ML11356A171

New York State (NYS) Revised Pre-Filed Evidentiary Hearing Exhibit NYSR0014G, UFSAR, Rev. 20 Indian Point Unit 2 (Submitted with License Renewal Application) (2007) (IP2 UFSAR, Rev. 20)

ML11356A171
Person / Time
Site:	Indian Point
Issue date:	12/22/2011
From:	- No Known Affiliation
To:	Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML11356A166	List: ML11356A167 ML11356A168 ML11356A169 ML11356A170 ML11356A171 ML11356A172
References
RAS 21611, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01
Download: ML11356A171 (188)

Text

{{#Wiki_filter:NYSR0014G Revised: December 22, 2011 IP2 FSAR UPDATE 9.3.2.4 Component Cooling Loop Components 9.3.2.4.1 Component Cooling Heat Exchangers The two component cooling heat exchangers are of the shell and straight tube type. Service water circulates through the tubes while component cooling water circulates through the shell side. Parameters are presented in Table 9.3-2. 9.3.2.4.2 Component Cooling Pumps The three component cooling pumps, which circulate component cooling water through the component cooling loop are horizontal, centrifugal units. The original pumps have casings made from cast iron (ASTM 48) based on the corrosion-erosion resistance and the ability to obtain sound castings. The material thickness indicates the high quality casting practice and the ability to withstand mechanical damage and, as such, is substantially overdesigned from a stress level standpoint. Carbon steel casing material (ASTM A216) has been evaluated and approved for replacement pumps. Parameters are presented in Table 9.3-2. 9.3.2.4.3 Auxiliary Cooling Water Pumps The component cooling pumps do not run during the injection phase of a loss of coolant accident with loss of offsite power. The CCW circulating water pumps provide cooling for the safety injection pump motors and the auxiliary component cooling water pumps provide cooling for the recirculation pumps during this phase, with heat absorbed by the thermal inertia of the component cooling system. Two motor-driven auxiliary component cooling water pumps are started during the injection phase to provide cooling flow to the recirculation pump motor coolers. A CCW circulating water pump is connected to the motor shaft of each safety injection pump to cool the safety injection pump bearings. Both the auxiliary component cooling water pumps and the CCW circulating water pumps are discussed in further detail in Section 6.2. 9.3.2.4.4 Component Cooling Surge Tank The component cooling surge tank, which accommodates changes in component cooling water volume is constructed of carbon steel. Parameters are presented in Table 9.3-2. In addition to piping connections, the tank has a flanged opening at the top for the addition of the chemical corrosion inhibitor to the component cooling loop. 9.3.2.4.5 Component Cooling Valves The valves used in the component cooling loop are standard commercial valves constructed of carbon steel with bronze or stainless steel trim. Since the component cooling water is not normally radioactive, special features to prevent leakage to the atmosphere are not provided. Self-actuated spring-loaded relief valves are provided for lines and components that could be pressurized beyond their design pressure by improper operation or malfunction. Cha pter 9, Page 42 of 99 Revision 20, 2006 OAG10000215_1129

IP2 FSAR UPDATE 9.3.2.4.6 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. The piping has been evaluated for the most limiting component cooling water temperatures under loss of coolant accident conditions and found to be acceptable 9.3.2.5 Residual Heat Removal Loop Components 9.3.2.5.1 Residual Heat Exchangers The two residual heat exchangers located within the containment are of the shell and U-tube type with the tubes welded 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. 9.3.2.5.2 Residual Heat Removal Pumps The two residual heat removal pumps are vertical, centrifugal units with special seals to prevent reactor coolant leakage to the atmosphere. All pump parts in contact with reactor coolant are austenitic stainless steel or equivalent corrosion resistant material. Cooling water is provided from the component cooling water system via flexible stainless steel hose. 9.3.2.5.3 Residual Heat Removal Valves The valves used in the residual heat removal loop are constructed of austenitic stainless steel or equivalent corrosion resistant material. 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. Two remotely-operated series stop valves at the inlet with a pressure interlock isolate the residual heat removal loop from the reactor coolant system. In addition the residual heat removal loop is isolated from the reactor coolant system by two series check valves and a remotely operated stop valve on the outlet lines. As depicted in Plant Drawing 227781 [Formerly UFSAR Figure 9.3-1, Sheet 1], overpressure protection in the residual heat removal loop is provided by a relief valve. 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. 9.3.2.5.4 Residual Heat Removal Piping All residual heat removal loop piping is austenitic stainless steel. The piping is welded with flanged connections at the pumps and at valve 741A. Cha pter 9, Page 43 of 99 Revision 20, 2006 OAG10000215_1130

IP2 FSAR UPDATE 9.3.2.5.5 Low Pressure Purification System The system is used to clean reactor coolant water when the primary system is depressurized during an outage. The system has a 100-gpm canned purification pump, a line that bypasses the volume control tank and charging pumps of the chemical and volume control system and associated valves as shown in Plant Drawing 208168 [Formerly UFSAR Figure 9.2-1, sheet 2]. The system is designed for 600 psi operation. 9.3.2.6 Spent Fuel Pit Loop Components 9.3.2.6.1 Spent Fuel Pit Heat 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. 9.3.2.6.2 Spent Fuel Pit Pumps One of two spent fuel pit pumps circulates water in the spent fuel pit cooling loop. The second pump is on standby. All wetted surfaces of the pumps are austenitic stainless steel, or equivalent corrosion resistant material. The pumps are operated manually from a local station. 9.3.2.6.3 Refueling Water Purification Pump When it is required to clean up the refueling water storage tank water, 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. 9.3.2.6.4 Spent Fuel Pit Filter The spent fuel pit filter removes particulate matter larger than 5 f.L from the spent fuel pit water. The filter cartridge is synthetic fiber and the vessel shell is austenitic stainless steel. 9.3.2.6.5 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. 9.3.2.6.6 Spent Fuel Pit Demineralizer The demineralizer is sized to pass 5-percent 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. In addition, it is used for purification of the refueling water storage tank water. 9.3.2.6.7 Spent Fuel Pit Skimmer (Deleted) Cha pter 9, Page 44 of 99 Revision 20, 2006 OAG10000215_1131

IP2 FSAR UPDATE 9.3.2.6.8 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 pit water are austenitic stainless steel or equivalent corrosion resistant material. 9.3.2.6.9 Spent 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 pump, heat exchanger, and filter to facilitate maintenance. 9.3.3 System Evaluation System performance has been evaluated for service water temperatures up to 95°F for normal operating modes, loss of offsite power and loss of coolant accident conditions. 9.3.3.1 Availability And Reliability 9.3.3.1.1 Component Cooling Loop For component cooling of the reactor coolant pumps, the excess letdown heat exchanger and the residual heat exchangers inside the containment, most of the piping, valves, and instrumentation are located outside the primary system concrete shield at an elevation above the water level in the bottom of the containment at postaccident conditions. (The exceptions are the cooling lines for the reactor coolant pumps and reactor supports, which can be secured following the accident.) In this location the systems in the containment are protected against credible missiles and from being flooded during postaccident operations. Also, this location provides shielding, which allows for maintenance and inspections to be performed during power operation. Outside the containment, the residual heat removal pumps, 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. Replacement of one pump or one heat exchanger is practicable while the other units are in service. The wetted surfaces of the component 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 and is housed in structures of the same classification. The components are designed to the codes given in Table 9.3-1 and the design pressures given in Table 9.3-2. In addition, the components are not subjected to any high pressures or stresses. Hence, a rupture or failure of the system is very unlikely. In the event of a loss-of-offsite power, the plant emergency diesel generators are immediately started and the component cooling water pumps are automatically loaded (in sequence) onto the emergency buses and started. Component cooling water to the reactor coolant pump thermal barrier heat exchanger is thus automatically restored to provide reactor coolant pump seal cooling and prevent seal failure for at least a 2-hr period following a loss-of-offsite power. Cha pter 9, Page 45 of 99 Revision 20, 2006 OAG10000215_1132

IP2 FSAR UPDATE An alternate power supply is also provided for one of the component cooling water pumps from the 13.8-kV normal offsite power through Unit 1 switchgear. If normal offsite power is not available, this pump can be energized using any of the three available gas turbines. During the recirculation phase following a loss-of-coolant accident, one of the three component cooling water pumps is required to deliver flow to the shell side of one of the residual heat exchangers. 9.3.3.1.2 Residual Heat 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 operable, safe operation is governed by Technical Specifications and safe shutdown of the plant is not affected; however, the time for cooldown is extended. The function of this equipment following a loss-of-coolant accident is discussed in Section 6.2. Alternate power can be supplied to one residual heat removal pump from the 13.8-kV normal outside power through Unit 1 switchgear. The time to cool down using the auxiliary safe shutdown components (1 RHR pump and heat exchanger, 1 component cooling pump, and 1 service water pump supplying flow to non-essential header) has been determined 1 . Conditions assumed were an initial core power of 102% of 3216 MW and service water temperature of 95°F. The analysis shows that the RCS can be brought to the cold shutdown mode (temperature less than 200°F) within 72 hours. 9.3.3.1.3 Spent Fuel Pit Cooling Loop This manually controlled loop may be shut down safely for time periods, as shown in Section 9.3.3.2.3, for maintenance or replacement of malfunctioning components. 9.3.3.2 Leakage Provisions 9.3.3.2.1 Component Cooling Loop Water leakage from piping, valves, and equipment in the system inside the containment is not considered to be generally detrimental unless the leakage exceeds the makeup capability. With respect to water leakage from piping, valves, and equipment outside the containment, welded construction is used where possible to minimize the possibility of leakage. 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, the sampling, or the auxiliary coolant systems, or a leak in the thermal barrier cooling coil for the reactor coolant pumps. Tube or coil leaks in components being cooled would be detected during normal plant operations by the leak detection system described in Sections 4.2.7 and 6.7. Such leaks are also detected at any time by a radiation monitor that samples the component cooling pump discharge downstream of the component cooling heat exchangers. 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 temperatures are stable. Cha pter 9, Page 46 of 99 Revision 20, 2006 OAG10000215_1133

IP2 FSAR UPDATE The component, which is leaking can be located by sequential isolation or inspection of equipment in the loop. If the leak is in one of the component cooling water heat exchangers it can be isolated and repaired within the limitations of the Technical Specifications. Overall leakage within the containment is limited to the value given in the Technical Specifications. Should a large tube-side to shell-side leak develop in a residual heat exchanger, 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 in the component cooling loop. If the leaking residual heat exchanger is not isolated from the component cooling loop before the inflow completely fills the surge tank, the relief valve on the surge tank lifts. The discharge of this relief valve is routed to the auxiliary building waste holdup tank. The severance 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 after 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 lines downstream from each reactor coolant pump protect the downstream piping and thermal barrier cooling coils from overpressure should cooling water be isolated to the thermal barrier coil when the reactor coolant pumps are still operating. The valves set pressure equals the design pressure of the reactor coolant system. The relief valves on the cooling water lines downstream from the sample, excess letdown, seal water, nonregenerative, 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 will allow the component cooling system to be a closed system under accident conditions, even at 100-percent of containment design pressure. The over-pressurization incident, which results from a passive failure of a reactor coolant pump seal cooling coil coincident with the failure of the high flow cutoff valve would result in a maximum component cooling water pressure of 185 psig. This pressure is allowed in the component cooling water system in accordance with its design code of 831.1, 1967 edition, par 102.2.4(2), addressing permissible variation and allowable stress value for a limited time. 9.3.3.2.2 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-coolant accident condition, water from the containment recirculation sump is recirculated through a loop inside the containment using the recirculation pumps and the residual heat exchangers. The residual heat removal pumps (which are located outside of the containment) serve as backup to the internal recirculation pumps. Chapter 9, Page 47 of 99 Revision 20, 2006 OAG10000215_1134

IP2 FSAR UPDATE Each of the two residual heat removal pumps is located in a shielded compartment with a floor drain. Piping conveys the drain water to a common sump. Two redundant sump pumps, each capable of handling the less than 50 gpm flow, which would result from the failure of a residual heat removal pump seal, discharge to the waste holdup tank. 9.3.3.2.3 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 bypass 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 mode would be from 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 180°F to 212°F with a full core in storage is at least 1.8 hr. Makeup water can be supplied within this time from the primary water storage tank, the refueling water storage tank and/or the fire protection system. The maximum required makeup rate for boiloff is 62 gpm (for a full core). Spent fuel pit temperature and level instrumentation would warn the operator of an impending loss of cooling. A local flow indicator is available to support operation of the Spent Fuel Pit Pumps. 9.3.3.3 Incident Control 9.3.3.3.1 Component Cooling Loop In the unlikely event of a pipe severance in the component cooling loop, backup is provided for postaccident heat removal by the containment fan coolers. Should the break occur outside the containment the leak could either be isolated by valving or the broken line could be repaired, depending on the location 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 one of the primary makeup water pumps. If the loop drains completely before the leakage is stopped, it can be refilled by a primary makeup water pump in less than 2 hr. If the break occurs inside the containment on a cooling water line to a reactor coolant pump, the leak can be isolated. Each of the cooling water supply lines to the reactor coolant pumps contains a check valve inside and a common remotely operated valve outside the containment wall. Each return line (combined oil coolers and combined thermal barrier coolers) has a common remotely operated valve outside the containment wall. The cooling water supply line to the excess letdown heat exchanger contains a check valve inside the containment wall and both supply and return lines have automatically isolated valves outside the containment wall. Cha pter 9, Page 48 of 99 Revision 20, 2006 OAG10000215_1135

IP2 FSAR UPDATE Flow indication is provided on the component cooling return lines from the safety injection and residual heat removal pumps. Each of the component cooling supply lines to the residual heat exchangers has a normally closed remotely-operated valve. If one of the valves fails to open upon a safety injection signal, the valve, which does open supplies a heat exchanger with sufficient cooling to remove the heat load during long term postaccident recirculation. The portion of the component cooling loop located outside the containment is considered to be a part of the reactor building isolation barrier. Except for the normally closed makeup line the primary water and city water emergency cooling lines, and equipment vent and drain lines, there are no direct connections between the cooling water and other systems. The primary water make-up and SIS/RHR Emergency Cooling Lines have manual valves that are normally closed unless required for their design function or testing. The city water emergency cooling line contains two normally closed isolation valves with an open tell-tale connection between them. The tell-tale prevents the potential contamination of a potable water source with component cooling water corrosion inhibitor chemicals. 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. 9.3.3.3.2 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 both valves independently 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 operating condition. 9.3.3.3.3 Spent Fuel Pit Cooling Loop The most serious failure of this loop is complete loss-of-water in the storage pool. To protect against this possibility, the spent fuel storage pool cooling connections enter near the water level so that the pool cannot be either gravity drained or inadvertently drained. For this same reason care is also exercised in the design and installation of the fuel transfer tube. The water in the spent fuel pit below the cooling loop connections could be removed by using a portable pump. Instrumentation is provided that will activate an alarm in the control room if the level in the spent fuel pit is at a preset level deviation above or below normal. Operators normally observe the level in the pool on a regular basis. 9.3.3.4 Malfunction Analysis A failure analysis of pumps, heat exchangers and valves is presented in Table 9.3-5. 9.3.4 Minimum Operating Conditions Minimum operating conditions for the auxiliary coolant system are specified in the Technical Specifications. Cha pter 9, Page 49 of 99 Revision 20, 2006 OAG10000215_1136

IP2 FSAR UPDATE 9.3.5 Tests and Inspections Tests and inspections of the auxiliary coolant system are specified in the Technical Specifications. The portion of the Residual Heat Removal System that is outside of containment, and not tested in accordance with Technical Specifications, shall be tested at least once each 24 months either by use in normal operation or by hydrostatically testing at 350 psig. The piping, between the residual heat removal pump suction and the containment isolation valves in the residual heat removal pump suction line from the containment sump, shall be hydrostatically tested once each 24 months at no less than 100 psig. Visual inspection of the system components shall be performed during these tests and any significant leakage shall be measured by collection and weighing or by another equivalent method. Repairs or isolation shall be made as required to maintain leakage from the Residual Heat Removal System components located outside of the containment per Technical Specification 5.5.2. REFERENCES FOR SECTION 9.3

1. Letter (with attachment, WCAP-12312) from S. Bram, Con Edison, to NRC,

Subject:

Application for License Amendment to Increase the Design Basis Inlet Temperature of the Service Water System, dated July 13, 1989. TABLE 9.3-1 Auxiliary Coolant System Code Requirements Component Component cooling heat exchangers ASME VIII Component cooling surge tank ASME VIII Component cooling loop piping and valves USAS B31.1 Residual heat exchangers side ASME VIII, shell side ASME III, Class C, tube Residual heat removal piping and valves USAS B31.1 Spent fuel pit filter ASME III, Class C Spent fuel heat exchanger side ASM E VIII, shell side ASME III, Class C, tube Spent fuel pit loop piping and valves USAS B31.1 Cha pter 9, Page 50 of 99 Revision 20, 2006 OAG10000215_1137

IP2 FSAR UPDATE TABLE 9.3-2 (Sheet 1 of 2) Component Cooling Loop Component Data Component Cooling Pumps Parameters Quantity 3 Type Horizontal centrifugal Rated capacity (each), gpm 3600 Rated head, ft H2 0 220 Motor horsepower, hp 250 Material (pump casing) Cast iron or Carbon steel Design pressure, psig 150 Design temperature, OF 200 Component Cooling Heat Exchangers Quantity 2 Type Shell and straight tube Design heat transfer, Btu/hr 31.4 x 106 Shell side (component cooling water) Operating inlet temperature, OF 100.1 Operating outlet temperature, OF 88.2 Design flow rate, Ib/hr 2.66 x 106 Design temperature, OF 200 Design pressure, psig 150 Material Aluminum-bronze Tube side (service water) Operating inlet temperature, OF 75 1 Operating outlet temperature, OF 81.9 Design flow rate, Ib/hr 4.55 x 106 Design temperature, OF 200 Design pressure, psig 150 Material Copper-nickel (90-10) Cha pter 9, Page 51 of 99 Revision 20, 2006 OAG10000215_1138

IP2 FSAR UPDATE TABLE 9.3-2 (Sheet 2 of 2) Component Cooling Loop Component Data Component Cooling Surge Tank Quantity Volume, gal 2000 Normal water volume, gal 1000 Design pressure, psig 100 Design temperature, OF 200 Construction material Carbon steel Relief valve setpoint, psig 52 Auxiliary Component Cooling Water Pumps Quantity 2 Type Vertical centrifugal Rated capacity, gpm 80 Rated head, ft H2 0 100 Motor horsepower, hp 5 Casing material Cast steel Design pressure, psig 150 Design temperature, OF 200 CCW Circulating Water Pumps (Safety Injection Pumps) Quantity 3 Type Centrifugal Rated capacity, gpm 20 Rated head, ft H2 0 115 Casing material Stainless Steel Design pressure, psig 225 Design temperature, OF 200 Component Cooling Loop Piping and Valves Design pressure, psig 150 Design temperature, OF 200 Notes:

1. Operation is acceptable up to 95°F.

Cha pter 9, Page 52 of 99 Revision 20, 2006 OAG10000215_1139

IP2 FSAR UPDATE TABLE 9.3-3 (Sheet 1 of 2) Residual Heat Removal Loop Component Data Reactor coolant temperature at startup of heat removal, OF 350 Time to cool reactor coolant system from 1 350°F to 200°F, hr (all equipment operational) 48 1 350°F to 140°F, hr (all equipment operational) 113.6 Refueling water storage temperature, OF Ambient Decay heat generation at 10 hrs after shutdown condition, Btu/hr 85.6 X 106 1 Reactor cavity fill time, hr 1 Reactor cavity drain time, hr 4 Residual Heat Removal Pumps Quantity 2 Type Vertical centrifugal Rated capacity (each), gpm 3000 Rated head, ft H2 0 350 Motor, hp 400 Material Stainless steel Design pressure, psig 600 Design temperature, OF 400 Cha pter 9, Page 53 of 99 Revision 20, 2006 OAG10000215_1140

IP2 FSAR UPDATE TABLE 9.3-3 (Sheet 2 of 2) Residual Heat Removal Loop Component Data Residual Heat Exchangers Quantity 2 Type Shell and U-tube Design heat transfer (each), Btu/hr 30.8 x 106 Shell side (component cooling water) Operating inlet temperature, OF 88.3 Operating outlet temperature, OF 100.8 Design flow rate, Ib/hr 2.46 x 106 Design temperature, OF 200 Design pressure, psig 150 Material Carbon steel Tube side (reactor coolant) Operating inlet temperature, OF 135 Operating outlet temperature, OF 113.5 Design flow rate, Ib/hr 1.44 x 106 Design temperature, OF 400 Design pressure, psig 600 Material Stainless steel Residual Heat Removal Loop Piping and Valves

1. Isolated loop Design pressure, psig 600 Design temperature, OF 400

2. Loop Isolation Design pressure, psig 2485 Design temperature, OF 650 Notes:

1. Aligned to RHR system at 20 hours after shutdown, 95°F Service Water Cha pter 9, Page 54 of 99 Revision 20, 2006 OAG10000215_1141

IP2 FSAR UPDATE TABLE 9.3-4 (Sheet 1 of 3) Spent Fuel Cooling Loop Component Data Spent fuel pit heat exchanger Quantity 1 Type Shell and U-tube Design heat transfer, Btu/hrs1 7.96 x 106 Shell side (component cooling water) Normal operating inlet temperature, °F1 100 Normal operating outlet temperature, °F1 105.7 Design flow rate, Ib/hr 1.4 x 106 Design temperature, OF 200 Design pressure, psig 150 Material Carbon steel Tube side (spent fuel pit water) Normal operating inlet temperature, °F1 120 Normal operating outlet temperature, °F1 112.8 Design flow rate, Ib/hr 1.1 x 106 Design temperature, OF 200 Design pressure, psig 150 Material Stainless steel Spent fuel pit skimmer pump Retired in place Refueling water purification pump Quantity 1 Type Horizontal centrifugal Rated capacity, gpm 100 Rated head, ft H2 0 150 Design pressure, psig 150 Design temperature, OF 200 Material Stainless steel Chapter 9, Page 55 of 99 Revision 20, 2006 OAG10000215_1142

IP2 FSAR UPDATE TABLE 9.3-4 (Sheet 2 of 3) Spent Fuel Cooling Loop Component Data Spent fuel pit cooling loop piping and valves Design pressure, psig 150 Design temperature, OF 200 Spent fuel pit skimmer loop piping and valves Retired in place Refueling water purification loop piping and valves Design pressure, psig 150 Design temperature, OF 200 Spent fuel pit pump Quantity 2 Type Horizontal centrifugal Material Stainless steel Rated capacity, gpm 2,300 Rated head, ft H2 0 125 Motor, hp 100 Design pressure, psig 150 Design temperature, OF 200 Spent fuel storage pool Volume fe Typical Boron concentration, ppm boron min Tech Spec Boron concentration, ppm boron min Spent fuel pit filter Quantity 1 Internal design pressure of housing, psig 200 Design temperature, OF 250 Rated flow, gpm 100 Maximum differential pressure across filter 5 element at rated flow (clean cartridge), psi Maximum differential pressure across filter 20 element prior to removing, psi Filtration requirement 98-percent retention of particles down to 5 /-l Cha pter 9, Page 56 of 99 Revision 20, 2006 OAG10000215_1143

IP2 FSAR UPDATE TABLE 9.3-4 (Sheet 3 of 3) Spent Fuel Cooling Loop Component Data Spent fuel pit strainer Quantity 1 Rated flow, gpm 2,300 Maximum differential pressure across the strainer 1 element at rated flow (clean), psi Perforation, in. ~0.2 Spent fuel pit demineralizer Quantity 1 Type Flushable Design pressure, psig 200 Design temperature, OF 250 Flow rate, gpm 100 Resin volume, fe 30 Spent fuel pit skimmers Deleted Spent fuel pit skimmer strainer Retired in place Spent fuel pit skimmer filter Retired in place Notes: 1.0riginal design. Cha pter 9, Page 57 of 99 Revision 20, 2006 OAG10000215_1144

IP2 FSAR UPDATE TABLE 9.3-5 Failure Analysis of Pumps, Heat Exchangers, and Valves Components Malfunction Comments and Consequences

1. Component cooling Rupture of a The casing and shell are designed for 150 psi and water pumps pump casing 200°F, which exceeds maximum operating conditions. Pump is inspectable and protected against credible missiles. Rupture is not considered credible. However, each unit is isolable. Two of the three pumps are needed to carry total pumping load.

2. Component cooling Pump fails to One operating pump supplies sufficient cooling water pumps start water for emergency core cooling during recirculation.

3. Component cooling Manual valve This is prevented by pre-startup and operational water pumps on a pump checks. Further, during normal operation, each suction line pump is checked on a periodic basis, which would show if a valve is closed.

4. Component cooling Normally open The valve is checked open during periodic water valve valve operation of the pumps during normal operation.

5. Component cooling Tube or shell Rupture is considered improbable because of low heat exchanger rupture operating pressures. Each unit is isolable. Both units may be required to carry total heat load for normal operation at 95°F Service Water.

6. Demineralized water Sticks open The check valve is backed up by the manually-makeup line check operated valve. Manual valve is normally closed.

valve

7. Component cooling Left open This is prevented by pre-startup and operational heat exchanger vent checks. On the operating unit such a situation is or drain valve readily assessed by makeup requirements to system. On the second unit such a situation is ascertained during periodic testing.

8. Component cooling Fails to open There is one valve on each outlet line from each water outlet valve to heat exchanger. One heat exchanger remains in residual heat service and provides adequate heat removal exchanger during long-term recirculation. During normal operation the cooldown time is extended.

Cha pter 9, Page 58 of 99 Revision 20, 2006 OAG10000215_1145

IP2 FSAR UPDATE 9.3 FIGURES Figure No. Title Figure 9.3-1 Sh. 1 Auxiliary Coolant System - Flow Diagram, Sheet 1, Replaced with Plant Drawing 227781 Figure 9.3-1 Sh. 2 Auxiliary Coolant System - Flow Diagram, Sheet 2, Replaced with Plant Drawing 9321-2720 Figure 9.3-1 Sh. 3 Auxiliary Coolant System - Flow Diagram, Sheet 3, Replaced with Plant Drawing 251783 9.4 SAMPLING SYSTEM 9.4.1 Design Basis 9.4.1.1 Performance Requirements This system provides for analysis of liquid and gaseous samples obtained during normal operation and postaccident conditions. The containment atmosphere postaccident sampling system is discussed in Sections 6.8.2.2 and 6.8.2.3. Sampling of the primary and secondary coolant systems is discussed below. Primary samples include the following:

1. Reactor coolant system hot-leg loops 21 and 23.

2. Pressurizer steam space and liquid space.

3. Residual heat removal loop.

4. Safety injection system accumulators 21, 22, 23, and 24.

5. Recirculation pumps 21 and 22 discharge.

6. Chemical and volume control system letdown lines at demineralizer inlet and outlet.

7. Holdup tanks.

8. CVCS holdup tank transfer pumps discharge.

9. Chemical drain pump 21 discharge.

10. Waste evaporator feed pump 21 discharge.

11. High-radiation sampling system collection tank discharge.

These samples are obtained at the high-radiation sampling system panels and evaluated by the online analysis systems or manual analysis. Secondary samples are obtained from the secondary sampling system, which is separate from the high-radiation sampling system. Postaccident sampling of the primary system is an extension of the use of the high-radiation sampling system for routine sampling. The NRC approved 3 the removal of the requirements and administrative controls for the postaccident sampling system from the Technical Specifications and accepted regulatory commitments to maintain:

1. contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere; Cha pter 9, Page 59 of 99 Revision 20, 2006 OAG10000215_1146

IP2 FSAR UPDATE

2. the capability for classifying fuel damage events at the Alert threshold within the Emergency Plan Implementing Procedures (EPIPs); and

3. the capability for monitoring radioactive iodines that have been released to offsite environs within the EPIPs.

Sampling system discharge flows are limited under normal and anticipated fault conditions (malfunctions or failure) to preclude any fission product releases beyond the limits of 10 CFR

20. Shielding has been provided to minimize operator exposure to any radiation during the sampling procedures.

The primary coolant sampling system was evaluated by the NRC against the criteria in Item II.B.3 of NUREG-0737 and found acceptable. 1,2 9.4.1.2 Design Characteristics The design characteristics of the high-radiation sampling system include the following:

1. Control of background radiation and operator exposure to radiation.

2. Rapid sampling and analysis.

3. Sampling and transfer of undiluted samples.

In addition, the system is capable of the following:

1. The system can be used for both routine and postaccident sampling and has the capability to obtain an undiluted reactor coolant sample under accident conditions for transport offsite for independent analyses.

2. Inline measurement of the reactor coolant specific conductivity, pH, and dissolved oxygen, hydrogen, chlorides, and boron under both routine and postaccident conditions.

3. Additional sample connections are available for more flexibility in selecting sample points; redundant sample connections allow for further expansion if needed to ensure sample acquisition under postaccident conditions.

4. Methods for cooling and depressurizing all high temperature-high pressure fluids for gas sampling and inline analyses.

5. Specially designed shielded transfer casks minimize operator radiation exposure when obtaining diluted and undiluted liquid samples. A small aliquot of reactor coolant system liquid or containment air samples is transferred as required to designated areas for analyses using a holder to maintain adequate distance and provide low operator radiation exposure.

Flow paths are also provided for boron concentration, and isotopic inline analysis. Sampling of other process coolants, such as tanks in the waste disposal system, is accomplished locally. Equipment for sampling secondary and nonradioactive fluids is separated from the equipment provided for reactor coolant samples. Leakage and drainage resulting from Cha pter 9, Page 60 of 99 Revision 20, 2006 OAG10000215_1147

IP2 FSAR UPDATE the sampling operations are collected and drained to tanks located in the waste disposal system. 9.4.1.3 Primary Sampling Two types of samples are obtained by the primary sampling system: high temperature-high pressure reactor coolant system and steam generator blowdown samples, which originate inside the reactor containment, and low temperature-low pressure samples from the chemical and volume control and auxiliary coolant systems. 9.4.1.3.1 High Pressure-High Temperature Samples A sample connection is provided from each of the following:

1. The pressurizer steam space.

2. The pressurizer liquid space.

3. Hot legs of loops 21 and 23.

4. Blowdown from each steam generator.

9.4.1.3.2 Low Pressure-Low Temperature Samples A sample connection is provided from each of the following:

1. The letdown demineralizers inlet and outlet header.

2. The residual heat removal loop, just downstream of the heat exchangers.

3. The volume control tank gas space.

4. The (safety injection system) accumulators 21, 22, 23, and 24.

5. Recirculation pumps 21 and 22 discharge.

9.4.1.4 Expected Operating Temperatures The high pressure-high temperature samples and the residual heat removal loop samples leaving the sample heat exchangers are cooled to minimize the generation of radioactive aerosols. 9.4.1.5 Secondary Sampling The secondary sampling system provides continuous sampling and analysis of the plant's secondary systems. This ensures the maintenance of proper water chemistry conditions in the secondary side piping and equipment. A sample connection is provided from each of the following:

1. Each of the four main steam lines.

2. Each condenser hotwell section.

3. Condensate pump discharge.

4. Outlet of the 26 feedwater heaters.

5. Drains collection tank inlet from primary water.

Chapter 9, Page 61 of 99 Revision 20, 2006 OAG10000215_1148

IP2 FSAR UPDATE 9.4.1.6 Codes and Standards System code requirements are given in Table 9.4-1. In addition, the high radiation sampling system was designed and installed to meet the provisions of NUREG-0737. These provisions include the following:

1. Provide postaccident sampling and analysis capability. The combined time for sampling and analysis is 3 hr or less from the time a decision is made to take a sample.

2. Provide capability to obtain and analyze a sample without radiation exposure to any individual exceeding the criteria of GDC 19 (10 CFR Part 50, Appendix A).

3. Provide means of measuring pH, conductivity, chlorides, dissolved hydrogen, dissolved oxygen, inline isotopic analysis, and boron analysis.

4. Provide means of safely obtaining pressurized samples, depressurized samples, and diluted and undiluted samples for laboratory analysis.

5. Safely store the sampled fluid until its disposal is determined.

6. Provide means of diverting to the containment the stored sample fluid.

7. Provide the capability to use the system on a continuous day-to-day basis.

8. Provide the capability to flush the sampled lines.

9. Provide the capability of drawing samples even when the reactor coolant system is depressurized (reactor coolant system, residual heat removal, and recirculation lines).

9.4.2 System Design and Operation 9.4.2.1 Primary Sampling System The primary sampling system consists of the high-radiation sampling system, which is shown in Plant Drawing 9321-2745. The high-radiation sampling system provides the representative samples for inline monitoring and laboratory analysis under normal or postaccident conditions. Analytical 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 concentration, evaluating fuel element integrity and mixed-bed demineralizer performance, and regulating additions of corrosion controlling chemicals to the systems. The high-radiation sampling system can be operated intermittently or on a continuous basis. Samples can be withdrawn under conditions ranging from full power to cold shutdown to postaccident conditions. Reactor coolant liquid, [Note - For postaccident conditions, the reactor coolant liquid sample may be taken from the reactor coolant system hot legs 21 and 23 or the recirculation pump Cha pter 9, Page 62 of 99 Revision 20, 2006 OAG10000215_1149

IP2 FSAR UPDATE discharge or the residual heat removal loop.], which is normally inaccessible or which requires frequent sampling, is sampled by means of permanently installed piping leading to either the inline isotopic analyzer, or the liquid sampling panel located in the sentry high-radiation sampling system room (formerly the waste evaporator room) at the 80-ft level of the primary auxiliary building. A seismic Class I concrete wall surrounds the high-radiation sampling system panel and a combination of lead shot and steel composes the shielding for the panel itself. These materials provide the shielding necessary to allow access to the high-radiation sampling system during and following accident conditions. Most of the primary sampling equipment is located in the sentry high-radiation sampling system room although some of it is located in other areas such as the pipe trench area of the 51-ft elevation and the 68-ft elevation of the mezzanine within the primary auxiliary building. The delay coils and remotely operated valves on the reactor coolant system hot-leg sample lines are located inside the reactor containment. Containment isolation valves are located immediately outside containment and are controlled, in an accident, from either the central control room or the sample system valve control panel. A line from the makeup water system has been installed to provide water for flushing of the sample lines. Reactor coolant hot-leg liquid, pressurizer liquid, and pressurizer steam samples originating inside the reactor containment flow through separate sample lines to the sentry liquid sampling panel. The samples pass through the reactor containment to the auxiliary building where they are cooled (pressurizer steam samples recondensed and cooled) in the sample heat exchangers. The reactor coolant samples are then routed through the inline isotopic analyzer where specific nuclides are identified. All samples then go to the sentry high-radiation sampling system panel. This consists of a liquid sampling panel, which is subdivided into a reactor coolant module, which includes the capability for dissolved gas analysis, a demineralizer sampling module, and a radwaste sampling module. Associated with the liquid sampling panel is the chemical analysis panel. These modules are discussed in detail later. The chemical analysis panel analytical results register on the chemical monitor panel in the sentry high-radiation sampling system room. There are remote readouts for the boron analysis in the radio chemistry laboratory and nuclear service building 1. Reactor coolant and demineralizer samples from the chemical and volume control system are depressurized and degasified in the reactor coolant module and demineralizer modules, respectively. From there they are sent to the chemical analysis panel, which can analyze for hydrogen, oxygen, chlorides, pH, and conductivity. Provisions are included in the primary sampling system to allow each sample to be purged through the sample lines and panel to ensure that representative samples are obtained. The sample volumes are routed to the high-radiation sampling system collection tank or chemical drain tank after completion of the task. The reactor coolant sample originating from the residual heat removal loop of the auxiliary coolant system has a motor-operated isolation valve located close to the sample source outside the containment. The sample line from this source intersects the sample line coming from the hot leg at a point ahead of the sample heat exchanger. This sample then follows the same flow Cha pter 9, Page 63 of 99 Revision 20, 2006 OAG10000215_1150

IP2 FSAR UPDATE path as that described for the reactor coolant system hot-leg samples. See Plant Drawings 9321-2745 and 227178 [Formerly UFSAR Figure 9.4-1]. A steam-generator sample line is taken from each blowdown line outside containment. The sample lines are routed to the blowdown tank room adjacent to the primary auxiliary building where the samples are cooled and are then passed through a radiation monitor as well as routed to cell 2 of the support facilities. These sample streams pass through additional local heat exchangers in cell 2 and subsequently through radiation, pH, conductivity, and chloride monitors. The sample waste under normal conditions is then routed to the river. Samples not suitable for release are diverted to the support facilities contaminated drain tank and waste disposal system. In the event of primary-to-secondary coolant leakage in one or more of the steam generators, the blowdown will be diverted to the support facilities secondary boiler blowdown purification system flash tank. This system cools the blowdown and either stores it in the support facilities waste collection tanks or purifies it. The purification process consists of filtering and demineralizing the blowdown. The filters will remove undissolved material of 25 /-l or greater. Mixed-bed demineralizers, which utilize cation and anion resin, remove isotopic cations and anions as well as nonradioactive chemical species. The effluents of the demineralizers are monitored and the specific activity is recorded on a two-pen recorder in the support facilities chemical system control room. 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. 9.4.2.1.1 Components A summary of principal component data is given in Table 9.4-2. 9.4.2.1.1.1 Sample Heat Exchangers Ten sample heat exchangers reduce the temperature of samples from the pressurizer steam space, the pressurizer liquid space, each steam generator, and the reactor coolant system liquid before samples reach the sample vessels and the sample sink. The tube side of the heat exchangers is austenitic stainless steel, the shell side is carbon steel. 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 from the auxiliary coolant system circulates through the shell side. 9.4.2.1.1.2 Delay Coil and Restriction Orifice The high-pressure reactor coolant sample line, which contains a delay coil consisting of coiled tubing and a restriction orifice, will provide at least 40 sec sample transit time within the containment and an additional 20 sec transit time from the reactor containment to the sampling station. This allows for decay of short-lived isotopes to a level that permits normal access to the sampling room. Cha pter 9, Page 64 of 99 Revision 20, 2006 OAG10000215_1151

IP2 FSAR UPDATE 9.4.2.1.2 Liquid Sampling Panel The liquid sampling panel valves and components are arranged in three modules installed in a common panel shield:

1. Module 1 - Reactor coolant sampling module (RC).

2. Module 2 - Demineralizer sampling module (OM).

3. Module 3 - Radwaste sampling module (RW).

Sample tubing and components are mounted behind the shielded panel within a plenum. Any gas leakage is vented to a local prefilter and HEPA filters and finally to existing ventilation ducts containing charcoal filters. A vessel at the bottom of the plenum collects any minor liquid leakage, which is pumped to radwaste. This provides containment of radioactivity during sampling operations. As a safety measure, the liquid sampling panel has a hooded splash box to contain any accidental liquid spill or gaseous release during normal sampling of pressurized reactor coolant or liquid grab sampling from all three modules. Each system can be purged through the sample lines and panel to ensure representative samples will be obtained. The purge flow can be directed back to the containment to chemical drain tank 21 and the associated waste disposal system or to the shielded high-radiation sampling system waste collection tank. All lines of the liquid sampling panel can be flushed with demineralized water following each sampling operation. Provisions are included for eliminating water from the gas expansion vessel and drying the gas lines of the panel. Included as part of the liquid sampling panel are carts, shielded casks, and other specialized equipment for sampling under accident conditions. After sampling, the shielded casks can be removed to provide samples for backup in-house analyses or stored for subsequent offsite analysis. The viewing window and sampling compartment for alignment of the cart and cask are located in the lower right section of the liquid sampling panel. The types of samples that can be obtained from the liquid sampling panel during normal operation are:

1. Undiluted, depressurized liquid grab samples from the reactor coolant, demineralizer, and radwaste modules.

2. Removable 75-ml pressurized liquid samples from the reactor coolant module, for subsequent analysis in the chemical analysis panel.

3. Inline pressurized liquid samples from the reactor coolant module.

Additional functions of the liquid sampling panel during normal operation include:

1. Purging of lines with sample to ensure representative samples will be obtained.

2. Reduction of pressure and control of flow rate of the primary coolant as it flows to the chemical analysis panel.

3. Routing of stripped gas from the pressurized liquid sample to the chemical analysis panel gas chromatograph.

Cha pter 9, Page 65 of 99 Revision 20, 2006 OAG10000215_1152

IP2 FSAR UPDATE The types of samples that can be obtained from the liquid sampling panel during accident conditions are:

1. Undiluted liquid samples from the reactor coolant and radwaste modules in cart/cask.

2. Diluted (1 to 1000) liquid samples from the reactor coolant and radwaste modules in cart/cask.

3. Inline pressurized liquid sample from the reactor coolant module.

4. Diluted (1 to 15,000) stripped gas sample from the reactor coolant pressurized liquid sample.

Additional functions of the liquid sampling panel during accident conditions include:

1. Purging of lines with sample to ensure representative samples will be obtained.

2. Capability for back-flushing the inline filters of the reactor coolant and radwaste modules.

3. Capability for flushing all lines and sample bottles on an individual section basis to control radiation levels as necessary.

4. Routing of stripped gas from the pressurized reactor coolant sample to the chemical analysis panel gas chromatograph.

5. Reduction of pressure and control of flow rate of the primary coolant as it flows to the chemical analysis panel.

9.4.2.1.3 Isotopic Analyzer Isotopic analyses may be performed on the following samples obtained from the liquid sampling panel:

1. Pressurized reactor coolant sample (gas and liquid) in removable sample flask for normal sampling.

2. Undiluted grab samples from the reactor coolant, demineralizer and radwaste modules of the liquid sampling panel for normal sampling.

3. Diluted liquid samples from the reactor coolant and radwaste modules of the liquid sampling panel for accident sampling.

4. Undiluted liquid samples from the reactor coolant and radwaste modules of the liquid sampling panel for offsite analyses during accident conditions.

5. Diluted stripped gas samples from the reactor coolant module of the liquid sampling panel for accident sampling.

Isotopic analyses are performed using a Ge(Li) detector gamma spectroscopy system using previously established counting geometries. 9.4.2.1.4 Boron Analyzer Backup boron analyses may be performed on the following samples from the liquid sampling panel for analysis in the onsite laboratory.

1. Undiluted grab samples from the reactor coolant, demineralizer, and radwaste modules of the liquid sampling panel for normal sampling.

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IP2 FSAR UPDATE

2. Diluted liquid samples obtained from the liquid sampling panel shielded cart/cask from the reactor coolant and radwaste modules of the liquid sampling panel for accident sampling.

The primary sampling system provides that both the routine and accident sample analyses of undiluted samples are performed online using a mannitol titration boron analyzer. It periodically samples an identical line from the chemical analysis panel from which conductivity, dissolved oxygen, and pH are measured. The range of the accident procedure is from 0.5 to 6.0 ppm boron. The estimated precision at the 95-percent confidence level is +13-percent, -3.3-percent at the 2-ppm boron level. 9.4.2.1.5 Cart and Casks The cart and casks associated with the liquid sampling panel are used for removal of samples obtained from the reactor coolant and radwaste modules during accident conditions. The shielded casks are mounted on a cart, which moves the cask into position for sampling from the liquid sampling panel. The carts permit access to the casks to obtain a laboratory sample or for storage in a remote area upon completion of the sampling operation. 9.4.2.1.6 Chemical Analysis Panel The chemical analysis panel receives an undiluted liquid sample stream and stripped gas from the reactor coolant module of the liquid sampling panel. The chemical analysis panel is divided into three major sections:

1. Flow control and cell section, consisting of the appropriate tubing, valves, and sensing elements.

2. Chromatograph section, containing two ion chromatographs for liquid analysis and a gas chromatograph for gas analysis.

3. Calibration section, where the solutions required for calibrating the pH, specific conductivity, and dissolved oxygen monitors, and ion chromatograph are available for use.

Valves, tubing, cells, and transmitters are mounted on the back of the panel shield within a plenum. Any gas leakage from the liquid sampling panel, chemical analysis panel, or boron analyzer is vented to a pre-filter and HEPA filter and subsequently to the primary auxiliary building ventilation ducts containing charcoal filters. Drip pans are mounted beneath the flow control/cell section and ion chromatograph to collect any minor leakage and to protect other equipment. The ion and gas chromatographic equipment, which contacts radioactive liquid or gas is mounted behind the shield to minimize operator exposure during the sampling/analysis process. The chemical analysis panel gas chromatograph and ion chromatograph sampling operations are controlled from the chemical monitor panel. The chemical analysis panel provides the capability for inline determination of the pH, specific conductivity, dissolved oxygen, temperature, and chloride content of a reactor coolant sample flowing from the liquid sampling panel during normal or accident conditions. In addition, the gas chromatograph permits determination of the hydrogen concentration of the stripped gas from Chapter 9, Page 67 of 99 Revision 20, 2006 OAG10000215_1154

IP2 FSAR UPDATE the reactor coolant. Remote readouts of the instrumentation measuring the chemical parameters are on the chemical monitor panel. Flushing lines are provided to flush all internal liquid and gas panel lines, and sample lines connecting the chemical analysis panel to the liquid sampling panel. Reagent calibration tanks may be flushed with nitrogen. 9.4.2.1.7 Chemical Monitor Panel The chemical monitor panel is an auxiliary recorder/monitor panel, which contains the indicating and recording equipment for the cells and analyzers, which are mounted in the chemical analysis panel. The panel permits the operator to work with and observe indicating and recording equipment from a remote location, to reduce exposure under accident conditions. Prior to sampling, the operator performs instrument zero and calibration adjustments of the monitors and evaluates chromatograms during the process of calibrating the instrumentation. This is accomplished prior to the chemical analysis panel receiving reactor coolant liquid or stripped gas from the liquid sampling panel. The monitor indicator readings include conductivity, pH, and dissolved oxygen measurements. The dissolved oxygen monitor, for low level routine analysis, includes a meter indication while the oxygen/temperature monitor provides a recording during accident conditions for higher levels of dissolved oxygen. A three-channel recorder records the chromatograms from the ion and gas chromatograph. The ion chromatogram is evaluated to determine the chloride concentration in the reactor coolant. Dissolved hydrogen concentration in the reactor coolant is determined by evaluating the gas chromatogram. Control of the sample injection to the chromatographs is provided by controls on the front of the panel. 9.4.2.1.8 High Radiation Sampling System Collection Tank After analysis, the liquid and gaseous samples are routed to the high radiation sampling system collection tank. A nitrogen line to the tank provides a pressurized noncombustible atmosphere. A vent line is provided for the venting of excess gases. There is also a line running back to the high radiation sampling system panel for analysis of the contents of the tank. If the level of radiation is too high following an accident the samples in the tank can be routed back to containment; otherwise the samples will be routed to the chemical drain tank. 9.4.2.1.8.1 Chemical Drain Tank During normal operation the liquid and gaseous samples are routed to the chemical drain tank. This tank is then pumped to the Unit 2 waste holdup tank. A sample can be directed to the radwaste module, if analysis is required prior to transfer. 9.4.2.1.8.2 Piping and Fittings All liquid and gas sample lines are austenitic stainless steel tubing and are designed for high pressure service. With the exception of the sample pressure vessel quick-disconnect couplings and compression fittings at the sample sink and at the liquid sampling panel sump and pump connections, socket-welded joints are used throughout the sampling system. Lines are so located as to protect them from accidental damage during routine operation and maintenance. Cha pter 9, Page 68 of 99 Revision 20, 2006 OAG10000215_1155

IP2 FSAR UPDATE 9.4.2.1.8.3 Valves Remotely-operated stop valves are used to isolate all sample points and to route sample fluid flow inside the reactor containment. Manual or motor-operated 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. All valves in the system are constructed of austenitic stainless steel or equivalent corrosion resistant material. Isolation valves are provided outside the reactor containment, which trip closed upon generation of the containment isolation signal. 9.4.2.2 Secondary Sampling System The secondary sampling system is shown in Plant Drawing 9321-7020 [Formerly UFSAR Figure 9.4-2]. This system is used to determine steam and condensate/feedwater quality and chemical addition requirements. The steam and water analysis station is located in the turbine building. It consists of a local panel where various controls, alarms, recorders and indicators are located; racks for the sample coolers and analyzers; and, a sample sink where grab samples can be obtained. The main steam can be analyzed for various additives, contaminants or isotopes. The condensate and/or feedwater can be analyzed for salinity, pH, conductivity, dissolved oxygen, residual hydrazine, and various additives and contaminants. High salinity is indicative of river water leakage into the condenser or makeup carryover. Conductivity is measured to determine the degree of possible dissolved solids entrainment into the systems. Because of its corrosive effects, dissolved oxygen is measured and recorded and used as a guide in determining the proper amount of hydrazine to be added to the condensate. The six individual condenser hotwells are provided with specific conductivity analyzers. These instruments are used to identify the specific condenser sextant that has salt water ingress. 9.4.3 System Evaluation 9.4.3.1 Availability and Reliability Automatic action is not required of the sampling system during an emergency or to prevent an emergency condition. In a postaccident situation, after proper safeguards are instituted between the central control room and the liquid sample control panels 1 and 2, permission could be granted for operators to activate specific valve combinations on these panels. This would permit selective use of the inline isotopic analyzer and associated high radiation sampling system liquid sampling panel. Cha pter 9, Page 69 of 99 Revision 20, 2006 OAG10000215_1156

IP2 FSAR UPDATE 9.4.3.2 Leakage Provisions Leakage of radioactive reactor coolant from this system within the containment is evaporated to the containment atmosphere and removed by the cooling coils of the containment fan coolers. Leakage of radioactive material from the most likely places outside the containment is collected by running a ventilation line from the high radiation sampling system panel to an existing exhaust duct in the old sampling room. This duct has a diffuser with a damper. During normal operation, air from the room is taken in through the diffuser; during accident conditions the damper is closed and air is taken into the ventilation system only from the high radiation sampling system panel ventilation. The gases from the panel pass through a pre-filter, a HEPA filter, a 450 cfm exhaust fan, and then into the existing ventilation system, which contains a charcoal filter. This system is seismic Class I. Liquid leakage from the sentry liquid sampling panel, chemical analysis panel, and boron analyzer valves within the common vented system is drained to the liquid sampling panel sump and pumped to either the chemical drain tank 21 or the high radiation sampling system collection tank. 9.4.3.3 Incident Control The system operates on a continuous basis for isotopic analysis, conductivity, dissolved oxygen, pH, and during steam-generator blowdown sampling. The inline dissolved hydrogen, chloride and boron concentrations can be obtained periodically from the sentry high radiation sampling system room. 9.4.3.4 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 9.4-3. From this evaluation it is concluded that proper consideration has been given to station safety in the design of the system. 9.4.3.5 High Radiation Sampling System Evaluation The high radiation sampling system is an independent system to provide information to plant operators. It is separate from other safety and non-safety systems. It is located in an area served by the primary auxiliary building ventilation system. The high radiation sampling system has the capability of handling both low and high radiation sampling without exceeding personnel exposure guidelines. Sufficient shielding is provided on the high radiation sampling system liquid sampling panel to allow personnel access for postaccident sampling. REFERENCES FOR SECTION 9.4

1. Letter from S. A. Varga, NRC, to J. D. O'Toole, Con Edison,

Subject:

Postaccident Sampling at the Indian Point Unit 2, Safety Evaluation Report, dated June 28, 1984.

2. Letter from S. A. Varga, NRC, to J. D. O'Toole, Con Edison,

Subject:

Postaccident Sampling at the Indian Point Unit 2, Safety Evaluation Report, dated December 12, 1984. Cha pter 9, Page 70 of 99 Revision 20, 2006 OAG10000215_1157

IP2 FSAR UPDATE

3. Letter from P.D Milano, NRC to M.R. Kansler, Entergy,

Subject:

Indian Point Nuclear Generating Unit No. 2 - Amendment Re: Deletion of Technical Specifications for the Post Accident Sampling System (PASS) using the Consolidated Line Item Improvement Process (TAC No. MB2991). Dated January 30, 2002. TABLE 9.4-1 Sampling System Code Requirements Code Sample heat exchanger ASME 111,1 Class C, tube side ASM E VIII, shell side Piping and valves USAS B31.12 Notes:

1. ASME III - American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Nuclear Vessels.

2. USAS B31.1 - Code for pressure piping and special nuclear cases where applicable.

TABLE 9.4-2 Primary Sampling System Components Sample Heat Exchanger Number 10 Type Coiled tube in shell Heat exchanged (each), 2.14 x 105 Btu/hr Surface area (each), ft2 3.73 Shell Design pressure, psig 150 Design temperature, of 350 Component cooling water flow (nominal), gpm 17 Flow,lb/hr 20,000 Component cooling water inlet temperature, of 105 outlet temperature, of 130 Material Carbon steel Tubes Tube diameter in., 0.0. 3/8 Design pressure, psig 2485 Design temperature, of 680 Flow,lb/hr 209 Inlet temperature (saturated steam), of 653 Outlet temperature, of 127 Material Austenitic stainless steel Cha pter 9, Page 71 of 99 Revision 20, 2006 OAG10000215_1158

IP2 FSAR UPDATE TABLE 9.4-3 Malfunction Analysis of Sampling System Sample Chains Malfunction Comments and Consequences Pressurizer steam Remotely operated Diaphragm or motor-operated valve outside space sample, sampling valve inside the reactor containment closes automatically pressurizer liquid reactor containment on containment isolation signal or by operator space sample, or hot- fails to close. action from the control room. leg sample. Any sample train. Sample line break Same as above. inside containment. 9.4 FIGURES Figure No. Title Figure 9.4-1 Sh. 1 Primary Sampling System - Flow Diagram, Sheet 1, Replaced with Plant Drawing 9321-2745 Figure 9.4-1 Sh. 2 Primary Sampling System - Flow Diagram, Sheet 2, Replaced with Plant Drawing 227178 Figure 9.4-2 Secondary Sampling System - Flow Diagram, Replaced with Plant Drawing 9321-7020 9.5 FUEL HANDLING SYSTEM 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 postirradiation cooling. The system is designed to minimize the possibility of mishandling or maloperations that could cause fuel damage and potential fission product release. The fuel handling system consists basically of:

1. The reactor cavity, which is flooded only during plant shutdown for refueling.

2. The spent fuel pit, which is kept full of water and is always accessible to operating personnel.

3. The fuel transfer system, consisting of an underwater conveyor that carries the fuel through an opening between the areas listed in the discussion of plant containment.

Chapter 9, Page 72 of 99 Revision 20, 2006 OAG10000215_1159

IP2 FSAR UPDATE 9.5.1 Design Basis 9.5.1.1 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, boron concentration is maintained at not less than that required to shutdown the core to a keff = 0.95. Periodic checks of refueling water boron concentration ensure the proper shutdown margin. 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 fuel racks and spent fuel storage pit have accommodations as defined in Table 9.5-1. In addition, the spent fuel pit has the required spent fuel shipping area. The spent fuel storage pit is filled with borated water at a concentration to match that used in the reactor cavity and refueling canal during refueling. The fuel is stored vertically in an array with sufficient center-to-center distance between assemblies to assure Keff <1.9E:!"en if unborated water was used to fill the pit and ::;0.95 when filled with water borated ;:::: 4QQQ ppm boron. Limits on enrichment and burnup of fuel in the spent fuel storage pit are given in the Technical Specifications. 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. In lieu of maintaining a monitoring system capable of detecting a criticality as described in 10CFR70.24, IP2 has chosen to comply with the seven criteria of 10CFR50.68(b). 9.5.1.2 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 a reliable and adequate cooling medium for spent fuel transfer and heat removal from the spent fuel pit. Overall this is provided by an auxiliary cooling system. Natural radiation and convection is adequate for cooling the holdup tanks. 9.5.1.3 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. (GDC 68) Adequate shielding for radiation protection is provided during reactor refueling by conducting all spent fuel transfer and storage operations underwater. This permits visual control of the operation at all times while maintaining radiation levels as low as reasonably achievable for the period of 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 Chapter 9, Page 73 of 99 Revision 20, 2006 OAG10000215_1160

IP2 FSAR UPDATE provided for waste handling and storage facilities to permit operation within requirements of 10 CFR 20. Gamma radiation is continuously monitored in the auxiliary building. A high level signal is alarmed locally and is annunciated in the control room. 9.5.1.4 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 radioactivity. (GDC 69) All fuel and waste storage facilities are contained and equipment designed so that accidental releases of radioactivity directly to the atmosphere are monitored and do not exceed the applicable limits. 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 seismic Class I structures so that the liner prevents leakage even in the event the reinforced concrete develops cracks. All vessels in the waste disposal system, which are used for waste storage are designed as seismic Class I equipment. 9.5.2 System Design and Operation The reactor is refueled with equipment designed to handle the spent fuel underwater 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 kept full of water 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 a fuel transfer tube, which penetrates 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 overhead hoist. 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 inserts normally associated with the replacement of one-third of a core. Chapter 9, Page 74 of 99 Revision 20, 2006 OAG10000215_1161

IP2 FSAR UPDATE 9.5.2.1 Major Structures Required for Fuel Handling 9.5.2.1.1 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 during fuel assembly transfer. The reactor vessel flange is sealed to the bottom of the reactor cavity by a Presray seal, which prevents leakage of refueling water from the cavity. This seal is installed after reactor cooldown but prior to flooding the cavity for refueling operations. Following refueling operations and prior to return to power, this seal is removed. The cavity is large enough to provide 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. 9.5.2.1.2 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 containment and protrudes through the end of the canal. Canal wall and floor linings are similar to those for the reactor cavity. 9.5.2.1.3 Refueling Water Storage Tank The normal duty of the refueling water storage tank is to supply borated water to the refueling canal and reactor cavity for refueling operations. In addition, the tank provides borated 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 minimum volume of water and the minimum amount of boration of the water in the refueling water storage tank is defined in the Technical Specifications. Heating is provided to maintain the temperature above freezing. The tank design parameters are given in Chapter 6. 9.5.2.1.4 Spent Fuel Storage Pit The spent fuel storage pit is designed for the underwater storage of spent fuel assemblies, failed fuel cans if required, and control rods after their removal from the reactor. The pit accommodations are listed in Table 9.5-1. 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 and is seismic Class I design. This structure was analyzed to determine compliance with ACI-31S(77), and SRP 3.S of NUREG-OSOO. In addition to the mechanical loadings, the pool structure was also analyzed to Chapter 9, Page 75 of 99 Revision 20, 2006 OAG10000215_1162

IP2 FSAR UPDATE include the temperature induced loadings. For this purpose, the thermal boundary conditions were conservatively specified as 180°F pool water temperature and OaF outside ambient. The thermal moments computed by the finite element analyses were combined with those due to mechanical loads. The results of these analyses show that there are large margins between the factored loads and corresponding design strengths. The pit is lined with a leak-proof stainless steel liner. All welds were vacuum-box tested during construction to assure a leaktight membrane. The effect of a thermal gradient would be to compress the liner. A review of the stress factors resulting from the finite element analyses demonstrates that an adequate design margin exists for the spent fuel pit liner walls and basemat. Storage racks are provided to hold spent fuel assemblies and are erected on the pit floor. Fuel assemblies are held in a square array, and placed in vertical cells. Fuel inserts are stored in place inside the spent fuel assemblies. 9.5.2.1.5 Storage Rack High density fuel storage racks have been designed to provide a maximum storage capacity of 1376 locations. The arrangement of the fuel storage racks in the spent fuel storage pool is shown in Figure 9.5-2. The fuel storage rack arrangement contains two types of storage rack arrays. Region 1, consisting of three racks that use the flux trap design, can store 269 new or irradiated fuel assemblies. The flux trap design used in Region 1 uses spacer plates in the axial direction to separate the cells. Boraflex absorber panels are held in place adjacent to each side of the cell by picture-frame sheathing. The spacer plates between cells form a flux trap between the boraflex absorber panels. Region 1 racks were originally designed to store new fuel with enrichments up to 5.0 wlo U235 . Region 1 is subdivided into two regions (Region 1-1 and Region 1-2): Region 1-1 is assumed to have sustained a 100% loss of Boraflex (i.e., none of the boraflex in the panels is assumed to be available). Technical Specifications show the fuel assembly criteria that will meet the requirements of 10 CFR 50.68(b)(4) if stored in Region 1-1. The maximum initial enrichment that can be stored in Region 1-1 with no burnup is 1.95 wlo U235 . Region 1-2 is assumed to have sustained a 50% loss of Boraflex (i.e., 50% of the boraflex in the panels is assumed to be available). Region 1-2 can accommodate unirradiated fuel up to 5.0 wlo U235 assuming the presence of a minimum number of IFBA rods. The maximum initial enrichment that can be stored in Region 1-2 when there are no IFBA rods is 4.50 wlo U235 . Each Region I storage cell, as shown in Figure 9.5-3, is a square box with an 8.75 inch inside dimension. Boraflex poison is held in place adjacent to each side of the box by "picture-frame" sheathing. The boxes are assembled into racks with an east-west pitch of 10.765 inches (center-to-center) and a north-south pitch of 10.545 inches, as shown in Figure 9.5-4. A 1/2 inch thick base plate is provided at the bottom of the rack. Adjustable leg supports are welded to the underside of the base plate. A six-inch diameter flow hole is provided in the base plate for each storage cell, and two one-inch holes are provided for cross flow at the bottom of each cell. Chapter 9, Page 76 of 99 Revision 20, 2006 OAG10000215_1163

IP2 FSAR UPDATE Region 2, consisting of nine racks that use the egg-crate design, can store 1105 fuel assemblies and two failed fuel canisters. Region 2 racks consist of boxes welded into a "checkerboard" array with a storage location in each square. One Boraflex absorber panel is held to one side of each cell wall by picture frame sheathing. Region 2 racks were originally designed to store fuel assemblies that have undergone significant burnup (e.g., ::;5.0 wlo U235 with a burnup of at least 40,900 megawatt days per metric ton (MWD/MT)) or fuel assemblies with a relatively low initial enrichment and low burnup (i.e., ::;1.764 wlo U235 at zero burnup). Region 2 is subdivided into two regions (Region 2-1 and Region 2-2): Region 2-1 is assumed to have sustained a 100% loss of Boraflex (i.e., none of the boraflex in the panels is assumed to be available). The maximum initial enrichment that can be stored in Region 2-1 with no burnup is 1.06 wlo U235 . Region 2-2 is assumed to have sustained only a 30% loss of Boraflex (i.e., 70% of the boraflex in the panels is assumed to be available). "Peripheral" Cells, consisting of six select cells along the SFP west wall in Region 2-2, may be used to store fuel that meets the requirements for storage in any other location in the SFP. Cells between and adjacent to the "peripheral" cells may be filled with fuel assemblies that meet the requirements for storage in Region 2-2). The two prematurely discharged fuel assemblies meet the requirements and qualify for storage in the "peripheral" cells. The storage racks are positioned on the floor so that adequate clearances are provided between racks and between the rack and pool structure to avoid impacting of the sliding racks during seismic events. The horizontal seismic loads transmitted from the rack structure to the pool floor are only those associated with friction between the rack structure and the pool liner. The vertical deadweight and seismic loads are transmitted directly to the pool floor by the support feet. 9.5.2.1.6 New Fuel Storage New fuel assemblies and control rods are stored in a separate area with a location that 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. Criticality analyses have been performed assuming the fully loaded racks are flooded with water. The analyses demonstrated that Keff is less than 0.95 for fuel with Integral Fuel Burnable Absorbers (lFBA) and enrichments in the range 4.5 wlo to 5.0 w/o. Keff is also less than 0.95 for fuel enriched to 4.5% or less with no absorbers. 9.5.2.2 Major Equipment Required for Fuel Handling 9.5.2.2.1 Reactor Vessel Stud Tensioner Stud tensioners are used to make up the head closure joint and during this process all studs are stretch tested to more than nominal working loads at every refueling. The stud tensioner is a hydraulically-operated (oil as the working fluid) device provided to permit preloading and unloading of the reactor vessel closure studs at cold shutdown conditions. A Cha pter 9, Page 77 of 99 Revision 20, 2006 OAG10000215_1164

IP2 FSAR UPDATE stud tensioner was chosen in order to minimize the time required for the tensioning or unloading operations. Three tensioners are provided and they are normally applied simultaneously to 0 three studs 120 apart. One hydraulic pumping unit operates the tensioners, which are hydraulically connected in parallel. The studs are tensioned to their operational load in a number of steps to prevent high stresses in the flange region and unequal loadings in the studs. A relief addition, micrometers are provided to measure the elongation of the studs after tensioning. 9.5.2.2.2 Reactor Vessel Head Lifting Device The reactor vessel head lifting device consists of a welded and bolted structural steel frame with suitable rigging to enable the crane operator to lift the head and store it during refueling operations. The lifting device is permanently attached to the reactor vessel head. 9.5.2.2.3 Reactor Internals Lifting Device The reactor internals lifting device is a fixture providing the means to grip the top of the reactor internals package and to transfer the lifting load to the crane. The device is lowered onto the guide tube support plate of the internals and is manually 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 engage guide studs mounted on the vessel flange to provide close guidance during removal and replacement of the internals package. 9.5.2.2.4 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 out of the mast to grip the fuel assembly. The gripper tube is long enough so the upper end is still contained in the 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. Controls for the manipulator crane are located inside the control console mounted on the trolley platform. Bridge, trolley and hoist positions are electronically displayed via encoders on the control console. The drives for the bridge, trolley and hoist are variable speed. Crane interlocks and limit switches are monitored by a Programmable Logic Controller (PLC). In an emergency the bridge trolley and hoist can be operated manually. An electronic load cell located on the trolley platform monitors the suspended weight on the gripper tool. This load cell sends a low voltage signal to a PLC and to a display located on the control console. This load is electronically displayed on the control console. An overload condition stops the hoist drive from 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: Chapter 9, Page 78 of 99 Revision 20, 2006 OAG10000215_1165

IP2 FSAR UPDATE

1. Encoders provide feedback pertaining to the bridge, trolley and hoist positions.

Bridge, trolley, and hoist positions are displayed to the operator on the control console.

2. Only the bridge and trolley are allowed to simultaneously operate at the same time. Bridge and trolley motion will be prohibited if hoist is in motion. Likewise, hoist motion will be prohibited if the bridge and trolley are already in motion.

3. Encoders determine the position of the mast, which will prohibit bridge and trolley movement based on the gripper height. The hoist also has a mechanical limit switch serving as a redundant mast "full up" limit.

4. A mechanical weight actuated lock in the gripper prevents operation of the gripper under load even if air pressure is applied to the operating cylinder. As backup protection to the mechanical interlock, an electrical interlock prevents the opening of a solenoid valve in the air line to the gripper except when the gripper is unloaded as indicated by a load cell.

5. Hoist load monitoring components detect overload conditions which will prohibit hoist raise motion when loading is excessive.

6. The PLC monitors the status of the gripper selector switch. Crane motion will not be allowed if the gripper indicator shows that the gripper is in transition or both conditions are activated (between OPEN and CLOSED).

7. The systems encoders along with the Crane's PLC will establish a boundary zone within the pool area. Crane motion is prohibited through these established boundary zones unless the bypass mode has been selected. Motion speeds will be decreased when operating in the bypass mode.

8. When the gripper is loaded with an assembly the mast must be in the full up position before bridge and trolley motion are allowed. With an empty gripper, bridge and trolley motion are prohibited until the "Gripper in Mast" elevation is present (full up is not required to traverse with an empty gripper).

9. Hoist load monitoring components detect underload conditions which will prohibit hoist lower motion. This prevents continued hoist motion if an assembly is hung up while being inserted between other fuel assemblies.

10. An encoder positioning system displays to the operator the precise position of the manipulator crane over each row of core coordinates for bridge, trolley and hoist movement over the reactor and the transfer canal.

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 design basis earthquake. 9.5.2.2.5 FSB Fuel Handling Bridge Crane Cha pter 9, Page 79 of 99 Revision 20, 2006 OAG10000215_1166

IP2 FSAR UPDATE 9.5.2.2.6 Fuel Transfer System The fuel transfer system, shown in Figure 9.5-1, is a cable driven system that traverses the conveyor car carriage on tracks extending from the refueling canal through the transfer tube and into the spent fuel pit. The conveyor car receives a fuel assembly in the vertical position from the manipulator crane. The fuel assembly is then lowered to a horizontal position for passage through the tube, and then is raised to a vertical position in the spent fuel pit. During plant operation, the conveyor car is stored in the refueling canal inside the containment. A blank flange is bolted on the transfer tube on the reactor side and a gate valve closed on the spent fuel pit side (see Figure 5.2-5) to seal the reactor containment. The blind flange is supplied with a double o-ring seal and is pressurized by the WCCPP System during normal operation to assure containment isolation. 9.5.2.2.7 Rod Cluster Control Changing Fixture A fixture is mounted on the reactor cavity wall for removing rod cluster control (RCC) elements from spent fuel assemblies and inserting them into new fuel assemblies. 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 lifts it out of the fuel assembly. By repositioning the carriage, a new fuel assembly is brought under the guide tube and the gripper lowers the RCC element and Cha pter 9, Page 80 of 99 Revision 20, 2006 OAG10000215_1167

IP2 FSAR UPDATE releases it. The manipulator crane loads and removes the fuel assemblies into and out of the carriage. 9.5.2.2.8 Lower Internals Support Stand A support stand for the lower internals package is installed in the lower internals laydown area at the east end of the refueling canal. The stand is to be used to rest the lower internals package to facilitate access to the internal surfaces of the reactor vessel. 9.5.3 System Evaluation Underwater transfer of spent fuel provides essential ease and corresponding safety in handling operations. Water is an effective, economic, and transparent radiation shield and a reliable cooling medium for removal of decay heat. Basic provisions to ensure the safety of refueling operations are:

1. Gamma radiation levels in the containment and fuel storage areas are continuously monitored. These monitors provide an audible alarm at the initiating detector indicating an unsafe condition. Continuous monitoring of reactor neutron flux provides immediate indication and alarm of an abnormal core flux level in the control room.

2. Violation of containment integrity is not permitted when the reactor vessel head is removed unless the shutdown margin is maintained greater than 5-percent L1k/k.

3. Whenever fuel is added to the reactor core, a reciprocal curve of source neutron multiplication is recorded to verify the sub-criticality of the core.

4. A Boraflex surveillance program was established when the high density racks utilizing Boraflex were installed. This program now includes coupon surveillance and monitoring of silica level (which is indicative of Boraflex degradation) in the spent fuel pit water.

9.5.3.1 Incident Protection Direct communication between the control room and the refueling cavity manipulator crane operator is available whenever changes in core geometry are taking place. This provision allows the control room operator to inform the manipulator crane operator of any impending unsafe condition detected from the main control board indicators during fuel movement. This provision shall be satisfied with fuel in the reactor and when: 1) the reactor head is being moved, or 2) the upper internals are being moved, or 3) loading and unloading fuel from the reactor, or 4) heavy loads greater than 2300 Ibs (except for installed crane systems) are being moved over the reactor with the reactor vessel head removed. If direct communication between the control room and the refueling cavity manipulator cannot be met, suspend any and all of these operations. Suspension of these operations shall not preclude completion of movement of the above components to a safe conservative position. Cha pter 9, Page 81 of 99 Revision 20, 2006 OAG10000215_1168

IP2 FSAR UPDATE 9.5.3.2 Malfunction Analysis Various potential failures, which could create paths for drainage from the refueling cavity have been considered. A plant procedure defines actions to deal with these postulated events. All credible failures result in drainage to safe storage. An analysis evaluating the environmental consequences of a fuel handling incident is presented in Section 14.2.1.1. Inadvertently locating an unirradiated fuel assembly of 5.0-percent enrichment in a region II storage location has been analyzed. The analysis shows that the array would be subcritical even with no soluble boron poison in the water in the fuel storage pool. With a boron concentration of 350 ppm the shutdown margin would be more than Ei~p~rcent. The technical specifications require that the boron concentration be maintained at gOGO ppm or more at all times. 9.5.4 Minimum Operating Conditions Minimum operating conditions are specified in the facility Technical Specifications. In addition, when fuel is in the reactor vessel and the reactor head bolts are less than fully tensioned the reactor Tavg shall be less than or equal to 140°F. 9.5.5 Tests and Inspections During preoperational testing, the Presray seal (which seals the reactor vessel flange to the bottom of the reactor cavity) was deflated with a full head of water in the cavity. No leakage was observed. 9.5.6 Control of Heavy Loads The control of heavy loads in the fuel storage building and the movement of loads over spent fuel in the spent fuel pit are discussed in the Technical Requirements Manual. Cha pter 9, Page 82 of 99 Revision 20, 2006 OAG10000215_1169

IP2 FSAR UPDATE TABLE 9.5-1 Fuel Handling System Data NEW FUEL STORAGE PIT Core storage capacity 1/3 Equivalent fuel assemblies 72 Center-to-center spacing of assemblies, in. 20.5 Maximum Keffwith unborated water 0.95 SPENT FUEL STORAGE PIT Equivalent fuel assemblies1 1376 Number of space accommodations for failed fuel cans 2 Number of space accommodations for spent fuel shipping cask 1 Center-to-center spacing of Regions 1-1,1-2 assembly storage 10.545(N-S) cells, in 10.765(E-W) Center-to-center spacing of Regions 2-1, 2-2 assembly storage 9.04 cells, in Maximum Keff with borated water (Regions 1-1, 1-2 and Regions  ::;0.95 2-1, 2-2) Maximum Keff with unborated water (Regions 1-1, 1-2 and <1.0 Regions 2-1, 2-2) MISCELLANEOUS DETAILS Width of refueling canal, ft 3 Wall thickness for spent fuel storage pit, ft 3 to 6 Weight of fuel assembly with rod cluster control (dry), Ib 1,580 Quantity of water required for refueling, gal 300,000 Notes:

1. After reracking.

9.5 FIGURES Figure No. Title Figure 9.5-1 Fuel Transfer System Figure 9.5-2 Spent Fuel Storage Rack Layout Figure 9.5-3 Spent Fuel Storage Cell Region 1 Figure 9.5-4 Region I Cell Cross-Section Figure 9.5-5 Region II Cross-Section Cha pter 9, Page 83 of 99 Revision 20, 2006 OAG10000215_1170

IP2 FSAR UPDATE 9.6 FACILITY SERVICE SYSTEMS 9.6.1 Service Water System 9.6.1.1 Design Basis The service water system is designed to supply cooling water from the Hudson River to various heat loads in both the primary and secondary portions of the plant. Provision is made to ensure a continuous flow of cooling water to those systems and components necessary for plant safety either during normal operation or under abnormal and accident conditions. Sufficient redundancy of active and passive components is provided to ensure that cooling is maintained to vital loads for short and long periods. The design of the essential header is to provide cooling water in the event of a single failure of any active component used during the injection phase of a loss-of-coolant accident. The system also provides water required for cleaning the traveling screens. 9.6.1.2 System Design and Operation The service water system flow diagram is shown in Plant Drawings 9321-2722 and 209762 [Formerly UFSAR Figure 9.6-1, sheets 1 and 2]. Six identical vertical, centrifugal sump-type pumps, each having a capacity of at least 5000 gpm at 220-ft total design head, supply service water to two independent discharge headers; each header may be supplied by three of the pumps. Two pumps are required for design flow in each header. A rotary-type strainer is in the discharge of each pump, and is designed to remove solids down to 1/16-in. diameter. Each header is connected to an independent supply line. Either of the two supply lines can be used to supply the essential loads, with the other line feeding the nonessential loads. The essential loads are those, which must have an assured supply of cooling water in the event of a loss of offsite power and/or a loss-of-coolant accident. The cooling water for these loads is supplied by the designated essential service water header. The nonessential loads are those, which are supplied with cooling water from the designated nonessential service water header by manually starting a service water pump when required following a loss-of-coolant accident. The essential and nonessential service water requirements are listed in Table 9.6-1. The nonessential loads are the component cooling heat exchangers, the turbine lube oil coolers, the main boiler feed pump lube oil coolers, and the remaining steam generation plant services. By manual valve operation, the essential loads can be transferred to the supply line carrying the nonessential loads and vice versa. Connections have been provided so the turbine generator lube oil coolers and other non-safety related loads can be supplied from the Unit 1 river water system. Water is drawn from the river and passes under a debris wall, through two racks in parallel and finally two traveling screens. Each pump in the circulating water system is installed in an individual chamber while the service water pumps are in a common chamber with two intakes. Each intake is provided with a traveling screen. Openings are also provided between the main circulating water pump chambers and the service water pump chamber. These two openings can be closed by gates. One gate is normally open. The service water pumps can therefore obtain water through four separate intakes each equipped with means to prevent debris from entering the pumps, and each capable of supplying all the water required for the service water pumps. Electric heaters are provided in the traveling screens 27 and 28 to prevent icing of the screens. Even if the main circulating pump intake Cha pter 9, Page 84 of 99 Revision 20, 2006 OAG10000215_1171

IP2 FSAR UPDATE were gO-percent blocked, that intake alone would be capable of supplying all water required for the service water pumps at design conditions. Service water is chlorinated by the addition of sodium hypochlorite solution as required to control micro-organism fouling of the system. The intake structure is designed as seismic Class I, and is therefore not subject to collapse under earthquake loading. During normal operation, the essential loads are supplied by at least one of the three pumps provided and the nonessential loads are normally supplied by two of the three pumps provided. Following a simultaneous incident and loss of offsite power, the cooling water requirements for all five fan cooling units and the other essential loads can be supplied by any two of the three service water pumps on the header designated to supply the nuclear and essential secondary load supply lines. Any two of these three pumps can be powered by the emergency diesels as described in Chapter 8. These emergency powered pumps are those necessary and sufficient to meet blackout and emergency conditions. Either one of the two sets of three pumps can be placed on the diesel starting logic. The containment ventilation cooling units are supplied by individual lines from the containment service water header. Each inlet line is provided with redundant motor-operated shutoff valves and drain valves. Similarly, each discharge line from the cooler is provided with redundant motor-operated shutoff valves and a manual balancing valve. This allows each cooler to be isolated individually for leak testing of the system or to be drained and maintained open to atmosphere during the integrated leak tests of containment. The ventilation cooler and motor cooler discharge lines will be monitored for radioactivity by routing a small bypass flow from each through redundant radiation monitors. Upon indication of radioactivity in the effluent, each cooler discharge line would be monitored individually to locate the defective cooling coil. This feature has been incorporated into the design since the service water system pressure at locations inside the containment with the system in the incident mode alignment could be below the containment post-accident design pressure of 47 psig. Thus, there could be outleakage of radioactivity to the environment if a break occurred in the service water system. However, 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 with the system in the incident mode alignment is below the containment design pressure of 47 psig. During normal plant operation, flow through the cooling units will normally be throttled for containment temperature control purposes by a valve on the common discharge header from the cooling units. Two independent, full-flow isolation valves open automatically in the event of a safety injection signal to bypass the control valve. Both valves fail in the open position upon loss of air pressure and either valve is capable of passing the full flow required for all five fan cooling units for accident mitigation. An 18-in. bypass line has been installed around the flow control valves in the service water return line from the fan cooler units. The line containing a flow limiting orifice and a butterfly valve can provide manual control for optimal service water flow rate through the fan cooler units during normal plant operation. Should there be a failure in the piping or valves at the header supplying water to the containment cooling coils, one of the two series header isolation valves in the center of the header can be manually closed and service will continue on the side of the header opposite the failure. The supply line attached to Chapter 9, Page 85 of 99 Revision 20, 2006 OAG10000215_1172

IP2 FSAR UPDATE this side of the header now supplies the essential loads, whether or not it did so before the failure. Likewise, operation of at least one component cooling heat exchanger is ensured despite the failure of any single active or passive component in the system from the service water pumps to the heat exchangers themselves. Following a simultaneous incident and blackout, the component cooling heat exchangers are not needed during the injection phase: thus they are normally fed from the nonessential supply header. At the beginning of the recirculation phase at least one of the service water pumps on the nonessential header is manually started to supply at least 2500 gpm of service water to each of the component cooling heat exchangers. The emergency diesel-driven generator units are supplied with cooling water from the essential supply line on a continuous basis. One of the two parallel modulating control valves in the common discharge line from the diesel coolers is flow-controlled during normal operation, and on a safety injection signal, both valves open fully to ensure a sufficient supply of cooling water to the diesels. The inlet valving is arranged so that each of the three diesels can be served by either of the supply headers and, furthermore, the failure of a single passive or active component will not result in the loss of all diesel power. 9.6.1.3 Design Evaluation The nonessential portion of the service water system is not required for the maintenance of plant safety immediately following an accident. The essential portion of the service water system is designed to provide cooling water in the event of a single failure of any active component used during the injection phase of the safety injection system (Section 6.2). Sufficient pump capacity is included to provide design service water flow under all conditions and the headers are arranged in such a way that even loss of a complete header does not jeopardize plant safety. In response the NRC Generic Letter 96-06, the containment fan cooler units and their associated service water piping were evaluated for susceptibility to waterhammer or two-phase flow. In the event of a loss of offsite power, the flow of essential service water will be interrupted until the emergency diesel generators start and restore power to the essential service water pumps. The pressure in the cooling coils and service water piping will drop to subatmospheric and a vapor pocket will form in the region of the fan coolers. When the essential service water pumps restart, the pocket will close and a water hammer will occur. The magnitude of waterhammer is approximately 394 psig. Dynamic analysis of the piping and supports shows that stresses meet the criteria for upset and faulted conditions, respectively. In the case of loss of offsite power and a loss of coolant accident, water trapped in the tubes and piping will be heated and vaporized. When the service water pumps are restarted, rapid condensation of trapped steam and collapsing of the void causes a waterhammer pressure pulse, with a magnitude less than that discussed in the preceding paragraph. The potential for two-phase flow conditions has also been evaluated. If it is assumed that there is no fouling of the fan cooler tubes, there will be flashing and two-phase flow in the discharge piping. However, analyses show that, although the flow will be reduced, the clean fan cooler units will exchange enough heat to meet required removal rates. Cha pter 9, Page 86 of 99 Revision 20, 2006 OAG10000215_1173

IP2 FSAR UPDATE 9.6.1.4 Tests and Inspections Each service water pump underwent a hydrostatic test in the shop in which all wetted parts were subjected to a hydrostatic pressure of one-and-one-half times the shutoff head of the pump. In addition, the normal capacity versus head tests were made on each pump. Valves in the portions of the service water system essential to safety underwent a shop hydrostatic test of 250 psi on the body and 175 psi on the seat. The service water system design pressure is 150 psig. All service water piping was hydrostatically tested in the field at 225 psig or one-and-one-half times design. The welds in shop-fabricated service water piping were 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 are tested periodically. 9.6.2 Fire Protection System Criterion: Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and the control room. Fire detection and protection systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Fire fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components. (GDC 3, Appendix A to 10 CFR 50) This criterion (GDC 3, 10 CFR 50 Appendix A) represents a revised design basis for the Indian Point Unit 2 fire protection system as was established for the original plant design and initial license application. In 1976 at the request of the NRC, Con Edison initiated a review and evaluation of the station fire protection system to this new criterion; modifications were subsequently proposed by Con Edison to the overall fire protection program. On January 31, 1979, the NRC approved the Indian Point Unit 2 overall fire protection program as providing additional assurance that safe shutdown can be accomplished and that the plant can be maintained in a safe condition during and following potential fire situations. This NRC approval was made as Amendment No. 46 to the facility operating license. Additional fire protection regulations were issued in 10 CFR 50.48 and Appendix R to Part 50 on November 19,1980, with an effective date of February 17,1981. These regulations established requirements for utilities to implement a fire protection program, and backfitted certain requirements in Appendix R to all utilities. For Con Edison, these included the various separation and protection requirements contained within Section III.G, emergency lighting requirements as stipulated by III.J, and oil collection system requirements for reactor coolant pumps as contained in Section 111.0. Additionally, Section III.L established performance requirements for alternative shutdown systems. Subsequent to the regulations established in 10 CFR 50.48, various NRC generic letters and guidance documents have been issued to provide clarification of the Appendix R requirements. Chapter 9, Page 87 of 99 Revision 20, 2006 OAG10000215_1174

IP2 FSAR UPDATE 9.6.3 City Water System The functions of the city water system are:

1. To provide the water supply for the fire protection system.

2. To provide an emergency supply of water to the suction of the auxiliary boiler feed pumps.

3. To provide makeup water to various systems.

4. To provide cooling water to various components.

5. To provide water to areas where hose connections are located for general usage.

City water for the Indian Point Unit 2 comes from the city water main on Broadway via the Unit 1 mains and storage tanks. Unit 2 is tied to this system primarily through piping connections at two locations on the low pressure header (see Plant Drawings 192505, 192506, and 193183 [Formerly UFSAR Figure 9.6-5]). One connection is in the vicinity of the Unit 1 superheater building on the south side of the header. This connection provides water for:

1. Emergency makeup to the house service boilers.

2. Cooling the house service boiler water samples.

3. General usage at the house service boilers.

4. Makeup to the expansion tank of the conventional plant closed cooling system.

5. Cooling and general usage at the steam and water analysis station.

The second connection is at the north side of the header. This connection provides water for:

1. Makeup to the expansion tanks of the diesel-generator jacket water cooling system.

2. Emergency feed to the auxiliary boiler feed pumps.

3. Makeup to the expansion tank of the instrument air compressor closed cooling system.

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4. General usage via hose connections inside the primary auxiliary building and waste holdup tank pit.

5. Emergency makeup to the isolation valve seal-water supply tank.

6. Spray water to the steam-generator blowdown tank.

A backup water supply is also provided for the circulating water pump seals and bearings. There are also emergency city water connections in the primary auxiliary building that can be used for the charging pumps, residual heat removal pumps, and safety injection pumps. 9.6.4 Compressed Air Systems 9.6.4.1 Instrument Air System The instrument air system is designed such that the instrument air shall be available under all operating conditions; all essential systems requiring air during or after an accident shall be self supporting; all controls shall fail to a safe position on loss of power; and, after an accident, the air system shall be re-established. The system is shown in Plant Drawing 9321-2036 [Formerly UFSAR Figure 9.6-6]. To meet the design criteria the following design features have been incorporated. Duplicate compressors are installed with duplicate dryers and filters throughout. In addition, alternate supplies are provided from the Unit 2 station air system, and Unit 1 station air system. A connection has been provided in the station air system to allow a backup supply of air from portable compressed air equipment. Those items essential for safe operation and safe cooldown are provided with air reserves or gas bottles. These supplies enable the equipment to function in a safe manner until the air supply is reestablished. The controls are specified to fail to a safe position on loss of air or electrical power. The compressors, filters and air dryers are located on the ground floor of the control building, a seismic Class I structure, and they, along with other essential sections of the air supply system, have been designed to operate after a seismic event. In the event of a break in the non-essential portion of the system, a flow restrictor in the supply line to the non-essential portion will limit flow to the capacity of one instrument air compressor. The system is served by two 225-scfm Worthington teflon-ring compressors, which discharge into a common air receiver. The instrument air from the receiver passes through one of two full-capacity heatless dryers. These heatless dryers are rated at 750 scfm, dewpoint compatible with the lowest expected outdoor temperature, and are dual-tower type dryers, with one of the dryers in service and one on standby. However, in the event that the transfer mechanism should fail during cycling of the dryer, the other dryer can be brought in to service. Each dryer is basically a stand-alone system, with dual prefilter, dryer and afterfilter units, and with local alarms and category alarms to the control room. An alternate air supply line from the station air system is provided, and has its own pair of full-capacity heatless regenerative dryers. The instrument air compressors may be operated in two modes. One mode provides for the compressors to be in standby and to come on automatically in the event of low pressure in the common air receiver. During this mode, air is supplied by the station air system. The other mode of operation provides for simultaneous running of both compressors in order to provide continuity of service to Class I areas in the event of outage of the conventional plant instrument air header. A restriction orifice is provided so as to limit the flow to the capacity of one instrument air compressor into a possible line break in the secondary plant air header. Cha pter 9, Page 89 of 99 Revision 20, 2006 OAG10000215_1176

IP2 FSAR UPDATE Upon notification of this break, a valve is provided to isolate the secondary plant and prevent pressure decay in the primary plant header. Valving has been installed to provide flexible operations as related to the alternate station air supply and to maintain proper isolation capabilities. All air and oil filters are dual type to provide maintenance during operation. 9.6.4.2 Station Air System The station air system shown in Plant Drawing 9321-2035 [Formerly UFSAR Figure 9.6-7] is supplied by a Worthington Corporation two-stage 650-scfm compressor located in the turbine building. The air is discharged through an aftercooler and moisture separator at 100 psig and 110°F. The maximum discharge pressure will be 125 psig. The cooling water for the aftercooler and compressor jacket is supplied from a closed cooling water system, which contains treated city water. The compressor is controlled by the solenoid unloader valves, which are energized through a pressure switch arrangement in automatic or hand (manual) modes. In the automatic mode, the compressor will run in single- or two-stage operation and unload at a predetermined pressure setting with motor and compressor stopped. In manual mode, the compressor runs continuously and is loaded and unloaded at predetermined pressure settings. High-water and high-air temperature switches are connected to the control annunciator. This system is alternatively supplied by the Unit 1 service air system through a manually operated valve interconnection to the Unit 2 air receiver. The size of the connection is equal to the Unit 2 supply pipe. The station air system can also serve as an alternate supply to the Unit 2 instrument air system. In addition, an automatic emergency supply is supplied to the containment building weld channel and penetration pressurization system. Valve position lights in the control room advise the operator as to the status of emergency makeup control valve PCV-1140. A manual local reset solenoid valve is provided at the emergency valve. 9.6.5 Heating System The heating system for Unit 2 represents an extension of the heating system for the Indian Point Unit 1. Package boilers have been installed to supply steam for Unit 2 and are interconnected with the distribution header of the boilers for Unit 1. The main steam header from these boilers links the existing steam header to Unit 2 and also to Unit 3, so that output from any of the package boilers may be made available for the heating requirements of Unit 1, Unit 2, or Unit 3. With respect to Unit 2, there are separate piping circuits for the unit heater steam supply to the east side and the west side of the turbine hall, including the heater bay. An extension from the circuit to the east side of the turbine hall serves the turbine oil storage tanks for both clean and dirty oil storage. Other heating services extend to the fan room, the fuel storage building, the containment building, the primary auxiliary building, the primary water storage tank, and the refueling water storage tank. Cha pter 9, Page 90 of 99 Revision 20, 2006 OAG10000215_1177

IP2 FSAR UPDATE Provision is made for the following heating services:

1. Containment building.

a. Steam unit heaters.

b. Valves with hose bibs for maintenance purposes.

2. Primary auxiliary building.

a. Electric strip heaters.

b. Steam unit heaters.

c. Air makeup steam tempering units.

3. Purge system containment building.

a. Air makeup steam tempering units.

4. Fuel storage building.

a. Steam unit heaters for standby heating.

b. Air makeup steam tempering units. (Steam supply isolated)

5. Fan room.

a. One steam unit heater.

9.6.6 Plant Communications Systems For discussion of the facility communications systems, see Section 7.7.4. REFERENCES FOR SECTION 9.6

1. Letter from Donald S. Brinkman, NRC, to Stephen B. Bram, Con Edison,

Subject:

Emergency Amendment to Increase the Service Water Temperature Limit to 90°F (TAC 73764), dated August 7, 1989. TABLE 9.6-1 Minimum Essential Service Water Requirement Under Accident Conditions Service Flow each Number Total Flow (gpm) (gpm) Containment Recirculation Fan Coolers 1600 5 8000 Containment Recirculation Fan Coolers Motors 17 5 85 Emergency Diesel Generators 400 3 1200 Instrument Air Compressor Heat Exchangers 65 2 65 Radiation Monitor Sample Coolers 10 3 30 Service Water Pump Strainer Blowdown 100 3 300 (750)1 Minimum Non Essential Service Water Requirements Post LOCA Recirculation Service Flow each Number Total Flow (gpm) (gpm) Cha pter 9, Page 91 of 99 Revision 20, 2006 OAG10000215_1178

IP2 FSAR UPDATE Component Cooling Water Heat Exchangers 2500 2 5000 Service Water Pump Strainer Blowdown 100 3 300 (750)1 Note:

1. Each strainer is mechanically set for 225 +/- 25 gpm backflush flow, 750 gpm total (max).

9.6 FIGURES Figure No. Title Figure 9.6-1 Sh. 1 Service Water System - Flow Diagram, Sheet 1, Replaced with Plant Drawing 9321-2722 Figure 9.6-1 Sh. 2 Service Water System - Flow Diagram, Sheet 2, Replaced with Plant Drawing 209762 Figures 9.6-2 Through Deleted 9.6-4 Figure 9.6-5 Sh. 1 City Water System - Flow Diagram, Sheet 1, Replaced with Plant Drawing 192505 Figure 9.6-5 Sh. 2 City Water System - Flow Diagram, Sheet 2, Replaced with Plant Drawing 192506 Figure 9.6-5 Sh. 3 City Water System - Flow Diagram, Sheet 3, Replaced with Plant Drawing 193183 Figure 9.6-6 Instrument Air - Flow Diagram, Replaced with Plant Drawing 9321-2036 Figure 9.6-7 Station Air - Flow Diagram, Replaced with Plant Drawing 9321-2035 9.7 EQUIPMENT AND SYSTEM DECONTAMINATION 9.7.1 Design Basis Activity outside the core can 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 being decontaminated. The products of water activation are not long lived and may be removed by natural decay during reactor cool-down and subsequent flushing procedures. Activated corrosion products are the primary source of the remaining activity. 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 (Fe304) with oxides of other metals 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 components. The contamination can be spread by various means when access is required. Contact while working on primary system components can result in contamination of the equipment, tools and clothing Chapter 9, Page 92 of 99 Revision 20, 2006 OAG10000215_1179

IP2 FSAR UPDATE of the personnel involved in the maintenance. Also, leakage from the system during operation or spillage during maintenance can contaminate the immediate areas and contribute to the contamination of the equipment, tools, and clothing. 9.7.2 Methods 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 nonporous materials. Personnel and their clothing are decontaminated according to the standard health physics requirements. Those areas of the plant, which are susceptible to spillage of radioactive fluids are painted with a sealant to facilitate decontamination that may be required. Generally washing and flushing of the surface 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. Portable components and tools can be cleaned by the use of a liquid abrasive bead decontamination unit, an ultrasonic unit, a sandblast unit or a Freon degreaser unit installed in Unit 1. 9.7.3 Decontamination Facilities Decontamination facilities onsite consist of an equipment pit and a cask pit located adjacent to the spent fuel storage pit. In the stainless steel-lined equipment pit, fuel handling tools and other tools can be cleaned and decontaminated. In the cask decontamination pit, the outside surfaces of the shipping casks are decontaminated, if required, by using steam, 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 the personnel, a decontamination shower and washroom is located adjacent to the radiation control area locker room. Personnel decontamination kits with instructions for their use are in the radiation control area locker room. 9.8 PRIMARY AUXILIARY BUILDING VENTILATION SYSTEM 9.8.1 Design Basis The primary auxiliary building ventilation system is designed to accomplish the following:

1. Provide sufficient circulation of filtered air through the various rooms and compartments of the building to remove equipment heat and maintain safe ambient operating temperatures.

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2. Control flow direction of airborne radioactivity from low activity areas toward higher activity areas and through monitored exhaust paths.

3. Provide purging of the building to the plant vent for dispersion to the environment.

The air exhausted by the system is filtered, monitored, and diluted so that offsite dose during normal operation will not exceed Offsite Dose Calculation Manual (ODCM). 9.8.2 System Design and Operation The primary auxiliary building ventilation system (See Plant Drawing 9321-4022 [Formerly UFSAR Figure 5.3-1]) is composed of the following systems:

1. Makeup air handling system complete with fan, filters, heating coils, and supply ductwork.

2. Exhaust system complete with fans, ductwork, roughing filters, HEPA filters, and charcoal filters.

3. Outside air intake for the waste storage tank pit area.

Design parameters for the system components are given in Table 9.8-1. Branch supply ducts direct makeup air to the various floors at the east end of the building, from where it flows to the rooms and compartments. Air is exhausted from each of the building compartments through ductwork designed to make the supply air sweep across the room as it travels to the room exhaust register. The air then flows to the exhaust fan inlet plenum, and is drawn by the operating exhaust fan through roughing filters, HEPA filters, and charcoal filters before discharge to the plant vent. The exhaust system has been designed to ensure that air flows from the "clean" end of the building through the "hot" areas. Ventilating air exhausted from the waste storage tank pit is arranged to bypass the primary auxiliary building system and flow directly into the exhaust fan inlet plenum. There are four fans in the containment building purge system and primary auxiliary building ventilation system. The two exhaust fans (containment building purge and/or primary auxiliary building exhaust fans 21 and 22) are common to both the containment building purge system and primary auxiliary building ventilation system. The supply fan in each of the ventilation systems operates only in its individual ventilation system. The primary auxiliary building supply fan normally runs, along with either or both of the exhaust fans. The containment building purge supply fan runs with either of the exhaust fans, with the other exhaust fan as a backup. All four fans may also run simultaneously. The interlocking for the fans is such that in no event will the number of supply fans operating be greater than the number of exhaust fans operating. However, operation of an exhaust fan without a supply fan running is acceptable. Fans are manually selected. All four fans can be started and stopped by four discrete control switches located on the fan room control panels. Each fan has indicating lights on the fan room control panel and in the main control room. An auto trip alarm is also provided. In addition, each of the fans have a "jog" pushbutton located on the fan room control panel for testing. Cha pter 9, Page 94 of 99 Revision 20, 2006 OAG10000215_1181

IP2 FSAR UPDATE TABLE 9.8-1 Primary Auxiliary Building Ventilation System Component Data Units Required Units Units for System Installed Capacity Normal Operation Exhaust1 Fans, standard conditions 2 55,500 cfm 1 Fan pressure 10.3 in. H2O Fan motors 2 125 hp 1 Plenums 2 55,500 cfm 1 Roughing filters 2 55,500 cfm 1 HEPA filters 2 55,500 cfm 1 Carbon Filters 1 55,500 cfm 1 Supply Tempering Unit (Primary Auxiliary Building) Fans, standard conditions 1 50,400 cfm 1 Fan pressure 1 2.5-in. H2O Fan motor 1 50 hp 1 Filters 1 50,400 cfm 1 Coils 1 50,400 cfm 1 Outside Air Intake 1 5100 cfm 12 (Waste Storage Tank Pit Area) Notes:

1. These two exhaust fans are used interchangeably and/or as backup for:

(1) ventilation of primary auxiliary building, (2) containment building purge system.

2. Outside Air Intake may be covered during cold weather conditions.

9.9 CONTROL ROOM VENTILATION SYSTEM 9.9.1 Design Basis The control room heating, ventilation, and air conditioning system is designed to accomplish the following:

1. Maintain 75°F dry bulb and approximately 50-percent relative humidity in the control room at outside design conditions at 93°F dry bulb and 75°F wet bulb.

2. Permit cleanup of airborne particulate radioactivity entering the control room with normal makeup air flow and by infiltration.

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IP2 FSAR UPDATE 9.9.2 System Design and Operation The Unit 2 control room ventilation system is composed of the following equipment:

1. A direct expansion air conditioning unit complete with fan, steam heating coil and roughing filter. The design capacity of the unit is 9200 cfm. A backup fan of the same design capacity has been installed in parallel with the air conditioning unit.

2. A filter unit consisting of case, HEPA filters, charcoal filters, post-filters and booster fans with a capacity of approximately 2000 cfm.

3. Duct system complete with dampers and controls to allow three system operating modes.

The Unit 1 control room ventilation equipment for the central control room has been modified for recirculation mode only. The control room ventilation systems are shown on Plant Drawings 252665 and 138248 [Formerly UFSAR Figure 9.9-1]. The Unit 2 control room ventilation system can be operated as follows:

1. Normal Conditions

a. With outside air makeup will supply cooling or heating for the control room atmosphere as required, using fresh outside air makeup and with the charcoal filter unit bypassed. (Mode 1)

2. Incident Conditions

a. On safety injection and/or high radiation signal, with outside air makeup filtered the booster fan will start and dampers will be positioned to permit outside air to flow through the charcoal filter unit. (Mode 2)

b. On toxic gas and/or smoke signal, the outside makeup air will be isolated and the carbon filter booster fan will not operate, the system will be in 100%

recirculation mode. (Mode 3) All these operations can be performed manually from the control room. However, in the event of a safety injection signal and/or high radiation signal, the control room dampers will automatically reposition and start the booster fan to place the charcoal filter unit in service, for system operating mode 2. A redundant toxic chemical and radiation monitor for central control room air intakes has been installed. For additional discussion of this system, see Section 7.2. 9.9 FIGURES Figure No. Title Figure 9.9-1 Central Control Room HVAC (Heating, Ventilation, and Air Conditioning), Replaced with Plant Drawings 252665

                                  & 138248 Cha pter 9, Page 96 of 99 Revision 20, 2006 OAG10000215_1183

IP2 FSAR UPDATE 9.10 FUEL STORAGE BUILDING VENTILATION SYSTEM 9.10.1 Design Basis The fuel storage building ventilation system is designed to perform the following functions:

1. Maintain the fuel storage building at negative pressure so as to prevent unmonitored releases.

2. Provide sweep ventilation of the building, across the spent fuel pool, from areas of low potential contamination to areas of higher potential contamination.

3. Filter particulates and iodine through HEPA and charcoal filters to reduce the postulated offsite dose, which may result from a dropped fuel rod. NRC SER dated July 27, 2000 approved a fuel handling accident analysis that took no credit for filtration to reduce offsite dose so this design feature is no longer required for accident mitigation.

4. Remove normal building heat.

9.10.2 System Design And Operation The fuel storage building ventilation system, shown in Figure 5.3-1, consists of two air supply units (whose fans have been retired in place) and one exhaust system. In addition, an axial spot cooling fan circulates 3000 cfm of air to the spent fuel pit heat exchanger room. The power and control circuits for the fuel storage building (FSB) air supply fans and dampers, and dampers for the FSB exhaust fan, have been retired-in-place. Each supply unit has manually-operated outlet dampers that allow the exhaust fan to draw air through the building. Each also has a tempering (heating) coil which have been retired in place. Steam supply to the heating coils have been isolated and retired in place and the condensate line isolated. The exhaust system consists of registers, ductwork, a filter bank, and a fan. Three exhaust registers are located near the pool surface level, at the north end, and a fourth is near the ceiling at the north end of the building. The registers near the pool surface are intended to provide a sweep flow over the pool. Air from the registers is ducted to a plenum chamber, which contains the filter banks. It flows sequentially through filter banks, consisting of roughing filters, HEPA filters, and charcoal filters, and then to the exhaust fan. Air from the exhaust fan is discharged to the plant vent. The exhaust fan is the centrifugal type, belt-driven by 100 hp 480-V motor. The system provides an air flow rate of nominally 20,000 cfm. The system is balanced to divide the exhaust air flow equally between the exhaust registers and to maintain the building at a slight negative pressure. The exhaust fan is operated and controlled from a single local control room. As a result of IP2 Operating License Amendment No. 229 (dated June 5, 2002), the limiting conditions for operation and the surveillance requirements for the fuel storage building air Cha pter 9, Page 97 of 99 Revision 20, 2006 OAG10000215_1184

IP2 FSAR UPDATE filtration system were relocated from the Technical Specifications to the UFSAR. These relocated requirements have been modified to reflect the assumptions used for the fuel handling accidents approved by the Technical Specification Amendment 211 (July 27, 2000). These are contained in UFSAR Sections 9.10.3 and 9.10.4 below. 9.10.3 Limiting Conditions for Operation (Fuel Storage Building Air Filtration System) The fuel storage building ventilation system is assumed to be operating whenever spent fuel movement is taking place within the spent fuel storage areas, allowed after the fuel has had a continuous 100 hour decay period. 9.10.4 Surveillance Requirements (Fuel Storage Building Air Filtration System) Amendment 211 recognized the fuel storage building ventilation system would be operating for an accident even though the assumptions were to release the source term over a 2 hour period at ground level (FSAR Section 14.2). The fuel storage building ventilation system does not have to be demonstrated operable in the assumed configuration each refueling, prior to refueling operations, and prior to handling fuel. The fuel storage building air filtration system shall be periodically tested (a 25% allowance is allowed consistent with the philosophy of Technical Specification SR 3.0.2) to assure continued compliance with 10 CFR 50, Appendix I and design criteria in accordance with ASME N510-1989, as follows:

1. verifying that the pressure drop across the combined H EPA filters and charcoal adsorber banks is less than 6 inches water gauge while operating the system at ambient conditions and at a flow rate of 20,000 cfm +/-10% at least once each 24 months during aerosol or leak rate system tests.

2. verifying that the system maintains the spent fuel storage pool area at a pressure less than that of the outside atmosphere during system operation at least once each 24 months.

3. A visual inspection of the normal atmosphere cleanup system and all associated components should be performed in accordance with Section 5 of ASME N510-1989.

4. In-place aerosol leak tests, in accordance with Section 10 of ASME N510-1989, for HEPA filters upstream from the carbon adsorbers in normal atmosphere cleanup systems should be performed: at least once each 24 months; after each partial or complete replacement of a HEPA filter bank; following detection of, or evidence of, penetration or intrusion of water or other material into any portion of a normal atmosphere cleanup system that may have an adverse effect on the functional capability of the filters; and, following painting, fire, or chemical release in any ventilation zone communicating with the system that may have an adverse effect on the functional capability of the system. The leak test should confirm a combined penetration and leakage (or bypass) of the normal atmosphere cleanup system of less than 0.05% of the challenge aerosol at rated flow +/-10%. A filtration system satisfying this condition can be considered to warrant a 99% removal efficiency for particulates.

5. In-place leak testing, in accordance with Section 11 of ASME N510-1989, for adsorbers should be performed: at least once each 24 months; following removal of an adsorber sample for laboratory testing if the integrity of the adsorber section is affected; after each partial or complete replacement of carbon adsorber in an adsorber section; following detection of, or evidence of, penetration or intrusion of water or other material into any Cha pter 9, Page 98 of 99 Revision 20, 2006 OAG10000215_1185

IP2 FSAR UPDATE portion of a normal atmosphere cleanup system that may have an adverse effect on the functional capability of the adsorbers; and, following painting, fire, or chemical release in any ventilation zone communicating with the system that may have an adverse effect on the functional capability of the system. The leak test should confirm a combined penetration and leakage (or bypass) of the adsorber section of 0.05% or less of the challenge gas at rated flow +/-1 0%.

6. The efficiency of the activated carbon adsorber section should be determined by laboratory testing of representative samples of the activated carbon exposed simultaneously to the same service conditions as the adsorber section in accordance with ASTM 03803-1989 at a face velocity of 50 ftlmin, a temperature of 89F, and a 95% relative humidity. Sampling and analysis should be performed: at intervals of approximately 24 months; following painting, fire, or chemical release in any ventilation zone communicating with the system that may have an adverse effect on the functional capability of the carbon media; and, following detection of, or evidence of, penetration of water or other material into any portion of the filter system that may have an adverse effect on the functional capability of the carbon media. The acceptance criteria is a methyl iodide penetration of less than 7.5%.

Cha pter 9, Page 99 of 99 Revision 20, 2006 OAG10000215_1186

   °Yo1r'i9~;m 1~~~6fXZti~M V~nu1t19~9~n1~~q8y6N~~I~~~*~~gheast _ Indian Point Enerqy Center CONVEYOR CAR VlINCH                                        PLAIIT COITA IIER UPEIDING FR.uce IIII1CH MOTOR OR I VEIl PLAlfOPM U'EIIDIIIG FRANE
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• PERIPHERAL CELL INDIAN POINT UNIT No. 2 UFSAR FIGURE 9.5-2 SPENT FUEL STORAGE RACK LAYOUT MIC. No. 1999MC3887 REV. No. 178 OAGI0000215_1188

INDIAN POINT UNIT No. 2 UFSAR FIGURE 9.5-3 SPENT FUEL STORAGE CELL REGION I MIC. No. 1999MC3888 REV. No. 17 A OAGI0000215_1189

INDIAN POINT UNIT No. 2 UFSAR FIGURE 9.5-4 REGION I CELL CROSS SECTION MIC. No. 1999MC3889 I REV. No. 17 A OAGI0000215_1190

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I I 3/4"-T58 53 G. BHALLA 12/10/02 2 co L_.J cO L_.J cO L_.J cO L_.J THIS DRAWING CONTAINS ITEMS WHICH 2 v v v v MUST BE CONTROLLED WITHIN ENTERGY AS: 10 e-, '">-, 10 10 e-,

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0Jf') PER THE QAPD OWG. SIZE A ~;pGE COMPANY 3/4"-CH-250IR 3/4"-CH-250IR 3/4"-CH-250IR 3/4"-CH-2501 STATION i i i CON 251 A 2"-C58 REACTOR COOLANT 251B 2"-C58 REACTOR COOLANT 251C 2"-C58 REACTOR COOLANT i 251D 2"-C58 REACTOR COOLANT EDISON INDIAN POINT PUMP (LOOP I) PUMP (LOOP 2) PUMP (LOOP 3) PUMP (LOOP 4) NOTES BORO: WESTCHESTER C- 16 C- 18 C-20 C-23 AoVALVE FAILS WITH FLOW TO VOLUME CONTROL TANK TITLE: FLOW DIAGRAM [ B.SPECIAL VALVE-FUNCTIONS AS BOTH ISOLATION CHEMICAL & VOLUME CONTROL SYSTEM I I ** I ** & RELIEF VALVE" 251E 2"-C58 i C-17 R C S 251F 2"-C58 L i

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C- 19 II R C S I fA 251G 2"-C58 i ~ C-21 J R C S 251H 2"-C58 Li

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C-22 J R C s I fA Co'" INDICATES CONTROL VALVE HAS ADDITIONAL ASSOCIATED CONTROL EQUIPMENT & IS REPRESENTED UFSAR FIGURE Noo 9.2-1 (SHT. 2) APPROVALS B. B ON CONTROL VALVE HOOK-UP DETAIL DRAWING CP 9321-F-7056. ENGINEERING WORK THIS DRAWING WITH D.FOR CONTINUATION SEE DWG. 9321-F-2734 (TYP. FOR 12 LOCo) ENGINEER DWG. 9321-F-2734 MANAGER: CP

                                                              --                         --                                --                            CP
                                                                                                                                                               --                                              --                                       --                                  --          CP
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INSIDE REACTOR CONTAINMENT OUTSIDE REACTOR CONTAINMENT FOR CONT. SEE DWG. 9321-F-2736

                                                                                                                                                                                                                                                                                                                                                                                                    --                            --

CP

                                                                                                                                                                                                                                                                                                                                                                                                                                                                       --                                -                                                                                               AND DWGo 9321-F-2736 E.THE QUALITY GROUP A,B,C AND SEISMIC BOUNDARIES EXTEND TO THE FIRST SEISMIC SUPPORT/RESTRAINT BEYOND THE BOUNDARIES SHOWNo DISCIPLINE ENGINEER:

DESIGN 1 z DESIGN MANAGER: WoJo KING 11-6-87 REF. DWG. ~------------------~lV DESIGN 0 EVERYTHING ON THIS DRAWING 9321-C-2016 FLOW DIAGRAM SYMBOLS SUPERVISOR: FA 8-17-90 CO (EXCLUDING THOSE PORTIONS WITH 9321-F-2538 CONTAINMENT BLDG PRIMARY COOLANT DASHED LINES) IS PART OF THE P---'- DRAWN CHEMICAL AND VOLUME CONTROL PRESSURIZER PIPING PLAN ISOUTH HALF) BY: MARZULLO 8-17-90 OJ SYSTEM leVCS) LISTED IN EXHIBIT A. CI-240-1 EXCEPT AS 9321-F-2734 PIPING AT REACTOR COOLANT PUMPS ~------------------~Ico SPECIFICALLY INDICATED HEREIN DESIGN L.K 8-17-90 I ~ I RCS > REACTOR COOLANT SYSTEM CHECKER:

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w Z I "'-X42D H W A. DIAPHRAGM SEAL.

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I"-CH- R D. ADDITIONAL VENTS & DRAINS MAY BE REQUIRED BY THE EVERYTHING ON THIS DRAWING (EXCLUDING THOSE PORTIONS LINE 04 PI 191 53 E. PIPE RUN GLOBE VALVES ARE NORMALLY INSTALLED WITH FLOW UNDER THE SHOWN WITH DASHED LINES) IS PART OF THE CHEMICAL AND 1133 1155 Q: f-.-v1----" P SEAT. EXCEPTIONS ARE NC, (LATER) VOLUME CONTROL SYSTEM (CVCS) LISTED IN INE #104 A235309 I "-X42p' I-C36

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E UPDATED DWG, TO SHOW RETIRED BORON RECOVERY  ?/9/RR V EQUIPTMENT PER CRS #199806527. 8 A* G . ---+-------------1 CHEMICAL & VOLUME CONTROL L9QJ~~ INDIAN POINT f - I I DES RELEASE AS CONSTRUCTED, / 88 SYSTEM - SH. NO.2 - S G. BHALLA PIN 69901-AF I 02/09/01 D.B_/YJ, 2/09/01 SUPV. DEESNIGGN DISC. ENG. UFSAR FIGURE No.9. 2- I I SHT. 3 I o DATE ENG. MGR. DRWN ISCALE REC'O DWG

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• PW-612 I "-X42D PW-613 I "-X42D SEE DWG. t SEE DWG 9321 F 2719 ;::.. 22J ..

CLASS" A" S.C.~ NON-CLASS

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                                                                                                                                     ,y,t**---(        INSIDE              OUTSIDE                                                                                                                                                                                                                                                                                                                                                                   5383 SEE NOTE NO.3           --~,

1463 CONTAINMENT CONTAINMENT

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If) a) a) If) t FOR CONTINUATION SEE '"

                                                                                                                        ~~ 1/4"XI/4"                                                                                                                                                                                                                                                                                                                                                                                                                          If)       f-
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                       ,,*:'2               I")     f-                                                                                                                                                                   DWG. 9321-F-2728
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              " '" '" ri' '"" '",

I '" SOV

                                                                                                                        * ~                            955E
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              '" '                     "        a)

If) ........ If) 1 llJ _ '" - '- .--.--" In 976 3/8"LINE #297(2505R)

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                                                        ~I 4" T58 __-==-~~~F~.~C;.~~~~~::~S~L:-~2:5~O~5~R~
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                                                                                                                                                                                                                                                                                                                                         "    FOR CONTINUATION SEE                                                                                                                                                                                                     1 "                                                                    ) . .
                                            ~
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      ,

NO. 24 OWG. 9321-F-2736 3/8"T58 102635091 955F *** PRV

                                                                                                                                                                                                                                                                                                                                                                                                    ~                                                                                                                               T 6

I ZC ~C 3/S"TA5SR S 955E 955E 955E. VOLUME CONTROL (l_.__.__._._.**_-_._.__._- 3/4"LINE NO.297 (15IR) 3/4" 3/8" J 3/8"LINE #297(2505R) .i. 3"LINE #132

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          ~--:r-.--,-.--.----.--.--.--.-.--.--f
                                                                         -IS           -2S                                           [>"

147~SEE A242688 TANK GAS SPACE (CVCS) 975 3/4"T58 ~ '" 977 3/8"T58 I l' 954E F. C. SL -2505R NO. 23 2 ~.~~~--[~~-;~"::;;;~~:: y/4"T58 ID2635081 955E ** LEDT~MOIWNN OLUITNLEEtT~._......._..__..._....... ,-----::r 1 3/8"LINE NO.295 ISL-2505R) r ~ I TO SUMP TANK #21

                                                        ,                 ZC              ZC 3/8"TA58R S                                                                                                                                                                                                                                        ICVCS)                                    974A                                      ft,                                                                                                                                                                                                                        5462                                                                                                              989E 9550          9550                                           1,~5~                                            PRV            SOV                                                     PRV                                                                                                                         3/4"T58                                                                                                                                                                                                                                                             3/8"T58                                                                                                           3/8"T58                                                   3/8"T58 LE6~~~~ i~~~lT~._......._..__..._.......
                                                                         - 1S          -2S                                           ,><1 ...... ~        PRY                       956G           956G                                                    956H                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            5155 9550                                                                 T                                                                                      T                                                                         1                                                                              3/8"UNE NO.296 ISL-2505R) 9540                                                 F.e.                SL-2505R                                                                                                                                         L\ 1'1          Vii'                                           (eves)                                    9746                                                                                                                                                                                                                                                                 546 I                                                                                                               988
                                            ~

ac v ~ NO. 22 ~ SEE A242688 IA-1581 IVSWS IA-IS8, L! 3/4"T58 3/8"T58 5336 3/8"T58 L'J 3/4"T58 ID2635071 9550 u. ~

   --                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              SAMPLE )                          ~ III                                 ~   III                                                                                                                                                                                    r-
                                                        ,             C                  ZC 3/8"TA58R
                                                                                                                                                                                                      ~~6                                                                               ~     c                   k\NOTE A                                                                                                                                                                                                                                                                                                                                           SINK          (                                                                                     I w§~

956H 956H NOTES, 955C 955C EL.80'W.G.COMPRESSOR RM.

                                                                                                                                                                                                                                                                                           - I         -2        ,
                                                                         - IS           - 2S
                                                                              ...............
                                                                                                         ..c~--{:kJ~--lS;h'l-:><}"---1 1476                 LINE NO.69                                                                         y SEE NOTE NO.3

_ ________ ~~~---3~/~8~'~'L-I~N~E~N~0~.~6~9~I~S~L~2~5~0~5~R~)L_ __________________________________________________________________________________ ~3~/~8~"~T~5:8 5478 ________~~___ I. ALL TUBING IS 3/8" 0.0. EXCEPT q"'i. C".~ ______ ~~~ ______ ~~~~ ____ 1-~~~_~-2;5~O~5R I [:<<3------ 989F ~, AS DETAILED.

2. ALL VALVES INSTALLED WITH FLOW NO. 21 ~""",,"';tJ PRV 956G 3/8" TA58R

                                                                                                                                                                                                                                    ***                                              SOV
                                                                                                                                                                                                                                                                                                                 ':1JOM 3/8"TA58R
                                                                                                                                                                                                                                                                                                                          ***                                                                                                                                                                                                                                                                                                                                                                                                                                                 3/8"T58                                                UNDER SEAT EXCEPT VALVE NUMBERS
                                                               '00 SOV 955C 3/8" TA58R
                                                                                                              ***                                            1\ ~CP 956A                                                                                            956B         lS                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  962A, 962B. 962C, 968, 970A. 978, 980.

ACCUMULATOR SAMPLES (SIS) 951 T T 3. 3/4" NOMINAL PIPE 0.0. X 3/8" 0.0. IA-2420~1/2"PENETRATION TUBING INSERT.

            "                                                                                                                                                                                                                                              <---l><1                               ,
                                                                                                                                                                                                                                                                                                                                  ~

DWG. A235296 9TOI , 4. ADDITIONAL VENTS AND DRAINS MAY BE ZC IA-15B IVSWS IA-15S ID,

                                                                                                                                            ~

If) , 951 , i '"'" ACS \()

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                '"'"'         ACS
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                         '"'"

REQUIRED BASED ON PIPE LAYOUT.

                                                                                                                                                                                                                                  ~
                                                                                  -2S                                                                                                                                              -                                                        ze                          \NOTE A                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                5. ALL CHECK VALVES SHOULD BE "Y"

________ IA-1587 ~: 956B '", ,'" '"~,

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            ,   3/8"T58 TYPE LIFT CHECK.

6. DELAY COIL --- 125 FT. X 3/8" 0.0.

5 ~"--~~~:::I

                                                                                                                                                                                                                      /"'-.",- ....0:

8  ! _I

                                                                                                                                                                                                                                                                                           ~--,===--~ ID263501 I                                                                                                                                                                                                                                                                                                            '"                   '"                                                                                                      '"                                       4388                               TUBING X 0.245 1.0. OR EQUIVALENT.                                                                 ,
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                         ~                                                                                                 ~

956A *** 956B *** 3/8"LINE NO.25 (SL-

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                         ")

9S-Sv 1 3/8" 7 . * * '" INDICATESS CONTROL VALVE HAS ADDITIONAL ASSOCIATED EQUIPMENT PRV 3/8"TA58R PRV 3/8"TA58R 3/8"T58 AND IS REPRESENTED ON CONTROL 956C 9560 S VALVE HOOK-UP DETAIL DRAWING PRESSURE STM.SPACE PRESSURE STM.SPACE 9321-F-7056. T T SAMPLE HT.EXCH.NO.21 SAMPLE HT.EXCH'.NO.22 8. THE QUALITY GROUP A,B,C AND SEISMIC PENETRA TION S -IAC_><15:}8*i2:f------=t:JI\:}-~~, IVSWS IA-158 ~ 9Tcl5 BOUNDARIES EXTEND TO THE FIRST SEISMIC SUPPORT/RESTRAINT BEYOND

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If); If);

[ i:' ~

ACS ACS

                                                                                                                                                                                                                                  ~

THE BOUNDARIES SHOWN.

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                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            "                                                                                                                                                                   3/8"T58
                                                                                                     ~9 Vii~

fRESSURIZER LIQUID SPACE = I IIA-1588 ~ M ' (\J! N! 0J! 0J! i 4379 REFERENCE DRAWING, IRCS) OWG. ';J')21-t="-2738 ?8  ! ID263S021

                                                                                                      -:-1 F.C.
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                         ~                                                                                                 ~
                                                                 ~                                                       SL -        R             LINE NO.26                                                                     ~                                                                                                                                                                                                                                           3/8"LINE NO.26 ISL-2505R)                                                                                                                                                                                    ~         3/8"                                                                                                                      evcs - CHEMICAL VOLUME CONTROL SYSTEM
   --                                       r'         952 3/4"T58 ID26350s1 953 3/8"TA58R
                                                                                                              ***

1/4"XI/4":2

                                                                                                                                      ~

CP 956C 3/8"TA58R

                                                                                                                                                                                                                                    ***                                                                          9560 3/8"TA58R
                                                                                                                                                                                                                                                                                                                         ***                                                                                                                                                                                                                                          3/8"T58                                                                                            3/8"T58 U.E.&

U.E.& RCS - C. DWG. ------ 9321-F-2736 C. OWG. ------ 9321-F-2737 REACTOR COOLANT SYSTEM r-

                                                                                                                .-------t1~ ~ ~                                                                                                                                                                                                                                                                                                                                                                                                                                                                             PRESSURE LIQUID SPACE SAMPLE HT.EXCH'.NO.21 PRESSURE LIQUID SPACE SAMPLE HT,EXCH.NO.22 U.E.&

STEAM C. DWG. ------ 9321-F-2738 GEN. BLOWNOOWN SYS. PENETRA TION U.E.& C. DWG. ------ 9321-F-2729

                                                                                                                         ~3/8I1XI/4" ~                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    J/8"LINE #557 I 2505R ) TYP                                                           1                                                              ACS -

U.E.& AUXILIARY COOLANT SYSTEM C. DWG. ------ 9321-F-2720

                                                                                                                                      '"                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  FDR CONT. SEE OWG. A2084 79 ~""-"'T............--.~<j                                         [:<<3--------.,                                         WDS -

U.E.& WASTE DISPOSAL SYSTEM C. DWG. ------ 9321-F-2719 58 T 5160 U.E.& C. DWG. ------ 9321-F-2730 i

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                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                       ;>                                          ~

SAMPLE EL.80'W.G.COMPRESSOR RM. SINK 5334

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                           .....

l 1 5172 517 3/8"T58 SIS - U.E.& IVSWS U.E.& SAFETY INJECTION SYSTEM C. DWG. ------ 9321-F-2735

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          - ISOLATION VALVE SEAL WATER SYS C. DWG. ------ 9321-F-2746 RESTRICTION --,                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              5173           5156
                                                                                                                                                                                                                                                                                                                                                                                                                                                   ~                              COMPONENT COOLING WATER                           (ACS)                                                                                                                                                5194                                                         3/8"T58 3/8"T58                                                                 REFERENCES, 4

SEE OWG.9321-F-2738 ORIFICE IVSWS-3/8"-LINE NO.543

                                                                                                                                                                                                                                                                                                                                                                                                                                                        ~,                   REF OWG NO. 9321-F-2720                                                                                                                                                                                                                                                                                                                                                                                                                                    ,

954B 4546-X L' "':?~'Il* NOTE A "c-' "'" , "c- , '"

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      ",
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                             )

5193 PROCESS FLOW DIAGRAM W-OWG.540F902

                                                             ).         ,*~,_~~c_2505R ,                                                                                                                                                                                                                                                                                                                                                                                                                                                              ,,'

• HOT LEG LOOP 3 c-' 4376 3/8"T58 DEFINITION OF SYMBOLS,U.E.&C. SPEC.

IRCS)

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3/8"T58 PART B. 9321-01-248-18

                                                                                                                                                                                                                                                                                                                                                                                                                                   ~!                    ",                                                                                           ",

T1

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                                                                                                                                                                                                                                                                                                                                                                                                                                                         "                                                                                            "                                                            J                                                        9TO~   \------t<<:i---                                                                                                                              INSTALLATION OF INSTRUMENTATION T0                                                                                                                                                                                                                      I")                   I")                                                                 I")                      I")                                                                                                                                                    5202 1/2"T58 5159                     5158                                                   W-PRDC.SPEC. CAP - 294367 REV. I it                                                                                                                                                                                                                                                                    1/4"LINE NO.59
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                     .                 5218                                                                           3/8"T58
                                                                                                                                                                                                                                                                                                                                                                                                                                ~                                                                                         ~

LINE NO.59 3!R'/1 TI NO.59(SL-2505R)~ ~ 1/4'1 SL-2505R 1/411 (s/8" 5332 --+ (TYP) MATERIAL SPEC. PIPE AND FITTINGS,

                                                                                                                                                        -
                                                                                                                                                         !Ji ---

956E 3/4"TM58 956F 3/4"TM58 l : t 991 4389 3/8"C58 l: [::<<:] 989 3/8"T58

t r REACTOR COOLANT

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                         ...                             REACTOR COOLANT LJ                    ,

4398 5466 3/8"C58 5467 3/8"T58 3/8"T58 5203 1/2"T58 ID 551 I _3(~'T58151 57] U.E.&C. SPEC. 9321-01-248-18 9321-F-2016 FLOW DIAGRAM SYMBOLS NUCLEAR EQUIPMENT DRAINS 9321-F-2738 PENETf 3/8"C58 SAMPLE HEAT SAMPLE HEAT I") I") r-

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                               ~
   --                         HOT LEG LOOP                                                                                                                                                                                                                                                                                                                                                                                   EXCHANGER NO.21                                                                             EXCHANGER NO.22
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        'V ID                                                                                    55-SAMPLING SYSTEM DWG.A227178 (RCS)                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                   u
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      <                                                                                                                        POST ACCIDENT CONTAINMENT AIR SAMPLING SYSTEM DWG.A208479
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                               ~ ....

994A 1-PRIMARY WATER SYS

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                               "'"'"

o INLINE 3/8"T58 cf:t8 ~I UWli. Y...521 -I- -2 /~4 PW-15IN SL-2505R

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4394 3/4"TM58 W

                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      ,                                  HOSE ISOTOPIC ANALYZER LEGEND z
                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                     >                                                                                                                ,                                                      CANBERRA) 3/8/1
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IC)-INDICATES INTERFACE WITH PANEL MFR. if J 3/8"T58 NO.595(SL-2505R) < \t'8r VS -VENTILATION SYSTEM

                                                                                                                                                                                                                                                                                                                                                        ~
                                                                                                                                                                                                                                                                                                                                                                                                                                                                **                      r:N I,

994B 3/4"LINE #103 ... N2 OVERFLOW ~ 3/8"T58 " 4395 I") (FLUSH WATER) PW-506 4391 3/4"TM58 4390 I") 3/8"C58 3/8"C58 I n 3/4"LINE #782-I") SAMPLE

                                                                                                                                                                                                                                                                                                                                                                                                                                                    *i,,~ ,_. . . ....I-._.                    -(

If) 5162 3/ 8 " T58 3/8" T58 T FOR CONT.SEE 3 POINT .-L *

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v*,r __ __ __ III [:<<3 5 I6 I x..--- OWG.9321-F-2696

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3/4"LINE#782 MW-501 3/4"G22

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SL-2505R

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H PW-507 3/4"X 420

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ID I") iii 3/8"C58 3/8"T58 3/8"T58 DEMINERALIZER UNIT 3/4"T58 EXCESS FLOW VALVE ft I"LINE #462 FROM CVCS g:: 1-1/2"IA 2-------------------------------r---------------------( GAS FLOW _ _-.l 855 "' f-

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RESIDUAL HEAT If) o J-.w

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I") ID REMOVAL LOOP 9900 I-- 4887 5494 HOLD UP TKS VIA LINES '" INSTRUMENT AIR '" ! ~ INDICATOR / REGULATOR LIQUID If) o 1/2" +- (ACS) (A251783) #462 & 117 (9321-F-2737) W 3/8"LINE #799 (2505) "'! PI

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5339 W

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