ML20259A204

Defueled Safety Analysis Report, Chapter 3, Facility Design and Operation

ML20259A204
Person / Time
Site:	Indian Point
Issue date:	09/14/2020
From:	Entergy Nuclear Operations
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20259A199	List: ML20259A200 ML20259A201 ML20259A202 ML20259A203 ML20259A204 ML20259A205 ML20259A206 ML20259A207 ML20259A208 ML20259A209 ML20259A210 ML20259A211 ML20259A212 ML20259A213 ML20259A214 ML20259A215 ML20259A216 ML20259A217 NL-20-066, Engineering Report IP-RPT-11-00032, Revision 5, Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center
References
NL-20-066
Download: ML20259A204 (126)

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IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-1 of 3-126 Revision 0, 2020 CHAPTER 3 FACILITY DESIGN AND OPERATION 3.0 Introduction Section 3.1 discusses the nuclear fuel.

Sections 3.2 through 3.11 discuss the auxiliary systems required to ensure the safe operation or servicing of the spent fuel pit.

These sections consider systems in which component malfunctions, inadvertent interruptions of system operation, or a partial system failure may lead to a hazardous or unsafe condition. The extent of information provided for each system is proportional to the relative contribution of, or reliance placed upon, each system with respect to the overall facility operational safety.

The following systems are considered under this category:

Chemical and Volume Control System This system supports the transferring and storage of liquid radwaste.

Auxiliary Coolant System This system provides for transferring heat from the stored spent fuel and other components to the service water system and consists of the following two loops:

1. The spent fuel pit loop removes decay heat from the spent fuel pit.

2. The component cooling loop cools the spent fuel pit water.

Sampling System This system provides the equipment necessary to obtain liquid and gaseous samples from facility systems.

Facility Service Systems These systems include fire protection, service water, and auxiliary building ventilation.

Fuel Handling System This system provides for handling fuel assemblies.

Equipment and Decontamination Processes These processes provide for the removal of radioactive deposits from system surfaces.

Primary Auxiliary Building Ventilation System This system maintains ambient operation temperatures and provides purging of the auxiliary building to the plant vent.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-2 of 3-126 Revision 0, 2020 Control Room Ventilation System This system maintains the required environment in the control room.

Circulating Water System This system provides dilution flow for liquid waste discharges.

Section 3.12 discusses the leakage provisions regarding the cooling loops.

Section 3.13 discusses information displays and alarms.

Section 3.14 discusses the communications systems.

Section 3.15 discusses the Electrical Systems.

Section 3.16 discusses the containment systems.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-3 of 3-126 Revision 0, 2020 3.1 Nuclear Fuel The fuel rods are cold worked Zircaloy-4 or ZIRLOTM tubes containing slightly enriched uranium dioxide fuel.

The fuel assembly is a can-less type with the basic assembly consisting of the rod cluster control guide thimbles attached to the grids and the top and bottom nozzles. The fuel rods are held by the spring clip grid in this assembly, which provides support for the fuel rods.

High parasitic (HIPAR) fuel was used for the initial fuel and reload fuel through Cycle 4. Low parasitic (LOPAR) fuel was loaded for Cycles 5 through 9, and optimized fuel assemblies (OFA) were loaded for Cycles 10, 11, and 12. For Cycles 13, 14 and 15, 15x15 VANTAGE+ fuel assemblies were loaded as the feed fuel. For Cycle 16, 15x15 VANTAGE+ fuel assemblies with Performance+ enhancements were loaded as feed fuel. For Cycles 17 through 24, 15x15 Upgraded fuel design assemblies were loaded as feed fuel.

Rod cluster control assemblies and wet annular burnable absorber rods are inserted into the guide thimbles of the fuel assemblies. The absorber sections of the control rods are fabricated of silver-indium-cadmium alloy sealed in stainless steel tubes.

3.1.1 Fuel Assemblies The fuel assemblies are designed to perform satisfactorily throughout their lifetime. The assemblies are structurally designed to maintain sufficient integrity to permit safe removal from the core, subsequent handling during cooldown, shipment, and fuel reprocessing.

The fuel rods are supported at nine locations along their length within the fuel assemblies by grid assemblies, which are designed to maintain control of the lateral spacing between the rods through the design life of the assemblies. The magnitude of the support loads provided by the grids is established to minimize possible fretting without overstressing the cladding at the points of contact between the grids and fuel rods and without imposing restraints of sufficient magnitude to result in buckling or distortion of the rods. In addition, there are three Intermediate Flow Mixing (IFM) grids spaced along the fuel assembly and a protective grid (P-grid) on the bottom of the assembly. These grids do not provide any support function.

The fuel rod cladding is designed to maintain encapsulation of the fuel throughout the design life.

3.1.2 Rod Cluster Control Assemblies The criteria used for the design of the cladding on the individual absorber rods in the rod cluster control assemblies are similar to those used for the fuel rod cladding. The cladding is designed to be free standing under all operating conditions and will maintain encapsulation of the absorber material throughout the absorber rod design life.

3.1.3 Mechanical Design The fuel is in the form of slightly enriched uranium dioxide ceramic pellets. The pellets are stacked to an active height of 144-in. (previously 142 in.) within ZIRLOTM (previously Zircaloy-4) tubular cladding, which is plugged and seal welded at the ends to encapsulate the fuel.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-4 of 3-126 Revision 0, 2020 The control rods, designated as rod cluster control assemblies, consist of groups of individual absorber rods, which are held together by a spider assembly at the top end and actuated as a group. In the inserted position, the absorber rods fit within hollow guide thimbles in the fuel assemblies. The guide thimbles are an integral part of the fuel assemblies and occupy locations within the regular fuel rod pattern where fuel rods have been deleted. Figures 3.1-1 and 3.1-2 show a typical rod cluster control assembly in a fuel assembly.

3.1.3.1 Fuel Assembly The assemblies are square in cross section, nominally 8.426-in. on a side, and have an overall height of approximately 159.975 inches. The fuel rods in a fuel assembly are arranged in a square array with 15 rod locations per side and a nominal centerline-to-centerline pitch of 0.563-in.

between rods. Of the total possible 225 rod locations per assembly, 20 are occupied by guide thimbles for the rod cluster control rods and one for incore instrumentation. The remaining 204 locations contain fuel rods. In addition to fuel rods, a fuel assembly is composed of a top nozzle, a bottom nozzle, grid assemblies, 20 absorber rod guide thimbles, and one instrumentation thimble. Additional information regarding the fuel rods and fuel assemblies is provided in Table 3.1-1.

The guide thimbles in conjunction with the grid assemblies and the top and bottom nozzles comprise the basic structural fuel assembly skeleton. The grid assemblies are bulge attached to the guide thimbles at each location along the height of the fuel assembly at which lateral support for the fuel rods is required. Within this skeletal framework the fuel rods are contained and supported and the rod-to-rod centerline spacing is maintained along the assembly.

The original fuel design for Indian Point 2 was the Westinghouse High Parasitic (HIPAR) fuel assembly (Figure 3.1-4). This consisted of Zircaloy-4 clad fuel rods, nine Inconel grids and stainless steel instrumentation and guide thimbles. Burnable absorbers used were Pyrex glass.

Starting with Cycle 5, the Westinghouse Low Parasitic (LOPAR) fuel assembly (Figure 3.1-5) was introduced. This design consisted of Zircaloy-4 clad fuel rods, nine Inconel grids and Zircaloy-4 instrumentation and guide thimbles.

For Cycle 8, Wet Annular Burnable Absorbers (WABA) were introduced.

For Cycle 10, the Westinghouse Optimized Fuel Assembly (OFA) (Figure 3.1-6) was introduced.

This consisted of Zircaloy-4 clad fuel rods, two Inconel grids (top & bottom), seven Zircaloy-4 grids and Zircaloy-4 instrumentation and guide thimbles. In addition, thimble plugs were removed from the core this cycle based on analysis performed to support removal. The assembly top nozzle design was changed to a Reconstitutable Top Nozzle (RTN) design to facilitate reconstitution of failed fuel.

For Cycle 11, the OFA fuel assembly design incorporated Debris Filter Bottom Nozzles (DFBN) and Integral Fuel Burnable Absorbers (IFBA).

For Cycle 13, the Westinghouse VANTAGE+ fuel design was introduced (see Figures 3.1-3 and 3.1-7). This design included ZIRLO clad fuel rods, two Inconel grids, seven low pressure drop (LPD) Zircaloy-4 grids, three Zircaloy-4 Integral Flow Mixing grids (IFM), ZIRLO instrumentation and guide thimbles, annular axial blankets along with the DFBN, IFBA and RTN. Use of WABAs was continued.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-5 of 3-126 Revision 0, 2020 For Cycle 15, the VANTAGE+ fuel assembly design incorporated Performance+ features of ZIRLO grids and IFMs and a hardened coating of zirconium oxide on the bottom section of the fuel rod clad to increase debris resistance.

For Cycle 16, the VANTAGE+ fuel assembly design was further enhanced with Performance+

features that include debris mitigation features of an additional grid located at the bottom end plug of the fuel rod, a longer fuel rod end plug and a revised DFBN. Other Performance+

enhancements include longer fuel rods and longer annular axial blanket (see Figure 3.1-15).

In addition to the above fuel design changes, the maximum rod average burnup was increased from 60,000 MWD/MTU to 62,000 MWD/MTU starting in Cycle 16.

For Cycle 17, the 15x15 Upgraded fuel assembly design was used (Figure 3.1-16). This design has features to address grid-to-rod fretting fuel failures. These include I-spring mid-grids, enhanced IFMs and balanced mixing vanes. In addition, the tube-in-tube thimble design was incorporated with a single-dashpot, which improves straightness.

For Cycle 18, solid axial blanket pellets were introduced for the non-IFBA fuel rods.

For Cycle 19, the top nozzle spring design was changed from the VANTAGE+ design to the standard spring design.

Cycle 20 and 21 fuel was the same as Cycle 19, there were no fuel design changes.

Cycle 22 uses the 15x15 Upgraded design with changes to the bottom nozzle and the protective grid. Five flow holes on each side of the bottom nozzle were removed to eliminate possible debris intrusion into the fuel through the holes. It is now the modified Debris Filter Bottom Nozzle (mDFBN). The manufacturing of the protective grid was changed to prevent dimple cracking. It is now the Robust Protective Grid (RPG). In addition, secondary sources were removed from the core.

Cycle 23 introduced the Westinghouse Integral Nozzle (WIN) top nozzle design replacing the previous RTN design that is susceptible to stress-corrosion cracking of the hold-down screws.

Bottom Nozzle Two types of nozzle designs were used for the HIPAR fuel assemblies. One type, which is square in cross section, is fabricated from type 304 stainless steel consisting of four side plates, 12 cross bars and four pads or feet. The side plates are welded together at the corners to form a plenum for inlet coolant to the fuel assembly. The cross bars are welded at each end to the top edges of the side plate and function as the bottom end support for the fuel rods. The bottom support surface for the fuel assembly is formed by the four pads, which are welded to the side plates in the corners. This design was used in a majority of the first core fuel assemblies. The previously used LOPAR and OFA fuel incorporate an equivalent bottom nozzle design utilizing a square perforated plate rather than the cross bars and side plate. On both designs, their respective cross bars and perforated plate prevent the fuel rods from falling through the bottom nozzles of the assembly.

Coolant flow to the fuel assembly is directed from the plenum in the bottom nozzle upward to the interior of the fuel assembly and to the channel between assemblies.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-6 of 3-126 Revision 0, 2020 Axial loads imposed on the assembly, as well as the weight of the assembly are distributed through the guide thimbles and the bottom nozzle to the lower core support plate. Indexing and positioning of the fuel assembly in the core is controlled through two holes in diagonally opposite pads, which mate with locating pins in the lower core plate. Lateral loads imposed on the fuel assembly are also transferred to the core support structures through the locating pins.

The OFA and VANTAGE+ bottom nozzle used the reconstitutable feature found on the previously installed LOPAR fuel design, which uses a locking cup to lock the thimble screws on the guide thimble assembly, instead of the lockwire used in earlier LOPAR designs. The OFA nozzle assembly is shorter when compared to the previously installed LOPAR assembly to enhance fuel rod growth allowances.

The two bottom nozzle designs used in the OFAs are both square in cross section and fabricated from 304 stainless steel. The design used in earlier regions consists of a perforated plate, four angle legs, and four pads of feet. The angle legs are fastened to the plate forming a plenum space for the coolant inlet to the fuel assembly.

The remaining OFA regions and the VANTAGE+ and 15x15 Upgraded fuel regions (starting with Cycle 13, Region 15) incorporate an equivalent bottom nozzle design denoted as the Debris Filter Bottom Nozzle (DFBN). This nozzle adds side plates or "skirts" to the previous design increasing structural capability for abnormal loads and providing a more defined plenum space below the nozzle. Additionally, the relatively large adapter plate flow holes of the earlier design are replaced with a new pattern of smaller flow holes. The decrease in size of the holes provides a "screen" for larger debris particles, which would otherwise cause damage if allowed to pass into the assembly.

In both designs, the adaptor plates prevent accidental downward ejection of the fuel rods from the fuel assembly. The nozzles are fastened to the assembly guide tubes by stainless steel screws, which penetrate through the nozzle and mate with a threaded plug in each guide tube (Figure 3.1-8). The screw possesses a circular locking cup around the screw head, which is crimped into mating detents (lobes) on the bottom nozzle, preventing the screw from loosening.

The DFBN was modified starting with Cycle 22 to eliminate five holes on each side (mDFBN) to eliminate the potential for intrusion of debris.

Top Nozzle The Reconstitutable Top Nozzle (RTN) used in OFA, VANTAGE+ and 15x15 Upgraded fuel assemblies is a box-like structure, which functions as the fuel assembly upper structural element and forms a plenum space where the heated fuel assembly discharge coolant is mixed and directed toward the flow holes in the upper core plate. The nozzle is comprised of an adaptor plate enclosure, top plate, clamps, hold-down leaf springs and assorted hardware. Each nozzle has four sets of leaf springs. All parts, with the exception of the springs and their hold-down bolts/screws, are constructed of type 304 stainless steel. The springs are made from age hardenable Inconel 718 and the bolts/screws from Inconel 600 for Region 16 and earlier regions, and from shot peened Inconel 718 for Regions 17 through 24. The bolts/screws were eliminated with the WIN design.

The adaptor plate portion of the nozzle is square in cross section, and is perforated by machined slots to provide for coolant flow through the plate. At assembly, the top ends of the LOPAR thimble stainless sleeves are fitted through individual bored holes in the plate and welded to the plate

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-7 of 3-126 Revision 0, 2020 around the circumference of each hole. In the OFA removable top nozzle design, a groove is provided in each thimble thru-hole in the nozzle plate into which a stainless steel nozzle insert is mechanically connected by means of a preformed circumferential bulge near the top of the insert.

Thus, the adaptor plate acts as the fuel assembly top end plate, and provides a means of distributing evenly among the guide thimbles any axial loads imposed on the fuel assemblies.

The nozzle enclosure is actually a square tubular structure, which forms the plenum section of the top nozzle. The bottom end of the enclosure is pinned and welded to the periphery of the adaptor plate and the top end is welded to the periphery of the top plate.

The top plate is square in cross section with a square central hole. The hole allows clearance for the rod cluster control absorber rods to pass through the nozzle into the guide thimbles in the fuel assembly and for coolant exit from the fuel assembly to the upper internals area. Two pads containing axial through-holes, which are located on diametrically opposite corners of the top plate provide a means of positioning and aligning the top of the fuel assembly. As with the bottom nozzle, alignment pins in the upper core plate mate with the holes in the top nozzle plate. Hold-down forces of sufficient magnitude to oppose the hydraulic lifting forces on the fuel assembly are obtained by means of the leaf spring sets, which are mounted on the top plate. The springs are fastened in pairs to the top plate at the two corners where alignment holes are not used and radiate out from the corners parallel to the sides of the plate. Prior to the WIN design, fastening of each pair of springs is accomplished with a clamp, which fits over the ends of the springs and two bolts/screws (one per spring set), which pass through the clamp and spring, and thread into the top plate. At assembly, the spring mounting bolts/screws are torqued sufficiently to preload against the maximum spring load and then lockwelded to the clamp, which is counterbored to receive the bolt/screw head. The spring load is obtained through deflection of the spring pack by the upper core plate. The spring pack form is such that it projects above the fuel assembly and is depressed by the core plate when the internals are loaded into the reactor. The free end of the spring pack is bent downward and captured in a key slot in the top plate to guard against loose parts in the reactor in the event (however remote) of spring fracture. The capture of the loose end has been deleted in latter designs.

Starting with Cycle 14, Region 16, the fuel has a cast top nozzle. This is a two-piece design incorporating a machined stainless steel adapter plate welded to a low-cobalt investment casting.

The cast top nozzle is functionally interchangeable with the previous design and meets design criteria for the top nozzle.

Westinghouse has developed a top nozzle design to eliminate the potential of generating a loose part from fracture of the fuel assembly holddown spring screws. This design is called the Westinghouse Integral Nozzle (WIN). The WIN uses the same basic two piece nozzle as the standard design except that the top nozzle casting has been modified to include an integral pad in place of the previously separate clamp. As the name implies, these pads are cast as integral parts of the top nozzle casting. The WIN springs includes manufacturing process modifications for added margin against primary water stress corrosion cracking. Unlike previous bolted designs, the WIN design provides a wedged rather than a clamped joint. The tails of the spring packs are a slip-fit interface in the respective clamp pad cavities. Once preloaded, each spring pack is effectively locked in place during normal operation by the reaction forces generated at its tail. The flow holes, thermal characteristics, and method of attachment are all unchanged from the previous top nozzle design. The design was first introduced with the Cycle 23 Region 25 15x15 Upgraded fuel assemblies.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-8 of 3-126 Revision 0, 2020 In addition to its plenum and structural functions, the nozzle provides a protective housing for components, which mate with the fuel assembly. In handling a fuel assembly with a control rod inserted, the control rod spider is contained within the nozzle. During operation in the reactor, the nozzle protects the absorber rods from coolant cross flows in the unsupported span between the fuel assembly adaptor plate and the end of the guide tube in the upper internals package.

Plugging devices, which fill the ends of the fuel assembly thimble tubes at unrodded core locations, neutron source rods and burnable absorber rods are all contained within the fuel top nozzle.

For the RTN and WIN designs, a stainless steel nozzle insert is mechanically connected to the top nozzle adaptor plate (Figure 3.1-10) via the engagement of the preformed circumferential bulge near the top of the insert and the mating groove in the wall of the adapter plate thimble tube through-hole. The insert has four equally spaced axial slots, which allow the insert to deflect inwardly at the elevation of the bulge, thus permitting the installation and removal of the nozzle.

The insert bulge is positively held in the adapter plate mating groove by placing a lock tube with a uniform OD identical to that of the thimble tube into the insert. The lock tube is secured in place by a top flare, which creates a tight fit and six non-yielding projections on the OD, which interface with the concave side of the insert to preclude escape during core component transfer. The adaptor plate acts as the fuel assembly top end plate and provides a means of evenly distributing any axial loads imposed on the fuel assemblies to the guide thimbles.

Guide Thimbles The control rod guide thimbles in the fuel assemblies provide guided channels for the absorber rods during insertion and withdrawal of the control rods. Up to and including Region 18 (VANTAGE+), they are fabricated from a single piece of tubing, which is drawn to two different diameters. The OFA thimbles are made of Zircaloy-4 and the VANTAGE+ thimbles are made of ZIRLOTM. The larger inside diameter at the top provides a relatively large annular area for rapid insertion during a reactor trip and accommodates a small amount of upward cooling flow during normal operations. The bottom portion of the guide thimble has two sections of reduced diameter producing a "double dashpot" action when the absorber rods near the end of travel in the guide thimbles during a reactor trip. The transition zones at the dashpot sections are conical in shape so that there are no rapid changes in diameter in the tube.

Starting with Region 19 (15x15 Upgraded fuel design), the guide thimbles incorporate the tube-in-tube dashpot design. The tube-in-tube design utilizes a separate dashpot tube assembly that is inserted into the guide thimble assembly pulled to a press fit over the thimble end plug and bulged into place. To maintain the same diametrical clearance between the guide thimble ID and the dashpot OD, the 15x15 upgraded nominal dashpot thickness was reduced from 0.0165 to 0.0160 inches. As the dashpot tube in the design can provide additional lateral support in that bottom thimble span, it is expected that there will be additional resistance to lateral deformation and Incomplete Rod Insertions as a result of the design modification. The 15x15 Upgraded fuel thimbles are made of ZIRLOTM.

Flow holes are provided just above the first dashpot transition to permit the entrance of cooling water during normal operation, and to accommodate the outflow of water from the dashpot during reactor trip.

The dashpot is open at the bottom by means of the drainage hole in the thimble screws that secure the bottom nozzle to the welded end plugs of the guide thimbles. This geometry is shown in Figure 3.1-8.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-9 of 3-126 Revision 0, 2020 The top ends of the thimble tubes are mechanically attached to the sleeve of the top grid. An insert is also bulge attached to the thimble and the insert upper end is in turn mechanically attached to the top nozzle as shown in Figure 3.1-10.

VANTAGE+ Grids Prior to Region 18, the VANTAGE+ assembly has twelve grids. Starting with Region 18, a thirteenth grid, the protective grid (P-grid), was added to the VANTAGE+ assembly. The top and bottom grids, as in the OFA assembly, and the P-grid are Inconel 718 non-mixing vane grids.

Low Pressure Drop (LPD) Zircaloy-4 grids are used for the middle grids with Zircaloy-4 IFMs located in the three uppermost middle grid spans. The VANTAGE+ fuel assembly with Performance+ options has ZIRLO grids for the three IFMs and seven mid grids. The LPD grids have mixing vanes, diagonal springs and a reduced grid height, relative to the OFA grids. The LPD grid cells use the standard four dimples and two springs per cell for support locations. The IFMs provide mid-span flow mixing in the hottest fuel assembly spans. Each IFM cell contains four dimples, which are designed to prevent midspan channel closure and fuel rod contact with the mixing vanes. With the additional Performance+ enhancements added to the fuel starting with Region 18, a new Protective Bottom Grid (PBG) has been added. The PBG is a wider, extra grid at the very bottom of the fuel assembly that protects the fuel from debris. Its purpose is to filter out debris and hold it at an elevation below the bottom of the active core. The PBG is not a structural grid. The bottom of the PBG lies below the tops of the lower end plugs within the fuel rod. This means that any debris caught in the PBG cannot fret through the cladding and expose fuel pellets.

All VANTAGE+ outside grid straps contain mixing vanes, which also act as guides during fuel handling. The grids are also attached to the thimble tubes via the bulging mechanism as shown in Figure 3.1-14. Top grid nozzle attachment is shown in Figure 3.1-10. All grids employ the anti-snag outer strap design. A mixing vane grid is shown in Figure 3.1-11.

15x15 Upgraded Design Grids The 15x15 Upgraded fuel design still contains twelve grids with the top and bottom grids unchanged from the VANTAGE+ design. The thirteenth grid, the protective grid (P-grid) also remains the same as the VANTAGE+ design. The middle grids have changed to an I-spring design. The changes were made to improve fuel rod fretting margin. In addition to the spring change the size of the dimples was increased. The strap thickness was decreased to help offset the pressure drop increase due to the I-spring and increased grid strap height. The strap height increased to create space to accommodate the increased dimples and the I-spring. The IFM grid design was enhanced to increase contact area also.

The 15x15 Upgraded design protective grid has been redesigned for Cycle 22 to reduce stresses that caused dimple cracking. The Robust Protective Grid (RPG) dimensions changed and vibration mitigation features were added.

Fuel Rods The fuel rods consist of uranium dioxide ceramic pellets in slightly cold worked ZIRLOTM tubing, which is plugged and seal welded at the ends to encapsulate the fuel. Sufficient void volume and clearances are provided within the rod to accommodate fission gases released from the fuel, differential thermal expansion between the cladding and the fuel, helium released from poison

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-10 of 3-126 Revision 0, 2020 burnup (IFBA rods), and fuel swelling due to accumulated fission products without overstressing of the cladding or seal welds. Shifting of the fuel within the cladding is prevented during handling or shipping prior to core loading by a stainless steel helical compression spring, which bears on the top of the fuel.

At assembly, the pellets are stacked in the cladding to the required fuel height. The compression spring is then inserted into the top end of the fuel and the end plugs pressed into the ends of the tube and welded. A hold-down force of approximately four times the weight of the fuel is obtained by compression of the spring between the top end plug and the top of the fuel pellet stack. All fuel rods are internally pressurized with helium in order to minimize compressive clad stresses and creep due to coolant operating pressures.

The fuel pellets are in the form of a right circular cylinder and consist of slightly enriched uranium dioxide powder, which is compacted by cold pressing and sintering to the required density. The ends of each pellet are dished slightly to allow the greater axial expansion at the center of the pellets to be taken up within the pellets themselves and not in the overall fuel length. The 15x15 Upgraded fuel has mid-enriched annular (IFBA) and solid (non-IFBA) pellets in the axial blanket region of the fuel rod and optimized plenum spring to maximize the available plenum volume for increased burnup. The 15x15 Upgraded fuel has a longer fuel rod to allow higher fission gas release due to longer cycles. This is allowable due to the ZIRLO cladding, which has less rod growth on irradiation.

Each fuel rod is marked with a permanent traceability code. This aids in ensuring that rods of the proper enrichment will be loaded into each fuel assembly. The identification numbers on the fuel assembly top nozzles will then maintain the enrichment identity.

3.1.3.2 Rod Cluster Control Assemblies The rod cluster control assemblies remain inserted in fuel assemblies stored in the Spent Fuel Pit (SFP) or Independent Spent Fuel Storage Installation (ISFSI).

The control rods or rod cluster control assemblies consist of a group of individual absorber rods fastened at the top end to a common hub or spider assembly. These assemblies, one of which is shown in Figure 3.1-1, were provided to control the reactivity of the core under operating conditions. The absorber material used in the control rods is silver-indium-cadmium alloy, which is essentially "black" to thermal neutrons and has sufficient additional resonance absorption to increase its worth significantly. The alloy is in the form of extruded single-length rods, which are sealed in stainless steel tubes to prevent the rods from coming in direct contact with the coolant.

Additional information regarding the rod cluster control assemblies is provided in Table 3.1-1.

The overall control rod length is such that when the assembly has been withdrawn through its full travel, the tips of the absorber rods remain engaged in the guide thimbles so that alignment between rods and thimbles is always maintained. Since the rods are long and slender, they are relatively free to conform to any small misalignments with the guide thimble. Prototype tests have shown that the rod cluster control assemblies are very easily inserted and not subject to binding even under conditions of severe misalignment.

The spider assembly is in the form of a center hub with radial vanes supporting cylindrical fingers from which the absorber rods are suspended. Handling detents and detents for connection to the drive shaft are machined into the upper end of the hub. A spring pack is assembled into a skirt integral to the bottom of the hub to stop the rod cluster control assembly and absorb the energy

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-11 of 3-126 Revision 0, 2020 from the impact at the end of a trip insertion. The radial vanes are joined to the hub and the fingers are joined to the vanes by furnace brazing. A center-post, which holds the spring pack and its retainer is threaded into the hub within the skirt and welded to prevent loosening in service.

All components of the spider assembly are made from type 304 stainless steel except for the springs, which are Inconel X-750 alloy and the retainer, which is of 17-4 PH material.

The absorber rods are secured to the spider so as to ensure trouble free service. The rods are first threaded into the spider fingers and then pinned to maintain joint tightness, after which the pins are welded in place. The end plug below the pin position is designed with a reduced section to permit flexing of the rods to correct for small operating or assembly misalignments.

In construction, the silver-indium-cadmium rods are inserted into cold-worked stainless steel tubing, which is then sealed at the bottom and the top by welded end plugs. Sufficient diametral and end clearance are provided to accommodate relative thermal expansions and to limit the internal pressure to acceptable levels.

The bottom plugs are made bullet-nosed to reduce the hydraulic drag during a reactor trip and to guide smoothly into the dashpot section of the fuel assembly guide thimbles. The upper plug is threaded for assembly to the spider and has a reduced end section to make the joint more flexible.

Stainless steel clad silver-indium-cadmium alloy absorber rods are resistant to radiation and thermal damage thereby ensuring their effectiveness under all operating conditions. Rods of similar design have been successfully used in a number of operating nuclear plants.

3.1.3.3 Neutron Source Assemblies Neutron source assemblies remain inserted in fuel assemblies stored in the SFP or ISFSI.

Six neutron source assemblies were utilized in the first cycle core. These consisted of two assemblies with four secondary source rods each, and four assemblies with one secondary source rod and one primary source rod each. The rods in each assembly were fastened to a spider at the top end. The spider for the four secondary source rod assemblies was similar to the rod cluster control assembly spiders, while the primary source assembly spider was similar to that of the burnable poison and plugging device assemblies. Various source assembly designs are used in the reload cycles.

In the first cycle core, the neutron source assemblies were inserted into the rod cluster control guide thimbles in fuel assemblies at unrodded locations. The location and orientation of each of the assemblies in the core is shown in Figure 3.1-17.

The primary and secondary source rods both utilize the same type of cladding material as the absorber rods (cold-worked type 304 stainless steel tubing). The secondary source rods contain Sb-Be pellets. The primary source rods contained capsules of Pu-Be source material in the initial core loading; for reload cores, this material may vary. Design criteria similar to that for the fuel rods is used for the design of the source rods; i.e., the cladding is free standing, internal pressures are always less than reactor operating pressure, and internal gaps and clearances are provided to allow for differential expansions between the source material and cladding.

Starting in Cycle 22, secondary sources were removed from the core.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-12 of 3-126 Revision 0, 2020 3.1.3.4 Plugging Devices Plugging devices remain inserted in fuel assemblies stored in the SFP or ISFSI.

In order to limit bypass flow through the rod cluster control guide thimbles in fuel assemblies, which do not contain either control rods, source assemblies or burnable absorber rods, the fuel assemblies at those locations were fitted with plugging devices. The plugging devices consist of a flat plate with short rods suspended from the bottom surface and a spring pack assembly. The plugging devices fit within the fuel assembly top nozzles and rest on the adaptor plate. The short rods project into the upper ends of the thimble tubes to reduce the bypass flow area. The spring pack is compressed by the upper core plate when the upper internals package is lowered into place. Similar short rods are also used on the source assemblies to fill the ends of all vacant fuel assembly guide thimbles. All components in the plugging device, except for the springs, are constructed from type 304 stainless steel. The springs are wound from an age hardenable nickel base alloy to obtain higher strength.

Coincident with implementation of the Indian Point Unit 2 OFA transition, removal of thimble plugging devices from the core was allowed. This included the removal of the thimble plugs from the OFA assemblies, previously installed LOPAR assemblies, and all new core component clusters (burnable absorbers and sources).

Thimble plugs were reinstalled for all assemblies without a designated insert (e.g. RCCA, WABA, or secondary source) in Cycles 17, 18, 19 and 20 to satisfy stretch power uprate conditions.

Starting in Cycle 21, thimble plugs were again removed from the core.

3.1.3.5 Burnable Absorber Rods Burnable absorber rods remain inserted in fuel assemblies stored in the SFP or ISFSI.

The burnable absorber rods are statically suspended and positioned in vacant rod cluster control thimble tubes within the fuel assemblies at nonrodded core locations. The absorber rods in each fuel assembly are grouped and attached together at the top end of the rods by a flat plate, which fits with the fuel assembly top nozzle and rests on the top adaptor plate.

The plate with the absorber rods is held down and restrained against vertical motion with a spring pack, which is attached to the plate and is compressed by the upper core plate when the reactor upper internals package is lowered into the reactor. Historically, this ensured that the absorber rods cannot be lifted out of the core by flow forces.

The absorber rods used during Cycles 1 through 7 consisted of borated Pyrex glass tubes contained within type 304 stainless steel tubular cladding, which was plugged and seal welded at the ends to encapsulate the glass (Figures 3.1-18 and 3.1-19). The glass was also supported along the length of its inside diameter by a thin-wall type 304 stainless steel tubular inner liner.

Starting in Cycle 8, Wet Annular Burnable Absorber (WABA) rods were used. As shown in Figures 3.1-20 and 3.1-21, WABA rods are composed of annular pellets containing aluminum oxide-boron carbide (Al2O3 - B4C) burnable absorber material contained within two concentric Zircaloy-4 tubes.

The Zircaloy-4 tubes are plugged and seal welded at the ends to enclose the annular stack of absorber material. The tubes are also the inner and outer cladding of the annular burnable absorber rod. A hold-down device is placed on top of the pellet stack to hold the stack in position and to allow for pellet stack growth. The hold-down device is a C-shape Zircaloy polygonal ring

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-13 of 3-126 Revision 0, 2020 clip. Within the rod is an annular plenum to allow for helium gas release from the absorber material during boron depletion. During operation, reactor coolant flowed through the inner tube and outside the outer tube of the annular rod. The annular rods are grouped and attached at the top end to a hold-down assembly and retaining plate in the same way as the borosilicate glass absorber rod. WABA rods are used in preference to standard BPRAs to provide smaller residual burnup penalty.

Starting with Cycle 11, Integral Fuel Burnable Absorbers (IFBA) were used in conjunction with the WABA rods. The IFBA features a zirconium diboride coating on the fuel pellet surface on the central portion of the enriched UO2 pellets. IFBA provided power peaking and moderator temperature coefficient control.

Additional information regarding the burnable absorber rods is provided in Table 3.1-1.

3.1.4 Evaluation 3.1.4.1 Fuel Evaluation To assure that manufactured fuel rods meet a high standard of excellence from the standpoint of functional requirements, many inspections and tests are performed both on the raw material and the finished product. These tests and inspections include chemical analysis, tensile testing of fuel tubes, dimensional inspection, X-ray of both end plug welds, ultrasonic testing, and helium leak tests.

3.1.4.2 Fuel Assembly and Rod Cluster Control Assembly Mechanical Evaluation Axial and lateral bending tests have been performed in order to simulate mechanical loading of the assembly. Although the maximum column load expected to be experienced in service is approximately 1000 lb., the fuel assembly was successfully loaded to 2200 lb. axially with no damage resulting. This information is also used in the design of fuel handling equipment to establish the limits for inadvertent axial loads during fuel handling.

3.1.5 Quality Assurance Program The quality assurance program plan of the Westinghouse Nuclear Fuel Division is summarized in Reference 3.1-1.

The program provides for control over all activities affecting product quality, commencing with design and development and continuing through procurement, materials handling, fabrication, testing and inspection, storage, and transportation. The program also provides for the indoctrination and training of personnel and for the auditing of activities affecting product quality through a formal auditing program.

Westinghouse drawings and product, process, and materials specifications identify the inspections to be performed.

3.1.6 Quality Control Quality control philosophy is generally based on the following inspections being performed to a 95-percent confidence that at least 95-percent of the product meets specification, unless otherwise noted.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-14 of 3-126 Revision 0, 2020

1.

Fuel system components and parts The characteristics inspected depend upon the component parts; the quality control program includes dimensional and visual examinations, check audits of test reports, material certification, and nondestructive examination, such as X-ray and ultrasonic.

All material used in the fuel assembly is accepted and released by quality control.

2.

Pellets Inspection is performed for dimensional characteristics such as diameter, density, length, and squareness of ends. Additional visual inspections are performed for cracks, chips, and surface conditions according to approved standards.

Density is determined in terms of weight per unit length and is plotted on zone charts used in controlling the process. Chemical analyses are performed on a specified sample basis throughout pellet production.

3.

Rod inspection The fuel rod inspection consists of the following nondestructive examination techniques and methods, as applicable:

a. Each rod is leak tested using a calibrated mass spectrometer, with helium being the detectable gas.

b. Rod welds are inspected by ultrasonic test or X-ray in accordance with a qualified technique and Westinghouse specification.

c. All rods are dimensionally inspected prior to final release. The requirements include such items as length, camber, and visual appearance.

d. All fuel rods are inspected by gamma scanning or other approved methods to ensure proper plenum dimensions.

e. All fuel rods are inspected by gamma scanning, or other approved methods to ensure that no significant gaps exist between pellets.

f.

All fuel rods are active gamma scanned to verify enrichment control prior to acceptance for assembly loading.

g. Traceability of rods and associated rod components is established by quality control.

4.

Assemblies Each fuel assembly is inspected for compliance with drawing and/or specification requirements.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-15 of 3-126 Revision 0, 2020

5.

Other inspections The following inspections are performed as part of the routine inspection operation:

a. Tool and gauge inspection and control, including standardization to primary and/or secondary working standards. Tool inspection is performed at prescribed intervals on all serialized tools. Complete records are kept of calibration and conditions of tools.

b. Audits are performed of inspection activities and records to ensure that prescribed methods are followed and that records are correct and properly maintained.

c. Surveillance inspection, where appropriate, and audits of outside contractors are performed to ensure conformance with specified requirements.

6.

Process control To prevent the possibility of mixing enrichments during fuel manufacture and assembly, strict enrichment segregation and other process controls are exercised.

The uranium dioxide powder is kept in sealed containers. The contents are fully identified both by descriptive tagging and preselected color coding. A Westinghouse identification tag completely describing the contents is affixed to the containers before transfer to powder storage. Isotopic content is confirmed by analysis.

Powder withdrawal from storage can be made by only one authorized group, which directs the powder to the correct pellet production line. All pellet production lines are physically separated from each other and pellets of only a single nominal enrichment and density are produced in a given production line at any given time.

Finished pellets are placed on trays identified with the same color code as the powder containers and transferred to segregated storage racks within the confines of the pelleting area. Samples from each pellet lot are tested for isotopic content and impurity levels prior to acceptance by quality control. Physical barriers prevent mixing of pellets of different nominal densities and enrichments in this storage area. Unused powder and substandard pellets are returned to storage in the original color-coded containers.

Loading of pellets into the clad is performed in isolated production lines, and again only one enrichment and density loaded on a line at a time.

A serialized traceability code is placed on each fuel tube to provide unique identification.

The end plugs are inserted and then inert-welded to seal the tube. The fuel tube remains coded and traceability identified until just prior to installation in the fuel assembly.

At the time of installation into an assembly, the traceability codes are removed and a matrix is generated to identify each rod in its position within a given assembly. The top nozzle is inscribed with a permanent identification number providing traceability to the fuel contained in the assembly.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-16 of 3-126 Revision 0, 2020 REFERENCES FOR SECTION 3.1

1.

J. Moore, "Nuclear Fuel Division Quality Assurance Program Plan;" WCAP-7800, Revision 5, Westinghouse Electric Corporation, November 1979.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-17 of 3-126 Revision 0, 2020 TABLE 3.1-1 Core Mechanical Design Parameters1 Fuel assemblies Rod array 15 x 15 Rods per assembly 2042 Rod pitch, in.

0.563 Overall dimensions 8.426 x 8.426 HIPAR/LOPAR 8.424 x 8.424 OFA 8.426 x 8.426 VANTAGE+/15x15 Upgraded Number of grids per assembly (HIPAR/LOPAR/OFA)

(VANTAGE+)

(15x15 Upgraded) 9 12 or 13 13 Number of instrumentation thimbles 1

Number of guide thimbles 20 Diameter of guide thimbles, upper part, in., HIPAR 0.545 O.D. x 0.515 I.D.

Diameter of guide thimbles, lower part, in., HIPAR 0.484 O.D. x 0.454 I.D.

Diameter of guide thimbles, upper part, in., LOPAR 0.546 O.D. x 0.512 I.D.

Diameter of guide thimbles, lower part, in., LOPAR 0.489 O.D. x 0.455 I.D.

Diameter of guide thimbles, upper part, in., OFA/V+

0.533 O.D. x 0.499 I.D.

Diameter of guide thimbles, lower part, in., OFA/V+

0.489 O.D. x 0.455 I.D.

Diameter of guide thimbles, upper part, in., 15x15 Upgraded 0.533 O.D. x 0.499 I.D.

Diameter of guide thimbles, lower part, in., 15x15 Upgraded 0.487 O.D. x 0.455 I.D.

Fuel rods Outside diameter, in.

0.422 Diametral gap, in.

0.0075 Clad thickness, in.

0.0243 Clad material Zircaloy-4 (HIPAR/LOPAR/OFA)

ZIRLOTM (VANTAGE+/15x15 Upgraded)

Overall length 148.6, HIPAR 151.9, LOPAR 152.17, OFA 152.55, VANTAGE+

152.88, V+ w/P+

Enhancements/15x15 Upgraded Length of end cap, overall, in.

0.688, HIPAR 0.265, LOPAR 0.357, OFA/V+ (TOP) 0.430, OFA/V+ (BOTTOM) 0.350, V+w/P+/15x15 Upgraded (TOP) 0.810, V+w/P+/15x15 Upgraded (BOTTOM)

Length of end cap, inserted in rod 0.250, HIPAR 0.200, LOPAR 0.130, OFA/V+/15x15 Upgraded Active fuel length, in.

142, HIPAR

144, LOPAR

144, OFA/V+/15x15 Upgraded

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-18 of 3-126 Revision 0, 2020 TABLE 3.1-1 (Cont.)

Core Mechanical Design Parameters1 Rod cluster control assemblies Neutron absorber 5-percent Cd, 15-percent In, 80-percent Ag Cladding material Type 304 SS - cold worked Clad thickness, in.

0.019 Number of control rods per cluster 20 Length of rod control, in.

156.436 (overall)

Length of absorber section, in.

142.00 Wet Annular Burnable Absorber (WABA) Rods Pellet Material Al2O3-B4C Boron Loading (Natural) 0.0243 g/cm (B-10) 0.0060 g/cm Pellet O.D. /I.D.

0.318"/0.278" Tube material Zircaloy-4 Outer tube O.D. /I.D.

0.3810"/0.3290" Inner tube O.D. /I.D.

0.2670"/0.2250" Notes:

1. All dimensions are for cold conditions. Data is for all fuel types unless otherwise stated.

2. Twenty-one rods are omitted: Twenty provide passage for control rods and one contains incore instrumentation.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-19 of 3-126 Revision 0, 2020 3.1 FIGURES Figure No.

Title Figure 3.1-1 Typical Rod Cluster Control Assembly Figure 3.1-2 Rod Control Cluster Assembly Outline Figure 3.1-3 Fuel Assembly and Control Cluster Cross Section -

HIPAR, LOPAR, OFA, and VANTAGE+

Figure 3.1-4 HIPAR Fuel Assembly Figure 3.1-5 LOPAR Fuel Assembly Figure 3.1-6 OFA Fuel Assembly Figure 3.1-7 VANTAGE+ Fuel Assembly Figure 3.1-8 Guide Thimble to Bottom Nozzle Joint Figure 3.1-9 LOPAR Top Grid to Nozzle Attachment Figure 3.1-10 OFA And VANTAGE+ Top Grid to Nozzle Attachment Figure 3.1-11 Spring Clip Grid Assembly Figure 3.1-12 Mid-Grid Expansion Joint Design Plan View Figure 3.1-13 Elevation View - LOPAR Grid to Thimble Attachment Figure 3.1-14 Elevation View-VANTAGE+ Grid to Thimble Attachment Figure 3.1-15 VANTAGE+ Fuel Assembly with Performance+

Enhancements Figure 3.1-16 15x15 Upgraded Fuel Assembly Figure 3.1-17 Cycle 1 - Neutron Source Locations [Historical]

Figure 3.1-18 HIPAR Burnable Poison Rod Figure 3.1-19 LOPAR Burnable Poison Rod Figure 3.1-20 Comparison of Borosilicate Glass Absorber Rod with WABA Rod Figure 3.1-21 Wet Annular Burnable Absorber Rod

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-20 of 3-126 Revision 0, 2020 3.2 Chemical and Volume Control System Portions of the chemical and volume control system will continue to be utilized in the permanently defueled condition to process liquid radwaste.

3.2.1 System Design and Operation The following chemical and volume control system (CVCS) equipment will remain in service in the permanently defueled condition to support the processing of liquid radwaste:

21, 22, and 23 CVCS Hold Up Tanks (HUT) 21, 22, and 23 CVCS Hold Up Tank Transfer Pumps Hold Up Tank Recirculation Pump In addition, the following equipment is part of the Waste Disposal System.

Waste Hold Up Tank (WHUT)

Spent Resin Storage Tank Waste Hold Up Transfer Pump Table 3.2-1 defines the ASME code class for the CVCS. As liquid enters the holdup tanks, the nitrogen cover gas is displaced to the gas decay tanks in the waste disposal system through the waste vent header. A recirculation pump is provided to transfer liquid from one holdup tank to another.

Liquid effluent in the holdup tanks is processed by demineralization or as radwaste.

The resin fill tank is used to process resin from the demineralizers. The tank is made of austenitic stainless steel.

Basic material of construction is stainless steel for all valves.

All chemical and volume control system piping handling radioactive liquid is austenitic stainless steel. All piping joints and connections are welded, except where flanged connections are required to facilitate equipment removal for maintenance and hydrostatic testing.

3.2.2 Purpose Portions of the system are used to collect effluents and transfer them to the waste disposal system. Effluents are initially collected in the chemical and volume control system holdup tanks.

As fluid enters the holdup tanks, released gases (hydrogen and fission gases) mix with the nitrogen cover gas and are eventually drawn off to the waste gas system.

A holdup tank low pressure interlock will trip the CVCS holdup tank transfer pumps upon low pressure in the holdup tank. This interlock reduces the potential for creating a negative pressure condition in the holdup tanks during drain down of the tank.

Three holdup tanks contain radioactive liquid. The contents of one tank are normally being processed while another tank is being filled. The third tank is normally kept empty to provide additional storage capacity when needed. The total liquid storage sizing basis for the holdup

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-21 of 3-126 Revision 0, 2020 tanks is given in Table 3.2-2. The tanks are constructed of austenitic stainless steel. The three holdup tank transfer pumps are used to transfer water to waste collection tanks in unit 1. These centrifugal pumps are constructed of austenitic stainless steel.

TABLE 3.2-1 Chemical and Volume Control System Code Requirements Component Code Holdup tanks ASME III, Class C Piping and valves USAS B31.12 Notes:

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

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-22 of 3-126 Revision 0, 2020 TABLE 3.2-2 Chemical and Volume Control System Principal Component Design Data Summary Quantity Type

Capacity, gpm Head, ft or psi Design Pressure, psig Design Temperature, °F Pumps Holdup tank recirculation 1

Centrifugal 500 100-ft 75 200 Primary water makeup 2

Centrifugal 150 210-ft 150 Ambient Holdup Tank Transfer Pump 22 1

Centrifugal 25 63-ft 150 200 Holdup Tank Transfer Pump 21

& 23 2

Centrifugal 25 63-ft 150 200

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-23 of 3-126 Revision 0, 2020 3.3 Auxiliary Coolant System 3.3.1 Design Basis The auxiliary coolant system consists of two loops as shown in Drawings 227781, 9321-2720, and 251783 [Formerly Figure 3.3-1, Sheets 1, 2, and 3] the component cooling loop and the spent fuel pit cooling loop.

3.3.1.1 Performance Objectives 3.3.1.1.1 Component Cooling Loop The component cooling loop is designed to provide cooling to dissipate waste heat from various facility components. It also provides cooling for spent fuel pit components.

The loop design provides for detection of radioactivity entering the loop from the spent fuel pit and also provides means for isolation.

3.3.1.1.2 Spent Fuel Pit Cooling Loop The spent fuel pit cooling loop is designed to remove from the spent fuel pit the heat generated by stored spent fuel elements.

The loop design consists of two pumps, a heat exchanger, a filter, a demineralizer, piping, and associated valves and instrumentation. Alternate cooling capability can be made available under anticipated malfunctions or failures (expected fault conditions).

Loop piping is so arranged that the failure of any pipeline does not drain the spent fuel pit below the top of the stored fuel elements.

The thermal design basis for the loop provides for all fuel pool rack locations to be filled at the end of a full core discharge.

3.3.1.2 Design Characteristics 3.3.1.2.1 Component Cooling Loop Normally one pump and at least one component heat exchangers are operated to provide cooling water for the components located in the auxiliary building. At elevated CCW supply temperatures two pumps may be required. The water is normally supplied to the SFP cooling system even though one of the components may be out of service.

Makeup water is taken from the primary water treatment plant, as required, and delivered to the surge tank. A backup source of water is provided from the primary water makeup transfer pumps.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-24 of 3-126 Revision 0, 2020 The operation of the loop is monitored with the following instrumentation:

1. A pressure indicator on the line between the component cooling pumps and the component cooling heat exchangers.

2. A temperature indicator, flow indicator, and radiation monitor in the outlet line from the heat exchangers.

3. A temperature indicator on the main inlet line to the component cooling pumps.

3.3.1.2.2 Spent Fuel Pit Cooling Loop The spent fuel pit contains spent fuel discharged from the Unit 2 and Unit 3 reactors. Spent fuel cooling loop performance has been analyzed for operation at a core power level of 102% of 3216 MWt and at service water temperatures up to 95°F. When a refueling load of approximately 88 freshly discharged assemblies (plus previously discharged assembles) are present, the pump and spent fuel heat exchanger will handle the load and maintain a bulk pit water temperature less than 140°F. When a full core of 193 assemblies is freshly discharged, the bulk pit water temperature is maintained below 180°F.

Two criteria must be met before spent fuel can be discharged to the spent fuel pit:

1. Spent fuel cannot be discharged to the spent fuel pit until at least 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> after shutdown to satisfy the assumptions of the spent fuel handling accident analysis as discussed in Section 6.2.1. This requirement will be met prior to the implementation of the original version of the Defueled Technical Specifications and Defueled Safety Analysis Report. Thus, it will essentially be a historical requirement, because the facility will be permanently shut down and defueled.

2. An additional delay time limit prior to spent fuel discharge is administratively controlled by operating procedures to ensure that the total spent fuel heat load is within the capacity of the spent fuel cooling loop to satisfy the bulk pit water temperature limits discussed above. This is a variable time limit primarily dependent upon service water temperature, and cooling capacity without supplemental cooling.

3.3.1.3 Codes and Classifications All piping and components of the auxiliary coolant system are designed to the applicable codes and standards listed in Table 3.3-1. The component cooling loop water contains a corrosion inhibitor to protect the carbon steel piping. Austenitic stainless steel piping is used in the remaining piping systems that contain borated water without a corrosion inhibitor.

3.3.2 System Design and Operation 3.3.2.1 Component Cooling Loop Component cooling is provided for the following heat sources:

1. Waste gas compressors (waste disposal system).

2. Spent fuel pit heat exchanger (auxiliary coolant system).

Typically, one component cooling pump and at least one component cooling heat exchanger can accommodate the heat removal loads. Two CCW pumps are in stand-by and at least one heat

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-25 of 3-126 Revision 0, 2020 exchanger is utilized. At elevated CCW supply temperatures two CCW pumps may be required.

Three pumps and two heat exchangers can be used to remove the residual and sensible heat.

The surge tank accommodates expansion, contraction and inleakage of water, and ensures a continuous component cooling water supply until a leaking cooling line can be isolated. Makeup to the surge tank is provided from the primary water makeup system. The surge tank is normally vented to the atmosphere. In the unlikely event that the radiation level in the component cooling loop reaches a preset level above the normal background, the radiation monitor in the component cooling loop annunciates in the control room and closes a valve in the surge tank vent line.

Parameters for components in the component cooling loop are presented in Table 3.3-2.

3.3.2.2 Spent Fuel Pit Cooling Loop The spent fuel pit cooling loop removes residual heat from fuel placed in the pit for long term storage. The loop can safely accommodate the heat load from all of the assemblies for which there is storage space available.

The spent fuel pit cooling loop consists of two pumps, a heat exchanger, filter, demineralizer, piping and associated valves and instrumentation. One of the pumps draws water from the pit, circulates it through the heat exchanger and returns it to the pit. Component cooling water cools the heat exchanger. Redundancy of this equipment is not required because of the large heat capacity of the pit and the slow heatup rate.

The clarity and purity of the spent fuel pit water is maintained by passing approximately 5-percent of the loop flow through a filter and demineralizer. The spent fuel pit pump suction line, which is used to draw water from the pit, penetrates the spent fuel pit wall above the fuel assemblies. The penetration location prevents loss of water as a result of a possible suction line rupture.

Parameters for components in the spent fuel cooling loop are presented in Table 3.3-3.

3.3.2.3 Component Cooling Loop Components 3.3.2.3.1 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 3.3-2.

3.3.2.3.2 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 3.3-2.

3.3.2.3.3 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 3.3-2. In addition to

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-26 of 3-126 Revision 0, 2020 piping connections, the tank has a flanged opening at the top for the addition of the chemical corrosion inhibitor to the component cooling loop.

3.3.2.3.4 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.

3.3.2.3.5 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 3.3.2.3.6 Primary Water Storage Tank A single 165,000-gal primary water storage tank is provided to store the demineralized water used by the primary water makeup system shown in Drawing 9321-2724 [Formerly Figure 3.3-2]. The storage tank is constructed of type 304 stainless steel.

Chemical addition to the tank, if required, can be accomplished via a 3-in. blind flange connection located near the top of the tank, directly off the pressure-vacuum relief valve. A local sample point is provided on the bottom of the tank in addition to a tank drain and a loop seal overflow.

This loop seal will prevent the entrance of air. To ensure that this loop seal is filled with water a valved line is provided from the tank drain to the loop seal.

Besides these lines into the primary water storage tank, there is a feed from the primary water makeup pump recirculation. Lines carrying heating steam to and from the tank also enter it near its bottom. All of these connections and lines entering the tank are heat traced to prevent them from freezing. A large inspection port is provided on the side of the tank.

3.3.2.3.6.1 Primary Water Storage Tank Level Measurement Level in the tank is measured and indicated locally and in the central control room. In addition, high level and low level are alarmed in the central control room.

3.3.2.3.6.2 Primary Water Storage Tank Temperature Control Temperature in the tank is indicated locally. An additional temperature measurement is made at the tank, on the suction line to the makeup pumps.

The temperature element will sense a representative fluid temperature. This temperature measurement is used to control steam flow to the coils located at the bottom of the storage tank.

The steam coils will maintain the water in the storage tank at a sufficiently high temperature to prevent freezing of the tank contents. The walls of the tank are insulated and all lines connected to the tank and exposed to the environment are electrically heat traced to prevent freezing.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-27 of 3-126 Revision 0, 2020 In addition, the external instrument cabinet is heated and weatherproofed to help ensure a controlled temperature for the tank level instrumentation. Low temperature alarms alert the operator of any instrument heat trace failure or low temperatures in the instrument enclosure.

3.3.2.3.7 Primary Water Makeup Pumps Two primary water makeup pumps are provided and normally take their suction from the primary water storage tank. The pumps are constructed of type 316 austenitic stainless steel. Each can supply 150 gpm of water at a total dynamic head of 210-ft.

Control of both pumps is provided from the central control room. No local control of the pump is provided.

Normally one pump will be selected to run continuously; the second will be in auto. A limited flow recirculation line is provided and remains open in case makeup water is not required at a given time anywhere in the facility. An orifice in this line limits the recirculation flow.

Each pump is also provided with a discharge pressure gauge. Operation of the pumps without a suction head is prevented.

3.3.2.4 Spent Fuel Pit Loop Components 3.3.2.4.1 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.

3.3.2.4.2 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.

3.3.2.4.3 Spent Fuel Pit Filter The spent fuel pit filter removes particulate matter larger than 5 from the spent fuel pit water.

The filter cartridge is synthetic fiber and the vessel shell is austenitic stainless steel.

3.3.2.4.4 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.

3.3.2.4.5 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.

µ

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-28 of 3-126 Revision 0, 2020 3.3.2.4.6 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.

3.3.2.4.7 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.

3.3.3 System Evaluation System performance has been evaluated for service water temperatures up to 95°F for normal conditions and loss of offsite power.

3.3.3.1 Availability and Reliability 3.3.3.1.1 Component Cooling Loop The portion of the component cooling loop supplying containment components (reactor coolant pumps, excess letdown heat exchanger, and residual heat removal heat exchangers) is permanently isolated and made non-functional following plant shutdown and final removal of the fuel from the core to the spent fuel pit.

The portion of the component cooling loop outside containment, which includes the spent fuel pit heat exchanger, the component cooling water pumps and heat exchangers, and associated valves, piping, and instrumentation, is maintained functional following permanent plant shutdown and de-fueling for the purposes of providing a cooling method for the fuel in the spent fuel pit. The components of this loop section are capable of being repaired or replaced as necessary. 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. This loop section was originally designed to Seismic Class I criteria and maintains the robustness of the design although no credit is taken for this in the permanently defueled condition. The loop components continue to be housed in structures that were also originally designed to meet Seismic Class I requirements. The components are designed to the codes given in Table 3.3-1 and the design pressures given in Table 3.3-2. In addition, the components are not subjected to any high pressures or stresses. Hence, a rupture or failure of the system is very unlikely. Should cooling of the stored spent fuel by the component cooling loop be lost for some reason, other means of covering and cooling the fuel are available.

The Component Cooling Water Pumps are powered from appropriate electrical sources that are deemed to be reliable. These sources include primary and alternate off-site power supplies as well as available diesel generators.

3.3.3.1.2 Spent Fuel Pit Cooling Loop This manually controlled loop may be shut down safely for time periods, as shown in Section 3.3.3.2.2, for maintenance or replacement of malfunctioning components.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-29 of 3-126 Revision 0, 2020 3.3.3.2 Leakage Provisions 3.3.3.2.1 Component Cooling Loop With respect to water leakage from piping, valves, and equipment serving auxiliary coolant systems, 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 auxiliary coolant systems.

Tube or coil leaks in components being cooled would be detected during normal facility operations by the leak detection system described in Section 3.12. 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 site personnel would check that temperatures are stable.

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.

The atmospheric vent on the component cooling surge tank is automatically closed in the event of high radiation level in the component cooling loop. If 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 relief valves on the cooling water lines downstream from the spent fuel pit, 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. Historically, the over-pressurization incident resulted in a maximum component cooling water pressure of 185 psig from an event that is no longer possible. This pressure is allowed in the component cooling water system in accordance with its design code of B31.1, 1967 edition, par 102.2.4(2), addressing permissible variation and allowable stress value for a limited time.

3.3.3.2.2 Spent Fuel Pit Cooling Loop A leaking fuel assembly in the spent fuel pit can result in 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

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-30 of 3-126 Revision 0, 2020 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 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.

3.3.3.3 Incident Control 3.3.3.3.1 Component Cooling Loop In the unlikely event of a pipe severance in the component cooling loop, 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 primary water storage tank by one of the primary 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.

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 has 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.

3.3.3.3.2 Spent Fuel Pit Cooling Loop The most serious failure of this loop is complete loss-of-water in the SFP. To protect against this possibility, the SFP cooling connections enter near the water level so that the SFP cannot be either gravity drained or inadvertently drained. The water in the SFP 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 SFP on a regular basis.

3.3.3.4 Malfunction Analysis A failure analysis of pumps, heat exchangers and valves is presented in Table 3.3-4.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-31 of 3-126 Revision 0, 2020 TABLE 3.3-1 Auxiliary Coolant System Code Requirements Component Code Component cooling heat exchangers ASME VIII Component cooling surge tank ASME VIII Component cooling loop piping and valves USAS B31.1 Spent fuel pit filter ASME III, Class C Spent fuel heat exchanger side ASME VIII, shell side ASME III, Class C, tube Spent fuel pit loop piping and valves USAS B31.1

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-32 of 3-126 Revision 0, 2020 TABLE 3.3-2 (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 H2O 220 Motor horsepower, hp 250 Material (pump casing)

Cast iron or Carbon steel Design pressure, psig 150 Design temperature, °F 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, °F 100.1 Operating outlet temperature, °F 88.2 Design flow rate, lb/hr 2.66 x 106 Design temperature, °F 200 Design pressure, psig 150 Material Aluminum-bronze Tube side (service water)

Operating inlet temperature, °F 751 Operating outlet temperature, °F 81.9 Design flow rate, lb/hr 4.55 x 106 Design temperature, °F 200 Design pressure, psig 150 Material Copper-nickel (90-10)

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-33 of 3-126 Revision 0, 2020 TABLE 3.3-2 (Sheet 2 of 2)

Component Cooling Loop Component Data Component Cooling Surge Tank Quantity 1

Volume, gal 2000 Normal water volume, gal 1000 Design pressure, psig 100 Design temperature, °F 200 Construction material Carbon steel Relief valve setpoint, psig 52 Component Cooling Loop Piping and Valves Design pressure, psig 150 Design temperature, °F 200 Notes:

1.

Operation is acceptable up to 95°F.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-34 of 3-126 Revision 0, 2020 TABLE 3.3-3 (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, lb/hr 1.4 x 106 Design temperature, °F 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, lb/hr 1.1 x 106 Design temperature, °F 200 Design pressure, psig 150 Material Stainless steel Spent fuel pit skimmer pump Retired in place

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-35 of 3-126 Revision 0, 2020 TABLE 3.3-3 (Sheet 2 of 3)

Spent Fuel Cooling Loop Component Data Spent fuel pit cooling loop piping and valves Design pressure, psig 150 Design temperature, °F 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, °F 200 Spent fuel pit pump Quantity 2

Type Horizontal centrifugal Material Stainless steel Rated capacity, gpm 2,300 Rated head, ft H2O 125 Motor, hp 100 Design pressure, psig 150 Design temperature, °F 200 Spent fuel pit Volume ft3 37,300 Typical Boron concentration, ppm boron

>2,000 min Tech Spec Boron concentration, ppm boron

>2,000 min Spent fuel pit filter Quantity 1

Internal design pressure of housing, psig 200 Design temperature, °F 250 Rated flow, gpm 100 Maximum differential pressure across filter element at rated flow (clean cartridge), psi 5

Maximum differential pressure across filter element prior to removing, psi 20 Filtration requirement 98-percent retention of particles down to 5 µ

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-36 of 3-126 Revision 0, 2020 TABLE 3.3-3 (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 element at rated flow (clean), psi 1

Perforation, in.

0.2 Spent fuel pit demineralizer Quantity 1

Type Flushable Design pressure, psig 200 Design temperature, °F 250 Flow rate, gpm 100 Resin volume, ft3 30 Notes:

1. Original design.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-37 of 3-126 Revision 0, 2020 TABLE 3.3-4 Failure Analysis of Pumps, Heat Exchangers, and Valves Components Malfunction Comments and Consequences

1. Component cooling water pumps Rupture of a pump casing The casing and shell are designed for 150 psi and 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.

2. Component cooling water pumps Pump fails to start One operating pump supplies sufficient cooling water for SFP cooling.

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

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

5. Component cooling heat exchanger Tube or shell rupture Rupture is considered improbable because of low 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 makeup line check valve Sticks open The check valve is backed up by the manually-operated valve. Manual valve is normally closed.

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

8. Component cooling water outlet valve to residual heat exchanger Fails to open There is one valve on each outlet line from each heat exchanger. One heat exchanger remains in service and provides adequate heat removal to support safe storage of spent fuel in the spent fuel pit.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-38 of 3-126 Revision 0, 2020 3.3 FIGURES Figure No.

Title Figure 3.3-1 Sh. 1 Auxiliary Coolant System - Flow Diagram, Sheet 1, Replaced with Drawing 227781 Figure 3.3-1 Sh. 2 Auxiliary Coolant System - Flow Diagram, Sheet 2, Replaced with Drawing 9321-2720 Figure 3.3-1 Sh. 3 Auxiliary Coolant System - Flow Diagram, Sheet 3, Replaced with Drawing 251783 Figure 3.3-2 Primary Water Makeup System - Flow Diagram, Replaced with Drawing 9321-2724

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-39 of 3-126 Revision 0, 2020 3.4 Sampling System 3.4.1 Design Basis 3.4.1.1 Performance Requirements This system provides for analysis of liquid and gaseous samples obtained during normal conditions. Sampling of the following systems is discussed below:

1. Holdup tanks.

2. CVCS holdup tank transfer pumps discharge.

3. Chemical drain pump 21 discharge.

These samples are obtained at the high-radiation sampling system panels and evaluated by the online analysis systems or manual analysis.

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 site personnel exposure to any radiation during the sampling procedures.

3.4.1.2 Design Characteristics The design characteristics of the high-radiation sampling system include the following:

1. Control of background radiation and site personnel 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 routine sampling.

2. 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.

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. Leakage and drainage resulting from the sampling operations are collected and drained to tanks located in the waste disposal system.

3.4.1.3 Primary Sampling Low temperature-low pressure samples are obtained by the primary sampling system from the chemical and volume control and auxiliary coolant systems.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-40 of 3-126 Revision 0, 2020 3.4.1.4 Codes and Standards System code requirements are given in Table 3.4-1. In addition, the system meets the following provisions:

1. 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).

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

3. Provide means of safely obtaining diluted and undiluted samples for laboratory analysis.

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

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

6. Provide the capability to flush the sampled lines.

3.4.2 System Design and Operation 3.4.2.1 Primary Sampling System The primary sampling system consists of the high-radiation sampling system, which is shown in Drawing 9321-2745 (Figure 3.4-1). The high-radiation sampling system provides the representative samples for inline monitoring and laboratory analysis under normal conditions.

Analytical results provide guidance in the operation of the auxiliary coolant and chemical and volume control systems. Analyses show both chemical and radiochemical conditions. Typical information obtained includes fission product radioactivity level, hydrogen, oxygen, and fission gas content, corrosion product concentration, and chemical additive concentration.

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.

3.4.2.1.1 Components 3.4.2.1.1.1 Liquid Sampling Panel The liquid sampling panel valves and components are arranged in two modules installed in a common panel shield:

1. Module 2 - Demineralizer sampling module (DM).

2. 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.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-41 of 3-126 Revision 0, 2020 As a safety measure, the liquid sampling panel has a hooded splash box to contain any accidental liquid spill or gaseous release during reactor coolant or liquid grab sampling from the 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 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.

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 conditions are undiluted, depressurized liquid grab samples from the demineralizer, and radwaste modules.

An additional function of the liquid sampling panel during normal conditions is the purging of lines with sample to ensure representative samples will be obtained.

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

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

2. Diluted liquid samples from the radwaste modules of the liquid sampling panel.

3. Undiluted liquid samples from the radwaste modules of the liquid sampling panel for offsite analyses.

3.4.2.1.1.3 Boron Analyzer Backup boron analyses may be performed on undiluted grab samples from the demineralizer and radwaste modules of the liquid sampling panel for normal sampling for analysis in the onsite laboratory.

The primary sampling system provides that the routine sample analyses of undiluted samples are performed 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.

3.4.2.1.1.4 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.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-42 of 3-126 Revision 0, 2020 3.4.2.1.1.5 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.

3.4.2.1.1.6 Valves 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.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-43 of 3-126 Revision 0, 2020 TABLE 3.4-1 Sampling System Code Requirements Code Piping and valves USAS B31.11 Notes:

1.

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

3.4 FIGURES Figure No.

Title Figure 3.4-1 Sh. 1 Primary Sampling System - Flow Diagram, Sheet 1, Replaced with Drawing 9321-2745 Figure 3.4-1 Sh. 2 Primary Sampling System - Flow Diagram, Sheet 2, Replaced with Drawing 227178

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-44 of 3-126 Revision 0, 2020 3.5 Fuel Handling System The fuel handling system provides a safe, effective means of transporting and handling fuel until it leaves the facility after post-irradiation 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 of the spent fuel pit, which is kept full of water and is always accessible to personnel.

3.5.1 Design Basis 3.5.1.1 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)

The spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations. The spent fuel storage pit has accommodations as defined in Table 3.5-1. In addition, the spent fuel pit has the required spent fuel shipping area. The spent fuel storage pit is filled with borated water. The fuel is stored vertically in an array with sufficient center-to-center distance between assemblies to assure Keff <1.0 even if unborated water was used to fill the pit and 0.95 when filled with water borated 2000 ppm boron. Limits on enrichment and burnup of fuel in the spent fuel storage pit are given in the Technical Specifications.

Both IP2 and IP3 irradiated fuel assemblies may be handled and stored in the IP2 spent fuel storage pit. The above stated spent fuel storage requirements are being applied to both IP2 and IP3 irradiated fuel assemblies in the IP2 spent fuel storage pit. Detailed instructions are available for use by personnel handling irradiated fuel assemblies. 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 irradiated fuel handling 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).

3.5.1.2 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 spent fuel pit cooling 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.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-45 of 3-126 Revision 0, 2020 3.5.1.3 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 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 personnel. Pit water level is indicated, and water removed from the pit must be pumped out since there are no gravity drains. Shielding is provided for waste handling and storage facilities to permit operation within requirements of 10 CFR 20.

Gamma radiation is continuously monitored in the auxiliary building. A high level signal is alarmed locally and is annunciated in the control room.

3.5.1.4 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 spent fuel storage pit is a reinforced concrete structure with a seam-welded stainless steel plate liner. This structure is designed to withstand the anticipated earthquake loadings as a seismic Class I structure 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 III equipment.

3.5.2 System Design and Operation The spent fuel pit is kept full of water and is always accessible to personnel.

3.5.2.1 Major Structures Required for Fuel Handling 3.5.2.1.1 Spent Fuel Storage Pit The spent fuel storage pit is designed for the underwater storage of spent fuel assemblies, failed fuel cans if required, control rods and other non-fuel hardware inserts after their removal from both the Unit 2 reactor and the Unit 3 reactor.

The pit accommodations are listed in Table 3.5-1.

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.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-46 of 3-126 Revision 0, 2020 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-318(77), and SRP 3.8 of NUREG-0800. In addition to the mechanical loadings, the pool structure was also analyzed to include the temperature induced loadings. For this purpose, the thermal boundary conditions were conservatively specified as 180°F pool water temperature and 0°F 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 from both Units 2 and 3.

3.5.2.1.2 Storage Rack High density fuel storage racks have been designed to provide a maximum storage capacity of 1374 locations. The arrangement of the fuel storage racks in the SFP is shown in Figure 3.5-1.

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 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. Note: Boraflex is no longer credited for neutron absorption, but is still physically present in a degraded state.

Each Region I storage cell, as shown in Figure 3.5-2, 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 3.5-3. 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.

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.

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 SFP 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 SFP floor by the support feet.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-47 of 3-126 Revision 0, 2020 3.5.2.2 Major Equipment Required for Fuel Handling 3.5.2.2.1 FSB Fuel Handling Bridge Crane The PaR Systems, Inc. Crane is a wheel-mounted platform, which spans the East-West (E-W) direction of the Spent Fuel Pit (SFP) and travels in the North-South (N-S) direction. The PaR Crane is secured to the crane rails on the FSB El. 95-0, via seismic hold-down brackets and associated bolting. An Encoder Tracking Device is mounted on the FSB West Walkway, El. 96-6, which positions the crane in the N-S direction. The crane mounted computer will position the Crane Trolley-Tower Structure in the E-W direction. This equipment will position the Crane Motorized Hoist over a pre-assigned Spent Fuel Assembly (SFA) within the SFP. In addition, the PaR Crane controls interface with the existing FSB Up-Ender Control Console No. 21 (PK1). This computerized control feature provides assurance that the PaR Crane will not interfere with the FSB Up-Ender Assembly, which is located in the Fuel Transfer Canal.

The Motorized Hoist-Sheave Assembly is attached to a Trolley Structure, which is located on the wheel-mounted platform. The Motorized Hoist design incorporates a single lifting cable, which has a safety factor of 11.49:1. This safety factor exceeds the design criteria (10:1) for single lifting cables, as outlined in NUREG-0612. The Tower Structure is mounted on a Motorized Trolley, which travels in the E-W direction on the wheel-mounted work platform. The Motorized Hoist-Sheave Assembly, which has a 1-Ton rated capacity, will transfer SFAs within the SFP via long-handled tools suspended from the hoist hook. The hoist travel and tool length are designed to limit the maximum lift of a SFA and maintain a safe shield depth below the water surface of the SFP. A load weighing system will sense overload and underload conditions. This system will stop the upward movement of a SFA when it senses a load greater than a programmed set-point.

In addition, this system will stop the downward movement of a SFA when it senses a slack cable condition.

A 480V, 3-phase, 50 AMP power feed (normal supply) is provided from Distribution Panel No.

EP57 to the PaR Crane. In addition, a 480V, 3-phase, 100 AMP power feed (alternate supply) is provided from MCC27 to the PaR Crane. Transfer Switch No. EDA57 is provided so that the reliable power feeds can be provided by Distribution Panel No. EP57 or MCC27.

3.5.2.2.2 Shield Transfer Canister (STC) and HI-TRAC Transfer Cask The NRC has issued Amendment 268 for the inter-unit transfer of spent fuel from Unit 3 to Unit 2 (Reference 3.5-1). The Amendment is based on evaluations conducted for each aspect of the inter-unit transfer of fuel as documented in the Licensing Report (Reference 3.5-2). The non-proprietary version of the Licensing Report is incorporated by reference in the DSAR.

The STC is a thick-walled vessel with a removable top lid capable of transferring up to twelve spent fuel assemblies and associated non-fuel hardware. For inter-unit spent fuel transfer operations between the Unit 3 SFP and the Unit 2 SFP, the STC is used in conjunction with the HI-TRAC transfer cask. During STC closure activities and spent fuel transfer operations, the STC shielding is supplemented with the HI-TRAC shielding (steel, lead and water) and the water contained in the annulus space located between the STC and the HI-TRAC. For inter-unit spent fuel transfer operations, the HI-TRAC uses a solid lid and a centering assembly that keeps the STC centered inside the HI-TRAC cavity. The centering assembly forms an annular region inside the HI-TRAC which remains mostly full of water during loading and transfer operations. An air space is left in the HI-TRAC above the STC top flange to allow the STC lid operations to occur unhindered by water and provide an expansion volume for the water inside the HI-TRAC cavity.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-48 of 3-126 Revision 0, 2020 During spent fuel transfer operations the STC is mostly full of borated water and is steam blanketed to remove air from the STC. The STC includes a removable bolted lid with vent and drain ports for steam blanketing and water filling/draining purposes. The STC lid is coated on the top and sides to protect the carbon steel surfaces from corrosion. Should the coating system be damaged during wet fuel transfer operations, the damaged coating is removed and replaced with N-5000 or vacuum grease to prevent corrosion. The STC lid has lifting devices that can be remotely or manually actuated to engage trunnions on the STC body to lift the STC body when the STC lid bolting is removed. The STC lid also has threaded lid lifting points which provide a means to attach the STC and lid to overhead cranes.

THE STC is moved between Units 3 and 2 vertically in the HI-TRAC. Neither the HI-TRAC nor the STC are handled in the horizontal orientation when loaded with spent fuel assemblies and associated non-fuel hardware. In addition to the water in the STC cavity and the water in the annulus space between the STC and the HI-TRACs inner shell, the Hi-TRACs water jacket is also filled with water. These three discrete zones of water provide shielding and aid in heat transfer.

3.5.3 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.

Gamma radiation levels in the fuel storage area are continuously monitored. These monitors provide an audible alarm at the initiating detector indicating an unsafe condition.

An analysis evaluating the environmental consequences of a fuel handling incident is presented in Section 6.2.1.

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 SFP. With a boron concentration of 350 ppm the shutdown margin would be more than 5-percent. The technical specifications require that the boron concentration be maintained at 2000 ppm or more at all times.

3.5.4 Minimum Operating Conditions Minimum operating conditions are specified in the facility Technical Specifications.

3.5.5 Control of Heavy Loads 3.5.5.1 Introduction / Licensing Background A generic letter dated December 22, 1980, required responses to the guidelines of NUREG-0612 Control of Heavy Loads at Nuclear Power Plants. In response, the IP2 provisions for handling and control of heavy loads at Indian Point Unit 2 were addressed by letters June 22, 1981, September 30, 1982, January 31, 1983, and January 20, 1984. The NRC Safety Evaluation Report in letter dated February 19, 1985, concluded that the guidelines of NUREG-0612, Sections 5.1.1 and 5.3 have been satisfied and the Phase I of this issue for Indian Point Unit 2 is acceptable.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-49 of 3-126 Revision 0, 2020 The NRC Safety Evaluation Report in letter dated November 21, 2005 authorized the use of a single-failure-proof gantry crane for spent fuel cask handling operations up to 110 tons in weight.

Additional information was provided in letter dated July 12, 1996 in response to NRC Bulletin 96-

02.

NEI-08-05 R0, Industry Initiative on Control of Heavy Loads documented the industry initiative to address NRC staff concerns regarding the interpretation and implementation of regulatory guidance associated with heavy load lifts, was endorsed in Regulatory Issues Summary 2008-28 and has been addressed at Indian Point 2. This supersedes prior head drop analyses.

3.5.5.2 Safety Basis NUREG 0612 has two basic approaches available to demonstrate compliance: demonstrate adequate load handling reliability, or demonstrate that load drop consequences are within the limits of Criteria I-IV listed in Section 5.1 of the NUREG. Both approaches have been utilized in performing the evaluations described in the following sections.

In situations where a demonstration of handling system reliability was employed, the guidelines of NUREG-0612, Section 5.1.6, Single-Failure Proof Handling Systems, were utilized. The Ederer crane for cask handling was designed as a single failure proof crane.

In situations where a demonstration of limited load drop consequences was employed, a combination of system analyses and structural analyses was utilized. The specific approach chosen was based on the completeness of the available information, and a preliminary assessment of the likelihood of success of the possible approaches.

3.5.5.3 Scope of Heavy Load Handling Systems The following cranes and hoists were determined to be capable of handling heavy loads based on the criteria of NUREG-0612:

Fuel Handling crane (40/5-ton) 110t Ederer crane (110-ton)

The following discuss the results of our evaluations and submittals and are controlled using commitments A-873, A-887, A-1010, A-1015, A-1207, A-2467, A-2491, A-2492, A-2493, A-3174, A-3175, A-3176, A-3179, A-3180 and A-3465.

3.5.5.4 Response to NUREG 0612, Phase I Elements A defense-in-depth approach was used to ensure that all load handling systems are designed and operated so that their probability of failure is appropriately small. The basis for the approach was the Staff guidelines tabulated in Section 5 of NUREG-0612 and the program initiated to ensure that these guidelines are implemented. These guidelines consist of the following criteria from Section 5.1.1 of NUREG-0612:

Guideline 1 - Safe Load Paths Guideline 2 - Load Handling Procedures Guideline 3 - Crane Operator Training

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-50 of 3-126 Revision 0, 2020 Guideline 4 - Special Lifting Devices Guideline 5 - Lifting Devices (Not Specially Designed)

Guideline 6 - Cranes (Inspection, Testing, and Maintenance)

Guideline 7 - Crane Design Satisfaction of these guidelines for the Fuel Handling Crane and the Ederer Crane is shown in Table 3.5-2.

Guideline 1 - Safe Load Paths To ensure that crane operators remain knowledgeable of load handling precautions, annual refresher training is conducted to identify exclusion areas and to review load handling procedures.

In addition to the above procedures, additional structural and systems analyses were performed to determine the consequences of a load drop indicate that suitable system redundancy and structural integrity exist so that the consequences of a load drop would not exceed the criteria of NUREG-0612, Section 5.1.

Fuel Storage Building Ederer Crane The Ederer 110-ton design rated gantry crane is used to move spent fuel casks up to 110 tons into and out of the spent fuel pit by lifting a fully loaded Holtec HI-TRAC 100 spent fuel transfer cask and its associated components. The HI-STORM cask system utilizes the HI-TRAC 100 transfer cask for transporting a multi-purpose canister (MPC) from the spent fuel pit, and for inter-cask MPC transfers required for on-site storage. However, this crane is restricted from handling casks over spent fuel in the spent fuel pit and will only be utilized for other loading activities in the FSB.

Safe load paths have been determined, analyzed and documented in procedures for control of heavy loads handled by the Ederer gantry crane. Deviations from the safe load paths will require written alternative procedures reviewed and approved in accordance with IP2 procedures.

The Ederer gantry crane (by design) is unable to move spent fuel casks over any area of the spent fuel pit where the spent fuel is stored.

Fuel Handling Crane The Fuel handling Crane may be used to transport equipment, such as inspection rigs or electronics, to the Spent Fuel Pit area. For equipment handling, the crane is utilized to transport loads of no greater than 2000 lbs over the pool area. For fuel handling, the crane may carry a load no heavier than the weight of a fuel assembly containing a control rod assembly, plus the tool and small load block.

No object weighing more than 2,000 pounds may be moved over any region of the spent fuel pit when the pit contains spent fuel, unless a technical analysis has been performed consistent with the requirements of NUREG-0612 establishing the necessary controls to assure that a load drop accident could damage no more than a single fuel assembly. Administrative and procedural controls to protect fuel and fuel racks may include path selection to prevent loads from passing over or near fuel. For cases in which very heavy loads (>30,000 pounds) are transported over the spent fuel pit, the loads cannot under any circumstances pass over irradiated fuel. In all cases where loads >2,000 pounds are carried over the pit, the ventilation system must be functional.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-51 of 3-126 Revision 0, 2020 All standard modes of failure have been considered in the design of the Fuel Handling Crane.

These modes of failure were provided for by utilization of a minimum safety factor of 5 based on the ultimate strength of the material used in the design of cables, shafts and keys, gear teeth and brakes.

All crane equipment was sized to handle the single heaviest load realized during facility operation.

All lifts are made by qualified personnel. The equipment is properly maintained and periodically inspected by qualified personnel. An analysis of impact loading of the spent fuel cask into the spent fuel storage pool is provided in Section 3.5.3.

Mechanical stops incorporated on the bridge rails of the Fuel Handling Crane make it impossible for the bridge of the crane to travel further north than a point directly over the spot in the spent fuel pit that is reserved for the spent fuel cask. Therefore, it will be impossible to carry any object over the spent fuel storage areas north of the spot in the pit that is reserved for the cask with either the 40 or 5-ton hook of the Fuel Handling Crane. However, to further minimize the potential for a heavy load impacting irradiated fuel in the spent fuel pit, load paths will be defined in procedures and shown on equipment layout drawings.

The mechanical stops may be removed under administrative controls and the crane moved over spent fuel storage areas, provided that the fuel storage building ventilation system is functional, the spent fuel pit boron concentration is at least 2000 ppm and there is no heavy load carried.

This allows operations over the spent fuel pit with the 5-ton hoist. The 40-ton hoist may not carry any load over the SFP since the load block is a one-ton load and has not been fully evaluated for heavy loads.

The existing 40-ton non-single-failure-proof Fuel Handling Crane does not have the capacity to handle the HI-TRAC 100 transfer cask and its associated components. Performance of the crane satisfies the objectives of NUREG-0612 and the intent of NUREG-0554 with regard to maintaining the potential for a load drop extremely small.

Guideline 2 - Load Handling Procedures A series of operating procedures have been developed for operation of load handling equipment at Indian Point Unit 2.

Load handling procedures provide for the movement of all heavy loads in the vicinity of irradiated fuel or systems and equipment required for decay heat removal, and that load designation was based on the generic load identified in Table 3-1 of NUREG-0612. Further, these procedures contain the precautionary information required by NUREG-0612, Guideline 2. These procedures comply with the commitments made for safe load handling.

The 110T Ederer gantry crane operating procedures utilized for cask and cask component lifts include: identification of required equipment; inspection and acceptance criteria required before load movement; the steps and proper sequence to be followed in handling the load; defining the safe load path; and other precautions. A specific cask loading and handling procedure will provide additional details for controlled movement during cask handling operations.

Guideline 3 - Crane Operator Training

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-52 of 3-126 Revision 0, 2020 A qualification program for the qualification and training of crane operators at Indian Point Unit 2 have been developed to meet the provisions of ANSI B30.2-1976, with no exceptions taken.

Crane operator training and qualification is addressed in the qualification program and include precautions and instructions to assure proper operator conduct.

This qualification program meets the requirements of Chapter 2-3 of ANSI B 30.2-1967, Operation - Overhead and Gantry Cranes, as developed by the American National Safety Code for Cranes, Derricks, Hoists, Jacks and Slings.

Guideline 4 - Special Lifting Devices The HI-TRAC lifting yoke used with the Ederer crane is the only special lifting device that is required to meet the guidelines of ANSI N14.6-1993 and the additional guidelines of NUREG-0612, Section 5.1.6(1)(a).

Guideline 5 - Lifting Devices (Not Specially Designed)

Facility procedures require that sling selection and use for all loads requiring sling lifting devices be in accordance with ANSI B30.9.

Other lift components utilized with the Ederer Crane and HI-STORM 100 cask system meet ANSI B30.9-1971 requirements, including the additional guidelines of NUREG-0612, Section 5.1.6(1)(b).

Guideline 6 - Cranes (Inspection, Testing, and Maintenance)

The 110T Ederer gantry crane is inspected, tested and maintained in accordance with Chapter 2-2 of ANSI B30.2-1976 and the additional guidance contained in NUREG-0612, Section 5.1.1(6) regarding frequency of inspections and test.

Guideline 7 - Crane Design A design analysis of each handling system using the design criteria of the applicable standards has been performed.

The 40-ton Fuel Handling Crane was built prior to the issuance of ANSI B30.2-1976 and CMAA-70. However, a detailed point-by-point comparison has been performed, comparing information from the manufacturer with the criteria of these standards. Analysis was performed for only those components that are load bearing or are necessary to prevent conditions which could lead to a load drop. This review indicates that the crane complies with all requirements with the exception of Specification 3.2 of CMAA-70 and Section 2.1.4.1 of ANSI B30.2-1976. These specifications require that welding be performed in accordance with AWS D1.1, Structural Welding Code, and AWS D14.1, Specifications for Welding Industrial and Mill Cranes. The welding procedures used are equivalent to current welding criteria based on the following:

a) welding was performed in accordance with the then-current code AWS Dl,l, Structural Welding Code b) practices and procedures used for welding are equivalent to those in AWS D14.l, which was not issued at the time

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-53 of 3-126 Revision 0, 2020 c) welders were qualified to existing AWS criteria d) all welds were visually inspected Section 3.5.6, Fuel Storage Building (FSB) Dry Cask Storage (DCS) Operations, provides more detail on the FSB 110-Ton Ederer Single Failure Proof Gantry Crane and Section 3.5.2.2.1 provides more detail on the FSB Fuel Handling Bridge crane.

Additional specific information concerning design compliance with the more restrictive requirements of CMAA-70 is contained in the safety evaluation report.

The 110-ton Ederer gantry crane is installed on a crane rail system. The crane rail system for the Ederer crane consists of crane rail, rail pad, rail clip, sole plate assembly, and sole plate anchor embedments. The sole plate assembly consists of 2" thick steel plate which is held to the concrete slab with 1" diameter rod anchor embedments. The crane rail is attached to the sole plate assembly by rail clips with a rail pad between the crane rail and the sole plate assembly. The crane rail and the concrete slab of the reconstructed truck bay are designed and built to withstand seismic loads, as well as the static loads.

The Ederer gantry crane was designed with a telescopic tower and automated folding cantilever arms to avoid interference with either the existing overhead crane or the refueling bridge crane.

During dry cask loading operations the gantry crane will be in its raised position and the existing overhead crane will remain in the south position and de-energized to prevent accidental movement. Once the cask loading operation is completed, the gantry crane will be stored in its far west position, with the tower lowered and the arms folded. This will allow unobstructed use of both the existing overhead and refueling bridge cranes.

The cantilevered girder for the main hoist trolley will extend over the spent fuel pit cask laydown area. The girders are equipped with a retraction mechanism, accomplished via lead screw actuators that allow them to be folded back in order to permit unobstructed use of the existing overhead and refueling bridge cranes. Because of the cantilevered design, the gantry crane requires provisions to ensure stability against overturning. This is accomplished via a floor anchorage system with fixed-in-place hold down features that oppose crane uplift forces. To provide a foundation system capable of resisting these uplift forces, the design includes a steel ballast box filled with steel plates that will act as a counterbalance. The ballast box foundation consists of a 2-foot thick reinforced concrete slab founded on bedrock, and its primary function is to transmit all bearing loads from the weight of the ballast box directly to the underlying bedrock.

The Ederer gantry crane movements are governed by a series of limit and proximity switches that are controlled by a programmable logic controller (PLC) which ensures that: (1) movement of the trolley towards the spent fuel pit is only permitted if turnbuckles are attached to the crane tie down points, cantilever arms are extended and locked in place, and the main transfer hoist is operating at an elevation that allows the HI-TRAC to clear the south wall of the spent fuel pit; (2) limit switches on the trolley rails limit excessive movement of the trolley to the north and prohibit lowering of the load until a minimum northward travel is reached; and (3) main transfer hoist operation is prohibited until the Ederer gantry crane tower is in its raised position and pinned in place.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-54 of 3-126 Revision 0, 2020 3.5.5.5 Single Failure Proof Cranes for Spent Fuel Casks Sections 3.5.5.4, Response to NUREG 0612, Phase I Elements, and 3.5.6, Fuel Storage Building (FSB) Dry Cask Storage (DCS) Operations, provide more detail on the FSB 110-Ton Ederer Single Failure Proof Gantry Crane.

3.5.5.6 Safety Evaluation The controls implemented to address NUREG-0612 Phase 1 elements make the risk of a load drop very unlikely. The use of increased safety factors for load path elements makes the risk of a load drop extremely unlikely and acceptably low. In the event of a postulated load drop, the consequences are acceptable, as demonstrated by system analyses or the load drop analysis.

Restrictions on load height, load weight, and medium under the load are reflected in facility procedures. The risk associated with the movement of heavy loads is evaluated and controlled by station procedures.

The design and use of the Ederer single-failure-proof gantry crane is in accordance with NUREG-0554 and satisfies the guidelines of NUREG-0612. The crane enables the use of the HI-TRAC transfer cask and associated components with very low risk to irradiated fuel stored in the spent fuel pit. The use of the Ederer single-failure-proof gantry crane for cask handling operations for loads up to 110 tons is approved.

3.5.6 Fuel Storage Building (FSB) Dry Cask Storage (DCS) Operations The 100-Ton Dry Cask Storage System (HI-TRAC, Multi-Purpose Canister (MPC) and HI-STORM Overpack), FSB 110-Ton Single Failure Proof Gantry Crane, FSB Low Profile Transporter (LPT)

System, and Vertical Cask Transporter (VCT) facilitate removal of Spent Fuel Assemblies (SFAs) from the Spent Fuel Pool (SFP). During FSB Dry Cask Storage Operations, SFAs are transferred from the SFP with the HI-TRAC / MPC, inserted into the HI-STORM Overpack. The LPT System transfers the HI-STORM Overpack from the FSB to the east side of the PAB / MOB Crossover Walkway. The Vertical Cask Transporter transports the HI-STORM Overpack to the IPEC Independent Spent Fuel Storage Installation (ISFSI) Facility.

3.5.6.1 FSB 110-Ton Ederer Single Failure Proof Gantry Crane The 110-Ton Ederer Crane was designed to withstand normal operating loads, rated loads, seismic loads and extraordinary loads. The design of the 110-Ton Ederer Crane satisfies the design requirements and safety factors of CMAA-70, NUREG-0554 and Regulatory Guide 1.29.

The 110-Ton Ederer Crane conforms to the single failure proof requirements addressed in NRC NUREG-0554 and will retain and control a suspended critical load during and following a Safe Shutdown Earthquake (SSE).

The 110-Ton Ederer Crane is normally located on the west side of the FSB Truck Bay Floor, EL 77-6 in its lowered position and rests on rails that are installed within the FSB Truck Bay Floor.

In order to lift and transfer loaded and unloaded spent fuel casks into and out of the SFP, the 110-Ton Ederer Crane is raised to its upper position. Once in the upper position, the East and West Crane Girder Assemblies, and the North Tie-End Crane Girder Assembly are fully-extended (cantilevered position) out over the SFP. The 110-Ton Ederer Crane is then in position for visually controlled lifting and transferring of loaded and unloaded spent fuel casks into and out of the SFP Cask Pit Area.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-55 of 3-126 Revision 0, 2020 The 110-Ton Ederer Crane is connected, via tie-down turnbuckles, to the 30-Ton Ballast Box, which is embedded within the FSB Truck Bay Floor. The Ballast Box is filled with ~200-Tons of counter weight steel plates. The Ballast Box and tie-down turnbuckles provide stability and restraints for the 110-Ton Ederer Crane during a seismic event. The 110-Ton Ederer Crane is provided with counter weight to reduce the uplift (tension) loads in the tie-down turnbuckles, when the Crane Trolley travels onto the East and West Crane Girder Assemblies towards the SFP Cask Pit Area.

The 110-Ton Ederer Crane is provided with numerous limit switches, which control movement of the 110-Ton Ederer Crane. The limit switches are mounted on the Trolley, Girder Assemblies, Jacking Screw System, and Gantry Crane Structure. The limit switches control the speed of the equipment, limit travel distances and provide interlock signals to defeat some crane functions, when the 110-Ton Ederer Crane is not in the correct configuration. In addition, a Seismic Accelerometer automatically de-energizes the power feed to the 110-Ton Ederer Crane during a seismic event.

3.5.6.2 FSB Low Profile Transporter (LPT) System The FSB LPT System was designed to withstand normal operating loads, maximum vertical -

lateral track loads, maximum static stack-up loads, maximum transit loads, and Safe Shutdown Earthquake (SSE) loads. The design of the FSB LPT System satisfies the design requirements and safety factors of AISC, IPEC and Hilman-Rollers.

The FSB Gantry Crane transports the fully-loaded HI-TRAC / MPC towards the east for stake-up onto the Hi-STORM Overpack. The LPT Assembly, Chain Drive Assembly and Chain Drive Control Panel control internal FSB movements in the East-West direction.

The FSB LPT System transports the fully-loaded HI-STORM Overpack towards the east and south, so that, the HI-STORM Overpack can exit the FSB. The LPT Transporter Assembly, Transfer Table, Hydraulic Cylinders, and Transfer Table Hydraulic Cart control internal FSB movements in the North-South direction.

The FSB LPT System transports the fully-loaded HI-STORM Overpack through the FSB Truck Bay Roll-Up Door Opening, into the FSB Alleyway Trench and to the east side of the IP2 PAB /

MOB Crossover Walkway. The HI-STORM Overpack is in position to be lifted and transported by the Vertical Cask Transporter to the IPEC ISFSI Facility. The LPT Assembly, Track Assemblies, Guide Bars and Aircraft Tugger control external FSB movements in the East-West direction. Empty HI-STORM Overpacks will be transported into the FSB in the reverse order from above.

3.5.7 Inter-Unit Spent Fuel Transfer Operations The NRC has issued Amendment 268 for the inter-unit transfer of spent fuel from Unit 3 to Unit 2 (Reference 3.5-1). The Amendment is based on evaluations conducted for each aspect of the inter-unit transfer of fuel as documented in the Licensing Report (Reference 3.5-2). The non-proprietary version of the Licensing Report is incorporated by reference into the DSAR.

In preparation for inter-unit spent fuel transfer operations between the Spent Fuel Pool (SFP) in the IP3 Fuel Storage Building (FSB) and the SFP in the IP2 FSB, the HI-TRAC top lid is removed and the empty shielded transfer canister (STC) is placed inside the HI-TRAC transfer cask. The HI-TRAC / STC Centering Assembly centers the STC inside of the HI-TRAC. The HI-TRACs

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-56 of 3-126 Revision 0, 2020 solid top lid is installed to prevent any spilling of the water during the transfer process. Movement of the HI-TRAC (containing the STC) is performed using the Vertical Cask Transporter (VCT),

and using the IP2 Low Profile Transporter (LPT), or using Air Pads at IP3.

THE VCT moves the HI-TRAC containing the empty STC outside the IP3 FSB truck bay door.

The HI-TRAC is lowered onto Air Pads and the VCT releases the HI-TRAC. The IP3 FSB truck bay door is opened and the HI-TRAC is positioned inside the IP3 FSB truck bay beneath the FSB cask handling crane using the IP3 Air Pads. The HI-TRAC top lid is removed and the annulus between the STC and HI-TRAC is filled with demineralized water to the required level. The STC lid nuts and washers are removed and the STC is filled with SFP water.

The FSB cask handling crane is positioned over the STC and the STC Lift Lock is fastened to the STC lid and attached to the FSB cask handling crane. The STC is removed from the HI-TRAC and positioned over the cask loading area of the SFP. A set of remotely (or manually) actuated STC Lifting Devices attach the STC lid to the STC lifting trunnions. The STC is lowered into the cask loading area and the lid is removed.

For each fuel transfer cycle, up to twelve IP3 spent fuel assemblies including associated non-fuel hardware are loaded into the STC. The STC lid is positioned over the STC and installed. The STC Lifting Devices attach the lid to the STC lifting trunnions. After the Lifting Device arms are properly engaged to the lifting trunnions, the STC is raised to the surface of the SFP and any standing water on the lid is removed. A small amount of water is removed from the STC to avoid spilling during handling. Under the direction of Radiation Protection personnel radiological controls are established and surveys taken as the STC is raised and removed from the SFP, sprayed with demineralized water and placed directly into the HI-TRAC in the IP3 truck bay. The STC lid, nuts and washers are installed with the nuts left loose. The STC Lift Lock is disconnected from the STC top lid and removed. Free flow verification through the STC lid vent and drain lines is performed. The STC lid nuts are torqued and the STC seals are tested in accordance with ANSI N14.5 to assure that the STC is properly assembled for transfer operations. The required STC water level is established by blowing steam into, and water out of, the STC cavity thereby creating a compressible water vapor space. The STC top lid radiation level is measured to verify compliance with Technical Specification requirements. As required by the Technical Specifications the pressure inside the STC is monitored for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to demonstrate that there is not a significant amount of air in the STC and that a fuel misload has not occurred.

Following completion of the pressure test the STC lid vent and drain port cover plates are installed and the seals are testing in accordance with ANSI N14.5. The HI-TRAC top lid is installed and the bolts are tightened and the seal is tested in accordance with ANSI N14.5. The HI-TRAC side radiation levels are measured to verify compliance with Technical Specification requirements.

The IP3 FSB truck bay door is opened and the loaded HI-TRAC is moved outside the IP3 FSB to the VCT on Air Pads using the Prime Mover.

The VCT travels inside the Protected Area on the approved haul route between IP3 and IP2. Prior to each transfer of spent fuel assemblies, the haul route is visually inspected and repaired as necessary.

The HI-TRAC containing the loaded STC is lowered from the VCT onto the IP2 LPT and moved into the IP2 FSB. Inside the IP2 FSB, the HI-TRAC is positioned beneath the 110-Ton Ederer Crane. A drain line containing a pressure gauge is connected to the HI-TRAC top lid vent port and opened relieving any internal pressure. The HI-TRAC top lid bolts are removed and the HI-TRAC top lid is removed. The drain line is then attached to the vent port connection located on

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-57 of 3-126 Revision 0, 2020 the lid of the STC and opened relieving any internal STC pressure. STC lid nuts and washers are removed.

The Lift Cleats (with the Lift Cleat Adapter) are attached to the STC lid (the STC Lifting Devices already are installed on the STC lid). The 110-Ton Ederer Crane is attached to the STC through the Lift Cleat Adapter. The STC lifting device arms are engaged with the STC trunnions. Under the direction of Radiation Protection personnel the STC is raised out of the HI-TRAC and positioned directly over the SFP cask loading area and lowered into the pool. IP2 Technical Specification 3.7.12 requires that boron levels in the IP2 SFP have a concentration of greater than 2000 ppm which is also required for the STC spent fuel unloading activities.

With the STC in the SFP cask loading area, the STC Lifting Devices are released from the STC lifting trunnions and the STC lid is removed. The spent fuel assemblies and associated non-fuel hardware are removed from the STC and placed in the SFP racks in accordance with the requirements of IP2 Technical Specification 3.7.13. The STC lid is positioned over the STC and installed. The lids STC Lifting Devices are attached to the STC lifting trunnions and the STC is raised to the surface of the SFP. Any standing water on the lid is removed. Under the direction of Radiation Protection personnel the STC is raised and removed from the SFP, sprayed with demineralized water, and the water inside the STC is lowered before the STC is placed into the HI-TRAC. The STC lid studs and nuts are installed and the lid studs and nuts are tightened. The Lift Cleats are disconnected from the STC top lid and the Lift Cleats are Lift Cleat Adapter are removed. The HI-TRAC top lid is installed, the bolts are tightened, and the HI-TRAC containing the empty STC is then ready to be returned to the IP3 FSB.

REFERENCES FOR SECTION 3.5

1.

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.

268 to Facility Operating License No. DPR-26, July 13, 2012.

2.

Holtec Report HI-2094289, Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at Indian Point Energy Center, Revision 10.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-58 of 3-126 Revision 0, 2020 TABLE 3.5-1 Fuel Handling System Data SPENT FUEL STORAGE PIT Equivalent fuel assemblies1 1374 Number of space accommodations for failed fuel cans 2

Number of space accommodations for spent fuel shipping cask 1

Center-to-center spacing of Region cells, in 10.545(N-S) 10.765(E-W)

Center-to-center spacing of Region cells, in 9.04 Maximum Keff with borated water (Region) 0.95 Maximum Keff with unborated water (Region)

<1.0 MISCELLANEOUS DETAILS Wall thickness for spent fuel storage pit, ft 3 to 6 Weight of fuel assembly with rod cluster control (dry), lb 1,580 Notes:

1.

After re-racking.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-59 of 3-126 Revision 0, 2020 TABLE 3.5-2 NUREG-0612 Compliance Matrix Heavy Loads Weight or Capacity (tons)

Guideline 1 Safe Load Paths Guideline 2 Procedures Guideline 3 Crane Operator Training Guideline 4 Special Lifting Devices Guideline 5 Slings Guideline 6 Crane - Test and Inspection Guideline 7 Crane Design Interim Measure 1 Technical Specifications Interim Measure 6 Special Attention

1.

Fuel Handling Crane 40 R

R C

++

C C

R R

++

2. 110t Ederer Crane 110 C

C C

C C

C C

++

++

C - Action complies with NUREG-0612 Guideline.

R - Revisions/modifications designed to comply with NUREG-0612 Guideline.

++ - Not applicable.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-60 of 3-126 Revision 0, 2020 3.5 FIGURES Figure No.

Title Figure 3.5-1 Spent Fuel Storage Rack Layout Figure 3.5-2 Spent Fuel Storage Cell Region 1 Figure 3.5-3 Region I Cell Cross-Section Figure 3.5-4 Region II Cross-Section

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-61 of 3-126 Revision 0, 2020 3.6 Facility Service Systems 3.6.1 Service Water System 3.6.1.1 Design Basis The service water system is designed to supply cooling water from the Hudson River to various heat loads to support the storage and handling of spent fuel. The system also provides water required for cleaning the traveling screens.

3.6.1.2 System Design and Operation The service water system flow diagram is shown in Drawings 9321-2722 and 209762 [Formerly Figure 3.6-1, 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. 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 loads. The loads are those, which are supplied with cooling water from the designated service water header by manually starting a service water pump when required. The minimum flow requirements for the service water system are met by one or more pumps supplying at least 5000 gpm. This ensures that the following loads will be provided with sufficient cooling:

Spent fuel cooling via the CCW heat exchangers TWS wash water and CWP bearing cooling 22 Standby Diesel Generator (referred to as 22 Emergency Diesel Generator in site documents)

Condenser waterbox degassing pumps Appendix R/SBO Diesel Generator Zurn strainer blowdown 13 FWCHX for CENTAC cooling 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 will be left 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.

The loads are normally supplied by one or more pumps provided.

The standby diesel-driven generator unit is supplied with cooling water from the service water header. One of the two parallel modulating control valves in the common discharge line from the diesel coolers is closed and will be manually opened to run the standby diesel generator. The inlet valving is arranged so that the diesel can be served by either of the supply headers.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-62 of 3-126 Revision 0, 2020 The cooling water supply to the Appendix R Diesel Generator Heat Exchanger may be manually realigned from city water to service water.

3.6.1.3 Design Evaluation Sufficient pump capacity is included to provide design service water flow to support the storage of spent fuel in the SFP.

3.6.1.4 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.

3.6.2 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)

License Condition 2.K of Facility License DPR-26 for IP2 regarding the Fire Protection Program was eliminated in License Amendment No. 294 to reflect the permanently defueled condition of the facility. After the certifications required by 10 CFR 50.82(a)(1) were docketed for IP2, the 10 CFR Part 50 license no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the fire protection program was revised to take into account the decommissioning facility conditions and activities. IP2 continues to utilize the defense-in-depth concept, placing special emphasis on detection and suppression in order to minimize radiological releases to the environment.

License Condition 2.K, which was based on maintaining an operational fire protection program in accordance with 10 CFR 50.48, with the ability to achieve and maintain safe shut down of the reactor in the event of a fire, is no longer be applicable at IP2. In addition, Appendix R to 10 CFR 50 is no longer be applicable to IP2. However, many of the elements that are applicable for the operating plant fire protection program continue to be applicable during facility decommissioning.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-63 of 3-126 Revision 0, 2020 During the decommissioning process, a fire protection program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard.

The Indian Point 2 Fire Protection Program description is provided separately in the following documents:

IPEC Fire Protection Program Plan IP2 Fire Hazards Analysis Report These documents provide a complete description of the Indian Point 2 Fire Protection Program including a description of fire areas, fire suppression and detection as well as other fire protection features credited to limit the effects of fires.

3.6.3 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 makeup water to various systems.

3. To provide cooling water to various components.

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

5. To provide cooling water to the SBO / Appendix R Diesel Generator Heat Exchanger.

City water for the Indian Point Unit 2 comes from the city water main on Broadway via the Unit 1 mains and storage tanks and is under cathodic protection where the piping crosses the Algonquin Gas pipes. Unit 2 is tied to this system primarily through piping connections at two locations on the low-pressure header (see Drawings 192505, 192506, and 193183 [Formerly Figure 3.6-2]).

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

2. Cooling to the Appendix R Diesel Generator Heat Exchangers.

The second connection is at the north side of the header. This connection provides water for general usage via hose connections inside the primary auxiliary building and waste holdup tank pit.

A backup water supply is also provided for the circulating water pump seals and bearings.

3.6.4 Compressed Air Systems 3.6.4.1 Instrument Air System The instrument air system is designed such that the instrument air shall be available under all conditions. The system is shown in Drawing 9321-2036 [Formerly Figure 3.6-3] and is provided by the 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.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-64 of 3-126 Revision 0, 2020 3.6.4.2 Station Air System The station air system shown in Drawing 9321-2035 [Formerly Figure 3.6-4] is 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.

3.6.5 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 primary auxiliary building, and the primary water storage tank.

Provision is made for the following heating services:

1. Primary auxiliary building.

a. Electric strip heaters.

b. Steam unit heaters.

2. Purge system containment building.

a. Air makeup steam tempering units.

3. Fuel storage building.

a. Steam unit heaters for standby heating.

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

4. Fan room.

a. One steam unit heater.

REFERENCES FOR SECTION 3.6

1.

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

Subject:

Emergency Amendment to Increase the Service Water Temperature Limit to 90oF (TAC 73764), dated August 7, 1989.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-65 of 3-126 Revision 0, 2020 3.6 FIGURES Figure No.

Title Figure 3.6-1 Sh. 1 Service Water System - Flow Diagram, Sheet 1, Replaced with Drawing 9321-2722 Figure 3.6-1 Sh. 2 Service Water System - Flow Diagram, Sheet 2, Replaced with Drawing 209762 Figure 3.6-2 Sh. 1 City Water System - Flow Diagram, Sheet 1, Replaced with Drawing 192505 Figure 3.6-2 Sh. 2 City Water System - Flow Diagram, Sheet 2, Replaced with Drawing 192506 Figure 3.6-2 Sh. 3 City Water System - Flow Diagram, Sheet 3, Replaced with Drawing 193183 Figure 3.6-3 Instrument Air - Flow Diagram, Replaced with Drawing 9321-2036 Figure 3.6-4 Station Air - Flow Diagram, Replaced with Drawing 9321-2035

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-66 of 3-126 Revision 0, 2020 3.7 Equipment And System Decontamination 3.7.1 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-or n-p reactions on the water or impurities in the water, and activated corrosion products.

Fission products in the reactor coolant 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 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 coolant system or SFP components can result in contamination of the equipment, tools and clothing of the personnel involved in the maintenance. Also, leakage or spillage from these systems can contaminate the immediate areas and contribute to the contamination of the equipment, tools, and clothing.

3.7.2 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 facility, 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.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-67 of 3-126 Revision 0, 2020 3.7.3 Decontamination Facilities Decontamination facilities onsite consist of an equipment pit and a cask pit located adjacent to the spent fuel storage pit inside Unit 1 on the 70 Fuel Handling floor. In the stainless steel-lined equipment pit, fuel handling tools and other tools can be cleaned and decontaminated.

For the personnel, a decontamination shower and washroom is located on the 72 inside Unit 1 NSB. Personnel decontamination kits with instructions for their use are in the radiation control area decon room.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-68 of 3-126 Revision 0, 2020 3.8 Primary Auxiliary Building Ventilation System 3.8.1 Design Basis The primary auxiliary building ventilation system is designed to accomplish the following:

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

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 monitored and diluted so that offsite dose will not exceed Offsite Dose Calculation Manual (ODCM).

3.8.2 System Design and Operation The primary auxiliary building ventilation system (See Drawing 9321-4022 is composed of the following systems:

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

2. Exhaust system complete with fans and ductwork.

Design parameters for the system components are given in Table 3.8-1.

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, 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 three fans in the primary auxiliary building ventilation system. The two exhaust fans (primary auxiliary building exhaust fans 21 and 22) and the supply fan.

The primary auxiliary building supply fan normally runs, along with either or both of the exhaust fans. 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 fans can be started and stopped by 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.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-69 of 3-126 Revision 0, 2020 TABLE 3.8-1 Primary Auxiliary Building Ventilation System Component Data System Units Installed Units Capacity Units Required for 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

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

Coils 1

50,400 cfm 1

Notes:

1.

These two exhaust fans are used interchangeably for the ventilation of primary auxiliary building.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-70 of 3-126 Revision 0, 2020 3.9 Control Room Ventilation System 3.9.1 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. 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. Duct system complete with dampers and controls.

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 Drawings 252665 and 138248 [Formerly Figure 3.9-1]. The Unit 2 control room ventilation system can be operated as follows:

1.

Normal Condition With outside air makeup will supply cooling or heating for the control room atmosphere as required, using fresh outside air makeup.

2.

Incident Condition On toxic gas and/or smoke signal, the outside makeup air will be isolated, the system will be in 100% recirculation mode.

These operations are performed manually from the control room. A redundant toxic chemical and radiation monitor for central control room air intakes has been installed.

3.9 FIGURES Figure No.

Title Figure 3.9-1 Central Control Room HVAC (Heating, Ventilation, and Air Conditioning), Replaced with Drawings 252665 & 138248

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-71 of 3-126 Revision 0, 2020 3.10 Fuel Storage Building Ventilation System 3.10.1 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 SFP, from areas of low potential contamination to areas of higher potential contamination.

3. Remove normal building heat.

3.10.2 System Design and Operation The fuel storage building ventilation system consists of an exhaust system. In addition, an axial spot cooling fan circulates 3000 cfm of air to the spent fuel pit heat exchanger room.

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 SFP surface are intended to provide a sweep flow over the SFP.

Air from the registers is ducted to a plenum chamber 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.

3.10.3 Limiting Conditions for Operation 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.

3.10.4 Surveillance Requirements The fuel storage building ventilation system does not have to be demonstrated functional in the assumed configuration prior to handling fuel. The fuel storage building ventilation system shall be periodically tested to verify 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.

3.11 Circulating Water System The circulating water system provides dilution flow for liquid waste discharges.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-72 of 3-126 Revision 0, 2020 Hudson River water is used for the condenser circulating water. River water flows under the floating debris skimmer wall, through traveling screens, and into six separate screenwells. The traveling screens, which operate continuously, are designed to reduce the potential for fish and debris from entering the circulating water pumps. Each screenwell is provided with stop logs to allow dewatering of any individual screenwell for maintenance purposes.

The water from each individual screenwell flows to a motor-driven, vertical, mixed flow condenser circulating water pump. Each of the six condenser circulating water pumps provides 140,000 gpm and 21-ft total dynamic head when operating at 254 rpm and 84,000 gpm and 15-ft total dynamic head when operating at 187 rpm. Each pump is located in an individual pump well, thus tying a section of the condenser to an individual pump. The circulating water is piped to the condensers and is discharged back into the river.

A surface-type, single-pass, radial flow condenser (#22) with a bolted divided water box at both ends is provided. Fabricated steel water box and shell construction is used. Condenser #22 uses titanium tubes and tube sheets. Water box manholes are provided for access. The design parameters for Condenser #22 are given in Table 3.11-1.

The pressure-retaining components or compartments of components comply, as a minimum, with the codes detailed in Table 3.11-2.

TABLE 3.11-1 Design Parameters for Condenser #22 Type Radial flow, single-pass, divided water box, deaerating Number 1

TABLE 3.11-2 Codes and Classifications System pressure vessels and pump casing ASME Boiler and Pressure Vessel Code,Section VIII System valves, fittings, and piping USAS Section B31.1 Power Piping Code (1955) ASA, USAS, ANSI Pressure Testing of Repairs and Modifications USAS Section B31.1 Power Piping Code (1992) 3.11 FIGURES Figure No.

Title Figure 3.11-1 Condenser Air Removal and Water Box Priming - Flow Diagram, Replaced with Drawing 9321-2025

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-73 of 3-126 Revision 0, 2020 3.12 LEAKAGE DETECTION AND PROVISIONS FOR THE AUXILIARY COOLANT LOOPS 3.12.1 Design Bases Monitoring Radioactivity Releases Criterion:

Means shall be provided for monitoring the containment atmosphere and the facility effluent discharge paths for radioactivity released from normal operations, from anticipated transients, and from accident conditions. An environmental monitoring program shall be maintained to confirm that radioactivity releases to the environs of the plant have not been excessive. (GDC 17)

IP2 has been permanently shut down and defueled. As a result, there are no abnormal operations, transient, or accidents that credited the containment for isolation. The component cooling loop liquid is monitored for radioactivity concentration during normal operation. The Offsite Dose Calculation Manual (ODCM) provides the methodology to calculate radiation does rates and dose to individual persons in unrestricted areas in the vicinity of Indian Point due to the routine release of liquid effluents to the discharge canal. The ODCM also provides setpoint methodology that is applied to effluent monitors and optionally to other process monitors.

3.12.2 Systems Design and Operation For relevant systems located outside the containment, leakage is determined by one or more of the following methods:

1. For systems containing radioactive fluids, leakage to the atmosphere would result in an increase in local atmospheric activity levels and would be detected by either the plant vent monitors or by one of the area radiation monitors. Similarly, leakage to other systems that do not normally contain radioactive fluids would result in an increase in the activity level in that system.

2. For closed systems such as the component cooling system, leakage would result in a reduction in fluid inventory.

3. All leakage would collect in specific areas of the building for subsequent handling by the building drainage systems, e.g., leakage from the service water loop would collect in the sumps provided, and would result in the operation, or increased operation, of the associated sump pumps and increased inventory in the liquid waste processing system.

The relevant fluid systems for which no special leak detection outside containment is provided include the following:

1. Component cooling.

2. Service water.

3. Waste disposal.

Various methods are used to detect leakage from the auxiliary loops. Although described to some extent under each system description, all methods are included here for completeness.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-74 of 3-126 Revision 0, 2020 3.12.3 Components 3.12.3.1 Component Cooling Liquid Monitor This channel continuously monitors the component cooling loop of the auxiliary coolant system for activity indicative of a leak from the spent fuel pool cooling system. A scintillation detector is installed in the local radiation monitor skid assembly. This assembly is located in the primary auxiliary building and receives sample flow from the component cooling pump discharge downstream of the component cooling heat exchangers. The detector assembly output is amplified by a preamplifier, processed and transmitted to the radiation monitoring system console, the display console and a recorder in the control room. The activity is indicated on digital displays.

High-activity alarm indications are displayed on the control board annunciator and the display console.

The measuring range of this monitor is 10-5 to 10-2 µCi/cm3.

3.12.3.2 Component Cooling Loop Leakage outside containment depending on location will be diverted by floor drains to either the PAB sump tank or PAB sump where it will then be transferred to the Waste Holdup Tank.

3.12.3.3 Service Water System Leakage outside containment depending on location will be diverted by floor drains to either the PAB sump tank or PAB sump where it will then be transferred to the Waste Holdup Tank.

3.12.4 Leakage Provisions Provisions are made for the isolation and containment of any leakage.

3.12.4.1 Design Basis The provisions made for leakage are designed to prevent uncontrolled leaking of auxiliary cooling water. This is accomplished by routing the leakage to various sumps and holdup tanks.

3.12.4.2 Design and Operation Various provisions for leakage avert uncontrolled leakage from the auxiliary coolant loops.

3.12.4.3 Component Cooling Loop Leakage outside containment depending on location will be diverted by floor drains to either the PAB sump tank or PAB sump where it will then be transferred to the Waste Holdup Tank.

Other provisions made for leakage from the component cooling loop are discussed in Section 3.3.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-75 of 3-126 Revision 0, 2020 3.12.4.4 Service Water System Leakage outside containment depending on location will be diverted by floor drains to either the PAB sump tank or PAB sump where it will then be transferred to the Waste Holdup Tank.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-76 of 3-126 Revision 0, 2020 3.13 Information Display and Recording Alarms and annunciators in the central control room provide site personnel with warning of abnormal facility conditions that might lead to the damage of components, fuel, or other unsafe conditions. Other displays and recorders are provided for indication of routine conditions and for the maintenance of records.

Control and display equipment for station auxiliary systems are located on the control board.

The auxiliary electrical system controls required for manual switching between the various power sources described in Section 3.15.1.2 are provided to the left of the control board.

Controls and indications Primary Auxiliary Building and Fuel Service Building ventilation systems are located on CCR panel SL.

Audible alarms will be sounded in appropriate areas throughout the station if high-radiation conditions are present.

A process computer system is installed with color graphic displays in the central control room that monitors facility data as well as easily accessible sets of key facility safety parameters. It also provides data links with the technical support center, the emergency operations facility and the Alternate emergency operations facility. It has the capability of long-term data storage and retrieval.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-77 of 3-126 Revision 0, 2020 3.14 Communications Facility communications are conducted via telephone, radio, and Public Address (paging) systems.

The facility telephone and radio communications systems include PBX electronic switches, backup phone lines and a UHF radio system.

The public address system for Indian Point Unit 2 consists of "Page" and "Party" communications, which are common to both the primary (nuclear) and secondary (conventional) portions of Units 1 and 2. The Page and Party communications are also monitored at a speaker panel located in the CCR. Radio channels are available at the Indian Point Unit 2 control room. These radio channels are as follows:

1. Radios provide the central control room with radio communication to site personnel.

2. Indian Point area radios provide the central control room with radio communication to the emergency response facilities and offsite monitoring teams.

If the control room were to become inaccessible, communications would be conducted with the use of portable radios. This in-house radio system is also provided for communicating with site personnel.

3.14.1 Central Control Room Communication Facilities The central control room is provided with telephone-radio-page/party communication consoles and page/party handset stations.

A State/County Radiological Emergency Communication System (RECS) hotline is available.

The NRC Emergency Notification System (ENS) hotline is available in a separate location.

A separate printer and its telephone modem are also available for meteorological data reception.

3.14.2 Radio Communication The radio channels are available at the radio line consoles in the central control room.

Repeaters for the radio channels are located onsite. Wired audio/control pairs connect the transceivers with the communication consoles in the central control room for remote operation.

3.14.3 Page/Party Line Communication Page or Party line communication can be initiated in the CCR from either communication consoles or from handset stations.

An emergency alarm switch is provided in the CCR to connect and actuate the existing alarm oscillators to the Page system for the "Evacuation," "Fire," or "Air Raid" alert signals.

Another switch is provided on the central control room desk, which allows all outdoor speakers of the Indian Point 2 facility to be turned off at night.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-78 of 3-126 Revision 0, 2020 3.14.4 Backup Power for Communications The facility radio and telephone communications systems are automatically supplied from a back-up power source, upon failure of the normal power source. In addition, each PBX is provided with back-up battery capability of eight (8) hours of operation. The page/party system is powered from the DC system (through an inverter) with backup power from a bus.

3.14.5 In-house Radio System An in-house radio system provides communications between personnel at the facility. Field units are low-wattage, hand-held units, which are not to be used in areas containing equipment, which is potentially sensitive to radio-frequency interference.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-79 of 3-126 Revision 0, 2020 3.15 Electrical Systems The function of the auxiliary electrical system is to provide reliable power to those auxiliaries required during any normal facility conditions. Sufficient independence and isolation between the various sources of electrical power is provided in order to guard against concurrent loss of all auxiliary power. The facility is supplied with normal and standby power sources. Facility power is provided by a 13.8-kV / 6.9-kV autotransformer.

A diesel-generator set supplies standby power to the facility in the event of a loss of AC auxiliary power. There are no automatic bus ties associated with these buses. The Station Blackout (SBO) / Appendix R diesel-generator is installed in the Unit 1 Turbine Building and is used to supply standby power for the facility.

The standby diesel-generator set is located in the Diesel Generator Building adjacent to the Primary Auxiliary Building and supplies a source of standby power. It may be started manually upon the occurrence of an undervoltage condition on any 480-V switchgear bus. The standby diesel is adequate to provide standby power to ensure the safe storage of spent fuel at the facility in lieu of the SBO / Appendix R diesel generator.

All electrical systems and components that were historically vital to plant safety, including the standby diesel generator, were classified as seismic Class I and were designed so that their integrity was not impaired by the design-basis earthquake, certain wind storms, floods, or disturbances on the external electrical system.

The 138-kV outside source of power and the 13.8-kV / 6.9-kV autotransformer are adequate to run all of the facility auxiliary loads.

The 125-V DC power supply consists of a battery charger and a battery panel. Under all conditions, the battery charger supplies all DC loads. DC control power for the 480-V Switchgear and the standby diesel generator is supplied via the 125-V DC system. This design ensures that adequate DC power is available.

The 6.9-kV system is arranged as six buses. The buses receive power from the offsite 13.8 kV source. Buses 2, 3, 5, and 6 each serve one of the four 6900-V / 480-V station service transformers. Normal and offsite power to the 480-V switchgear buses is supplied through these station service transformers.

The 480-V system is arranged as four switchgear buses. Each 480-V switchgear bus supplies several 480-V motor control center buses for power distribution throughout the facility. The 480-V switchgear buses are supplied from the 6.9-kV buses as follows: 2A from 2, 3A from 3, 5A from 5, and 6A from 6. Tie breakers are provided between 480-V switchgear buses 2A and 3A, 2A and 5A, and 3A and 6A.

Power for instrumentation and control is provided by four 118-V AC Instrument Supply Systems.

Each system consists of one manual bypass switch, two 118-V AC buses, and associated interconnections. The four systems are powered from transformers supplied from 480-V MCCs.

Several sources of offsite power are available to Indian Point Unit 2. These consist of three separate underground feeders from the Buchanan 13.8-kV substation. The 13.8-kV line is rated

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-80 of 3-126 Revision 0, 2020 19.8 MVA at 13.8-kV. The 13.8-kV / 6.9-kV autotransformer is rated 20 MVA. No safety or emergency power is required from these sources for the retired Indian Point Unit 1.

The source of offsite power from the 13.8-kV distribution system at Buchanan is available to 6.9-kV buses 5 and 6 through supply breakers GT-25 and GT-26. The transfer from the normal to the reserve supply (or vice versa) must be accomplished manually.

The diversity and redundancy inherent in the combination of offsite electrical systems minimize the probability of losing electric power from any of the remaining sources as a result of, or coincident with, the loss of power from the transmission network, or the loss of onsite power sources.

3.15.1 ELECTRICAL SYSTEM DESIGN 3.15.1.1 Network Interconnections The external transmission system provides auxiliary power as required by the facility.

The electrical one-line diagram for the Indian Point Station is presented in Drawing 250907

[Formerly Figure 3.15-1]. Power is supplied to the facility from the Buchanan 138-kV Substation.

Several power flow paths exist to connect 13.8-kV buses through 13.8-kV / 6.9-kV autotransformers to Buses 5 and 6.

3.15.1.1.1 Reliability Assurance Two external sources of power are available to Indian Point Unit 2. They are the 138-kV tie from the Buchanan 345-kV / 138-kV autotransformer and the 138-kV Buchanan-Millwood ties. These 138-kV ties normally power the 13.8-kV sources via 138-kV / 13.8-kV transformers in the Buchanan switchyard. Loss of any of these 138-kV sources will not affect the other. Substantial flexibility and alternate paths exist within each source.

The 138-kV supply from the Buchanan substation with its connections to the Con Edison 345-kV system provides a dependable source of station auxiliary power. Upon loss of 345-kV / 138-kV auto-transformer supply at Buchanan, two 138-kV ties are designed to provide additional auxiliary power from the Millwood 138-kV substation. A further guarantee of reliable auxiliary power, independent of transmission system connections, is provided by the SBO / Appendix R Diesel.

The SBO / Appendix R Diesel, associated switchgear and breakers minimum operating requirements are specified in Appendix B of the DSAR (the Unit 2 Technical Requirements Manual (TRM)). A minimum quantity of fuel for the SBO / Appendix R Diesel shall be available at all times the SBO / Appendix R Diesel is considered functional. Support systems for cooling include the City Water Storage Tank and the Service Water System (SWS) (first the city water and then a switch to the SWS). If these requirements cannot be met, then the SBO / Appendix R diesel generator is considered non-functional and the TRM requirements are followed.

The fuel supply consists of two onsite 30,000-gal fuel oil tanks. A minimum amount of fuel is maintained available and dedicated for the SBO / Appendix R Diesel. This minimum fuel inventory ensures that the SBO / Appendix R Diesel will be capable of supplying the maximum electrical load for the Indian Point Unit 2. Commercial oil supplies and trucking facilities exist to ensure deliveries of additional fuel within one day's notice.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-81 of 3-126 Revision 0, 2020 3.15.1.2 Distribution System The auxiliary electrical system is designed to provide a simple arrangement of buses requiring a minimum of switching to restore power to a bus in the event that the normal supply is lost.

The basic components of the facility electrical system are shown on the electrical one-line diagrams (See Drawings 208377, 231592, 208088, 9321-3004, 249956, 9321-3005, 208507, 249955, 208241, 9321-3006, 248513, 208500, 208502, 208503, 208501, 9321-3008, and Figure 3.15-4 [Formerly Figures 3.15-3, and 3.15-5 through 3.15-16]), which include the 13.8-kV, the 6.9-kV, the 480-V, the 118-V AC instrument, and the 125-V DC systems.

3.15.1.2.1 Gas Turbine Autotransformer and Station Service Transformers Power to the auxiliaries on the 6.9-kV buses is supplied by a 13.8-kV / 6.9-kV two-winding autotransformer connected to an offsite supply. Power to the 480-V buses is supplied from four 6900-V / 480-V, air-insulated, dry-type station service transformers.

These transformers were designed and constructed in accordance with ANSI C57.11, as the applicable standard of record at the time of fabrication. During engineered safeguards loading and operation, these transformers are loaded within their ratings. Manufacturer shop tests of the transformers were conducted in accordance with the American Standard Test Code C 57.12.90.

This series of tests consisted of the following:

1. Resistance measurements of all windings.

2. Ratio tests.

3. Polarity and phase relation tests.

4. No-load losses.

5. Exciting current.

6. Impedance and load loss.

7. Temperature test.

8. Applied potential tests.

9. Induced potential tests.

The normal source of power to the 480-V buses is supplied from the 138-kV switchyard.

3.15.1.2.2 6.9-kV System The 6.9-kV system is arranged as six buses that receive power from the 13.8-kV system by bus main breakers and the 13.8-kV / 6.9-kV autotransformer. Buses 2, 3, 5, and 6 each serve one 6900-V / 480-V station service transformer.

3.15.1.2.3 480-Volt System The 480-V system is arranged as Switchgear buses 2A, 3A, 5A, and 6A and numerous motor control center buses. The 480-V switchgear buses are supplied from the 6.9-kV buses as follows:

2A from 2, 3A from 3, 5A from 5, and 6A from 6 (buses 2A and 3A are within the same power train). Tie breakers are provided between 480-V Switchgear buses 2A and 3A, 2A and 5A, and 3A and 6A.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-82 of 3-126 Revision 0, 2020 All four 480-V switchgear buses supply power to systems and equipment. A single source of DC control power is provided for control of 480-V breakers, protective circuits and other devices.

The preferred and alternate sources of DC control power for the breakers are:

Transfer Switch Associated Bus Preferred Source Alternate Source EDD1 6A DC PP #24 DC PP #22 EDD2 2A DC PP #22 DC PP #24 EDD3 3A DC PP #23 DC PP #21 EDD4 5A DC PP #21 DC PP #23 3.15.1.2.4 125-V DC Systems There is a single 125-V DC system serving the various DC loads throughout the facility. The system consists of one battery, one battery charger and one or more DC panels.

The battery charger is supplied from a 480-V motor control center bus. The battery charger supplies the DC loads.

3.15.1.2.5 118-V AC Instrument Supply Systems There are four 118-V AC instrument supply systems serving the various instrumentation and control systems throughout the facility. Each system consists of one manual switch and two 118-V AC instrument buses (See Drawing 250970 [Formerly Figure 3.15-2] for system arrangement and connections to power sources). All four systems are supplied from transformers supplied from 480-V MCCs. The 118-V AC system manual transfer switch is mounted in a separate enclosure and will bypass the static transfer switch and provide power from the step-down transformers directly to the 118-V AC buses. Voltage drop calculations demonstrate that equipment supplied from Buses 21 and 21A are operable with the postulated minimum source voltage. This is typical of all instrument buses.

3.15.1.2.6 Evaluation of Layout and Load Distribution Electrical distribution system equipment is located to minimize the exposure of relevant circuits to physical damage as a result of natural phenomena. To a certain extent the Diesel-Generator Building is protected from tornados and major tornado generated major missiles because it is situated between large buildings as shown in the site plot plan (Drawing 9321-1002). The diesel-generator installation is considered redundant to other lines of power supply. As described in Section 3.15, there are alternate power supplies. In the case of a tornado, reliance is placed on power supply redundancy and not solely on the diesel installation.

The 6.9-kV buses are housed in two metal-clad switchgear units. The enclosures for switchgear 21 and 22 are located at elevation 15 ft in the turbine building. Each breaker is mounted in a separate compartment. Switchgear 21 and 22 have a solid top with cable penetrations and some openings on the side. The cable openings at the top are sealed to minimize bus exposure to fire, water, and other physical damage. An overcurrent condition on any of the 6.9-kV buses actuates the associated bus protection lockout relays, which isolate the bus by tripping and locking out both the normal supply breaker and the 6.9-kV tie breaker for that bus.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-83 of 3-126 Revision 0, 2020 The 480-V buses are housed in two metal-enclosed switchgear units located at the 15-ft elevation of the Indian Point Unit 2 control building. The switchgear structure provides protection to minimize exposure from mechanical, fire, and water damage. Buses 5A and 2A are contained in switchgear enclosure 21; Buses 6A and 3A constitute switchgear enclosure 22. The switchgear contains the buses, the bus supply breakers, the tie breakers, the load (feeder) breakers, the station service transformers, and the potential transformers for synchronizing and under-voltage relay protection. The normal 480-V switchgear supply breakers 52/2A, 52/3A, 52/5A, and 52/6A are tripped under the following conditions:

1. Loss of voltage (~46-percent) on bus 5A or 6A.

2. Actuation of manual trip pushbuttons on each breaker.

3. Actuation of control switches in the Central Control Room.

4. Actuation of control switches in the Diesel-Generator Building.

5. Individual breaker overcurrent protection.

A separate category alarm and bullet lights in the central control room will alert site personnel when any 480-V switchgear bus voltage falls to 94-percent. These are primarily intended to alert site personnel to sustained degraded voltages that result from problems on the offsite power system.

Remote manual and automatic control of the 480-V switchgear breakers and associated relays requires 125-V DC control power.

Each 480-V switchgear breaker is equipped with a Westinghouse "Amptector 1A"* solid-state overcurrent trip unit to protect the auxiliary equipment supplied by the breaker (including cables) and the associated switchgear. The settings of the solid-state overcurrent trip unit are based on the supplied load. The solid-state trip unit is provided with an instantaneous and/or short-time setting(s) to protect against fault conditions, and long-time setting to protect against over-load conditions.

Each circuit breaker is tripped on overcurrent conditions (overload or short circuit) by the combined operations of three components:

1. Sensors

2. Amptector solid-state trip unit *

3. Actuator All necessary tripping energy (for a breaker trip on an overcurrent condition only) is derived from the load current flowing through the sensors; no separate power source is required. The tripping characteristics for a specific breaker rating, as established by the sensor rating, are determined by the continuously variable settings of the Amptector* static trip unit. This unit supplies a pulse of tripping current (when preselected conditions of current magnitude and duration are exceeded) to the actuator, which produces a mechanical force to trip the breaker.

If an overcurrent condition occurs on one of the 480-V switchgear buses while the bus is supplied from the normal source, lockout relays trip (if required) and prevent the closing of the alternate supply breakers (standby diesel and bus ties) associated with the bus. These relays must be manually reset after the overcurrent condition is cleared to allow these breakers to close.

The 480-V motor control centers are located in the areas of electrical load concentration.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-84 of 3-126 Revision 0, 2020

• Note that the Amptector may be replaced by the equivalent Westinghouse device, Westector.

The Indian Point Unit 2 Cable Raceway System is comprised of 4 raceway systems. 6.9kV cables are routed in their own raceway system independent of the other raceway systems. 480 VAC and 125 VDC cable 350 mcm and larger are routed in the heave Power Raceway. Those cables smaller than 350 mcm and over 65VAC are routed in the Small Power and control Raceway.

Instrument cables 65VAC and less are run in the Instrument Raceway. Instrument cables less than 65VAC are typically routed in the Instrument Raceway. On a case by case basis, cables have been routed in an alternate raceway however there is no mixing between the 6.9kV raceway and cables of lower voltages. Certain other cables such as thermocouple cable, public address, instrument power and fiber optics are routed in raceway as convenient.

The application and routing of control, instrumentation, and power cables minimize their exposure to damage from any source. All cables are designed using conservative margins with respect to their current carrying capacities, insulation properties, and mechanical construction. All cables are fire resistant.

Cable loading of trays and consequently heat dissipation of cable throughout the facility has been carefully studied and controlled to ensure that there is no overloading. The criteria for electrical loading were developed using IPCEA (now ICEA) Standard P-46-426, manufacturer recommendations, and good engineering practice.

Derating factors for cables in trays without maintained spacing are taken from Table VIII of the IPCEA publication. Derating factors for the maximum ambient temperature existing in any area of the facility are also taken from the IPCEA publication. These factors are applied against ampacities selected from appropriate tables in other portions of the standard.

For physical loading of trays, the following criteria are followed: for 6.9-kV power, one horizontal row of cables is allowed in a tray; for heavy power, two horizontal rows of cables are allowed; for medium power, small power & control or instrumentation, 70-percent of the cross-sectional area of a tray is the maximum fill, with the heavy power cables limited to two horizontal rows. During initial plant construction, a computer program monitored the loading and prevented the routing of anything greater than this amount.

To ensure that only fire-retardant cables are used throughout the facility, a careful study of cable insulation systems was undertaken early in the design of the plant. Insulation systems that appeared to have superior flame-retardant capability were selected and manufacturers were invited to submit cable samples for testing. An extensive flame testing program was conducted including ASTM vertical flame and Con Edison vertical flame and bonfire tests. A report summarizing the testing was prepared by Con Edison. These tests were used as one of the means of qualifying cables, and the specifications were written on the basis of the results.

The following tests were made to determine the flame-retardant qualities of the covering and insulations of various types of cables for Indian Point Unit 2:

1. Standard Vertical Flame Test - made in accordance with ASTM-D-470-59T, Tests for Rubber and Thermoplastic Insulated Wire and Cable.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-85 of 3-126 Revision 0, 2020

2. Five-Minute Vertical Flame Test - made with cable held in vertical position and 1750°F flame applied for 5 min.

3. Bonfire Test - consisted of exposing bundles of three or six cables to flame produced by igniting transformer oil in a 12-in. pail for 5 min. The cable bundles were supported horizontally over the center of the pail with the lowest cable 3 in. above the top of the pail.

The time required to ignite the cable and the time the cable continued to flame after the fire was extinguished were noted.

In those areas where the compressed instrument air system is near the essential 480-V switchgear, the following provision is incorporated to shield this switchgear and cabling from potential missiles or pipe whip:

1. The compressed instrument air lines in the vicinity of the switchgear are supported at the piping bends. This will resist any step loading of PA (which could occur in the event of an instantaneous circumferential rupture) without occurrence of a "plastic hinge." The possibility of pipe whip is eliminated.

These provisions ensure that no missile or whipping pipe originating from postulated failures in the compressed instrument air system will strike the switchgear.

3.15.1.3 Standby Power 3.15.1.3.1 Source Descriptions The source of offsite power is the Con Edison 138-kV system. A diesel-generator set provides a source of onsite standby power that can be used in lieu of the SBO / Appendix R diesel generator.

The diesel generator set is an Alco Model 16-251-E engine coupled to a Westinghouse 900 rpm, 3-phase, 60-cycle, 480-V generator. The unit has a capability of 1750 kW (continuous), 2300 kW for 1/2 hour in any 24-hour period, and 2100 kW for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in any 24-hour period.

The standby diesel unit, which is a backup to the normal standby AC power supply, is capable of supplying the power requirement for facility equipment. The unit is installed in a historically classified seismic Class I structure located near the Primary Auxiliary Building.

The standby diesel is manually started by two redundant air motors, each unit having a complete 53-ft3 air storage tank and compressor system powered by the 480-V distribution system. Each air receiver has sufficient storage for four normal starts. However, the standby diesel will consume only enough air for one automatic start during any particular power failure. Additionally, the engine control system is designed to shut down and lock out any engine that did not start during the initial try.

To ensure rapid start, the unit is equipped with water jacket and lube-oil heating. A pre-lube pump circulates the oil when a unit is not running. The unit is located in a heated room.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-86 of 3-126 Revision 0, 2020 Audible and visual alarms are located in the control room and in the diesel generator building.

Alarms on the electrical annunciator panel in the control room are:

1. Diesel-generator trouble.

2. Diesel-generator oil storage tank low level.

3. Diesel-Generator Trouble.

4. Diesel-Generator Service Water Flow Low.

The activation of the standby diesel generator trouble alarm in the control room will be caused by the initiation of any of the following alarms in the diesel generator building:

1. Low oil pressure.

2. Differential fuel strainer, secondary.

3. Over-crank.

4. High differential lube-oil strainer.

5. High water temperature.

6. High differential pressure lube-oil filter.

7. High-high jacket water temperature.

8. Overspeed.

9. Overcurrent.

10. Low fuel oil level, day tank.

11. Reverse power.

12. Low start air pressure.

13. Exciter field shutdown.

14. High/Low lube-oil temperature.

15. High differential pressure primary filter.

The diesel-generator oil storage tank low level alarm will be energized on a low level in the fuel-oil storage tank.

The alarm Diesel-Generator Trouble located on Panel SG in the Central Control Room will be activated respectively by the following conditions at the diesel general local control panel:

1. Loss of DC control power.

2. Engine control switch position (Off or Manual).

3. Breaker control switch position pulled-out [Note - the breaker control switch in the CCR will activate the Safeguards Equipment Locked Open alarm (Window 1-8 on Panel SB-1) in the CCR].

4. Engine stop solenoid energized.

5. Day tank level low, primary and backup fuel pump fails to start.

There are six electrical contacts, each of which when activated will energize a diesel-generator lockout relay. This lockout relay will, in turn, cause the standby diesel to shut down if it is operating. These contacts are activated by one of the following conditions:

1. Activation of the standby diesel emergency stop push-button in the diesel-generator building.

2. Activation of the overcurrent relay. A phase-to-phase fault or excessive loads on the standby diesel generator will operate this relay.

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3. Activation of the reverse power relay.

4. Activation of the over-crank relay. If the standby diesel engine fails to attain speed within 13 sec, this relay will be energized.

5. Activation of the overspeed relay. When the mechanical governor senses 1070 rpm, this relay will be energized.

6. Activation of the low oil pressure relay. This relay is energized by the coincident sensing of lube-oil pressure below 60 psi by two of the three oil pressure switches for the standby diesel.

An oil pressure timer is set to allow 20 sec to pass before tripping the standby diesel engine lockout relay. This circuit is designed to provide sufficient time for the oil pressure to build up following an engine start.

Shutdown permits corrective action to be taken before the engine is damaged, and the standby diesel generator can then be returned to normal operation. Once any of these six electrical contacts has been activated causing the standby diesel engine lockout relay to energize, the lockout relay must be manually reset locally before the diesel can be started.

3.15.1.3.2 Standby Fuel Supply The standby diesel generator has a 175-gal fuel-oil day tank plus an underground bulk storage supply tank and uses diesel oil Specification Number 2. The day tank is located within the diesel-generator building and supplies its engine-mounted fuel-oil pump. The day tank is automatically filled during engine operation from its separate underground storage tank located outside adjacent to the diesel-generator building. The storage tank has a capacity of 7700 gal and is provided with a motor-driven transfer pump mounted in a manhole opening above oil level. If a low level is detected in the day tank for the standby diesel generator, its transfer pump will automatically start to refill the tank to approximately 158 gal.

The diesel oil transfer pump stops automatically when 15.5-in. of oil remains in the underground tank which equates to a maximum of approximately 7000-gal of available fuel oil.

As previously mentioned in Section 3.15.1.1.1, commercial oil supplies and trucking facilities exist to ensure deliveries on one day's notice.

3.15.1.3.3 Standby Diesel Generator Location The standby diesel generator is located in a sheet metal, steel-framed building immediately South of the Primary Auxiliary Building. The engine foundation is surrounded by a 1-foot-high concrete curb containing sufficient volume to hold all the lube-oil or fuel released from a single engine in the event of an inadvertent spill or line break.

Diesel generator fire protection features necessary to meet the criteria of 10 CFR 50.48(f) are described in the document under separate cover entitled, "IP2 Fire Hazards Analysis." A control panel, which contains relays and metering equipment for the standby diesel generator is located on the west end of the building. A reinforced-concrete wall separates the standby diesel generator from the control panel.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-88 of 3-126 Revision 0, 2020 3.15.1.3.4 Loading Description The standby diesel-generator is manually started on the occurrence of an undervoltage on any 480-V switchgear bus.

On undervoltage on any bus, the engine is manually started and run at idle and can be connected to deenergized buses by site personnel from the control room or locally.

3.15.1.3.5 Batteries and Battery Chargers The battery installation is composed of individual lead-calcium storage cells connected to provide a nominal terminal voltage of 125-V DC. The battery is fed from a charger that is fed from a 480-V motor control center. The battery bus is equipped with a sensitive-type undervoltage relay, which provides alarm / indication of an undervoltage condition. Ground alarms are also provided on the chargers board. Improved status indication of the battery charger and the direct current system has been provided by a DC bus trouble alarm. Loads on the DC system are shown on Drawings 208501 and 9321-3008 [Formerly Figures 3.15-15 and 3.15-16].

3.15.1.3.6 Reliability Assurance The 480-V equipment is arranged on four buses. The 6.9-kV equipment is supplied from six buses.

The outside source of power is adequate to run all normal operating equipment. The 13.8-kV / 6.9-kV autotransformer can supply all the auxiliary loads.

A diesel generator (SBO / Appendix R diesel generator or standby diesel generator) has enough capacity to provide standby power to the facility.

A total loss of DC feed to the switchgear and associated equipment will not cause a loss of offsite power through an inadvertent tripping of the Indian Point Unit 2 light and power supply circuit breakers, because DC power is required to trip a breaker. Loss of DC feed to protective relaying will cause an alarm condition rather than initiation of a protective action. If necessary, the light and power circuit breakers in the Buchanan substation may be tripped manually at the breaker mechanisms.

The equipment arrangement in the Indian Point Unit 2 Central Control Room is discussed in Section 3.13.

3.15.2 TESTS AND INSPECTIONS The fuel supply and starting circuits and controls are continuously monitored and any faults are alarm indicated. An abnormal condition in these systems would be signaled without having to place the standby diesel generator on test. The standby diesel generator will be inspected in accordance with a licensee-controlled maintenance program. The maintenance program will require inspection in accordance with the manufacturer's recommendation for this class of standby service.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-89 of 3-126 Revision 0, 2020 The SBO / Appendix R diesel and support systems shall be tested and have surveillances in accordance with the TRM. These tests and surveillances are designed to assure that the SBO / Appendix R diesel will be available to provide power for operation of equipment, if required.

3.15 FIGURES Figure No.

Title Figure 3.15-1 Electrical One-Line Diagram, Replaced with Drawing 250907 Figure 3.15-2 Electrical Power System Diagram, Replaced with Drawing 250907 Figure 3.15-3 Main One-Line Diagram, Replaced with Drawing 208377 Figure 3.15-4 345-KV Installation at Buchanan Figure 3.15-5 6900-V One-Line Diagram, Replaced with Drawing 231592 Figure 3.15-6 480-V One-Line Diagram, Replaced with Drawing 208088 Figure 3.15-7 Single Line Diagram 480-V Motor Control Centers 21, 22, 23,25, 25A, Replaced with Drawing 9321-3004 Figure 3.15-7a Single Line Diagram - 480-V Motor Control Centers 24 and 24A, Replaced with Drawing 249956 Figure 3.15-8 Single Line Diagram - 480-V Motor Control Centers 27 and 27A, Replaced with Drawing 9321-3005 Figure 3.15-9 Single Line Diagram - 480-V Motor Control Centers 28 and 210, Replaced with Drawing 208507 Figure 3.15-9a Single Line Diagram - 480-V Motor Control Centers 29 and 29A, Replaced with Drawing 249955 Figure 3.15-10 Single Line Diagram - 480-V Motor Control Centers 28A and 211, Replaced with Drawing 208241 Figure 3.15-11 Single Line Diagram - 480-V Motor Control Centers 26A and 26B, Replaced with Drawing 9321-3006 Figure 3.15-11a Single Line Diagram - 480-V Motor Control Center 26C, Replaced with Drawing 248513 Figure 3.15-12 Single Line Diagram - 480-V Motor Control Centers 26AA and 26BB and 120-V AC Panels No. 1 and 2, Replaced with Drawing 208500 Figure 3.15-13 Single Line Diagram - 118-VAC Instrument Buses No. 21 thru 24, Replaced with Drawing 208502 Figure 3.15-14 Single Line Diagram - 118-VAC Instrument Buses No. 21A thru 24A, Replaced with Drawing 208503 Figure 3.15-15 Single Line Diagram - DC System Distribution Panels No. 21, 21A, 21B, 22, and 22A, Replaced with Drawing 208501 Figure 3.15-16 Single Line Diagram - DC System Power Panels No. 21 thru 24, Replaced with Drawing 9321-3008 Figure 3.15-17 Single Line Diagram of Unit Safeguard Channeling and Control Train Development, Replaced with Drawing 208376 Figure 3.15-18 Cable Tray Separations, Functions, and Routing, Replaced with Drawing 208761

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-90 of 3-126 Revision 0, 2020 3.16 CONTAINMENT STRUCTURES 3.16.1 Design Basis Historically, the reactor containment completely enclosed the entire reactor and reactor coolant system and ensured that essentially no leakage of radioactive materials to the environment would result even if gross failure of the reactor coolant system were to occur. The liner and penetrations were designed to prevent any leakage through the containment. The structure provided biological shielding for both normal and accident situations. In the permanently shut down and defueled condition, the reactor containment performs no active function. However, it must remain capable of withstanding natural phenomenon, so that it does not damage any Class I SSC.

The reactor containment is designed to safely withstand several conditions of loading and their credible combinations. The major loading conditions are an earthquake or wind.

3.16.1.1 Principal Design Criteria 3.16.1.1.1 Quality Standards Criterion:

Those systems and components of reactor facilities, which are essential to the prevention, or the mitigation of the consequences, of nuclear accidents, which could cause undue risk to the health and safety of the public shall be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes and standards pertaining to design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assurance programs, test procedures, and inspection acceptance criteria to be used shall be identified.

An indication of the applicability of codes, standards, quality assurance programs, test procedures and inspection acceptance criteria used is required. Where such items are not covered by applicable codes and standards, a showing of adequacy is required. (GDC 1)

The containment system structure is of primary importance with respect to its safety function by maintaining its structural integrity in the event of a natural phenomenon event, so it does not fail and cause damage to Class I SSCs.

Quality standards of material selection, design, fabrication, and inspection governing the above features conforms to the applicable provisions of recognized codes and good nuclear practice.

The concrete structure of the reactor containment conforms to the applicable portions of ACI-318-63. Further elaboration on quality standards of the reactor containment is given in Section 3.16.1.3.

3.16.1.1.2 Performance Standards Criterion:

Those systems and components of reactor facilities, which are essential to the prevention or to the mitigation of the consequences of nuclear accidents, which cause undue risk to the health and safety of the public shall be designed, fabricated, and erected to performance standards that enable such systems and

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-91 of 3-126 Revision 0, 2020 components to withstand, without undue risk to the health and safety of the public, the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon such as earthquake, tornado, flooding condition, high wind or heavy ice. The design bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been officially recorded for the site and the surrounding area and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design. (GDC 2)

All components and supporting structures of the reactor containment are designed so that there is no loss of function of such equipment in the event of maximum potential ground acceleration acting in the horizontal and vertical directions simultaneously. The dynamic response of the structure to ground acceleration, based on the site characteristics and on the structural damping, is included in the design analysis. Historically, the reactor containment was defined as a Class I structure for purposes of seismic design. Following the permanent shut down and defueling of the reactor, it was re-classified as a Class III structure (Section 1.7). Its structural members have sufficient capacity to accept, without exceeding specified stress limits, a combination of normal operating loads, functional loads due to a loss-of-coolant accident, and the loadings imposed by the maximum potential earthquake.

3.16.1.1.3 Fire Protection Criterion:

A reactor facility shall be designed to ensure that the probability of events such as fires and explosions and the potential consequences of such events will not result in undue risk to the health and safety of the public. Noncombustible and fire resistant materials shall be used throughout the facility wherever necessary to preclude such risk, particularly in areas containing critical portions of the facility such as containment, control room, and components of engineered safety features. (GDC 3)

Fire protection in all areas of the nuclear electric plant is provided by structure and component design that optimizes the containment of combustible materials and maintains exposed combustible material below the ignition temperature. The station is designed on the basis of limiting the use of combustible materials in construction by using fire-resistant materials to the greatest extent practical.

3.16.1.1.4 Records Requirement Criterion:

The reactor licensee shall be responsible for assuring the maintenance throughout the life of the reactor of records of the design, fabrication, and construction of major components of the plant essential to avoid undue risk to the health and safety of the public. (GDC 5)

Records of the design, fabrication, construction, and testing of the reactor containment are maintained throughout the life of the reactor, as modified in accordance with approved exemptions.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-92 of 3-126 Revision 0, 2020 3.16.1.1.5 Reactor Containment Criterion:

The containment structure shall be designed (a) to sustain, without undue risk to the health and safety of the public, the initial effects of gross equipment failures, such as a large reactor coolant pipe break, without loss of required integrity, and (b) together with other engineered safety features as may be necessary, to retain for as long as the situation requires, the functional capability of the containment to the extent necessary to avoid undue risk to the health and safety of the public.

(GDC 10).

The containment structure and all penetrations are designed to withstand, within design limits, the combined loadings of the historical design-basis accident associated with containment performance and design and maximum potential seismic conditions.

3.16.1.2 Loadings The following loadings are considered in the design of the containment:

1.

Structure dead load.

2.

Live loads.

3.

Equipment loads.

4.

Internal test pressure.

5.

Earthquake.

6.

Wind.

3.16.1.3 Codes and Standards The following is historical information. The design, materials, fabrication, inspection, and proof testing of the containment vessel complies with the applicable parts of the following codes and standards.

Code Title

1.

ASTM A-333, Gr. 1 Specification for Seamless and Welded Steel Pipe for Low Temperature Service

2.

ASTM A-181 Forged or Rolled Steel Pipe Flanges, Forged Fittings, and Valves and Parts for General Service

3.

ASTM A-300, Cl. 1, Firebox Specification for Notch Toughness Requirements for Normalized Steel Plates for Pressure Vessels

4.

ASTM A-201, Gr. B Specification for Carbon Silicon Steel Plates of Intermediate Tensile Ranges for Fusion Welded Boilers and other Pressure Vessels

5.

ASTM A-36 Specification for Structural Steel

6.

ASTM A-131, Gr. C Specification for Structural Steel for Ships

7.

ASTM A-240 Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet, and Strip for Fusion-Welded Unfired Pressure Vessels

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-93 of 3-126 Revision 0, 2020 Code Title

8.

ASTM A-312 Specification for Seamless and Welded Austenitic Stainless Steel Pipe

9.

ASTM A442, Grade 60 Specifications for Pressure Vessel Plates, Carbon Steel, Improved Transition Properties

10.

ASME Boiler and Pressure Nuclear Vessels Vessel Code-Section III Nuclear Vessels

11.

ASME Boiler and Pressure Unfired Pressure Vessels Vessel Code-Section VIII Unfired Pressure Vessels

12.

ASME Boiler and Pressure Welding Qualifications Vessel Code-Section IX Welding Qualifications

13.

ASTM C-33 Standard Specifications for Concrete Aggregates

14.

ASTM C-150 Standard Specifications for Portland Cement

15.

ASTM C-172 Standard Method of Sampling Fresh Concrete

16.

ASTM C-31 Standard Method of Making and Curing Concrete Compression and Flexure Test Specimens in the Field

17.

ASTM C-39 Standard Method of Test for Compressive Strength of Molded Concrete Cylinders

18.

ASTM-C-350 Specifications for Fly Ash for Use as an Admixture in Portland Cement Concrete

19.

ASTM C-94 Specifications for Ready Mixed Concrete

20.

ASTM C-42 Standard Methods of Securing, Preparing, and Testing Specimens from Hardened Concrete for Compressive and Flexural Strengths

21.

ASTM C-494 Specifications for Chemical Admixtures for Concrete

22.

ASTM A-305 Specifications for Minimum Requirements for Deformations of Deformed Steel Bars for Concrete Reinforcement

23.

ASTM A-408 Specifications for Special Large Size Deformed Billet-Steel Bars for Concrete Reinforcement

24.

ASTM A-432 Specification for Deformed Billet Steel Bars for Concrete Reinforcement with 60,000 psi Minimum Yield Strength

25.

Research Council of Riveted and Bolted Structural Joints of the Engineering Foundation Specification for Structural Joints Using ASTM A-325 Bolts

26.

ACI-613 Recommended Practice for Selecting Proportions for Concrete

27.

ACI-306 Recommended Practice for Winter Concreting

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-94 of 3-126 Revision 0, 2020 Code Title

28.

ACI-318, Part IV-B Structural Analysis and Proportioning of Members-Ultimate Strength Design

29.

ACI-318 Building Code Requirements for Reinforced Concrete

30.

ACI-505 Specification for the Design and Construction of Reinforced Concrete Chimneys

31.

ACI-315 Manual of Standard Practice for Detailing Reinforced Concrete Structures

32.

ASA N6.2 Safety Standards for the Design, Fabrication and Maintenance of Steel Containment Structures for Stationary Nuclear Power Reactors

33.

ASA A58.1 American Standard Code Requirements for Minimum Design Loads in Buildings and Other Structures

34.

State Building and Construction Code for the State of New York

35.

SSPC-SP-6 Commercial Blast Cleaning 3.16.2 Containment Structure Design 3.16.2.1 General Description The reactor containment structure is a reinforced concrete vertical right cylinder with a flat base and hemispherical dome. A welded steel liner with a minimum thickness of 0.25-in. is attached to the inside face of the concrete shell to ensure a high degree of leaktightness. The design objective of the containment structure is to retain its structural integrity during normal conditions and natural phenomenon events.

The structure, as shown on Drawings 9321-2501, 9321-2502, 9321-2503, 9321-2506, 9321-2507, 9321-2508, and Figure 3.16-1 consists of side walls measuring 148-ft from the liner on the base to the springline of the dome, and has an inside diameter of 135-ft. The side walls for the cylinder and the dome are 4-ft 6-in. and 3-ft 6-in. thick respectively. The inside radius of the dome is equal to the inside radius of the cylinder so that the discontinuity at the springline due to the change in thickness is on the outer surface. The cylindrical part of the liner is substantially round. The difference between the minimum and maximum inside diameters at any selected cross section does not generally exceed 0.25-percent of the nominal diameter at that elevation. Between elevations 43-ft and 95-ft, the maximum diameter of any cross section is 135-ft 2-in., and the minimum diameter is 134-ft 10-in. except at the liner closing the temporary opening in the northwest quadrant where a minimum diameter of 134-ft 8-5/8-in. was measured. This portion of the liner was erected after all exterior concrete work was completed and is within the local buckle allowance of the liner plates. Above elevation 95 ft the tolerance on inside diameter does not exceed 0.50-percent of the nominal diameter of the selected cross section. The liner is erected true and plumb so that the deviation does not exceed 1/500 of the height at the selected cross section (allowing for 2-in. local buckling of the liner plates).

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-95 of 3-126 Revision 0, 2020 Particular care is taken in matching edges of cylindrical and hemispherical sections to ensure that all joints are properly aligned. Maximum permissible offset of completed joints is 25 percent of nominal plate thickness. Plates buckled beyond acceptable limits are cut out and replaced with new plates.

The flat concrete base mat is 9-ft thick with the bottom liner plate located on top of this mat. The bottom liner plate is covered with 3-ft of concrete, the top of which forms the floor of the containment.

Where uplift from pressure occurs at the outer areas of the mat, the 9-ft thick mat has sufficient flexural capacity to resist the uplift.

No hydraulic uplift exists since the bottom elevation of the mat is considerably higher than that of the high water level.

The large mass of the containment including interior concrete and equipment makes the structure inherently stable from overturning due to seismic motion.

In addition, keying action from the reactor pit and sumps, plus friction between the concrete and rock, prevents a sliding of the structure from horizontal ground motion.

The basic structural elements considered in the design of the containment structure are the base slab, side walls, and dome acting as one structure under all possible loading conditions. The liner is anchored to the concrete shell by means of stud anchors. The lower portions of the cylindrical liner are insulated to avoid thermal deformation of the liner under accident conditions.

The containment structure is inherently safe with regard to common hazards such as fire, flood, and electrical storm. The thick concrete walls are invulnerable to fire and only an insignificant amount of combustible material, such as lubricating oil in pump and motor bearings, is present in the containment.

Internal structures consisted of equipment supports, shielding, reactor cavity and canal for fuel transfer, and miscellaneous concrete and steel for floors and stairs. All internal structures are supported on the mat with the exception of equipment supports secured to the intermediate floors.

A 3-ft thick concrete ring wall serving as a missile and partial radiation shield surrounds the reactor coolant system components and supports the polar-type reactor containment crane. A 2-ft thick reinforced concrete floor covers the reactor coolant system with removable gratings in the floor provided for crane access to the reactor coolant pumps. The four steam generators, pressurizer, and various piping penetrate the floor. Spiral stairs provide access to the areas below the floor.

The refueling canal connects the reactor cavity with the fuel transport tube to the spent fuel pool.

The floor and walls of the canal are concrete, with wall and shielding water providing the equivalent of 6-ft of concrete.

The refueling canal floor is 5-ft thick. The concrete walls and floor are lined with 0.25-in. thick stainless steel plate. The linings provide a leakproof membrane that is resistant to abrasion and damage during fuel handling operation.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-96 of 3-126 Revision 0, 2020 Waterproofing is provided in the areas of the containment in contact with backfill to prevent ground-water seepage. This consists of a coat of bitumastic No. 50, a 0.625-in.-thick layer of hardboard insulation, and a second coat of bitumastic No. 50. Fill for innermost 5-ft from containment walls is crushed rock of maximum size of 6-in. and minimum amount of fines. All fill is free of vegetable matter.

3.16.2.2 Design Load Criteria The following loads are considered to act upon the containment structure creating stresses within the component parts.

1.

Dead load consists of the weight of the concrete wall, dome, liner, insulation, base slab, and the internal concrete. Weights used for dead load calculations are as follows:

a. Concrete 150 lb/ft3

b. Reinforcing steel 490 lb/ft3 using nominal cross-sectional areas of reinforcing as defined in ASTM for bar sizes.

c. Steel lining 490 lb/ft3 using nominal cross-sectional area.

d. Insulation 6 lb/ft3 including stainless steel jacket.

2.

Live load consists of snow and construction loads on the dome and major components of equipment in the containment. Snow and ice loads are assumed to be applied uniformly to the top surface of the dome at an estimated value of 20 lb/ft2 of horizontal projection of the dome. This loading represents approximately 2-ft of snow, which is considered to be a conservative amount since the slope of the dome will tend to cause much of the snow that falls on it to slide off. A construction live load of 50 lb/ft2 has been used on the dome, but will not be considered to act concurrently with the snow load. Equipment loads are considered as specified on the drawings supplied by the manufacturers of the various pieces of equipment.

Design live loads inside the containment building are as follows:

a. Elevation 68-ft-0-in.

10-ft strip adjacent to crane wall = 600 psf Remaining strip = 100 psf

b. Elevation 95-ft-0-in.

Concrete slab = 500 psf Grating areas = 100 psf

3.

For the free volume of 2,610,000-ft3 within the containment, the design pressure is 47 psig.

4.

The ground acceleration for the design earthquake has been determined to be 0.1g applied horizontally and 0.05g applied vertically. These values have been resolved as conservative numbers based upon recommendations from Dr. Lynch, Director of Seismic Observatory, Fordham University.

A dynamic analysis is used to arrive at equivalent design loads. Additionally, a hypothetical ground acceleration of 0.15 g horizontal and 0.10 g vertical is used to

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-97 of 3-126 Revision 0, 2020 analyze for the no-loss-of-function. This is discussed in Section 3.16.3.11, Seismic Design.

Due to symmetry of the containment structure, torsional loads generated by an earthquake are insignificant and have not been considered.

Tornado loads have not been considered in the design of the Unit 2 containment; however, the seismic bars provide a more than adequate mechanism to withstand the torsional effect if it were to occur. An evaluation of the effect of tornado loads on the containment structure is presented in Appendix B of the Containment Design Report.

5.

The American Standards Association "American Standard Code Requirements for Minimum Design Loads in Buildings and Other Structures" (A58.1-1955) designates the site as being in a 25 psf zone for wind loads. In this code, for height zones between 100 and 499-ft, the recommended wind pressure on a flat surface is 40 psf. Correcting for the shape of the containment by using a shape factor of 0.60, the recommended pressure becomes 24 psf. The state building and construction code for the State of New York stipulates a wind pressure up to 30 psf on a flat surface for heights up to 300 feet. For design, a 30 psf basic wind load has been used from ground level up.

6.

Internal pressure was applied to test the structural integrity of the containment shell up to 115-percent of the design pressure. For this structure, the test pressure is 54 psig. The containment is also structurally designed to withstand an external pressure 2.5 psig higher than the internal pressure.

3.16.2.3 Material Specifications Basically, five materials are used for the construction of the containment structure.

These are:

1.

Concrete.

2.

Reinforcing steel.

3.

Plate steel liner.

4.

Insulation.

5.

Protective Coating.

Basic specifications for these materials are as follows:

1.

Concrete is a dense, durable mixture of sound coarse aggregate, fine aggregate, cement, and water. Cement conforms to ASTM, Specification C-150-65 "Standard Specification for Portland Cement," Type I (Normal), or Type II (moderate heat of hydration) requirements. Whenever high early strength is required, Type III Cement is used. Water is free from any injurious amounts of acid, alkali, salts, oil, sediment, or organic matter. The concrete has a minimum density of 150 lb/ft3.

The 28-day standard compressive strength of the concrete is 3000 psi. Adequate means of control are used in the manufacture of the concrete. To ensure the values of compressive strength are attained as a minimum, concrete samples are tested in accordance with the following ASTM Standards:

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-98 of 3-126 Revision 0, 2020 ASTM C-172 - Standard Method of Sampling Fresh Concrete ASTM C Standard Method of Making and Curing Concrete Compression and Flexure Test Specimens in Field ASTM C Standard Method of Test for Compressive Strength of Molded Concrete Cylinders All making and testing of concrete samples have been performed by Vacca Testing Laboratory and Research Company, Inc.

At certain specifically evaluated locations, non-structural surface type cracks and delaminations in the containment concrete have been repaired by injection of engineering approved epoxy grout. Although non-structural in nature, these repairs were performed in accordance with the requirements of IWL-4210 of the 1992 ASME Boiler and Pressure Vessel Code,Section XI, as applicable.

2.

Reinforcing steel for the dome, cylindrical walls and base mat is high-strength, deformed billet steel bars conforming to ASTM Designation A432-65 "Specification for Deformed Billet Steel Bars for Concrete Reinforcement with 60,000 psi Minimum Yield Strength." This steel has a minimum yield strength of 60,000 psi, a minimum tensile strength of 90,000 psi, and a minimum elongation of 7-percent in an 8-in. specimen. Reinforcing bars No. 11 and smaller in diameter are lapped spliced in the mat for flexural loadings and spliced by the Cadweld process in the walls and dome for tension loading. Bars No. 14S and 18S are spliced by the Cadweld process only. A certification of physical properties and chemical content of each heat of reinforcing steel delivered to the job site has been issued from the steel supplier. The splices used to join reinforcing bars have been tested to ensure that they will develop at least 125-percent of the minimum yield point stress of the bar. The test program required cutting out, at random, approximately 3-percent, completed splices and testing to determine their breaking strength.

3.

The plate steel liner is carbon steel conforming to ASTM Designation A442-65 "Standard Specification for Carbon Steel Plates with Improved Transition Properties," Grade 60. This steel has a minimum yield strength of 32,000 psi and a minimum tensile strength of 60,000 psi with an elongation of 22-percent in an 8-in. gauge length at failure.

The liner is 0.25-in. thick at the bottom, 0.50-in. thick in the first three courses, except 0.75-in. thick at penetrations, a minimum of 0.34-in. in the general area at elevation 46-ft. due to past corrosion, and 0.375-in. thick for remaining portion of the cylindrical walls and 0.50-in. thick in the dome. The 0.34-in. minimum thickness affects the calculated stress levels presented in the Containment Design Report and the Containment Liner Stress Analysis Report. However, evaluation of the reduced minimum thickness has concluded that no design criteria are exceeded. The liner material has been tested to ensure an NDTT more than 30°F lower than the minimum operating temperature of the liner material.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-99 of 3-126 Revision 0, 2020 Impact testing has been done in accordance with Section N331 of Section III of the ASME Boiler and Pressure Vessel Code. A 100-percent visual inspection of liner anchors was made prior to pouring concrete.

4.

The material used for the original insulation of the liner plate was polyvinylchloride with stainless steel jacket. The carbon steel liner with an inorganic zinc protective coating makes contact with the polyvinylchloride insulation, the stainless steel, and the sealant. However, these materials do not react with each other.

Manufacturer's tests on the polyvinylchloride insulation indicated that the insulation was capable of withstanding periodic compression at 60 psig at temperatures from 40°F to 120°F and a single compression under accident conditions without any detriment or change to the insulation properties. The manufacturer's analog transient analysis indicated only a 5°F rise in liner temperature 1000 sec after an exposure to 310°F for the entire duration of the analysis. This provides a factor of safety of approximately 15 on specified tolerable temperature rise in the liner. A factor of safety of 2 is provided on specified insulation performance versus tolerable temperature rise in liner.

5.

One 3 mil shop coat of Carbozinc No. 11 primer and one 4 mil minimum finish coat of Phenoline No. 305 as manufactured by the Carboline Company have been applied to the liner, as well as essentially all painted surfaces in containment, in accordance with the manufacturer's recommendations.

The effect of the historical post-accident environment on protective coatings was conservatively evaluated for Indian Point Unit 2. The coatings showed no deterioration after a number of cycles. A more thorough discussion on the qualifications of the protective coatings applied during construction is presented in WCAP-7198-L.1 In addition, various areas inside containment have been repaired and recoated with other DBA qualified coatings approved for use at Indian Point 2. Protective coatings used inside the containment are procured, applied, and maintained in compliance with Regulatory Guide 1.54 (June 1973), Quality Assurance Requirements for Protective Coatings Applied to Water Cooled Nuclear Power Plants. New quality requirements will be developed based on its provisions, but specific requirements, such as documented site meetings, field demonstrations, substrate priming, applicator reporting, inspection reporting and report forms will be considered on a job-by-job basis.

Quality of both materials and construction of the containment structure was ensured by a continuous program of quality control and inspection by Con Edison, and/or its field representatives, and Westinghouse Atomic Power Division, and United Engineers and Constructors Inc., as described in Section 3.16.2.5.

3.16.2.4 Design Stress Criteria This analysis, including the consideration of accident pressure loads, is retained, because it is conservative with respect to the containment structures function in the permanently shut down and defueled condition.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-100 of 3-126 Revision 0, 2020 The design is based upon limiting load factors that are used as the ratio by which loads will be multiplied for design purposes to ensure that the loading deformation behavior of the structure is one of elastic, tolerable strain behavior. The load factor approach is being used in this design as a means of making a rational evaluation of the isolated factors, which must be considered in ensuring an adequate safety margin for the structure. This approach permits the designer to place the greatest conservatism on those loads most subject to variation and which most directly control the overall safety of the structure. In the case of the containment structure, therefore, this approach places minimum emphasis on the fixed gravity loads and maximum emphasis on accident and earthquake or wind loads. The loads utilized to determine the required limiting capacity of any structural element on the containment structure are computed as follows:

1.

C = 1.0D +/- 0.05D + 1.5 P + 1.0 (T + TL)

2.

C = 1.0D +/- 0.05D + 1.25 P + 1.0 (T' + TL') + 1.25E

3.

C = 1.0D +/- 0.05D + 1.0P + 1.0 (T" + TL") + 1.0E' Symbols used in these formulae are defined as follows:

C

=

Required load capacity of section.

D

=

Dead load of structure and equipment loads.

P

=

Historical Accident pressure load as shown on historical pressure-temperature transient curves.

T

=

Load due to maximum temperature gradient through the concrete shell and mat based upon temperature associated with 1.5 times historical accident pressure.

TL

=

Load exerted by the liner based upon temperatures associated with 1.5 times historical accident pressure.

T'

=

Load due to maximum temperature gradient through the concrete shell and mat based upon temperatures associated with 1.25 times historical accident pressure.

TL'

=

Load exerted by the liner based upon temperatures associated with 1.25 times historical accident pressure.

E

=

Load resulting from either design earthquake or wind, whichever is greater.

T"

=

Load due to maximum temperature gradient through the concrete shell and mat based upon temperatures associated with the historical accident pressure.

TL"

=

Load exerted by the liner based upon temperatures associated with the historical accident pressure.

E'

=

Load resulting from assumed hypothetical earthquake.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-101 of 3-126 Revision 0, 2020 A chart for allowable versus actual stresses has been included in the Containment Design Report.

Load condition (1) indicates that the containment will have the capacity to withstand loadings at least 50-percent greater than those calculated for the historical postulated loss-of-coolant accident alone. Results of analysis using load condition (1) are shown in Figure 3.16-2.

Load condition (2) indicates that the containment will have the capacity to withstand loadings at least 25-percent greater than those calculated for the historical postulated loss-of-coolant accident with a coincident design earthquake. Results of analysis using load condition (2) are shown in Figure 3.16-3.

Load condition (3) indicates that the containment will have the capacity to withstand loads at least equal to those calculated for the historical postulated loss-of-coolant accident with a coincident hypothetical earthquake defined in Section 3.16.2.2. Results of analysis using load condition (3) are shown in Figure 3.16-4.

The mat has been analyzed using load conditions (1), (2) and (3) as shown in Figures 3.16-5 through 3.16-7 and also for loads occurring only at operating and test pressure conditions. For loads, see Table 3.16-1, Flooded Weights-Containment Building.

The loads resulting from wind on any portion of the structure do not exceed those resulting from earthquake.

The capacity of all structural components, with the minor exceptions of outer rebar at large containment openings addressed in Section 3.4.4 of the Containment Design Report, exceeds or is equal to the capacity required by the most severe loading combination. The loads resulting from the use of these equations will hereafter be termed "factored loads.

The load factors used in these equations are based upon the load factor concept employed in Part IV-B, "Structural Analysis and Proportioning of Members Ultimate Strength Design" of ACI 318-63. Because of the refinement of the analysis and the restrictions on construction procedure, the load factors in the design primarily provide for a safety margin on the load assumptions.

The design includes the consideration of both primary and secondary stresses. The design limit for tension member (i.e., the capacity required for the design load) is based upon the yield stress of the reinforcing steel.

The theoretical load carrying capacity of steel reinforced concrete cross-sections are reduced by a capacity reduction factor "", which provides for the possibility that small adverse variations in material strengths, workmanship, dimensions, and control, while individually within required tolerances and the limits of good practice, occasionally may combine to result in under-capacity.

For tension members, the factor "" has been established as 0.95. The factor "" is 0.90 for flexure and 0.85 for diagonal tension, bond, and anchorage.

For principle compression and tension, the liner stresses are maintained below 0.95 specified minimum yield at normal operating temperature (i.e., =0.95). For shear, the liner stresses are maintained below 0.6 yield.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-102 of 3-126 Revision 0, 2020 The liner is designed to assure that no strains greater than the strain at the guaranteed yield point will occur at the factored loads. In regions of local stress concentrations or stresses due to localized secondary load effects, the liner is permitted to yield but the maximum liner strain is limited to 0.5-percent. Sufficient anchorage is provided to ensure elastic stability of the liner. The basic design concept for the liner stud anchorage is the ductility of the anchorage that assures stud failure due to shear, tension or bending stress without the stud connection causing failure or tear of the liner plate. References 3.16-2 and 3.16-3 provide information on design of stud connection. The studs in the 0.50-in. plate are installed on 24-in. horizontal and 28-in. vertical grid and in the 0.375-in. plate on a 24-in. horizontal and 14-in. vertical grid. Studs are centered between vertical bars. In the dome, 5-ft by 5-ft panels are anchored in the center by studs and by Tbars at the edges. The 0.50-in. diameter bent welding studs are 9-in. long minimum and 9.50-in. long maximum with a 2-in. 90-degree hook at the end. An arc stud welding process was used on all bent welding studs. The arc stud welding process produces a circular weld around the 0.50-in. diameter stud with a diameter (outside to outside of weld) equal to 0.678-in. and a height equal to 0.157-in. The design considers the possibility of daily stress reversals due to ambient temperature changes for the life of the plant, and fatigue limit of the studs exceeds the design requirements. However, to accommodate possible fatigue failure in the plate-to-stud weldment, the depth of penetration to the liner plate is controlled to avoid impairment of liner integrity.

The boundary conditions in the cylinder are determined by assuming a buckling model (shown in Figures 3.16-8 through 3.16-10) in which the studs form the low points and the center of the panels form the high points of a series of peaks and valleys thus forming a set of panels whose edges represent points of inflection. The analytical procedure used is a simply supported plate under biaxial compression. A Mohr's circle analysis is used to find the normal and shear stresses on this simply-supported plate. The critical buckling stress is derived considering a plate whose length is equal to one-half of the diagonal distance between studs. This critical buckling load is 38.1 ksi for the 0.375-in. liner and 38.4 ksi for the 0.50-in. liner, which is higher than the yield strength of the liner, 32 ksi; therefore, the liner plate will begin to yield before the critical buckling stress is reached, and buckling failure does not control the design. Since shear reduces the stability of a plate subjected to compressive stresses, critical shear is considered and it was found that critical buckling is controlled by normal stresses rather than shear stresses. This is determined by considering the magnitude of both the normal and the shear stresses on the panel.

The magnitude of the shear is so low that it shows no effect on the previously stated critical buckling stresses.

In the dome the liner will be considered clamped at the stiffeners forming a 5-ft by 5-ft grid panel pattern. The center of each panel is fixed by a stud. Assuming points of inflection at the one-quarter point a distance of 1-ft 3-in. occurs between points of simple support. The critical buckling load is 58.1 ksi, which is also higher than the yield strength of the liner.

At maximum strain in the liner, the studs will not fail. This maximum strain due to an unbalanced load would occur in a panel adjacent to a buckled panel. Since this adjacent stud will not fail, no zipper effect will occur and massive buckling of the liner and mass failure of anchors is not credible.

The anchorages can fail by failure of the studs in shear or tension, by studs pulling out from the concrete, or by studs separating from the liner plate. The most likely mode of failure is by tensile failure of the stud. The anchors are designed so that failure occurs in the anchor rather than the plate, thereby ensuring that the leaktight integrity of the containment liner will be maintained.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-103 of 3-126 Revision 0, 2020 If failure should develop, it would be a random stud failure due to poor workmanship during stud attachment. This failure would not impair the liner integrity nor would it cause progressive failure.

The anchor must resist tensile and shearing loads. Tests have indicated that the lateral load needed to prevent column buckling is 1-percent of the axial yield load. Conservatively doubling this value to account for uncertain field conditions, a value of 2-percent is used.4 The total load per plate would be 24-in. x 0.50-in. x 32,000 psi = 384,000 lb. Therefore, the tensile load per anchor is 384,000 lb x 0.02 = 7680 lb, which yields a stress of 7680/0.2 = 38,400 psi.

This compares with a yield value of 50,000 psi and a tensile strength of 60,000 psi in the studs.

This does not consider the internal pressure, which provides further stability against buckling.

The shear load on the anchor is due to the strain in the liner. Assuming the liner approaches its yield strain of 0.1-percent, the anchor deflection would be 28-in. x.001 =.028-in. Tests on the stud anchor have shown a maximum deflection of about 0.1-in. can be tolerated before failure of the stud.

3.16.2.5 Quality Control This section is historical, and is retained for information only.

To ensure a high degree of confidence in plant design, construction, workmanship, materials, and performance, a quality control program has been in effect for this project in which the following principal organizations have their respective responsibilities:

1.

Consolidated Edison Company of New York, Inc. as initial owner and operator of the plant.

2.

Westinghouse Electric Corporation as the turnkey plant contractor and supplier of major equipment.

3.

United Engineers and Constructors Inc. as architect-engineer, construction manager, and constructor.

The function and responsibility in the quality control program of each of the above organizations is as follows:

3.16.2.5.1 Consolidated Edison Company of New York, Inc. (Con Edison) - Initial Licensee A qualified field representative was assigned to the field during the construction period. His responsibilities included continuous inspection of the construction of the containment building to ensure that all materials used and work performed was strictly in accordance with the plans and specifications. The Con Edison representative, through instructions received from the home office, had the power to stop the construction until any discrepancies were corrected and the work once more was in compliance with the specifications and plans.

The Con Edison representative was in constant communication and consultation with the construction superintendent in matters regarding quality control. In addition, personnel from U.S.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-104 of 3-126 Revision 0, 2020 Testing Laboratories were assigned to this project to monitor the inspection of the construction and obtain samples of the materials for testing.

3.16.2.5.2 Westinghouse Electric Corporation For the assurance of plant integrity and quality, Westinghouse performed the following functions regarding the containment building:

1.

Reviewed and approved the containment design criteria, material specifications and detail design concepts before they were released for construction. This work was done by qualified structural engineers at the company's home office.

2.

Reviewed the construction and inspection methods employed by United Engineers and Constructors Inc.

Westinghouse Pressurized Water Reactor Division, Nuclear Power Services Group had a field quality assurance representative in residence during the construction period. His function was the same as the Con Edison representative mentioned above. He reported discrepancies to the Westinghouse Construction and Services resident engineer who had the authority to stop the work until the discrepancy was resolved.

In addition to this, he audited the construction files, and verified that records were complete, accurate, and adequate for quality assurance.

Nuclear Power Service Headquarters quality assurance engineers also made trips to the site to audit, monitor, and review the project with regard to site quality assurance. Construction practices were observed for conformance to codes, specifications, and approved procedures.

3.16.2.5.3 United Engineers and Constructors Inc.

The responsibilities of United Engineers and Constructors Inc. in the quality control of the containment building were as follows:

1.

They inspected all materials delivered to the job site, and examined the suppliers' certified test reports of physical and chemical properties for those components furnished by them.

2.

They inspected fabrication of major components of the containment structure in the shop. Trip reports are available at the site.

3.

They maintained an adequate force of qualified supervisory personnel at all times.

4.

They supervised and were fully responsible for the quality of work performed by their subcontractors and for the craft labor employed and supervised by them.

5.

They maintained as part of their field engineering force, qualified personnel who performed a thorough inspection of each construction operation.

No changes in design or specifications were allowed without the approval of the engineer in charge of design.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-105 of 3-126 Revision 0, 2020 3.16.3 Containment Stress Analysis 3.16.3.1 General The structural design of the containment meets the requirements established by 1961 edition of "The State Building and Construction Code for the State of New York" so far as these provisions are applicable. All concrete structures have been designed, detailed, and constructed in accordance with the provisions of "Building Code Requirements for Reinforced Concrete" (ACI 318-63) so far as these provisions are applicable.

3.16.3.2 Method of Analysis Basically, three separate structural components have been analyzed, each in equilibrium with loads applied to it and with constraints occurring at the juncture of the structures. The three components are:

1.

The 135-ft ID hemispherical dome.

2.

The 135-ft ID cylinder.

3.

The base slab.

Mathematically, the dome and cylinder have been treated as thin-walled shell structures, which results in a membrane analysis. Since the thickness of the dome and cylinder is small in comparison with the radius of curvature (1/20 and 1/15) and there are no discontinuities such as sharp bends in the meridonal curves, the stresses due to pressure and wind or earthquake are calculated by assuming that they are uniformly distributed across the thickness.

Since the concrete is not assumed to resist any tensile or shear forces, radial shear reinforcing has been introduced in the lower portion of the wall in the form of hooked diagonal stirrups and diagonally bent bars as shown in Figure 3.16-1. Diagonal shear reinforcing, at 45° and 135° to the circumferential direction, are placed in the center of the cylinder wall for the full height of the wall and a distance above the springline into the dome to resist earthquake shears. The diagonal bars are discontinued in the upper area of the dome (beyond about 30 degrees above the springline), where the seismic shears are small and are carried by the dome reinforcing steel lying in the plane of principal tension.

The base slab has been treated as a flat circular plate supported on a rigid nonyielding foundation.

The limiting cases in the design of the wall for discontinuity moments and shears were considered.

One case considered an uncracked wall and the other considered a cracked wall with the steel acting as a spring constant. The value of µc varied from zero in the cracked case to.14 in the uncracked case. In the uncracked case, variations in Ec will have no effect on the answer since Ec appears in both the numerator and the denominator of the stiffness formulation. For the above variation in Ec and µc, the values of discontinuity moment and shear vary by 14-percent and 7-percent respectively at the base. These are the maximum deviations of the wall forces since the wall will actually vary from uncracked to cracked with an increase in containment height rather than be cracked or uncracked for the total height.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-106 of 3-126 Revision 0, 2020 In the area of thermal stress, the entire wall section will be cracked and no variation in Ec or µc need be considered. The liner stresses depend on the strains of the reinforcing steel and are not related to the concrete properties.

Shrinkage and creep effects will be relieved by cracking during the pressure test and will not be included in accident design considerations.

The finite element computer program has the capabilities of taking into account variations in µc and Ec and axisymmetric loads. However, it is not necessary to take into account the variations in µc and Ec for the reasons stated above.

The computer program used to study the general behavior of the structure and to generate boundary conditions was the axisymmetric shell structure program. This computer program, developed by Franklin Institute Research Laboratories, is designed to handle arbitrarily shaped shells of revolution subjected to axisymmetric as well as nonaxisymmetric loadings. The method of analysis consists of subdividing the shell into elements having continuous meridians with continuous first and second derivatives so that the first and second fundamental forms of the resulting shell elements are continuous throughout the element. By expanding the dependent variables in Fourier series in the circumferential direction, and assigning unspecified functions for the meridional variation, the independent variables are separated and a system of ordinary differential equations results for the dependent variables in terms of the meridional independent variable. Particular and complementary solutions of these ordinary differential equations are then found for each of the elements and each of the circumferential harmonics individually. The matching of the elements is achieved by writing the required boundary conditions.

The idealized section used with the axisymmetric shell structure program consists of five layers whose moment of inertia is equal to that of the actual section. The wall section is considered as cracked with the reinforcing carrying all loads.

A finite element program, with the capability to incorporate thermal loads, was used to analyze the containment shell considering the effect of the equipment hatch opening.

The shell was idealized into 10 layers with alternate layers of steel and concrete. Section 3.16.3.10 provides more information on the finite element analysis.

The computer program can handle the loads in the form of either surface traction or edge loads or both.

Analysis of the liner is presented in the Containment Liner Stress Analysis Report. The report also contains a description of analytical procedures arriving at forces, shears, and moments in the structural shell.

3.16.3.3 Dome Analysis The analysis of the hemispherical dome has been performed by the super-position of membrane forces resulting from gravity, historical accident pressure, and historical accident thermal loads.

In addition, earthquake or wind loading create both direct and shear stresses in the dome, and the historical operating temperature of the liner creates tension and compression. All of the combined direct stresses are developed in the reinforcing steel encased in the concrete. In the

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-107 of 3-126 Revision 0, 2020 upper area of the dome (about 30 degrees above the springline), where the seismic shears are small, seismic shears are carried by dome reinforcing steel lying in the plane of the principal tension. The dome reinforcing is spliced to the vertical steel in the cylindrical concrete wall, so that a continuity between the dome and the cylinder is realized. See Figure 3.16-11 for a section of wall, dome and for reinforcing in the dome.

3.16.3.4 Cylinder Analysis The analysis of the cylinder is by superposition of membrane forces resulting from gravity, historical pressure and thermal loads, overturning due to earthquake or wind and shears due to earthquake or wind. The concrete has been reinforced circumferentially using steel hoops and vertically by straight bars. Diagonal bars have been placed to resist the horizontal and vertical shears due to earthquake or wind. The required capacity of the diagonal bars has been designed so that the horizontal component per foot of the diagonals is equal to the maximum value of shear flow. A check was made to ensure that no net compressive force results in the diagonal bars because of the combination of seismic shear load and internal pressure load. Although, in the cylinder, the liner has some capacity available to resist the seismic shears, no credit is taken for this capacity.

For all of the cylinder and the lower areas of the dome, the diagonal reinforcing has been designed to accommodate all seismic shears. No credit has been taken for the dowel action of the vertical and horizontal bars in resisting seismic shear.

Only in the upper area of the dome (beyond about 30 degrees above the spring line) where the seismic shears are small is the liner counted on to resist shear. For all of the cylinder and the lower areas of the dome, the diagonal reinforcing has been designed to accommodate all seismic shears. No credit has been taken for the dowel action of the vertical and horizontal bars in resisting seismic shear.

3.16.3.5 Base Mat Analysis The base slab was treated as a flat circular plate supported on a rigid non-yielding foundation.

For loads applied uniformly around the slab, the analysis considers a 1-ft wide beam fixed at a point where the vertical shear is equal to zero. This is the point where the downward pressure on the mat and the dead weight overcome the uplift at the containment wall base mat juncture from pressure and earthquake loadings. Radial and circumferential reinforcing is provided at the top and bottom of the mat to resist moments in the areas where uplift occurs. Temperature steel was added in other areas to meet requirements of the (ACI-318) Code. Diagonal tension reinforcement was added to meet requirements of ACI-318 Code. See Figure 3.16-14 for base slab reinforcing detail.

Moments and shears were calculated by writing equations for moment and shear in terms of X using the containment wall-base slab juncture as the origin with X increasing toward the center of the containment building. The point along the circumference of the containment wall chosen as the end of the beam is a point where the maximum tension from the earthquake will exist. Since the containment structure is considered a beam in all earthquake analyses, the maximum uplift for which the mat is designed will occur at only one point on the circumference and will represent the worst possible uplift on the mat.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-108 of 3-126 Revision 0, 2020 All stresses were calculated using Part IV-B Structural Analysis and Proportioning of Members -

Ultimate Strength Design of the Building Codes Requirements for Reinforced Concrete (ACI318-63). No rebar stresses exceed 0.90 fy.

A gradient with an operating temperature of 120°F inside the containment (historical condition) and a 50°F temperature at the mat-rock interface was considered and stresses were negligible.

Ambient accident temperatures have no appreciable effect on the base slab. The maximum operating temperature of the containment is 130°F. The effect of elevated operating temperature on the structural elements was evaluated in 1987 and was found acceptable.

It is not possible to show that the design on nonyielding rock is more conservative than assuming the rock to be elastic. However, due to the installation of temperature reinforcing, the design is conservative. Reinforcing and concrete stresses are very low when considering the rock to be elastic.

To substantiate the above statement, the following studies were performed:

1.

The foundation modulus were determined using the expression:15 kZ =

4 1

Gro

µ where:

kZ = The vertical spring constant of a circular base supported on an elastic foundation G

E 2

=

+

(

)

1

µ rO = Radius of Foundation

µ = Poisson's Ratio To obtain the foundation modulus, kZ is divided by the area of the circular base to yield

(

)

ko kz A

G o

=

x

4 1

µ r

Substituting for G o

k =

E o

2 2

µ r

(1-

)

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-109 of 3-126 Revision 0, 2020

2.

The first case examined was that of a rectangular strip loaded with 1.5 times historical design accident pressure plus dead load using conservative properties for the Dolomitic limestone:7,14 E = 6.0 x 106 psi

µ = 0 Applying these values ko = 4370 lbs/in.3 The "characteristic" is defined as:6

=

1/4 k

4 EI

Where:

E is the modulus of elasticity of the structural base (concrete),

I is the moment of inertia of the structural base, k = ko b, (b = width of base) using base properties

= 7.56 x 10-3-in.-1 Where > beams may be considered as infinite in length.6 Taking the length of beam as being the base diameter

= 13.1 >

The beam was then analyzed as a beam of unlimited length loaded over an area equal to the base diameter with an 80 psi uniform load.

The solution to this problem gives

(

)

(

)

(

)

b a

b a

b a

C C

4 q

Qc B

B 2

4 q

c M

D D

2 k

2 q

c y

=

+

=

=

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-110 of 3-126 Revision 0, 2020 where yC is deflection of point being considered MC is the moment at point being considered QC is shear at point being considered q

is the uniform load a

is the distance from point under consideration to end of load b

is distance from point under consideration to other end of load.

Bx = e-x sin x Cx = e-x(cos x - sin x)

D = e-x cos x Maximum moment occurs at mid-point of load and is equal to 352-in.-lbs/in.

For the area of the mat where there is only temperature reinforcing, the maximum moment would cause a stress of 30 psi in the reinforcing.

The maximum shear would occur at the ends and is equal to 2.64 kips/in. This shear would cause a shear stress in an unreinforced concrete section of 26.4 psi.

3.

A second case examined was for the foundation material being less rigid than the concrete base. The model was the same for the first case:

Assumed Erock = 2.6 x 106 psi

µ = 0 For this case, the following were determined:

ko = 1890 lb/in.3

= 6.2 x 10-3-in.-1 Mmax = 3.66-in.-kips/in.

Qmax = 3.23 kips/in.

Srebar = 312 psi conc = 32.3 psi

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-111 of 3-126 Revision 0, 2020 As a final study, the maximum deflection as calculated in the first case was imposed as a settlement of the base mat for the outer portion and a section of the mat was analyzed for this settlement. A 30-ft section was used with fixity at the reactor pit, the remainder cantilevered from the pit.

The resulting moment and shear are as follows:

M = 142-in.-kips/in.

q = 396 lbs resulting in a rebar stress of 12.2 ksi and a shear stress of 4.0 psi.

From the above, it can be seen that the assumption that a foundation on rock is a rigid unyielding foundation is a valid assumption and that temperature reinforcing provides much greater resistance than required to accommodate the effects of any elastic deformation of the subgrade.

3.16.3.6 Analysis of Liner and Reinforcing Steel Approximately 67-percent of the inclined bars, provided to resist radial shear at the base of the containment wall, are secondary vertical bars, which are inside the primary vertical bars on the outside face and inside face of the wall. These bars are continuous and are bent across the wall where reinforcing is required to resist the radial shear. The remaining 33-percent of the required steel area is provided by stirrups that are hooked around the vertical bars by means of a 90-degree hook. Only one-third of the shear reinforcing at a particular elevation is made up of these hooked bars, which occur at four elevations up the wall. See Figure 4.16 of the Containment Design Report.

Since the stud anchors are hooked around reinforcing bars, concrete stresses for pull out loads are negligible. For high shear loads, which would be caused if a stud anchor should fail or be missing, local crushing of the concrete occurs; however, integrity of the anchor and liner plate is not impaired. See Figures 3.16-12 and 3.16-13.

The lowest elevation at which these hooked bars are used is at a point where only 65-percent of the maximum shear at the base is present. The remaining three levels are in regions where the shear is less than 25-percent of maximum base shear. Since the large majority of the shear is resisted by continuous vertical bars, a minimal amount of load must be transmitted to the vertical bars. The hooked stirrups will mechanically transmit the small amount of shear, which they carry.

The main function of the stirrups is to contain the formation of the diagonal tension crack. The mechanical anchorage of the stirrups is sufficient for this purpose.

There are no significant structural loadings, which must be transferred through the liner such as those required for crane brackets or machinery equipment mounts. Miscellaneous spray system piping, instrumentation, conduit, and insulation, which are attached to the liner can be supported by the free-standing liner without inducing significant stresses in the liner or liner anchorage.

Liner stress is imposed on the cylindrical penetration as a circular uniform load acting around the circumference of the penetration. The liner plate is locally thickened at the penetrations to take care of additional stresses.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-112 of 3-126 Revision 0, 2020 The liner can accommodate any shear it will see due to thermal expansion or earthquake.

An investigation was made on the thermal effects, based on the conservative assumptions that the base mat was fully fixed against any thermal movement thereby restraining the liner from movement. The 3-ft fill slab was then subjected to thermal growth. No excessive forces were introduced into the liner and the welds on the test channels were found to be sufficient to prevent any shear failure of the test channels from the liner due to movement of the 3-ft fill mat.

Seismic shear of the interior concrete is resisted by the keying action of the reactor pit and the sump for the recirculation pumps in addition to the weld channels. Considerable resistance is also provided by friction between the liner and the 3-ft slab.

Jet forces cannot remove the liner panels since the forces will be compressing the insulation panels against the liner and exterior wall. The panels are anchored to the liner with 3/16-in.

diameter stainless steel studs. The consequence of an insulation panel being displaced from the liner during or as a consequence of an accident is that the exposed liner would tend to expand.

The unequal strain between the exposed and unexposed portions of the liner causes a shear load on the liner anchor, and a local yielding in compression of the exposed portion of the liner. The liner anchor stud has the capacity to accommodate much greater strains than would be experienced at yield strain in the liner.

3.16.3.7 Containment Interior Structure The interior structure may be separated into five main structural components. They are:

1.

3-ft thick fill slab.

2.

3-ft thick crane wall.

3.

4-ft to 6-ft thick refueling canal.

4.

2-ft thick operating floor slab.

5.

Primary shield wall.

The method of design, stress analysis, critical stresses and locations are as follows:

1.

3-ft thick fill slab - The controlling loads on the 3-ft slab are the reactions are from the primary equipment supports due to various historical postulated pipe breaks.

The slab was designed as a series of radial beams running under the equipment supports and spanning between the reactor support wall and the crane wall.

Stresses in reinforcing were limited to 0.9 fy. Maximum stresses occur immediately below the primary equipment supports.

2.

3-ft thick crane wall - The crane wall was designed for a 7 psi differential pressure occurring immediately after a historical primary pipe break and prior to pressure equalization.

Although the stress levels associated with this pressure differential were sufficiently low to establish that the concrete could resist the pressure loading, sufficient reinforcing was provided to resist all membrane forces without any contribution from the concrete. Stresses were limited to 0.9 fy. The membrane hoop stress was 33 ksi and the axial vertical rebar stress was 14.3 ksi.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-113 of 3-126 Revision 0, 2020 A two dimensional finite element analysis was performed to determine the effect of the jet forces associated with the pipe break on the crane wall.

The jet force associated with a historical pipe break has been based on the static force PA where P is the primary system operating pressure and A is the cross sectional area of the coolant pipe.

The analysis indicated that in local areas (at the application of the force) yielding of the crane wall rebar will occur. The load was assumed to act at the mid-height of the wall, thus causing maximum bending moment. The ability of the wall to support the dead load of the crane was checked, considering the yielded area indicated by the computer analysis as unable to carry load.

A beam 12-ft long and 5-ft deep (the underside of the operating floor to the top of the potential yield portion of the crane wall) was found to provide more than twice the ultimate capacity required. This analysis was very conservative for three reasons:

1.

A jet force load at this location would cause little yielding since it is not located at mid span.

2.

The haunch at the underside of the operating floor was not considered.

3.

The membrane effect of the circular crane wall was not taken into account.

Further stability of the crane wall was demonstrated by determining the ultimate failure load by means of a yield line analysis. This analysis indicated that the structure has the capacity, through strain energy of structural response, to resist the uniform jet force load of 1500 kips or 975 kips with the 7 psi pressure differential without failure.

The containment internal concrete is essentially rigid; (fundamental frequency 18.6 cps) therefore, seismic loads were calculated using the maximum ground acceleration (0.15g).

The crane wall was initially considered as a cantilever beam with a frequency of approximately 13 cps and the base shear was determined by the response spectrum approach. The base shear was distributed to the individual nodes by the formula:

Fx =

h W

V h

W x

x

Where V

=

base shear Wx

=

weight of node under consideration hx

=

distance from base to section under consideration.

3 Wh = Summation of the product of weights and heights of all nodes The moment at the base was determined and the uplift calculated by considering a circular ring of thickness equal to the area of steel per in. This maximum uplift, which occurs at one point at the base of the structure stresses the rebar to 5.2 ksi.

The crane wall was also designed to resist steam and feed water pipe break reactions of 340 kips and 200 kips where supports are connected to the wall. The extra steel provided for pipe break loads is available in the form of steel buttresses to resist pressure, jet force, and seismic loads; however, it was not considered in the analysis.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-114 of 3-126 Revision 0, 2020

3.

4-ft to 6-ft thick refueling canal - The refueling canal was designed for the 7 psi pressure differential. The wall resists the pressure by spanning vertically between the refueling floor and the operating floor. Stresses were limited to 0.9 fy.

An analysis was performed to check the effects of the jet force load the cross section was found to be sufficient to provide stability. A yield line analysis was performed and provided the basis for the above.

The seismic load was determined by the same procedure used for the crane wall. The average load in kips/ft was distributed over the wall and the vertical span was conservatively assumed to carry the entire load. The resulting bending moment was found to be well within the capacity of the wall.

4.

2-ft thick operating floor slab - Because of the many openings in the floor for equipment, the floor was designed as a series of beams. Principal loadings were D.L. + 500 psf live load and 7 psi upward pressure differential + D.L. The first loading (D.L. + 500 psf live load) was designed in accordance with Part IV-B of ACI 318. Stresses for the pressure differential case were limited to 0.9 fy.

The operating floor was investigated. The following is retained for its historical context. There appears to be very little area of the operating floor, which could be reached by the expanding jet of water from a break in the reactor coolant system.

The jet will be greatly dispersed in the distance between the primary coolant piping and the underside of the operating floor. The only area of the floor, which could be struck by a jet spans between areas of the floor heavily reinforced as beams.

The span cross section consists of a T-beam with the 2-ft thick floor acting as the flange and the 7-ft high biological shielding wall as the web. This section can resist the jet force load within 0.9 fy stress limit on the rebar.

5.

Primary Shield Wall - This was designed for two loading conditions due to a split in the reactor. The stress in the reinforcing was limited to the tensile strength of the bars. The first load considered was a 1-ft wide longitudinal split along the length of the reactor. The vessel is assumed accelerated through a 6-in. distance against the support wall by the jet force caused by a 2200 psi pressure acting through a 26.4-ft long by 1-ft wide longitudinal vessel rupture, which results in an impact load of 650 k/ft. This load is imposed by considering an impact factor of two. The maximum rebar stress is 69.5 ksi. The second load considered a pressure buildup of 1000 psi inside the pit due to release of reactor contents. This produces a rebar stress of 86 ksi. The rebar used is ASTM A 432 with specified yield of 60 ksi and ultimate tensile strength of 90 ksi.

To protect the containment base liner, an average of 2-ft of concrete above the containment liner plus a 1-in. liner plate embedded on top of the concrete was provided at the bottom of the containment reactor cavity pit. Below the containment liner plate is 4.5-ft of structural concrete poured on rock.

The following is retained for its historical context. Temperature differential conditions as a result of a LOCA are considered to be of such short duration that the effects were not used in the design of interior structures for stress analysis. A sketch of the design conditions is given in Figure 3.16-15.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-115 of 3-126 Revision 0, 2020 The following is retained for its historical context. During normal operations, the only significant transient temperature gradients occur during startup. The minimum containment internal temperature is limited to 50°F. The maximum operating containment internal temperature is 130°F. Forced movement of containment air is used to limit the concrete temperature surrounding the reactor vessel. This forced air movement of the containment air as well as normal convection and radiation is expected to limit the concrete temperature differentials in the range of 5°F to 10°F. To demonstrate the large margin available in the concrete crane wall and the primary shield wall, a conservative assumption of a 30°F temperature gradient has been evaluated. The evaluation included the gradient effect through the crane wall, the 6-ft thick portion of the primary shield wall below the reactor coolant pipe nozzle, the 5-ft thick portion of the primary shield wall where the nozzles penetrate the wall, and the 4-ft thick wall above the shield wall.

The maximum rebar stress was found to be 4500 psi and occurs in the vertical rebar in the crane wall. The maximum compressive concrete stress was found to be 226 psi and occurs in the hoop direction in the 5-ft portion of the primary shield wall. These stresses are approximately 20-percent of the allowable working stress values and will have no significant effect on the design adequacy of the structures analyzed.

3.16.3.8 Soil Pressure Portions of the containment structure are subjected to the effects of backfill bearing against the containment wall. The effects on the structure are:

1.

Shear and overturning effects due to seismic response and interaction between the soil and structure.

2.

Discontinuity effects caused by the soil restraining deformation of the structure under accident pressures.

To determine the shear and overturning effects two limiting cases were investigated. The first was the case where the structure and soil move out of phase. It was assumed that the structure was subjected to the passive pressure of the soil with the mass of soil, within the shear failure envelope, accelerated against the structure with ground acceleration. In the second case the soil and structure move in phase. For this case it was assumed that the structure was subjected to the active pressure of the soil with the mass of soil, within the shear failure envelope, accelerated with the structure at ground acceleration.

These loads were then treated as external loads on the structure. See Section 3.1.5 of the Containment Design Report for additional information.

To determine the discontinuity effects caused by soil restraint, the structure was analyzed for the passive pressure case. The restraint of the deformation of the structure due to the soil was calculated. Vertical and circumferential bending moments due to this restraint were then

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-116 of 3-126 Revision 0, 2020 determined. Reinforcing bar stresses were calculated and found to be minor. This analysis was then verified by a finite element analysis.

In this analysis, full contribution of the backfill was assumed. During the course of construction it became necessary to build a retaining wall in a substantial area of the backfill, to facilitate construction. The retaining wall extends over 50-ft in plan and includes all of the high fill points assumed in the analysis and design. It can therefore be concluded that the analysis was conservative in that the backfill effects on the completed structure would be only a fraction of that assumed in the original design.

3.16.3.9 Thermal Stresses Temperature effects on the containment structure are due to a thermal gradient through the wall.

The reinforced-concrete wall restrains the liner from growing, resulting in compression in the liner and additional tension in the reinforcing.

Calculation of gradient stresses is based on method of analysis outlined in ACI 505-54, "Specification for the Design and Construction of Reinforced Concrete Chimneys."9 The gradient used is linear with 120°F on the inside and 0°F exterior concrete temperature (-5°F ambient). The maximum operating temperature of the containment is 130°F. The effect of elevated operating temperature (up to 150°F) on the structural elements was evaluated in 1987 and was found to be acceptable.

The ACI method assumes a cracked section in which the concrete carries no tension. The neutral surface (surface at which no thermal stress exists) is determined. Stresses in the liner and reinforcing are calculated based on the assumption that there is no distortion of the wall; i.e.,

variation of strain through the wall thickness is linear.

3.16.3.10 Analysis of Openings The methods followed in design of large openings are described in Section 3.4 of the Containment Design Report (CDR). Included are descriptions of the safety factors used in design. Sample calculations are provided, listing all the criteria and analyzing the effects of all pertinent factors, such as cracking. Also addressed in the CDR is how the existence of biaxial tension in concrete (cracking) has been taken care of in the design, and how the normal and shear stresses due to axial load, two-directional bending, two-directional shear, and torsion are combined. Additionally, the criteria for the design of the thickened part of the wall around the openings is stated.

The methods used to check the design of the thickened stiff part of the shell around large openings and its effect on the shell, torsional stresses, and shrinkage considerations are also addressed in Section 3.4 of the Containment Design Report. This section also describes how deformations and forces are handled around the large openings and in the transition zones into the main portion of the structure.

In the cylindrical section of the containment, where there are large openings for access hatchways and penetrations, the reinforcing bars (hoop, vertical and diagonal) are continued without interruption around the openings.

No bar terminates at any openings as illustrated around the penetration in Figure 3.16-1. Also, additional bars have been furnished locally to take the stresses developed around large openings.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-117 of 3-126 Revision 0, 2020 Concrete is locally thickened at the equipment access hatchway area to accommodate all the reinforcing bars required in this area.

A finite element analysis is performed on the large openings. Representation of the structure is by rectangular elements; each element consists of ten layers of orthotropic, elastic material to represent the reinforcement, concrete and the liner. About 1000 degrees of freedom are considered in the model. This analysis is used as a check on the adequacy of the large openings.

Results appear in the Containment Design Report.

A finite element analysis of the equipment hatch area indicated local liner plastic deformations during the pressure test. For the order of magnitude and location of these stresses, see Section 3.4 of the Containment Design Report. These deformations have no influence on the structure during the pressure test due to the ductility of the studs and liner plate.

The limiting elastic liner deformations during test pressure will be from tensile stresses. During an accident loading they will be from compressive stresses. Therefore, a relationship between the pressure and accident loads cannot be determined directly. However, the test pressure demonstrates the ductile behavior of the liner.

Since the containment is not subject to accident temperatures during the testing, no direct correlation between test and accident conditions can be made in evaluating thermal stresses at large openings.

The liner is stressed beyond the yield point in very local areas adjacent to the transition from the thickened equipment hatch boss to the cylinder wall. The maximum stress is equal to 39.28 ksi for the 1.5P loading condition. The strain corresponding to this stress (0.17-percent) is below the limits (0.5-percent) stated in Section 2.2.4 of the Containment Design Report. The average liner stress in the cylinder for the 1.5P load combination is approximately -15 ksi in the vertical direction and -2.0 ksi in the horizontal direction.

The maximum rebar stress associated with the 1.5P load combination is approximately 66 ksi in the 4'-6" portion of the containment wall cylinder.

For a complete discussion of liner stresses, see the Containment Liner Stress Analysis Report.

For a detailed discussion of liner stresses in the equipment hatch area and further justification of the stresses noted above, see Section 3.4.4 of the Containment Design Report.

All reinforcing is continuous around penetrations. Steps have been taken to ensure that no local crushing of concrete will occur. From Reference 3-16-16, it has been determined that in order to prevent local crushing of the concrete, a minimum bend diameter of 31 times the bar diameter is required when the reinforcing is stressed to yield. The angle of bend in the rebar determines the force that will be transmitted to the concrete in the event the bar tries to straighten out due to tension. For this reason most bars are bent at 10 degrees except at large penetrations including the equipment hatch, personnel lock, main steam and feedwater, and air purge penetrations, where the deviation of the bar from its centerline is too large to permit a 10-degree bend. In these cases the bars have been bent at 30 degrees but a tie-back system is used, which prevents a buildup of forces. To prevent this buildup, (in all cases except the equipment hatch penetration),

the line of force makes an angle of one-half of the angle of bend, from a horizontal line from the vertical bars and from a vertical line for the horizontal bars and is tangent to the outside of the penetration.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-118 of 3-126 Revision 0, 2020 At the personnel and equipment hatches a large void will be carried since, due to the large offset of the bars from their centerline, it will take the bars longer to return to their centerline after passing the penetration. To prevent any cracking and spalling of concrete and to add lost strength to the cross-section, these voids have been filled with added rebar, which achieves bond by means of mechanical anchorage.

The same precautions mentioned above have been taken with the seismic bars. See Figure 3.16-

16.

For penetrations between 9-in. and 18-in. in diameter, all the reinforcing bars including primary and secondary vertical bars and diagonal bars have been grouped around the penetrations. Due to the continuity of the bars and the relatively small opening size, no special provisions need be made to resist normal, shear, and bending stresses. The penetrations are keyed into the concrete, thus creating an edge loading, which will put torsion into the wall. The loads are small and the rebar will feel little effects from this torsional loading.

For penetrations greater than 18-in. up to 48-in. in diameter, the bars are continuous. Due to the large angle of bend of these bars, a tie-back system is used, which offers additional resisting strength to shear, bending, and torsional stresses.

3.16.3.11 Seismic and Wind Design The design of the containment, which is a Class I structure (see Section 1.7), is based on a "response spectrum" approach in the analysis of the dynamic loads imparted by earthquake. The seismic design takes into account the acceleration response spectrum curves as developed by G. Housner. Seismic accelerations have been computed as outlined in TID-702410 and Portland Cement Association Publication.11 The following damping factors have been used:

Percent Component Critical Damping

1.

Containment structure 2.0

2.

Concrete support structure of reactor vessel 2.0

3.

Steel assemblies:

a. Bolted or riveted 2.5

b. Welded 1.0

4.

Vital piping systems 0.5

5.

Concrete structures above ground:

a. Shear wall 5.0

b. Rigid frame 5.0 As indicated in Section 3.16.2.2, ground accelerations used for design purposes are 0.1g applied horizontally and 0.05g applied vertically. The natural period of vibration is computed by the Rayleigh method; in this method, the containment structure is analyzed as a simple cantilever intimately associated with the rock base and with broad base sections of adequate strength to

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-119 of 3-126 Revision 0, 2020 assure full and continued elastic response during seismic motions. Further, both bending and shear deformations are considered.

The structure is divided into sections of equal length and loaded laterally by dead weight of the section and any equipment and live load occurring at the section. Deflections caused by shear and moments are then determined, and the end deflection is given the value ' = 1.0 with corresponding values determined for other sections. The natural period of vibration for the structure is then determined by setting potential energy equal to kinetic energy and solving for the period.

T = 2

dm g

dm 2

Y0 1/2 where Y0 = maximum actual deflection

= deflection of section under consideration maximum actual deflection g = acceleration due to gravity dm = weight of section under consideration T = period in sec.

Based on an uncracked concrete section, the period is determined to be 0.241 sec. A more realistic calculation for a cracked section, using reinforcing steel and liner as the resisting elements, yields a period 0.936 sec.

Using the derived period and entering the acceleration spectral curves, Figures 1.7-1 and 1.7-2 of Section 1.7, and applying a 2-percent critical damping, a spectral acceleration for the containment was selected. This value was derived to determine the base shear. The distribution of base shear is a triangular loading assumption.

This assumption yields a load distribution pattern with zero loading at the base to a maximum loading at the spring line of the dome. Above this line, the loading decreases due to a change in section and consequently change in weight. This load distribution allows the determination of shears and moments at any critical section through the containment from which the appropriate unit stresses are obtained.

Seismic shears are resisted by diagonal reinforcing except in the upper areas of the dome. No credit is taken for the reinforcing in compression.

From 30 degrees above the springline, where the seismic shears are small, the shears are carried by dome reinforcing steel lying in the plane of principal tension

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-120 of 3-126 Revision 0, 2020 A finite element analysis was performed on the basemat using loads determined for the three basic loading conditions specified in the Containment Design Report. Maximum hoop moment caused by lack of symmetry of the seismic loading was found to be 454 in.kips/in. This compares with a capacity of 690-in.-kips/in. for the in-place hoop reinforcing.

Tornado loads have not been considered in the design of the Indian Point Unit 2 Containment Building; however, similarity in design of Indian Point Unit 3 (where such loads are considered) indicates that seismic reinforcement bars provide a more than adequate mechanism to withstand the torsional effect of Tornado loads.

The torsional effect results from wind striking the containment building at an angle from the normal, as shown in Figure 3.16-17. The torsional force is due to the component of the wind tangential to the surface of the containment building and is equal to:

Ft = ACD (q) (sin )

Where A = surface area of the containment CD = 0.5 from A.S.C.E. Paper 3269 - "Transactions of the A.S.C.E.," Vol. 126 Part II 1961, p. 1165 (coefficient of drag) q = 0.002558 V2 (wind pressure)

= 45 degrees This assumption is conservative in that the actual tangential force would be the result of skin friction and the effects would be negligible.

This component of torsional force is computed from a direct wind loading as based on A.S.C.E.

Paper 3269.

Torsional shear is a maximum at the juncture of the walls and base slab and varies to zero at the top of the dome.

The torsional effect can be converted to a shear per lineal foot around the circumference of the containment by distributing the shear over the circumference of the seismic reinforcing.

The seismic bars provide a more than adequate mechanism to withstand this torsional effect. The maximum stress in the bars under this loading is 17 ksi. See Figure 3.16-17.

3.16.3.12 Cathodic Protection (Historical Information)

During the initial Licensing process, a complete survey and tests to determine the need for cathodic protection on Indian Point Unit 2 was made by the A. V. Smith Engineering Company of Narberth, Pennsylvania. Electrical resistivity measurements and a visual inspection of the area away from the river, where the turbine generator building, reactor building, primary auxiliary building and associated facilities are located indicated that the environment is mostly rock with

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-121 of 3-126 Revision 0, 2020 areas of dry sandy clay. The electrical resistivity of the soil ranged from 3,500 to 30,000 ohm-cm with the majority of the readings being above 10,000 ohm-cm. On this basis, it was determined that cathodic protection was not required on underground facilities in areas away from the river or the containment building liner, although a protective coating on pipes was recommended to eliminate any random localized corrosion attack. An analysis of Hudson River water data, obtained from the Con Edison plant chemist, showed the electrical resistivity of the water to vary over an extremely wide range due to salt intrusion from the ocean. The range of resistivity has been from 59 to 10,000 ohm-cm with a large number reading in the 300 ohmcm area. This value was considered to be extremely corrosive and the following structures in the area near the river were placed under cathodic protection:

1.

Circulating water lines.

2.

Service water lines.

3.

Bearing piles.

4.

Sheet piling (earth and water side) and wing wall anchorage system.

5.

Metallic structures inside intake structure (traveling screens, bar racks, circulating water pump suction, service water pump suction).

In 2008, a cathodic protection field survey and assessment of underground structures at Indian Point Unit 2 was performed by PCA Engineering of Pompton Lakes, New Jersey. A positive shift in pipe potential was found where the City Water supply piping from the City Water Tank crosses the Algonquin Gas pipes. As a result the City Water supply piping in the vicinity of the gas pipes was placed under cathodic protection.

In 2009, a guided wave assessment of buried piping at Indian Point Unit 2 was performed by Structural Integrity Associates, Inc. of Centennial, Colorado. The assessment identified minor corrosion indications on the Unit 2 CST Condensate supply and return piping in the vicinity of the AFW Pump Building. As a result this piping was placed under cathodic protection.

The cathodic protection system for the Circulating Water lines and the Service Water lines were found not to be functional and the rectifiers were removed. In order to assure the lines will perform their functions, the buried pipes are inspected as part of the Underground Piping and Tank Program. Inspections of buried piping are initially performed using Guided Wave (GW) ultrasonic inspection techniques to locate potential areas of degradation. If significant degradation is detected during the GW inspections, excavation is performed to uncover the affected sections of piping and a direct visual inspection and UT thickness measurements are performed. Repairs and / or replacements are implemented as required to restore degraded piping sections to within the required structural margins of safety.

In addition to the inspections performed as part of the Underground Piping and Tank Program, the historically categorized nuclear safety related portion of the service water piping is further subjected to pressure and / or flow testing as required by ASME XI, Subsection IWA-5244. Visual inspections on the inside surface of the SW piping are also performed under the GL 89-13, Service Water program. Based on the results of the inspections and testing, the Service Water system is structurally adequate to perform its required safety function.

The cathodic protection system for the Traveling Water Screens and Bar Racks were found not to be effective and the installed cathodic protective systems were retired. The original Traveling Water Screens which were carbon steel were upgraded to stainless steel frames, baskets, and chains. The splash housings are also stainless steel. The Bar Racks, replaced in the mid-1990s,

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-122 of 3-126 Revision 0, 2020 are of carbon steel construction and are epoxy coated with a tar epoxy to provide corrosion protection. The guides for the Screens and Racks are carbon steel channels mounted in a concrete through. The rate of corrosion is slow and the Screens and Racks are on a regular PM cycle that checks for degraded conditions.

The Service Water and Circulating Water pumps suction are not cathodically protected. Rather, the Service Water Pumps suction is inspected and refurbished as part of the Service Water Pumps Preventing Maintenance (PM) activities. The Circulating Water Pumps are inspected and refurbished according to Preventive Maintenance program requirements.

3.16.3.13 Containment - Shear Crack The arrangement of reinforcing bars in the containment shell is such that a reinforcing bar crosses any potential crack plane. Any cracks resulting from diagonal tension caused by shearing forces will be carried by reinforcing bars, which span across the crack. Thus all shears will be carried by the reinforcing bars and none by the concrete.

The reinforcing bars are almost all continuous throughout the containment structure; however, where a bar terminates this is accomplished by means of a 180-degree hooked bar. In no case are bars simply terminated without providing means for additional anchorage.

Throughout the cylinder, the meridional reinforcing is continuous. Beyond the springline, the bars extend radially toward the center of dome. As the bars reach a 6-in. spacing, which is one-half the required spacing, alternate bars have been dropped off by means of reinforcing splice plates.

The splice piece consists of a plate with two Cadweld sleeves welded on the incoming side and one sleeve welded on the outgoing side. Thus, the number of bars present is halved and the spacing is increased to the required 12-in.

This is repeated to the top of the dome where a three-layered grid pattern has been used to maintain the continuity of the rebars. The bars in the grid pattern have been Cadwelded to the same type reinforcing splice plates described above, but the Cadweld is beveled to obtain the desired direction of the grid.

At the base in the area of high discontinuity stresses, additional No. 18S bars have been provided.

At the point where they were no longer needed, they have been Cadwelded to a No. 11 bar, which is terminated with a 180-degree hook.

All seismic bars have been terminated in a 180-degree hook. In no case was a No. 18S bar terminated in this way since the minimum 180-degree hook could not be provided in a 4-ft 6-in.

thick wall.

Radial shear reinforcing stirrups were terminated by hooking around vertical bars.

3.16.5 Longitudinal Splitting The cavity wall is designed to withstand the forces and internal pressurization associated with a longitudinal split without gross damage. See Section 3.16.3.7 for a discussion of the analysis of this assumed historical accident condition.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-123 of 3-126 Revision 0, 2020 3.16.6 Containment Structure Design Evaluation 3.16.6.1 Performance Capability Margin The containment structure is designed based upon limiting load factors, which are used as the ratio by which historical postulated accident and earthquake loads are multiplied for design purposes to ensure that the load/deformation behavior of the structure is one of elastic, low strain behavior. This approach places minimum emphasis on fixed gravity loads and maximum emphasis on accident and earthquake loads. Because of the refinement of the analysis and the restrictions on construction procedures, the load factors primarily provide for a safety margin on the load assumptions. Load combinations and load factors used in the design, which provide an estimate of the margin with respect to all loads, are tabulated in Section 3.16.2.

3.16.7 Preoperational Tests After the containment building was complete with liner, concrete structures, and all electrical and piping penetrations, equipment hatch and personnel locks were in place, the following tests were performed.

3.16.8 Strength Test A pressure test was made on the completed building using air at 54 psig. This pressure was maintained on the building for a period of at least 1 hr. During this test, measurements and observations were made to verify the adequacy of the structural design. For a description of observations, cracks, strain gauges, etc., refer to Reference 3.16-18.

REFERENCES FOR SECTION 3.16

1.

"Evaluation of Protection Coatings for Use in Reactor Containment," WCAP-7198-L, April 1968.

2.

J.B. Stellmeyer, W. H. Munse, and E. A. Selby, Fatigue Tests of Plates and Beams with Stud Shear Connections, Highway Research Record, No. 76.

3.

Robert C. Singleton, "The Growth of Stud Welding," Welding Engineer, July 1963.

4.

L. S. Beedle, "Plastic Design of Steel Frames," P. 131, Copyright 1958, Second print, 1961.

5.

Design Data, Nelson Concrete Anchor, Printed August 1, 1961.

6.

M. Hetenyi, "Beams on Elastic Foundation," Ann Arbor Press, 1955.

7.

S. Timoshenko, and D. H. Young, "Elements of Strength of Materials," D. Van Nostrand Company, Inc., 1962.

8.

S. Timoshenko, and S. Woinowsky-Krieger, Theory of Plates and Shells, Second Edition, McGraw-Hill, 1959.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-124 of 3-126 Revision 0, 2020

9.

American Concrete Institute, Specification for the Design and Construction of Reinforced Concrete Chimneys ACI-505-54, ACI Manual of Concrete Practices, Part 2, 1967.

10.

United States Atomic Energy Commission, "Nuclear Reactors and Earthquakes,"

TID-7024, 1963.

11.

J. Blume, N. Newmark, and L. Corning, "Design of Multistory Reinforced Concrete Buildings for Earthquake Motions," Portland Cement Association, 1961.

12.

Not Used.

13.

Not Used.

14.

Raymond J. Roark, "Formulas for Stress and Strain," McGraw-Hill Book Company, Fourth Edition.

15.

R. V. Whitman, and F. E. Richart, Jr. "Design Procedures for Dynamic Loaded Foundations," A.S.C.E., Journal of the Soil Mechanics and Foundations Division, Vol. 93, No. SM6, 1967.

16.

Paul F. Rice, Detailing and Placing of Reinforcing Bars, Concrete Construction, January 1965.

17.

S. Timoshenko and Goodier, Theory of Elasticity, Second Edition, McGraw-Hill Book Co., 1961.

18.

Wiss, Janney, Elstner & Associates, Inc., Consulting & Research Engineers, Structural Response of Secondary Containment Vessel During Structural Integrity Test at Indian Point Power Generating Station Unit No. 2, prepared for Wedco Corporation, April 22, 1971.

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-125 of 3-126 Revision 0, 2020 TABLE 3.16-1 Flooded Weights - Containment Building Item Flooded Operating Weight, lbs.

Pressurizer - 1 346,000 Steam generators - 4 3,746,000 Reactor - 1 (a)

Vessel 868,000 (b)

Internals 420,000 (c)

Piping 1,000,000 Reactor pumps - 4 824,000 Accumulator tanks - 4 529,000 175-ton polar crane - 1 650,000 Ventilation fans - 5 656,000 Reactor coolant drain tank - 1 20,000 Pressure relief tank - 1 100,000 Other miscellaneous equipment 100,000 Total 9,259,000

IP2 DEFUELED SAFETY ANALYSIS REPORT Page 3-126 of 3-126 Revision 0, 2020 3.16 FIGURES Figure No.

Title Figure 3.16-1 Containment Structure Figure 3.16-2 Cylinder and Dome-Load Condition (A) - 1.5P Figure 3.16-3 Cylinder and Dome-Load Condition (B) - 1.25P Figure 3.16-4 Cylinder and Dome-Load Condition (C) - 1.0P Figure 3.16-5 Loading Diagram in Mat-Load Condition (A) - 1.5P Figure 3.16-6 Loading Diagram in Mat-Load Condition (B) - 1.25P Figure 3.16-7 Loading Diagram in Mat-Load Condition (C) - 1.0P Figure 3.16-8 Weld Stud Connection at Panel Low Point Figure 3.16-9 Weld Stud Connection at Panel Low Point Figure 3.16-10 Weld Stud Connection at Panel Center Figure 3.16-11 Wall Section Figure 3.16-12 Cylinder Base Slab Liner Juncture Figure 3.16-13 Typical Base Mat Liner Detail Figure 3.16-14 Base Slab Reinforcing Detail Figure 3.16-15 Reactor Cavity Pit Figure 3.16-16 Equipment Hatch Personnel Lock, Main Steam and Feedwater, Air Purge - Rebar Figure 3.16-17 Torsional Effects