Channels 6 and 7 shall' provide'logarith- ) mic neutron flux level and period information from source { 1evel to rated power. The principal components in each-J chanac1 shall be a neutron detector, high-voltage power supply, current pulse amplifier, a five decade log count ) rate meter with low level period amplifier, log. count rate j i indicator, and log count rate recorder. Gas-filled Boron-lO J b lined proportional counters with a sensitivity of approxi-p rr.tely 12 counts /nv shall be used as detectors. Provisions E shall be made for remotely positioning the detectors. By 1 moving the detectors away from the midplane of the core, 'l i I their effective range msy be extended several decades. A i l short period on either channel shall be annunciated in the control rocaa. 6.1.2.2 Intermediate Ranae Channels - Channels k and 5 provide I loge.rithmic neutron flux level and period inferination frost l ,I approxinastely 10-k% to rated power. The principe.1 com _ ponents in each channel shall be a neutron detector, dual ~ hig.1-voltage power supply, Log-R and period amplifier, Iog-N I iniicator, period iniirator, and Icg-N recorder, he detectors shall be gama compensated ion chrabers with a sensitivity of i j approximately 4 x 10-14 amperes /nv. 6.1.2 3 Power Range Channels - Channels 1, 2 and 3 shall' providt. l linear neutron flux. level information from epproximately 10' % to 125% rated power. The principal components in each channel shall be a neutron detector, dual power supply, picoammeter with console range switch and power level indicator, and power level recorder. he detectors shall be gamma compensated ion chsrabers with a sensitivity of k x 10-14 amperes /ny, connected to the reactor safety system. 6.1.2.4 In-Core Flux Monitors - An in-core flux monitoring syste::: consisting of eight channels shall be provided to indicate and record linear neutron flux level at twenty-four dicerete positions within the reactor core. Each okanrel abr., ' consist of e vertical string of three detectors approvinate.1p evenly spaced axially within the core. h e detectors shall be g miniature fission chambers with a sensitivity of 3 x 10 aareres/nv. Each fission chamber shall be conasA ed to an unplifier and indicator with a 0 to 150% power seale. A I separate power supply shall be provided each channel, he monitoring system shall be calibrated so that a 100% power indication corresponds to the maximum rated heat flux at 157 Mwt. An indication of 110% power on any of the 24 monitors shall be annunciated in the control rotaa. e I b-5 s

e 6.1.2.5 Neutron hbnitoring Range Switch - The range switches are used in conjunction with the flux amplifiers to provide a selection of seven ranges of percent power. The range switch unit contains a selection switch which has a combination of resistors and capacitors connected to it to form several resistance-capacitance circuits. The unit also contains a trip circuit which is essentially identical to the trip circuit in the flux amplifier unit and is used for downscale trip. 6.1.3 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. A key lock reactor mode switch shall be provided with the five positions ; " shutdown," " refuel," " bypass dump tank," " start-up" and "run." These positions shall have the following functions: Mode Selector Switch Position Trip Functions Bypassed Run None Start-up None (e) Bypass Dump Tank (a) Low Steam Drum Water Level Recirculation Waterline Valves Closed Steam Line Beckup Isolation Valve Closed Ifigh Water Level in Scram Dump Tank (b) liigh Condenser Pressure Refuel (d) Low Steam Drum Water Level Recirculation Waterline Valves i Closed Steam Line Backup Isolation Valve Closed High Condenser Pressure i Shutdown None (c) (a) Control rod withdrawal is prevented by interlock while switch is in this mode position. (b) Bypass of this trip function is necessary to enable emptying the dump tank after a scram. l 6-6

o m 6.1.3 (Contd) (c) With the mode switch in the " shutdown" position, both the scram circuit and the control rod withdrawal circuit are open. The ventilatius duct circuit power supply is trans-ferred to a point which provides penetration closu:e pro-tection through signals from "high containment sphere pressure" and " low water 1cvel in reactor vessel." This pereits normal ventilation in the containment sphere during shutdown when the control rods are held in the full-in position. Hone of the reactor' safety systco signals are bypassed since there is no need to withd.w control rods. (d) dith the mode switch in the refuel position and the crane positioned over the reactor vescet, cranc operation is prevented if any one rod is eithdrawn from full-in position. (e) High condenser pressure trip is autocatically bypassed any time reactor pressure is below 350 psig. 6.1.4 Related Systems (a) Core Spray System Control The core spray system shall be automatically actuated by simultaneous tripping of the reactor safety eystem sensc: " low reactor water level" along with the " low reactor pressure" trip device. The " low reactor pressure" t:1p device consists of a pressure switch interlock which prcvents admission valve opening while the reactor pressure is cbeic 200 psig. (b) Emergency Condenser Contrel_ A pressure switch shall initiate automatic operaticn of th2 emergency condenser if the reacter pressure reaches 10Q:10 psi above the reactor operating pressure. 6.1.5 Operating Requirements (c) The reactor safety systcm shall be operable during power operation as indicated in Section 6.1.2 This systen shall be functionally tested at each major refueling shutdown, but not less frequently than once a year. All portiens of this system which can be made temporarily inoperative with-out either requiring plant shutdown or jeopardy to nuclec: safety shall be checked at least once a month. (b) The core spray system and emergency condenser control initiation censors shall be functionally tested at eech major refueling shutdown, but not less frequently than once a year. 6-7

a 6.15 (contd) (c) Both start-up range neutron monitoring channels shall be operable and measuring flux from the core during'the reactor start-up, prior to the power level at which the intermediate or power range channels become operative. However, if one of the two start-up range channels becomes inoperative.during the course of start-up and prior to the power level at which the intermediate or power range channels become operative, then it will be permissible to hold the control rod pattern and power level attained until both channels are again operable. (d) There shall be a minimum of two intermediate range channels providing logarithmic neutron flux level informaticn and proximately 10~gtectionduringreactorstart-upfromap-period scram pr % to approximately 5% of rated power. For reactor power operation above approximate 5% of rated power, the logarithmic neutron flux level information and period scram protection are not required in accordance with the specified effective arrangement of the reactor safety system given in Paragraph 6.1.2. (e) Any one of the three power range flux monitors may be taken out of service for maintenance during reactor opera-tion. If one monitor is out of service, a trip on either of the two remaining monitors shall scram the reactor. When maintenance is necessary, no major changes in power level, flux distribution or the control rod pattern shall be made. (f) Normally, during power operation at power levels above 180 Myt, at least 12 of the 24 in-core flux monitors shall be operating including at least one monitor in the central channels (2 and 4) or at least one monitor-in any adjacent pair of six outer channels (1, 3, 5, 6, 7 and 8). (g) Protection against a'" cold-water accident" is provided by recirculation pumps and valve interlocking. The valves on either side of the recirculating pumps are interlocked with pump power such that each va3ve must be in its proper position before the pump motor can be started. If the suction valve to the pump is closed, the motor will be tripped. If the discharge valve and bypass valve are closed, the motor vill be tripped. (h) Minimum nuclear instrumentation in operation during all shutdown conditions (except cold shutdown defined in Section 1.2) shall be the same as established for re-fueling operations (defined in Section 1.2), except that only one start-up range monitor is required. .6-8 s .,m....

3 6.2 00FTROL ROD WI'ItIDRAWAL PERMISSIVE SYSTD4 6.2.1 Interlocks Interlocks shall prevent control rod withdrawal when any of the following conditions exist: (h) When any two of the thirty-two scram accumulators are at pressure below 700 psig. (b) When two of the three power range channels read below 5 Percent on their 0 to 125 percent scales (or below 2 percent on their 0 k0 percent scales) when reactor power is above the minimum operating range of these channels. This interlock may be bypassed when all three of the power range channels are set on the minimum operable range. (c) When the scram dump tank is bypassed. (d) When the mode selector switch is in the shutdown position. 6.2.2 Operating Requirements The control rod withdrawal permissive interlocks shall always be operable. No further withdrawal of control rods vill be permitted if one of these circuits is found to be inoperable. Permissive circuits shall be functionally tested at each major refueling, but not less frequently than once a .en. year. 63 REFUELING OPERATION INTERLOCK SYSTD4 631 Reactor Refueling System All of the trip devices not bypassed by the mode selector switch in the refuel position shall be operative during all refueling operations. This shall include the sensors and trip devices of the reactor safety system as specified for power operation as follows: High Reactor Pressure Low Reactor Water Level High Containment Sphere Pressure High Scram Dump Tank Level Loss of Auxiliary Itver Supply High Neutron Flux Short Period Manual Scram o-9

i 6.3.2 Refueling Operation Controls (a) Interlocks shall be provided to prevent all motion with any of the refueling crancs (nar IV, jib cran e, trenn er c cask winch, and monorail crane) which are positioned over the reactor vessel whenever any control rod is not fulle inserted in the core and the mode selector switch is in the " refuel" mn' tloa. (b) Keylocks or interlocks shall be provided within the speed controls of the refueling cranes (jib crane, transfer cask winch, and monorail crane) to limit their lowering speeds at elevations 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 each niajor re-fuelivg shutdown and shall be maintained in the speci-fied condition during all refueling operations, (b) The refueling operation controls including position interlocks and travel speed controls shall be functionally tested at each major refueling shutdown. 6.4 PLAVT MONITORING SYSTEMS The plant monitoring systems include the process radiation monitoring systems and the area monitoring system. 6.4.1 Process Radiation Monitoring Systems The process radiation monitoring systems consist of the air ejector off-gas monitoring system including the fuel rupture detection system, stack-ns monitoring system, the emergency condenser vent monitor, and process liquid monitor system. (a) Air Ejector Off-Gas Monitoring System Continuous monitoring of the air ejector off-gas radio-activity shall be provided by two single-channel gamma scintillation spectrometer systems which shall be designed to detect noble gas fission prodcuts. One of these monitoring systems is an operational spare which may be used for scanning the entire energy spectrum. The sampling system shall be designed to hold up the gas sn";.le to allow time for the decay of Nitrogen-16 and other short-lived activation gases. The off-gas monitoring channels shall be calibrated so that the indicated and 6-10

. + ~ - ~ ' 7-f l2 4 ~6.b.1 (Contd) i recorded count rate output of the channel in service 4 can be evaluated in terms of microcuries of "ission r gas 1per unit volume of off gas. The off-cas flow rate r ehall be measured and recorded. The: radiation recorder r

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and off-gas flow rate' recorder shall: permit determination i of the fission gases discharge rate from 10-3 to 100 i curies per,second as measured at the stack. i At a release rate calculated to be equivalent to one I curie per second, (at time of emission from stack).a trip circuit in the. air ejector off-gas monitor shall ( cause an alarm to be sounded in the control room. At a release rate similarly calculated.to be equivalent to ten curies per second, the air ejector off-gas monitor trip circuit shall. initiate action of a time' i 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 i minutes.) (b) Stack-Gas Monitoring System 1 There shall be an isokinetic probe, permanently fixed in the stack approximately one-third up from the base to collect stack gas and particulate samples that are withdrawn with a gas pump through flowmetering and regulating equipment. The sample shall be passed through a particulate filter and a filter for iodine r sampling, both of which1are' located upstream from the-detector. These filters shall be removed and checked for radioactive contaraination as specified in 6.k.3 (b). After filtering, the continuous flow gas sample is presented to a enn+1.nucus monitoring single channel gamma spectrometer.. The equipment shall permit deter-mination of fission gas activities and gaseous activation products from 10-3 to 100 curies per second. The system shall have a high radiation alarm which shall be annunciated in the control room. The alarm setting is specified in 6.k.3(b). (c) Dmergency Condenser Vent Monitor The emergency condenser vent shall be monitored to detect a "significant release of radioactive material. Monitoring shall be supplied by two independent gamma sensitive in-strumentation channels employing scintillation crystal sensing devices. These channels shall have a range of 0.1 to 100 mr/hr and shall be provided with an. alarm which shall annunciate in the control room to inform the operator of a release of radioactive material. The alarms shall be as specified in 6.4 3 (c). i 6-11 i ,e. e4 e *

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6.k.1 (C;at i) (d) A process liquid monitor system employing gamma scintillation detector channels shall be provided to give indication of radioactivity trends in process liquid streams normally con-taining radioactive liquids and also to warn the operator of radioactivity in process liquid streams that do not nonmally contain radioactive liqpids. Alarms on monitors (1), (iv) and (v) shall be set so as to warn the control room operator via a common annunciator when concentrations are present which exceed these listed in Column II, Table II, Appendix B of 10.CFR 20. The remaining alarms shall also annunciate on this common annunciator and shall be set to alert the operator to unexpected changes in radioactivity levels. These set points will be based on experience. The procees liqpid streams which shall be monitored are as follows: (1)

Radioactive Waste System Effluent to Canal (ii)

Reactor Enclosure Cooling Water (iii) Main Condensate Demineralizer Influent (iv) ' Circulating Water Discharge (v) Service Water Return From Reactor Enclosure The radioactive vaste system effluent to canal channel.shall be used in conjunction with the liqpid vaste disposal system. Each batch of liqpid wastes to be discharged shall be analyzed prior to discharge as described'in 6.5 This monitor shall provide an additional means etrehecking the activity of the wastes being discharged. The circulating water discharge monitor s' hall monitor the . main' stream of plant effluent prior to its discharge into Imke Michigan, and ser.ve as a backup to the other liquid monitors. In addition, a continuous sample is drawn from the discharge canal for periodic analysis as specified in f .3 (d). 6.4.2 Area Monitoring System (a) Nineteen fixed gamma monitors employing scintillation type + detectors shall be installed throughout the plant, and each shall have the following: (1). A range consistent with expected radiation levels in the area to be monitored. (0.01 mr to 10 mr or O.1 mr to 100 mr or 1 mr to 1,000 mr.) (ii) An output indicated and recorded in the control room. (iii) An adjustable high radiation alarm which shall be annunciated in the control room. Alara settings that shall be as' indicated in 6.4 3 (e). f F e 6-12 k t' h a m-e --~e= + - -

e-

i e 6.h.2 (Ocntd) (b) Two of these nineteen area monitors shall be located in the vicinity of the fuel storage areas to provide gamma monitoring of the fuel storage areas and refueling op-erations. Iccal alarms shall be provided for these mon-itors, and alarm settings shall be in accordance with the provisions of 10 CFR 70. (c) At least five environmental film monitoring statione shall be provided for determining the integrated gamma dose rate in the site environs. These stations shall be placed on an arc of about 1,350 meters from the stack. 6.b.3 Operating Requirements (a) At least one of the two air ejector off-gas monitoring systems shall be in service during power operation and cet to initiate closure of the off-gas isolation valve as described below. Alarms normally shall be set to annunciate in the control room if the off-gas radioactiv-ity reaches a level that corresponds to a stack release rate of 0.1 curie per second. At stack release rates above 0.1 curie per second, the alarm shall be set approximately a factor of two above the expected off-gas release rate but in no event above that level corresponding to a stack release rate of 10 curies per second. The monitors shall be set to initiate closure of the off-gas isolation valve (after a time adjustable from 0 to 15 minutes) if the off-gas radioactivity reaches a level that would correspond to a stack release rate of ten curies per second. The cali-bration of the system and the automatic closure function of the isolation valve shall be checked at least monthly during power operation. (b) The stack-gas monitoring system shall normally be in service. Adequate spare parts shall be on hand to allow necessary repairs to be made promptly. The alarm normally shall be set to annunciate in the control room at a level that cor-responds to a stack release rate of 0.1 curie per seccnd. At stack release rates above 0.1 curie per second, the alarm shall be set approximately a factor of two above the expected stack release rate, but in no event above 10 curies per second. The calibration of the system shall be checked at least monthly. The particulate filter and iodine filter shall be analyzed at least weekly. 6-13 \\* Y

.3 b dh$ (ny.%O ~ h

it) One W the eneracncy condeneer vent monitors shall be in The monitors servics at all times during power operation.

' shall be set to alarm at approximately 10 mr above the marinum expected background during operation of the emer-A I fD The calibration shall be checked at

I.T gency condenser.

M.', least ao.thly. n .M, (d) The process liquid monitors shall normally be in service. gy,[*, Adequate spare parts shall be on hand to allow necessary Alarms shall be set as speci-repairs to be made promptly. Calibration of the 'hdioactive Weste ".,.. -,.g e fled in 6.4.1 (d). .....,i. System Effluent to Canal" monitor shall be checked at least "'[7 Calibration of the remaining monitors shall L, once a month. Each day an j' be checked at least once every three months. .g., analysis shall be made of the previous 24-hour collection of discharge canal vater. The area monitoring system shall normally be in operation; (e) however, individual monitors may be taken out of service for maintenance and repairs. Adequate spare parts shall be on Durine hand to allow necessary repairs to be made promptly. monitor outages in normally accessible areas, temporary moni-j toring shall be provided if the remainin6 area monitors do Calibration of monitors shall not provide adequate coverage. Alarm trip points shall be set be checked at least monthly. at a radiation level approximately twice the normal maximum indicated radiation level, but normally not less than one decade above the lowest scale reading. Two films, each with a minimum sensitivity of 10 mr, shall be provided at r.ach site environmental monitoring station. During operation at etack release rates of 0.1 curie per second or less, at least five monitoring stations shall be One film at each station shall be replaced and provided. analysed at least monthly..The other film shall be replaced and analyzed quarterly. Operation at stack release rates above 0.1 curie per second shall not exceed 48 hour 3 vithout at least fifteen film Two of these stations shall monitoring stations in service.The remaining additional statione be on-site near the stack. shall complement the permanent stations but shall be located One film at each statio at greater distances from the stack. The shall be replaced and analyzed at least every two weeks. y second film shall be replaced and analyzed quarterly. Under all stack release conditions, the film processor shall be instructed to report within 24 hours on all films that might indicate abnormal exposures. Gamma dose-rate measuring instruments and neutron dose-rate (f) measuring instruments shall be provided for establishing per-These instrumente vill be routinely , eissible working limits. The. instruments shall be repaired by qualified personnel. calibrated at least monthly. 6-14 8 e

MO'.3f'iUYd "E IO M D.it,E S.U. D I 6.5 1 Airborne Radioactive We { The airborne radioactive waste disposal system shall process ventilation air and other plant exhaust gases which may con-tain radioactive contaminants. These gaseous vastes shall be e discharged co the atmosphere through a 240-foot high stack. An automatically controlled damper shall admit outside air to the suction of the ventilation fans as necessary to main-tain a minimum stack flow of approximately 30,000 cfm. The several systems and their associated equipment for han-dling the normal sources of gaseous vastes are as fcllows: Noncondensable gases shall normally be removed from (a) the main condenser by one of two steam jet air ejec-The off-gas shall travel to the stack through tors. a buried holdup pipe sized for approximately 30 minutes holdup and then through a high efficiency particulate An air-operated isolation valve shall be pro-filter. vided between the filter and the stack. This valve shall be automatically closed by a signal from the off-gas monitoring system as described in Section 6.4.1 (a) and Section 6.4 3 (a). (b) Noncondensable gases shall be removed from the turbine steam seals by the gland seal exhauster. The gases shall be discharged from the exhauster to the stack through a buried holdup line sized for approximately. 90 seconde holdup. All ventilation air from the reactor containment sphere (c) and portions of the turbine building shall flow thrcugh The appropriate ducting and be discharged up the stack. equipment provided for this system is described in 4.2 7 Other minor potential sources of gaseous radioactive (a) vastes, such as vents, shall be discharged' to the stack. Ventilation air from the Chemical Laboratory and Counting Room shall be filtered and released directly to the atmosphere. 652 Liquid Radioactive Vastes All contaminated or potentially contaminated liquid vastes shall be collected in tanks and then sempled and analyzed on a batch basis to determine the concentration of radioactivity. The minimum treatment of vastes collected in the " dirty" re-ceiver tanks and the " clean" receiver tanks shall be filtra. Wastes, if released from the tion through the radwaste strainer. shall be pumped to the circulating water discharge canal

plant, 6-15 e,

---w r .w ,w,c ,..u -a, e

_~ _ ' } d,5.2 ( httif for dilution to permissible concentration if required. Where further treatment 1s necessary or desirable, the following methode shall be available: l (a) Boldup to permit decay of. radioactivity. (b) Removal of radioactivity by demineralization. i (c) Concentration of radioactivity by evaporation, (The concentrated slurry shall be stored in the concentrated vaste storage tanks for ultimate off-site disposal by a licensed contractor.) Liquid radioactive vaste system equipment shall include: two 5,000-gallon " dirty" receiver tanks, two 5,000-i gallon " clean"' receiver tanks, two 5,000-gallon vaste t' hold tanks, one 5,000-gallon chemical receiver tank, radvaste pumps,, concentrator feed pump, treated vaste pump, concentrator, strainer, filter, demineralizer and necessary instrumentation and controls. Sumps and smaller tanks for laundry draias, laboratory drains and decontonination pit drains shall also be provided. 653 Bolid Radioactive wastes-Spent demineralizer resins vill be sluiced to a shielded 10,000-gallon storage tank. An underground solid vaste storage vault shall be provided for other solid radio-active vastes. Disposition, as necessary, of vastes from the storage facility shall be via a licensed contractor. 6 5.4 Operating Requirements (a) The annual' average stack release rate for fission and activation gases shall not exceed one curie per I second. The maximum permissible stack release rate (for periods in excess of fifteen minutes) shall be-l

10. curies per second.

If the annual average stack release of one curie per second is exceeded for more than a week, steps shall be taken to reduce the re-f lease rate to the annual average. The annual average stack release rate for halogens and particulate matter (expressed in units of ricro-curies per second) shall not exceed the permissible 81**" airconcentrationsforunrestricted15'"m"3"se"cond. l / in 10 CFR 20, multiplied by 1.2 x 10 ~ c Iodine and particulate sample filters shall be re-i moved and analyzed at least weekly, i 6-16 i ,_...,~,_._._,__u.-.. __,_..._.a_.. m.. . ~ -

'l a A E.5,k (conta) Stack release rates for halogens and particulate matter shall be based on these analyses, which sha.ll be performed not sooner than 48 hours nor later than 72 hours after filter removal. (b) The liquid radioactive vastes may be released if the gross activity of plant origin in the effluent from the circulttinB vater discharge canal can be regulated so that it does not exceed, on an annual average, the limits given in 10 CTR 20. The inventory of liquid radioactive vastes in the liquid radioactive vaste disposal system shall not exceed 5,000 curies. (c) The inventory of solid radioactive vastes shall not exceed 40,000 curies. a h t .i e

O 6-17 h

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't O On e,Mn _ 6'.t ' M. : This section deecribee those plant operating procedures and procedure.1 enfeguania which have a potential effect on safety. Op*ratinn principles and proceduree 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 deyslopment program, and for operational testing of the nuclear safeguards' systems of the plant. 71 BASIC OpEFM ING PRINCIPLE 3 7 1.1 operation and contrcl of the reactor and most of the process equi;rer.t shall be centralized in the control room, which shall be located in the turbine building. 7 1.2 There shall be at least two operations personnel (one of whom shall be an AEC licensed operator) in the control room for start-up and shutdown of the plant. Then shall be at least one ADO licensed operator in the control room at other times durinc power operation and also during, refueling. No licensed operator shall be required in the control room when reactor is in the cold shutdown condition as defined in Section 1.2 3 The minimum shift complement shall consist of a Shift Supervisor and two operations personnel. 713 operators may perform certain operatinc functions at control panels and valve'rabke outside of the control room but only at the direction of er with prior knowledge of the operator in the control rocm. 7 1.4 Radiation monitoring by fixed or portable instrumentation shall be perfonned before entry into any radiation zones. 715 All personnal leaving radiation sones, and all equipment being removed from such zones, shall be surveyed to an extent @ equate for contrp}.of contamination. 71.6 start-ub' Normal shutdovn, and all other apetitive operations which may involve nuclear safety shall be perfonned in accordance vith specific written procedures. 717 Routine maintenance of protective devices and critical operating equipment shall be done in.acconlance with established written echedules. 7 1.8 In the event of any situation which may compromise the safety of continued operation, it shall be required procedure to shut the plant down and to take other planned emergency action to protect persons and property. i 7,1 s -r--

g 7 3.. C ' Insidente and a:tc having a potential for detrimental effect on j nu cle t..* safety shall be investigated to effect procedures to pl veJ,t recurrence. - Initial review shall be by the Plant Super-intendent. Further review will be by the General Office reviet' group under the Nuclear Plant Operations Supervisor. T.l.10 The Plant Superintend,ent shall have the over-all on-site respon-sibility for the plant.. Technical eupport within the-plant organization shall include personnel with training and experience in the areas of reactor engineering, instrumentation, chemistry and radiation protection. 72 PROCEDURAL SAFIDUARDS The following procedural safeguards have been established for the operating safety of the plant. 7 2.1 Detailed Operating and Emergency Procedures (a) Written procedures for normal and emergency operations which may involve nuclear safety shall be prepared and issued prior to start-up of the plant. 1 The above procedures shall be reviewed and approved by responsible persons on the Plant Operating staff and by appropriate representatives of the company's General Office in Jackson. These procedures shall conform to-the Technical Specifications. Copies of the Site Energency Plan will be kept in the Control Room, Auxiliary Equipment Room, Information Center and the Company's General Office. (b) Instructions for normal operation shall consist of detailed procedures required for the operation of plant equipment. hadiation control procedures shall be compiled to cover aspects of the plant's radiation protection program. 7 2.2 Administrative Procedural Controls The following controls shall be employed to promote safety of the plant: - (sMWaining' t)f the operating staff so that each employee' is acquainted with his specific duties and responsibilities and the action to be taken in the event of off-standard ' conditions. .., 1: ; :+ - (b) Training of new personnel to a level consistent with their specific duties and responsibilities. (c) Periodic management review for strict adherence to the operating and emergency procedures 'and the radiation ~control procedures. 7-2 r ,.-m ,,---,.._,,,,_.-.-m_-,%m,~ m.---.,,,,s ,,.e,.,-,,_m,,- ...m.. . ~... - m.-.

} i 1 7 2.2 (Centd) (d) Periodic management review for strict adherence to the operating limits and rcquirerents for the plant. i (e) Periodic management review for strict control of access to the olant. l (f) Periodic management review for strict adherence to tne urocedure for investigating and reporting unusual or unexpected incidents. 7.2.3 operational Review Procedures Review of day-to-day plant operating and maintenance orocedurec, onerating excerience and plant incidents shall be revieued by engineers of the General Office ataff under the direct 1on of the Nuclear Plant Operations Sunervisor. Recommencations with rcspect to both nuclear safety and equinment nrottetion shall be made to the Plant Sunerintenc'ent and to senior Company management. 7.3 NORMAL OPEDJLTION 7.3.1 Cone:ral Ictailed opercting nrocedures for cach nor-al moce of plant operation shall be arenared nrior to operrtion. The following is an outline of the princinal normal ooeration procedures having a potential effect on the safe operation of the plant. 7.3.2 Cold Start-Uo Af ter Extended Shutdown A cold start-up shall occur each time the reactor is returned to service following an extended shutdown. The crocedure for a normal cold start-up shall be as follows: (a) A start-up check list shall be followed nrior to beginning the actual start-up so that apolicable equipment and systems shall be in condition for start-up. Containment schere integrity provisions shall be in effect. (b) Each control rod shall be exercised and scranmed as a check of the control rod nydraulic systen and the reactor safety system. A couplinc verification check shall be included prior to or during start-up. (c) The start-up eneck list shall te revic. ;c L t aanroved by the Shift Supervisor prior to ctart-up. 7-3

(d) The log count rate meter shall indicate a minimum of three counts per second with a signal-to-noise ratio of 3 to 1. This will be accomplished by withdrawing the start-up chamber to a region of lower flux and observing the reduc-tion in count rate. (e) The reactor shall be brought critical by witndrawal of control rods to their bank positions either singly or in symmetric pairs following a prescribed withdrawal pattern. Whenever k-effective is greater than 0.970, any increase in reactivity (Q)* Creater than 0.001 shall result in a frutional increase in count rate ( o c ) c of not less than 55 of the fractional change in reactivity ( o (). R. (f) The power shall be adjusted once criticality is reacted to maintain a reactor vessel tem,erature rise rate not to exceed 1000 F per hour. (g) The turbine shaft sealing system shall ce placed in service as soon as sufficient steam pressure is available. (Aparox-imately 150 psig.) (h) The condenser shall be evacuated with the mechanical vacuun pump and the air-ejector will be placed in service. (i) Turbine heating shall be started during this operation sequence. After turbine heating is completed, and the reactor reaches rated pressure, the turbine shall be gradually brought up to speed. (j) The mode of turbine control shall be transferred to the initial pressure regulator. (k) The control rocs shall be adjusted to provice the ces2. red power distribution within the core. 7.3.3 Hot Start-Up Whenever the plant has been shut down for a period of time with the reactor vessel and auxiliaries remaining pressuriced, a hot start-up procedure shall be followed to return the plant to service. This procedure will be essentially indeocncent of the cause of shutdown assuming that the cause is recogniced and any nonstandard conditions have been corrected. The reactor instrumentation shall be reset and downscaled and a hot start-up check list shall be completed prior to the witndrawal of control rods. l ( ( = 1-keff keff 7-4

g 7 3.3 (Coned) 1 A coucling integrity eneck shall be nade in accordance with Section 5 3.2(d). The start-up shell trien proceed in accordr nce uith Paragraats (d) thraugh (k) of 7.3.2 of the normal cold start-up aroceduro outlined above. ) 7.3.h rormal Power oneration Lt. ring norr.a1 nouer oneration, thc init,1cl orcusuro rcCuJctor shall naintain the renetor precoure et ' tu rate r? vnluc ! r, - eperating the turbine admission vrives. The tur'.)ine-gener.n Mr locc. chall be establ*.ched by the cuntral roc oasiti>ns. The principal function of the operating personnel curing this period shall be as follows: (a) fhe maintenance of a continunes watch in the control room for prompt attention to any annuncietcd alarns. (b) The ad iustment of the control rod 1Lt te rn to acco-maro Lc changes in rcr.ctivdy anc: to naintun tne rcaired pouer distribution. (c) The evalurtion of abnormal conditionc cnc the initint:.on of corrective action cs required. 7.3.5 Extenced Shutdown An extended siutdown sht11 be accanoochc/ : follous: (a) Reactor oower shall be reduced by nnnipulation of the control rods, anc' the main generator loac' shall bc c'c-creased simultaneously. The turhine-g nrrator sncil be senarcted from the system. (b) All control rods shall be insertrd. (c) The remnal of rrrowr decay br-t. and the ! < c'uu llo. ~ reactor oressure shell be accornliched.37 controlling reactor steam flow. The rate of cooling of the ri.tetor vessel shall not be alloucd to exceed 1003 F ner hour. (d) The reactor shutdown cooling system shall be placed in operation whenever reactor nressure drops belou a nres-sure sufficient to maintain turbine seals. This systen will complete the cooling of the rcactor water to 1250 F. (e) A mininum of one start-up channel and one pouer range channel shall be left in operation. All instrumentation portaining to control of activity release shall be Irrt in operation. 7-5 i W--"' m-swr p 3 -e, t ,r-vsa-- r-

4 7.3.6 Short Duracier Shutdown A shutdown of short duration may be accomolished uhile main-taining system oressure. The turbine-generator shall bc un-loaded and separated from the system. Reactor heat shall be accommodated by systen losses or bypassing steam to the rain condenser. 7.lt RF.FUFLING OERATION The refueling operation shall be cunducted in accor6cnce with the following basic principles: (a) Written procedures snall be available prior to each re-fueling outage. (b) The insertion and renoval of fuel bundles and channels shall be done through the top of the recctor vessel after openinC rcactor vessel head closures as aonropriate. Water shielcing shall be provided by flooding t!.c r. retor vessel anc the refueling extension tank. Fuci cund.ico and channels shall be handled by means of a gra 1c, transfer cask, and crane. Puel movement shall follow the following scrur'20c for each fuel essembly renlaccd: (i) Rcnoval of selected asscmbly from core end trans-fer to surnt fuel storare. (ii) Reshuffling of remaining assemblies in corc as desired. (iii) Inscrtion of new assenbly in the vacant nosition. Shutdown margin checks chall be as ccscribcd in 5 3.5. Asscnbly reolacement shall prococd es icscrrced abovo until the desired number of fuel assenblics have been changed. (c) The tri, ctevices soccified in Section G.3.1 shall be in servicc and connected to the reactsr safcty syst" durirc all refueling operations. A ninimum of two low-level neutron monitors shall, durlag all refueling operations, be in servic: so as to indier.ta any change in neutron flux resulting from chcnce in reac-tivity. 7-6

t 7.., (Contd) In addition, both start-up nuclear instrumentation channels shall be_in service and measuring neutron flux during all refueling operations. The procedure which shall be used for core alterations (d) which increase reactivity shall be as outlined in 5 3 5 (a). Communications between the control room and the loading area shall exist during all core alterations. (e) The liquid poison system stall be available and ready for use. (f) Containment sphere integrity provisions shall be in effect during refueling operations. (g) Unirradiated fuel shall normally be stored in air in a new fuel storage area within the containment sphere. (h) Irradiated fuel and irradiated channels shall be stored in the spent fuel storage pool. ~ (1) The minimum refueling crew during refueling operations shall be four men. There shall be a licensed operator in the control room at all times, and the Shift Supervisor shall be in charge. 75 MAINTEMANCE The following basic principles shall guide the maintenance program at the plant: 7 5.1 Damaged or defective equipment shall be repaired or replaced. 75.2 Maintenance check lists shall 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. A system of equipment history records shall be kept in which 7 53 Vill be recorded the extent of and type of repair, the regular preventive maintenance actions, as wel1~as any nonroutine maintenance which is required. 7 5.h The preventive maintenance program shall include a schedule for exercising of nonnally idle components.

7. S.5 Instrumentation and control systems, especially the neutron power level instrumentation and the. reactor safety system, can 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 operation should it be necessary. 7-7

,lr.t - 756 Radiological protection penetices shall be observed in mainten-ance activlties. 757 It shall be permissible to remove a control rod drive from the core when the reactor is in the cold xenon-free condition. The core shutdown margin of 0 3% 21keff/keff with the strongest rod out of the core, shall be met. The mode switch shall be locked in the shutdown position and all associated equipment properly tagged. A spare control rod drive mechanism shall be used to replace the removed drive immediately upon removal of the defective drive. 76 OPERATIONAL TESTING OF NUCLFAR SAFEGUARD SYSTEMS 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 shall be as tabulated. t 0 7-8 0 = e 9 4 .e-g w,,, a w n

/' Reference Procedure bystus or Function Frequency of Within These Undergoing Test Routine Tests Specifications Contahent sphere 6 moeths or less acceso airlocks leaksco rate Post-incident scray system At each major refueling Section 3 5 2 automatic control operation shutdown

Control rod performance At each major refueling Section 5 3 2 shutdown and a,t least once occh quarter during operation Liquid poisen system Two months or less during Section 5 3 3 component operability power operation. One squib test-fired crch 12 months Reaster secot.y system At each major refueling Section 6.1 5 scram circuits requiring shutdown
plant shutdown to check Reactor safety system scram One month or less Section 6.1 5 circuits not requiring plant shutdown to check Containment sphere isola-At each major refueling Section 6.1 5 tion trip circuits shutdown
Reactor emergency At each major refueling Section 6.15 cooling-systems shutdown
r-trip circuits

.l Control rod withdrawal At each major refueling Beetion 6.2.2 permissive interlocks shutdown

function Refueling operation At each major refueling Section 6 3 3 controls function shutd;wn Calibration and functional One month or less Section 6.4.3 test of air ejector off-gas and stack-gas monitors Calibration of emergency One month or less Section 6.4 3 condenser vent monitors Calibration of process One month or less -

Section 6.4 3 liquid monitors " Effluent to Canal" monitor Three months or less - Section 6.4.3 " Remaining monitors" Calibration of area One wanth or less Section 6.4 3

1-monitoring system
But not less frequently than once a yeare 7-9

c.0 REFEA %F AND DEVELOFMENT FROGRAM Af ter completion'of Phase I research and development testing at the Big Rock Point Nuclear Plant, Phase 11 of the Research and Developman: Program will be initiated. All normal operating limitations will be observed as specified in section 8.3. The program will include irradiation of developmental fuel bundles, core performance testing, stability and transient perforcance testing, power distribution and physics testing, and operation of a process computer system. The latter will measure plant variables, process the data, and record the resulting information, but will carry no control or safety function in the Big Reck Point Nuclear Plant. The range of operating conditions will be covered in the Phase II tests in a stepwise sequence in order to give maximum knowledge priot-to proceeding to t3e next condition of operation. 8.1 FUEL IRRADIATION FROGRAM 8.1.1. Development Fuel Design Features Fifteen developmental fuel bundles have been designed for use in the Addi-Big Rock Point reactor core at the commencement of Phase II. tional developmental fuel bundles may be provided at a later datc in the Phase 11 effort. Although the details of these additional fuel bundles may differ somewhat from those described here for the original 15 fuel bundles, their principal design features will be similar. All fuel bundles are of a similar mechanical design typified by the bundle drawing in Figure 8.1. The design of the Phase 1 and Phase II developmental fuel is such at to give nuclear characteristics e~quivalent to the original core fuel. The priucipal difference between Phase I and Phase 11 fuel is in the higher reactivity of Phase II bundles to permit operation t - a burnup of 15,000 Mwd /T. b.ur different cladding materials are employed in the fuel designu. Enrichments for these bundles have be:m chosen to yield essentially the same reactivity for each of the type r. These reactivities are: 8-1

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i, j FIGURE 8.2 60 KW/'L CORE /' 2 o+4 i / o /' to -*- + + .A. 5 58 - - D D 7 / a u 57 - - D I D D A n / 56 - - D / 1 \\, I 55 - - \\ i A F 54 - - D i 53 - - D D D \\, \\ E .s 52 - - D D \\, - \\ D N 51 - - '\\ l l C l l i I I i i i i 1 1 2 3 4 5 6 \\, 7 8\\ B 9 10 \\A CONTROL IN-C?Pf MONITORS d PHASE ll R 6 0 ASSEMBLIES D PHASE il(NSTRUMENTED ASSEMBLIES I i 8-3 l ) l i )

s. 3.1,1 (D n t( ) k eo 20 C 1.32 2880 C 1.32 2880 C + 20% Void 1.30 Void Coefficient -27 x 10-4 Ak/% Void Moderator Coefficient + 0.1 x 10-4/C (Based on Critical Leakage) The design of the research and development bundles shall be such that the f0el rods may be replaced with rods containing poison material to effect either power shaping or reactivity control. Such rods may be tested for power flattening as part of the Phase II Research and Development Program. The effect of these rods shall be to reduce the "k" values and to make the temperature coef ficient more negative. The principal design f eatures of the Phase I and Phase II devclopmental fuel are essentially as follows: Geometry - Fuci Rod Array 11 x 11 No. of Standard Rods 109 No. of Special Corner Rods 12 Standard Fuci Rod Diameter, Inches 0.425 Special Fuel Rod Diameter, Inches 0.320 Spacers per Bundle 7 Average Heat Flux G 45 Kw/L, Stu/Hr-Ft 116,500 Hydraulic Diameter, Inches 0.58 Core Volume per Bundle, Liters 63 Phase I Phoe. TT Clad Material 104 SS 304 SS Zr-2 Inconel Incol ov Incolov No. of Bundles 4 (Powder) 4 3 3 3 '6 (Powder) Clad Thickness, Inches 0.010 0.010 0.030 0.019 0.019 0.011 Nominal Clad-UO2 Cap, Mils 0 5 7 6 6 0 Enrichment, Percent 2.7 2.7 2.8 4.5 4.1 3.3 UO2 Density, Percent 91 95 95 95 95 90 Water to Fuel Ratio 2.4 2.4 2.96 2.63 2.63 2.4 2 76.5 76.5 76.0 74.6 74.6 76.5 Heat Transfer Area, Ft Active Fuel Length, Inches 70 70 69.3 68 68 '. 7 0 UO2 Veight/ Bundles, Lb 370 385 302 336 336 374 Max. Fuci Cladding Stress, 72,200 32,400 14,600 24,200 26,400 70,000 Psi 8-4 y p -w g -e cy-e ~-+4 r ,, + w

.- A k 8.1.2 Instrumented Assembly Design One or more instrumented bundles will be utilized for selected tests during Phase 11. Their design will be identical to that of the orig - inal core fuel with modifications to accommodate the instrumentation. An instrument probe will be directed through a reactor vessel head penetration and will be inserted into the center of the fuel bundle where four fuel rods have been removed. Instrumentation associated with the probe includes a turbine type flowmeter to measure channel coolant flow rate, thermocouples.to measurefinlet-and cutlet coolant temperatures, and ion chambers to measure axial neutron flux levels. The lower orifice and upper handle will be elongated to incorporate the inlet and exit flowmeters. 8.2 PERFORMANCE TESTING The tests wi3) consist primarily of bringing the reactor to a set of specified operating conditions and measuring the nuclear, thernal and hydrodynamic characteristics in each of those modes of operation. Physics tests of core power distribution will also be performed utilizing gamma scan and wire irradiation methods. 8.2.1 Core Performance and Transient T-sts The Phase 11 core performancc, ;ransient and stability testing will be conducted over a range of variables as indicated by the following: Variable Range Core Size, No. of Bundles 40 to 86 Reactor Power Level, Mwt Up to 240 6 Recirculation Flow Rate 6 x 10 lb/hr to Full 2-Pump Flow Reactor Pressure, Psia 800 to 1500 Core Inlet Subcooling Rated to Maximum 8-(

8.2.1 (Contd) Variation in control rod programming and core orificing may also be applied. Recirculation flow rate shall be set by positioning each of the butterfly flow control valves at appropriate points between the " full open" and the "50% of rated flow" positions. The combined recirculation flow rate will, in no case, be set such as to violate the specific reactor operating lLnits imposed in dection 8.3. Detailed thermal hydraulic analyses have been performed for the range of Phase 11 tests that are outlined. Stability analyses of half-rates rated, and overpower operation of the rated 240 Mwt core have been performed. Any tests in the parametric range specified, prior to their performance, shall be shown by these and similar analyses to be within the established operating limits as given in Section 8.3. The calculated operating condition for a typical test is presented in Section 8.2.3. Treasient hydrodynamic performance and stability will be studied by introducing controlled disturbances such as reactivity os-cillation, turbine load changes, and pressure set point chan8es at selected reactor operating conditions. Stepwise increases in reactor power level, and decreasce in core flow rate will be observed in each test series to evaluate the progress and safety. 8.2.2 Sequence of Testing In addition to the irradiation of developmental fuel, a sequence of operational tests to provide deteiled knowledge of the steady state and transient thermal-bydraulic characteristics will be conducted. 8-6

s 8.2.2 (Contd) 1. Full Pover Coeratien The plant power shall be increased to approxirmtely 240 Mwt in approximately 20 hkt steps from the 157 Mwt rated condition. Satisfactory performance with respect to safety shall b'e established at each step before proceeding to the next step. The test program shall ultirately demonstrate 75 Mwe operation at a thermal power and core configuration capable of meeting the . established reactor operating limits. Tests and observations 'to be made at each of the incremental power increases shall be typified by the following: a. Power Calibration of Nuclear Instruments - Picon= meter readings shall be compared with thermal power level calcu-lations. b. Calibration of the In-Core Ion Chamber System - Flux vire data shall be used to calibrate the in-core chamber readings. c. Radiation Surveys - Portable and fixed instruments shall be used to measure neutron and gamma radiation in various plant areas. 4 d. Radiochemical Analysis -. Grab sampics from appropriate stations shall be analyzed. e. Bypass Valve Test - The characteristics of the bypass valve system shall be rechecked. It is the intent of such testing to assure that the bypass valve. control system wi11' control the reactor vessel pressure ddequately under manual and auto- ~ matic control conditions,c,,, r ' is ". - n.. f. Initial Pressure Regulator Tests - The operation of th'e initial pressure regulator system shall be tested at various set points and its effect on reactor operation recorded. 2. Rich Power Density De-onctrat' ion k'ith a reduced size core, typ[fie) y, Figure 8.2, a stepwisa increase inpowershallbemadefromappro~)imately115Mwttoapproxicately x 157 Met where the target power density of 60 Kv/L average is to be achieved. The approach to 157 Mwt from 115 Mwt shall be made in at least three approximate'ly equal steps. Core performance evaluation shall be made at each step increase to assure the satisfaction of all operating limits such as burnout ratic, heat flux, kilowatts per foot, and, stability. 87 + 6

.-... ~. .. - - - _. = ~ .C N .i j 8.2.2 (Contd)

3.. Performance Testinn As.a part of.the Phase II; developmental program, tests similar to

-[ Phase I tests shall be conducted-at escalated conditions compatible i with the new core operating IcVels. Wire irradfa; ion'and~ gamma scans shall be performed at appropriate points pre. ceding or during i a sequence of Phase II tests to provide'the necessary core power distribution data and to assure that operational'11mits are satis-' fled. In addition, shutdown margin checks, and temperature co-l efficient measurements shall be made whenever necessary or whenever significant changes have occurred in.the core.. New developmental fuel bundles shall be subjected.to fuel verification testing. l .j The Phase II core performance, transient and stability testing shall I be conducted over a range of-variables as indicated in Section 8.2.1. In addition, recirculation pump trip tests may be performed under selected conditions, shown by prior analysis.to be within operating limits, in order to evaluate coast-down characteristics, power decay characteristics, etc. The combinations of the variables'shall alvays be chosen and analyzed to satisfy the reactor operating limits specified in Section 8.3.. Maximum subcooling shall be that corresponding to full bypass of-steam around.the. turbine. Butterfly flow control valves in the two i recirculation lines shall be utilized to control flow to 507. of the full flow from a given pump. Other variations which may be imposed during the' testing sequence' would involve alternate control.ro,d programs and variations in core f inlet orificing, incleding removal of orifices. Prior to execution of the individual tests chosen within the parametric range described in Section 8.2.1, all limiting conditions shall be established by thermal-hydraulic and stability evaluations. i Transient hydrodynamic performance and stability shall be studied by $ntroducing controlled disturbances such as reactivity oscillation, i pressure set point changes, and flow variations.at selected reector operating conditions. Stepwise increases in reactor power Icvel ~j and decreases in core flow rate shall be observed in each test to evaluate progress and safety. l 4, Developmental Test Secuence-During and af ter the steps followed to achieve full power, the J Phase II developmental test sequence shall be approximately as follows: 8-8 i J i l l ._._.,_._._2 u.

's 1 8.2.2 (Contd) a. Core size, 84 bundles. b. Increase power from 157 Mwt to 240 Mwt stepwise at full flow and rated subcooling. c. At selected increasing power level steps, decrease flow from full two pump flow in a stepwise manner at rated subcooling. d. Repeat selected tests from "b" and "c" for a! condition of maximum subcooling. e. Decrease core size to approximately 41 bundles, f. Increase power stepwise from approximately 60% of the ultimate power until that power level is attained. g. Repeat tests described in (c) above with the small core. h. Increase subcaoling and repeat selected tests from (f) and (g) sbove. 8.2.3 Analyses of Typical Tests: Analyses of conditions anticipated at the full power and highest power density operations result in the expected core thermal and hydraulic characteristics given in Table 8.1. All reactor operating limits are met under these conditions. Other Phase 11 development tests, though differing in specific variables which may result in limiting conditions, will comply with the operating limitations specified in Section 8.3. 8-9 ~ -

t TABLE 8.1 ' ~ i. PilASE II CALCULATED TilERMAL-IlYDRAULIC OIARACTERISTICS ) .c b Peak Thermal Power PeakPowerDengty 4 4 j Number of Fuel Bundles in Core 84 41 Reactor Pressure, Psia 1250 1250 i Reactor Thermal Power, Mwt(a) 24 0 157 Average Power Density, Kw/L 45 60 I Fuel Type Development Initial - Development Initial Peaking Factors Local 1.407 1.407 1.418 1.418. Gross 2.01 1.823 1.943 1.762 { Overpower 1.22 1.22. 1.22 1.22 Total (Product) 3.45 3.13 3.36 3.05 llcat Flux, Btu /llr-Ft2 i Core Average at Rated Power 116,500I -116,500 -157,110 157,110~ 4 EE Maximum at Rated Power 329,000 299,000 434,000 393,000 Maximum at 122% Overpower 402,000 364,600 _528,000 478,900 s i Kw per Foot of Rod Maximum at Rated Power 10.7 8.9 14.2 11.7 lbximum at 122*s Overpower 13.1 10.9 17.2 14.2 3 Minimum Burnout Ratio j Steady State Overpouer 1.83 2.00 1.52 1.64 i Loss of Pump - 1.50 1.50 1.80 _'1.98 Total Recirculation Flow, Lb/llr x 106 12.5 12.5 9.86 9.86 i. ~ Core Inlet Subcooling, Btu /Lb 19.5 19.5 19.5 19.5-(a) At natural circulation with the 84-but.dle core, a rated power Icyc1 of.190 Mwt is calculated to have a burnout ratio above 1.5 ~ At natural circulation with the 41-bundic core, a rated power IcVel of 120 Mut is calculated to i have a burnout ratio above 1.5. b 1 i - ~,.,.. _ -. _ ~ s...-

,= I 'I j A 8.i 3[39T00 OPERATING LIMITS + Reactor operation'during the R&D program shall be determined by the most restrictive of the following limits. The test'pregram will include development fuel irradiation, reactor core performance tests, stability and transient tests, and power distribution and' t physics tests. (c) The burnout ratio at overpower shall be shown by calculation to be at least 1.5 for all reactor operation test corditior. The burnout ratio is defined as the minimum ratio cf the design burnout limit heat flux

to the reactor heat flux as calculated using appropriate allowances for flux distributions, flok dis-tributions, effects of power trcnsients and uncertainties.

(L) The transient burnout natio in the event of loss of recircu-lation pumps from rated power shall be shown by calculation to be at least 1.5 for all reactor operation test conditions. 2 (c) The heat flux shall not exceed 530,000 Btu /hr-ft and the fuel rod power generation per unit length shall not exceed 17.2~kv/ft. (evaluated at overpower conditions). These quantities shc11 te determined by use of the same allowances described in tbc burr.- out ratio calculation. (d) Test conditions shall be such that the measured steady stste zero to peak flux amplitudes shall not exceed 20% of the average operating flux level for a given test..in addition, th. zero to peak flux amplitude shall not exceed 20% of the operstip level during any stability test which may be conducted. thra rod oscillation is a part of the test, the initial amplitudc c f disturbance will be limited to one control rod notch or equiv-alent. When results of this limited controlled disturbance pi c evidence that the equivalent of multiple-rod or cultiple-notch oscillation will not exceed the stated flux criteria, these larger disturbances may be applied as part of the stability 2 testing. The operating limits specified in Section 5.3.1 (b) shall apply te all test operations during Phase 11 of the Research and Develop.Tir: Program whether utilizing original core fuel, instrumented fuel, Phase 1 developmental fuel, or ' Phase 11 developmental fuel.

"Eurnout Limit Curves for Boiling Water Reactors," by E. Janssen Land b. Levy, L?hD-3892.

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,1 A 6.4 OPERATING PROCEDURE Written procedures shall be prepared and issued prior to the initiation of each test involving the operation of the. plant. These procedures shall provide the information for performing each test, detailed to the extent appropriate, according to the outline as follows: (a) The objective of each particular test. (b) The step-by-sceo operational method for the test including specific terminal aspects of the test. (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 event that an ab - normal response occurs during the test will be defined. The responsible individual having authority to direct j the continuance or discontinuance of a particular test will be specified. (d) In addition to the details covered above in this outlire, there will be instructive information, prr."cided with at., g special equipment that may be employed in the courre of testing. 8.5 SPECIAL REVIEW PROCEDURES An R&D Review Committee shall be established to provide advice and consultation to the Plsnt Superintendent to further assure nuclear safety. This committee willreview as frequently as the situtation requires, and primarily at the request of the Plant Superintendent, all matters involving nuclear safety associated with the Phase II of the Research and Development Program at the Big Rock Point Nuclear Plant. 8-12

APPENDIX B ) CONSUMERS POWER COMPANY Estimated Schedules of Transfers of Special Nuclear Material From the Commis-sion to Consumers Power Company and to the Commission from Consumers Power Company: t (1) (2) (3) (4) (5) Returns by Net Yearly - Cumulative Consumers to AEC Distribution Distribution Date of Transfers Kas U-235 Including Including Transfer From AEC to Cumulative Cumulative (Fiscal Consumers Cold Spent Losses-Losses Yec ) Kgs U-235 Scrap Fuel Kgs U-235 Kas U-235 1962 423.0 34.0 0 389.0 399.0 1963 60.0 7.0 0 53.0 442.0 1964 54.0 10.0 0 44.0 486.0 1965 510.9 15.0 151.5 344.4 830.4 1966 291.8 29.9 151.5 110.4 940.8 1967 437.7 30.0 202.0 205.7 1,146.5 1968 291.8 20.0 289.5 (17.7) 1,128.8 1969 291.8 30.0 175.0 86.8 1,215.6 1970 350.3 23.9 210.1 116.3 1,331.9 1971 350.3 23.9 210.1 116.3 1,448.2 1972 350.3 23.9 210.1 116.3 1,564.5 1973 350.3 23.9 210.1 116.3 1,660.8 1974 350.3 23.9 210.1-116.3 1,797.1 1975 350.3 23.9 210.1-116.3 1,913.4 1976 350.3 23.9 210.1 116.3 2,,029.7 19)7 350.3 23.9 210.1 116.3 2,146.0 1978 350.3 23.9 210.1 116.3 2 262.3 3 1979 350.3 23 d 210.1 116.3 2,378.6 1980 350.3 23.9 210.1 116.3 2,494.9 1951 350.3 23.9 210.1 116.3 2,611.2 1982 350.3 23.9 210.1 116.3 2,727.5 1933 350.3 23.9 210.1 116.3 2,843.8 1984 350.3 23.9 210.1 116.3 2,960.1 1985 350.3 23.9 210.1 116.3 3,076.4 19f6 350.3 23.9 210.1 116.3 3,192.7 1967 350.3 23.9 210.1 116.3 3,309.0 1988 350.3 23.9 210.1 116.3 3,425.3 1589 350.3 23.9 210.1 116.3 3,541.6 1990 350.3 23.9 210.1 116.3 3,657.9 1991 350.3 23.9 210.1 116.3 3,774.2 1992 350.3 23.9 210,1 116.3 3,890.5 1993 350.3 23.9 210.1 116.3 4,006.8 1994 350.3 23.9 210.1 116.3 4,123.1 1995 350.3 23.9 210.1 116.3 4,239.4 1996 350.3 23.9 210.1 116.3 4,355.7 1997 350.3 23.9 210.1 116.3 4,472.0 1998 350.3 23.9 210.1 116.3 4,588.3 1999 350.3 23.9 210.1 116.3 4,704.6 2000 262.7 ___17.9 157.6 87.2 4.791.8 13,132.7 1,110.8 7,429.6 4,586.1 ---}}

ML20126E613

Text

1.0 INTRODUCTION

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