ML20058C447

Chapter 9 to Hb Robinson FSAR, Auxiliary & Emergency Sys

ML20058C447
Person / Time
Site:	Robinson
Issue date:	05/08/1970
From:	CAROLINA POWER & LIGHT CO.
To:
References
NUDOCS 8207260350
Download: ML20058C447 (153)

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SUMMARY

OF OFF-SITE DOSES FROM LOSS-OF-COOLANT ACCIDENT THYROID DOSES - REM Site Boundary - 2 Hours LPZ - 30 Days 10CFR100 Guidelines 300 300 teaign Basis Accident 1.3 0.45 Hypothetical Accident 116 26 3 WHOLE BODY DOSES - REM Site Boundary - 2 Hours LPZ - 30 Days 10CFR100 Guidelines 25 25 Design Basis Accident 0.009 0.004 Hypothetical Accident 4 0.7 8207260350 700508 PDR ADOCK 05000261 K PDR Amendraeat 3

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a. r ^ :,..~; _ ^~ ~.. :.GL :.~. 1 i APPE:: DIX 14A 4 H. ".. Pobin x Unit 'o. 2 Supplermnt The FSAM for tim li. B. Robin.:on Unit ? o. 2 plant presented a cor plete analyni:. for the lom:-of-coolant accident (14.3.2) at a peah lc7/ft of 1 19.1, which corronpends to 102% of the maximum calculated thermal l 7 rat inn of 2300 ' lt. The conservatisan inherent in this analysis were disc us ;ed on pap 14.3.2-12. A maximum clad temperature of 2450 F uas 1 predicted to occur for the double ended cold leg break. At a peak kw/ft of 18.3, thich corresponds to 102% of the initia! rating of 2200 K t, I the predicted po.' clad temperature would be reduced to a value of 2289*f. In Amendnent 13 Tcb Ih to the FSAR it was shown that the design per.h clad temperature and retal water reactions calculated were subatantial.'y belou veceptable va l u." e determined by the Westinghouse Experimental Rod Quench Tents. I t u., therefore concluded that the quench mechanism during the reflooding ph:u2 of a LOCA does not lead to rod shattering or loss of integrity over the range of conditions conservatively estimated for the li. D. Robine.on Unii ::0. 2 LOCA analysis. It has been <:uggested that the capability of Emergency Core Cooling Systen should be incre.:ed to reduce the maxinun clad temperatures. It should be emphasized that additional margins are available by virtue of the conser-i vat is~, in the analysis. Any changes to the Emergency Core Cooling Systcm are therefore n:r. a r ran ted. l l l Improved analytical nodels, which have been developed and experinentally verified,(1) nw cake it posrible to quantify the addi.tional conservatis, that it clai: ed in the design of the Emergency Core Cooling Systen. i J. U. Da r ry cot t and J. She f chech "PWR Core Behavior Following a Loss of Coolant bcident", 1: CAP 7422-L, January 1970. (Westinghoirse Proprie tary) 14A-1 Amendment 7 ~

3 APPENDIX 14B STEAM GENERATOR TUBE RUPTURE Justification of 70,000 Pounds Carryover to the Secondary Side The break flow following a complete tube failure is calculated to be approximately 80 lbs per second at a pressure differential of 1500 psi between the primary and secondary systems, and it is assumed that as the pressure varies the resultant flow is proportional to the square root of the pressure differential. Immediately after the postulated accident, pressurizer pressure and level will decrease until the reactor trip point is reached in about 3 minutes. During this period the average break flow is about 74 lbs per second corresponding to the reduced primary to secondary pressure differential and the total mass transfer in the first 3 minutes is therefore 13,300 lbs. 10 () Following reactor trip, the pressurizer level will fall rapidly and the safety injection actuation signal will be generated leading to actuation of the safety injection pumps. In the first two or three minutes af ter plant trip, the primary system pressure will dip through a minimum and eventually reach stability at the pressure where the flai through the break is balanced by incoming safety injection flow. This flow is about 270 gpm or 37 lbs/ l second. Since the reactor pressure passes through a minimum before reaching equilibrium at 1400 psia, the 37 lbs/second is the average break flow in the twenty-seven minute interval subsequent to plant trip. The total mass transfer in this interval is therefore 60,000 lbs. l Thus, in the absence of any Operator action, approximately 73,300 lbs of primary system fluid could be transferred to the secondary system in the 30 minute period after the accident and prior to the time the faulty steam generator is isolated. l The mass transfer estimated above is pessimistic in two respects.

First, lllI no consideration is given to the fact that the safety injection pump would 14B-1 Amendment 10 1

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be regulated when water level returns in the pressurizer, as suggested in the Emergency Operating Instructions. This would result in an average break flow of less than the 37 lbs/second assumed above. Secondly, as decay heat and core stored heat sources gradually reduce after plant trip, the reactor coolant temperature will gradually decrease. Therefore in order to maintain an equilibrium pressure, the break flow will be less than the incoming safety injection flow during the twenty-seven minutes subsequent to plant trip. Thus the /0,000 pounds was taken as a conservative mass carryover for this postulated accident. g(. / Amendment 10 'E

TABLE OF CONTENTS Section Title Page ,~ 9 AUXILIARY AND EMERGENCY SYSTEMS 9-1 9.1 General Design Criteria 9.1-1 9.1.1 Related Criteria 9.1-1 Reactivity Control System Malfunction 9.1-1 Engineered Safety Features Performance Capability 9.1-1 Containment Heat Removal Systems 9.1-2 9.2 Chemical and Volume Control System 9.2-1 9.2.1 Design Bases 9.2-1 Redundancy of Reactivity Control 9.2-1 Reactivity Hot Shutdown Capability 9.2-2 Reactivity Shutdown Capability 9.2-2 Reactivity Hold-Down Capability 9.2-3 Codes and Classifications 9.2-4 9.2.2 Systen Design and Operation 9.2-5 Expected Operating Conditions 9.2-9 Reactor Coolant Activity Concentration 9.2-9 Reactor Make-up Control 9.2-11 Automatic Make-up 9.2-12 Dilution 9.2-13 Boration 9.2-13 Alarm Functions 9.2-14 I 'T Charging Pump Control 9.2-14 1._ / Components 9.2-15 Regenerative Heat Exchanger 9.2-15 Letdown Orifices 9.2-16 Non-Regenerative Heat Exchanger 9.2-16 Mixed Bed Demineralizers 9.2-17 Cation Bed Demineralizers 9.2-17 Resin Fill Tank 9.2-18 Reactor Coolant Filter 9.2-18 Volume Control Tank 9.2-18 Charging Pumps 9.2-19 Charging Pump Accumulators 9.2-19 Chemical Mixing Tank 9.2-19 Excess Letdown Heat Exchanger 9.2-20 Seal Water Heat Exchanger 9.2-20 Seal Water Filter 9.2-21 Seal Water Injection Filters 9.2-21 Boric Acid Filter 9.2-21 Boric Acid Tanks 9.2-22 Boric Acid Tank Heaters 9.2-22 Batching Tank 9.2-22 Boric Acid Transfer Pumps 9.2-23 Boric Acid Blender 9.2-23 l 9-1

TABLE OF CONTENTS (Cont'd) Section Title Page Recycle Process 9.2-24 Holdup Tanks 9.2-24 Holdup Tank Recirculation Pump 9.2-24 Gas Stripper Feed Pumps 9.2-24 Base and Catin Ion Exchanger 9.2-24 Ion Exchanger Filters 9.2-25 Cas Stripper Equipment 9.2-25 Boric Acid Evaporator Equipment 9.2-26 Evaporator Condensate Demineralizers 9.2-26 Condensate Filters 9.2-27 Monitor Tanks 9.2-27 Monitor Tank Pumps 9.2-27 Primary Water Storage Tank 9.2-27 Primary Water Storage Pumps 9.2-27 Concentrates Filter 9.2-28 Concentrates Holding Tank 9.2-28 Concentrates Holding Tank Transfer Pumps 9.2-28 Deborating Demineralizers 9.2-28 Electrical Heat Tracing 9.2-29 Valves 9.2-30 Piping 9.2-30 9.2.3 System Design Evaluation 9.2-31 Availability and Reliability 9.2-31 Control of Tritium 9.2-31 Leakage Prevention 9.2-32 Incident Control 9.2-33 Malfunction Analysis 9.2-33 Galvanic Corrosion 9.2-35 9.3 Auxiliary Coolant System 9.3-1 9.3.1 Design Bases 9.3-1 Performance Objectives 9.3-1 Component Cooling Loop 9.3-1 Residual Heat Removal Loop 9.3-1 Spent Fuel Pit Cooling Loop 9.3-2 Design Characteristics 9.3-2 Component Cooling Loop 9.3-3 Residual Heat Removal Loop 9.3-3 Spent Fuel Pit Cooling Loop 9.3-4 Codes and Classifications 9.3-4 9.3.2 System Design and Operation 9.3-4 Component Cooling Loop 9.3-4 Residual Heat Removal Loop 9.3-6 Spent Fuel Pit Cooling Loop 9.3-6 Component Cooling Loop Components 9.3-7 Component Cooling Heat Exchangers 9.3-7 Component Cooling Pumps 9.3-8 Component Cooling Surge Tank 9.3-8 Chemical Pot Feeder Tank 9.3-8 Component Cooling Valves 9.3-8 Component Cooling Piping 9.3-9 9-11 X

i e > { TABLE OF CONTENTS (Cont'd) %r Section Title Page Residual Heat Removal Loop Components 9.3-8 Residual Heat Exchangers-9.3-8 Residual Heat Removal Pumps .9.3-9 Residual Heat Removal Valves 9.3-9 Residual Heat Removal Piping 9.3-10 Spent Fuel Pit Loop Components 9.3-10 l l Spent Fuel Pit Heat Exchanger 9.3-10 Spent Fuel Pit Pump 9.3-10 Refueling Water Purification Pump 9.3-11 Spent Fuel Pit Strainer 9.3-11 Spent Fuel Pit Filter 9.3-11 Spent Fuel Pit Demineralizer 9.3-11 i Spent Fuel Pit Skimmer 9.3-11 i Spent Fuel Pit Valves 9.3-11 Spent Fuel Pit Piping 9.3-12 9.3.3 System Evaluation 9.3-12 Availability and Reliability 9.3-12 Component Cooling Loop 9.3-12 Residual Heat Removal Loop 9.3-13 j Spent Fuel Pit Cooling Loop 9.3-13 i Leakage Provisions 9.3-13 Component Cooling Loop 9.3-13 Residual Heat Removal Loop 9.3-15 Spent Fuel Pit Cooling Loop 9.3-16 Incident Control 9.3-17 3 Component Cooling Loop 9.3-17 Residual Heat Removal Loop 9.3-18 Spent Fuel Pit Cooling Loop 9.3-18 i l Malfunction Analysis 9.3-19 l 9.3.4 Test and Inspection Capability .9.3-19 9.4 Sampling System 9.4-1 j 9.4.1 Design Basis 9.4-1 Performance Requirements 9.4-1 Design Characteristics 9.4-1 i High Pressure - High Temperature Samples 9.4-2 Low Pressure - Low Temperature Samples 9.4-2 l Expected Operating Temperatures 9.4-2 i Codes and Standards 9.4-2 l 9.4.2 System Design and Operation 9.4-3 Components 9.4-5 1, Sample Heat Exchangers 9.4-5 j Delay Coil 9.4-6 l Sample Pressure Vessels 9.4-6 Sample Sink 9.4-6 Piping and Fittings 9.4-7 Valves 9.4-7 0 4 9-111 1 . _,. _ _ _. _ _ _, _ _ _ _ _. _ ~. _.. _ _ _ _ _ _ _,

TABLE OF CONTENTS (Cont'd) Section Title Page 9.4.3 System Evaluation 9.4-8 Leakage Provisicns 9.4-8 Incident Control 9.4-8 Malfunction Analysis 9.4-8 9.5 Fuel Handling System 9.5-1 9.5.1 Design Basis 9.5-1 Prevention ot Fuel Storage Criticality 9.5-1 Fuel and Waste Storage Decay Heat 9.5-2 Fuel and Waste Storage Radiation Shielding 9.5-2 Protection Against Radioactivity Release From Spent Fuel and Waste Storage

9. 5-3 9.5.2 System Design and Operation 9.5-4 Major Structures Required for Fuel Handling 9.5-5 Reactor Cavity 9.5-5 Refueling Canal 9.5-5 Refueling Water Storage Tank 9.5-6 Spent Fuel Storage Pit 9.5-6 New Fuel Storage 9.5-7 Major Equipment Required for Fuel Handling 9.5-7 Reactor Vessel Stud Tensioner

9. 5-7 Reactor Vessel Heading Lifting Device 9.5-8 Reactor Internals Lifting Device 9.5-8 Manipulator Crane 9.5-8 Spent Fuel Pit Bridge 9.5-10 Fuel Transfer System 9.5-11 Rod Cluster Control Changing Fixture 9.5-11 Refueling Sequence of Operation 9.5-11 Preparation 9.5-11 kefueling 9.5-13 Reactor Reassembly 9.5-13 9.5.3 System Evaluation 9.5-16 Incident Protection 9.5-16 Malfunction Analysis 9.5-16 9.5.4 Test and Inspection Capability 9.5-17 9.6 Facility Services 9.6-1 9.6.1 Fire Protection System 9.6-1 Design Bases 9.6-1 System Design and operation 9.6-2 Water Supply 9.6-3 Alarm System 9.6-5 9

9-iv x

l i 1 ( i j TABLE OF CONTENTS (Cont'd) l Section Title Page 3 I l 9.6.2 Service Water System 9.6-5 ,1 Design Basis 9.6-5 System Design and Operation 9.6-6 Tests and Inspections 9.6-9 I 4 l 9.7 Equipment and System Decontamination 9.7-1 9.7.1 Design Basis 9.7-1 9.7.2 Methods of Decontamination 9.7-2 9.7.3 Decontamination Facilities 9.7-3 I O 1 l l i i l 4 i i i 1 !till I l i l i i 1 I i i I I j 9-v i i

LIST OF TABLES O ( Table Title w 9 AUXILIARY AND EMERGENCY SYSTEMS 9.2-1 Chemical and Volume Control System Code Requirements 9.2-2 Chemical and Volume Control System Performance Requirements 9.2-3 Principle Component Data Summary 9.2-4 Parameters Used in the Calculation of Reactor Coolant Fission Product Activities 9.2-5 Reactor Coolant System Equilibrium Activities 9.2-6 Tritium production in the Reactor Coolant 9.2-7 Malfunction Analysis of Chemical and Volume Control System 9.3-1 Component Cooling Loop Component Data 9.3-2 Residual Heat Removal Loop Component Data 9.3-3 Spent Fuel Cooling Loop Component Data 9.3-4 Auxiliary Coolant System Code Requirements 9.3-5 Failure Analysis of Pumps, Heat Exchangers and Valves 9.4-1 Sampling System Code Requirements 9.4-2 Sampling System Components 9.4-3 Malfurction Analysis 9.5-1 Fuel Handling Data \\m_/ 9.6-1 Service Water System Design Flows t I v 9-vi

LIST OF FIGURES Figure Title 9 AUXILIARY AND EMERGENCY SYSTEM 9.2-1 Chemical and Volume Control System, Sheet 1 9.2-2 Chemical and Volume Control System, Sheet 2 9.2-3 Chemical and Volume Control System, Sheet 3 9.3-1 Auxiliary Coolant System 9.3-2 Auxiliary Coolant System - Residual Heat Removal 9.3-3 Auxiliary Coolant System - Spent Fuel Pit Cooling 9.4-1 Sampling System 9.5-1 Fuel Transfer System 9.6-1 Fire Protection System 9.6-2 Service Water System G O 9-vii x

9. AUXILIARY AND EMERGENCY SYSTEMS ( ( ) The Auxiliary and Emergency Systems are supporting systems required to insure the safe operation or servicing of the Reactor Coolant System (detailed in Section 4). In some cases, the dependable operation of several systems is required to protect the Reactor Coolant System by controlling system conditions within specified operating limits. Certain systems are required to operate under emergency conditions. This section considers systems in which component malfunctions, inadvertent interruptions of system operation, or a partial system failure may lead to a hazardous or unsafe condition. The systems considered under this category are: Chemical and Volume Control System This system provides for nuclear poison fluid injection, chemical additions for corrosion control, reactor coolant b (_,/ clean-up and degasification, reactor coolant make-up, reprocessing of water letdown from the Reactor Coolant System, and reactor coolant pump seal water injection. Auxiliary Coolant System This system provides for transferring heat from reactor plant waters to the service water system and consists of the following three loops: The residual heat removal loop removes residual and sensible heat from the core and reduces the temperature of the Reactor Coolant System during the second phase of plant cooldown. The spent fuel pit loop removes from the spent fuel pit the heat generated by stored spent fuel elements. O 9-1

The component cooling loop removes residual and sensible heat frcm the Reactor Coolant System, via the residual heat removal loop, during plant shutdown, cools the spent fuel pit wter and the letdown flow to the Chemical and Volume Control System during power operation and provides cooling to dissipate waste heat from various primary plant components. Sampling System This system provides the equipment necessary to obtain liquid and gaseous samples frotr the reactor plant systems. Facility Service Systems These systems include fire protection and service water systems. Fuel llandling System This system provides for handling fuel assemblies, control rod assemblies, core structural components and material irradiation specimens. Equipment and System Decontamination These procedures provide for the removal of radioactive deposits from primary system surfaces, shipping casks and tools. O 1 i l I i i O 9-2

4 ) 9.1 CENERAL DESIGN CRITERIA ~ \\ The criteria which apply prim'rily to other systems discussed in other a Sections are listed and cross-referenced because details of directly related systems and equipment are given in this Section. Those criteria which are specific to one of the Auxiliary and Emergency Systems are listed and discussed in the appropriate system design basis section. 9.1.1 RELATED CRITERIA Reactivity Control Systems Malfunction Criterion: The reactor protection systems shall be capable of p'rotecting against any single malfunction of_the reactivity control systim, such as unplanned continuous withdrawal (not ejection or dropout) of a control rod, by limiting reactivity transients to avoid exceeding acceptable fuel damage limits. (GDC 31) U As described in Section 7 and justified in Section 14, the Reactor Protection Systems are designed to limit reactivity transients to DNBR > 1.30 due to any single malfunction in the deboration controls. Engineered Safety Features Performance Capability Criterion: Engineered Safety Features such as the emergency core cooling system and the containment heat removal system shall provide suf ficient performance capability to accommodate the failure of any single active component without resulting in undue risk to the health and safety of the public. (GDC 41) Each of the auxiliary cooling systems which serves an emergency function provides sufficient capability in the emergency operational mode to accommodate any single failure of an active component and still function in a manner to avoid undue risk to the health and safety of the plant personnel and the public. O V 9.1-1

I O Containment Heat Removal Systems Criterion: Where an active heat removal system is needed under accident conditions to prevent exceeding containment design pressure this system shall perform its required function, assuming failure of any single active component. (GDC 52) Each of the auxiliary cooling systems which serves an emergency function to prevent exceeding containment design pressure, provides sufficient capability in the emergency operational mode to accommodate any single failura of an active component and still perform its required function. O O 9.1-2

9.2 CHEMICAL AND VOLUME CONTROL SYSTEM () The Chemical and Volume Control System a) adjusts the concentration of chemical neutron absorber for chemical reactivity control, b) maintains the proper water inventory in the Reactor Coolant System, c) provides the required seal water flow for the reactor coolant pump shaft seals, d) processes reactor coolant letdown for reuse of boric acid and reactor makeup water, e) maintains the proper concentration of corrosion inhibiting chemicals in the reactor coolant and f) maintains the reactor coolant and corrosion activitier to within design levels. The system is also used to fill and hydrostatically test the Reactor Coolant System. During normal operation, this system also has provisions for supplying: 1) Hydrogen to the volume control tank 11) Nitrogen as required for purging the volume control tank lii) Hydrazine and lithium hydroxide, as required, via the chemical mixing A l f tank to the charging pumps suction. V 9.2.1 DESIGN BASES Redundancy of Reactivity Control Criterion: Two independent reactivity control systems, preferably of different principles, shall be provided. (GDC 27) In addition to the reactivity contro; achieved by the rod cluster control (RCC) as detailed in Section 7, reactivity control is provided by the Chemical and Volume Control System which regulates the concentration of boric acid solution neutron absorber in the Reactor Coolant System. The systela is designed to orevent uncontrolled or inadvertent reactivity changes which might cause system parameters to exceed design limits. Cm) d 9.2-1

Reactivity llot Shutdown Capability Criterion: The reactivity control system provided shall be capable of making and holding the core suberitical from any hot standby or hot operating condition. (GDC 28) The reactivity control systems provided are capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes. The maximum excess reactivity expected for the core occurs for the cold, clean condition at the beginnir.g of life of the initial control rods and soluble neutron absorber (boron). The full length Rod Cluster Control (RCC) assemblies are divided into two categories comprising control and shutdown groups. The control group, used in combination with chemical shim, provides control of the reactivity changes of the core throughout the life of the core at power conditions. This group of RCC assemblies is used to compensate for short term reactivity changes at power such as those produced due to variations in reactor power requirements or in coolant temperature. The chemical shim control is used to compensate for the more slowly occurring changes in reactivity throughout core life such as those due to fuel depletion and fission product buildup and decay. Reactivity Shutdown Capability Criterion: One of the reactivity control systems provided shall be capable of making the core subcritical under any anticipated operating condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margin should assure subcriticality with the most reactive control rod fully withdrawn. (GDC 29) The reactor core, together with the reactor control and protection system is designed so that the minimum allowable DNBR is no less than 1.30 and there is no fuel melting during normal operation including anticipated transients. O 9.2-2

The shutdown groups are provided to supplement the control group of RCC assemblies to make the reactor at Icast one per cent subcritical (k = 0.99) following gg C trip from any credible operating condition to the hot, zero power condition assuming the most reactive RCC assembly remains in the fully withdrawn position. Sufficient shutdown capability is also provided to maintain the core subcritical for the most severe anticipated cooldown transient associated with a single active failure, e.g., accidental opening of a steam bypass or relief valve. This is achieved with combination of control rods and automatic boron addition via the Safety Injection System with the most reactive rod assumed to be fully withdrawn. Reactivity Hold-Down Capability, Criterion: The reactivity control systems provided shall be capable of making the core suberitical under credible accident conditions with appropriate margins for contingencies and limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public. (GDG 30) ( )I Normal reactivity shutdown capability is provided by control rods, with w. boric acid injection used to compensate for the long term xenon decay transient and for plant cooldown. Any time that the plant is at power, the quantity of boric acid retained in the boric acid tanks and ready for injection will always exceed that quantity required for the normal cold shutdown. This quantity will always exceed the quantity of boric acid required to bring the reactor to hot shutdown and to compensate for subsequent xenon decay. The boric acid solution is transtarred from the boric acid tanks by boric acid pumps to the suction of the charging pumps which inject boric acid into the reactor coolant. Any charging pump and any boric acid transfer 9.2-3

pump can be operated from diesel generator power on loss of off-site AC power. Boric acid can be injected by one charging pump and one boric acid transfer pump at a rate which shuts the reactor down with no rods inserted in less than sixteen minutes. In sixteen additional minutes, enough boric acid can be injected to compensate for xenon decay although xenon decay below the equilibrium operating level will not begin until approximately 12-15 hours after shutdown. If two boric acid pumps and two charging pumps are available, these time periods are halved. Additional boric acid is employed if it is desired to bring the reactor to cold shutdown conditions. On the basis of the above, the injection of boric acid is shown to afford backup reactivity shutdown capability, independent of control rod clusters which normally serve this function in the short term situation. Shutdown for long term and reduced temperature conditions can be accomplished with boric acid injection using redundant compenents Codes and Classifications All pressure retaining components (or compartments of components) which are exposed to reactor coolant, comply with the following codes: a) System pressure vessels - ASME Boiler and Pressure Vessel Code, Section III, Class C, including para. N-2113. b) System valves, fittings and piping - USAS B31.1, including nuclear code cases. System component code requirements are tabulated in Table 9.2-1. The tube and shell sides on the regenerative heat exchanger and the tube side of the excess letdown heat exchanger are designed as ASME III, Class C. This designation is based on the following considerations: (a) each exchanger is connected to the primary coolant system by lines equal to or less than 3", and (b) each is located inside the reactor containment. Analyses show that the accident associated with a 3" line break does not result in clad damage or failure. Additionally, previously contaminated primary coolant, escaping from the primary coolant system during such accident is confined to the reactor containment building and no public hazard results. 9.2-4

/N 9.2.2 SYSTEM DESIGN AND OPERATION O The Chemical and Volume Control System, shown in Figures 9.2-1 through 9.2-3, provides a means for injection of control poison in the form of boric acid solution, chemical additions for corrosion control, and reactor coolant cleanup and degasification. This system also adds makeup water to the Reactor Coolant System, reprocesses water letdown from the Reactor Coolant System, and provides seal water injection l to the reactor coolant pump seals. Materials in contact with the reactor coolant are austenitic stainless steel or equivalent corrowion resistant materials. System components whose design pressure and temperature are less than the Reactor Coolant System design limits are provided with overpressure protective devices. System discharges from overpressure protective devices (safety valves) Og and system leakages are directed to closed systems. Effluents removed from such closed systems are monitored and discharged under controlled conditions. During plant operation, reactor coolant flows through the letdown line from a loop cold leg on the discharge side of the pump and, after processing is returned to the cold leg of another loop on the discharge side of the pump via a charging line. An alternate charging connection is provided on a loop hot leg. An excess letdown line is also provided for removing coolant from the reactor coolant system. 1 9.2-5

Each of the connections to the Reactor Coolant System has an isolation valve located clo.se to the loop piping. In addition, a check valve is located downstream of each charging line isolation valve. Reactor coolant entering the Chemical and Volume Control System flows through the shell side of the regenerative heat exchanger where its temperature is reduced. The coolant then flows through a letdown orifice which reduces the coolant pressure. The cooled, low pressure water leaves the reactor containment and enters the auxiliary building where it undergoes a second temperature reduction in the tube side of the non-regenerative heat exchanger followed by a second pressure reduction by the low pressure letdown valve. After passing through one of the mixed bed demineralizers, where ionic impurities are removed, coolant flows through the reactor coolant filters and enters the volume control tank through a spray nozzle. Ilydrogen is automatically supplied, as determined by pressure control, to the vapor space in the volume control tank, which is predominantly hydrogen and water vapor. The hydrogen within this tank is supplied to the reactor coolant for maintaining a low oxygen concentration. Fission gases are periodically removed from the system by venting the volume control tank to the Waste Disposal System. From the volume control tank the coolant flows to the charging pumps which raise the pressure above that in the Reactor Coolant System. The coolant then enters the containment, passes through the tube side of the regenerative heat exchanger, and is returned to the Reactor Coolant System. The cation bed demineralizer, located downstream of the mixed bed demineralizers, is used intermittently to control cesium activity in the coolant and also to remove excess lithium which is formed from B10 (n, a) Li reaction. 9 9.2-6

Boric acid is dissolved in hot water in the batching tank to a concentration of approximately 12 percent by weight. The lower portion of the batching tank is jacketed to permit heating of the batching tank solution with low pressure steam. A transfer pump is used to transfer the batch to the boric acid tanks. Small quantities of boric acid solution are metered from the dincharge of an operating transfer pump for blending with primary water as makeup for normal leakage or for increasing the reactor coolant boron concentration during normal operation. Electric immersion heaters maintain the temperature of the boric acid tanks solution high enough to prevent precipitation. Excess liquid effluents containing boric acid flow from the Reactor Coolant System through the letdown line and are collected in the holdup tanks. As liquid enters the holdup tanks, the nitrogen cover gas is displaced to the gas decay tanks in the Waste Disposal System through the waste vent header. The concentration of boric acid in the holdup tanks varies throughout core life from the refueling concentration to essentially zero at the end of the core cycle. A recirculation pump is provided to transfer liquid from O) one holdup tank to another. Liquid effluent in the holdup tanks is processed as a batch operation. This liquid is pumped through the evaporator feed ion exchangers which primarily remove lithium hydroxide an fission-products such as long-lived cesium. It then flows through the ion exchanger filter and into the gas stripper where dissolved gases are removed from the liquid. The gases are vented to the Waste Disposal System. The liquid effluent from the gas stripper I enters the boric acid evaporator. i I l AU i 9.2-7 1

The vapcr produced in the boric acid evaporator leaves the evaporator condenser and is pumped through a condensate cooler where the distillate is cooled to the operating temperature of the eva9 orator condensate demineralizers. After non-volatile evaporator carry over is removed by one of the two evaptc. tor condensate demineralizers, the condensate flows through the condensate filter and accumulates in one of two monitor tanks. The dilute boric acid solution originally in the boric acid evaporator remains as the bottoms of the distillation process and is concentrated to approximately twelve per cent boric acid. Subsequent handling of the condensate is dependent on the results of sample analysis. Discharge from the monitor tanks may be pumped to the primary water storage tank, recycled through the evaporator condensate demineralizers, returned to the holdup tanks for reprocessing in the evaporator train or discharged to the environment with the condenser circulating water when within the allowable activity concentration as discussed in Section 11. If the sample analysis of the monitor tank contents indicates that it may be discharged safely to the environment, two valves must be opened to provide a discharge path. As the effluent leaves, it is continuously monitored by the waste dispcsal system liquid effluent monitor. If an unexpected increase in radioactivity is sensed, one of the valves in the discharge line to the condenser circulating water closes automatically and an alarm sounds in the control room. Boric acid evaporator bottoms are discharged through a concentrates filter to the concentrates holding tank. Solution collected in the concentrates holding tank is sampled and then transferred to the boric acid tanks if analysis indicates that it meets specifications for use as boric acid makeup. Otherwise the solution is pumped to the holdup tanks for reprocessing by the evaporator train. The concentrated solution can also be pumped from the evaporator to the Waste Disposal System and finally placed in containers and mixed with cement. These containers can then be stored at the plant site for ultimate shipment off-site for disposal. O 9.2-8

O The deborating demineralizers can be used intermittently to remove boron from the reactor coolant near the end of the core life. When the deborating demineralizers are in operation, the letdown stream passes from the mixed bed demineralizers and then through the deborating demineralizers and into the volume control tank after passing through the reactor coolant filter. During plant cooldown when the residual heat removal loop is operating and the letdown orifices are not in service, a flow path is provided to remove corrosion impurities and fission products. A portion of the flow leaving the residual heat exchangers passes through the non-regenerative heat exchanger, mixed bed demineralizers, reactor coolant filter and volume control tank. The fluid is then pitmped, via the charging pump, through the tube side of the regenerative heat exchanger into the Reactor Coolant System. O Expected Operating Conditions Tables 9.2-2, 9. 2-3, and 9.2-5 list the system performance requirements, data for individual system components and reactor coolant equilibrium activity concentration. Table 9.2-4 supplements Table 9.2-5. Reactor Coolant Activity Concentration The parameters used in the calculation of the re' actor coolant fission product inventory, including pertinent information concerning the expected coolant cleanup flow rate and demineralizer effectiveness, are presented in Table 9.2-4. The results of the calculations are presented in Table 9.2-5. In these calculations one percent defects in the fuel rods are assumed to be present at initial core loading and are uniformly distributed throughout the core and the fission product escape rate coefficients are therefore based upon an average fuel temperature. O 9.2-9

The fission product activity in the reactor coolant during operation with small cladding pinholes or cracks in 1% of the fuel rods is computed using the following differential equations: For parent nuclides in the coolant, dN U d i C i i B - tB' wi for daughter nuclides in the coolant, dN" dt j C (A + Rn + ) N + AN Dv N = j j B - tB' wj iw f where: N = population of nuclide D = fraction of fuel rods having defective cladding R = purification flow, coolant system volumes per sec. B = initial boron concentration, ppm g B' = boron concentration reduction rate by feed and bleed, ppm per sec n = removal efficiency of purification cycle for nuclide A = radioactive decay constant v = escape rate coefficient for diffusion into coolant Subscript C refers to core Subscript w refers to coolant Subscript i refers to parent nuclide Subscript j refers to daughter nuclide Tritium is produced in the reactor from ternary fission in the fuel, irradiation of boron in the burnable poison rods (during initial fuel cycle only) and irradiation of boron, lithium and deuterium in the coolant. The deuterium contribution is less than 0.1 curie per year and may be neglected. The parameters used in the calculation of tritium production rate are presented in Table 9.2-6. O 9.2-10

r React.or Makeup Control Makeup for normal plant leakage 19 regulated by the reactor makeup control which is set by the opcrator to blend water from the primary water storage tank with concentrated boric acid to match the reactor coolant boron concentration. The makeup system also provides concentrated boric acid or primary water to l decrease the boric acid concentration in the Reactor Coolant System. To maintain the reactor coolant volume constant, an equal amount of reactor coolant at existing reactot coolant boric acid concentration is letdown to the holdup tanks. Should the letdown line be out of service during operation, suf ficient volume exists in the pressurizer to accept the amount of boric acid necessary for cold shutdown. Makeup water to the Reactor Coolant System is provided by the Chemical and Volume Control System f rom the following sources: a) The primary water storage tank, which provides water for dilution when the reactor coolant boron concentration is to be reduced OO b) The boric acid tanks, which supply concentrated boric acid solution when reactor coolant boron concentration is to be increased c) The refueling water storage tank, which supplies borated water for emergency makeup d) The chemical mixing tank, which is used to inject small quantities of solution when additions of hydrazine or pH control chemical are necessary. The reactor makeup control is operated from the control room by manually pre-selecting makeup composition to the charging pump suction header or the volume control tank in order to adjust the reactor coolant boron con-centration for reactivity control. Makeup is provided to maintain the desired operating fluid inventory in the Reactor Coolant System. The operator can stop the makeup operation at any time in any operating mode by l remotely closing the makeup stop valves. One primary water makeup pump and one boric acid transfer pump are normally operated. 9.2-11 Amendment 1 2

A portion of the high pressure charging flow is injected into the reactor coolant pumps between the pump impeller and the shaft seal so that the seals are not exposed to high temperature reactor coolant. Part of the I flow is the shaft seal leakage flow and the remainder enters the Reactor Coolant System through a labyrinth seal on the pump shaft. The shaft seal leakage flow cools the lower radial bearing, passes through the seals, is cooled in the seal water heat exchanger, filtered, and returned to the volume control tank. Seal water inleakage to the Reactor Coolant System requires a continuous letdown of reactor coolant to maintain the desired inventory. In addition, bleed and feed of reactor coolant are required for removal of impurities and adjustment of boric acid in the reactor coolant. Automatic Maxeup The " automatic makeup" mode of operation of the reactor makeup control provides boric acid solution preset by the operator to match the boron concentration in the Reactor Coolant System. The automatic makeup compensates for minor leakage of reactor coolant without causing significant changes in the h coolant boron concentration. Und,r normal plant operating conditions, the mode selector switch and makeup stop valves are set in the " Automatic Makeup" position. A preset low level signal from the volume control tank level controller causes the automatic makeup control action to open the makeup stop valve to the charging pump suction, open the concentrated boric acid control valve and the primary water makeup control valve. The flow controllers then blend the makeup stream according to the preset concentration. Makeup addition to the charging pump suction header causes the water level in the volume control tank to rise. At a preset high level point, the makeup is stopped; the prima ry water makeup control salve closes, the concentrated boric acid control valve closes and the makeup stop valve to charging pump suction closes. O 9.2-12 6,,,

Dilution The " dilute" mode of operation permits the addition of a pre-selected quantity v of primary water makeup at a pre-selected flow rate to the Reactor Coolant System. The operator sets the makeup stop valves to the volume control tank and to the charging pump suction in the closed position, the mode selector switch to " dilute", the primary water makeup flow controller set point to the desired flow rate, and the primary water makeup batch integrator to the desired quantity. If the dilution flow deviates 15 gpm from the preset flow rate, an alarm indicates the deviation. Opening the makeup stop valve to the volume control tank starts a primary water makeup pump. Makeup water is added to the volume control tank and then goes to the charging pump suction header. Excessive rise of the volume control tank water level is prevented by automatic actuation (by the tank level controller) of a three-way diversion valve, which routes the reactor coolant letdown flow to the holdup tanks. When the. preset quantity of primary water makeup has been added, the batch integrator causes the reactor makeup water pump to stop and the primary water makeup control valve to close. Boration The " borate" mode of operation permits the addition of a pre-selected quantity of concentrated boric acid solution at a pre-selected flow rate to the Reactor Coolant System. The operator sets the makeup stop valves to the volume control tank and to the charging pump suction in the closed position, the mode selector switch to " borate", the concentrated boric acid flow controller set point to the desired flow rate, and the concentrated boric acid batch l integrater to the desired quantity. If the boration flow deviates 0.5 gpm from the preset flow rate, an alarm indicates the deviation. Opening the makeup stop valve to the charging pumps suctions starts the selected boric l acid transfer pump, and permits the concentrated boric acid to be added to the charging pump suction header. The total quantity added in mos t cases is so small that it has only a minor effect on the volume control tank level. When the preset quantity of concentrated boric acid solution has been added, the batch integrater causes the concentrated boric acid transfer pump to stop and the concentrated boric acid control valve to close. O 9.2-13 l c

o O The capability to add boron to the reactor coolant is sufficient so that no limitation is imposed on the rate of cooldown of the reactor upon shut-down. The maximum rates of boration and the equivalent coolant cooldown rates are given in Table 9.2-2. One set of values is given for the addition of boric acid from a boric acid tank with one transfer and one charging pump operating. The other set assumed the use of refueling water but with two of the three charging pumps operating. The rates are based on full operating temperature and on the end of the core life when the moderator temperature coefficient is most negative. Alarm Functions The reactor makeup control is provided with alarm functions to call the operator's attention to the following conditions: a) Deviation of primary water makeup flow rate from the control set point b) Deviation of concentrated boric acid flow rate from the control set point c) Low level (makeup ini' lation point) in the volume control tank wl. a reactor makeup control selector is not set for the automstic makeup control mode. Charging Pump Control Three positive displacement variable speed drive charging pumps are used to supply charging flow to the Reactor Coolant System. O 9.2-14

The speed of each pump can be controlled manually or automatically. During normal operation, only one of the three pumps is automatically controlled. () During normal operation, only one charging pump is operating and the speed is modulated in accordance with pressurizer level. During load changes the pressurizer level set point is varied automatically to compensate partially for the expansion or contraction of the reactor coolant associated with the T,y changes. T,yg conpensates for power changes by varying the pressurizer level set points in conjunction with pressurizer level for charging pump control. The level set points are varied between 20 and 60 percent of the adjustable range depending on the power level. Charging pump speed does not change rapidly with pressurizer level variations due to the reset action of the pressurizer level controller. If the pressurizer level increases, the speed of the pump decreases, likewise if the level decreases, the speed increases. If the charging pump on automatic control reaches the high speed limit, an alarm is actuated and a seecond charging pump is manually started. The speed of the second pump is manually regulated. If the speed of the charging pump on automatic control does not decrease and the second charging pump is operating at maximum speed, the O(G third charging pump can be started and its speed manually regulated. If i the speed of the charging pump on automatic control decreases to its minimum value, an alarm is actuated and the speed of the pumps on manual control is reduced. Components A summary of principal component data is given in Table 9.2-3. Regenerative Heat Exchanger The regenerative heat exchanger is designed to recover the heat from the letdown stream by reheating the charging stream during normal operation. This exchanger also limits the temperature rise which occurs at the letdown orifices during periods when letdown flow exceeds charging flow by a greater margin than at normal letdown conditions. 9.2-15

The letdown stream flows through the shell of the regenerative heat exchanger and the charging stream flows through the tubes. The unit is made of austenitic stainless steel, and is of all-welded construction. The exchanger is designed to withstand 2000 step changes in shell side fluid temperature f rom 130*F to 552.2 during the design life of the unit. Letdown Orifices One of the three letdown orifices controls flow of the letdown stream during normal operation and reduces the pressure to a value compatible with the non-regenerative heat exchanger design. Two of the letdown orifices are each designed to pass normal letdown flow. These orifices are used in parallel to pass maximum purification flow at normal Reactor Coolant System operating pressure. The remaining orifice is designed to pass three-fourths of the normal letdown flow. The orfices are placed in and taken out of service by remote manual operation of their respective isolation valves. One or both of the standby orifices may be used in parallel with the normally operating orifice in order to increase letdown flow when the Reactor Coolant System pressure is below normal. This arrangement provides a full standby capacity for control of letdown flow. Each orifice consists of bored pipe made of austenitic stainless steel. Non-Regenerative (letdown) Heat Exchanger The non-regenerative heat exchanger cools the letdown stream to the operating temperature of the mixed bed demineralizers. Reactor coolant flows through the tube side of the exchanger while component cooling water flows through the shell. The letdown stream outlet temperature is automatically controlled by a temperature control valve in the component cooling water outlet stream. The unit is a multiple-tube-pass heat exchanger. All surfaces in contact with the reactor coolcr.t are austenitic stainless steel, and the shell is carbon steel. O 9.2-16

a Mixed Bed Demineralizers Two flushable mixed bed demineralizers maintain reactor coolant purity. A lithium-7 cation resin and a hydroxyl form anion resin are initially charged into the demineralizers. Both forms of resin remove fission and corrosion products, and in addition, the reactor coolant causes the anion resin to be converted to the borate form. The resin bed is designed to reduce the concentration of ionic isotopes in the parification stream, except for cesium, yttrium, and molybdenum, by a minimum factor of 10. Each demineralizer is sized to accommodate the maximum letdown flow. One demineralizer serves as a standby unit for use should the operating demineralizer become exhausted during operation. The demineralizer vessels are made of austenitic stainless steel, and are provided with suitable connections to facilitate resin replacement when required. The vessels are equipped with a resin retention screen. Each g demineralizer has suf ficient capacity af ter operation for one core cycle with one per cent defective fuel rods to reduce the activity of the primary coolant to refueling concentration. l Cation Bed Demineralizer A flushable cation resin bed in the hydrogen form is located downstream of the mixed bed demineralizers and is used intermittently to control the 10 concentration of lithium-7 which builds up in the coolant from the B (n, a) Li reaction. The demineralizer also has suf ficient capacity to naintain the cesium-137 concentration in the coolant below 1.0 uc/cc with one percent defective fuel. The demineralizer would be used intermittently to control cesium. The demineralizer is made of austenitic stainless steel and is provided with suitable connections to facilitate resin replacement when required. The vessel is equipped with a resin retention screen. 9.2-17

Resin Fill Tank The resin fill tank is used to charge fresh resin to the demineralizers. The line f rom the conical bottom of the tank is fitted with a dump valve and may be connected to any one of the demineraliter fill lines. The demineralizer water and resin slurry can be sluiced into the demineralizer by opening the dump valve. The tank, designed to hold approximately one-third the resin volume of one mixed bed demineralizer, is nade of austenitic stainless steel. Reactor Coolant Filter The filter collects resin fines and particulates larger than 25 microns from the letdown stream. The vessel is made of austenitic stainless steel, and is provided with connections for draining and venting. Design flow capacity of the filter is equal to the maximum purification flow rate. Dis-posable synthetic filter elements are used. Bases being considered to determine when the reactor coolant filter will be replaced are: (1) a high pressure differential across the filter, (2) a set time limit after which the filter will be replaced, and (3) when a portable radiation monitor shows radiation in excess of established limits. Volume Control Tank The volume control tank collects the excess water released from zero power to full power, that is not accommodated by the pressurizer. It also receives the excess coolant release caused by the deadband in the reactor control temperature instrumentation. Overpressure of hydrogen gas is maintained in the volume control tank to control the hydrogen concentration in the reactor coolant at 25 to 35 cc per kg of water (standard conditions). A spray nozzle is located inside the tank on the inlet line from the reactor coolant filter. This spray nozzle provides intimate contact to equilibrate the gas and liquid phases. A remotely operated vent valve discharging to the Waste Disposal System permits removal of gaseous fission products which are stripped from the reactor coolant and collected in this tank. The volume control tank also acts as a head tank for the charging pumps and a reservoir for the leakage from the reactor coolant pump controlled leakage seal. The tank is constructed of austenitic stainless steel. 9.2-18

Charging Pumps 1 ( ( j Three charging pumps inject coolant into the Reactor Coolant System. The pumps are the variable speed positive displacement type, and all parts in contact with the reactor coolant are fabricated of austenitic stainless steel and other material of adequate corrosion resistance. These pumps have mechanical packing followed by a leakoff to collect reactor coolant before it can leak to the outside atmosphere. Pump leakage is piped to the drain header for disposal. The pump design precludes the possibility of lubricating oil contaminating the charging flow, and the integral discharge valves act as check valves. Each pump is designed to provide the full charging flow and the reactor coolant pump seal water supply during normal seal leakage. Each pump is designed to provide rated flow against a pressure equal to the sum of the Reactor Coolant System maximum pressure (existing when the pressurizer power operated relief valve is operating) and the piping, valve and equipment pressure losses of the charging system at the design charging flows. r( )\\ One of the three charging pumps can be used to hydrotest the Reactor Coolant System. The pumps are normally energized manually from the control room, and flow is automatically controlled by pressurizer level. Charging Pump Accumulators i A charging pump accumulator is attached to each charging pump outlet line l to substantially reduce the outlet pressure pulses and' reduce piping vibration. l Chemical Mixing Tank The primary use of the chemical mixing tank is in the preparation of caustic i solutions for pH control and hydrazine for oxygen scavenging. O 9.2-19

The capacity of the chemical mixing tank is determined by the quantity of 35 per cent hydrazine solution necessary to increase the concentration in the reactor coolant by 10 ppm. This capacity is more than sufficient to prepare the solution of pH control chemical for the Reactor Coolant System. The chemical mixing tank is made of austenitic stainless steel. Excess Letdown Heat Exchanger The excess letdown heat exchanger cools reactor coolant letdown flow until the flow rate is equal to the nominal injection rate through the reactor coolant pump labyrinth seal, if letdown through the normal letdown path is blocked. The unit is designed to reduce the letdown stream temperature from the cold leg temperature to 195*F. The letdown stream flows through the tube side and component cooling water is circulated through the shell side. All surfaces in contact with reactor coolant are austenitic stainless steel and the shell is carbon steel. All tube joints are welded. The unit is designed to withstand 12,000 step changes in the tube fluid temperature from 80*F to the cold leg temperature. Seal Water Heat Exchanger The seal water heat exchanger removes heat from the reactor coolant pump seal water returning to the volume control tank and reactor coolant discharge from the excess letdown heat exchanger. Reactor coolant flows through the tubes and component cooling water is circulated through the shell side. The tubes are welded to the tube sheet to prevent leakage in either direction, resulting in undesirable contamination of the reactor coolant or component cooling water. All surfaces in contact with reactor coolant are austenitic stainless steel and the shell is carbon steel. O 9.2-20

r The unit is designed to cool the excess letdown flow and the seal water flow to the temperature normally maintained in the volume control. tank if all the reactor coolant pump seals are leaking at the maximum design leakage rate. Seal Water Filter The filter collects particulates from the reactor coolant pump seal water return and from the excess letdown heat exchanger flow. The filter is designed to pass the sum of the excess letdown flow and the maximum design leakage f rom the reactor coolant pump seals. The vessel is constructed of austenitic stainless steel and is provided with connections for draining and venting. Disposabic synthetic filter cartridges are used. Seal Water Injection Filters Two filters are provided in parallel, each sized for the injection flow. They collect particulates from the water supplied to the reactor coolant pump seal. T ) Boric Acid Filter The horic acid filter collects particulates from the boric acid solution being pumped to the charging pump suction line. The filter is designed to pass the design flow of two boric acid pumps operating simultaneously. The vessel is constructed of austenitic stainless steel and the filter elements are disposable synthetic cartridges. Provisions are available for venting and draining the filter. G 9.2-21 s

Boric Acid Tanks O The boric acid tank capacities are sized to store sufficient boric acid solution for refueling plus enough boric acid solution for a cold shutdown shortly af ter initial full power operation is achieved. In addition, each tank has sufficient boric acid solution to achieve cold shutdown if the most reactive RCC is not inserted. One tank supplies boric acid for the boron injection tank and reactor coolant makeup while recycled solutions f rom the concentrates holding tank is accumulated in the other tank. The concentration of boric acid solution in storage is maintained between l 11.5 and 13% by weigh t. Periodic manual sampling is performed and corrective action is taken, if necessary, to ensure that these limits are maintained. As a consequence, measured amounts of boric acid solution can be delivered to the reactor coolant to control the chemical poison concentration. The combination overflow and breather vent connection has a water loop seal to minimize vapor discharge during storage of the solution. The tanks are constructed of austenitic stainless steel. O Boric Acid Tank lleaters Two 100% capacity electric immersion heaters located near the bottom of each boric acid tank are designed to maintain the temperature of the boric acid solution at 165*F with an ambient air temperature of 40 F thus ensuring a temperature in excess of the solubility limit (for 20,000 ppm boron this is 130*F). The temperature is monitored and low temperature is alarmed in the control room. The heaters are sheathed in austenitic stainless steel. Batching Tank The batching tank is sized to hold one week's makeup supply of boric acid solution for the boric acid tank. The basis for makeup is reactor coolant leakage of 1/2 gpm at beginning of core life. The tank may also be used for solution storage. A local sampling point is provided for verifying the solution concentration prior to transferring it to the boric acid tank or for draining the tank. Amendment 1 9.2-22

The tank manway is provided with a removable screen to prevent entry of foreign particles. In addition, the tank is provided with an agitator to improve mixing during batching ciserations. The tank is constructed of austenitic stainless steel, and is not used to handle radioactive substances. The tank is provided with a steam jacket for heating the boric acid solution to 165*F. Boric Acid Transfer Pumps Two 100% capacity canned centrifugal pumps are used to circulate or transfer chemical solutions. The pumps circulate boric acid solution through the boric acid tanks and inject boric acid into the charging pump suction header and boron injection tank in the Safety Injection System. Although one pump is normally used for boric acid batching and transfer and the other for boric acid injection, either pump may function as standby for the other. The design capacity of each pump is equal to the normal letdown flow rate. The design head is sufficient, considering line and valve losses, to deliver rated flow to the charging pump suction header Q(.\\ when volume control tank pressure is at the maximum operating value (relief valve setting), All parts in contact with the solutions are austenitic stainless steel and other adequately corrosion-resistant material. The transfer pumps are operated either automatically or manually from the main control room or from a local control peint. The reactor makeup control operates one of the pumps automatically when boric acid solution is required for makeup or boration. i Boric Acid Blender l l The boric acid blender promotes thorough mixing of boric acid solution and primary water from the primary water supply circuit. The blender consists of a conventional pipe fitted with a perforated tube insert. All material is austenitic stainless steel. The blender decreases the pipe length required to homogenize the mixture for taking a representative local sample. l O 9.2-23

Recycle Process Holdup Tanks ihree holdup tanks contain radioactive liquid which enters the tank from the letdown line. The liquid is released from the Reactor Coolant System during startup, shutdowns, load cha.iges and from boron dilution to compensate for burnup. The contents of one tant are normally being processed by the gas stripper and evaporator train while another tank is being filled. The third tank is normally kept empty to provide additional storage capacity when needed. Each liquid storage tank size is based on 2/3 of the primary system volume, and the holdup tank capacity is given in Table 9.2-3. The tanks are constructed of austenitic stainless steel. Holdup Tank Recirculation Pump The re "rculation pump is used to mix the contents of a holdup tank or transfer the contents of one holdup tank to another holdup tank. The wetted surface of this pump is constructed of austenitic stainless steel. Gas Stripper Feed Pumps The two gas stripper feed pumps supply feed to the gas stripper boric acid evaporator trains from a holdup tank. The non-operating pump is a standby and is available for operation in the event the operating pump malfunctions. These canned centrifugal pumps are constructed of austenitic stainless steel. Base and Cation Ion Exchangers Three flushable base and cation ion exchangers remove anions and cations (primarily cesium and molybdenum) from the holdup tank effluent. The resin is initially in the hydrogen form. Experiments performed by Westinghouse indicate that the decontamination factor for cesium (see table 9.2-4) is conservative. The demineralizer vessel is constructed of austenitic stainless steel and contains a resin retention screen. l L 9.2-24

Ion Exchanger Filters /S V These filters collect resin fines and particulates from the cation ion exchanger. The vessel is made of austenitic stainless steel and is provided with connections for draining and venting. Disposable synthetic filter cartridges are used. The design flow capacity is equal to the boric acid evaportor flow rate. Gas Stripper Equipment Two gas strippers are provided. Each removes nitrogen, hydrogen, and fission gases from the holdup tank effluent. The gas stripper consists of a preheater, stripping column with a reflux condenser and associated pumps, piping, and instrumentation. The gas stripper preheater, located upstream of the gas stripper, heats the liquid effluent f rom the holdup tanks from ambient temperature to approximately 205'F using the gas stripper bottoms. The bottoms are cooled in the ) preheater from approximately 220*F to 120*F. The preheater is a regenerative / type shell and tube unit constructed of austenitic stainless steel. The gas strippers consist of a hot well with heating coil to store stripped water, a stripping section packed with pall rings, a spray type liquid inlet header and an overhead integral reflux condenser. Liquid flowing to the gas strippers is controlled to constant rate by a flow controller. The gas strippers are designed for the same flow rate as the evaporator and are 5 designed to reduce the influent gas concentration by a factor of 10, Two gas stripper bottom pumps per gas stripper, operated from level control, transfer effluent f rom the gas stripper hot wells to the boric acid evaporator via the gas stripper preheaters. Each centrifugal pump is rated at the evaporator processing rate. The pumps are austenitic stainless steel and one is an installed standby for the operating pump. OO 9.2-25

I Boric Acid Evaporator Equipment O Two boric acid evaporators concentrate boric acid for reuse in the Reactor Coolant System. Borated water enters the evaporator and the liquid is concentrated to approximately 12 weight per cent boric acid. Vapors leave the evaporator and are condensed. The solide decc ttamination factor between the condensate and the bottoms is approximately 10. All evaporator equipment is constructed of austenitic stainless steel and is supplied as a unit. The boric acid evaporator equipment consists of the boric acid evaporator feed tank, two boric acid evaporator concentrates pumps, boric acid evaporator, boric acid evaporator condenser, two boric acid evaporator condensate pumps, boric acid evaporator condensate cooler, vacuum pumps and associated piping and instrumentation. The boric acid evaporator feed tank has sufficient capacity to hold one day's production of 12 per cent boric acid solution produced from refueling concen-tration feed. The evaporator and condenser heat transfer area is suf ficient to maintain the required feed rate. The evaporator is steam heated. Component cooling water flows through the tube of the condenser. The boric acid distillate cooler reduces the te.nperature of the condensate to approximately 100"F. The condensate flows through the tubes and component cooling water through the shell. Evaporator Condensate Demineralizers Two anion demineralizers remove any boric acid contained in the evaporator condensate. Hydroxyl based ion-exchange resin is used to produce evaporator condensate of high purity. Facilities are provided for regeneration of the resin. When regeneration is no longer feasible, the resin is flushed to the spent resin storage tank. The resin volume in each demineralizer is selected to keep resin regenerations to an average of once per month during a core cycle with a flow rate equal to the evaporator flow rate. 9 9.2-26

Condensate Fi l te rs AV Two filters collect resin fines and particulates from the boric acid evaporator condensate streams. Each vessel is made of austenitic stain-less steel, and is provided with a connection for draining and venting. Disposable synthetic filter elements are used. The design flow capacity of each filter is equal to the boric acid evaporator flow rate. Monitor Tanks Two monitor tanks permit continuous operation of the evaporator trains. When one tank is filled, the contents are analyzed and either reprocessed, discharged to the Waste Disposal System, or pumped to the primary water ' storage tank. These tanks contain a diaphragm membrane and are stainless steel. Monitor Tank Pumps (3 (s,) Two monitor tank pumps discharge water 'from the monitor tanks. The pumps are sized to empty a monitor tank in 2.0 hours. The pumps are constructed of austenitic stainless steel. Primary Water Storage Tanks The primary water storage tank is used to store makeup water which is supplied from the monitor tanks and the water treatment plant. Makeup water from the tank discharges to the suction of the primary water makeup pumps. The tank contains a diaphragm membrane and is made of stainless steel. 1 Primary Water Makeup Pumps Two primary water makeup pumps take suction from either the monitor tanks or the primary water storage tank. These pumps are used to feed dilution water to the boric acid blender and are also used to supply makeup water for intermittent flushing of equipment and piping. 9.2-27 Amendment 1

r Each pump is sized to match the maximum letdown flow. One pump serves as a standby for the other. These centrifugal pumps are constructed of austenitic stainless steel. Concentrates Filter A disposable synthetic cartridge type filter removes particulates f rom the evaporator concentrates. Design flow capacity of the filter is equal to the boric acid evaporator concentrates transfer pump capacity. The vessel is made of austenitic stainless steel. Concentrates llolding Tank The concentrates holding tank is sized to hold the production of concentrates f rom one batch of evaporator operat ion. The tank is supplied with an electrical heater which prevents boric acid precipitation and is constructed of austenitic stainless steel. Co_ncentrates lloiding Tank Transfer Pumps Two holding tank transfer pumps discharge boric acid solution from the concentrates holding tank to the boric acid tanks. The canned centrifugal pumps are sized to empty the concentrates holding tank in 20 ninutes. The wetted surfaces are constructed of austenitic stainless steel and other adequately corrosion-resistant material. Deborating Demineralizers When required, two anion demineralizers remove boric acid from the Reactor Coolant System fluid. The demineralizers are provided for use near the end of a core cycle, but can be used at any time, liydroxyl based ion-exchange resin is used to reduce Reactor Coolant System boron concentration by releasing a hydroxyl ion when a borate ion is absorbed. Facilities are provided for regeneration. When regeneration is no longer feasible, the resin is flushed to the spent resin storage tank. 4 9.2-28

Each demineralizer is sized to remove the quantity of boric acid that must be removed from the Reactor Coolant System to maintain full power i(,j\\ operation near the end of core life without the use of the holdup tanks or evaporators. Electrical lleat Tracing Electrical heat tracing is installed under the insulation on all piping, valves, line-mounted instrumentation, and components normally containing concentrated boric acid solution. The heat tracing is designed to prevent boric acid precipitation due to cooling, by compensating for heat loss. Exceptions are: a) Lines which may transport concentrated boric acid but are subsequently flushed with reactor coolant or other liquid of low boric acid concentration during normal operation b) The boric acid tanks, which are provided with immersion heaters ) v c) The batching tank, which is provided with a steam jacket d) The concentrates holding tank, which is provided with an immersion heater. Duplicate tracing on sections of the Chemical and Voltme Control System normally containing boric acid solution provides standby capacity if the operating tracing malfunctions. Lines which are provided with heat tracing are shown on Figures 9.2-1 through 9.2-3. ,O b) 9.2-29

Valves Valves that perform a modulating function are equipped with two sets of packing and an intermediate leakoff connection that discharges to the Waste Disposal System. All other valves have stem leakage control. Globe valves are installed with flow over the seats when such an arrangement reduces the possibility of leakage. Basic material of construction is stainless stael for all valves except the batching tank steam jacket valves which are carbon steel. Isolation valves are provided at all connections to the Reactor Coolant System. Lines entering the reactor containment also have check valves Inside the containment to prevent reverse flow from the containment. Relief valves are provided for lines and components that might be pres-surized above design pressure by improper operation or component malfunction. Pressure relief for the tube side of the regenerative heat exchanger is provided by the auxiliary spray line isolation valve which is designed j to open when pressure under the seat exceeds reactor coolant pressure by 250 psi. Relief valves settings and capacities are given in Table 9.2-3. Piping All Chemical and Volume Control System piping handling radioactive liquid is austenitic stainless steel. All piping joints and connections are welded, except where flanged connections are required to facilitate equipment removal for maintenance and hydrostatic testing. Piping, valves, equipment and line-mounted instr amentation, which normally contain concentrated boric solution, are heated by duplicate electrical tracing to ensure solubility of the boric acid. l l @1 l 9.2-30 l 1

9.2.3 SYSTEM DESIGN EVALUATION

O Availability and Reliability V

A high degree of functional reliability is assured in this system by providing standby components where performance is vital to safety and by assuring fail-safe response to the most probable mode of failure. Special provisions include duplicate heat tracing with alarm protection of lines, valves, and components normally containing concentrated boric acid. The system has three high pressure charging pumps, each capable of supplying the normal reactor coolant pump seal and makeup flow. I The electrical equipment of the Chemical and Volume Control System is arranged so that multiple items receive their power from various 480 volt buses (See Figure 8.2-4). Each of the three charging pumps are powered from separate 480 volt buses. The two boric acid transfer pumps are also powered from separate 480 volt buses. One charging pump and one boric acid t ansfer pump l are capable of meeting cold shutdown requirements shortly after full-power operation. In cases of loss of a-c power, a charging pump and a boric acid transfer pump can be placed on the emergency diesels if necessary. Control of Tritium { I The Chemical and Volume Control System is used to control the concentration of tritium in the Reactor Coolant System. Essentially all of the tritium is in chemical combination wi'.h oxygen as a form of water. Therefore, any leakage of coolant to the containment atmosphere carries tritium in the same proportion, as it exists in the coolant. Thus, the level of tritium in the containment atmosphere, when it is sealed from outside air ventilation, is a function of tritium level in the reactor coolant, the cooling water temperature at the cooling coils, which determines the dew point temperature of the air, and the presence of leakage other than reactor coolant as a source of moisture in the containment air. O 9.2-31

There are tuo major considerations with regard to the presence of tritium: a) Possible plant personnel hazard during access to the containment. Leakage of reactor coolant during operation with a closed containment causes an accumulation of tritium in the containment atmosphere. It is desirable to limit the accumulation to allow containment access. b) Possible public hazard due to release of tritium to the plant environment. Neither of these considerations is limiting in this plant. The concentration of tritium in the reactor coolant is maintained at a level which precludes personnel hazard during access to the containment. This is achieved by discharging part of the condensate from the boric acid recovery process to the plant circulating cooling water. Leakage Prevention Quality control of the material and the installation of the Chemical and Volume Control valves and pipings, which are designated for radioactive service, is provided in order to essentially eliminate leakage to the atmosphere. The components designated for radioactive service are provided with welded connections to prevent leakage to the atmosphere. However, flanged connecticas are provided in each charging pumps suction and discharge, on each boric acid pump nuction and discharge, on the relief valves inlet and outlet, on three-way valves and on the flow meters to permit removal for maintenance. The positive displacement charging pumps stuffing boxes are provided with leakoffs to collect reactor coolant before it can leak to the atmosphere. All valves which are larger than 2 inches and which are designated for radioactive service at an operating fluid temperature above 212*F are provided with a stuffing box and lantern leakoff connections. Leakage to the atmosphere is essentially zero for these valves. All control valves are either provided with stuffing box and leakoff connections or are totally enclosed. Leakage to the atmosphere is essentially zero for these valves. 9.2-32

Diaphragm valves are provided where the operating pressure and the operating temperature permit the use of these valves. Leakage to the' atmosphere is essentially zero for these valves. Incident Control The letdown line and the reactor coolant pumps seal water return line penetrate the reactor containment. The letdown line contains air-cperated valves inside the reactor containment and two air-operated valves outside the reactor containment which are automatically closed by the containment isolation signal. The reactor coolant pumps seal water return line contains one motor-operated isolation valve outside the reactor containment which is automatically closed by the containment isolation signal. I The three seal water injection lines to the reactor coolant pumps and the charging line are inflow lines penetrating the reactor containment. Each line contains two check valves inside the reactor containment to provide isolation of the reactor containment when a break occurs in these lines outside the reactor containment. i Malfunction Analysis To evaluate system safety, failures or malfunctions were assumed concurrent with a loss-of-coolant accident and the consequences analyzed and presented in Table 9.2-7. As a result of this evaluation, it is concluded that proper consideration has been given to plant safety in the design of the system. If a rupture were to take place between the reactor coolant loop and the first isolation valve or check valve, this incident would lead to an uncontrol-led loss of reactor coolant. The analysis of loss of coolant accidents is discussed in Section 14. O 9.2-33 a

Should a rupture occur in the Chemical and Volume Control System outside the containment, or at any point beyond the first check valve or remotely operated isolation valve, actuation of the valve would limit the release of coolant and assure continued functioning of the normal means of heat dissipation from the core. For the general case of rupture outside the containment, the largest source of radioactive fluid subject to release is the contents of the volume control tank. The consequences of such a release are considered in Section 14. When the reactor is subcritical; i.e., during cold or hot shutdown, refueling and approach to criticality, the relative reactivity status (neutron source multiplication) is continuously monitored and indicated by BF C "" "#8 8 3 count rate indicators. Any appreciable increase in the neutron source multi-plication, including that caused by the maximum physical boron dilution rate (See Table 9.2-2), is slow enough to give ample time to start a corrective action (boron dilution stop and/or emergency boron injection) to prevent the core from becoming critical. The maximum dilution rate is based on the abnormal condition of two charging pumps operating at full speed delivering unborated primary water to the Reactor Coolant System at a particular time during refueling when the boron concentration is at the maximum value and the water volume in the system is at a minimum. At least two separate and independent flow paths are available for reactor coolant boration; i.e., the charging line, or the reactor coolant pumps labyrinths. The malfunction or failure of one component will not result in the inability to borate the Reactor Coolant System. An alternate flow path is always available for emergency boration of the reactor coolant. As backup to the boration system the operator can align the refueling water storage tank outlet to the suction of the charging pumps. Boration during normal operation to compensate for power changes will be indicated to the operator from two sources; (a) the control rod movement and (b) the flow indicators in the boric acid transfer pump discharge line. When the emergency boration path is used, three indications to the operator 9.2-34

are available. The primary indication is a flow indicator in the emergency ('] boration line. The charging line flow indicator will indicate boric acid flow since the charging pump suction is aligned to the horic acid transfer pump suction for this mode of operation. The change in boric acid tank level is another indication of boric acid injection. On loss of seal injection water to the reactor coolant pump seals, seal water flow may be reestablished by manually starting a standby charging pump. Even if the seal water injection flow is not reestablished, the plant can be operated indefinitely since the thermal barrier cooler has sufficient capacity to cool the reactor coolant flow which would pass through the thermal barrier cooler and seal leakoff from the pump volute. Galvanic Corrosion The only types of materials which are in contact with each other in borated water are stainless steels, Inconel, Stellite valve materials and Zircaloy fuel element cladding. These materials have been shown( } to exhibit (7 only an insignificant degree of galvanic corrosion when coupled to each other. For example, the galvanic corrosion of Inconel versus 304 stainless steel resulting from high temperature tests (575*F) in lithiated, boric acid solution was found to be less than -20.9 mg/dm for the test period of 9 days. Further galvanic corrosion would be trivial since the cell currents at the conclusion of the tests were approaching polarization. Zircaloy versus 304 stainless steel was shown to polarize at 180 F lithiated, boric acid solution in less than 8 days with a total galvanic attack of -3.0 gm/dm. Stellite versus 304 stainless steel was polarized in 7 days at 575*F in lithiated boric acid solution. The total galvanic 2 corrosion for this couple was -0.97 mg/dm, As can he seen from the tests, the effects of galvanic corrosion are insignificant to systems containing borated water. )WCAP 1844 "The Galvanic Behavior of Materials in Reactor Coolants" D. G. Sammarone, August, 1961. m i 9.2-35

1 TABLE 9.2-1 (g/ '~' CHEMICAL AND VOLUME CONTROL SYSTEM CODE REQUIREMENTS Regenerative heat exchanger ASME III*, Class C Non-regenerative heat exchanger ASME III, Class C, tube side, ASME VIII, shell side Mixed bed demineralizers ASME III, Class C Reactor coolant filter ASME III, Class C Volume control tank ASME III, Class C Seal water heat exchanger ASME III, Class C, tube side, ASME VIII, shell side Excess letdown heat exchanger ASME III, Class C, tube side ASME VIII, shell side Chemical mixing tank ASME VIII Deborating demineralizers ASME III, Class C f') Cation bed demineralizer ASME III, Class C M Seal water injection filters ASME III, Class C l Holdup tanks ASME III, Class C Boric acid filter ASME III, Class C Gas stripper package ASME III, Class C Evaporator condensate demineralizers ASME III, Class C Concentrates filter ASME III, Class C Cation ion exchanger ASME III, Class C Ion exchanger filter ASME III, Class C Condensate filter ASME III, Class C Piping and valves USAS B31.l**

• ASME III - American Society of Mechanical Engineers, Boiler and Pressure Vessel Code Section III, Nuclear Vessels.
  • USAS B31.1 - Code for Pressure Piping, and special nuclear cases where fg

) applicable. (~3 a

I TAI 1LE 9.2-2 CllEMICAL AND VOLUME CONTROL SYSTEM PERFORMANCE REQUIREMENTS

• 40 Plant design life, years Seal water supply flow rate, gpm**

24 l 9 Seal water return flow rate, gpm l 60 Normal letdown flow rate, gpn i l 120 Maximum letdown flow rate, gpm Normal charging pump flow (one pump), gpm 69 l 45 Normal Charging line flow, gpm l Maximum rate of boration with one transfer and one charging pump, ppm / min, (from initial l RCS concentration of 1800 ppm) 23.8 Equivalent cooldown rate to above rate of

boraticn, F/ min 6.8 Maximum rate of boron dilution (two charging pumps) ppm / hour (from initial RCS concentration of 2500 ppm) 350 lwo-pump rate of boration, using refueling water, ppm / min (from initial RCS concentration of 10 ppm) 6.2 Equivalent cooldown rate to above rate of boration, *F/ min 1.7 Temperature of reactor coolant entering system at full power, F (design 555.0 Temperature of coolant return to Reactor Coolant System at full power, F (design) 493.0 Normal coolant discharge temperature to holdup tanks, F

127.0 t An.ount cf 11 1/2 boric acid solution required to meet. cold shutdown requirements shortly ~'~s after full power operation, gallons (including ecnsideration for one stuck rod) J 2640 ' ~. - Reactor coolant water quality is given in Table 4.2-4. Volumetric flow rates in gpm are based on 130 F and 2350 psig. Amendment 1

O O O TABLE 9.2-3 PRINCIPAL COMPONENT DATA

SUMMARY

Heat Letdown Letdown Design Design Transfer Flow AT Pressure Temperature Quantity Btu /hr lb/hr

• F psig,shell/ tube
• F, shell/ tube Heat Exchangers 6

Regenerative 1 8.65 x 10 29,826 265 2485/2735 650/650 6 Non regenerative 1 14.8 x 10 29,826 163 150/600 250/400 6 Seal water 1 2.17 x 10 126,756 17 150/150 250/250 6 Excess letdown 1 4.75 x 10 12,400 360 150/2485 250/650 Capacity Design Design Each Pressure Temperature Quantity Type gpm Head psig

• F Pumps Charging 3

Pos. displ. 77-2385 psi 3000 250 Boric acid 2 canned 60 235 ft 150 250 Holdup tank recirculation 1 Centrifugal 500 100 ft 150 200 Primary water makeup 2 Centrifugal 150 300 ft 150 250 Monitor tank 2 Centrifugal 100 150 ft 150 200 Concentrates holding tank transfer 2 Canned 20 150 ft 75 250 Gas stripper feed 2 Canned 12.5 200 ft 150 200 Gas stripper bottom 2 Canned 12.5 93 ft 75 300 Design Design Pressure Temperature Quantity Type Volume, Each psig "F Tanks 3 Volume control 1 Vert. 300 ft 75 Int /15 Ext 250 3 Charging pump accum. 3 Vert. 100 in 3000 250 Boric acid 2 Vert. 7500 gal Atmos. 250 Chemical mixing 1 vert. 5.0 gal 150 250 Batching 1 Jacket Btm. 400 gal Atmos. 250 3 Holdup 3 Horizontal 6,500 ft 15 200

TABLE 9.2-3 (Cont'd) Design Design Pressure Temperature Quantity Type Volume psig

• F 1

Primary water storage 1 Diaphragm 150,000 gal Atmos. 125 Concentrates holding 1 Vertical 925 gal Atmos. 250 Monitor 2 Diaphragm 10,000 gal Atmos. 150 Resin Design Design Vo}ume Flow Pressure Temperature Quantity Type ft gpm_ psig

• F Demineralizers Mixed bed 2

Flushable 30 109 200 250 Cation bed 1 Flushable 20 60 200 250 Base and cation ion exchangers 3 Flushable 30 25 150 250 Evaporator condensate 2 Fixed 30 25 200 250 Deborating 2 Fixed 43 120 200 250 Relief Pressure Quantity psig Capacity Relief valves Charging pump 3 2735 100 gpm Holdup tank 3 12 120 gpm Letdown line (intermediate 240 gpm pressure section 1 600 Letdown line (low pressure section) 1 200 165 gpm Seal water return line 1 150 Batching tank heating 1 320 lb/hr 20 jacket 170 gpm Volume control tank 1 75 O O O

O O O TABLE 9.2-3 (Continued) Design Design Temperature Pressure Quantity

• F psig Filters Reactor coolant 1

250 200 Seal water 1 250 150 Boric acid 1 250 150 Seal water injection 2 200 2735 Concentrates 1 250 100 Condensate 2 250 150 Ion exchanger 2 250 200

TABLE 9.2-4 O PARAMETERS USED IN THE CALCULATION OF REACTOR COOLANT FISSION PRODUCT ACTIVITIES 1. Core thermal power, MWt 2300 2. Fraction of fuel containing clad defects 0.01 3. Reactor coolant liquid volume, cu ft 9400 4. Reactor coolant average temperature, *F 574.1 5. Purification flow rate (normal), gpm 60 6. Effective cation demineralizer flow, gpm 6 7. Volume control tank volumes a. Vapor, cu ft 180 b. Liquid, cu ft 120 8. Fission product escape rate coefficients: -8 a. Noble gas isotopes, sec~ 6.5 x 10 b. Br, I and Cs isotopes, sec~ 1.3 x 10~ c. Te isotopes, sec~ 1.0 x 10~ ~ -9 d. Mo isotopes, sec 2.0 x 10 ~1 Sr and Ba isotopes, sec 1.0 x 10 e. f. Y, La, Ce and Pr isotopes, sec~ 1.6 x 10~ 9. Mixed bed demineralizer decontamination factors: a. Noble gases and Cs-134,136,137, Y-90 and Mo-99 1.0 b. All other isotopes 10.0 10. Cation bed demineralizer decontamination factor for Cs-134, 136, 137, Y-90 and Mo-99 10.0 11. Volume control tank noble gas stripping fraction (closed system): Isotoye Stripping Fraction _ -5 Kr-85 2.3 x 10 ~1 Kr-85m 2.7 x 10 Kr--8 7 6.0 x 10~ Kr-88 4.3 x 10~ Xe-133 1.6 x 10~ -2 Xe-133m 3.7 x 10 -1 Xc-135 1.8 x 10 Xe-135 m 1.0

TABLE 9.2-5 Q \\ ) REACTOR COOLANT SYSTEM EQUILIBRIUM ACTIVITIES (COOLANT TEMPERATURE 583*F) Activation Products uc/cc Mn-54 3.64 x 10~ -2 Mn-56 7.85 x 10 Co-58 1.09 x 10~ Fe-59 2.52 x 10~ -3 Co-60 1.29 x 10 -5 Cr-51 8.10 x 10 Non-Volatile Fission Products (Continuous Full Power Operation) pc/cc pc/cc 5 l 11-3 2.5 (max) 1-131 1.50 ~1 \\Y Br-84 2.6 x 10~ Te-132 1.65 x 10 ~1 Rb-88 2.48 I-132 5.54 x 10 Rb-89 5.8 x 10~ I-133 2.42 ~3 -2 Sr-89 2.5 x 10 Te-134 1.86 x 10 -5 ~l Sr-90 7.5 x 10 I-134 3.40 x 10 -5 ~1 Y-90 9.4 x 10 Cs-134 1.60 x 10 -3 Sr-91 1.15 x 10 1-135 1.30 -2 Y-91 4.4 x 10~ Cs-136 2.30 x 10 -3 -1 Y-92 4.06 x 10 Cs-137 8.69 x 10 -4 Zr-95 5.22 x 10 Cs-138 4.44 x 10~ ~4 Nb-95 5.19 x 10~ Ba-140 5.37 x 10 l -5 -4 Zr-97 1.60 x 10 La-140 5.53 x 10 -3 Mo-99 1.97 Cc-144 1.91'x 10 5 Ru-105 4.5 x 10 ' Pr-144 2.06 x 10' l O ,O Amendment 5 t

O TABLE 9.2-5 (Continued) Gaseous Fission Products uc/cc Kr-85 4.21 Kr-85m 1.01 Kr-87 0.693 5 Kr-88 2.49 Xe-133 1.66 x 10 Xe-135 4.57 Xe-138 3.18 x 10~ O l l l l l I I e l Amendment 5 l

1 TABl.E 9.2-6 ,r " ) N_) TRITIUM PRODUCTION IN Tile REACTOR COOLANT Basic Assumptions: Plant Parameters: 1. Core thermal power, MWt 2300 2. Coolant water volume, ft 9,400 3. Core volume, ft 937.3 4. Core volume fraction a. UO 2990 2 b. Zr + SS .0933 c. 11 0 .6077 2 5. Plant full power operating times a. Initial cycle 78 weeks (18 months) b. Equilibrium 49 weeks (11.3 months) ) 6. Boron Concentrations (Peak hot full power x quilibrium Xe) a. Initial cycle, ppm 890 b. Equilibrium cycle, ppm 825 7. Burnable poison boron content (total-all rod s ), Kg 13.4 8. Fraction of tritium in core (ternary fission + burnable boron) diffusing thru clad 0.30* -5 9. Ternary fission yield 8 x 10 atoms / fission

• The assumption that 30% of the ternary produced tritium diffuses into the coolant is based on the analysis made nf fuel retention in the Saxton and the Yankee stainless clad fuel. This analysis indicated that the fuel re-tained 68% of the tritium produced in the fuel. Although data is not currently available on zircaloy clad fuel operating at the specific power anticipated for these reactors, it is reasonably certain that a significant portion of the tritium released by the fuel will not diffuse through the zircaloy possibly because of the formation of zirconium tricide.

Shippingport data on zircaloy clad fuel indicates LL't less than 1% of ternary tritium produced is released to the coolant. Although this data cannot be used directly, it does indicate that zircaloy will reduce tritium diffusion.

r TABLE 9.2-6 (Cont'd) 9 10. Nuclear cross-sections and neutron fluxes B (n, 2a) T o ; mb (nv; n/cm -sec) 1 Mev < E < 5 Mev = 31.59 (Spectrum weighted) 5.04 x 10 E > 5 Mev = 75 7.4 x 10 Li (n, na) T (99.9% purity) 1 3 Mev < E < 6 Mev = 39.1 (Spectrum weighted) 2.14 x 10 E > 6 Mev = 0.4 '2.76 x 10 Li (n, a) T (99.9% purity Li ) o = 675 barns; av = 2.14 x 10 n/cm -sec 11. Cooling water flow: 4. 8 x 10 gpm = 9.6 x 10 cc/yr II CALCULATIONS curies / year curies / year A. Tritium from Core Initial Cycle Equilibrium Cycle 1. Ternary Fission 8,180 8,180 2. B (n, 2a) T (in poison rods) 592 N.A. 3. B (n, a) Li (n, na) T 1,110 N.A. 4. Release fraction (x 0.30) Total release to Coolant 2,965 2,455 B. Tritium from Coolant 1. B (n, 2a) T 843 582 2. Li (n, na) T (limit 2.2 ppm Li) 6.6 6.6 6 (n, a) T (purity of Li 3. Li = 99.9%) 6.6 6.6 4. Release Fraction (1.0) 5. Total Release to Coolant 856.2 595.2 C. Total Tritium in Coolant 3821 3050 .b

TABLE 9.2-6 (Cont'd) em / 10. Nuclear cross-sections and neutron fluxes B (n, 2a) T o ; mb (nv; n/cm -sec)' 1 Mev 1 E 15 Mev = 31.59 (Spectrum weighted) 5.04 x 10 / 12/ E > 5 Mev = 75 7.4 x 107 Li (n na) T (99.9% purity) 3F v < E < 6 Mev = 39.1 (Spectrum weighted) 2.14 x 10 E > 6 Mev = 0.4 76 x 10 7 6 (n, a) T (. 9% purity Li ) Li / 13 2 o = 675 barn.- nv = 2.14 x 10 n/cm -sec gpm = 9.6 x 10 ' cc/yr 11. Cooling water flow: 4. x 10 II CALCULATIONS curies / year curies / year A. Tritium from Core Initial Cycle Equilibrium Cycle (_f' l. Ternary Fission 8,180 8,180 10 (n, 2a) T (in poison /r ds) 592 N.A. 2. B 0 (n, a) Li (n, na T 1,110 N.A. 3. B 4. Release fraction (x,,0.30)_ Total release to,C'oolant 9,965 2,455 B. Tritium from Coo 1. B (n, 2a) T 843 582 7 2. Li (n, ria) T (limit 2.2 ppm / Li) 6.6 6.6 [n,a)T(purityofLi 3. Li = 99.9%) 6.6 6.6 R' lease Fraction (1.0) 4. e 5. Total Release to Coolant 856.2 595.2 . Total Tritium in Coolant 3821 3050 \\

TABLE 9.2-7 MALFUNCTION ANALYSIS OF CHEMICAL AND VOLUME CONTROL SYSTEM Component Failure Comments and Consequences

1) Letdown line Rupture in the The remote air-operated valve located near the line inside the main coolant loop is closed on low pressurizer reactor contain-level to prevent supplementary loss of coolant ment through the letdown line rupture. The contain-ment isolation valves in the letdown line out-side the reactor containment and also the orifice block valves are automatically closed by the containment isolation signal initiated by the concurrent loss-of-coolant accident. The closure of that valve prevents any leakage of the reactor containment atmosphere outside the reactor con-tainment.

2) Normal and See Above.

The check valves located near the main coolant alternate loops prevent supplementary loss of coolant charging line through the line rupture. The check valves located at the boundary of the reactor containment prevent any leakage of the reactor containment atmosphere outside the reactor containment. 3) Seal water See Above. The motor-operated isolation valve located out-return line side the containment is manually closed or is automatically closed by the containment isola-tion signal initiated by the concurrent loss-of-coolant accident. The closure of that valve prevents any leakage of the reactor containment atmosphere outside the reactor containment.

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The loop design provides for detection of radioactivity entering the loop from reactor coolant source and also provides for isolation means. Residual Heat Removal Loop The residual heat removal loop is designed to remove residual and sensible heat from the core and reduce the temperature of the Reactor Coolant System during the second phase of plant cooldown. During the first phase of cool-down, the temperature of the Reactor Coolant System is reduced by transferring heat from the Reactor Coolant System to the Steam and Power Conversion System. O) \\s_- 9.3-1 All active loop components which are relied upon to perform their function are redundant. The loop design precludes any significant reduction in the overall design reactor shutdown margin when the loop is brought into operation for residual heat removal or for emergency core cooling by recirculation. The loop design includes provisions to enable hydrostatic testing to applicable code test pressures during shutdown. Loop components, whose design pressure and temperature are less than the Reactor Coolant System design limits, are provided with overpressure protective devices and redundant isolation means. Spent Fuel Pit Cooling Loop The spent fuel pit cooling loop is designed to remove from the spent fuel pit the heat generated by stored spent fuel elements. Loop design does not incorporate redundant active components. Alternate cooling capability can be made available under anticipated malfunctions or failures (expected fault conditions). Loop piping is so arranged that failure of any pipeline does not drain the spent fuel pit below the top of the stored fuel elements. The design basis for the loop provides the capability to totally unload the reactor vessel for maintenance or inspection at the time that 1/3 of a core already occupies the spent fuel storage pool. O 9.3-2 N 43 ' j jl. Design Characteristics r Component Cooling Loop ^ i One pump and one component heat exchanger are non ally operated to provide;,,, cooling water for various components located in the auxiliary and containme$t ~ buildings. The water is normally supplied to all components being cooled ' s,,( 'ven though one of the components may be out of service. s sA_ Makeup water is taken from the primary wa'ter treatment plant, as required and delivered to the surge tank. A backup source of water is provided'from? the primary water make-up pumps. s A The operation of the loop is-monitored with the following instrumentation: a) Temperature detectors in (he inlet and outlet lines'for the component cooling heat exchangers b) A pressure detector on,the lidet bbtween the component cooling pumps, " I and the component cooling heat exchbngers i ^' t ^ ss s \\ f ? c) A temperature and flow indicator in the outlet header from the heat exchangers .q ) s s d) A radiation monitor og the inlet header to the compodent cooling' pumps.' { - t. Residual Heat Removal Lcgg Two pumps and two residual beat exchangers perform the decay heat cooling functions for the reactor unit. After the Reactor Coolant System temperature and pressure have been red 2ced to 350*F and 350 psig respectively, decay heat -s cooling is initiated by aligning the pumps to take suction from the<feactor outlet line and discharge through the heat exchangers into the reactor inlet line. If only one pump and one heat exchanger are available reduction of reactor coolant temperature is accomplished at a lower rate. ~ . s =,, 4 kb 9.3-3 l The equipment utilized for decay heat cooling is also used for emergency core cooling during loss-of-coolant accident conditions. This is described in Section 6. Spent Fuel Pit Cooling Loop During normal conditions 1/3 of a core is stored in the pool. When 1/3 of a core is present, the pump and spent fuel heat exchanger will handle i the load and maintain a pit water temperature less than 120"F. When 1-1/3 cores are stored, the pit is maintained below 150*F. The pool is initially filled with water from the refueling water storage tank. Codes and Classifications All piping and components of the Auxiliary Coolant System are designed to the applicable codes and standards listed in Table 9.3-4. The component cooling loop water contains a corrosion inhibitor to protect the carbon steel piping. Austenitic stainless steel piping is used in the residual heat removal loop, ubich contains reactor coolant, and in the spent fuel pit cooling loop, which contains water without corrosion inhibitor. 9.3.2 SYSTEM DESIGN AND OPERATION , Component Cooling Loop t Component cooling is provided for the following heat sources: a) Residual heat exchangers (Auxiliary Coolant System, ACS) b) Reactor coolant pumps (Reactor Coolant System) c) Non-regenerative heat exchanger (Chemical and Volume Control System, CVCS) d) Excess letdown heat exchanger (CVCS) e) Seal water heat exchanger (CVCS) f) Boric acid recycle evaporator and condensate coolers (CVCS) O 9.3-4 g) Sample heat exchangers (Sampling System) h) Waste evaporator condenser (Waste Disposal System) i) Waste gas compressors (Waste Disposal System) j) Residual heat removal pumps (ACS) k) Safety injection pumps (Safety Injection System, SIS) 1) Containment spray pumps m) Spent fuel pit heat exchanger (ACS) n) Charging pumps (CVCS) o) Control rod drive air-water heat exchanger At the reactor coolant pump, component cooling water removes heat from the bearing oil and the thermal barrier. Since the heat is transferred from the component cooling water to the service water, the component cooling loop serves as an intermediate system between the reactor coolant and service water cooling system. This double barrier arrangement reduces the probability of leakage of high pressure, potentially radioactive coolant to the service water system. During normal full power operation, one component cooling pump and one component cooling heat exchanger accommodate the heat removal loads. One of the two standby pumps provides 100% backup and a heat exchanger provides 100 per cent backup during normal operation. Three pumps and two heat ex-changers are utilized to remove the residual and sensible heat during plant shutdown. If one of the pumps or one of the heat exchangers is not operative, safe shutdown of the plant is not affected, however, the time for cooldown is extended. The surge tank accommodates expansion, contraction and in-leakage of water, and ensures a continuous component cooling water supply until a leaking cooling line can be isolated. Because the tank is normally vented to the atmosphere, a radiation monitor in the component cooling pump inlet header annunciates in the control room and closes a valve in the surge tank vent line in the unlikely event that the radiation level reaches a preset level above the normal background. 9.3-5 s Residual llent Removal Loop The residual heat removal loop consists of heat exchangers, pumps, piping and the necessary valves and instrumentation. During plant shutdown, coolant flows from the Reactor Coolant System to the residual heat removal pumps, through the tube side of the residual heat exchangers and back to the Reactor Coolant System. The inlet line to the residual heat removal loop starts at the hot leg of one reactor coolant loop and the return line connects to the low head Safety Injection System Piping to the three cold legs. The residual heat exchangers are used to cool the water during the latter phase of Safety Injection System operation. These duties are defined in Section 6. The heat loads are transferred by the residual heat exchangers to the component cooling water. During plant shutdown, the cooldown rate of the reactor coolant is controlled by regulating the flow through the tube side of the residual heat exchangers. A by-pass line and an automatic flow control valve arcund the residual heat exchangers are used to maintain a constant flow through the residual heat removal loop and to control cooldown. Double, remotely operated valving in the inlet line is provided to isolate the residual heat removal loop from the Reactor Coolant System. When Reactor Coolant System pressure exceeds the design pressure of the residual heat removal loop, an interlock between the Reactor Coolant System wide range pressure channel and the first inlet valve prevents the valve from opening. A remotely operated valve and two check valves isolate each line to the Reactor Coolant System cold legs from the residual heat removal loop. Over pressure in the loop is prevented by a relief valve which discharges to the pressurizer relief tank. Spent Fuel Pit Cooling Loop The spent fuel pit cooling loop removes residual heat from fuel stored in the spent fuel pit. The loop is normally required to handle the heat load from 1/3 of the core freshly discharged from the reactor but it can safely 9.3-6 w \\ accommodate the heat load from 1-1/3 cores, for which there is storage space available. The spent fuel is placed in the pit during refueling and is stored until it is shipped to a reprocessing facility. The spent fuel pit is located outside the reactor containment and is not affected by any loss-of-coolant accident in the containment. The water in the pit is isolated by a valve from that in the refueling canal during most of the refueling operation. Only a very small amount of interchange of water occurs as fuel assemblies are transferred during refueling. The spent fuel pit cooling loop consists of a pump, heat exchanger, filter, demineralizer, piping and associated valves and instrumentation. The pump draws water from the pit, circulates it through the heat exchanger and returns it to the pit. A second pump is useu to circulate refueling water through the demineralizer and filter for purification. Component cooling water cools the heat exchanger. Redundancy of this equipment is not required because of the large heat capacity of the pit and the slow heat up rate as () shown in Table 9.3-3. However, in the event of failure of the spent fuel pump, alternate connections are provided for connecting a temporary pump to the spent fuel pit loop. This consists of blind flange connections in the suction and discharge piping. The clarity and purity of the spent fuel pit water is maintained by passing approximately 5 per cent of the loop flow through a filter and demineralizer. The spent fuel pit pump suction line, which is used to drain the pit, pene-trates the spent fuel pit wall above the fuel assemblies. The penetration location prevents loss of watar as a result of a possible suction line rupture. Component Cooling Loop Components Component Cooling Heat Exchangers The component cooling heat exchangers are of the shell and straight tube type. Service water circulates through the tubes while component cooling water ,,s k_,) circulates through the shell side. Parameters are presented in Table 9.3-1. 9.3-7 Component Cooling Pumps The three component cooling pumps which circulate component cooling water through the component cooling loop are are horizontal, centrifugal units. The pump casings are made from cast iron (ASTM 48) based on the corrosion-erosion resistance and the ability to obtain sound castings. The material thickness is dictated by high quality casting practice and ability to withstand mechanical damage and as such are substantially overdesigned from a stress level standpoint. Parameters are presented in Table 9.3-1. Component Cooling Surge Tank The component cooling surge tank which accommodates changes in component cooling water volume is constructed of carbon steel. Parameters are pre-sented in Table 9.3-1. In addition to piping connections, the tank has a flanged opening at the top which can be used, if required, for the addition of the chemical corrosion inhibitor to the component cooling loop. Chemical Pot Feeder Tank The chemical pot feeder tank provides for the direct addition of corrosion additive to the component cooling water. Parameters are listed in Table 9.3-1. Component Cooling Valves The valves used in the component cooling loop are constructed of carbon steel with bronze or stainless steel trim. Since the component cooling water is not normally radioactive, special valve features (such as special leakoff times to the Waste Disposal System) to prevent leakage to the atmosphere are not provided. Self-actuated spring loaded relief valves are provided for lines and com-ponents that could be pressuri:ed to their design pressure by improper operation or malfunction. O 9.3-8 Component Cooling Piping All component cooling loop piping is carbon steel with welded joints and connections except at components which might need to be removed for maintenance. Residual Heat Removal Loop Components Residual Heat Exchangers The two residual heat exchangers located within the auxiliary building are of the ahell and U-tube type with the tubes wel'.ed to the tube sheet. Reactor coolant circulates through the tubes, while component cooling water circulates through the shell side. The tubes and other surfaces in contact with reactor coolant are austenitic stainless steel and the shell is carbon steel. Residual Heat Removal Pumps The two residual heat removal pumps are in-line, centrifugal units with special seals to prevent reactor coolant leakage to the atmosphere. All pump parts in contact with reactor coolant cre austenitic stainless steel or equivalent corrosion resistant material. Residual Heat Removal Valves The valves used in the residual heat removal loop are constructed of austenitic stainless steel or equivalent corrosfon resistant material. Manual stop valves are provided to isolate equipment for maintenance. Throttle valves are provided for remote and manual control of the residual heat exchanger tube side flow. Check valves prevent reverse flow through the residual heat removal pumps. Double, remotely operated series stop valves are provided at the inlet to isolate the residual heat removal loop from the Reactor Coolant System. x 9.3-9 When Reactor Coolant System pressure exceeds the design pressure of the residual heat removal loop, an interlock between the Reactor Coolant System j wide range pressure channel and the first inlet valve prevents the valve from opening. A remotely operated stop valve and two series check valves isolate each line to the Reactor Coolant System cold legs from the residual heat removal loop. Overpressure in the residual heat removal loop is prevented by a relief valve which discharges to the pressurizer relief tank. Valves that perform a modulating function are equipped with two sets of packing and an intermediate leakoff connection that discharges to the Waste Disposal System. Manually operated valves have backseats to facilitate repacking and to limit the stem leakage when the valves are open. Residual lleat Removal Piping All residual heat removal loop piping is austenitic stainless steel. The piping is welded except for flanged connections at the control valves. Spent Fuel Pit Loop Components Spent Fuel Pit IIcat Exchanger The spent fuel pit heat exchanger is of the shell and U-tube type with the tubes welded to the tube sheet. Component cooling water circulates through the shell, and spent fuel pit water circulates through the tubes. The tubes are austenitic stainless steel and the shell is carbon steel. Spent Fuel Pit Pump The spent fuel pit pump circulates water in the spent fuel pit cooling loop. All wetted surfaces of the pump are austenitic stainless steel, or equivalent corrosion resistant material. Pump operation is manually controlled from a local station. 9.3-10 Refueling Water Purification Pump The refueling water purification pump circulates water in a loop between the refueling water storage tank and the spent fuel pit demineralizer and filter. All wetted surfaces of the pump are austenitic stainless steel. The pump is operated manually from a local station. Spent Fuel Pit Strainer A stainless steel strainer is located at the inlet of the spent fuel pit loop suction line for removal of relatively large particles which might otherwise clog the spent fuel pit demineralizer. Spent Fuel Pit Filter The spent fuel pit filter removes particulate matter from the spent fuel pit water. The filter cartridge is synthetic fiber and the vessel shell ['V} is austenitic stainless steel. Spent Fuel Pit Demineralizer The demineralizer is sized to pass 5% of the loop circulation flow, to provide adequate purification of the fuel pit water for unrestricted access to the working area, and to maintain optical clarity. Spent Fuel Pit Skimmer A skimmer pump and filter are provided for surface skimming of the spent fuel pit water. Flow from this pump is returned to the spent fuel pit. Spent Fuel Pit Valves Manual stop valves are used to isolate equipment and lines, and manual throttle valves provide flow control. Valves in contact with spent fuel v 9.3-11 a pit water are austenitic stainless steel or equivalent corrosion resistant material. Syent Fuel Pit Piping All piping in contact with spent fuel pit water is austenitic stainless steel. The piping is welded except where flanged connections are used at the pumps, heat exchanger and control valve to facilitate maintenance. 9.3.3 SYSTEM EVALUATION Availability and Reliability Componen t Cooling Loop For continued cooling of the reactor coolant pumps, and the excess letdown heat exchanger, most of the piping, valves, and instrumentation are located outside the primary system concrete shield at an elevation well above the anticipated post-accident water level in the bottom of the containment. (The exception is the cooling lines for the reactor coolant pumps which can be isolated by two valves in series following the accident.) In this annular area the component cooling equipment is protected against credible missiles and from being flooded during post-accident operation. Also, this location l provides radiation shielding which allows for maintenance and inspections i to be performed during power operation. I l Outside the containment, the residual heat removal pumps, the residual heat l l exchangers, the spent fuel heat exchanger, the component cooling pumps and heat exchangers, and associated valves, piping and instrumentation are maintainable and inspectable during power operation. System design provides for the replacement of one pump or one heat exchanger while the other units are in service. O 9.3-12 r Several of the components in the componert cooling loop are fabricated from carbon steel. The component cooling water contains a corrosion inhibitor to protect the carbon steel. Welded joints and connections are used except where flanged closures are employed to facilitate maintenance. The entire system is seismic Class I design. The components'are designed to the codes given in Table 9.3-4. In addition the components are not subjected to any high pressures (See Table 9.3-1) or stresses, llence a rupture or failure of the system is very unlikely. During the recirculation phase following a loss-of-coolant accident, one of the three component cooling water pumps delivers flow to the shell side of one of the residual heat exchangers. Residual lleat Removal Loop Two pumps and two heat exchangers are utilized to remove residual and sensible heat during plant cooldown. If one of the pumps and/or one of the heat exchangers is not operative, safe operation of the plant is not affected; -~ s Jr however, the time for cooldown is extended. The function of this equipment s following a loss-of-coolant accident is discussed in Section 6. i Spent Fuel Pit Cooling Loop l l This manually controlled locp may be shutdown safely for reasonable time l periods, as shown in Table 9.3-3, for maintenance or replacement of mal-1 functioning components. i Leakage Provisions Component Cooling Loop l Welded construction is used where possible throughout the component cooling l loop piping, valves and equipment to minimize the possibility of leakage, l ('~) I %, ) l 9.3-13 i -,a -m r The component cooling water could become contaminated with radioactive water due to a leak in any heat exchanger tube in the Chemical and Volume Control, h the Sampling, or the Auxiliary Coolant Systems, or a leak in the cooling coil for the reactor coolant pump thermal barrier. Tube or coil leaks in components being cooled would be detected during normal plant operation by the leak detection system described in Sections 4.2.7 and 6.5. Such leaks are also detected anytime by a radiation monitor located on the main return header. Leakage from the component cooling loop can be detected by a falling level in the component cooling surge tank. The rate of water level fall and the area of the water surface in the tank permit determination of the leakage rate. To assure accurate determinations, the operator would check that tempe rat ures a re s t ab le. The component which is leaking can be located by sequential isolation or inspection of equipment in the loop. If the leak is in the on-line component lh cooling water heat exchanger, the standby exchanger would be put on line and the leaking exchanger isolated and repaired. During nonnal operation the leaking exchanger could be left in service with leakage up to the capacity of the makeup line to the system f rom the primary water treatment plant. By manual transfer, emergency power is available for makeup pump operation. 1 An incredible cooling water temperature increase of about 250 F would be required to overfill the component cooling surge tank. However, should l a large tube side to shell side leak develop in a residual heat e xchange r, the water level in the component cooling surge tank would rise, and the operator would be alerted by a high water alarm. The atmospheric vent on the tank is automatically closed in the event of high radiation level at the component cooling water pump suction header. If the leaking residual I heat exchanger is not isolated from the component cooling loop before the inflow completely fills the suron tank, the relief valve on the surge tank l lifts. The discharge of this relief valve is routed to the auxiliary building waste holdup tank. O Amendment 1 9.3-14 The severence of a cooling line serving an individual reactor coolant pump cooler would result in substantial leakage of component cooling water. However, the piping is small as compared to piping located in the missile protected area of the containment. Therefore, the water stored in the surge tank af ter a low level alarm together with makeup flow provides ample time for the closure of the valves external to the containment to isolate the leak before cooling is lost to the essential components in the component cooling loop. The relief valves on the component cooling water header downstream from each of the reactor coolant pumps are designed with a capacity equal to the maxi-mum rate at which reactor coolant can enter the component cooling loop from a severence type break of the reactor coolant pump thermal barrier cooling coil. The valve set pressure equals the design pressure of the component cooling piping. The relief valves on the cooling water lines downstream from the sample, excess letdown, seal water, non-regenerative, spent fuel pit and residual heat exchangers are sized to relieve the volumetric expansion occurring if the exchanger shell side is isolated when cool, and high temperature coolant flows through the tube side. The set pressure equals the design pressure of the shell side of the heat exchangers. The relief valve on the component cooling surge tank is sized to relieve the maximum flow rate of water which enters the surge tank following a rupture of a reactor coolant pump thermal barrier cooling coil. The set pressure equals the design pressure of the component cooling surge tank. Initial protection is provided by an isolation valve which automatically closes on high flow in the event of a thermal barrier coil rupture. Residual Heat Removal Loop During reactor operation all equipment of the residual heat removal loop is idle, and the associated isolation valves are closed. During the loss-of-9.3-15 coolant accident condition, water from the containment sump is recirculated through the exterior piping system. To obtain the total radiation dose to the public due to leakage from this system, the potential leaks have been evaluated and discussed in Sections 6 and 14. Each of the two residual heat removal pumps is located in a shielded com-partment with a floor drain. In each compartment the leakage drains to a sump and is then pumped to the waste holdup tank by sump pumps. Two 100 gpm sump pumps are provided and each is capable of handling the flow which results from the failure of a residual heat removal pump seal. Each sump has a level indicator which will warn the operator of high water level. Both of the lines from the containment sump to the individual residual heat removal pumps has two remotely operated isolation valves in series. Spent Fuel Pit Cooling Loop Whenever a leaking fuel assembly is transferred from the fuel transfer canal to the spent fuel storage pool, a small quantity of fission products may enter the spent fuel cooling water. A small purification loop is provided for removing these fission products and other contaminants from the water. The probability of inadvertently draining the water from the cooling loop of the spent fuel pit is exceedingly low. The only means of draining the cooling loop is through such actions as opening a valve on the cooling line and leaving it open when the pump is operating. In the unlikely event of the cooling loop of the spent fuel pit being drained, the spent fuel storage pit itself cannot be drained and no spent fuel is uncovered since the spent fuel pit cooling connections enter near the top of the pit. With no heat removal the time for the spent fuel pit water to rise from 120 F to 180 F with 1/3 core stored in the pit is approximately 20.0 hours. The temperature and level indicators in the spent fuel pit would warn the operator of the loss of cooling. This slow heatup rate of the spent fuel pit would allow O 1 l l 9.3-16 l sufficient time to take any necessary action to provide adequate cooling using the emergency cooling connections provided while the cooling cap-ability of the spent fuel pit cooling loop is being restored. Incident Control Component Cooling Loop Containment isolation valves are automatically closed on a high containment pressure signal. The cooling water supply header to the reactor coolant pumps contains a check valve inside and two remotely operated valves outside the containment wall. The cooling water supply line to the excess letdown heat exchanger contains a check valve inside the containment wall which is closed during normal operation. Except for the normally closed makeup line and equipment vent and drain lines, there are no direct connections between the cooling water and other systems. The equipment vent and drain lines outside the containment have manual valves which are normally closed unless the equipment is being vented or drained for maintenance or repair operations. The vent lines are also capped as an additional safety feature. Following a loss-of-coolant accident, one component cooling pump and one component cooling heat exchanger accommodate the heat removal loads. If either a component cooling pump or component cooling heat exchanger fails, one of the two standby pumps and the standby heat exchanger provide 100% backup. Valves on the component cooling return lines from the safety injec-tion, containment spray and residual heat removal pumps are normally open. Each of the component cooling return lines from the residual heat exchangers has a ncrmally closed remotely operated valve. If one of the valves fails to open at initiation of long-term recirculation, the valve which does open sup-plies a heat exchanger with sufficient cooling to remove the heat load. If the break of a cooling line occurs inside the containment, adequate valving is available outside the containment on the component cooling supply and return lines to isolate the leak (see Figure 9.3-1). None of the D 9.3-17 components inside the containment require conponent cooling water during recirculation. If the break occurs outside the containment, the leak could either be isolated by valving or the broken line could be repaired, depen-ding on the position in the loop at which the break occurred. Once the leak is isolated or the break has been repaired, makeup water is supplied from the reactor makeup water tank by either one of the reactor makeup water pumps. Residual Heat Removal Loop The residual heat removal loop is connected to the reactor outlet line on the suction side and to the reactor inlet line on the discharge side. On the suction side the connection is through two electric motor-operated gate valves in series with the first valve interlocked with reactor coolant system pressure. On the discharge side the connection is through two check valves in series with an electric motor operated gate valve. All of these are closed whenever the reactor is in the operatine ondition. Spent Fuel Pit Cooling Loop The most serious failure of this loop is complete loss-of-water in the storage pit. To protect against this possibility, the cooling pump suction connection penetrates the pit wall and terminates near the normal water level so that a break in the pipe will not gravity drain the pit. The pit drain piping penetrates the pit wall at an elevation 6 feet above the top of the fuel assemblies. Complete siphon draining of the pit by a break in this line is prevented by a normally closed valve located near the pit wall at the same elevation as the penetration. A break in this line upstream of the valve will only drain the pool to an elevation 6 feet above the fuel assemblies. The cooling water return line does not penetrate the pit wall and is prevented from siphon draining the pit by a 0.5 inch hole in the pipe located 6 inches below the normal water level. O 9.3-18 1 In the event of failure of the spent fuel pit pump or loss of cooling to the heat exchanger, an alternate means for cooling the spent fuel pit water is provided. Alternate cooling connections are provided in the spent fuel pit loop and component cooling loop piping for connecting temporary piping to the component cooling heat exchanger and for connecting a temporary pump to the spent fuel pit loop. Malfunction Analysis A failure analysis of pumps, heat exchangers and valves is presented in Table 9.3-5. 9.3.4 TEST AND INSPECTION CAPABILITY The residual heat removal pumps flow instrument channels can be calibrated during shutdown. The active components of the Auxiliary Coolant System are in either continuous or intermittent use during normal plant operation, thus no ad-ditional periodic tests are required. Periodic visual inspections and preventative maintenance can be conducted as necessary, l O 9.3-19 Amendment 1 1 TABLE 9.3-1 ,,Ag COMPONENT COOLING LOOP COMPONENT DATA Component Cooling Pumps Quantity 3 Type llorizontal centrifugal Rated capacity, gpm 6000 Rated head, f t 11 0 180 2 Motor horsepower, hp 350 Casing material Cast iron Design pressure, psig 150 Design temperature, 'F 200 Component Cooling lleat Exchangers ,G Quantity 2 b Type Shell and straight tube 6 IIcat transferred, Btu /hr (Shutdown Condition) 29.4 x 10 Shell side (component cooling water) - Inlet temp., *F 115 Outlet temp., *F 108 0 Design flow rate, lb/hr 4.46 x 10 Design temperature, *F 200 Design pressure, psig 150 Material Carbon steel Tube side (service water) - Inlet temperature, *F 95 Outlet temperature, "F 101 6 Design flow rate, lb/hr 4.96 x 10 Design pressure, psig 15 0 Design temperature, *F 200 Material Admiralty OI v l TABLE 9. 3-1 (Continued) Component Cooling Surge Tank Quantity 1 Volume, gal 2000 Normal water volure, gal. 1000 Design pressure, psig 100 Design temperature, *F 200 Construction material Carbon steel Chemical Pot Feeder Tank Quantity 1 Volune, gal. 3 Design pressure, psig 150 Design temperature, *F 150 Component Cooling Loop Piping and Valves Design pressure, psig 150 Design temperature, F 200 0 O TABLE 9. 3-2 MIDUAL llEAT REMOVAL IDOP COMPONENT DATA Reactor coolant temperature at startup of residual heat removal,

• F 350 Time to cool reactor coolant system f rom 350*F to 140*F, hr (all equipment operational) 20 Refueling water storage temperature. *F Ambient Decay heat generation at 20 hrs after shutdown 6

condition, Btu /hr 45 x 10 Approximate reactor cavity fill time, hr 1 Approximate reactor cavity drain time, hr 4 11 B0 concentration in refueling water storage 3 3 2000-2500 tanks, ppm boron Residual heat removal pumps Quantity 2 Type in-line Centrifugal Rated capacity, gpm 3750 Rated head, f t 110 225 2 Motor horsepower, hp 300 Material SS Design pressure, psig 600 Design temperature, 'F 400 Residual Heat Exchangers Quantity 2 Type Shell and U-tube 6 Heat transfer, Btu /hr 29.4 x 10 TABLE 9. 3-2 (Continued) Shell side (component cooling water) Inlet temperature, F* 108 Outlet temperature, *F 115 6 Design flow rate, Ib/hr 4.29 x 10 Design pressure, psig 150 Design temperature, F 200 Mate rial Carbon steel Tube side (reactor coolant) l Inlet temperature, *F 140 Outlet tempe rat u re, *F 124 6 Design flow rate, lb/hr 1.88 x 10 Design pressure, psig 600 Design temperature, F 400 Material Stainless steel O I l l 9 TABLE 9.3-3 SPENT FUEL COOLING LOOP COMPONENT DATA System Cooling Capacity, Btu /hr 6 Normal (1/3 core) 6.54 x 10 2 0 Maximum (1-2/3 cores) 8.12 x 10 Pit Water Heat Inertia, No Heat Removal Time to heat from 120 to 180*F, 2/3 core, hr 30 from 134(1} to 180 F, 1-1/3 core, hr 19 Time to heat Spent fuel pit heat exchanger Quantity 1 Type Shell and U-tube 6 transfer ( , Btu /hr 7.96 x 10 Heat Shell side (component cooling water) Inlet temperature, *F 100 Outlet temperature, *F 106 6 Design flow rate, lb/hr 1.4 x 10 Design pressure, psig 150 Design temperature, *F 200 Material Carbon steel Tube side (spent fuel pit water) Inlet temperature, *F 120 Outlet temperature, *F 113 Design flow rate, lb/hr 1.1 x 10 Design pressure, psig 150 Design temperature, *F 200 Material Stainless steel Note 1: The temperature of equilibrium with cooling system operating. Note 2: Assumed pool water to heat exchanger at 120*F and cooling water to heat exchanger at 100*F. Amendment 2 TABLE 9.3-3 (Continued) Sheet 2 of 4 Spent Fuel Pit Pump Data Quantity 1 Type Horizontal centrifugal Minimum developed head, f t 110 125 2 Motor horsepower 100 Design pressure, psig 150 Design temperature, "F 200 Material Stainless steel Spent Fuel Storage Pool volume, ft 37,000 Boron concentration, ppm baron 2000 to 2500 Spent Fuel Pit Filter Quantity 1 Type Replaceable cartridge Internal design pressure of housing, psig 150 Design temperature, *F 200 Rated flow, gpm 100 Manimum differential pressure across filter element at rated flow (clean cartridge), psi 5 Manimum differential pressure across the filter element prior to cartridge replacement, psi 20 l l Spent Fuel Pit Demineralizer Quantity 1 Type Flushable Design pressure, psig 200 Design temperature, *F 250 l Design flow rate, gpm 100 Resin volume, cu. ft. 30 vessel volume, cu. ft. 43 l l l TABI.E 9.3-4 AUXT I.I ARY '._001. ANT SYSTEM Col >E !<l;quIIti:MENTS Component' cooling heat exchangers ASME VIII* Component cooling surge tank ASME VIII Component cooling loop piping and valves ASA B31.l** 5 Residual heat exchangers ASME III***, Class C, tube side, ASME VIII, 'S l shell side Residual heat removal piping and valves USAS B31.1 Spent fuel pit filter ASME III, Class C Spent fuel heat exchanger ASME III, Class C, tube side, ASME VIII, !S shell side Spent fuel pit demineralizer ASME III, Class C f5 Spent fuel pit loop piping and valves ASA B31.1 e ASME VIII - American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section VIII ASA B31.1 - Code for Pressure Piping, and special nuclear cases l3 where applicable ASME III - American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Nuclear Vessels. /% Amendment 5 J F TABLE 9.3-5 ]AILL'RE M ALYSIS OF PUMPS, HEAT EXCHM:CERS, A*;D VALVES Components Malfunction Comments and Consequences 1. Component cooling water Rupture of a pump The casing is designed for 150 psi and 200*F which pumps casing exceeds maximum operating conditions. Pu p is inspectable and protected against missiles. Rupture due to missiles is not considered credible. Each unit is isolable. Any of the three pumps can carry the total pumping load. 2. Component cooling Pump fails to start One operating pump supplies sufficient water for cooling. water pumps 3. Component cooling Manual valve on This is prevented by prestartup and ope acional checks, water pumps a pump suction Further, during nomal operation, each pump is checked line closed on a periodic basis which would show if a valve is closed. 4. Component cooling Valve on discharge The valve is checked open during periodic operation line sticks closed of the pumps during nomal operation. water pumps 5. Component cooling Tube or shell Rupture is considered improbable because of low heat exchange r rupture operating pressures. Each unit is isolable. 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AUXILIARY C00LAST SYSTEM - SPENT FUEL PIT C00Lh G FIGURE 9.3 3 O 9.4 SAMPLING SYSTEM 9.4.1 DESIGN BASES P_erformance Requirements This system provides samples for laboratory analysis to evaluate reactor coolant, and other reactor auxiliary systems chemistry during normal operation. It has no active emergency function. This system is normally isolated at the containment boundary. Sampling system discharge flows are limited under normal and anticipated 4 fault conditions (malfunctions or failure) to preclude any fission product releases beyond the limits of 10 CFR 20. Design Characteristics v The system is capable of obtaining reactor coolant samples during reactor operation and during cooldown when the system pressure is low and the residual heat removal loop is in operation. Access is not required to the containment. Sampling of other process coolants, such as tanks in the Waste Disposal System, is accomplished locally. Equipment for sampling secondary and non-radioactive fluids is separated from the equipment provided for reactor coolant samples. Leakage and drainage resulting from the sampling operations are collected and drained to tanks located in the Waste Disposal System. l Two types of samples are obtained by the system: high temperature -high pressure Reactor Coolant System samples which originate inside the reactor containment, and low temperature -low pressure samples from the Chemical and Volume Control and Auxiliary Coolant Systems. i v 9.4-1 O liigh Pressure - High Temperature Samples A sample connection is provided from each of the following: np a) The pressurizer steam space b) The pressurizer liquid space c) Hot legs of loops 2 and 3 d) Blowdown lines from each steam generator Low Pressure - Low Temperature Samples A sample connection is provided from each of the following: a) The mixed bed demineralizer inlet header b) The mixed bed demineralizer outlet header c) The residual heat removal loop, just downstream of the heat exchangers d) The volume control tank gas space e) The accumulators Expected Operating Temperatures The high pressure, high temperature samples and the residual heat removal loop samples leaving the sample heat exchangers are held to a temperature of 130 F to minimize the generation of radioactive aerosols. Codes and Standards Sys tem component code requirements are given in Table 9.4-1. O 9.4-2 9.4.2 SYSTEM DESIGN AND OPERATION s / T d (u s The :;ampling Systen, shown in Figure 9.4-1, provides the representative samples for laboratory analysis. Analysis results provide guidance in the operation of the Reactor Coolant, Auxiliary Coolant, Steam and Chemical and Volume Control Systems. Analyses show both chemical and radiochemical conditions. Typical information obtained includes reactor coolant boron and chloride concentrations, fission product radioactivity level, hydrogen, oxygen, and fission gas content, corrosion product concentration, and chemical additive concentration. The information is used in regulating boron concentrations adjustments, evaluating fuel element integrity and mixed bed demineralizer performance, and regulating additions of corrosion controlling chemicals to the systems. The Sampling System is designed to be operated manually, on an intermittent basis. Samples can be withdrawn under conditions ranging from full power to cold shutdown. (Q Y I' Reactor coolant liquid lines, which are normally inaccessible and require frequent sampling, are sampled by means of permanently installed tubing leading to the sampling room. l l Sampling System equipment is located inside the auxiliary building with most of it in the sampling room. The delay coil and sample lines with remotely operated valves are located inside the reactor containment. Reactor coolant hot leg liquid, accumulator liquid, pressurizer liquid and pressurizer steam samples originating inside the reactor containment flow through separate sample lines to the sampling room. Each of these connections to the Reactor Coolant System has a remote operated isolation valve located close to the sample source. The samples pass through the reactor containment compartment, to the auxiliary building, and into the sampling room, where they are cooled (pressurizer stea a samples condensed and cooled) in the sample heat exchangers. The samp le stream pressure is reduced by a manual throttling valve located downstream of each sample pressure vessel. The 9.4-3 sample steam is purged to the volume control tank in the Chemical and Volume Control System until suf ficient purge volume has passed to permit collection of a representative sample. After sufficient purging, the sample p ressure vessel is isolated and then disconnected for laboratory analysis of the contents. Alte rna tely, liquid samples may be collected by bypassing the sample pressure vessels. After sufficient purge volume has passed to permit collection of a representative sample, a portion of the sample flow is diverted to the sample sink where the sample is collected. The reactor coolant sample originating f rom the residual heat removal loop of the Auxiliary Coolant System has a remote operated, normally closed isolation valve located close to the sample source. The sample line from this source is connected into the sample line coming from the hot leg at a point ahead of the sample heat exchanger. Samples from this source can be collected either in the sample pressure vessel or at the sample sink as with hot leg samples. Liquid samples originating at the Chemical and Volume Control System letdown line at demineralizer inlet and outlet pass directly through the purge line to the volume control tank. Samples are obtained by diverting a portion of the flow to the sample sink. If the pressure is low in the letdown line, the purge flow is directed to the chemical drain tank. The sample line from the gas space of the volume control tank delivers gas samples to the volume control tank sample pressure vessel in the sampling room. Purge flow for these samples is discharged to the vent header in the Waste Disposal System. Because the pressurizer steam phase, the reactor coolant dissolved gas and volume control tank gas phase samples may contain accumulated radioactive gases, the respective sample vessel stations are located in small, well ventilated and shielded cubicles within the sampling room. O 9.4-4 Amendment 1 /} Samples of the steam generator liquid are obtained from the blowdown lines. These sample lines are routed by separate lines from each steam generator blowdown line into the sample room. These lines are equipped with two remote operated isolation valves in each line immediately outside the containment. These valves are automatically closed upon receipt of a signal from the blowdown sample radiation monitor or the containment isolation system. The sample lines are routed to the sample room where the liquid is cooled and the pressure reduced. Each individual sample is then split into two routes: one goes to the sample sink to provide periodic samples for chemical analysis, the second goes to a conductivity cell, a radiation monitor and then to blowdown flash tank. This second line handles a continuous flow for a constant reading of conductivity and a constant monitoring for radiation. The sample sink, which is contained in the laboratory bench as a part fg of the sampling hood, contains a drain line to the Waste Disposal System. U Local instrumentation is provided to permit manual control of sampling operations and to ensure that the samples are at suitable temperatures and pressures before diverting flow to the sample sink. Components A summary of principal component data is given in Table 9.4-2. Sample Heat Exchange rs Six sample heat exchangers reduce the temperature of samples from pressurizer steam space, pressurizer liquid space, each steam generator and the reactor coolant to 130*F before samples reach the sample vessels and sample sink. The tube side of the heat exchangers is austenitic stainless steel, the shell side is carbon steel. O 1 l 9.4-5 i The inlet and outlet tube sides have socket-weld joints for connections to the high pressure sample lines. Connections to the component cooling water lines are socket-weld joints. The samples flow through the tube side and component cooling water f rom the Auxiliary Coolant System circulates through the shell side. Delay Coil The sample line contains a delay coil, consisting of coiled tubing, which has sufficient length to provide at least a 40 seconds sample transit time within the containment and an additional 20 seconds transit time from the reactor containment and an additional 20 seconds transit time from the reactor containment to the sampling hood. This allows for decay of short lived isotopes to a level that permits normal access to the sampling room. Sample Pressure Vessels The high pressure sample trains, the residual heat removal loop sample train and the volume control tank gas space sample train each contain sample pressure vessels which are used to obtain liquid or gas samples. The hot leg and the residual heat removal loop sample lines have a single sample pressure vessel in common. Integral isolation valves are furnished with the vessel and quick-disconnect coupling valves containing poppet-type check valves, are connected to nipples extending f rom the valves on each end. The vessesl, valves and couplings are austenitic stainless steel. Sample Sink The sample sink is located in a hooded enclosure which is equipped with an exhaust ve n t i l at o r. The work area around the sink and the enclosure is large enough for sample collection and storage for radiation monitoring equipment. The sink perimeter has a raised edge to contain any spilled liquid. 9.4-6 ' Piping and Fittings t All liquid and gas sample lines are austenitic stainless steel tubing and l are designed for high pressure service. With the exception of Lthe sample vessel quick-disconnect couplings and compression fittings at the sample sink, socket welded joints are used throughout the Sampling System. Lines are so' located as to protect them from accidental damage'during routine j ' operation and maintenance. l i l Valves Remotely operated stop valves are used to isolate all sample ~ points and to route sample fluid flow inside the reactor containment. Manual stop valves are provided for component isolation and flow path control at all normally accessible Sampling System locations. Manual throttle valves are provided to adjust the sample flow rate as indicated-on Figure 9.4-1. I ( Check valves prevent gross reverse flow of gas from the volume control tank into the sample sink. ) i All valves in the system are constructed of austenitic stainless steel 1 or equivalent corrosion resistant material. Isolation valves are provided outside the reactor containment which trip closed upon actuation of the containment isolation signal. l I l I i bV. 9.4-7 li.---.-.--.-.-. 9.4.3 SYSTEM EVALUATION Leakage Provisions Leakage of radioactive reactor coolant f rom this system within the containment is evaporated to the containment atmosphere and removed by the cooling coils of the Containment Air Recirculation and Cooling System. Leakage of radioactive material from the most likely places outside the containment is collected by placing the entire sampling station under a hood provided with an offgas vent to waste gas processing. Liquid leakage from the valves in the hood is drained to the chemical drain tank. Incident Control The system operates on an intermittent basis, and under administrative manual control. Malfunction Analysis To evaluate system safety, the failures or malfunctions are assumed concurrent with a loss-of-coolant accident, and the consequences analyzed. The results are presented in Table 9.4-2. From this evaluation it is concluded that l proper consideration has been given to station safety in the design of 1 the system. l 9 9.4-8 TABLE 9.4-L uJ SAMPLING SYSTEM CODE REQUIREMENTS Sample heat exchanger ASFE III*, Class C, tube side ASME VIII, shell side Sample pressure vessels ASME III, Class C Piping and valves USAS B31.l** kJ ASME III - American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Nuclear Vessels. USAS B31.1 - Code for Pressure Piping and special nuclear cases where applLcable. l O v l TABLE 9.4-2 SAMPLING SYSTEM COMPONENTS Sample Heat Exchanger General Number 6 Counter flow Type Design heat transfer rate (duty for 652.7*F 5 sat. steam to 127*F liquid), each, Btu /hr 2.12 x 10 Shell Design pressure, psig 150 Design temperature. *F 350 Component cooling water flow, gpm 40 P ressure loss at 40 gpm, psi 25 Operating cooling water temperature, in,

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Amendment 1 1 p TABLE 9.4-2 (Continued) U Sample Pressure Vessels Number, total 8 Volume, pressurizer steam sample, 2 supplied, ml 75 Volume, pressurizer liquid sample, 2 supplied, ml 75 Volume, reactor coolant hot leg sample, 2 supplied, ml 75 Volume, volume control tank sample, 2 supplied, ml 75 Design pressure, psig 2485 Design temperature, *F 680 Manual Throttic Valves Normal operating temperature, F 120-130 Design pressure, psig 2485 Body design temperature, *F 680 Piping g Liquid and gas sample line internal diameter, in. 0.245 k_ Design pressure, psig 2485 Design temperature, F 680 0 TABLE 9.4-3 MALRINCTION ANALYSIS OF SAMPLING SYSTEM Sample Chains Malfunction Comments and Consequences Pressurizer steam space Remote operated Two valves on each line sample, pressurizer sampling valve outside the containment liquid space sample, inside reactor are closed on containment accumulator or hot leg containment fails isolation signal. sample to close Any sample chain Sample line break Two valves on each line inside containment outside the containment upstream of remotely are closed on contain-operated valve ment isolation signal. Seal water is inserted between the two valves. 9 -.. - ~ _ ~. p%$ ICE OUT S DE c DN ta i m*4E hY C ONT Atha *E N T CUT $iDE lee S i DE SAMpteM, nocat s aMoL ie:G C OMPChE NT CCOU% NOOM WATE R ( Ac5) I rvswg cc CC To E U EN Sk 1121k1 T f T 90s sMs l i l g g orc 9C & 8 brc ps E S$vei2te } T T STE am & DACE Mm M a m e' 2 4 G56A 1568 (e c S) ab5 d %= T 53 he -t pS8 />s-ros. P&TDs. V.9 94 95 Dr Ts erE% asst e NOTEL L 2 scisi 2

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- 4ASTE 015P05AL 5YSTLM i I SHEIT I y DWG. 68 4.,72 I I 5HEIT 2 y DWG. e e *d9 r7 1 5 15 5AFETY INICTION SY5fLM e 'co" s DwG.een s7s t IV545 - 150 TAT 104 W ALWL 51AL nATER 5YSTE)4 e a 5 I t h4tco D=dr. G e90ao3 4,7 0 f DM l SAMPLING SYSTEM FIGURE 9.4-1 l l 9.5 FUEL HANDLING SYSTF31 (v\\ The Fuel Handling System provides a safe effective means of transporting and handling fuel from the time it reaches the plant in an unirradiated condition until it leaves the plant after post-irradiation cooling. The system is designed to minimize the possibility of mishandling or malopera-tions that cause fuel damage and potential fission product release. The Fuel Handling System consists basically of: a) The reactor cavity, which is flooded only during plant shutdown for refueling b) The spent fuel pit, which is full of water during and after the first refueling and is always accessible to operating personnel c) The Fuel Transfer System, consisting of an underwater conveyor that transports fuel assemblies between the reactor cavity and the f) spent fuel pit. v 9.5.1 DESIGN BASIS Prevention of Fuel Storage Criticality Criterion: Criticality in the new and spent fuel storage pits shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls. (GDC 66) During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concentration is maintained at not less than required to shut down the core to a k = 0.90 on the basis of all RCC gg assemblies inserted and a complete core installed. This shutdown margin maintains the core at a k = < 0.99, even if all control rods are withdrawn eff from the core. Weekly checks of the refueling water storage tank boron concentration ensure the proper shutdown margin. A check is made once l l (O per shift during core refueling operations and strict administrative controls ~~/ l are used to monitor any dilution of the refueling water in the reactor vessel. t l 1 l 9.5-1 l l O The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations. The new and spent fuel storage pits have accommodations as defined in Table 9.5-1. In addition, the spent fuel pit has an area set aside for accepting the spent fuel shipping casks. This operation is also done under water. Borated water is used to fill the spent fuel storage pit at a concentration to match that used in the reactor cavity and refueling canal during refueling operations. The fuel in the spent fuel and new fuel storage pits is stored vertically in an array with the sufficient center-to-center distance between assemblics to assure K 77 < 0.90 even if unborated water were to fill the space between the assemblies. Detailed instructions are available for use by refueling personnel. These instructions, the minimum operating conditions, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety. Fuel and Waste Storage Decay Heat Criterion: Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage facilities and to waste storage tanks that could result in radioactivity release which would result in undue risk to the health and safety of the public. (GDC 67) The refueling water provides reliable and adequate cooling medium for spent fuel transfer and heat removal from the spent fuel pit is provided by an auxiliary cooling system. Natural radiation and convection is adequate for cooling the holdup tanks. Fuel and Waste Storage Radiation Shielding Criterion: Adequate shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities. (CDC 68) 9.5-2 i Adequate shielding for radiation protection is provided during reactor refueling by conducting all spent fuel transfer and storage operations under water. This permits visual control of the operation at all times while maintaining low radiation levels, less than 2.5 mr/hr, for periodic occupancy of the area by operating personnel. Pit water level is indicated, and water removed from the pit must be pumped out since there are no gravity drains. Shielding is provided for waste handling and storage facilities to permit operation within requirements of 10 CFR 20. Camma radiation is continuously monitored in the auxiliary building. A high level signal is alarmed locally and is annunciated in the control room. Protection Against Radioactivity Release From Spent Fuel and Waste Storage Criterion: Provisions shall be made in the design of fuel and waste storage facilities such that no undue risk to the health and safety of the public could result from an accidental release of radio-j activity. (GDC 69) All fuel storage facilities are contained and equipment designed so that accidental releases of radioactivity directly to the atmosphere are monitored and do not exceed the guidelines of 10 CFR 100. The reactor cavity, refueling canal and spent fuel storage pit are reinforced concrete structures with a seam-welded stainless steel plate liner. These structures are designed to withstand the anticipated earthquake loadings as Class 1 structures so that the liner prevents leakage even in the event the reinforced concrete develops cracks. All operating areas in the fuel storage facilities contain ventilation systems. All vessels in the Waste Disposal System which are used for waste storage are Class I seismic design. Od 9.5-3 9.5.2 SYSTEM DESIGN AND OPERATION The reactor is refueled with equipment designed to handle the spent fuel under water from the time it leaves the reactor vessel until it is placed in a cask for shipment from the site. Boric acid is added to the water to ensure subcritical conditions during refueling. The Fuel Handling System may be generally divided into two areas: the reactor cavity which is flooded only during plant shutdown for refueling and the spent fuel pit which is full of water during and af ter the first refueling and is always accessible to operating personnel. These two areas are connected by the Fuel Transfer System consisting of an underwater conveyor that carries the fuel through an opening in the plant containment. The reactor cavity is flooded with borated water from the refueling water storage tank. In the reactor cavity, fuel is removed from the reactor vessel, transferred through the water and placed in the fuel transfer system by a manipulator crane. In the spent fuel pit the fuel is removed from the transfer system and placed in storage racks with a long manual tool suspended from an over-head hoist. After c sufficient decay period, the fuel is removed from storage and loaded into a shipping cask for removal from the site. Both the manipulator crane and the long handled tool can handle only one assembly at a time. New fuel assemblies are received and stored in racks in the new fuel storage area. New fuel is delivered to the reactor by lowering it into the spent fuel pit and taking it through the transfer system. The new fuel storage area is sized for storage of the fuel assemblies and control rods normally associated with the replacement of one-third of a core plus space for another one-third core. The fuel for the initial core loading is stored temporarily in the spent fuel storage pit. The pit is kept dry during this period. O 9.5-4 Major Structures Required for Fuel llandling Reactor Cavity The reactor cavity is a reinforced concrete structure that forms a pool above the reactor when it is filled with borated water for refueling. The cavity is filled to a depth that limits the radiation at the surface of the water to 2.5 milliroentgens per hour during fuel assembly transfer., The reactor vessel flange is scaled to 'the bottom of the reactor cavity by a bolted, gasketed seal ring which prevents leakage of refueling water j from the cavity. This seal is fastened and closed after reactor cooldown but prior to flooding the cavity for refueling operations. The cavity is large enough to prohl'de storage space for the reactor upper and lower internals, the control cluster drive shafts, and miscellaneous refueling tools. The floor and sides of the reactor cavity are lined with stainless steel. Refueling Canal The refueling canal is a passageway extending from the reactor cavity to the inside surface of the reactor containment. The canal is formed by two concrete shielding walls, which extend upward to the same elevation as the reactor cavity. The floor of the canal is at a lower elevation than the reactor cavity to provide the greater depth required for the fuel transfer system tipping device and the control cluster changing fixture located in the canal. The transfer tube enters the reactor con-tainment and protrudes through the end of the canal. Canal wall and floor linings are similar to those for the reactor cavity. L 9.5-5 Amendment 1 r-Refueling Water Storage Tank The normal duty of the refueling water storage tank is to supply borated water to the ref ueling canal for r efueling operat ions. In add ition, the tank provides horated water for delivery to the core following either a loss-of-coolant or a steam line rupture accident. This is described in Chapter 6. The capacity of the tank is based upon the requirement for filling the reactor cavity and refueling canal. The water in the tank is borated to a concentration which assures reactor i shutdown by at least 10% 6k/k when all RCC assemblies are inserted and when the reactor is cooled down for refueling. The tank design parameters are given in Chapter 6. Spent Fuel Storage Pit The spent fuel storage pit is designed for the underwater storage of spent fuel assemblies and control rods after their removal from the reactor. I The pit accommodations are listed in Table 9.5-1. Control rods are stored in the fuel assemblies. Spent fuel assemblies are handled by a long handled tool suspended from an overhead hoist and manipulated by an operator standing on the movable bridge over the pit. The spent fuel storage pit is constructed of reinforced concrete having thick walls and is Class I seismic design. The entire interior basin face and transfer canal is lined with stainless steel plate. llence, the probability of rupture of the pit is exceedingly low. O Amendment 1 9.5-6 () A storage rack is provided to hold spent fuel assemblies and is erected on the pit floor. Fuel assemblies are held in a square array, and placed in vertical cells. The racks are designed so that it is impossible to insert fuel assemblies in other than the prescribed locations, thereby ensuring the necessary spacing between assemblies. Control rod clusters are stored in place inside the spent fuel assemblies. New Fuel Storage New fuel assemblies and control rods are stored in a separate area whose location facilitates the unloading of new fuel assemblies or control rods from trucks. This storage vault is designed to hold new fuel assemblies in specially constructed racks and is utilized primarily for the storage of the replacement fuel assemblies required for cycled loading. The assemblies which make up the remaining part of the first core are stored in the spent fuel pit which otherwise remains unused until the time of first refueling. The new fuel assemblies are stored in racks arranged to space the fuel I) assemblies the same amount as for spent fuel. kJ Major Equipment Required for Fuel Handling Reactor Vessel Stud Tensioner l The stud tensioner is a hydraulically operated (oil as the working fluid) l l device provided to permit preloading and unloading of the reactor vessel closure studs at cold shutdown conditions. Stud tensioners were chosen in order to minimize the time required for the tensioning or unloading opera-tions. Three tensioners are provided and they are applied simultaneously l to three studs 120' apart. One hydraulic pum,)ing unit operates the tensioners which are hydraulically connected in parallel. The studs are tensioned to their operational load in two steps to prevent high stresses in the flange region and unequal loadings in the studs. Relief valves are provided on each tensioner to prevent overtensioning of the studs due to excessive pressure. ) 9.5-7 Charts indicating the stud elongation and load for a give oil pressure are included in the operating instructions. In addition, mic rometers lh are provided to measure the elongation of the studs after tensioning. Reactor Vessel Head Lif ting Device The reactor vessel head lif ting device consists of a welded and bolted structural steel frame with suitable rigging to enable the crane operator to li f t the head and store it during refueling operations. The lifting device is permanently attached to the reactor vessel head. Reactor Internals Lifting Device The reactor internals lif ting device is a fixture provided to remove the upper reactor internals package and to move it to a storage location in the refueling canal. The device is lowered onto the guide tube support plate of the internals and is manuallv bolted to the support plate by three bolts. The bolts are controlled by long torque tubes extending up to an operating platform on the lifting device. Bushings on the fixture lh engage guide studs mounted on the vessel flange to provide close guidance during removal and replacement of the internals package. This lifting device can also be used to remove the lower internals once the vessel has been cleared of all fuel assemblies. Manipulator Crane The manipulator crane is a rectilinear bridge and trolley crane with a vertical mast extending down into the refueling water. The bridge spans the reactor cavity and runs on rails set into the floor along the edge of the reactor cavity. The bridge and trolley motions are used to position the vertical mast over a fuel assembly in the core. A long tube with a pneumatic gripper on the end is lowered down out of the mast to grip the fuel assembly. The gripper tube is long enough so the upper end is still contained in the O 9.5-8 Amendment 1 g ?Q. mast when the gripper end contacts the fuel. A winch mounted on the trolley raises the gripper tube and fuel assembly up into the mast tube. The fuel is transported while inside the mast tube to its new position. The manipulator can lift only one fuel assembly at a time. All controls for the manipulator crane are mounted on a console on the trolley. The bridge is positioned on a coordinate system laid out on one rail. The electrical readout system on the console indicates the position of the bridge. The trolley is positioned with the aid of a scale on the bridge structure. The scale is read directly by the operator at the console. The drives for the bridge, trolley, and winch are variable speed and include a separate inching control on the winch. Electical interlocks and limit switches on the bridge and trolley drives protect the equipment. In an emergency, the bridge, trolley, and winch can be operated manually using a handwheel on the motor shaft. The suspended weight on the gripper tool is monitored by an electric load cell indicator mcunted on the control console. A load in excess of 110 per cent of a fuel assembly weight stops the winch drive frrm moving in the up direction. The gripper is interlocked through a weight sensing device and also a mechanical spring lock so that it cannot be opened when supporting a fuel assembly. Safety features are incorporated in the system as follows: a) Travel limit switches on the bridge and trolley drives b) Bridge, trolley, and winch drives which are mutually interlocked to prevent simultaneous operation of any two drives c) A position safety switch, the GRIPPER TUBE UP position switch, which prevents bridge and trolley main motor drive operation except when it is actuated. s. l 'a 9.5-9 d) An interlock which prevents the opening of a solenoid valve in the air line to the gripper except when zero suspended weight is indicated by a force gage. As back-up protection for this interlock, the mechanical weight actuated lock in the gripper prevents operation of the gripper under load even if air pressure is applied to the operating cylinder. e) The EXCESSIVE SUSPENDED WEIGitT switch, which opens the hoist drive { circuit in the up direction when the loading is excessive, f) An interlock on the hoist drive circuit in the up direction, which permits the hoist to be operated only when either the OPEN or CLOSED l indicating switch on the gripper is actuated, g) An interlock of the bridge and trolley drives, which prevents the bridge drive from traveling beyond the north edge of the core unless the trolley is aligned with the refueling canal centerline. The trolley drive is locked out when the bridge is beyond the north edge of the core. Suitable restraints are provided between the bridge and trolley structures and their respective rails to prevent derailing and the manipulator crane is designed to prevent disengagement of a fuel assembly from the gripper in the event of a maximum potential earthquake. Spent Fuel Pit Bridge The spent fuel pit bridge is a wheel-mounted walkway, spanning the spent fuel pit which carries an electric monorail hoist on an overhead structure. The fuel assemblies are moved within the spent fuel pit by means of a long handled tool suspended from the hoist. The hoist travel and tool length are designed to limit the maximum lift of a fuel assembly to a safe shielding depth. Amendment 1 9.5-10 m Fuel Trans f er System n ( l G' The fuel transfer system, shown in Figure 9.5-1, is an underwater air-motor driven conveyor car that runs on tracks extending f rom the refueling canal through the transfer tube and into the spent f uel pit. The conveyor car receives a fuel assembly in the vertical position f rom the manipulator crane. The fuel assembly is lowered to a horizontal position for passage through the tube, and then is raised to a vertical position in the spent f ue l pi t. During plant ope ra t i on, the conveyor car is stored in the refueling canal. The gate valve is closed and a blind flange is bolted on the transfer tube to seal the reactor containment. Rod Cluster Control Changing Fixture A fixture is mounted on the reactor cavity wall f or removing rod cluster control (RCC) elements from spent f uel assemblies and inserting them into r-(%) new fuel assemblics. The fixture consists of two main components; a guide ~./ tube mounted to the wall for containing and guiding the RCC element, and a wheel-mounted carriage for holding the fuel assemblies and positioning fuel assemblies under the guide tube. The guide tube contains a pneumatic gripper on a winch that grips the RCC element and lif ts it out of the f uel assembly. By repositioning the carriage, a new fuel assembly is brought under the guide tube and the gripper lowers the RCC element and releases it. The manipulator crane loads and removes the fuel assemblies into and out of the carriage. Ref ueling Sequence of Operat ion P re pa rat ion a) The reactor is shut down and cooled to ambient conditions, b) A radiation survey is made and the containment vessel is entered. [v) 9.5-11 Amendment 1 c) The control rod drive mechanism (CRDM) missile shield is removed to storage, d) CRDM cables and cooling air duct: are disconnected from the CRDM and removed to storage, c) Reactor vessel head insulation and instrument leads are removed. f) The reactor vessel head nuts are loosened with the hydraulic tensioners. g) The reactor vessel head studs are removed to storage. h) The canal drain holes are plugged. 1) Checkout of the fuel transfer device and manipulator crane is started. j) Guide studs are installed in three holes and the remainder of the stud holes are plugged. I k) The reactor vessel to cavity seal ring is bolted in place. 1) Final preparation of underwater lights and tools is made. Checkout of manipulator crane and fuel transfer system is completed. m) The fuel transfer tube flange is removed, n) The reactor vessel head is unseated and raised one foot with the plant crane. o) The reactor cavity is filled with water to the vessel flange. The water is pumped into the reactor cavity by the residual heat removal pumps f rom the refueling water storage tank through the reactor vessel. The normal Residual lleat Removal System inlet valves from the Reactor Coolant System are closed. p) The reactor vessel head is slowly lifted while water is pumped into the reactor cavity. The water level and vessel head are raised simultaneously keeping the water level just below the head. q) When the reactor is filled, the residual heat removal loop is restored to normal operation. r) The reactor vessel head-is taken to the storage pedestal. s) The control rod drive shafts are unlatched. t) The reactor vessel internals lifting rig is lowered into position by the plant crane and latched to the support plate. u) The reactor vessel flange protector ring is disengaged from the internals lifting rig. 9 .b.endment 1 9.5-12 = _ _. _. _.. _ _ - i i v) The reactor vessel internals are lifted out of the vessel and placed in the underwater storage rack. u) The core is now ready for refueling. i Refueling 1 l The refueling sequence is now started utilizing the manipulator crane. The sequence for fuel assemblies in non-control positions is as follows: i a) Spent fuel is removed from the core and placed into the fuel transfer system for removal to the spent fuel pit. ) i b) Partially spent fuel is rearranged in the core. j c) Replacement fuel assemblies are brought in from the spent fuel pit through the transfer system and loaded into the core. d) Whenever any fuel is added to the reactor core, a reciprocal curve of source neutron multiplication is recorded to verify the subcriti-cality of the core. 4 1 The refueling sequence is modified for fuel assemblies containing rod cluster control (RCC) elements, as required. If a transfer of the RCC elements between fuel assemblies is required, the assemblies are taken to the RCC 7 4 change fixture to exchange the RCC elements from one assembly to another. l Such an exchanger is required whenever a fuel assembly containing RCC elements is removed from the core and whenever a fuel assembly is placed 1 in or taken out of a control position during the refueling rearrangement. Reactor Reassembly 4 a) The fuel transfer car is parked and the fuel transfer tube isolation valve closed. i i b) The reactor vessel internals package is picked up by the plant crane and replaced in the vessel. The reactor vessel flange protector ring is attached to the internals lifting rig and the rig is removed to storage. j 9.5-13 Amendment 1 i _. _..., _ _ _ _. - _ - - -. ~ _. _ c) The control rod drive shafts are relatched to the RCC elements. l d) The manipulator crane is parked. g e) The old seal rings are removed from the reactor vessel head, the grooves cleaned and new rings installed, f) The reactor vessel head is picked up by the plant crane and positioned over the reactor vessel, g) The reactor vessel head is slowly lowered as the water level is lowered. The water level is lowered by opening a valve at the residual heat removal pump discharge and water is pumped from the reactor cavity into the refueling water storage tank until it reaches the vessel flange level. The normal residual heat removal line is closed, h) When the reactor vessel head is about one foot above the flange, the reactor cavity is completely drained. When the water in the reactor cavity reaches the vessel flange level, the valve at the residual heat removal pump is closed. The normal residual heat removal operation is restored and the remaining water in the reactor cavity is drained into the reactor coolant drain tank via the low point in the canal drain. The water is then pumped back into the refueling water storage I tank by the reactor coolant drain tank pumps. The flange surface is manually cleaned. l i) The reactor vessel head is seated. l l j) The guide studs are removed to t heir storage rack. The stud hole plugs are removed. l k) The head studs are replaced and retorqued. 1 1) The canal drain holes are unplugged and the fuel transfer tube flange is replaced. Amendmeat 1 9.5-14 i l I \\ l i ( ) m) Electrical leads and cooling air ducts are reconnected to the CRDM's. l j I j n) Vessel head insulation and instrumentation leads are replaced. i I o) The reactor vessel to cavity seal ring is unbolted. t p) A hydrostatic test is performed. i f I q) Control red drives are checked, l i l I r) The CRDM missile shield is picked up with the plant crane and replaced. \\ t I i i i s) Equipment access door is closed and scaled. l f 1 t) Pre-operational tests are performed. i i i l !e i l t i t t i l t I l f i t I l 4 l } 9.5-15 Amendment 1 [ l i i r 9.5.3 SYSTEM EVALUATION Underwater transfer of spent fuel provides essential case and corresponding f safety in handling operations. Water is an effective, economic and trans-parent radiation shield and a reliable cooling medium for removal of decay 5:ea t. liasic provisions to ensure the safety of refueling operations are: a) Gamna radiation levels in the containment and fuel storage areas are continuously monitored. These monitors provide an audible alarm at i the initiating detector indication an unsafe condition. Continuous monitoring of reactor neutron flux provides immediate indication and alarm in the control room of an abnormal core flux level. b) Violation of containment integrity is not permitted when the reactor vessel head is removed unless the shutdown margin is maintained greater then 10% t.h/k. c) Whenever any fuel is added to the reactor core, a reciprocal curve of lf scurce neutron multiplication is recorded to verify the suberiticality of the core. Incident Protection Direct communication between the control room cnd the refueling cavity i l manipulator crane is avuilable whenever changes in core geometry are taking ( place This provision allows the control room operator to inforn the manipulator l operator of any impending unsafe conditions detected from the main control board indicators during fuel movement. Malfunction Analysis An analysis is presented in Section 14 concerning damage to one complete outer row of fuel elements in an assembly, assumed as a conservative limit for evaluating environmental consequences of a fuel handling incident. L 9.5-16 Any suspected defective fuel assembly is placed in a failed fuel can and sealed to provide an isolated chamber for testing for the presence of fission products. The failed fuel cans are stainless steel cylinders with lids that can be bolted in place remotely. An internal gas space in the lid provides for water expansion and for collection and sampling of fission product gases. Various remotely operable quick-disconnect fittings permit connection of the can to sampling loops for continuous circulation through the can. If sampling shows the presence of fission products indicative of a cladding failure, the sampling lines are closed off by valves on the can and the encapsulated fuel assembly is removed to the spent fuel storage racks to await shipment. Design of the cans complies with federal regulation 10 CFR 71 so that the defective fuel can be stored and shipped while sealed in the failed fuel can. 9.5.4 TEST AND INSPECTION CAPABILITY Upon completion of core loading and installation of the reactor vessel l head, certain mechanical and electrical tests can be performed prior to initial l l criticality. The electrical wiring for the rod drive circuits, the rod position indicators, the reactor trip circuits, the in-core thermocouples, and the reactor vessel head water temperature thermocouple can be tested at the time of installation. The tests can be repeated on these electrical 1 items before initial plant operation. I 9.5-17 1, I i i I i i TAlil,E 9. 5-1 FUEL. IfANDI.ING DATA l ) l New Fuel Storage Pit l Core storage capacity 2/3 Equivalent fuel assemblies 105 l t Center-to-center spacing of assemblies, in. 21 l i l Maximum k with unborated water 0.90 eff I l Spent Fuel Storage Pit f Core storage capacity 1 1/3 i Equivalent fuel assemblies 236 Number of space accommodations for spent fuel shipping casks 1 Center-to-center spacing of assemblies, In. 21 I t Maximum k with unborated water 0.90 77 i MisceIlaneous Details Width of refueling canal, ft. 4 l 9 Wall thickness or spent fuel storage pit, ft. 3 to 6 Weight of fuel assembly with 1:CC (dry), Ib. %1580 Capacity of refueling water storage tank, gal. 350,000 f Minimum contents of refueling water storage j tank for Safety Injection or Spray System [ operability, gal. 285,000 Quantity of water required for refueling, gal. 285,000 I f S e i 1 l Amendment 1 ,m re w- -e . --- e e---> e wn mm ---e-=- i i i I i i UPENDING FRAME WINCH y b / REFUELlHG / WATER LEVEL f-N 6/ ll' REACTOR l CAVITY l REFUELING '.i l i CANAL l l Nd UPENDING CABLE FUEL ASS'Y 'u~. J CONTAINER dE 3..- UPENDING fl BLIND */2bo'. ~I ?>. FRAME ll \\ 5f}i;("[ FUEL FLANGE ,\\ y }*o CONVEYOR _ \\ S'Y ,S.(7 ".y b y l-d j ll II ll 11 OI J. - o 4,w..o. w.M t, me rnu.aw o .v

• REACTOR 6A

/ VESSEL CONVEYOR TRACKS AIR MOTOR i k j FUEL TRANSFEF i 1 m- - I I } PLArtT CONTAINER -UPENDING FRAME WINCH MOTOR DRIVEN PLATF0FM ICF - C FRAME w I WATER LEVEL FUEL ASS'Y CONTAINER (PERFORATED) FUEL SPENT FUEL ASS'Y STORAGE PIT / 3 CONVEYOR O }U g / CAR ,y \\ v K n=

1..n SPENT FUEL

,r STORAGE RACKS -N u l GATE l 5 VALVE NDING FRAME '/ / r'-l lc, ,,:,'x i ITT ll ll Il 11 1 L SECTION ,"AWgg,3g, m.. ...v cs m osms A-A FUEL TRANSFER TUBE i $YSTEM 1 1 I t FUEL TRANSFER SYSTEM FIG. 9.5-1 J 9.6 _F_ACILITY SERVICES O

V 9.6.1 FIRE PROTECTION SYSTEM Design Bases Criterion

The facility is designed so that the probability of fires and explosions and the potential consequences of such events does not result in undue risk to the health and safety of the public. Noncombustible and fire-resistant materials shall be used throughout the facility wherever necessary to preclude such risk, particularly in areas containing critical portions of the facility such as containment, control room, and components of engineered safety features.

(GDC 3) Fire prevention in all areas of the nuclear unit is provided by structure and component design which optimizes the containment of combustible materials and maintains exposed combustible materials below their ignition temperature in the design atmosphere. Fire control requires the capability to isolate or remove fuel from an igniting source, or to L) reduce the combustibles temperature below the ignition point, or to exclude the oxidant, and preferably, to provide a combination of the three basic control means. The latter two means are fulfilled by providing fixed or portable fire-fighting equipment of capacities proportional to the energy that might credibly be released by fire. This plant is designed on the basis of limiting the use of combustible materials in construction and of using fire-resistant materials to the greatest extent possible. The fire protection system has the design capability to extinguish any probable combination of simultaneous fires which might occur at the station. The fire protection system is shown schematically on Figure 9.6-1. System Design and Operation i j^ Normal fire protection is provided by a fixed water spray system, fire hydrants, hose lines, and portable extinguishers. 9.6-1 l A fixed automatic water spray and alarm system (manually tripped from the control room, or from the deluge valve area) with additional alarm indication in the control room provide protection for the turbine lube oil reservoir and purification system, and the hydrogen seal oil unit. Floors in the turbine area are pitched so that any oil spills will drain to proper drainage facilities, thus preventing the spread of an oil fire outside the turbine area. The unit auxiliary and station auxiliary transformers, main power trans-former, and the bulk oil storage area are protected by means of fixed automatic water spray and alarm systems, with additional indication in the control rocm. Readily accessible 1-1/2-inch rubber covered hose lines and continuous flow type hose reels are distributed throughout the plant so that all areas in the plant are within 20 feet of a spray nozzle when attached to not more than 100-foot lengths of hose. All nozzles are 1-1/2-inch spray nozzles pinned to the hose adapters with provision made to prevent the use of the straight stream position. Fire protection in the diesel generator room consists of the following: a) A fireproof wall between the two diesel engines to confine a fire to the space in which it originates. b) An automatic CO system with separate detectors in each diesel 2 compartment so that the compartment containing the fire will be the only one blanketed. c) A fire hose station with 50 feet of hose and nozzle is located adjacent to each compartment. Portable extinguishers are provided at strategic locations. Hand CO2 extinguishers are located in the control room, cable vaults and electrical equipment rooms. 9.6-2 'Fater Sunp1v The plant contains a ten-inch diameter fire water line loop-header This header is supplied uith a motor-driven fire pump and gas-driven emergenc fire pump and is continuously preocurized with a fire water booster pump A sufficient nunber of fire hydrants are installed in the loop to provide proper protection to the plant. Connection to the existing loop in the unit No. 1 area is cade with isolation valves which are normally locked closed. All fire protection piping which is not a Class I structure is designed to meet the Class I scismic criteria. An interior fire header is installed in the plant sized to delivery an adequate quantity of water throughout the plant at a sufficient pressure j lto service all outlets. For unit No. 2, the fire loop and header system l ' will normally be pressurized by the fire water boos ter pump of 50 gpm capacity at a 290 foot head. i 6 !9 j Upon reduction of header pressure in the fire water system loop for unit i l No. 2 an adequate supply of fire water from the cooling-vater lake is supplied to the system by an electric motor-driven pump of 2500 gpm capacity i at a 290 foot head. This pump is backed up by a separately located internal combustion engine-driven fire water pump of the same capacity Both an automatic starting with alarms in the control room. The combus tion engine local fuel supply capacity is designed for at least eight hours 6 I of operation. Power for the electrically-driven fire pump is taken from two different i sources to insure a source of power should one source fail. 2 l The individual fire pumps take their suction from the circulating water intake system on the clean side of the traveling screens between the

0 screens and the circulating water pumps.

i 9.6-3 Amendnent 6 s Fire protection it providr l t o the e::t erior plant areas by yard fire } ' hydrant <. The yard piping consists of an underground ten-inch header wh ich i n <;uppl ied f ron the fire water punps. Thi:. header it also connected to the existing fire protection synten for unit Io. 1 by t l inolation valve ubich are nor:. ally leched closed. Valved branches fron this underground synter supply interior fire protection systers in the enclosed sections of the plant, including the turbine building. [ l Sectionalizing valves in the yard piping systen are provided to pernit l partial pipe line 1: elation uithout interruption of service to the entire nysten during uaintenance or future extension of facilities. i I l A multiple opening, fire departnent connection is provided outside of i the building. i l The ntartup, unit auxiliary, and main transformers are protected by b separate uater npray systems, autcmatically operated by heat actuated device: The water supply for these systers is provided from the nain n ~ fire protection hcader in the turbine building. Each systen includes v, a deluge valve, heat actuated fire protection devices, spray nozzles and a fixed piping nysten. Since the reactor containnent has little combustible equipment, no special fire protection nyntems are required within the structure. The bearing oil syster., for the reactor coolant pumps are self-contained. In view of the nnall amount of combustible naterials in the containment, the naxinun surface tenperature of equipnent, and appropriate techniques for controlling fires, no renotely operated or special systems are deered nece: < ary. Portable dry chtmical extinguinher are provided for use 1 during nalntenance periods. The auxiliary building in also provided with dry-chcnical, portable fire extinguisher: in order to niniaine the spread of radioactive contamination l in the event of fire. 1 s* % rire protection for the fuel storage building con :ints of dry chemical ext inguisher: and hose connections. 1 N r-- 1y-.-,,i.,-------.n A nain fire prot ect ion hinde r in the turbine building supplie: the t urbine building pipiun loop and the deluge syatem-previonnly enumerated. The fire prevention design and facilitie:, located in the control reo are diocuused in Section 7-7.4. Alarn S, sten A fire warning nynt em in provided an an integral part of the autw.atic unter spray cyntom, to indicate an alarn in case of a fire in any protected area. Add i tional visual alarr:s are counted in the control room together I with the ability to start the motor driven fire water pienp. 6 Alarns pertinent to the two fire water pumps are on the control board. Rate of teuperature increase, fixed temperature and smoke detection alarn systens are provided for the relay room alarn systen in the area where all cablen fron the cable tunnel enter t. rays in the relay roca. This area in fully accencible for portable equipment or hose lines.

9.6.2 SERVICl

I?ATER SYSTEM Desfrn nanes The Service 1later System has been decIgned to provide redundant cooling water supplie with isolation valves to those components necensary for plant uafety either during normal operation or under accident conditior_. The Service L'ater Sys tem also supplies cooling vater t o various other heat loads in both the prinary and seconlary port ions of the plant. { Lake Robincon is the source of service water. 2 Ar. end. n t / . _ = _

4 4 =# O rM =4a m. r The systen is sired to ensure adequate heat removal based on highest u ! ( expected tenperatures of cooling water, maximum loadings and leakage { ( f allowances. I 1 System Design and Operati on l The Service Uater Systen flou diagran is shown in Figure 9.6-2. Four identical vertical, wet pit pumps, each having a capacity of 8000 gpn at 120 ft TDH, supply service water to two independent supply lines. l Either of the two supply lines can be used to provide cooling water to ( l ( all containment air recirculation cooling coils, the containment air f recirculation fan motor coolers, the turbine-driven auxiliary feedwater pump, and the diesel generators. Components in the nain turbine building can also be supplied with cooling water from either of the two j supply lines. Each of the supply lines provide water to a notor-driven auxiliary feedwater pump, an instrument air compressor and a component I I l (^\\f cooling heat exchanger. Vi i i Unter is drawn from the lake and passes through traveling screens. Two concrete walls separate the intake into three bays. A single service ( water pump is located in the tuo outside bays with the other two service { vater pumps located in the remaining bay. [ I The intake structure is designed as Seismic Class I, and is therefore* not subject to collapse under earthquake loading. The only part of the Service Water System which is not seismic Class I design is the section in the turbine building. This section of the system can be isolated [ by redundant renotely operated isolation valves in series located in the l i Class I auxiliary building. Two booster pumps (one on each of the supply lines Icading from the service water header in the auxiliary building to the containment air recirculation cooling units) provide adequate fjow of the service water b ~ -f ,m

i I through the cooling coils. Each of these pumps has a capacity of 3200 j () gpm at 100 f t Tull, v 1 The Service Water System is monitored and operated from the control room. 1 Isolation valves are incorporated in all service water lines penetrating the containment. All pump discharge valves, as well as supply header isolation valves are motor operated, and controlled remotely from the control room. i The containment ventilation cooling units are supplied by individual lines f rom the auxiliary building service water header. Each inlet line is provided with a motor operated shutof f valve and drain valve. Similarly, each discharge line from the cooler is provided with a motor operated shutoff valve. This allows each cooler to be isolated indi-vidua11y for leak testing of the system or to be drained and maintained open to the atmosphere during the integrated leak tests of the contain-ment. The ventilation cooler discharge lines will be monitored for radioactivity by routing a small bypass flow f rom each through a radiation (~) monitor. When excessive radioactivity in the effluent is indicated by an O alarm, each cooler discharge line would be monitored individually to locate the defective cooling coil. Ilowever, since the cooling coils and service water lines are completely closed inside the containment, no contaminated leakage is expected into these units. The service water system pressure at locations inside the containment is below the containment design pressure of 42 psig. l l l I C'} t> l 9.6-7 Amendment 1

._.__= 1 The motor-operated valves on the inlet and outlet service water lines for the fan coolers are equipped with indicating lights in the control i room. All service water piping, except for that downstream of the booster pumps, has mechanical joints. Joints on the service water piping downstream of the booster pumps are welded. The four service water pumps are connected to the diesel bus which can be supplied by either one of the two emergency diesels in the event of loss of all outside power. For this condit ion, the Service Water System is designed to supply cooling water to only the ? quired energency systems. Under the conditions of a concurrent loss-of-coolant accident and loss of offsite power, any two of four pumps using the emergency diesel power are capable of supplying the required cooling capacity. The emergency diesel-driven generator units are supplied with cooling water from a supply line on a continuous basis. Two modulating control valves, one on each supply line to the diesels, contro! flow during normal operation, and on a high containnent building pressure signal, both valves open fully to insure a sufficient supply of cooling water to each diesel. The inlet valving is arranged so that either of the two diesels can be served by either of the supply lines. During normal operation, the cooling loads are supplied by three of the four pumps available. Following a simultaneous loss of coolant accident anc' loss of offsite power, the cooling water requirenents for all four fan cooling units and the other essential loads can be supplied by any two of the four scr< ice water pumps during the inj ec t i on and long te rm recirculation phase of the Safety injection System (see Table 9.6-1). Service water to at least two containment air recircolation units is assured despite the passive failure of any pipe or valve body or active cemponent failure in the Service Water System from the service water i pumps to the containment air recirculation units themselves. c 9.6-8

Service water to at least one component cooling heat exchanger is assured despite the passive failure of any pipe or valve body or active system component failure in the system from the service water pumps to the heat exchangers themselves. Following a simultaneous loss of coolant accident and loss of offsite power, the component cooling heat exchangers are not needed during the injection phase. At the beginning of the recirculation phase at least one component cooling heat exchanger is placed in service. l The system is designed with two remotely operated isolation valves in the center of the supply header. Two service water pumps normally serve one of the two separate supply lines. The flow provided by two pumps is ensured even with a single passive failure of a pipe or valve body or an active system component failure in the supply system. i Tests and Inspections Each service water pump will undergo a hydrostatic test in the shop in which all wetted parts will be subjected to a hydrostatic pressure of one and one-half times the shut-off head of the pump. In addition, the normal capacity vs. head tests will be made on each pump. All valves in the service water system will undergo a shop hydrostatic test of three times the design pressure on the body and two and one-third times design pressure on the seat. Service water system design pressure is 150 psig. All service water piping, except for the piping downstream of the service water booster pumps, will be hydrostatically tested in the field of 225 l psig or one and one-half times design. Service water piping downstream from the service water booster pumps will be hydrostatically tested in the field at 150 psig or one and one-half times design. The welds in shop fabricated service water piping will be liquid penetrant or magnetic particle inspected in accordance with the ASME Boiler and Pressure Vessel Code, Section VIII. Electrical components of the service water system can be tested periodically. 9.6-9

TABLE 9.6-1 i SERVICE WATER REQUIREMENTS Normal Operation Flow Accident Flow (Number) Each (GPM) (Number) Each (GPM) Containment fan Containment fan cooling coils (4) 3,200 cooling coils (4) 3,200 Component cooling Component cooling heat heat exchanger (1) 10,000 exchanger (1) 10,000 Feedwater pump (2) 50 Auxiliary feedwater pump (1) 15 Air compressors (3) 26 Air compressors (2) 13 Auxiliary building Auxiliary building heating heating and ventilation 268 and ventilation 268 Condenser vacuum pumps 30 Emergency water supply to auxiliary feedwater pump 600 Equipment in turbine Diesel generators (2) 1,200 building 9,290 22,864 15,296 Normal Operation Accident Number of pumps required 3 2 Required pump capacity (each), gpm 7,621 7,648 ) Rated capacity per pump, gpm 8,000 8,000

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l G 190 199 SERVICE WATER SYSTEM FIGURE 9.6-2 a

O 9.7 EQUIPMENT AND SYSTEM DECONTAMINATION 9.7.1 DESIGN BASIS Activity outside the core could result from fission products from defective fuel elements, fission products from tramp uranium left on the cladding in small quantities during fabrication, products of n - y or n - p reactions on the water or impurities in the water, and activated corrosion products. Fission products in the reactor coolant associated with normal plant operation and tramp uranium are generally removed with the coolant or in subsequent flushing of the system to be decontaminated. The products of water activiation are not long lived and may be removed by natural decay during reactor cooldown and subsequent flushing procedures. Activated corrosion products are the primary source of the remaining activity. a The corrosion products contain radioisotopes from the reactor coolant which have been absorbed on or have diffused into the oxide film. The oxide film, essentially magnetite (Fe 0 ) with oxides of other metals 34 including Cr and Ni, can be removed by chemical means presently used in industry. Water from the primary coolant system and the spent fuel pit is the primary potential source of contamination outside of the corrosion film of the primary coolant system. The contamination could be spread by various means when access is required. Contact while working on primary system components could result in contamination of the equipment, tools and clothing of the personnel involved in the maintenance. Also, leakage from l the system during operation or spillage during maintenance could contaminate the immediate areas and could contribute to the contamination of the equipment, tools, and clothing. O(3 j 9.7-1 1

O 9.7.2 MET 110DS OF DECONTAMINATION Surface contaminates which are found on equipment in the primary system and the spent fuel pit that are in contact with the water are removed by conventional techniques of flushing and scrubbing as required. Tools are decontaminated by flushing and scrubbing since the contaminates are generally on the surface only of non-porous materials. Personnel and their clothing are decontaminated according to the standard health physics requirements. Those areas of the plant which are susceptable to spillage of radioactive fluids are painted with a sealant to facilitate decontamination that may be required. Generally washing and flushing of the surfaces are sufficient to remove any radioactivity present. The corrosion films generally are tightly adhering surface contaminates, and must be removed by chemical processes. The removal of these films is generally done with the aid of commercial vendors who provide both services and formulations. Since decontamination experience with reactors is continually being gained, specific procedures may change for each decontamination case. For corrosion films, the APAC (alkaline permanganate-diammonium citrate) treatment, or an organic acid variation of the APAC treatment is considred to be the most effective for removal. Portable components may be cleaned with a combination of chemical and ultrasonic methods if required. 1 O 9.7-2 4

9.7.3 DECONTAMINATION FACILITIES Decontamination facilities on site consist of a cask pit located adjacent to the spent fuel storage pit with a steel-lined floor. The outside surf aces of the shipping casks are decontaminated, if required, by using water detergent solutions, and manual scrubbing to the extent required. When the outside of the casks are decontaminated, the casks are removed by the auxiliary building crane and hauled away. For personnel, a decontamination shower and washroom is located adjacent 7 to the Radiation Control Area (RCA) locker room. Personnel decontamination f kits with instructions for their use are in the RCA locker room. i l i l 9.7-3 Amendment 1 l l .}}