ML19345E308
| ML19345E308 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/02/1959 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19345E307 | List: |
| References | |
| NUDOCS 8101060786 | |
| Download: ML19345E308 (17) | |
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1 Attachment No.1 to Amendment No.15 to License Applicatica AEC Docket 50-29 pd October 2, 1959 '
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!sy YANKEE ATOMIC ELECTRIC COMPANI NY[
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Technical Specifications fc
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A. -PURPOSE AND SCOPE 43/
In this document the design features and operating limits, procedures and principles are set forth which it is proposed shs11 constitute the technical specifications to be incorporated in the construction permit CPPR-5 and in any superseding license. Design features and operating limits, procedures and prin-ciples covered in the technical specifications may not be changed or modified except by amendment to the construction permit or superseding license. Changes and modifications in the design and operation of the plant which do not conflict s
with the technical specifications may be put into effect by following the proce-dure outlined in Section B.
y References to specific sections of the final Hazards Sumary Report which appear in parentheses opposite numbered captions in Section C of the tech-nical specifications are not to be taken as incorporating the section or sections into the specifications by reference but are included for convenience. References to sections of the final Hazards Summary Report which appear in the text of Sec-tions F, G and H are considered to incorporate these specific sections of the final Hazards Summary Report into the technical specifications.
B.
CHANGE PROCEDURES Proposed changes in the design and operation of the plant which do not violate these technical specifications will first be reviewed by the Plant Safeguards Committee which consists of the Plant Superintende'nt, the Chief Engi-neer, the Technical Manager and the Reactor Engineer. If the committee deter-mines that the proposed change involves hazards not greater than and not different from those analyzed in the final Hazards Summary Report and does not involve a material alteration of the facility, the proposed change may be made without further approval from the Commission. If the committee determines that the hazards involved are or may be greater than or different from those analyzed in the final Hazards Summary Report, or that the proposed change may involve a material alteration of the facility, and the reactor designers and company con-sultants concur in this view, Yanlee will provide the Commission with a descrip-
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tion and hazards evaluation report of the proposed change. If, within fifteen
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days after the date of acknowledgment by the Division of Licensing and Regulation 7}6/obd'774
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of receipt of such report, the Commission does not issue any notice to the pd contrary, Yankee may make such change without further approval. Within the fifteen day period the Commission :uy establish through notice to Yankee the procedure it wishes to follow in considering the proposed change.
C.
DESIGN EATURES 1.
Reactor Vessel (Section 230, Final Hazards Summary Report)
The reactor vessel is a vertical cylindrical pressure vessel fabri-cated primarily of SA-302, Grade B carbon steel and clad internally with Type 30h stainless steel. Top and bottom heads are hemispherical. The upper head is removable and is provided with a bolted and gasketed closure. There are four inlet and four outlet nozzles which are at the same elevation just above the reactor core and which alternate and are equally spaced around the circum-ference of the vessel. The vessel is designed, built and tested in accordance with the ASE Boiler and Pressure Vessel Code,Section VIII, Unfired Pressure Vessels,1956 Edition and latest addenda, and is certified under Code Case No.
123h.
Overall Height 31 ft. - 6 in.
Inside Diameter 9 ft. - 1 in.
Design Pressure 2500 psia Design Temperature 650 F.
2.
Main Coolant System (Section 201, Final Hazards Summary Report)
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The main coolant system consists of four closed loops connected in parallel to the reactor vessel. Each loop contains a canned-motor type circula-ting pump, a steam generating heat. exchanger, two gate type stop valves, a 5" crossover connection, and a check valve, as well as 20 and 2h inch 0.D. piping.
The major material of constmetion used throughout this system is Type 30h stainless steel.
Pumos Main coolant pumps are of the centrifugal type, driven by hermeti-cally sealed motors. Each pump has a nominal rating of 23,700 gpm against a head of 228.5 ft of water at h96 F and 2,000 psia. The pressure containing portions are designed to a pressure and temperature of 2,500 psia and 650 F and are in accordance with the ASME Code,Section VIII, where applicable.
Steam Generators The steam generators consist of a vertical shell and U-tube evapo-rator together with 3-stage moisture separating equipment. The primary side inlet and outlet nozzles are located at the bottom. The maiu coolant is the heating fluid and flows through the tubes.
The steam geac,rators are constmeted in accordance with the re-quirements of the ASE Code,Section VIII, Unfired Pressure Vessels, 1956 Q
Edition, Paragraph U-1-e, under which they are classed as an unfired steam I
y boiler. The primary side of the steam generators is designed to a pressure kJ of 2,500 psia and the shell side to 1,0$0 psia. The design temperatures for the primary side and the shell side are 650 F and 600 F, respecti.aly.
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Valves The two stop valves and the check valve in each main coolant loop are designed to pressure and temperature requirements (2,300 psia and 650 F) according to the 1957 edition of ASA B16.5, Standard for Steel Pipe Flanges and Flanged Fittings, and other applicable codes.
Pipe and Fittings The stainless steel main coolant piping is of the hollow forged and bored type. Both 20 and 2h inch 0.D. pipe is used in each loop. The de-sign and fabrication is in accordance with the requirements of the ASA B31.1, 19$$ edition, Code for Pressure Piping. The sizing of the pipe as to wall thickness is based on a design pressure and temperature of 2,300 psia and $50 F with provisions for variations in pressure and temperature as permitted by the Code for Pressure Piping.
L Pressure Control and Relief System (Section 202, Final Hazards Summary Report)
The pressure control and relief system consists of a pressurizer vessel normally containing water and steam, replaceable direct imersion heaters, safety and relief valves, a spray system, interconnecting piping, valves and in-strumentation.
Pressurizer The total volume of the pressurizer is 295 cu. ft. Under normal operating ccaditions the pressurizer contains about 90 cu. ft. of water. The pressurizer vessel is classified as an electrically heated steam generator and is designed, constructed and tested in accordance with the requirements of the AS'E Boiler and pressure Vessel Code,Section VIII (1956 Edition and latest addenda) and AS!E Code Case Nos.122h and 123h. The maximum allowable working pressure and temperature are 2,500 psia and 668 F.
Heaters Direct immersion electrical heaters, having a total output of 300 N, are installed in four flanged penetrations in the lower section of the pressur-izer vessel. During normal operation, one heater group (37.5 W) is placed on cycling control to maintain primary system prescure, and the remaining seven groups are placed on low pressure back-up control.
Safety and Relief Valves Two safety valves and a solenoid operated relief valve are provided to accommodate pressure surges which are beyond the over-pressure limiting capacity of the pressurizer and the spray system.
The safety valves acting in conjunction with the reactor negative temperature coefficients are capable l
of preventing system pressure from exceeding the design pressure of 2,500 psia by more than 60 The design is based on the ASME Boiler Code,Section I.
The solenoid operated relief valve operates at a lower pressure to minimize the G
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operating frequency of the code safety valves.
Each main coolant loop is Iv
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provided with a water relief valve to relieve over-pressure due to thermal ex-O pansion. Discharge of safety and relief valves is to the low pressure surge tank which acts as a blowdown arri quenching tank. If the blowdown capacity of the low pressure surge tank is exceeded, safety valves in the surge tank relieve to the vapor container. As a result all primary system relief discharges are confined within the vapor container.
Spray System an adjustable spray provides continuous circulation of water through i
the presemizer thereby maintaining equilibrium conditions between the water l
and stsam phases. In addition, a solanoid operated spray valve operates to limit positive pressure surges in the primary system.
h.
Initial Core (Sections 100,101,102 and 103, Final Hazards Summary Report)
The reactor core is composed of 76 fuel assemblies, arranged verti-cally between two support plates. The coolant flow is upward through the core.
The fhel is slightly enriched uranium dioxide contained in stainless steel tubes.
The enrichment is uniform throughout the core.
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Nominal Initial Core Design Data Average Core Diameter (cold) 75.35 in.
Average Core Length (cold) 91.86 1.
9 Number of Fuel Assemblies 76 UO in the core 52,015 lb.
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Uranium in the core 20,8h0 kg.
Uranium enrichment 3.h w/o U-235 in the core 708.6 kg.
The 76 fuel assemblies are made up of a number of full length, vertical stainless steel tubes containing solid, cylindrical pellets of com-pacted and sintered UO. Each fbel assembly is basically an 18 x 18 square 2
with rods omitted from this pattern as required to provide cruciform slots _ for -
the control rods. One central rod is also omitted from each assembly to provide for insertion of core instrumentation. As a result there are two types of fuel assemblies, containing 30h and 305 fuel rods respectively. Each of these fuel assemblies is made up of nine sub-assemblies with the individual fuel tubes spaced by tubular fermles brazed into place. The nine sub-assemblies which make up each fuel assembly are secured at each end to. identical end fittings and are held together at several points along their length by straps passing through the fuel rod array and welded at each face to rubbing strips.
Each fuel rod contains a number of perforated, stainless steel discs which serve to divide the column of UO into six sections. - No enclosing wrappers are utilized 4
2 on the individual fuel assemblies with the result that the overall-cort presents an open lattice for the coolant flow.
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Nominal Initial Fbel Design Data, (Cold Dimensions)
Tube Material Type 3h8 stainless steel Pellet Diameter 0.29h in.
Pellet Length 0.6 in.
Minimum Pellet Density 91.5% of theoretical Fuel Tube OD 0.3h0 in.
Fuel Tube ID u.298 in.
Pellet-Clad Diametrical Clearance 0.00h in.
Fuel Tube center to center pitch normal 0.h22 in.
in line with control rod vanes 0.h5h in.
5.
Control Rods and Drives (Section 101 av 103, Final Hazards Summary Report)
There are 2h control rods made of an alloy of 80% silver,15% indium, and 5% cadmium. These are electro-plated with nickel and heat-treated to provide a diffusion S nd.
The control rods are of 7.865 inch span, cruciform cross sec-tion and travel vertically in spaces left open in the fuel lattice at two oppo-site corners of each fuel assembly. 2h Zircaloy-2 control rod followers are provided, attached to the bottom of each control rod.
Fixed shim rods are placed in the eight extra control rod channels in the outer region of the core. These rods are similar in size and shape to
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the control rods, but will be either boron stainless steel or Zircaloy-2, depend-ing on reactivity requirements determined at the time of start-up testing.
Each of the 2h control rods is driven by an individual drive mechanism of the positive latch, magnetic jack type. These drives are mounted on the head of the reactor vessel and move the control rods upward or downward through the core. The pressure housing within which are located the drive shaft and all other moving parts is designed, fabricated and tested in accordance with the ASME Boiler and Pressure Vessel Code,Section VIII, Unfired Pressure Vessels, 1956 Edition and latest addenda. The operating coils are mounted outside the pressure housings. Power failure to the drive mechanisms de-energizes the mag-netic coils and allows all control rods to fall into the core by gravity. Auto-matic or manual scram is accomplished by similarly de-energizing the coils.
The control rods are divided into six groups of rods which normally move as units, although rods can be moved individually as required. The inner five groups, each mde up of either 2 or h control rods, move at 12 inches per minute or less, and the other group of 8 control rods moves at 6 inches per minute or less.
Active length of contr-a rods 92-7/8 in.
Maximum time from rer eipt of scram signal to 90% control rod insertion 2 sec.
6.
Chemical Shutdown System (Section 20h, Final Hazards Summary Report)
(Ov) absorbing boric acid solution to the main coolant system to bring the reactor The Chemical Shutdown System provides a means of adding a neutron from hot shutdown to the cold sub-critical condition.
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The system consists of a mixing and storage tank, a transfer pump and the necessay valves, piping, instrumentation and control. A 12 wt. %
solution of boric acid is prepared in the mixing and storage tank which is then added to the main coolant by charging pump operation while the reactor plant is at operating conditions. Main coolant pump circulation assures that the injected solution is well distributed throughout the main coolant system. The entire injection operation can be carried out by remote control of the motor operated inlet valve between the mixing and storage tank and the charging pump suction header. The system transfer puma transfers 12 wt. % boric acid solu-tion from the mixing and storage tank to the safety injection shield tank cavity water storage tank where it is diluted to a 1 wt. % boric acid solution for use in the Safety Injection System.
The Chemical Shutdown System also provides a means of removing boric acid from the main coolant system. During reactor start-up, initial boric acid concentration is reduced by dilution and recirculation until about 5% of the original concentration remains in the main coolant. The remaining amount is removed by bleeding the slightly borated main coolant to the low p;e.tsure surge tank, pumping with the purification pumps through chemical ion exchanprs to the charging pumps, which then return the de-borated water to the main coolant system.
7.
Shutdown Cooling System (Section 210, Final Hazards Sumnary Report)
The Shutdown Cooling System is provided to remove the heat generated by radioactive decay of fission products in the reactor core following shutdown.
This system is placed in operation after the main coolant has been reducad to i
approxirately 330 F and a pressure of less than 300 psi gage. It then reduces
- y the main coolant temperature to lho F or less and operates continuously to main-tain this temperature as long as required.
The system consists of a heat exchanger, circulating pump, piping, and necessay valves arranged in a lcw pressure auxiliary loop in parallel with one of the main coolant loops. The shutdown cooling pump takes suction from the hot leg of the main coolant piping on the reactor side of the stop valve, circu-lates the main coolant through the tube side of the shutdown cooler and back into the cold leg of the main coolant piping on the reactor side of the loop stop valves.
Complete back-up of the system is provided by the low pressure surge tank pump and heat exchanger which are identical units connect.ed in parallel. The en-trance and exit piping ce'nected to the main coolant system instae the vapor container up to and inclu 2ng the four motor operated isolation valves meet the 2,300 psi gage, $50 F main coo] ant piping design.
8.
Safety Injection System (Section 212, Firal Hazards Sur: mary Report)
The functions of the Safety Injection System are:
(1) To automatically supply borated water to the reactor vessel for cooling the core in the unlikely event of a maJer loss of coolant accident; and, (2) To supply water for flooding the shield tank cavity during l
refueling operations.
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The system consists of a 125,000 gal safety injection-shield tank cavity water storage tank, two 1800 gom pumps, miscellaneous piping, valves. id O
The system is started automatically by a coincidence circuit eviisist-controls.
ing of a vapor container high pressure signal and a main coolant 7 ow pressure signal. Both signals must occur before safety injection will take place. Manual initiation of the system may also be accomplished by operation of the control switch on the rain control board. Either manual or automtic initiation causes 1
safety injection pumps to start and valves to open or close resulting in correct valve line-up for safety injection. Operation of the two pumps at 3600 gpm fills the reactor vessel to the top of the core in approximately 2.1 minutes.
The safety injection pumps are supplied with power from two independ-ent ?h00 V buses, each bus supplied from its own station service transformer connected to one of two independent 115 KV transmission lines. The piping and valves inside tne vapor container me,c the main coolant piping design of 2,300 psi gage at $50 F.
9.
Component Cooling System (Section 206, Final Hazards Summry Report)
The Component Cooling System is provided to remove waste heat from various nuclear plant components and systems. The use of this closed indirect system permits removal of component heat without contamination of river water even if there is leakage of radioactive fluids from the cooled components.
The system consists of two coolers, two 2000 gpm circulating pumps, a h,000 gal, surge tank, system instrumentation and piping. Equipment is con-nected to two main piping headers located outside the vapor container. Independ-ent lines, provided with isolation valves outside the vapor container, connect the header to the components inside the vapor container. Each circulating pump and cooler can accommodate full heat removal loads.
Coolant flow is provided to the four main coolant pumps, the neutn 3 shield tank coolers and the sample cooler located inside the tapor container.
Each line to this equipment is provided with check and stop valves. Coolant is l
also supplied for the fuel pit cooler, the low pressure surge tank cooler, the l
shutdown cooler and for the waste disposal building. The component cooling system is designed to operate at a mximum design pressure of 110 psi gage and a maximum temperature of 150 F.
l 10.
Secondary Steam System (Section 219 and 23h, Final Hazards Summary Report)
The Secondary Steam System conducts steam in a lh in., Schedule 80, carbon steel pipe from each of the four steam generators located within the vapor container, through angle type non-return valves, into a common 2h' in.,
i Schedule 60, carbon steel pipe and then by two 18 in., Schedule 60, carbon steel pipes to the throttle valves of the turbine. The design pressure of this system is 1,035 psi gage at $50 F.
Output from the four steam generators is about 1,800,000 lbs per hour of steam with h50 psi gage steam pressure at the throttle valves of the turbine for rated load, lh5,000 IGT.
Two 2-} by 6 in, safety valves are pmvided on each of the four lh in. main steam lines just ahead of the non-return valves. One safety valve i -
in each line is set to open at a maximum of 935 psi gage and the other at a i
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maximum of 1,035 psi gage, discharge being made to the secondary stack. These O
valves prevent ver-pressure in the ste m generator shells when (1) a turbine throttle trip occurs without a reactor scram or control rod insertion, and, (2) loss of condenser vacuum or malfunctioning of the turbine steam bypass occurs during het standby or shutdown. The calculated safety valve discharge total flow is 817,000 lbs per hour which is mximum flow that can occur under the above con-ditions.
A 6" turbine steam bypass line is provided, allowing for steam bypass around the turbine to the condenser during startup, hot standby or shutdown and during transient conditions. This line branches off the main 2h" steam line and is provided with an automtically operated valve set to open at not less than 760 psi gage main steam line pressure. Manual control of this valve is also provided from the control room. This system is designed to relieve 100,000 lbs per hour with normal min steam line pressures.
The main condenser condenses steam from the turbine generator low pressure exhaust and from the turbine steam bypass line. The condenser is a single pass, 85,000 sq. ft. unit, consisting of 12,h95 - 7/8" OD tubes with a 30 ft. overall length. The condenser hotwell is of the deaerating type. The loss of condenser vacuum to 18 in. Hg will cause a turbine throttle valve trip whicn in turn will cause a reactor scram from loads above 15 MWelectric. Fur-ther decrease of vacuum to lh in. Hg will cause the turbine steam bypass valve to close.
- 11. Waste Disnosal System (Section 209, Final Hazards Summary Report) b Solid Wastes Combustible solid wastes will be handled in fiber drums and burned in a specially designed incinerator. Exhaust gases will be diluted, scrubbed, and filtered to remove entrained solid particles and then passed to the incin-erator stack. Ashes resulting from combustion will be mixed with water and cement in a $$ gal. steel drum, the mixture allowed to solidify, and the result-ant solid mterial held for shipment.
Non-combustible wastes will be immobilized and shielded in concrete in steel drums where practical. Where size makes such packaging impossible, material will be shielded and stored in a restricted area until arrangements can be completed for permanent storage or burial at an A.E.C. approved disposal area.
Liquid Wastes The largest part of the activity appearing as liquid wast e originates in the primry plant in the form of dissolved or suspended materials resulting from corrosion, fission product escape, or activation of. impurities in makeup water. Lesser quantities appear in laboratay wastes, and miscellaneous drains.
These wastes will be treated by passage through ion exchange resins or by evaporation to a more highly concentrated solution in a specially designed evaporator. Certain liquids miy also be held in storage for decay of short-lived activity either before or after treatment.
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The concentrated solutions will be mixed with cement in steel barrels and allowed to harden into a form suitable for long term storage or shipment off-O site.
Reclaimed water from the evaporator will be suitable for reuse in the plant or for discharge to the river after dilution 'sy the condenser circulating water flow of approximately lh0,000 gpm. The activity in the purified water be-fore dilution is calculated to be 2.03 x 10-7 microcuries per ml. The activity after dilution will not be allowed to exceed the limit of 1.0 x 10-7 microcuries per ml established in 10 CFR 20.
Ion Exchange Resins These resins are used for continuous cleanup of the bleed and feed stream of rnin coolant water during normal operation and also as part of the waste treatment facility. The resins are loaded into disposable steel c./linders for use at the plant. While in use these cylinders of resin are operated under several feet of water for protection of plant personnel against radiation from material which accumulates on the resin bed. After the resins are exhausted, they are not recharged. The cylinders are disconnected and moved into storage position - still under several feet of water - where the activity is allowed to decay. They are then removed, with such shielding as may be required, for dis-posal off-site. This design eliminates the handling of activated resins in bulk form or disposal of solutions which have been activated in the course of regener-ating resins.
Gaseous Wastes b
These are primarily fission product gases, resulting from failed fuel elements, mixed with hydrogen introduced into the primary system water for corro-sion control. These gases come cut of solution either in the low pressure surge tank or in the waste treatment evaporator and are pumped directly to a compressed gas surge drum. This drum acts as a reservoir of gases free from air or oxygen which are used to fill certain covered tanks in the waste disposal system when liquids containing dissolved hydrogen are removed.
The mixed gases colleet gradually in the waste gas surge drum and the excess over tank filling requirements will be removed once each month and stored in one of three gas decay drums for about 60 days to reduce the activity.
After decay, the gases are discharged hrough a filter to remove particulates, diluted by air from a 15,000 CFM purge fan, and vented to the atmosphere through the primary vent stack. The decayed gas is emitted at 26 fps and dilution will occur by entrainnent even under unfavorable meteorological conditions. By this means discharges will be kept below the proposed MPC of L.5 x 10-7 microcuries per ml allowed for two-thirds of the time in an unrestricted area.
12.
Vapor Containment (Section 231, Final Hazards Summary Report) l The vapor container is a 125 foot steel sphere which encloses all pressurized components of the Main Coolant System. It is designed, built and tested in accordance with the ASME Boiler and Pressure Vessel Code,Section VIII, Unfired Pressure Vessels,1956 Edition and latest addenda. It is not provided with a relief valve, in accordance with special ruling, Case No.1235.
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The 10% over-pressure, permitted by the code on the design pressure of 31.5 psi gage, results in an allowable pressure of 3h.5 psi gage which is slightly O
higher than the pressure developed by a complete loss of main coolant accident.
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i The leakage rate for the vapor container was specified to be less than 0.1 wt. % of the contained air during a 2h hour interval at 15 psi gage internal pressure, which corresponds to 70 cu ft per hr (S T P).
The vapor container will be operated at not more than 2 psi gage positive pressure aM leakage will be monitored during operation in order to establish the leakage rate and to guarG against any possibility of gross leakage through improper closure after opeM, the vapor container. This monitoring u ll be accomplished by use of the ame apparatus as was employed in the deter-mination of the initial leak rate Pipe lines penetra+1ng the vapor container, which are used only when the plant is not in operation, are provided with valves located outside the vapor container. These valves are closed whenever the reactor is critical or when the main coolant system is pressurized with nuclear fuel in place. Incoming pipe lines, used for operation of the plant, are provided with two check valves, one inside and one outside the vapor container. Outgoing lines, used for opera-tion of the plant, are each provided with a trip valve arranged to close auto-matically following a pressure rise in the vapor container.
A double door, air lock type personnel access opening is provided for entrance to and egress from the vapor container without opening it to the atmosphere.
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- 13. Radiation Shielding (Section 232, Final Hazards Summry Report)
Radiation shielding is provided to allow safe operation and mainten-ance of the plant and is mde up of the neutron shield tank surrounding the re-actor vessel, the concrete primary shield surrounding the neutron' shield tank, the concrete secoMary shield surrounding and separating all equipment within the vapor container, the fuel hanc' ling shielding and auxiliary shielding in the control room, waste disposal building and elsewhere.
The control room is shielded from the vapor container by a concrete wall four feet thick so as to be tenable even under postulated accident condi-tions.
4 The shielding thicknesses were designed to meet the following criteriar At h85 Fit -
Working stations 0.75 mr per hr Intermittently manned ground level areas 2
mr per hr During Refueling -
Average.
2 mr per.r v) i r
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lb. Reactor Protection System The neutron flux monitoring devices which are used in the operation 4
of the reactor are located external to the reactor vessel. The circuitry asso-ciated with these devices incorporates fali.-safe design throughout. Detectors for the high start-up rate and high flux level protection are located in the neutron shield tank surrounding the reactor vessel. High start-up rate protec-tion is provided by two neutron detecting channels in the source range and two neutron detecting channels in the intermediate range. High flux level scram normally has a coincidence feature which requires two out of three power range I
channels to initiate a scram, but may be operated to initiate a scram from a single channel.
The three intermediate and three power range nuclear detectors are spaced circumferentially around the core and at elevations above and below the center of the core. By observing changes in ralative nux readings, shown by these six detectors, axial or radial flux oscillations can be detected and cor-rected.
i Table I, below, lists the functions comprising the reactor protec-tun system, the number of channels associated with each Ibnction, the number of channels required to initiate a scram signal, and the set points for each function.
A permissive relay circuit is operated from a pressure switch actuated fron the first stage turbine steam pressure when the pressure is equiv-alent to 15 Welectric output. This circuit prevents automatic withdrawal of the control rods and bypasses the low flow and low pressure scrams when the power is below 15 Welectric corresponding to 10f, reactor power. At 15 Welec-tric and above, the low flow and low pressure scrams are reconnected, the start-up rate scram is bypassed, and the system is available for automatic control of the control rods. A permissive relay bypass switch is available to connect the scram signals directly to the shutdown amplifiers.
An emergency power supply consisting of an engine driven AC generator unit is provided to maintain service to vital equipment in the unlikely event i
of a sustained total loss of electrical power. The 75 W, h80 V generator ener-gizes two of the h80 V station busses and is so sized that one motor-generator i
battery charging set, pressurizer heaters, one charging pump and process instru-mentation may be operated to mintain the plant in a safe condition during a prolonged power outage.
D.
OPERATING PRINCIPLES l
The basic operating principles that will be adhered to in the operation of the plant are as follows:
l 1.
The safety of plant operation will be proven by comprehensive Startup and Power Operation tests before being placed in regular operation.
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2 Operation of the plant will be accomplished primarily from the centralized control room, manned by two operators at all times except during reactor refueling or other periods of cold shutdown.
TABC I REAC h t PR T TPODCS No. of Channels Punction Requi-eti to Trip Bypasse s Set Point Limits Remrks :
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High Startup Rate -
1 cut of 2 Auto typass 5 2 dec/ min - rax g
Reactor Scram Above 15 De High Neutron 2 out of 3 None
!!eutron flux corresponds to l
Flux Level -
120% rated power * - max h
b Reactor Scran E
Low Main Coolant 1 out of 2 Auto bypass 1600 psig - min This cetting is considered appropriate j.
Pressure -
Below 15 We for operation at any power level up to ?
Reacter Scran L85 W thernal.
p Lew Main Coolant 2 out cf14 Auto bypass 60% of nor al nain coolant Tiew -
Below 15 We flew - nin Reactor Scram E_
Farr2a1 Not Applicable None Available ary tine g
Reactor Scran P
1 out of 2 Auto bypass Usual Turbine Trip set points Turbine trip scrams reactor arxi trips O
Reacter Scram Below 15 1Me generator above 15 We.
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Generator Trip -
1 mt of 2 Auto bypass Usual Generator Trip set points Genera *.or trip scrans reactor arvi trips ll Reactor Scram Below 15 We turbirs above 15 PMe.
ll Fanual Turbine Not Applicable Auto bypass Available ary tine above 15 PMe
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Generatcr Trip -
Belcw 15 We i
Reactar Scram l
Safety Injection 2 out of h for None 270 psig main coolant pressure-min During normal operation system is set I
Cperation autoratic initiation.
5 psig vapor container pressure-max for automatic initiation.
Safety j
(Requires coincident injection may be initiated nanually at ;
cperation of 1 of 2 any time.
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sure signals and 1 of 2 high vapor con-tainer pressure signals.)
Pressurizer Not Applicable None 2h8$ psig, No.1 Valve - max Safety Valve 2560 psig, No. 2 Valve - rax Operation Vapor Container 1 out of 2 None 5 psig vapar container pressure-max Closes trip valves in outgoing lines Outgoing Line Trip-in vapor container.
Valve Operation l'ain Coolant Valve Not Applicab1?
None 300 F temperature differential - rax Prevents ar;y isolated loop from being Interlock placed into operation if the loop temperature is mora than 300 F lcwer than highest cold leg temperature in the remaining loops.
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" Rated power" is 32 Et or such higher power up to h85 PMt as may be authorized.
Notes Recctor scram always initiates turbine and generator tr$.
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3.
Some plant operation will be perfctmed at local panels under direction of the control room operators.
h.
Although all plant operation will be by written procedure, it is recognized that an operator error in possible. There-fore, plant design is such that no single operator or equip-ment failure can cause an accident of serious consequence.
5.
The plant will be divided into two general areas, one to be considered potentially contaminated and the other clean.
Access and operations in the potentially contaminated area vill be subject to restrictions in order to control exposure of personnel. Exposure records on all personnel will be maintained.
6.
Integrity of the vapor container will be maintained at all times when the reactor is critical or when the main coolant system is pressurized with the reactor core in place.
7.
Limited access to the vapor container will be provided when the nuclear plant is at operating temperature and pressure but with the reactor sub-critical. Access into the vapor container will not be allowed with the reactor critical, ex-cept for special test reasons.
8.
The nuclear plant and the waste disposal system will be opeo m
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ated in such a manner as to maintain a balanced water inventcry so that little or no liquid waste is discharged from the plant.
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9.
Maintenance work on the steam plant and portions of the nuclear plant outside of the vapor container will be performed with the plant in operation if the work can be carried out in a manner consistent with plant and personnel safety.
10.
Irradiated fuel and control rods will be removed, transported and stored by semi-remote methods under a sufficient depth of water to provide adequate cooling and shielding.
- 11. The plant will be operated in compliance with all applicable rules and regulations of the Commonwealth of Massachusetts and of the Federal Government, armi in a manner consistent with best industrial practices for large electric generating stations.
- 12. All significant unexpected incidents, u tsafe acts or incidents of excessive exposure to radiation will be investigated by the Plant Safeguards Committee with a view to altering procedures and thereby preventing recurrence.
- 13. In the event of any situation which may threaten the safety of continued operation, it will be required procedure to shut the plant down immediately and to take such other planned emergency actions as are necessary to protect persons and property.
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E.
REACTIVITY LIMITATIONS O
V 1.
The hot, clean, zero power reactor will be at least 3% dk/k sub-critical with all control rods inserted and without beric acid in the main coolant water.
2.
Sufficient boric acid will be added to the main coolant system prior to cold shutdown to maintain the cold core with all control rods inserted at least 5% dk/k sub-critical.
3.
The reactivity insertion rate due to withdrawal of the highest worth control rod group will not exceed 2.1 x 10-h dk/k/second.
F.
INITIAL C(PE LOADDD AND LOW PCWER NUCLEAR CORE TESTS 1.
Pre-Cort Loading Test _s The main coolant system and the auxiliary systems will be filled, cleaned, hydrostatically tested and Ihnetionally tested to establish their in-tegrity and performance according to the procedures and to the standards des-cribed in Sections $03D1 and 503D2 of tLs final Hazards Summary Report.
2 Initial Core Loading Upon completion of Pre-Core Loading Tests described in (1) above, loading of the initial core will begin according to procedures outlined in Sec-
[m) tion 503E1 of the final Hazards Summary Report. Fuel assemblies will be loaded
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ane by one in water sufficiently borated to render the fully loaded core at least 10% dk/k sub-critical at room temperature. Temporary BF3 detectors will be placed adjacent to the fuel and a plot of inverse source multiplication against number of fuel assemblies in the core will be made as loading progresses.
If at any point in the loading procedure the plot indicates that criticality might occur if the rnunber of fuel assemblies then in the core were doubled, the boron concentration will be increased sufficiently to correct this condition before additional fuel assemblies are inserted. In additica, emergency shutdown will be provided during the loading operation by having the chemical shutdown system lined up and ready for use under control of the fuel loading operator.
Negative reactivity can be added by this means at the rate of 1 x 10-4 Ak/k per second.
After completion of core loading, the response of the normal plant nuclear instrumentation to the core neutron sources wir be checked to establish performance in accordance with design specifications.
.4 remaining reactor in-ternals will be installed, the reactor vessel head wil. be bolted in place and the main coolant system filled with water of the same boron concentration as that already in the core.
3 <, Initial Crit _icality After installation of the pressure vessel head, the control rod drive opezating and position-indicating coils will be installed and tested and h
operation of the control rod drives will be checked one by one, according to
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procedures outlined in Section 503E2 of the final Mazards Summary Report.
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15.
The reactor will then be brought to criticality according to proce-dures outlined in Section 503E3 of the fiml Hazards Summry Report. During the O'
approach to criticality, the main coolant system will remain at ambient tempera-ture and pressure, with circulation of the borated water provided by the shutdown cooling system. The outside group of control rods will be partially withdrawn to pravide s&fety protection during rubsequent control rod motion. The remaining control rods will be withdrawn in increments so as to maintain the several rod groups at approximtely the same axial position in the core. Control rod with-drawal will be regulated by observing the normal nuclear instrumentation and by graphic plots of inverse core multiplication plotted against banked control rod position.
h.
Other Low Power Testing Following initial criticality a number of tests will be performed to determine experimentally such reactor characteristics as control rod and boron reactivity worths, and temperature, pressure and flow reactivity coefficients.
These tests and the procedures to be followed in performing them are outlined in Section 503E of the final Hazards Summary Report. During this series of tests reactor poweir level will not exceed 5 Wthermal.
G.
NUCLEAR PLANT PORER OPERATIONAL TESTS AND PU4ER OPERATION Following initial criticality, completion of low power nuclear core tests and verification that reactor characteristics are in close agreement with design parameters and coefficients, a series of power operational tests will be
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made as described in Section 503F of the final Hazards Summry Report.
During this series of tests the reactor and plant will be brought in incremental steps from zero power to the design rating of the initial core, 392 K4 thermal and 110 M4 net electrical. Power increments will be not more than 30 K4 electrical.
Reactor and plant characteristics as determined at each power level must be in satisfactory agreement with design parameters before making the next incremental increase to a higher power level.
At the conclusion of the series of tests which will demonstrate that reactor characteristics are f n close agreement with design calculations and that j
the plant is capable of opere'.ing at the initial core design rating, the plant will be operated for not less than 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, as continuously as possible, at 392 M4 thermal,110 M4 not electrical.
After 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> of operation at 392 W thermal, and verification of operating characteristics at this level, the power level will be increased above 392 W thermal in progressive steps of not more than 30 M4 thermal, provided that before making each step calculations and experim ntal data indicate that, at these higher power levels:
a.
Reactor accidents do not involve hazards greater than nor different frc:1 those already analyzed in the final Hazards Summary Report; and, b.
The heat flux at the point closest to burnout in the hottest channel is less than 50% of the burnout heat flux; and, g
V 16 c.
Bulk boiling does not occur at the outlet of the hottest f')
channel; and, o
d.
A sufficient margin is preserved between the maximum fuel center temperature and the limiting fuel center temperature that is established at that time:
In any event the ultimate pcwer level will not be in excess of h85 M4 thermal. Yankee will give the Commission 30 days advance notice of its intent to mke a step increase in power level above 392 X4 thermal and at the saw time will provide the Commission with full information in support of the proposed step increase in power level. Unless the Ccmmission advises Yankee otherwise during the 30 day period, Yankee may commence operation at the in-1reased power level specified in its notice.
H.
ROUTINE NUCLFAR TESTS The measurements made in Section G, above, will yield information on initial nuclear parameters which will provide a basis for comparison with results of similar experirents to be performed at intervals throughout core life. These routine and periodic tests are described in Section 509 of the final Hazards Summary Report. Included are measurements of the power coeffi-cient of reactivity and the moderator temperature coefficient of reactivity which will be made after each 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of full power operation of the initial c or e. These continuing tests will provide a means of detecting any change in the kinetic characteristics of the core due either to plutonium buildup or g
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uranium depletion.
Since changes will be gradual, these tests will allow d
evaluation of the situation at intervals throughout core life before reactor kinetic Jharacteristics have changed materially.
I.
MAINTENANCE Primry plant maintenance will be grouped into two categories:
1.
Major maintenance of equipment within the vapor container will not be performed until such time as the plant is shut down and depressurized.
2.
Minor maintenance of equipment outside of the vapor container will be performed with the plant in operation. Minor main-tenance or adjustment of equipmant inside of the vapor con-tainer, for which depressurization is not r quired and for which entrance is possible without hazard, sill also be per-forred with the main coolant system pressurized and the reactor sub-critical.
All maintenance operations on contaminated equipment or in contami-nated areas will be supervised by Technical Services personnel to assure that proper radioactivity safeguards and decontamination procedures are observed.
,m Secondary plant maintenance will be performed using conventional power plant methods and procedures as the appearance of radioactivity in this
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system is prevented by the layout of the plant, the design of equipment and
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N 17.
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f-administrative controls. Continuous mcnitoring and test pregrams further guard against the appearance of radioactivity in this systen.
Safety procedures and rules provide for proper clearance routines, working nethods, tool inspections and safety instruction for plant personnel.
J.
REPORTS (1)
Yankee shall nake an inmediate report in writing to the Commission of any sic d G : ant indication or occurrence of an unsafe condition relating to the operation of the facility.
(2)
Yankee shall report in writing to the Commission any changes in the ner.bership of the Plant Safeguards Committee within ten days after any such change is made.
(3)
Ninety days after completion of startup testing and the initial 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> of operation at 392 M4 thernal, Yankee will submit a report describing the result s of these operations.
(h)
Yankee shall submit to the Commission a quarterly report for each quarter during the first year after completion of the foregoing test program.
Such quarterly reports shall be filed within 30 days after the end of the quarter covered by the report. Thereafter, Yankee shall file an annual report. The first
/~~ 'y such annual report shall be filed within thirteen months after the filing of the
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fourth quarterly report referred to above. Each report filed under this para-graph shall include a description of operating experience pertinent to safety and changes in facility design, performance characteristics and operating proce-dures during the reporting period.
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