ML23292A138
| ML23292A138 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/12/2023 |
| From: | Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23291A105 | List:
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| References | |
| PLA-8081 | |
| Download: ML23292A138 (104) | |
Text
SSES-FSAR Text Rev. 79 9.1 FUEL STORAGE AND HANDLING 9.1.1 NEW FUEL STORAGE 9.1.1.1 Design Bases 9.1.1.1.1 Safety Design Bases 9.1.1.1.1.1 Safety Design Bases - Structural a) The new fuel storage racks containing a full complement of fuel assemblies are designed to withstand all credible static and dynamic loadings to prevent damage to the structure of the racks, and therefore the contained fuel, and to minimize distortion of the racks arrangement (See Table 3.9-2s).
b) The racks are designed to protect the fuel assemblies from excessive physical damage which may cause the release of radioactive materials in excess of 10 CFR 20 requirements under normal conditions.
c) The racks are constructed in accordance with the Quality Assurance Requirements of 10 CFR 50, Appendix B.
d) The new fuel storage racks are categorized as Safety Class 2 and Seismic Category I.
9.1.1.1.1.2 Safety Design Bases - Nuclear a) The new fuel storage racks are designed and maintained with sufficient spacing between new fuel assemblies to assure that the fully loaded array in dry storage or fully flooded conditions has a keff 0.95 including allowance for calculational biases and uncertainties.
b) The new fuel storage vault is covered by leak tight metal removable covers. These covers prevent an optimum moderator (e.g., fire-fighting foam) from reaching the new fuel. The movement of these covers is administratively controlled by approved plant procedures.
c) The new fuel storage vault criticality calculations assumed that the storage array was infinite in all directions. Since no credit is taken for leakage, the values reported as effective neutron multiplication factors are in reality infinite neutron multiplication factors.
d) The biases between the calculated results and experimental results as well as the uncertainty involved in the calculations are taken into account as part of the calculational procedure to assure that the specified keff limits are met. Also when fuel is in the new fuel storage vault two radiation monitors are utilized to detect criticality in the new fuel storage vault.
FSAR Rev. 71 9.1-1
SSES-FSAR Text Rev. 79 9.1.1.1.2 Power Generation Design Bases a) New fuel storage racks are supplied for 30% of the full core fuel load in each unit.
b) New fuel storage racks are designed and arranged so that the fuel assemblies can be handled efficiently during refueling operations.
9.1.1.2 Facilities Description The location of the new fuel storage facility within the station complex is shown in Section 1.2.
Each new fuel storage rack (Figure 9.1-1) holds up to 10 channeled or unchanneled assemblies in a row. Fuel spacing (7 inches nominal center-to-center within a rack, 12 inches nominal center-to-center between adjacent racks) within the rack and from rack-to-rack, coupled with limits on fuel lattice reactivity, will limit the effective multiplication factor of the array (keff) to not more than 0.95. The fuel assemblies are loaded into the rack through the top. Each hole for a fuel assembly has adequate clearance for inserting or withdrawing the assembly channeled or unchanneled. Sufficient guidance is provided to preclude damage to the fuel assemblies. The upper tie plate of the fuel element rests against the rack to provide lateral support. The design of the racks prevents accidental insertion of the fuel assembly in a position not intended for the fuel. This is achieved by abutting the sides of each casting to the adjacently installed casting.
In this way, the only spaces in the new fuel racks are those into which it is intended to insert fuel. The weight of the fuel assembly is supported by the lower tie plate which is seated in a chamfered hole in the base casting.
The floor of the new fuel storage vault is sloped to a drain located at the low point. This drain removes any water that may be accidentally and unknowingly introduced into the vault. The drain is part of the liquid radwaste collection system.
The area radiation monitoring equipment for the new fuel storage area is described in Subsection 12.3.4.
9.1.1.3 Safety Evaluation 9.1.1.3.1 Criticality Control New Fuel Vault criticality analyses demonstrate that General Design Criterion 62 requirements (Prevention of Criticality in Fuel Storage and Handling) are met if fuel is stored in the normal dry condition or if the abnormal condition of flooding occurs (Reference 9.1-14). Susquehanna complies with the requirements of 10 CFR 50.68(b).
The calculations of keff are based upon an infinite geometrical arrangement of the fuel array using Framatome methods (Reference 9.1-14). The arrangement of fuel assemblies in the fuel storage racks results in keff 0.95 for both the dry storage condition and the fully flooded condition, assuming the most reactive fuel and moderator temperatures.
The New Fuel Storage Vault has not been designed to preclude criticality at optimum moderation between dry and flooded conditions (e.g., fire-fighting foam). Watertight covers are used as measures to prevent an inadvertent criticality. As an added precaution, criticality monitors have been installed. Administrative controls restrict the use of foam on the Refueling Floor and on the Reactor Building roof during those times when new fuel is being stored in the new fuel storage vault.
FSAR Rev. 71 9.1-2
SSES-FSAR Text Rev. 79 9.1.1.3.2 New Fuel Rack Design a) The new fuel storage vault contains 23 sets of castings each of which may contain up to 10 fuel assemblies; therefore a maximum of 230 fuel assemblies may be stored in the fuel vault.
b) There are three tiers of castings which are positioned by fixed box beams. This holds the fuel assemblies in a vertical position and supported at the lower and upper tie plate with additional lateral support at the center of gravity of the fuel assembly.
c) The lower casting supports the weight of the fuel assembly and restricts the lateral movement; the center and top casting restricts lateral movement only of the fuel assembly.
d) The new fuel storage racks are made from aluminum. Materials used for construction are specified in accordance with ASTM specifications in effect in 1970. The material choice is based on a consideration of the susceptibility of various metal combinations to electrochemical reaction. When considering the susceptibility of metals to galvanic corrosion, aluminum and stainless steel are relatively close together insofar as their coupled potential is concerned. The use of stainless steel fasteners in aluminum to avoid detrimental galvanic corrosion in a predominantly air environment, is a recommended practice and has been used successfully for many years by the aluminum industry.
e) The minimum center-to-center spacing for the fuel assembly between rows is 11.875 inches. The minimum center-to-center spacing within the rows is 6.535 inches. Fuel assembly placement between rows is not possible.
f) Lead-in and lead-out of the casting, in the rack, provides guidance of the fuel assembly during insertion or withdrawal.
g) The rack is designed to withstand the impact force of 4000 ft-lbs while maintaining the safety design basis. This impact force could be generated by the vertical free fall of a fuel assembly from the height of 5.3 feet.
h) The storage rack is designed to withstand the pull-up force of 4000 lbs. and a horizontal force of 1000 lbs. There are no readily available forces in excess of 1000 lbs.
i) The storage rack is designed to withstand horizontal combined loads up to 222,000 lbs, well in excess of expected loads.
j) The maximum stress in the fully loaded rack in a faulted condition is 25.9 Kips (See Table 3.9-2s). This is lower than the allowable stress.
k) The fuel storage rack is designed to handle non-irradiated, low emission radioactive fuel assemblies. The expected radiation levels are well below the design levels.
FSAR Rev. 71 9.1-3
SSES-FSAR Text Rev. 79 l) The fuel storage rack is designed using non-combustible materials. Plant procedures and inspections assure that combustible materials are restricted from this area. Fire prevention by elimination of combustible materials and fluids is regarded as the prudent approach rather than fire accommodation and the need for fire suppressant materials which could negate criticality control assurances. Therefore, fire accommodation is not considered necessary.
m) The new fuel vault covers, which are carbon steel, are illustrated in Figure 9.1-2. The covers overlap the curb and have a protective lip that prevents direct impingement of water into the vault. The modified I-beams that span the vault provide mechanical support and direct water run-off from the covers.
9.1.2 SPENT FUEL STORAGE Spent fuel is stored both in the Reactor Building Spent Fuel Storage Pools and at the Independent Spent Fuel Storage Installation (ISFSI). This Section applies only to spent fuel storage in the Reactor Building. Spent fuel storage in the ISFSI is described in Section 11.7.
9.1.2.1 Design Bases 9.1.2.1.1 Safety Design Bases 9.1.2.1.1.1 Safety Design Bases - Structural a) The high density spent fuel storage racks containing a storage space sufficient for approximately 372% of one full core of fuel assemblies are designed to withstand all credible static and dynamic loadings to prevent excessive damage to the structure of the racks, and therefore the contained fuel (See Table 3.9-2(af)).
b) The racks are designed to protect the fuel assemblies from excessive physical damage which may cause the release of radioactive materials in excess of 10 CFR 20 requirements under normal or abnormal conditions.
c) The racks are constructed in accordance with the Quality Assurance Requirements of 10 CFR 50, Appendix B.
d) The spent fuel storage racks are categorized as Safety Class 2 and Seismic Category I.
e) The spent fuel pool structure and the anchorage system to the fuel storage racks are categorized as Seismic Category I.
9.1.2.1.1.2 Safety Design Bases - Nuclear The effective neutron multiplication factor (Keff) of the fuel array in any combination of any stored positions up to and including the fully loaded condition is less than or equal to 0.95. The positioning of the neutron poisoning material (boral) between adjoining fuel assemblies assures subcriticality by at least 5% K under all normal and abnormal conditions. Consideration has been given to the geometry of the racks, possible abnormal loading, and the density of the coolant/moderator. As an additional precaution, two radiation monitors will detect a criticality event in the spent fuel pool.
FSAR Rev. 71 9.1-4
SSES-FSAR Text Rev. 79 9.1.2.1.2. Power Generation Design Bases The spent fuel storage pool and fuel storage racks are designed to assure:
a) subcriticality, by at least 5% K b) decay heat from fuel assemblies/bundles will not adversely affect the fuel, racks, or pool walls.
c) radiation levels will be "As Low As Reasonably Achievable."
9.1.2.1.3 Storage Capacity Design Bases Each reactor unit has a spent fuel pool which has high density fuel storage racks providing a maximum storage capacity of 2840 fuel assemblies. These fuel locations also provide storage of used fuel channels, if needed. In addition, the fuel storage rack design provides storage of 10 various reactor internal components, such as:
a) control rods b) control rod guide tubes c) defective fuel storage containers d) "out-of-core" sipping containers This capacity provides each reactor unit storage space for off loading one-quarter (1/4) of a core for approximately ten (10) years, plus one complete core load of fuel.
Each reactor unit's spent fuel pool is interconnected, via a transfer canal. Spent fuel may be transferred safely, through this transfer canal, to the other pool. This capability provides greater flexibility for the stations storage of spent fuel, if the need ever arises.
Each reactor unit's spent fuel pool walls also have storage hangers for one hundred and thirty control rods. These hangers, empty or full of control rods, do not interfere with the storage of fuel or the other mentioned reactor internal components in this section. Extended sling assemblies may be used in conjunction with the hangers to store the control rods at a lower depth in the pool.
9.1.2.2 Facilities Description The location of the spent fuel storage facility within the station complex is shown in Dwgs.
M-246, Sh. 1 and M-256, Sh. 1. The racks are connected to wall embedments on the pool walls and shown in Figure 9.1-4. Each pool has 24 racks for a storage capacity of 2840 fuel assemblies plus 10 multipurpose cavities for storage of control rods, control rod guide tubes, and defective fuel containers, and out-of-core sipping containers.
The spent fuel liner plate is not a structural element (i.e., it is not load bearing). The fuel racks are attached to the pool walls by embeds and anchors, which are designed for all credible loads (see Table 9.1-7a). The liner plate is welded to these embeds. In addition, the liner plate is attached to the pool walls by a system of stiffeners and anchors. The racks, embeds and fuel pool walls and liner plate (including anchor system) are designed for all credible loads.
FSAR Rev. 71 9.1-5
SSES-FSAR Text Rev. 79 A leak detection system is provided for the collection of possible leakage through the pools' liner plate. The liner leakage detection system is segregated into sections that collect leakage at independent locations below the pools. Drainage paths are formed by welded channels behind the liner weld joints and are designed to permit free gravity flow to manual telltale valves.
This system is provided to:
a) Prevent pressure buildup behind the liner plate b) Prevent the uncontrolled loss of contaminated pool water to other cleaner locations within the secondary containment, and c) Provide expedient liner leak detection and measurement.
Both Units 1 and 2 share a common cask pit that accepts the spent fuel shipping cask and accommodates underwater fuel transfer to the cask from either unit through its respective transfer canals. Movements of the cask on the refueling floor are restricted as shown on Drawing C-1807, Shts. 1 & 2.
The evaluations of the consequences of a postulated accidental drop of a spent fuel assembly and the shipping cask are discussed in Chapter 15. The capability of the spent fuel pool storage facility to prevent missiles generated by high winds from contacting the fuel is discussed in Subsection 3.5.2.
The rack arrangement is designed to prevent accidental insertion of fuel bundles between adjacent racks.
The five (5) foot to six (6) foot thick spent fuel pool walls provide radiation shielding to 2.5 Mrem/Hr measured on the outside of the spent fuel pool walls. Normal water shielding over the stored fuel in the racks is approximately 23 feet and is sufficient to provide shielding for required building occupancy. Under the normal water level conditions, about 8.5' of water is above the active fuel when moved through the refueling channel. This depth of water provides shielding to assure less than 2.5 Mrem/Hr to the operators on the refueling platform.
Accidental droppage of heavy objects into the fuel pool is precluded by the use of administrative procedures, electrical interlocks to limit the reactor building crane travel over the spent fuel pool, and the use of guardrails and curbs around all pools and the reactor wells to prevent fuel handling and servicing equipment from falling into the pools.
The spent fuel pools, reactor wells, dryer-separator storage pools, and common shipping cask pool including all gates are designed to Seismic Category I requirements. All pools and wells are lined with stainless steel to minimize leakage and reduce corrosion product formation. The spent fuel pools are further designed so that they cannot be drained to a level that uncovers the top of the stored fuel. The normal water shielding over the stored fuel in the racks is approximately 23 ft. However, in the unlikely event that the pool gates fail to contain the pool water, the fuel racks and their contained fuel are assured of maintaining water coverage at all times.
Cooling water supply lines enter the spent fuel pool from above the normal water level and are provided with high point siphon breaking vent lines to prevent siphoning of water from the pools.
FSAR Rev. 71 9.1-6
SSES-FSAR Text Rev. 79 The superstructure of the reactor building serves as a low leakage barrier to provide atmospheric isolation of the spent fuel storage pool and associated fuel handling area.
The superstructure is composed of structural steel framing, metal siding and metal roof decking.
The superstructure is designed to Seismic Category I criteria.
Features to limit potential offsite exposures in the event of significant release of radioactivity from the spent fuel have been provided.
These include a ventilation exhaust system, isolation of the secondary containment on high radiation, air mixing, and a standby gas treatment system capable of maintaining the secondary containment at 1/4-in. water column negative pressure with respect to the outside ambient pressure. These features are discussed in Subsection 9.4.2.
The radiological considerations for the spent fuel storage arrangement are described in Chapter 12.
9.1.2.3 Safety Evaluation 9.1.2.3.1 Criticality Control Critically analyses have been performed for bounding reference fuel assemblies to demonstrate that storage of fuel assemblies of each design in the spent fuel pool high density racks results in a keff 0.95 (Reference 9.1-15). Storage of ATRIUM 11 has been evaluated for storage with all previously used fuel types and its evaluation bounds all previous fuel types (Reference 9.1-15).
Susquehanna complies with the requirements of 10 CFR 50.68(b).
Calculations for ATRIUM 11 were performed by Framatome using the KENO Monte Carlo Code and CASMO-4 (Reference 9.1-15).
The analysis is based upon an infinite array of fuel assemblies loaded in the spent fuel pool rack cell geometry. The bounding analysis for ATRIUM 11 was performed at 4oC (39.2oF). The analysis addresses the effects of water density (voids), calculational and manufacturing uncertainties, storage of assemblies with or without fuel channels, and bundle orientation within the racks.
The spent fuel storage pool has also been analyzed under abnormal and accident conditions.
These conditions included a fuel bundle placed vertically along the edge of the spent fuel pool, a fuel bundle laid horizontally on the top of the spent fuel pool racks, and a single missing Boral panel from the storage array. For all normal, abnormal, and accident conditions, the spent fuel pool rack keff remained less than 0.95.
9.1.2.3.2 High Density Fuel Storage Rack Design Spent fuel storage racks provide a place in the spent fuel pool for storing new and spent fuel.
The high density spent fuel racks contain a neutron-absorbing medium of natural boron carbide (B4C) in an aluminum matrix core clad with 1100 series aluminum. This neutron absorber is marketed under the trade name of Boral.
FSAR Rev. 71 9.1-7
SSES-FSAR Text Rev. 79 Boral slabs are manufactured under a proprietary qualified process. This process assures a uniform minimum B-10 areal density of 0.0233 gm/cm2 in the Boral slabs utilized in the construction of the Susquehanna Racks. Benchmark measurements of those slabs yield a neutron attenuation factor of 0.963 minimum.
The rack manufacturer assured that correct Boral locations and quantities were present in accordance with the design and procurement documents through a rigorous quality assurance program that was evaluated and approved by the AE. The construction of the rack assures that all adjacent storage cavities are separated by a Boral slab. The Boral is sealed within two concentric square aluminum tubes referred to as poison cans.
Boral panel placement in a rack cell is shown in Figure 9.1-5. Nominal dimensions for a fuel storage cell poison can are shown in Figure 9.1-6.
Figure 9.1-3 shows the structural design. Each rack module consists of six basic components:
- 1) top grid casting
- 2) bottom grid casting
- 3) poison cans
- 4) side plates
- 5) corner angle clips
- 6) adjustable foot assemblies Each component is anodized separately.
The top and bottom grid casting are machined to maintain a nominal fuel pitch (center-to-center spacing) of 6.625 inches. Within these machined areas, in a checkerboard pattern, Boral poison cans are nested. This ensures smooth entry and removal of fuel assemblies in each fuel cavity. This design also assures Boral is located between adjacent fuel assemblies. To complete the module, the grids are bolted and/or riveted together by four corner angles and four side shear panels. Adjustable foot assemblies are located at the four corners of each module to allow adjustment for variations of the pool floor level of +/-0.75 inches. To maintain a flat, uniform contact area, the leveling screw bearing pads are free to pivot.
Each module is level with each other module at the top. There is nominally seven inches of clearance from the bottom of the module to the pool floor. This assures adequate clearance for cooling water to enter each fuel cell and keep each fuel assembly cool through natural convection.
The modules are bolted together into four super modules. The perimeter modules have seismic bracing to embedments in the pool wall assuring structural integrity through all anticipated dynamic loads. The weight of the fuel assembly is supported in the chamfered hole in the bottom casting. Nominal center-to-center fuel spacing between modules is 9.375 inches.
a) The square poison cans are positioned in a top and bottom grid in a checkerboard pattern. Each poison can is pressure and vacuum leak tested for integrity.
FSAR Rev. 71 9.1-8
SSES-FSAR Text Rev. 79 b) The seismic restraints from the racks to the wall embedments consist entirely of a welded stainless steel construction. To reduce any galvanic corrosion, inconel pins are used between the wall seismic restraints and racks. The only interface of each module with the pool floor are four stainless steel pads attached to the rack leveling screws. A 1/4 inch ABS plastic material is volumetrically captured between this pad and the aluminum leveling screw to prevent galvanic corrosion with the pool floor stainless steel liner plate.
c) All materials used for construction are specified in accordance with the ASTM specifications, as applicable. The ASTM Standards used for stainless steel were A240-72b, A276-71 and A312-72a and for aluminum were B209-73, B26-74 and B211-74.
Traceability of major rack components to a heat lot are maintained.
In addition, the suppliers' quality assurance-quality are control program audited by the AE and user, in effect to ensure that the Boral has the required minimum B4C density and uniform B4C distribution in each sheet. Boral traceability is maintained.
d) A dimensional, visual, and functional (including testing with a dummy fuel assembly) inspection of the racks is performed prior to shipment by the rack manufacturer.
e) The rack materials have no significant degradation due to the total radiation doses expected in the spent fuel pool over the design life.
f) The minimum center-to-center fuel spacing within a rack assembly is 6.500". The minimum center-to-center fuel spacing between racks is 9.125". Fuel assembly placement between modules or cavities of a module are not possible.
g) The racks are designed to withstand the loading under the following loading conditions:
dead, live, jammed fuel assembly, dropped fuel assembly, thermal, OBE and DBE seismic, SRV, and LOCA or Chugging.
h) The racks are installed in the pool on four tension and compression quadrants to eliminate thermal loads resulting from confined expansion.
i) An inservice inspection (ISI) program will be in effect throughout the life of the racks to assure quality of the poisoned racks is maintained, as described in Subsection 9.1.2.3.3.
9.1.2.3.3 Inservice Inspection Sixteen test coupons are to be provided for an on-going in-service inspection program. Two coupons, one of which is vented and the other sealed, would be removed and analyzed at intervals of 1, 3, 5, 10, 15, 20, 30, and 40 years after installation. One set of coupons will be tested during the tenth or eleventh year after Unit 1 enters the Period of Extended Operation.
9.1.2.3.3.1 Test Coupon Description and Installation A typical test coupon is a shortened production-type can similar to the spent fuel rack. Four sheets of BORAL neutron poison are encapsulated between the inner and outer cans. After assembly, the entire coupon is anodized.
FSAR Rev. 71 9.1-9
SSES-FSAR Text Rev. 79 The sealed cans are pressure-checked through a hole in the outer can. This hole is then welded to prevent water from contacting the BORAL. The unsealed cans will also have a 13/64 inch hole which will not be welded closed.
Two test coupons, one vented and the other unvented, are tied together with a hanger. This hanger contains a handling eye so that they can be hung on the perimeter of the spent fuel rack.
9.1.2.3.3.2 Test Coupon Inspection a) The test coupon assembly will be removed from the spent fuel pool.
b) The test coupon will be drained of the entrapped water from the vented coupon and the pH of the water in the fuel pool will be determined.
c) The vented coupon will be disassembled sufficiently so that the neutron absorber can be extracted.
d) Upon disassembly, note whether there is water in the sealed coupon. If so, perform step #b above.
e) Visual inspection of the neutron absorber plates will be noted and any discoloration, corrosion damage or physical damage will be recorded. If corrosion or physical damage is noted, record depth and extent of damage.
f) The plates will be washed in a mild abrasive and detergent solution, then rinsed in clean water and/or acetone. The plates will be dried in a 175°F oven for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, followed by 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in a 300°F oven and 4 additional hours in a 500°F oven. The plate weight will be determined, at room temperature, following each drying interval. Drying may be discontinued when no further weight loss occurs.
g) Each plate will be weighed and determine weight change.
h) Reperform step #e.
i) Perform neutron attenuation testing on each plate.
j) All data will be recorded, including pH values, for future comparison.
9.1.3 SPENT FUEL POOL COOLING AND CLEANUP SYSTEM 9.1.3.1 Design Bases The Fuel Pool Cooling and Cleanup System (FPCCS) is designed and operated with the following considerations:
a) The FPCCS is designed to maintain the fuel pool water temperature below 125°F.
FSAR Rev. 71 9.1-10
SSES-FSAR Text Rev. 79 START HISTORICAL The heat load which served as the basis for the FPCCS design is based upon filling the pool with 2840 fuel assemblies from normal refueling discharges and transferred to the fuel pool within 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> after shutdown. Tables 9.1-2a and 9.1-2b show the originally assumed discharge schedule and heat load.
END HISTORICAL Table 9.1-2e shows an updated discharge schedule which was used as the basis for the Appendix 9A analysis only. This analysis was based upon filling the pool with 2850 fuel assemblies from normal refueling discharge and transferred to the fuel pool within 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> after shutdown.
b) During an emergency heat load (EHL) condition one RHR pump and heat exchanger are available for fuel pool cooling. The EHL condition occurs when one fuel pool is full including a full core unloaded.
START HISTORICAL The original design bases for full core unloaded conditions occurred when the spent fuel racks of one spent fuel pool were filled with 2840 fuel assemblies including a full core discharged to the pool within 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> after shutdown (control rods inserted).
Tables 9.1-2c and 9.1-2d show the original discharge schedule and heat load that was assumed for the systems design for this condition for Units 1 and 2.
END HISTORICAL Table 9.1-2f shows an updated discharge schedule. This updated schedule for full core unload conditions is based upon filling the pool with 2850 fuel assemblies including a full core discharge to the pool within 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> after shutdown.
The RHR Fuel Pool Cooling mode (RHRFPC) will maintain the isolated fuel pool water temperature, (with the heat load of 4.02 x 107 BTU/hr) at or below 125°F with or without assistance from the FPCCS under normal refueling conditions. When the decay heat load of the spent fuel drops to the level for which the FPCCS is designed, the RHR system may be disengaged. For crosstied spent fuel pools, the RHRFPC mode in one unit in combination with the normal Fuel Pool Cooling system of the other unit will maintain the crosstied fuel pools at or below 125°F with the EHL in one pool and fuel at the normal scheduled off load rate in the other pool.
c) Following a seismic event, the normal Fuel Pool Cooling system is postulated to be unavailable due to its Non-Seismic Category I, Non-Class 1E power design. If such an event were to occur the RHR Fuel Pool Cooling (RHRFPC) mode would be used to provide cooling to the spent fuel pools to prevent boiling.
FSAR Rev. 71 9.1-11
SSES-FSAR Text Rev. 79 All piping and components of the RHRFPC mode are Seismic Category I, Quality Group B or C constructed to ASME Section III standards. The RHR System is Class 1E powered and both loops have separate power supplies. The RHRFPC system is hardpiped and requires operation of several manual valves (which are accessible following a seismic event) to establish the flowpath. In addition, other manual and motor operated valves must be operated in order to assure proper operation of the RHRFPC mode. Proper operation of all active components in the RHRFPC mode is confirmed on a periodic basis in accordance with plant procedures.
The RHR pump suction path for the Fuel Pool Cooling mode is shared with the Shutdown Cooling mode of RHR. Consequently, Shutdown Cooling and Fuel Pool Cooling cannot be performed concurrently on a given unit. However, Alternate Shutdown Cooling and Fuel Pool Cooling can be performed concurrently since different suction sources are used.
Appendix 9A contains an evaluation of a boiling spent fuel pool for a Non-Seismic Category I Fuel Pool Cooling system. Boiling of the spent fuel pool(s) would not occur during a seismic event due to use of the RHR Fuel Pool Cooling system as a backup Seismic Category I Fuel Pool Cooling system. The RHRFPC mode can be placed into service well in advance of the postulated time to boil of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> (see Subsection 9.1.3.3).
d) The FPCCS is designed to maintain the water clarity and quality in the pools as follows to facilitate underwater handling of fuel assemblies and to minimize fission and corrosion product buildup that pose a radiological hazard to operating personnel:
Conductivity 3 micromho/cm at 25°C pH 5.3 - 7.5 at 25°C Chloride (as CL-) 0.5 ppm Heavy elements (Fe,Cu,Hg,Ni) <0.1 ppm Total insolubles <1 ppm 9.1.3.2 System Description Each reactor unit is provided with its own FPCCS as shown on Dwgs. M-153, Sh. 1, M-153, Sh. 2, and M-154, Sh. 1.
The system cools the fuel storage pool water by transferring the decay heat of the irradiated fuel through heat exchangers to the service water system. During Refueling Outages, when the service water system is shutdown the decay heat is transferred through the FPCCS heat exchangers to the temporary cooling towers/chillers located outside the units or by circulating river water through the heat exchangers. The River Water Make Up line tap provided at the special manhole near North Gate House supplies river water for this purpose and also returns this water back into the RWMU line. This RWMU line tap and the temporary cooling towers/chillers are connected to the heat exchangers through temporary pumping equipment and Supplemental Decay Heat Removal piping.
Water clarity and quality in the fuel storage pools, transfer canals, reactor wells, dryer-separator pools, and shipping cask pit are maintained by filtering and demineralizing.
FSAR Rev. 71 9.1-12
SSES-FSAR Text Rev. 79 The FPCCS consists of fuel pool cooling pumps, heat exchangers, skimmer surge tanks, filter demineralizers, associated piping, valves, and instrumentation.
Equipment Description Table 9.1-1 shows the design parameters of the FPCCS equipment. The seismic and quality group classifications of the FPCCS components are listed in Section 3.2.
One skimmer surge tank for each unit collects overflow water from skimmer drain openings with adjustable weirs at the water surface elevation of each pool and well. The common shipping cask pit water overflows to both units' skimmer surge tanks.
Wave suppression scuppers along the working side of the fuel pools also drain to the skimmer surge tanks. The skimmer openings in the pool liners are protected with a wire mesh screen to prevent floating objects such as the surface breaker viewing aids from entering the surge tanks.
The adjustable weir plates are set according to the required cooling flow, desired flow pattern, and water shielding needs.
The skimmer surge tank provides a suction head for the fuel pool cooling pumps and a buffer volume during transient flows in the normally closed loop FPCCS. It provides sufficient live capacity for three days' normal evaporative loss from the fuel pool without makeup from the condensate transfer system. A removable object retention screen in the tank is accessible through the flanged tank top. Tank level indication and alarms on a control panel on the refueling floor and/or the vicinity of the fuel pool cooling pumps announce when the remote manual makeup valves must be opened or water drained from the system.
The fuel pool cooling pumps are stopped upon a low tank level signal.
Three fuel pool heat exchangers piped in parallel are located in the reactor building below the surge tank bottom elevation. The shell side is subjected to the static head of the skimmer surge tank level only. This is a minimum of 5 psi lower than the tube side service water pressure, thus minimizing the possibility of radioactive contamination of the service water system (see Subsection 9.2.1) from a tube leak.
The number of heat exchangers in service depends on the decay heat load from irradiated fuel in the spent fuel pool. The common inlet and each heat exchanger outlet temperature are recorded and high temperature alarmed on a local control panel.
Three fuel pool cooling pumps piped in parallel are placed in service in conjunction with the heat exchangers. They take suction from the heat exchangers and develop sufficient head to process a partial system flow through the filter demineralizers and transfer it combined with the bypass flow to the diffuser pipes at the bottom of the pools.
The pump controls, discharge pressure indicators, flow indicator, and alarms for low flow and low discharge pressure are provided on a local control panel.
The pumps trip individually upon low NPSH. Three fuel pool filter demineralizers are piped in parallel. One fuel pool filter demineralizer is designated for use with each unit with a spare filter demineralizer shared by both units. The design flow per filter demineralizer is less than the total system flow. Part of the cooled water is therefore bypassing at a manually adjustable rate.
FSAR Rev. 71 9.1-13
SSES-FSAR Text Rev. 79 If the inlet temperature should exceed 150°F, the filter demineralizer must be manually bypassed to prevent degradation of the ion exchange resin.
The filter demineralizer units are designed to operate with water flowing at nominal 2 gpm/sq. ft.
filter area. Powdered ion-exchange resin or resin mixed with Solka-Floc (or other filtering aid) is used as a filter medium. The filter elements are stainless steel mesh, mounted vertically in a tube sheet and replaceable as a unit. Venting is possible from the upper head of the filter vessel to the reactor building ventilation system. The upper head is removable for installation and replacement of the filter elements. The filter demineralizer units are located separately in shielded cells. Sufficient clearance is provided to permit removal of the filter elements from the vessels. Each cell contains only the filter demineralizer and connecting piping. All inlet, outlet, recycle, vent, drain valves, and the holding pumps are located in a separate shielded room, together with necessary piping and headers, instrument elements, and controls. Penetrations through shielding walls are located so that shielding requirements are not compromised.
A post-strainer is provided in the effluent stream of each filter demineralizer to limit the migration of the filter material. The post-strainer element is capable of withstanding a differential pressure greater than the shut-off head for the system.
The ion exchange resin is a mixture of finely ground, 300 mesh or less (average particle size),
cation and anion resins in proportions determined by service. The cation resin is a strongly acidic polystyrene with a divinylbenzene cross-linkage. The resin is supplied in fully regenerated hydrogen form. The anion resin is a strongly basic, Type I, quaternary ammonium polystyrene with a divinylbenzene cross-linkage. The resin is supplied in a fully regenerated hydroxide form.
The resin is replaced when the pressure drop is excessive or the ion exchange resin is exhausted. Backwashing and precoating operations are controlled from a local control panel in the reactor building. The spent filter medium is backwashed from the elements with instrument air and condensate and transferred via a receiving tank to the RWCU sludge phase separator or to the reactor water clean-up phase separators in the radwaste building.
New ion exchange resin is mixed in a resin tank and transferred as a slurry by a precoat pump to the filter where it is deposited on the filter elements. A separate precoat tank is provided to allow precoating of the filter elements with Solka-Floc (or other filtering aid) only or prior to depositing ion exchange resins. Both tanks are furnished with an agitator for mixing the filter medium slurries. The precoat subsystem is common to both FPCCS and may also be used for chemical cleaning of the filter demineralizers.
The holding pump associated with each filter demineralizer maintains circulation through the filter in the interval between the precoating operation and the return to normal system operation, or upon decrease in process flow below a point where the precoating may fall off the filter elements.
The filter demineralizers are controlled from a panel in the reactor building of Unit 1. Differential pressure and inlet and outlet pressure instrumentation are provided for each filter demineralizer unit to indicate when backwash is required. Suitable alarms, differential pressure indicators, and flow indicating controllers are provided to monitor the condition of the filter demineralizer and the post effluent strainers.
FSAR Rev. 71 9.1-14
SSES-FSAR Text Rev. 79 The backwash and precoat operations are push-button initiated, automatically sequenced operations. The filter demineralizer inlet and outlet conductivity is recorded and 1.2 micromho/cm in the outlet is alarmed on the reactor building sample station cabinet via an alarm light and the process value recorded on the water chemistry data acquisition system.
Fuel pool high and low level alarms, temperature indication and high temperature alarms are provided on a refueling floor control panel. Level set points are adjustable over the skimmer weir range. In addition, independent instruments provide both fuel pool level and temperature indication on a recorder in the Control Room. The span of these independent instruments bounds the range of the alarm setpoints.
A high rate of leakage through the refueling bellows assemblies, drywell to reactor well seals, or the fuel pool and shipping cask pit double gates is alarmed on a refueling floor control panel.
All local alarms are duplicated individually or as group alarms in the main control room.
Operational Description During normal plant operation, the fuel pools are crosstied to the common shipping cask pit.
The fuel pool cooling pumps circulate the pool water in a closed loop, taking suction from the skimmer surge tank through the heat exchangers and discharging a partial flow through the filter demineralizer, the balance passing through a bypass line back to the fuel pool diffusers.
After the reactor has been shut down, the vessel head and one refueling gate is removed. Two refueling water pumps (see Subsection 9.2.10) transfer condensate from the refueling water storage tank through diffusers into the reactor well and dryer-separator pool. The water level rises from the RPV flange elevation to the fuel pool water level in approximately 4 hr. The second refueling gate is then removed and refueling operations continued.
As the heat load increases with additional spent fuel elements being transferred from the reactor core to the spent fuel pool, additional pumps and heat exchangers of the FPCCS are put into service to meet the design objectives. Part of the cooled water can be diverted to the reactor well through the filling diffusers assisting the RHR system in removing decay heat rising from the core to the water surface. At this time two fuel pool filter demineralizers may be used in conjunction with the reactor water cleanup system to maintain required water quality in the reactor, reactor well, dryer-separator pool, and fuel pool.
Most outages include a period for maintenance on the cooling towers and the service water system. Since the fuel pool cooling heat exchangers are cooled by the service water system, the outage unit's fuel pool cooling system can be operated by circulating river water through the heat exchangers or by using temporary cooling towers/chillers located outside the unit.
Supplemental decay heat removal piping can connect the tube side of the FPCC heat exchangers to either the temporary cooling towers/chillers or to the River Water Makeup line tap provided at the special manhole near North Gate House. During this period of the outage, the operating unit's fuel pool cooling system is also used to cool the pools. This in part, is accomplished by connecting the fuel pools via the cask storage pit. Prior to implementing this method of pool cooling, the shutdown unit's fuel pool temperature is monitored and calculations and tests are performed to assure that the capacity of the operating unit's fuel pool cooling system is adequate to maintain pool temperatures within acceptable limits. Once the shutdown unit's FPCCS becomes available, it is placed back into service.
FSAR Rev. 71 9.1-15
SSES-FSAR Text Rev. 79 After refueling has been completed, the refueling water pumps transfer the water from the reactor well and dryer-separator pool through a condensate demineralizer back to the refueling water storage tank. This is accomplished in approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Gravity draining of the refueling water to the refueling water storage tank is possible.
As the decay heat from the spent fuel decreases with time, the number of operating pumps and heat exchangers may be reduced to keep the fuel pool below the maximum normal design temperature.
The shipping cask storage pit is filled and drained in the same manner as the reactor well with one refueling water transfer pump. The shipping cask storage pit is interconnected with the fuel pool during cask loading operations of spent fuel for offsite disposal. A small stream of fuel pool cooling water may be diverted from the fuel pool cooling pumps to the filling diffuser of the shipping cask pit to remove decay heat and water impurities during cask loading operations.
This water returns over a skimmer weir to the skimmer surge tanks.
During periods when the heat in the pool is greater than the capacity of the fuel pool cooling system (such that acceptable fuel pool temperatures cannot be maintained), e.g., storing of a full core of irradiated fuel shortly after shutdown, the RHR system can be used to dissipate the decay heat. One RHR pump takes suction from an intertie line to the skimmer surge tank and discharges through one RHR heat exchanger to two independent diffusers at the fuel pool bottom. With the Spent Fuel Pool(s) filled to a height approximately 7.5 inches above the weirs, the skimmer surge tank provides sufficient suction head for an RHR pump in the RHR Fuel Pool Cooling (RHRFPC) mode.
Makeup water to replenish evaporative and small leakage losses from the pools is provided from the condensate transfer storage tank into the skimmer surge tank by opening a remote manual valve.
A Seismic Category I line from each of the two emergency service water loops is connected to the RHR intertie diffuser lines of each fuel pool, allowing for emergency makeup in support of RHRFPC or during postulated boiling of the pool water as described in Appendix 9A. The manual supply valves in these emergency makeup lines are accessible from elevations below the refueling floor.
9.1.3.3 Safety Evaluation At FPCCS design conditions where the pool heat load is 12.6 MBTU/HR and service water temperature is 95°F the FPCCS will maintain the fuel pool water less than 125°F. At improved service water temperature conditions, the FPCCS can maintain the fuel pool water less than 125°F with larger heat loads in the pool. This condition occurs during refueling outages. When this condition exists the pool is monitored to assure adequate FPCCS capacity exists. When the FPCCS cannot maintain the pool temperature less than 125°F and low enough so that the calculated time to boil is greater than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />, the RHR system in the Fuel Pool Cooling Mode (RHRFPC) can be connected to the pools to maintain pool temperatures below 125°F by the RHRFPC mode. AT EHL conditions (4.02 x 107 BTU/hr), RHRFPC can maintain the pool temperature below 125°F with spray pond water temperatures below Technical Specification limits. Pool configuration will be maintained during the outage sequence so that the calculated time to boil is greater than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. The exception to this are those periods of time when the spent fuel pools are connected to the reactor cavity and RHR Shutdown Cooling or RHRFPC is FSAR Rev. 71 9.1-16
SSES-FSAR Text Rev. 79 in operation. Under these circumstances Seismic Category I, Class 1E cooling is already in service, thereby eliminating the need to provide sufficient time to place RHR in service to prevent spent fuel pool boiling.
A Seismic Category I makeup is provided by a 2 in. line from each emergency service water (ESW) loop to the RHR fuel pool diffusers, thus providing redundant flow paths from a reliable source of water. The design makeup rate from each ESW loop is based on replenishing the postulated boil-off from the MNHL in each fuel pool for 30 days following the loss of the FPCCS capacity. This provides a capacity far in excess of what would be required by the RHRFPC mode in response to a loss of normal fuel pool cooling due to a seismic event.
All piping and equipment shared with or connecting to the RHR intertie loop are Seismic Category I, Quality Group C, or equivalent and can be isolated from any piping associated with the non-Seismic Category I Quality Group C fuel pool cooling system.
Due to its Non-Seismic Category I, Non-Class 1E power design, the consequences of a seismic event are required to be analyzed for the FPC system. In response to this event, the RHRFPC mode will be used to prevent boiling from occurring; however, a non-mechanistic evaluation of boiling of both spent fuel pools is contained in Appendix 9A in order to conservatively bound the radiological consequences.
The spent fuel pools are normally maintained in a crosstied configuration during dual unit operation and refueling outages. This assures that the time to boil following a loss of normal fuel pool cooling is a minimum of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />; however, in this configuration the time to boil is typically much greater than the minimum 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. The exception to this is shortly after a unit is shutdown for a refueling outage when the spent fuel pools are interconnected with the reactor vessel via the reactor cavity. During this time frame, the time to boil can be less than the 25 hour2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> criteria, due to the combined decay heat load in the spent fuel pools and the reactor vessel. It should be noted that this is largely due to the decay heat resent in the reactor vessel at the time that the spent fuel pools are connected to the reactor vessel. However, a time to boil of less than 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> is not relevant, since RHR Shutdown cooling is required to be in service prior to interconnecting the spent fuel pools with the reactor cavity, thereby eliminating the need to provide sufficient time to place Seismic Category I, Class 1E cooling in service to prevent spent fuel pool boiling. The crosstied configuration allows use of either unit's systems (normal SFP Cooling or RHRFPC) to cool the pools, thus providing fuel pool cooling redundancy.
Crosstied spent fuel pools also provide redundancy for the level instrumentation location in the control room. This instrumentation is designed to operate following an Operating Basis Earthquake under boiling spent fuel pool conditions, and is expected to remain functional.
While not classified as Class 1E equipment, the instruments receive power from independent Class 1E power supplies that are Diesel Generator backed.
Should a seismic event occur during dual unit power operation with crosstied pools, adequate reactor core cooling will be provided and spent fuel pool boiling will be prevented. Only one loop of RHR is necessary to provide long term decay heat removal per reactor vessel. Similarly, only one loop of RHR is necessary to provide long-term decay heat removal to crosstied spent fuel pools. Since either unit's RHR system can provide cooling to both units spent fuel pools with the pools crosstied, a failure of one loop RHR in one of the units would still allow a sufficient number of loops to cool both reactors and the spent fuel pools. In this case, the unit providing spent fuel pool cooling would utilize Alternate Shutdown Cooling for long-term decay heat removal from reactor. The other unit would utilize the normal Shutdown Cooling mode.
FSAR Rev. 71 9.1-17
SSES-FSAR Text Rev. 79 Certain specific plant evolutions, will require the pools to be isolated. These evolutions will be procedurally controlled to ensure that sufficient cooling systems are available given the plant configuration at the time of the evolution.
An evaluation of the impacts of operating the RHRFPC mode on the Ultimate Heat Sink (UHS) was performed as a separate evaluation of the minimum heat transfer case discussed in Subsections 9.2.7.3.1 and 9.2.7.3.6. The results of this evaluation indicate that the Spray Pond (UHS) will be maintained below the design maximum temperature under worst case accident conditions.
Additional details on the design of the RHRFPC mode are provided in Sections 5.4.7.1.1.6, 5.4.7.2.6c, and 9.1.3.1c.
Provisions to minimize and monitor leakage from the fuel pool are described in Subsection 9.1.2.3.
Makeup for evaporative and small leakage losses from the fuel pool is normally supplied from the condensate transfer system to the skimmer surge tanks of each unit. The intermittent flow rate is approximately 50 gpm to each surge tank.
The water level in the spent fuel storage pool is maintained at a height which is sufficient to provide shielding for required building occupancy. Radioactive particulates removed from the fuel pool are collected in filter demineralizer units in shielded cells. For these reasons, the exposure of station personnel to radiation from the spent fuel pool cooling and cleanup system is normally minimal. Further details of radiological considerations are described in Chapter 12.
An evaluation of the radiological effect of a boiling fuel pool is presented in Appendix 9A.
9.1.3.4 Inspection and Testing Requirements No special tests are required because at least one pump, heat exchanger, and filter demineralizer are continuously in operation while fuel is stored in the pool. The remaining components are periodically operated to handle increased heat loads during refueling.
The pool liner leak detection drain valves are periodically opened and the leak rate estimated by the volumetric method. Gas or dye pressure testing from behind the liner plate may be performed to locate a liner plate leak.
Routine visual inspection of the system components, instrumentation, and trouble alarms is provided to verify system operability. Components and piping of the FPCCS designed per ASME Boiler and Pressure Vessel Code,Section III, Class 3 are in-service inspected as described in Section 6.6.
The system was preoperationally tested in accordance with the requirements of Chapter 14.
FSAR Rev. 71 9.1-18
SSES-FSAR Text Rev. 79 9.1.4 FUEL HANDLING SYSTEM AND REACTOR SERVICING EQUIPMENT 9.1.4.1 Design Bases The fuel-handling system is designed to provide a safe and effective means for transporting and handling fuel from the time it reaches the plant until it leaves the plant after post-irradiation cooling. Safe handling of fuel includes design considerations for maintaining occupational radiation exposures as low as practicable during transportation and handling.
Design criteria for major fuel handling system equipment is provided in Tables 9.1-2 through 9.1-4 which list the safety class, quality group, and seismic category. Where applicable, the appropriate ASME, ANSI, Industrial and Electrical Codes are identified. Additional design criteria is shown below and expanded further in Subsection 9.1.4.2.
The transfer of new fuel assemblies between the railroad/truck bay and the new fuel inspection stand and/or the new fuel storage vault is accomplished using the Unit 1 or Unit 2 reactor building cranes or the refueling floor jib cranes equipped with a general purpose grapple, or nylon sling (1" minimum).
The Unit 1 or Unit 2 reactor building crane auxiliary hoist or a refueling floor jib crane is used with a general purpose grapple or nylon sling to transfer new fuel from the fuel inspection stand or the new fuel vault to the fuel pool. From this point on, fuel bundles will be handled by the telescoping grapple on the refueling platform. Individual fuel rods may be handled by the auxiliary hoists on the bridge.
The refueling platform including refueling platform rails, clamps, and clips are Safety Class 2 and Seismic Class 1 from a structural standpoint in accordance with 10 CFR 50, Appendix A and B. Allowable stress due to safe shutdown earthquake loading is 120 percent of yield or 70 percent of ultimate, whichever is least. A dynamic analysis is performed on the structures using the response spectrum method with load contributions being combined by the square root sum of the squares (SRSS) method.
The refueling platform structures are designed in accordance with the AISC Manual of Steel Construction. All parts of the hoist systems are designed to have a safety factor of five based on the ultimate strength of the material. The design of the fuel (main) hoist includes some redundant components such that no single probable event shall result in fuel bundle drop.
Maximum deflection limitations are imposed on the main structures to maintain relative stiffness of the platform. Welding of the platform is in accordance with AWS D14-1 or ASME Boiler and Pressure Vessel Code Section 9. Gears and bearings meet AGMA Gear Classification Manual and ANSI B3.5. Materials used in construction of load bearing members are to ASTM specifications. For personnel safety, OSHA Part 1910-179 is applied. Electrical equipment and controls meet ANSI CI, National Electric Code, and NEMA Publication No. IC1, MG1.
The general purpose grapple and the main telescoping fuel grapple have redundant hooks.
The fuel grapple has an indicator which confirms that the hooks are in the closed position.
The fuel grapple is used for lifting and transporting fuel bundles. It is designed as a telescoping grapple that can extend to the proper work level and in its normal up position state still maintains adequate shielding over fuel.
FSAR Rev. 71 9.1-19
SSES-FSAR Text Rev. 79 To preclude the possibility of raising radioactive material out of the water, the cables on the auxiliary hoists incorporate an adjustable, removable stop that will actuate a limit switch to prevent hoisting when the free end of the cable is at a preset distance below water level. In the event of limit switch failure, the stops are intended to jam the hoist. In addition, redundant electrical interlocks are a part of the grapple.
Provision of a separate cask loading pool, capable of being isolated from the fuel storage pool, will eliminate the potential accident of dropping the cask and rupturing the fuel storage pool.
Refer to Chapter 15 for accident considerations.
Refuel Floor Auxiliary Platform (RFAP) is used to perform ISI work on the Reactor Vessel and its components and to perform other work during refueling outages. RFAP is designed to maintain its structural integrity during and after a dynamic/seismic event. Design/Analysis of RFAP meets the Quality Assurance requirements of 10 CFR 50, Appendix B.
The 360 Degree Refuel Work Platform is another reactor cavity platform which enables personnel to perform ISI Work (and other activities) on the reactor vessel in parallel with fuel movements by the Refueling Platform. The 360 Degree Refuel Work Platform is designed to maintain its structural integrity during and after a SSE. Design/Analysis of the 360 Degree Refuel Work Platform meets the Quality Assurance requirements of 10 CFR 50, Appendix B.
9.1.4.2 System Description Table 9.1-5 is a listing of typical tools and servicing equipment supplied with the nuclear system.
The following paragraphs describe the use of some of the major tools and servicing equipment and address safety aspects of the design where applicable.
Table 9.1-5 also includes a listing of tools provided for the Refuel Floor Wetlift System. The Refuel Floor Wetlift System allows reactor vessel disassembly and reassembly work activities to be performed from the Refueling Platform. The Refuel Floor Wetlift System consists of subsystems, related to MSL Plugs with MSL Plugs Restraint Ring and Rigid Pole Handling System. These subsystems are described in the following Sections.
9.1.4.2.1 Spent Fuel Cask This Section applies to a spent fuel cask used to transfer spent fuel to an off-site storage or reprocessing facility. For the NUHOMS system, the On-Site Transfer Cask used to transfer spent fuel to the on-site Independent Spent Fuel Storage Installation (ISFSI) is described in Section 11.7.6.1. The Holtec equipment used to transfer spent fuel to the on-site ISFSI is described in Sections 11.7.6.3 and 11.7.6.4.
For the NUHOMS system only, the spent fuel cask is used to transfer spent reactor fuel assemblies from the spent fuel pool via the cask pit to a fuel storage or fuel reprocessing facility.
The NUHOMS cask may also be used for offsite shipment of irradiated reactor components such as control rod blades, in-core monitors, etc.
The maximum loaded weight and, hence, the capacity of the NUHOMS and Holtec casks is determined by the 125 tons lifting capacity of a reactor building crane. The maximum loading height, i.e., height of the open cask in the storage pit, is determined by the depth of the shipping cask pit from the gate bottom. This allows for a constant water depth over the fuel in transit from the reactor to the fuel pool and into the shipping cask.
FSAR Rev. 71 9.1-20
SSES-FSAR Text Rev. 79 For the NUHOMS system, the cask is designed to dissipate the maximum allowable heat load from contained irradiated fuel by natural convection at least from the time the cask pit is drained until the cooling system on the transport vehicle is connected.
For the Holtec system, the cask is designed to dissipate the maximum allowable heat load from contained irradiated fuel by natural convection at least from the time the cask water is drained until the cask processing and transport activities are complete.
For the NUHOMS system, the cask further allows underwater replacement of the lid and other operations that may pose unacceptable radiation hazards to personnel. Considerations facilitating decontamination of the cask are given in the design. The design of the NUHOMS cask meets all applicable regulations of the Department of Transportation and 10 CFR 71 with respect to shipping of large quantities of fissile materials. For the Holtec system, the design of the MPC canister meets all applicable regulations of the Department of Transportation and 10 CFR 71 with respect to shipping of large quantities of fissile materials. However, departure of spent fuel from the site utilizing the HI-STORM FW system will be addressed at a later date since there is currently no federal long term repository for storage of spent nuclear fuel.
At present, no specific type of cask has been chosen for off-site storage. Over the lifetime of the plant, several different sizes and models may be used which the fuel handling facilities can accommodate.
9.1.4.2.2 Cask Crane See Subsection 9.1.5 for discussion of reactor building cranes.
9.1.4.2.3 Fuel Servicing Equipment The fuel servicing equipment described below has been designed in accordance with the criteria listed in Table 9.1-2.
9.1.4.2.3.1 Fuel Prep Machine The fuel preparation machine, generally represented in Figure 9.1-9, is mounted on the wall of the fuel pool and can be used for stripping reusable channels from the spent fuel and for channeling or rechanneling of fuel. The machine is also used with the fuel inspection fixture to provide an underwater inspection capability, and with the defective fuel storage container to contain a defective fuel assembly for stripping of the channel.
The fuel preparation machine consists of a work platform, a frame, and a moveable carriage.
The frame and moveable carriage are located below the normal water level in the fuel storage pool, thus providing a water shield for the fuel assemblies being handled. The fuel preparation machine carriage has an up-travel-stop to prevent raising irradiated fuel above the safe water shield level. The up-travel-stop may be relocated for the purpose of new fuel handling in the fuel preparation machine. The moveable carriage is operated by a foot pedal controlled air hoist.
FSAR Rev. 71 9.1-21
SSES-FSAR Text Rev. 79 9.1.4.2.3.2 New Fuel Inspection Stand The new fuel inspection stand, generally represented in Figure 9.1-10, serves as a support for the new fuel bundles undergoing receiving inspection and provides a working platform for technicians engaged in performing the inspection.
The new fuel inspection stand consists of a vertical guide column, a lift unit to position the work platform at any desired level, bearing seats and upper clamps to hold the fuel bundles in position.
This inspection stand has been modified to include a mast guide assembly and jib arm to transfer channels from the new fuel channel up ender onto the fuel bundles.
9.1.4.2.3.3 Channel Bolt Wrench The channel bolt wrench, generally represented in Figure 9.1-11, is a manually operated device approximately 12 feet in overall length. The wrench is used for removing and installing the channel fastener assembly while the fuel assembly is held in the fuel preparation machine.
The channel bolt wrench has a socket which mates and captures the channel fastener capscrew.
9.1.4.2.3.4 Channel Handling Tool The channel handling tool, generally represented in Figure 9.1-12, is used in conjunction with the fuel preparation machine to remove, install, and transport fuel channels in the fuel storage pool.
The tool is composed of a handling bail, a lock/release knob, extension shaft, angle guides, and clamp arms which engage the fuel channel. The clamps are actuated (extended or retracted) by manually rotating lock/release knob.
The channel handling tool is suspended by its bail from a spring balancer on the channel handling boom located on the fuel pool periphery.
9.1.4.2.3.5 Fuel Pool Sipper The fuel pool sipper, generally represented in Figure 9.1-13, provides a means of isolating a fuel assembly to concentrate fission products for detection of defective fuel assemblies.
The fuel sipper head isolates individual fuel assemblies by sealing the top of the fuel channel and pumping water from the bottom of the fuel assembly, through the fuel channel, to a sampling station, and return to the primary coolant system. After a soaking period, a water sample is obtained and is radio-chemically analyzed.
9.1.4.2.3.6 Fuel Inspection Fixture The fuel inspection fixture, generally represented in Figure 9.1-14, is used in conjunction with the fuel preparation machine to permit remote inspection of fuel elements. The fixture consists of two parts: (1) a lower bearing assembly, and (2) a guide assembly at the upper end of the carriage. The fuel inspection fixture permits the rotation of the fuel assembly in the carriage, FSAR Rev. 71 9.1-22
SSES-FSAR Text Rev. 79 and, in conjunction with the vertical movement of the carriage, provides complete access for inspection.
9.1.4.2.3.7 Channel Gauging Fixture The channel gauging fixture, generally represented in Figure 9.1-15, is a go/no-go gauge used to evaluate the condition of a fuel channel, prior to rechanneling or when one is difficult to install.
The channel gauging fixture consists basically of a frame, gauging plate and gauging block.
The gauging plate is shimmed to correspond to the outside dimension of a usable fuel channel.
The gauging block conforms to the inside dimension of lower end of a usable fuel channel.
The channel gauging fixture is installed in the vertical position, between the two fuel preparation machines and hangs from the fuel pool curb.
9.1.4.2.3.8 General Purpose Grapple The general purpose grapple is a handling tool used generally with the fuel. The grapple can be attached to the reactor building auxiliary hoist, jib crane, and the auxiliary hoists on the refueling platforms. The general purpose grapple is used to remove new fuel from the vault, place it in the inspection stand, and transfer it to the fuel pool. It can be used to handle new fuel during channeling. Either a manually operated or air operated general purpose grapple can be used for this fuel handling. The manually operated and air operated general purpose grapples are generally represented in Figures 9.1-16 and 9.1-25, respectively. A nylon sling (1" minimum) may also be used to handle new fuel assemblies.
9.1.4.2.3.9 Fuel Grapple The fuel grapple on the Unit 1 Refueling Platform is a telescopic mast with a grapple head used to lift and orient fuel bundles for core and storage rack placement. The telescopic portion of the mast is made up of cylindrical stainless steel tube assemblies. The outer tube assembly is suspended from the platform at its upper end by means of a pin and hanger joint. The upper end of the inner tube assembly is suspended from the dual-cable of the platform's main hoist.
The grapple head is attached to the lower end of the inner tube assembly and has dual hooks, fail safe (closed) operation and sealed magnetic switches for grapple open and closed indication. The grapple head also has an internally mounted camera that provides the operator with a clear view directly through the open grapple. This allows a close-up verification of serial numbers, orientation, channel fastener condition, seating and grapple alignment.
The fuel grapple on Unit 2 refueling platform is a telescoping mast with a double hook grapple head used to lift and orient fuel bundles for core and storage rack placement. It is a triangular, open sectioned mast constructed of tubular stainless steel.
Unit 2 Fuel Grapple Mast section-to-section guidance is provided by nylon bearing pads.
Vertical motion is supplied by a dual wire rope cable hoist, which provides a redundant load path, and is mounted on the Refueling Platform Main Trolley. Hoist cable attachment to the inner-most grapple section is achieved through a rocker arm/clevis assembly which allows for load equalization in the hoist wire ropes. A redundant hook grapple head featuring individual hook engage switches and air cylinders consists of engage switches wired in series and interlocked with the main hoist load cell in a manner to prevent raising a fuel bundle with either hook disengaged. Figure 9.1-23 outlines the main fuel grapple.
FSAR Rev. 71 9.1-23
SSES-FSAR Text Rev. 79 9.1.4.2.3.10 Fuel Transfer Stand The fuel transfer stand, generally represented in Figure 9.1-24, is a passive device constructed using aluminum structural members. The transfer stand provides three positions for landing metal fuel shipping containers in a near vertical position to facilitate lifting the bundles without tilting more than 5 degrees from vertical. The transfer stand is equipped with ladders to allow easy access by refueling personnel for rigging the fuel bundles for removal from the shipping containers.
9.1.4.2.3.11 New Fuel Channel Up Ender The new fuel channel up ender, generally represented in Figure 9.1-26, serves as a hydraulic holder to upend each fuel channel from a horizontal to a vertical position in preparation for channeling of new fuel.
9.1.4.2.3.12 New Fuel Up Ending Stand The new fuel up ending stand, generally represented in Figure 9.1-27, serves as a work platform for refueling personnel and a frame for metal fuel shipping containers to be placed in a near vertical position. The stand is equipped with numerous platform levels to allow easy access by refueling personnel for rigging the fuel bundles for removal from the shipping containers.
9.1.4.2.4 Servicing Aids General area underwater lights are provided with a suitable reflector for illumination. Suitable light support brackets are furnished to support the lights in the reactor vessel to allow the light to be positioned over the area being serviced independent of the platform. Local area underwater lights and drop lights are used for illumination where needed.
Portable underwater closed circuit television cameras are provided. The cameras may be lowered into the reactor vessel and/or fuel pool to assist in the inspection and/or maintenance of these areas.
A general purpose, plastic viewing aid is provided to float on the water surface to provide better visibility. The viewing aid contains colored components, or is appropriately marked, to allow the operator to observe it in the event of filling with water and sinking. Portable, submersible type, underwater vacuum cleaners are provided to assist in removing crud and miscellaneous particulate matter from the pool floors, or the reactor vessel. The pump and the filter unit are completely submersible for extended periods. The filter "package" is capable of being remotely changed, and the filters will fit into a standard shipping container for off-site burial. Fuel pool tool accessories are also provided to meet servicing requirements.
9.1.4.2.5 Reactor Vessel Servicing Equipment The essentiality and safety classifications, the quality group, and the seismic category for this equipment are listed in Table 9.1-3. Following is a description of the equipment designs in reference to that table.
FSAR Rev. 71 9.1-24
SSES-FSAR Text Rev. 79 9.1.4.2.5.1 Reactor Vessel Service Tools These tools are used when the reactor is shut down and the reactor vessel head is being removed or reinstalled. Tools in this group are:
Stud Handling Tool Stud Wrench Nut Runner Stud Thread Protector Thread Protector Mandrel Seal Surface Protector Stud Elongation Measuring Rod Dial Indicator Elongation Measuring Device Head Guide Cap These tools are designed for a 40 year life in the specified environment. Lifting tools are designed for a safety factor of 5 or better with respect to the ultimate strength of the material used. When carbon steel is used, it is either hard chrome plated, parkerized, or coated with an acceptable paint.
9.1.4.2.5.2 Steam Line Plug 9.1.4.2.5.2.1 Steam Line Plug (REM*Light Model)
The steam line plugs are used during reactor refueling or servicing; they are inserted in the steam outlet nozzles from inside of the reactor vessel to prevent a flow of water from the reactor well into the main steam lines during servicing of safety relief valves, main isolation valves, or other components of the main steam lines, while the reactor water level is raised to the refueling level. The main steam line plugs are designed to withstand a differential pressure of 60 psid without ejecting from the main steam line nozzle, and have been tested to hold at 90 psid. The plug is equipped with a wire rope tether designed to limit the distance the plug can fall and prevent the plug from reaching the core support top guide.
The steam line plug design provides two seals of different types. Each one is independently capable of holding full head pressure. The equipment is constructed of non-corrosive materials.
The plug body is designed in accordance with the "Aluminum Design Manual" by the Aluminum Association.
9.1.4.2.5.2.2 Main Steam Line (MSL) Plugs (Spring Disk Model) and MSL Plugs Restraint Ring [Refuel Floor Wetlift System]
The Main Steam Line (MSL) Plugs are used during the reactor refueling or servicing; they are inserted in the steam outlet nozzles from inside the reactor vessel to prevent flow of water from the reactor into the main steam lines during servicing of the safety relief valves, main isolation valves or other components of the main steam lines, while Reactor Cavity is flooded. The MSL plugs are designed to withstand a differential pressure of 50 psi without ejecting from the main steam line nozzle. The MSL Plugs Restraint Ring provides a backup mechanical means to prevent ejection of the MSL Plugs.
FSAR Rev. 71 9.1-25
SSES-FSAR Text Rev. 79 9.1.4.2.5.3 Shroud Head Bolt Wrench 9.1.4.2.5.3.1 Shroud Head Bolt Wrench (Supplied with Nuclear System)
This is a hand held tool for operation of shroud head bolts. It is designed for a 40-year life, it is made of aluminum to be easy to handle and to resist corrosion. Testing has been performed to confirm the design.
9.1.4.2.5.3.2 Shroud Head Bolt Wrench (Refuel Floor Wetlift System)
This tool is used for operation of the shroud head bolts. The shroud head bolt wrench is attached to the rigid poles supplied with the Rigid Pole Handling System and can be manipulated from either the Rigid Pole Handling System hoist, or the Reactor Building Cranes' auxiliary hoists, or the Refueling Platforms auxiliary hoists.
9.1.4.2.5.3.3 Shroud Head Bolt Wrench (Lightweight supplied by Vendor)
This is a long extension wrench that is used to remotely loosen or tighten the shroud head bolts from the Refueling Platform. The wrench is operated manually and is supported on a bracket which is mounted on the Refueling Platform handrails. It is made of aluminum to be easy to handle and to resists corrosion. Testing has been performed to confirm the design.
9.1.4.2.5.3.4 Shroud Head Bolt Wrench Type HTC (Supplied by Vendor)
This tool is a long extension wrench that is used for remotely loosening shroud head bolts that are stuck during the removal of the shroud head / steam separator. The wrench is not used to tighten the shroud head bolts to avoid over torquing the shroud head bolts. The wrench is operated manually and is hoisted and moved via the Reactor Building crane. It is made of stainless steel to resist corrosion. The stainless steel assembly allows a maximum of 500 lb-ft to be applied to it. Testing has been performed to confirm the design.
9.1.4.2.5.4 Head Holding Pedestal Three pedestals are provided for mounting on the refueling floor for supporting the reactor vessel head. The pedestals have studs which engage three evenly spaced stud holes in the head flange. The flange surface rests on replaceable wear pads made of aluminum. When resting on the pedestals, the head flange is approximately 3 feet above the floor to allow access to the seal surface for inspection and O-ring replacement.
The pedestal structure is a carbon steel weldment, coated with an approved paint. It has a base with bolt holes for mounting it to the concrete floor. The structure is designed in accordance with "The Manual of Steel Construction" by AISC.
9.1.4.2.5.5 Head Nut and Washer Rack The RPV head nut and washer rack is used for transporting and storing up to 6 nuts and washers. The rack is a box shaped aluminum structure with dividers to provide individual compartments for each nut and washer. Each corner has a lug and shackle for attaching a 4-leg lifting sling.
FSAR Rev. 71 9.1-26
SSES-FSAR Text Rev. 79 The rack is designed in accordance with the "Aluminum Construction Manual" by the Aluminum Association, and for a safety factor of 5.
9.1.4.2.5.6 Head Stud Rack The head stud rack is used for transporting and storage of up to 6 reactor pressure vessel studs. It is suspended from the auxiliary building crane hook when lifting studs from the reactor well to the operating floor.
The rack is made of aluminum to resist corrosion.
9.1.4.2.5.7 Dryer and Separator Strongback The dryer and separator strongback is a lifting device used for transporting the steam dryer or the shroud head with the steam separators between the reactor vessel and the storage pools.
The strongback consists of a cruciform shaped structure which is suspended from a hook pin assembly. On the end of each arm of the cruciform is a socket with a pneumatically operated pin for engaging the four lift eyes on the steam dryer or shroud head. Also, it is used for installation and removal of the MSL Plugs Restraint Ring when required.
9.1.4.2.5.8 RPV Head Strongback-Carousel and Adapter The RPV Head Strongback-Carousel is an integrated piece of equipment consisting of a cruciform-shaped strongback, a circular monorail, and a circular storage tray. It is designed to perform functions formerly performed separately by Head Nut and Washer Racks, Head Strongback, and Stud Tensioner Monorail.
The strongback is a box beam structure with four (4) arms and a hook box in the center. An adapter is used as an interface between the hook box and the Reactor Building crane sister hook. Two (2) hook pins are used for hook box to adapter engagement, and two (2) more hook pins are used for adapter to crane sister hook engagement. On each of the four (4) arms, the strongback has a lift rod for engagement to the lift lugs on the RPV head. The circular monorail is mounted on the extensions of the strongback arms and four (4) additional arms equally spaced between the strongback arms. The monorail circle matches the stud circle of the reactor vessel and serves to suspend stud tensioners. The storage tray is suspended from the ends of the same eight (8) arms and surrounds the RPV head flange. It has provision for storage of nuts, washers and stud thread protectors.
All steel structure is designed in accordance with "The Manual of Steel Construction" by AISC or Crane Manufacturers Association of America (CMAA) Specification No. 70 as applicable. The welding is in accordance with the ASME Boiler and Pressure Vessel Code Section IX. A safety factor of 5 or greater with reference to the ultimate strength is used for the design. In addition, the strongback is designed such that one leg of the cruciform will support the maximum applied load without exceeding the stresses allowed by the specifications noted above (AISC or CMAA as applicable). The strongback lifting assembly is proof load tested to 125 (+/-2%) percent of the rated load. After the load test, all structural welds are magnetic particle tested. The aluminum structure for the nut rack is designed in accordance with the "Aluminum Construction Manual" by the Aluminum Association and for a minimum safety factor of 5 with reference to the ultimate strength.
FSAR Rev. 71 9.1-27
SSES-FSAR Text Rev. 79 9.1.4.2.5.9 Service Platform The service platform is not used and has been eliminated.
The service platform is non-Seismic Class I equipment, and it has been designed for 0.75 g horizontal and 0.00 g vertical. The physical size of the device is such that it cannot enter the reactor pressure vessel.
9.1.4.2.5.10 Service Platform Support The service platform support is not used and has been eliminated. A segmented aluminum seal surface protector supplied by CBI Nuclear serves as a protector for the reactor vessel flange.
9.1.4.2.5.11 Steam Line Plug Installation Tool 9.1.4.2.5.11.1 Steam Line Plug Installation Tool [REM*Light Model]
The main steam line plug with its integral installation tool assembly are remotely installed one at a time from the refueling platform using the monorail auxiliary hoist and the jet pump grapple.
The installation tool assembly, which remains in place, acts as a redundant restraint preventing the plug from inadvertently exiting the nozzle.
9.1.4.2.5.11.2 MSL Plugs Installation and Removal (I/R) Tool (Spring Disk Model) [Refuel Floor Wetlift System]
The MSL Plugs I/R Tool is attached to the rigid poles and suspended from the hoist of the Rigid Pole Handling System or the Reactor Building Crane auxiliary hoist for transporting, installing and removing the MSL Plugs in the steam line nozzles of the reactor vessel. The MSL Plugs I/R Tool can handle two MSL Plugs at a time and is designed to a factor of safety of 5 or greater in reference to the ultimate tensile strength of its materials.
9.1.4.2.5.12 Refuel Fuel Floor Auxiliary Platform (RFAP)
The Refuel Pool Auxiliary Platform is an electrically operated with manual backup, lightweight work platform. This is an engineered tool and it is operated on the existing refueling platform rails. It is a fully engineered clear span bridge for use as a portable work platform for personnel access over the reactor cavity, the spent fuel storage pools and the equipment pools.
9.1.4.2.5.13 360 Degree Refuel Work Platform The 360 Degree Refuel Work Platform is a steel structure platform with a trough (partially submerged in the reactor cavity) to allow personnel occupancy while fuel moves occur with the Refueling Platform overhead. It is engineered tool which is lifted in and out of the reactor cavity and its support feet rest on the cavity shield block support ledge. Designated removable work carriages and a jib hoist can be attached to the trough and used for vessel servicing/inspections.
FSAR Rev. 71 9.1-28
SSES-FSAR Text Rev. 79 9.1.4.2.5.14 Jet Pump Plugs The Jet Pump Plugs are used during reactor refueling or servicing; they are inserted in the Jet Pump nozzles from inside the reactor vessel to prevent flow of water from the reactor well into the Reactor Recirculation discharge lines. The Jet Pump Plugs are designed to withstand a differential pressure of 100 psi without ejecting. The Jet Pump Plugs are dynamically qualified to maintain their integrity during a seismic event.
9.1.4.2.6 In-Vessel Servicing Equipment The multiple LPRM strongback attached to the reactor building crane auxiliary hoist is used to support up to eight LPRM assemblies during transfer to the reactor vessel. The LPRM assemblies are then transferred one at a time to the instrument handling tool for insertion into the reactor vessel. Water Seal Caps are installed under vessel prior to removal of an LPRM assembly to prevent loss of reactor coolant water during LPRM replacement.
The instrument strongback attached to the Reactor Building crane auxiliary hoist is used for servicing neutron monitor dry tubes should they require replacement. The strongback supports the dry tube during transfer to the vessel. The in-core dry tube is then decoupled from the strongback and is guided into place while being supported by the Instrument Handling Tool.
Final in-core insertion is accomplished from below the reactor vessel. The instrument handling tool is attached to the refueling platform auxiliary hoist and is used for removing and installing fixed in-core dry tubes as well as handling neutrons source holders and the Source Range Monitor/Intermediate Range Monitor dry tubes.
Each in-core instrumentation guide tube is sealed by an O-ring on the flange and in the event that the seal needs replacing, an in-core guide tube sealing tool is provided. The tool is inserted into an empty guide tube and sits on the beveled guide tube entry in the vessel. When the drain on the Water Seal Cap is opened, hydrostatic pressure seats the tool. The flange can then be removed for seal replacement.
The auxiliary hoist on the refueling platform is used with appropriate grapples to handle control rods, flux monitor dry tubes, sources, and other internals of the reactor.
9.1.4.2.7 Refueling Equipment Fuel movement and reactor servicing operations are performed from a platform which spans the refueling, servicing, and storage cavities.
Following description of refueling platform applies to both Unit 1 and Unit 2 except where indicated otherwise.
Either platform can be operated over either unit's reactor cavity. However, Unit 1 Refueling Platform is usually used for refueling of both units since this platform has enhanced operational capabilities.
9.1.4.2.7.1 Refueling Platform The refueling platform is a gantry crane which is used to transport fuel and reactor components to and from pool storage and the reactor vessel (including pool to pool and core to core movements). The platform spans the fuel storage and vessel pools on rails bedded in the FSAR Rev. 71 9.1-29
SSES-FSAR Text Rev. 79 refueling floor. A telescoping mast and grapple suspended from a trolley system is used to transport and orient fuel bundles, for core, storage rack, or shipping cask placement. Control of the platform is from an operator station on the main trolley with a position indicating system provided to position the grapple over core locations. The platform control system includes indication to verify hook closed and grapple load and interlocks to prevent unsafe operation over the vessel during control rod movements, and limit vertical travel of the grapple. Two 1000-pound capacity auxiliary hoists, one main trolley mounted and one auxiliary trolley mounted, are provided for servicing such as LPRM replacement, fuel support replacement, jet pump servicing, and control rod replacement. The grapple in its normal up position provides about 8 feet 6 inches minimum water shielding over the active fuel of the bundle during transit.
The Unit 1 Refueling Platform has a computer based control system. This provides the capability to operate in Manual Control Mode, Semi-Automatic Mode, or Automatic Mode.
The system also provides Boundary Zone Integration/Checking capability thereby preventing Refueling Mast interference with reactor vessel walls, reactor cavity walls, reactor core shroud, fuel pool walls, cattle chute, cask loading area, control rod brackets, or any other equipment programmed into the exclusion zone. All Boundary Zone controls and interlocks are active during automatic, semi-automatic and manual operations.
When using the opposite units refueling platform on the refuel unit for fuel handling activities (U1 platform refueling U2 reactor and vice versa), the refuel units idle platform may be powered from an alternate source which does not have the RMCS refuel interlock interface. When powered from the alternate source the refuel units platform becomes an auxiliary work platform over the dryer-separator storage pool or reactor vessel. In this configuration, the Main Hoist on this work platform will be in a stowed position and therefore physically disabled from handling fuel. The Auxiliary Hoists (i.e., Frame and Monorail Hoists) on the work platform will be administratively controlled from operation in the vessel if the Steam Separator is removed. In addition to the RMCS refueling interlocks, any boundary zone or travel interlocks may also be defeated for the platform functioning as an auxiliary work platform.
The Refuel Floor Wetlift System's Rigid Pole Handling System is mounted on the monorail portion of either units Refueling Platforms and is a general purpose underwater pole system for use in the Refuel Floor Pools and Reactor Cavities. The Rigid Pole Handling System is based on a sectionally assembled, long handle tool for reactor vessel servicing that consists of a hoist, assembly work station, pole storage station, pole section and tools. The Rigid Pole Handling System provides the means for unlatching and latching of the Separator and supports the installation and removal of the MSL Plugs from the Refueling Platform with the Reactor Cavity flooded. The Rigid Pole Handling System is designed for a 2000 pounds load rating. The Rigid Pole Handling System is designed so that the assembly and its components retain structural integrity during seismic events to prevent its collapse into the Reactor Cavity or Refuel Floor Pools or dropping of suspended load. The Rigid Pole Handling System is designed to a factor of safety of 10 or greater in reference to the ultimate tensile strength of its materials.
9.1.4.2.8 Storage Equipment Specially designed equipment storage racks are provided. Additional storage equipment is listed on Table 9.1-5. For fuel storage racks description and fuel arrangement, see Subsections 9.1.1 and 9.1.2.
FSAR Rev. 71 9.1-30
SSES-FSAR Text Rev. 79 Defective fuel rods may be stored in a fuel rod storage basket. The storage basket has physical dimensions similar to a fuel assembly, stainless steel tubes for storing individual fuel rods, and a handle for transporting. The fuel rod basket storage is stored in the spent fuel storage pool racks.
Defective fuel assemblies are placed in defective fuel storage containers, as necessary. These containers are stored in the multi-purpose storage container which is a part of the high density spent fuel racks.
Defective fuel storage containers can be picked up and moved with a fuel bundle in them.
Channels can also be removed from the fuel bundle while in a defective fuel storage container.
Defective individual fuel rods may be stored in a failed fuel rod holder.
The Fuel Pool Sipper may be used for out-of-core wet sipping at any time. They are used to detect a defective fuel bundle. The containers cannot be used for transporting a fuel bundle.
The bail on the container head is designed not to fit into the fuel grapple.
9.1.4.2.9 Under Reactor Vessel Servicing Equipment The primary functions of the under reactor vessel servicing equipment are to: (1) remove and install control rod drives, (2) service thermal sleeve and control rod guide tube, (3) install and remove the neutron detectors. Table 9.1-4 lists the equipment and tools required for servicing.
The control rod drive handling system (CRDHS) is a pneumatically powered device, designed for use in the under vessel gallery for the purpose of exchanging control rod drives. The capabilities of this system include: transport of the control rod drive (CRD) into the under vessel gallery, erection to the vertical position and installation of the CRD, removal of a CRD and reposition to horizontal, and transport out of the under vessel gallery. The system will also accept a spud end shield sleeve to be installed directly after withdrawal of the CRD from the housing. The system consists of two primary parts, the carriage and the winch cart. A pneumatic control stand is separate from these two parts and is attached only by hoses to the other two parts. No electrical systems are incorporated into this design.
The winch cart is the heart of the motive forces of the system. Twin stainless steel aircraft cables are wound on and off twin cable drums. The cable drums are driven by an air motor through a self-locking worm gear box. These cables perform two functions: (1) raise/lower the elevator and (2) rotate the carriage from horizontal to vertical and back. The two cables provide a redundant safety feature; either cable can safely support the CRD in the unlikely event of a cable failure. The cables are retained in the grooves of the drums by the pressure of spring-loaded rollers. This prevents the cables from disengaging from the drums should they become slack. Thus the lifting components are equipped with adequate features to prevent uncontrolled movement upon loss of air or component failure.
The CRDHS was designed as non-safety related equipment in accordance with NES, inc.
Quality Assurance requirements as spelled out in NES Proposal 8660-283, dated 2/4/87.
There were no specific industry codes or standards identified in design and construction of this system.
FSAR Rev. 71 9.1-31
SSES-FSAR Text Rev. 79 The equipment handling platform is powered electrically and provides a working surface for equipment and personnel performing work in the under vessel area. It is a polar platform capable of 360° rotation. This equipment is designed in accordance with the applicable requirements of OSHA (Vol. 37, No. 202, Part 191 ON), AISC, ANSI-C-1, (National Electric Code).
The thermal sleeve installation tool locks, unlocks, and lowers the thermal sleeve from the control rod drive guide tube. The key bender is designed to install and remove the anti-rotation key that is used on the thermal sleeve.
The in-core flange seal test plug is used to determine the pressure integrity of the in-core flange O-ring seal. It is constructed of non-corrosive material.
9.1.4.2.10 Fuel Transfer Description 9.1.4.2.10.1 Arrival of Fuel on Site New fuel arrives in the railway bay of the reactor building Unit 1 either by railcar or truck or the reactor building Unit 2 by truck in the truck bay. The access doors are closed to maintain the secondary containment as required by Technical Specifications. Unloading of the shipping containers is done by the auxiliary hoist of the Unit 1 or Unit 2 reactor building cranes.
9.1.4.2.10.2 Refueling Procedure Fuel handling procedures are described below and shown visually in Figures 9.1-20, 9.1-21, and 9.1-22. The Refueling Floor Layout is shown on Drawing C-1807, Shts. 1 & 2, and component drawings of the principal fuel handling equipment are shown in Figures 9.1-9 through 9.1-16 and Figure 9.1-23.
The fuel handling process takes place primarily on the refueling floor above the reactor. The principal locations and equipment are shown on Drawing C-1807, Shts. 1 & 2. The reactor, fuel pool, and shipping cask pool are connected to each other by transfer slots.
The reactor cavity/fuel pool slot is opened to allow for vessel refueling. The fuel pool/cask storage pit slot is normally open to allow transfer of spent fuel bundles and other irradiated components from the fuel pool to a shipping cask or to the other fuel pool. At other times, the fuel pool slot is closed to hydraulically isolate the fuel pools from the reactor cavity via watertight gates. Additionally, radiological shielding from the spent fuel may be provided via the installation of removable blocks in the slots.
The handling of new fuel on the refueling floor is illustrated in Figure 9.1-20. The transfer of the bundles between the shipping container (C) and the new fuel inspection stand (D) and/or the new fuel storage vault (E) is accomplished using 5-ton auxiliary hoist of the Unit 1 or Unit 2 reactor building crane or a half-ton floor mounted refueling jib crane equipped with a general-purpose grapple or nylon sling. The fuel bundle cannot be handled horizontally without support, so the shipping container is placed in an almost vertical position before being opened. The container is opened, and the bundles removed in a vertical position.
FSAR Rev. 71 9.1-32
SSES-FSAR Text Rev. 79 The auxiliary hoist of the Unit 1 or Unit 2 reactor building crane or the jib crane are also used with a general-purpose grapple or nylon sling to transfer new fuel from the new fuel vault or inspection stand to the fuel pool. From this point on, fuel bundles are handled by the telescoping grapple on the refueling platform.
The storage racks in both the vault and the fuel pool hold the fuel bundles or assemblies in a vertical position.
The new fuel inspection stand holds one or two bundles in vertical position. The Inspector(s) ride up and down on a platform, and the bundles are manually rotated on their axes. Thus the inspectors can see all visible surfaces on the bundles. The general-purpose grapples and the fuel grapple of the refueling platform have redundant hooks, and an indicator which confirms that the grapple hooks are in the closed position.
The refueling platform uses a grapple on a telescoping mast for lifting and transporting fuel bundles or assemblies. The telescoping mast can extend to the proper work level; and, in its normal up position state, maintains adequate water shielding over the fuel being handled.
The reactor refueling procedure is shown schematically in Figure 9.1-21. The refueling platform (G) moves into position over a fuel assembly, lowers the grapple on the telescoping mast (H) and engages the bail on a fuel assembly. The assembly is lifted clear of the interferences and moved into position over its new location. The mast then lowers the assembly into the selected location and the grapple releases the bail. This same process is repeated to complete all required fuel moves. Fuel assemblies can be moved in this manner from one core (J) location to another if an in-core shuffle is being performed, between the core and a fuel pool location (F, K) or between the Unit 1 and Unit 2 fuel pools.
If the fuel assembly is to be dechanneled from the fuel prep machine, an operator, using a long-handled wrench, removes the screw(s) and springs from the top of the channel. The channel is then held, while a carriage lowers the fuel bundle out of the channel. The channel is then moved aside and stored. If required, the refueling platform grapple can carry the unchanneled bundle and place it in a storage rack. The channel handling boom hoist, (L) moves the channel to storage, if appropriate.
A channel rack is conveniently located near to the fuel prep machines, for storage of channels.
Refer to Section 9.1.4.2.10.2.12 for discussion on new fuel channeling.
To preclude the possibility of raising radioactive material out of the water, redundant electrical limit switches are incorporated in the auxiliary hoists of the refueling platform and the jib crane hoist, and interlocked to prevent hoisting above the preset limit. In addition, the cables on the hoists incorporate adjustable stops that are intended to jam the hoist in the event of limit switch failure.
When spent fuel is to be shipped, it is placed in a cask, as shown in Figure 9.1-22. The refueling platform grapples a fuel bundle from the storage rack in the fuel pools, lifts it, carries it through slot (B) into the shipping cask pool, and lowers it into the cask, (M). When the cask is loaded, the Unit 1 single failure proof reactor building crane sets the cask cover (N) on the cask.
After partially draining the shipping cask pool, the cask is lifted onto the refueling floor, decontaminated and lowered through the open hatchways, (P), onto the truck or railcar in the railway bay at grade level by the Unit 1 single failure proof reactor building crane only.
FSAR Rev. 71 9.1-33
SSES-FSAR Text Rev. 79 Provision of a separate cask loading pool, capable of being isolated from the fuel storage pool, eliminates the potential accident of dropping the cask and rupturing the fuel storage pool.
Additional detailed information is provided below.
9.1.4.2.10.2.1 New Fuel Preparation 9.1.4.2.10.2.1.1 Receipt and Inspection of New Fuel The incoming new fuel will be delivered to the site. The shipping containers should be unloaded from the transport vehicle and examined for damage during shipment. Each outer shipping container contains two fuel bundles supported by an inner metal container. The metal inner containers are removed from the outer shipping containers. The metal inner containers are then moved to the reactor building where they are lifted to the refueling floor. Both inner and outer shipping containers are reusable. Lifting of the metal inner containers to the refuel floor is to be accomplished by use of the reactor building crane extending down from the refueling floor through the appropriate equipment hatch.
9.1.4.2.10.2.1.2 Channeling New Fuel Typically new fuel is channeled in the inspection stand. If desired, new fuel can be channeled in the fuel pool using the fuel prep machine. Also, prior to shipment, fuel assemblies may have been channeled by the fuel manufacturer. The channeled new fuel is then stored in the new fuel vault or in the pool storage racks ready for insertion into the reactor.
9.1.4.2.10.2.1.3 Equipment Preparation Prior to use for refueling, all equipment must be placed in readiness. All tools, grapples, slings, strongbacks, stud tensioners, etc., should be given a thorough check and any defective (or well worn) parts should be replaced. Air hoses on grapples should be routinely leak tested. Crane cables should be routinely inspected. All necessary maintenance and interlock checks should be performed to assure no extended outage due to equipment failure.
The in-core flux monitors, in their shipping container, should be on the refueling floor. The channeled new fuel and the replacement control rods should be ready.
9.1.4.2.10.2.2 Reactor Shutdown The evolutions and sequence for reactor shutdown are controlled via appropriate plant procedure(s). Generally, the reactor sequence is as follows (Note: The following evolutions can be performed as required (i.e., series or parallel evolutions) as directed via the appropriate procedures):
a) Reactor shutdown with all control rods inserted (controlled shutdown and/or scram).
b) De-inert the Drywell and Suppression Chamber.
c) Remove reactor well shield plugs (2 layers) when plant conditions allow. (The layers do not have to be removed at the same time). The eight reactor well shield plugs are removed via the reactor building crane.
FSAR Rev. 71 9.1-34
SSES-FSAR Text Rev. 79 This operation can be immediately followed by removal of the three canal plugs and the three slot plugs. Thus, a total of 14 separate plugs must be removed and placed on the refueling floor. Refer to Drawing C-1807, Shts. 1 & 2 for placement of these plugs on the refueling floor.
9.1.4.2.10.2.2.1 Drywell Head Removal Immediately after removal of the reactor well shield plugs, the work to unbolt the drywell head can begin. The drywell head is attached by removable bolts protruding from the lower drywell flange. The nuts on top are merely loosened and the bolt heads swing outward. The bolts are then pulled upwards and supported with the nuts on a slotted lip of the head.
The main hoist hook of the Unit 1 or Unit 2 reactor building crane is attached to the hook box on top of the unbolted drywell head which is lifted to its appointed storage space on the refueling floor.
9.1.4.2.10.2.2.2 Reactor Well Servicing When the drywell head has been removed, an array of piping is exposed that must be serviced.
Various vent piping penetrations through the reactor well must be removed and the penetrations made water tight.
Vessel head piping and head insulation must be removed and transported to storage on the refueling floor. Water level in the reactor vessel is now brought to flange level in preparation for head removal.
9.1.4.2.10.2.3 Reactor Vessel Opening 9.1.4.2.10.2.3.1 Vessel Head Removal 9.1.4.2.10.2.3.1.1 Vessel Head Removal Using RPV Head Strongback/Carousel The RPV head strongback/carousel is transported by the Unit 1 or Unit 2 reactor building crane and positioned on the reactor vessel head. Each stud is detensioned and its nut loosened.
When the nuts are loose, they are backed off using a nut runner. The nuts and washers are placed in the integral nut rack on the strongback/carousel. With the nuts and washers removed, the vessel stud protectors and vessel head guide caps are installed. The four to six studs in line with the fuel transfer canal are removed from the vessel flange. The removed studs may be placed in the racks and transported to the refueling floor for storage. As an alternative, they may be stored in the vessel head and transported with the head.
The Unit 1 reactor building crane is used to lift the head and transport it to the head holding pedestals on the refueling floor. The head holding pedestals keep the vessel head elevated to facilitate inspection and "0" ring replacement.
FSAR Rev. 71 9.1-35
SSES-FSAR Text Rev. 79 9.1.4.2.10.2.3.2 Dryer Removal 9.1.4.2.10.2.3.2.1 Dryer Removal Dryer Removal Using Dryer and Separator Strongback With the Reactor Cavity flooded, or while flooding the Reactor Cavity, the Dryer and Separator Strongback is lowered by the Unit 1 and Unit 2 Reactor Building Crane and attached to the Dryer lifting lugs. The Dryer is lifted from the reactor vessel and transported underwater to its storage location in the Dryer-Separator Storage Pool adjacent to the Reactor Cavity by the corresponding units Reactor Building Crane.
9.1.4.2.10.2.3.3 Separator Removal 9.1.4.2.10.2.3.3.1 Separator Removal Using Dryer and Separator Strongback From the Refueling Platform, using the lightweight shroud head bolt wrench mounted on the handrail bracket, or shroud head bolt wrenches suspended from the Rigid Pole Handling Systems hoist or Reactor Building Cranes main or auxiliary hoists, the Separator is unlatched.
With the reactor Cavity flooded, the Dryer and Separator Strongback is lowered by the Unit 1 and Unit 2 Reactor Building Crane and attached to the Separator lifting lugs. The Separator is lifted from the reactor vessel and transported underwater to its storage location in the Dryer-Separator Storage Pool adjacent to the Reactor Cavity by the corresponding units Reactor Building Crane.
9.1.4.2.10.2.3.4 Fuel Bundle Sampling During reactor operation, the core off-gas radiation level is monitored. If an excessive rise in off-gas activity has been noted, the suspect assemblies can be sampled during shutdown to locate any leaking fuel assemblies. The fuel sampler or sipper rests on the channels of a four bundle array in the core. An air bubble is pumped into the top of the 4 fuel bundles and allowed to stay about 10 minutes. This stops water circulation through the bundles and allows fission products to concentrate if a bundle is defective. After 10 minutes, a water sample is taken for fission product analysis. If a defective bundle is found, it is taken to the fuel pool and if required, may be stored in a special defective fuel storage container to prevent the spread of contamination in the pool. Alternately, suspect assemblies can be sipped using the fuel pool sipper described in Section 9.1.4.2.3.5.
9.1.4.2.10.2.4 Refueling and Reactor Servicing The two fuel pool gates isolating the fuel pool from the reactor well are now removed thereby interconnecting the fuel pool, the reactor well, and the dryer-separator storage pool. The gate strong back is attached to the gate lifting lugs and the Unit 1 or Unit 2 reactor building crane lifts the gate and places it on the fuel pool gate storage lugs. The actual refueling of the reactor can now begin.
FSAR Rev. 71 9.1-36
SSES-FSAR Text Rev. 79 9.1.4.2.10.2.4.1 Refueling The actual fuel handling is done with the fuel grapple which is an integral part of the refueling platform. The platform runs on rails over the fuel pool and the reactor well. In addition to the fuel grapple, the refueling platform is equipped with two auxiliary hoists which can be used with various grapples to service other reactor internals.
To move fuel in the core, the fuel grapple is aligned over the fuel assembly, lowered and attached to the fuel bundle bail. The fuel bundle is raised out of the core, moved to another core location or moved through the refueling slot to the fuel pool, positioned over the storage rack and lowered to storage. Fuel is moved from the storage pool to the reactor vessel in the same manner. Although described sequentially, simultaneous movement in three axes is not prohibited (Unit 1 only).
9.1.4.2.10.2.5 Vessel Closure The evolutions and sequence for reactor reassembly are controlled by appropriate plant procedure(s). Generally, the sequence is as follows.
Vessel closure is performed with the equipment described above. The service platform supplied with the Nuclear System is not used and has been eliminated.
The following steps are a typical sequence that will return the reactor to operating condition.
The procedures are the reverse of those described in the preceding Sections. Many steps are performed in parallel.
a) Core verification. The core position of each fuel assembly must be verified to assure the desired core configuration has been attained.
b) Control rod drive tests. The control rod drive timing, friction and scram tests are performed.
c) Remove the MSL Plugs Restraint Ring.
d) Replace Separator with corresponding unit's Reactor Building Crane and latch Separator using Rigid Pole Handling System and a Refueling Platform or Reactor Building Crane hoist.
e) Remove MSL Plugs from reactor vessel using Rigid Pole Handling System.
f) Replace Dryer with corresponding unit's Reactor Building Crane.
g) Close Fuel Pool gates.
h) Drain Reactor Cavity and Dryer-Separator Storage Pool.
i) Decontaminate Reactor Cavity.
j) Remove seal surface protector from the vessel flange and remove refueling shield (cattle chute).
FSAR Rev. 71 9.1-37
SSES-FSAR Text Rev. 79 k) Decontaminate dryer-separator storage pool.
l) Replace vessel studs.
m) Install reactor vessel head with Unit 1 reactor building crane.
n) Install vessel head piping and insulation.
o) Replace equipment pool and refueling slot shield plugs.
p) Hydro-test vessel, if necessary.
q) Open drywell vents, install vent piping.
r) Install drywell head with Unit 1 or Unit 2 reactor building cranes.
s) Inert reactor drywell and suppression chamber.
t) Install reactor well shield plugs with Unit 1 or Unit 2 Reactor Building Cranes.
u) Startup tests. The reactor is returned to full power operation. Power is increased gradually in a series of steps until the reactor is operating at rated power. At specific steps during the approach to power, the in-core flux monitors are calibrated.
9.1.4.2.10.3 Departure of Spent Fuel from Site This Section describes the departure of spent fuel from the site for the NUHOMS Dry Fuel storage system. For the Holtec Dry Fuel Storage System, departure of spent fuel from the site utilizing the HI-STORM FW system will be addressed at a later date since there is currently no federal long term repository for storage of spent nuclear fuel. Therefore, the below statements for off-site shipping and transportation only apply to the NUHOMS system.
Departure of the spent fuel stored at the Independent Spent Fuel Storage Installation (ISFSI) from the site will require that the Dry Shielded Canister (DSC) be extracted from the ISFSI Horizontal Storage Module (HSM) and returned to the Reactor Building Spent Fuel Storage Pools. Subsequent departure of this spent fuel from the site is described below.
The spent fuel shipping cask arrives by railcar or truck in the railway bay of the reactor building Unit 1. It is lifted from there by the main hoist 125 ton hook of the Unit 1 single failure proof reactor building crane through the floor openings to the refueling floor and placed into the empty shipping cask pit between the fuel pools of the Unit 1 and Unit 2 Reactor Building.
The cask outside is decontaminated from road dirt and the lid removed by the Unit 1 single failure proof reactor building crane. If the cask pit gates are installed and the cask pit is not filled, one of the inner gates of the shipping cask pit is removed. After filling of the shipping cask pool, the second gate to one of the fuel pools is removed and loading of the cask with irradiated fuel commences. The refueling platform is used to transfer fuel bundles of sufficiently low decay heat level from the spent fuel storage racks underwater into the shipping cask.
FSAR Rev. 71 9.1-38
SSES-FSAR Text Rev. 79 Following replacement of the cask lid, the gates to the fuel pool are inserted, the shipping cask pit drained and the cask outside decontaminated. The Unit 1 single failure proof reactor building crane then transfers the cask from the storage pit onto the shipping vehicle where a cooling system dissipates the remaining decay heat of the fuel during transport.
9.1.4.3 Safety Evaluation 9.1.4.3.1 Spent Fuel Cask This Section applies to a spent fuel cask used to transfer spent fuel to an off-site storage or reprocessing facility. For the NUHOMS system, the On-Site Transfer Cask used to transfer spent fuel to the on-site Independent Spent Fuel Storage Installation (ISFSI) is described in Section 11.7.6.1. The Holtec equipment used to transfer spent fuel to the on-site ISFSI is described in Sections 11.7.6.3 and 11.7.6.4.
The NUHOMS spent fuel cask is equipped with dual sets of lifting lugs and yokes compatible with the Unit 1 reactor building crane redundant main hook, thus preventing a cask drop due to a single failure. An analysis of the spent fuel cask drop is therefore not required.
The Holtec MPC is equipped with lifting lugs and yokes which meet the requirements of single failure proof and are compatible with the Unit 1 Reactor Building Crane redundant main hook, thus preventing an MPC drop due to a single failure. An analysis of the MPC drop is therefore not required.
9.1.4.3.2 Reactor Building Crane See Subsection 9.1.5.3 for the reactor building crane safety evaluation.
9.1.4.3.3 Fuel Servicing Equipment Failure of any fuel servicing equipment listed in Table 9.1-2 poses no hazard beyond the effect of the refueling accident analyzed in Chapter 15.
Safety aspects (evaluation) of the fuel servicing equipment are discussed in Subsection 9.1.4.2.3.
9.1.4.3.4 Servicing Aids The small manual devices listed in Table 9.1-5 facilitate underwater viewing and handling of fuel. Failure of any servicing aid does not pose any hazard beyond the effect of the refueling accident.
9.1.4.3.5 Reactor Vessel Servicing Equipment The effects of postulated load drops of the steam dryer, steam separator and the vessel head have been analyzed in accordance with NUREG 0612. These analyses are documented in calculations which have been performed as part of the SSES Heavy Loads Program. The structural effects of the limiting postulated drops along the load paths utilized for these lifts were found to be within defined acceptance limits. The results of these evaluations conclude that the 4 general criteria identified in Section 5.1 of NUREG 0612 are satisfied. Use of the FSAR Rev. 71 9.1-39
SSES-FSAR Text Rev. 79 dryer/separator strongback, the vessel head strongback, and the head strongback/carousel is therefore in compliance with the NUREG.
The reactor vessel heads are transported as equivalent single failure proof lifts using the Unit 1 Reactor Building single failure proof crane. Refer to FSAR Section 9.1.6.4.2 The Refuel Floor Auxiliary Platform and the 360 Degree Refuel Work Platform are designed to preclude them from becoming Safety Impact Items. Lifting of the RFAP and the 360 Degree Refuel Work Platform to the Refuel Floor and on the Refuel Floor will be in accordance with the SSES Heavy Loads Program. Use of the jib hoist on the 360 Degree Refuel Platform is limited to loads less than 1000 pounds (not a Heavy Load) and does not pose any hazard beyond the effect of the refueling accident.
9.1.4.3.6 In-Vessel Servicing Equipment Failure of any in-vessel servicing equipment listed in Table 9.1-5 poses no hazard beyond the effect of the refueling accident analyzed in Chapter 15.
9.1.4.3.7 Refueling Equipment The most severe failure of the refueling platform and associated grapple and hoists results in the dropping of a fuel assembly onto the reactor core. This refueling accident is analyzed in Chapter 15.
Safety aspects of the refueling equipment are discussed in Subsection 9.1.4.2.7. A description of fuel transfer, including appropriate safety features, is provided in Subsection 9.1.4.2.10. In addition, the following summary safety evaluation of the fuel handling system is provided below.
The fuel prep machine can be used to remove and install channels with all parts remaining under water. Mechanical stops prevent the carriage from lifting an irradiated fuel bundle or assembly to a height where water shielding is less than 7 feet. Irradiated channels, as well as small parts such as bolts and springs, are stored underwater. The spaces in the channel storage rack have center posts which prevent the loading of fuel bundles into this rack.
There are no nuclear safety problems associated with the handling of new fuel bundles, singly or in pairs. Equipment and procedures prevent an accumulation of more than two bundles in any location.
The refueling platform and the Refuel Floor Wetlift System's Rigid Pole Handling System are designed to prevent them from toppling into the pools during a SSE.
The grapple utilized for fuel movement is on the end of a telescoping mast. At retraction of the mast to its normal hoist position for fuel transport, the top of the active fuel of the bundle is about 8.5' below the water surface, which provides water shielding consistent with radiation zone designations established in Chapter 12. The grapple is hoisted by redundant cables inside of the mast; and is lowered by gravity. A digital readout is displayed to the operator, showing him the exact coordinates of the grapple over the core.
The mast is suspended and gimbaled from the trolley, near its top, so that the mast can be swung about the axis of platform travel, in order to remove the grapple from the water for servicing and for storage.
FSAR Rev. 71 9.1-40
SSES-FSAR Text Rev. 79 The grapple has dual hooks designed for fail safe (closed) operation with hook closed position indicated to the operator. The mechanical design of the grapple hook is such that grapple disengagement is prevented until the fuel assembly is seated.
In addition to the main hoist on the trolley, there is an auxiliary hoist on the trolley, and another hoist on its own monorail. These three hoists are precluded from operating simultaneously, because control power is available to only one of them at a time.
The Refuel Floor Wetlift System's Rigid Pole Handling System hoist's power is supplied from the Refueling Platform's Monorail Auxiliary hoist. These two hoists can be operated at the same time if required.
The two auxiliary hoists have mechanically actuated electrical limit switches, which are set to maintain water shielding over irradiated loads consistent with radiation zone designations established in FSAR Chapter 12. Adjustable mechanical stops attached to the hoist cables are intended to jam the hoist in the event of limit switch failure.
In summary, the fuel handling system complies with Regulatory Guide 1.13 (12/75), General Design Criteria 2, 3, 4, 5, 61, 62, and 63, and applicable portions of 10 CFR 50.
A system-level, qualitative-type failure mode and effects analysis relative to this system is discussed in Subsection 15A.6.5.
9.1.4.3.8 Storage Equipment The safety evaluation of the new and spent fuel storage is presented in Subsections 9.1.1.3 and 9.1.2.3.
9.1.4.3.9 Under Reactor Vessel Servicing Equipment Failure of any under reactor vessel servicing equipment poses no hazard in excess of the effects of accidents analyzed in Chapter 15.
9.1.4.4 Inspection and Testing Requirements 9.1.4.4.1 Inspection Much of the refueling and servicing equipment is subject to the strict controls of quality assurance, incorporating the requirements of federal regulation 10 CFR 50, Appendix B.
Components defined as essential to safety, such as the fuel storage racks and refueling platform have an additional set of engineering specified, "quality requirements" that identify safety-related features which require specific QA verification of compliance to drawing requirements.
Prior to shipment, all quality requirements are reviewed by QA personnel and combined into a summary product quality checklist. The quality checklist provides confirmation of the quality requirements for each product.
FSAR Rev. 71 9.1-41
SSES-FSAR Text Rev. 79 9.1.4.4.2 Testing Prior to multi-unit fabrication, major pieces of refueling or servicing equipment are fabricated and tested as prototype units. These units are tested to specifications defined by the responsible design engineer and implemented by a test engineering organization. In many cases, a full design review of the product is conducted before and after the testing cycle.
Any design changes affecting function, that are made after the design review of the qualification testing has been completed, are reverified by test or calculation.
When the unit is received at the site, it is inspected by quality assurance personnel to ensure that no damage has occurred during transit or storage. Prior to site operation, the refueling or servicing equipment must undergo a sequence of preoperational functional tests, as defined by a site preoperational test specification.
Fuel handling and vessel servicing equipment was preoperationally tested in accordance with Chapter 14.
Tools and servicing equipment used for refueling are inspected and preoperationally performance tested prior to use.
9.1.4.5 Instrumentation Requirements The majority of the refueling and servicing equipment is manually operated and controlled by the operator's visual observations. This type of operation does not necessitate the need for a dynamic instrumentation system.
However, there are several components that are essential to prudent operation that do have instrumentation and control systems.
9.1.4.5.1 Refueling Platform The refueling platform has a non-safety related X-Y-Z position indicator system that informs the operator which core fuel cell the fuel grapple is accessing. Refueling Platform interlocks and a control room operator monitoring fuel moves are used to prevent the fuel grapple from operating in a fuel cell where the control rod is not fully inserted. Refer to Subsection 7.6.1.1 for discussion of refueling interlocks.
Additionally, there are a series of mechanically activated switches that provide indications on the operator's console for grapple limits, hoist and cable load conditions, and confirmation that the grapple's hook is either open or closed.
A series of load cells are installed which provide interlocks that restrict hoist movements whenever threshold limits are exceeded on either the fuel grapple or the auxiliary hoist units and the Rigid Pole Handling System hoist.
9.1.4.5.2 Fuel Support Grapple Although the Fuel Support Grapple is not essential to safety, it has an instrumentation system consisting of mechanical switches and indicator lights. This system provides the operator with a positive indication that the grapple is properly aligned and oriented and that the grappling FSAR Rev. 71 9.1-42
SSES-FSAR Text Rev. 79 mechanism is either extended or retracted. The control rod/fuel support piece (CR/FSP) combination grapple is similar to the fuel support grapple, with the exception that it uses mechanical indicators.
9.1.4.5.3 Other Refer to Table 9.1-5 for additional refueling and servicing equipment not requiring instrumentation.
9.1.4.5.4 Radiation Monitoring The area radiation monitoring equipment for the refueling area is described in Subsection 12.3.4.
9.1.5 REACTOR BUILDING CRANES Two reactor building cranes are provided for the Susquehanna SES. Unit 1 crane is a single failure-proof crane and is designed to handle the spent fuel cask as well as refueling and vessel service load requirements for Unit 1 or Unit 2. The Unit 2 crane is not single failure-proof and is designed to handle construction loads as well as refueling and vessel service load requirements for Unit 1 or Unit 2 except the reactor vessel heads, spent fuel casks, NUHOMS On-Site Transfer Cask and HI-TRAC VW Transfer Cask.
The Unit 2 reactor building crane, rated 125 tons (main hoist), 5 tons (auxiliary hoist) is capable of carrying any loads within its rated capacity. Load carrying over or within restricted areas of the refueling floor must be in accordance with the limits established for those areas. (Reference Drawing C-1807, Shts. 1 & 2).
Administrative controls are used to preclude the Unit 2 reactor building crane from being used for handling the reactor vessel heads and the spent fuel casks, NUHOMS On-Site Transfer Cask and HI-TRAC VW Transfer Cask when stored in the spent fuel shipping cask storage pit.
9.1.5.1 Design Bases The main purpose of the Unit 1 single failure proof reactor building crane is to handle the reactor vessel heads, the spent fuel casks, NUHOMS On-Site Transfer Cask and HI-TRAC VW Transfer Cask between the cask transport vehicles, the cask storage pit, and the wash-down area in the reactor building. Secondary purposes of the Unit 1 and Unit 2 reactor building cranes include:
a) Handling loads related to maintenance and replacement of equipment from the reactor building which are received or shipped through the railroad or truck bay access doors for either Unit 1 or Unit 2.
b) Handling of shield plugs, drywell heads, steam dryer and separator, etc., during refueling operations.
FSAR Rev. 71 9.1-43
SSES-FSAR Text Rev. 79 The reactor building crane is designed for the following ratings:
Main Hoist Capacity 125 tons Auxiliary Hoist Capacity 5 tons Speed of Main Hoist (Critical - at rated load) 5 fpm Speed of Main Hoist (Non-Critical) 7 fpm (see Note 1)
Speed of Auxiliary Hoist (Critical - at rated load) 20 fpm Speed of Auxiliary Hoist (Non-Critical) 28 fpm (see Note 1)
Speed of Trolley (Critical) 30 fpm Speed of Trolley (Non-Critical) 50 fpm Speed of Bridge 50 fpm Lift of Main Hook 173 ft (min.) (see Note 2)
Lift of Auxiliary Hook 173 ft (min.)
Crane Span 130 ft (approx.)
Length of Runway (between stops) 323 ft (approx.)
Uncontrolled Drop Main Hoist 0.5 in. (max.)
Auxiliary Hoist 8.55 in. (max.)
Note 1: The non-critical mode of operation provides the additional speed range of 5 to 7 fpm with administratively controlled load limits of 40 tons for the main hoist and 20 to 28 fpm and 2 tons for the auxiliary hoist.
Note 2: Unit 2 reactor building crane ratings are identical to those of the Unit 1 crane, except for the main hook lift, which is 68 ft. This, in addition to administrative controls, precludes inadvertent use of the Unit 2 crane for spent fuel cask handling, since the main hook cannot lower the spent fuel cask to the truck bay on Elevation 670.
The auxiliary hooks of both cranes are designed for use underwater, up to 50 ft. depth.
9.1.5.2 Equipment Design a) General The Unit 1 and Unit 2 reactor building cranes are designed, fabricated, installed, and tested in accordance with ANSI B30.2.0, CMAA-70, and OSHA regulations.
b) Structural The structural portions of the crane bridge and trolley are designed for (1) dead load plus rated lift load plus impact load of 15 percent of the total dead plus rated live loads, not to exceed allowable stresses; (2) dead load plus rated lift load plus a lateral load of 10 percent of the total dead plus rated live loads, not to exceed allowable stresses; (3) the operating basis earthquake (OBE) while lifting the rated load, the working stresses not to exceed 125 percent of the allowable stress; (4) the design basis earthquake (DBE) while lifting the rated load, the allowable stresses to be less than 90 percent in bending, 85 percent in axial tension, and 50 percent in shear of the material minimum yield stresses; (5) a tornado loading of 300 psf, without live load, the allowable stresses to be the same as for (4) above.
FSAR Rev. 71 9.1-44
SSES-FSAR Text Rev. 79 The structure of the crane bridge consists of welded box type girders with truck saddles and truck frames of welded steel construction. The trolley side frames, sheave frames, and truck frames are of structural steelwelded construction.
c) Mechanical-General The Unit 1 and Unit 2 reactor building cranes are of a single trolley top running electric overhead travelling bridge design.
d) Mechanical-Unit 1 The Unit 1 reactor building crane main hoist is provided with the following dual components preventing a single failure to result in a drop of the spent fuel shipping cask:
- 1) Dual sister hook (hook within a hook).
- 2) Dual reeving systems complete with redundant wire ropes, upper, lower, and equalizing sheaves.
- 3) Dual main hoist gear boxes with individual braking systems.
Each wire rope has a safety factor of five against breaking while lifting the rated capacity. In case of failure of one of the two reeving systems, the dynamic load transfer to the other system will not cause the rope load to exceed one-third of the rope breaking strength.
The following holding brakes are provided:
Main hoist Three, rated for 150 percent of the motor torque, with provision for manual operation to allow lowering of the load after a power failure.
Trolley Two, one rated at 100% and the other rated 150% of the trolley drive motor torque at the point of application.
Bridge One, rated for 100 percent of motor torque for each of the two bridge motors.
All holding brakes are ac or dc magnet operated and must be energized to release them.
e) Controls-Unit 1 Trolley AC Adjustable frequency scalar control (open loop) with inherent reversing plugging control.
Bridge Adjustable frequency vector control (closed loop) with inherent reversing plugging control.
Hoists AC Adjustable frequency vector control (closed loop) including dynamic braking, with minimum speed of less than 0.1% of rated speed.
Operation of the crane is from the bridge mounted cab or floor. The floor operation is by radio control. Control at any one time is from one point only.
FSAR Rev. 71 9.1-45
SSES-FSAR Text Rev. 79 f) Mechanical-Unit 2 The Unit 2 reactor building crane is provided with the following components and is not designed to any single failure criteria:
- 1) Single Hook.
- 2) One set of 6 part double reeving system composed of a single wire rope, upper and lower sheaves.
- 3) Single main hoist gear box with single braking system composed of a primary brake and a time delayed secondary brake.
The single wire rope has a safety factor of five against breaking while lifting the rated capacity. The wire rope type used for Unit 2 will be identical to that used for Unit 1.
The following holding brakes are provided:
Main Hoist Two brakes rated for 150 percent of the motor torque, with provision for manual operation to allow lowering of the load after a power failure.
Trolley Two, one rated at 100% and the other rated at 50% of the trolley drive motor torque at the point of application.
Bridge One brake rated for 100 percent of motor torque for each of the two bridge motors.
All holding brakes are AC or DC magnet operated and must be energized to release them.
g) Controls-Unit 2 Trolley AC Adjustable frequency scalar control (open loop) with inherent reversing plugging control.
Bridge Adjustable frequency vector control (closed loop) with inherent reversing plugging control.
Hoists AC Adjustable frequency vector control (closed loop) including Dynamic braking, with minimum speed of less than 0.1% of rated speed.
Operation of the crane is from the bridge mounted cab or floor. The floor operation is by radio. Control at any one time is from one point only.
9.1.5.3 Safety Evaluation As described in Subsection 9.1.5.2, the Unit 1 Reactor Building Crane main hoist is provided with dual main hoist components capable of holding the load in the event of a single failure.
An overspeed switch activating all spring set motor brakes in the lowering direction holds the load in suspension in the Critical and Non-Critical modes of operation.
FSAR Rev. 71 9.1-46
SSES-FSAR Text Rev. 79 See Section 3.13 for discussion of compliance with Regulatory Guides 1.104 and 1.13.
See Appendix 9B for a discussion of compliance with BTP ASB9-1.
A dillon load switch is provided under the equalizer to prevent the Unit 1 Reactor Building Crane from lifting loads in excess of 110% of its rated capacity.
The Unit 1 and Unit 2 reactor building cranes are provided with limit switches to prevent overtravel of the bridge and trolley and stop the main and auxiliary hooks in their highest and lowest safe positions.
Two limit switches, each of different design, are provided to limit the upward movement of the main and auxiliary hoist.
Two geared limit switches are provided for the main hoist, and one for the auxiliary hoist to limit the downward movement of the respective hoists except that the Unit 2 reactor building crane has only one geared limit switch to limit downward movement of the main hoist and auxiliary hoist.
When the 125-ton hook is not in the parked upper position, movement of the crane bridge and/or trolley will be stopped when entering the Zone B restricted areas shown on Drawing C-1807, Shts. 1 & 2. It can be used in the Zone A areas of Drawing C-1807, Shts. 1 & 2 when the key-locked zone bypass switch is used in accordance with administrative guidance.
The trolley and bridge motion of the crane is equipped with non-contact distance sensor(s).
Position information is obtained via measurement of the speed of light from the sensor head to a reflector and back to the sensor head. This information is serially input to a Programmable Logic Controller which in turn prevents entry into a controlled zone. The respective motion will be permitted to reverse out of the restricted zone.
Administrative controls prevent placing the hoist in the upper parked position with a load suspended.
A key locked bypass switch is provided in the cab, with an additional selector switch on the radio controller that is disabled or enabled by a locked bypass switch in the cab, to allow the use of the main hoist over the RPV area for handling shield plugs, RPV and drywell heads, steam dryer/separator, etc.
Crane motor overload protection is provided by an electrical cut-out on the hoist drive motor.
The results of a failure mode and effect analysis are presented in Tables 9.1-6a and 9.1-6b for the Unit 1 and Unit 2 reactor building cranes.
The Unit 1 and Unit 2 reactor building cranes are safety related and a quality assurance program has been established and implemented on their design, fabrication, erection, and testing.
FSAR Rev. 71 9.1-47
SSES-FSAR Text Rev. 79 The Unit 1 and Unit 2 reactor building cranes are designed to remain on the runway in a parked and restrained position (by tornado locks) with no load attached under the following tornado wind loadings:
a) 300 psf on the windward crane girder b) +/-200 psf on the leeward crane girder The Unit 1 and Unit 2 reactor building crane mechanical and structural components are qualified to Seismic Category I requirements. They may become and remain inoperative after the operating basis earthquake, but no parts or the load will dislodge or fall. Manual lowering of the main hoist load is provided.
9.1.5.4 Inspection and Testing Requirements The following Unit 1 and Unit 2 reactor building crane components tests were performed during the crane fabrication as follows:
a) Each hook: Ultrasonic tests 200 percent load test followed by dimensional check Dry powder magnetic particle test.
b) Wire rope: Rope sample destructive breaking test.
c) Gears, gear pinions, swivels, Magnetic particle tests.
load block frames, hook trunnions, seismic restraints, and tornado locks:
d) Major structural welds: 100 percent magnetic particle testing.
The crane hoists, trolley, and bridge drives were operated in the shop to demonstrate their operability and the trolley tracking.
After the Unit 1 and Unit 2 Reactor Building Cranes were erected, they were thoroughly tested, including the crane rating test in accordance with ANSI B30.2.0.
After the Unit 1 and Unit 2 Reactor Building Crane retrofit all motions were thoroughly tested, including the crane rating test in accordance with ANSI B30.2. NOTE: The following exception applies: The load test was limited to the range of motion available in the access hatch.
The Unit 1 and Unit 2 Reactor Building crane periodic operational tests are performed in accordance with applicable OSHA regulations, local codes, and ANSI B30.2.0.
9.1.5.5 Instrumentation Requirements The Unit 1 and Unit 2 Reactor Building cranes are furnished with devices and controls, as described in Subsection 9.1.5.3. The Unit 1 Reactor Building crane has dual devices and controls to prevent or detect a single crane failure and thus preclude dropping of the spent fuel cask.
FSAR Rev. 71 9.1-48
SSES-FSAR Text Rev. 79 9.1.6 Control of Heavy Loads 9.1.6.1 Introduction / Licensing Background Heavy loads are typically defined as loads that weigh more than 1000 pounds in the vicinity of spent fuel and/or safety related equipment. The transport of heavy loads during maintenance activities and other evolutions is controlled to minimize risk.
The Heavy Loads Program ensures the safe handling of heavy loads at the Susquehanna Steam Electric Station (SSES). PPL is committed to NUREG 0612 Section 5.1.1 requirements (Phase I) for all areas of SSES where damage to safety related equipment is possible during the transfer of heavy loads. For SSES Reactor Pressure Vessel (RPV) disassembly and reassembly and for the movement of the NUHOMS On-Site Transfer Cask and spent fuel cask, PPL is committed to NUREG 0612 Sections 5.1.1 through 5.1.6 (Phase I and Phase II) requirements.
Based on the favorable implemented actions of licensees during Phase I effort, the NRC withdrew the requirement to complete Phase II in Generic Letter 85-11 and encouraged licensees to implement safety significant actions they deemed appropriate. PPL continued with its commitment to comply with the Phase II requirements of NUREG 0612 for RPV disassembly and reassembly and for the movement of the NUHOMS On-Site Transfer Cask and spent fuel casks and Holtec HI-TRAC VW Transfer Cask and MPC.
NRC Bulletin 96-02 dealt with the movement of heavy loads over spent fuel, over fuel in the RPV, and over safety related equipment. PPL found the approach used to meet the Phase I requirements and the voluntary Phase II requirements for heavy loads is in accordance with the regulatory guidelines for the scope of NRC Bulletin 96-02. At SSES, the transfer of heavy loads immediately adjacent to or over irradiated fuel in the RPVs or spent fuel pools is generally prohibited; however there are a few exceptions that are simply unavoidable during RPV disassembly and reassembly. PPL performed load drop analysis calculations which demonstrated the acceptability of these exceptions.
Additional detail on the Heavy Loads Program is available in the referenced drawings, procedures, industry standards, specifications, NRC guidance documents, licensing documents, letters, etc. provided in FSAR Section 9.1.7.
9.1.6.2 Safety Basis This section describes the safety basis that ensures that the risk associated with load handling failures is acceptably low.
9.1.6.2.1 The risk associated with load handling failures is acceptably low based on meeting the Phase I requirements of NUREG 0612 (Section 5.1.1) for the transport of all heavy loads.
9.1.6.2.2 The risk associated with handling of heavy loads for RPV disassembly and reassembly is acceptable since SSES meets the more stringent requirements of Phase II of NUREG 0612 (Sections 5.1.2 through 5.1.6). The RPV heads are transported as equivalent single failure FSAR Rev. 71 9.1-49
SSES-FSAR Text Rev. 79 proof lifts using the Unit 1 Reactor Building single failure proof crane. Load drop analyses exist for the transport of other heavy load associated with RPV disassembly and reassembly.
Technical Requirements for Operation exist to control radiation releases while transferring heavy loads over irradiated fuel.
9.1.6.2.3 The NUHOMS On-Site Transfer Cask, which is used to transfer spent fuel to the Independent Spent Fuel Storage Installation, is single failure proof and is transported as a single failure proof lift using the Unit 1 Reactor Building single failure proof crane. The HI-TRAC VW Transfer Cask, which is used to transfer spent fuel to the Railroad Bay, is single failure proof and is transported as a single failure proof lift using the Unit 1 Reactor Building single failure proof crane. Refer to the discussion presented in FSAR Section 15.7.5.
9.1.6.3 Scope of Heavy Load Handling Systems The overhead cranes, monorails, and jib cranes included in the Heavy Loads Program are described in Specification M-1435.
9.1.6.4 Control of Heavy Loads Program The control of Heavy Loads Program consists of 1) PPLs commitment to NUREG-0612, Phase I elements, 2) PPLs commitment to NUREG-0612, Phase II elements for RPV disassembly and reassembly, and 3) NUHOMS On-Site Transfer Cask, HI-TRAC VW Transfer Cask and spent fuel cask lifts are performed as single failure proof lifts.
9.1.6.4.1 Commitments in Response to NUREG 0612, Phase I Elements PPL is committed to complying with NUREG 0612 Phase I requirements. The seven elements associated with the Phase I requirements are incorporated into procedure NDAP-QA-0505 and are discussed in this section.
9.1.6.4.1.1 Safe Load Paths Lifts associated with refuel floor activities have safe load paths presented on safe load path drawings. The drawings are based on heavy load drop analysis calculations. Per procedure NDAP-QA-0505, load paths shall be predetermined and discussed with work groups to minimize the height and length of travel.
9.1.6.4.1.2 Load Handling Procedures Procedure NDAP-QA-0505 provides the controls for the implementation of the Heavy Loads Program. All heavy loads shall be lifted in accordance with a plant approved procedure or by an approved Heavy Loads Worksheet.
Prior to moving a heavy load in the proximity of irradiated fuel, the completion of a checklist is required to assure that the necessary plant systems are operable to mitigate and control radiological consequences associated with a postulated drop accident.
FSAR Rev. 71 9.1-50
SSES-FSAR Text Rev. 79 9.1.6.4.1.3 Crane Operator Qualifications Rigger, Lifting Equipment Inspector, and Crane Operator qualification requirements are provided in procedure NDAP-QA-0505 and other station procedures.
9.1.6.4.1.4 Special Lifting Devices To comply with Phase I requirements of NUREG 0612, procedure NDAP-QA-0505 implements industry standard ANSI N14.6 1978, American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kg) or More for Nuclear Materials.
9.1.6.4.1.5 Other Lifting Devices To comply with Phase I requirements of NUREG 0612, procedure NDAP-QA-0505 implements industry standard ASME B30.9 - 2003, Slings for lifting devices that are not specially designed.
9.1.6.4.1.6 Inspection and Testing Inspection and testing requirements are presented in procedure NDAP-QA-0505. Non-installed lifting equipment (strongbacks, lifting beams, boxes/containers, chain hoists, come-alongs and other engineered components) shall be inspected, load tested and documented per the Work Order Program in accordance with station procedures. Load tests are performed on the Reactor Building cranes as specified in the ASME B30.2-2005, Overhead and Gantry Cranes.
Load tests were performed initially (when the crane was new). In addition, load tests are required when modifications are performed on the load carrying components to the crane.
9.1.6.4.1.7 Crane Design To comply with Phase I requirements of NUREG 0612, procedure NDAP-QA-0505 implements industry standard ASME B30.2 - 2005, Overhead and Gantry Cranes. Cranes are designed to Crane Manufacturers Association of America (CMAA) specification #70 and #74 depending on the type of crane.
9.1.6.4.2 Reactor Pressure Vessel Head Lifting Procedures PPL is committed to complying with NUREG 0612 Phase II requirements for RPV disassembly and reassembly.
9.1.6.4.2.1 The transport of the RPV heads is controlled by RPV disassembly and reassembly procedures.
These lifts are performed as equivalent single failure proof lifts using the Unit 1 Reactor Building single failure proof crane. These lifts are classified as equivalent single failure proof lifts rather than single failure proof lifts. The load transfer elements (adapter box, strong back, and lifting lugs) below the crane hook meet NUREG 0612 Phase I requirements and not NUREG 0612 Phase II requirements. The guidance for equivalent single failure proof lifts is provided in Section 3.2 of NEI 08-05 which states The lifting devices used below the hook to make the reactor head lift are required to meet the Phase I requirements as delineated in NUREG-0612, Section 5.1.1.(4).
FSAR Rev. 71 9.1-51
SSES-FSAR Text Rev. 79 9.1.6.4.2.2 Other heavy load lifts associated with RPV disassembly and reassembly meet NUREG-0612 Phase II requirements. The acceptability for performing the vast majority of these lifts has been demonstrated by performing load drop analyses. The load drop analyses resulted in restrictions on load height, load weight, load transfer boundaries, and medium under the loads. The restrictions are presented on the safe load path drawings. The Unit 1 Reactor Building single failure proof crane cannot be used exclusively due to the presence of the Unit 2 Reactor Building crane, which utilizes the same rails, and a structural requirement to keep the cranes a certain distance apart.
9.1.6.4.3 Single Failure Proof Crane for the NUHOMS On-Site Transfer Cask and HI-TRAC VW Transfer Cask As discussed in FSAR Sections 9.1.4.3.1, 11.7.10, and 15.7.5, the NUHOMS On-Site Transfer Cask, which is used to transfer spent fuel to the Independent Spent Fuel Storage Installation, is transferred as a single failure proof lift using the Unit 1 Reactor Building single failure proof crane. The NUHOMS On-Site Transfer Cask is not lifted over the spent fuel pool and is transported in accordance with safe load path drawings and station procedures. As discussed in FSAR Sections 9.1.4.3.1, 11.7.10, and 15.7.5, the HI-TRAC VW Transfer Cask, which is used to transfer spent fuel to the Railroad Bay, is transferred as a single failure proof lift using the Unit 1 Reactor Building single failure proof crane. The HI-TRAC VW Transfer Cask is not lifted over the spent fuel pool and is transported in accordance with safe load path drawings and station procedures.
9.1.6.5 Safety Evaluation The controls implemented by NUREG 0612 Phase I elements and the controls implemented by NUREG 0612 Phase II elements, for maintenance activities involving RPV disassembly and reassembly, make the risk of a load drop very unlikely.
Load drop analyses, performed for heavy load transfers associated with RPV disassembly and reassembly, have demonstrated that the consequences of postulated load drops are acceptable. Restrictions on load height, load weight, load transfer boundaries, and medium under the load are reflected in plant procedures and safe load path drawings.
The most critical lifts of RPV heads, the NUHOMS On-Site Transfer Cask, HI-TRAC VW Transfer Cask and spent fuel casks are performed as equivalent single failure proof lifts and single failure proof lifts using the Unit 1 Reactor Building single failure proof crane. This approach makes the risk of a load drop extremely unlikely and acceptably low.
FSAR Rev. 71 9.1-52
SSES-FSAR Text Rev. 79 9.1.7 References 9.1-1 Specification M-1435, General Specification for Heavy Loads Review 9.1-2 Procedure NDAP-QA-0505, Crane, Hoist, and Rigging Program 9.1-3 Procedure NDAP-QA-0507, Conduct of Refuel Floor 9.1-4 Procedure NDAP-QA-0653, Medical Testing Requirement For Other Than Licensed Operator (RO/SRO) Regulated Positions 9.1-5 Procedure MT-GM-014, Rigging and Lifting Inspection 9.1-6 Procedure NTP-QA-46.1, Susquehanna Crane Operator Certification Training Program 9.1-7 Procedure ME-0RF-023, Dry Fuel Storage 61BT Dry Shielded Canister 9.1-8 Procedure ME-0RF-179, Dry Fuel Storage Equipment List and Reference Information 9.1-9 "Design and Fabrication Criteria Spent Fuel Storage Racks for Susquehanna Steam Electric Station," PARSP/3157, P. 7-1 and Appendix I, Revision 6, Programmed and Remote Systems Corp., April 6, 1979.
9.1-10 Drawing C-2090 Reactor Building Units 1 and 2 Safe Load Paths for Transfer of Heavy Loads 9.1-11 Drawing C-2586 Unit 2 Refuel Floor Laydown Plan and Safe Load Paths 9.1-12 Drawing C-2592 Unit 1 Refuel Floor Laydown Plan and Safe Load Paths 9.1-13 NUREG 0612, Control of Heavy Loads at Nuclear Power Plants, issued in July 1980 9.1-14 Susquehanna Steam Electric Station New Fuel Storage Vault Criticality Safety Analysis for ATRIUM 11 Fuel, ANP-3764(P), Revision 0, Framatome Inc, July 2019.
9.1-15 Susquehanna Steam Electric Station Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel, ANP-3765(P), Revision 0, Framatome Inc, August 2019.
9.1-16 NUREG 0544, Single-Failure Proof Cranes for Nuclear Power Plants 9.1-17 Generic Letter 85-11, Completion of Phase II of Control of Heavy Loads at Nuclear Power Plants June 28, 1985 9.1-18 NRC Bulletin 96-02, Movement of Heavy Loads Over Spent Fuel 9.1-19 PLA-857, N. W. Curtis to NRC, dated 6/22/81, Unit 1 Phase-One Response FSAR Rev. 71 9.1-53
SSES-FSAR Text Rev. 79 9.1-20 PLA-937, N. W. Curtis to NRC, dated 9/24/81, Unit 1 Phase-Two Response 9.1-21 PLA-1110, N. W. Curtis to NRC, dated 6/4/82, Unit 1 Phase-Two Response for Special Lifting Devices 9.1-22 USNRC letter to N. W. Curtis, dated 5/7/82, Unit 1 Phase-One Draft Technical Evaluation Report 9.1-23 PLA-1332, N. W. Curtis to NRC, dated 11/18/82, Unit 1 Phase-One Response to Draft Technical Evaluation Report 9.1-24 USNRC letter to N. W. Curtis, dated 7/21/83, Unit 1 Phase-One Safety Evaluation Report 9.1-25 PLA-1752, N. W. Curtis to NRC, dated 7/22/83, Unit 2 Phase-One Response 9.1-26 PLA-1843, N. W. Curtis to NRC, dated 9/29/83, Unit 2 Phase-Two Response 9.1-27 PLA-1988, N. W. Curtis to NRC, dated 12/13/83, Unit 2 Phase-Two Supplement 9.1-28 PLA-2511, H. W. Keiser to NRC, dated 9/30/85, Proposed Amendment 24 to License NPF-22 9.1-29 PLA-3521, H. W. Keiser to NRC, dated 2/27/91 Control of Heavy Loads 9.1-30 USNRC letter to N. W. Curtis, dated 10/31/83, Unit 2 Phase-One Safety Evaluation Report 9.1-31 PLA-4460, RG Byram to NRC, dated 5/13/96, 30 Day Response to Bulletin 96-02 Movement of Heavy Loads 9.1-32 NEI 08-05, Industry Initiative on Control of Heavy Loads Rev. 0 9.1-33 USNRC letter to TC Houghton, dated 9/5/08, NRC Safety Evaluation and Endorsement of NEI 08-05 9.1-34 ANSI N14.6, 1978, American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kg) or More for Nuclear Materials 9.1-35 ASME B30.9, 2003, Slings 9.1-36 ASME B30.2, 2005, Overhead and Gantry Cranes 9.1-37 Crane Manufacturers Association of America (CMAA) specification #70 and #74 9.1-38 NUREG-0776, Safety Evaluation Report Related to Operation of SSES Units 1 and 2, Supplement 6 9.1-39 Generic Letter 80-113, "Control of Heavy Loads at Nuclear Power Plants" December 22, 1980 FSAR Rev. 71 9.1-54
SSES-FSAR Text Rev. 79 9.1-40 Generic Letter 81-07, "Control of Heavy Loads" February 3, 1981 9.1-41 NRC Regulatory Issue Summary 2005-25: Clarification of NRC Guidelines for Control of Heavy Loads, dated October 31, 2005 9.1-42 NRC Regulatory Issue Summary 2005-25 Supplement 1: Clarification of NRC Guidelines for Control of Heavy Loads, dated May 29, 2007 9.1-43 NRC Regulatory Issue Summary 2008-28: "Endorsement of Nuclear Energy Institute Guidance for Reactor Vessel Head Heavy Load Lifts," dated December 1, 2008 9.1-44 Procedure SSES-HPP-3032-0090, Equipment Mobilization and Preparation 9.1-45 Procedure SSES-HPP-3032-0100, MPC Pre-Operation Inspection at SSES 9.1-46 Procedure SSES-HPP-3032-0200, MPC Loading at SSES 9.1-47 Procedure SSES-HPP-3032-0300, MPC Processing (FHD) at SSES 9.1-48 Procedure SSES-HPP-3032-0400, MPC Stack-up and Transfer 9.1-49 Procedure SSES-HPP-3032-0500, HI-STORM Operations and Transport at SSES 9.1-50 Drawing FF61945 Sh. 320 General Arrangement Drawing for Dry Storage Equipment FSAR Rev. 71 9.1-55
SSES-FSAR TABLE 9.1-1 SPENT FUEL POOL COOLING AND CLEANUP SYSTEM COMPONENT DESCRIPTION EQUIPMENT TOH, PUMP POWER DESIGN COMPONENT NOS. TYPE QUANTITY SIZE EACH MATERIAL FLOW EACH FT. HX CAPACITY PRESSURE/
EACH TEMP.
PSIG/°F Fuel Pool Cooling Pumps 1P-211A,B,C Horiz. Cntr. 3 . ss 600 gpm 200 60 hp 150/155 Fuel Pool Cooling Pumps 2P-211 A,8,C Horiz . Cntr. 3 - ss 600 gpm 200 60 hp 150/155 Fuel Pool F/D Holding Pump OP, 1P,2P-205 Horiz. Cntr. 3 - ss 160 gpm 45 5 hp 150/200 Fuel Pool F/D Precoat Pump OP-201 Horiz. Cntr. 1 - ss 475 gpm 65 15 hp 150/200 Fuel Pool Skimmer Surge Tank 1T-208 Vert. Cyl. 1 8027 gal. ss .
- 1 s,200 Fuel Pool Skimmer Surge Tank 2T-208 Vert. Cyl. 1 8027 gal. ss 15/200 Fuel Pool F/D Resin Feed Tank OT-202 Vert. Cyl. 1 188 gal. ss . . Atm/1S0 Fuel Pool F/0 Precoat Tank OT-201 Vert. Cyl. 1 500 gal. ss . . . Atm/150 Fuel Pool Filter Demineralizer OF, 1 F,2F-202 Vert. Cyl. 3 325 ft 2 ss 650 gpm - 150/200 Pressure Precoat Fuel Pool Heal Exch. 1E-202A,B,C Shell and 3 1310 ft 2 Shell and Shell : 4.4 x 10 6 BIU/hr 150/220 2E-202A,B,C Straight Channels: CS 296000 lb/hr at 125/110°F Shell 95/104°F Tubes Tubes, Fixed 3 1310 ft 2 Tubes & Tubes: 4.4 x. 106 Btu/hr 1501200 Tube Sheets, Tube-Sheets: 496000 lb/hr at 125/110°F Counter ss Shell Flow 95/104°F Tubes Rev . 49, 04/96
SSES-FSAR Table Rev. 55 TABLE 9.1-2 FUEL SERVICING EQUIPMENT CLASSIFICATION Essential Safety Quality Seismic Component Classification Classification Group Category No. Identification (a) (b) (c) (d) 1 Fuel Prep Machine PE 3 E I 2 New Fuel Inspection Stand NE 0 E NA 3 Channel Bolt Wrench NE 0 E NA 4 Channel Handling Tool NE 0 E NA 5 Fuel Pool Sipper NE 0 E NA 6 Fuel Inspection Fixture NE 0 E NA 7 Channel Gauging Fixture NE 0 E NA 8 General Purpose Grapple PE 2 E I 9 Fuel Transfer Stand NE 0 E NA 10 New Fuel Channel Up Ender NE 0 E NA 11 New Fuel Up Ending Stand NE 0 E NA Table Notes (a) NE - Nonessential PE - Passive Essential (b) 0 - Other (c) B - ASME Code Section III Class-2 D - ANSI B31.1 E - Industrial Code Applies I - Electrical Codes Apply (d) NA - No Seismic Requirements FSAR Rev. 62 Page 1 of 1
SSES-FSAR Rev. 55 HlSTORICAt..***1*N*FORMATlON TABLE 9.1-2a ORIGINAL DESIGN BASIS DECAY HEAT OUTPUT UNDER NORMAL FUEL STORAGE CONDITIONS*
Year of No. of Total No. of Time After Decay Heat Decay Heat Discharge Assemblies Assemblies in Shutdown Fraction (MW)
Discharged the Pool 1982 248 248 14 years 7.23(1 o- ) 7.72(10-)
1983 200 448 13 years 7.40(10- ) 6.38(10- )
1984 184 632 12 years 7.58(10- ) 6.01(10-)
1985 184 816 11 years 8.15(1 o- ) 6.46( 1o- )
1986 184 1000 10 years 8.35 (10- ) 6.62(1 o- )
1987 184 1184 9 years 8.55 (10- ) 6.78(10-)
1988 184 1368 8 years 8.78 (10-) 6.96(1 o- )
1989 184 1552 7 years 9.03(1 o- ) 7.16(1 o- )
1990 184 1736 6 years 9.35 (10- ) 7.41(10-)
1991 184 1920 5 years 9.83 (10- ) 7.80(10- )
1992 184 2104 4 years 1.07(1 o- ) 8.50(1 o- )
1993 184 2288 3 years 1.26(10-) 1.00(10-)
1994 184 2472 2 years 1. 73(1 o- ) 1.37(1 o- )
1995 184 2656 1 years 3.16(1 o- ) 2.51(10-)
1996 184 2840 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> 3.10(1 o- ) 2.44 Total Decay Heat= 3.70 MW 1.26(107 ) Btu/hr
- The first three batches have an exposure of 25,500 MWd/MTU while all subsequent batches have an of 500 MWd/MTU.
Rev. 60 Page 1 of 1
SSES-FSAR
=
I TABLE 9.1-2b ORIGINAL DESIGN BASIS DECAY HEAT OUTPUT UNDER NORMAL FUEL STORAGE CONDITIONS Year of No. of Total No. of Discharge Assemblies Assemblies in the Time After Decay Heat Decay Heat Discharged Pool Shutdown Fraction (MW) 1983 260 260 14 years 7.23(1 o*S) s.10,10* 2 1 1984 192 452 13 years 7.40(1 o-~; 6.13(10" 21 1985 180 632 12 years 7 .58110*5 ) 5.88(10* 1 )
1986 184 816 11 years 8. 15{10' 5 l 6.46(10* 2 1 1987 184 1000 10 years 8.35(10" 5 1 6.62{1 0* 21 1988 184 1184 9 years 8,55(10" 5) 6. 78{10*21 1989 184 1368 8 years 8.78(10" 5 ) 6.96(10*2 1 1990 184 1552 7 years 9.03(10.!:l 7 .16(10-.. )
1991 184 1736 6 years 9.35(10- 5 ) 7.41(10' 2) 1992 184 1920 5 years 9.83(10" 5 ) 1.so,10*2 )
1993 184 2104 4 years 1.07(10*4 ) 8.50(10* 2 1 1994 184 2288 3 years 1.26(10" 4 ) 1,QQ(10' 1 )
1995 184 2472 2 years 1. 73(1 o* l 4
1,37(10" 1 }
1996 184 2656 1 year 3. t 6(1 0*4 } 2.s1 (1 o*'}
1997 184 2840 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> 3.10(10' 1 ) 2.44 Total Decay Heat = 3.70 MW 1.26( 107 ) Btu/hr
-- ,_-__ /\(:*>:\-\.\ !\d'.::.-_::;.:t\<**--:i=,=*, =* --,-= .. *:
, V;i~
1 *:
'.Nb/:=~.:-:-:.:, ;f~*;;B~'.t~'.~hngiI/:::?
-* : *:*:,::*_-*--..:, *:... -;:-:.- ::'.:->::,:::::*_::,,: ::,:* ... : . .:*. :.= ,,... :::, :*.*. *:*-
Rev. 54, 10/99 Page 1 of 1
- SSES-FSAR TABLE 9 . 1-2c ORIGINAL DESIGN BASIS DECAY HEAT OUTPUT UNDER FULL CORE UNLOADED CONDITIONS*
Year of No. of Total No. of Discharge Assemblies Assemblies in the Time After Decay Heat Decay Heat Discharged Pool Shutdown Fraction (MWl 1982 220 220 1 l years 7.76110'~} 7 .36(1 o *Z) 1983 200 420 10 years 7 .95(10 '5 ) 6.86!10*2 l 1984 184 604 9 years a.1511 o-~l 6.4 7(1 0"2 )
1985 184 788 8 years 8.77(10*5 1 6 .96( 10*2 1 1986 184 972 7 years 9 .03(10*5 1 7 .1 en 0* 2
)
1987 184 1156 6 years 9.34( 10* 5 1 7.41(10" 2
)
1988 184 1340 5 years 9.82(10' 5 l 7.7sno*2 1 1989 184 1524 4 years 1 .01,10*4 } 8.49{10"2 )
1990 184 1708 3 years 1 .26(1 O-"I 1.00(10*1 1 1991 184 1892 2 years 1. 1211 o-~} 1.3611 Q' 1 )
1992 184 2076 1 year 3.13(10"') 2.48110' ~~
1993 764 2840 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> 2.sa(10*1 l 8.49 Total Decay Heat = 9.56 MW 3.26(1 0 7 ) Btu!hr
+ The first three batches have an exposure of 25,500 MWd/MTU while all subsequent batches have an exposure of 28,500 MWd/MTU .
./< :* i\ :j::/; :*:t)ii:-:*):;):,\),:..*: ...:;**; . . :>::-.:'t:::: ; .* :....
. Uriit =Nb.':, F~\ 1.9.93:RMueling.: ) :::**.: .. ' :
- -:.'- : *~;-~_:, .:*-i.'.-:ti: ;*\/:\!i::i\i/(.i:*!*:/{;*:. *., . .;**:. .** .\:!; : ... .:, ....... :.. *:,., -. ;/_
Rev. 54, 10/99 Page 1 of 1
SSES-FSAR TABLE 9.1-2d ORIGINAL DESIGN BASIS DECAY HEAT OUTPUT UNDER FULL CORE UNLOAD CONDITIONS Year of No. of Total No. of Discharge Assemblies Assemblies in the Time After Decay Heat Decay Heat Discharged Pool Shutdown Fraction (MW) 1983 232 232 11 years 1. 76(10* 5) 7.76110' 2 }
1984 192 424 10 years 7.95(10" 5 } 6.58(10* 2 1 1985 180 604 9 years 8.15(10" 5) 6.33(10 2
)
1986 184 788 8 years 8.77110' 5
) 6.96(10* 2}
1987 184 972 7 years 9.03(10' 5 ) 7.16(10- 2 )
1988 184 1156 6 years 9.34( 10'5 ) 7.41{10* 7 )
1989 184 1340 5 years 9.82{10' 5 ) 7.79110" 2) 1990 184 1524 4 years t .07(10' 4 ) 8.49(, 0* 2 ,
1991 184 1708 3 years 1 .26{ 1 ff 4 ) 1.00(1 o*l}
1992 184 1892 2 years 1.72(10' 4 ) 1.36(10*;1 1993 184 2076 1 year 3.13(10*4 } 2.48(10' 1 )
1994 764 2840 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> 2.ss110* 3 1 8.49 Total Decay Heat = 9.56 MW 3.26(10 7 ) Btu/hr
.. ,* ..-=.,:*-:**::::: -:-::*:::....,_:****:>.: : .:*.... *. ..-:
i': Unit =No'.=. .2<'-'.1 'as*4::~~_tu'e/ing= ::)>
.. ::.: _. *.:. :.::,:::=\:-\<:(=:;;<;-;:/Y:*?*::/:.::;: :*.
Rev. 54, 10/99 Page 1 of 1
SSES-FSAR Table Rev. 56 TABLE 9.1-2e UPDATED DESIGN BASIS DECAY HEAT OUTPUT UNDER NORMAL FUEL STORAGE CONDITIONS Cycle of No. of Total No. of Time After Decay Heat Discharge* Assemblies Assemblies in the Pool Shutdown (MW)
Discharged X-8 290 290 16 years 0.0826 X-7 316 606 14 years 0.0957 X-6 316 922 12 years 0.1045 X-5 316 1238 10 years 0.1105 X-4 316 1554 8 years 0.1166 X-3 316 1870 6 years 0.1404 X-2 316 2186 4 years 0.1868 X-1 316 2502 2 years 0.3696 X 348 2850 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> 5.3083 Total Decay Heat = 6.5150 MW 2.2229 (107) Btu/hr
- The results bound any discharge cycle, X, and are based on power uprate conditions.
FSAR Rev. 64 Page 1 of 1
SSES-FSAR Table Rev. 56 TABLE 9.1-2f UPDATED DESIGN BASIS DECAY HEAT OUTPUT UNDER FULL CORE UNLOAD CONDITIONS Cycle of No. of Total No. of Time After Decay Heat (MW)
Discharge* Assemblies Assemblies in the Shutdown Discharged Pool X-7 190 190 14 years 0.0578 I X-6 316 506 12 years 0.1045 I X-5 316 822 10 years 0.1104 I X-4 316 1138 8 years 0.1165 I X-3 316 1454 6 years 0.1402 I X-2 316 1770 4 years 0.1862 I X-1 316 2086 2 years 0.3677 I X 764 2850 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> 10.6952 I Total Decay Heat = 11.7785 MW 4.0188(107) Btu/hr
- The results bound any discharge cycle, X, and are based on power uprate conditions.
FSAR Rev. 64 Page 1 of 1
SSES-FSAR Table Rev. 62 TABLE 9.1-3 REACTOR VESSEL SERVICING EQUIPMENT CLASSIFICATION COMPONENT IDENTIFICATION ESSENTIAL SAFETY QUALITY SEISMIC NO. CLASSIFICATION(a) CLASSIFICATION(b) GROUP(C) CATEGORY(d) 1 Reactor Vessel Service Tools NE O E NA 2a Steam Line Plug PE O E I (REM*Light Model) 2b Main Steam Line (MSL) Plugs PE O E I (Spring Disk Model) (Wetlift) 2c MSL Plugs Restraint Ring (Wetlift) PE O E 3a Shroud Head Bolt Wrench NE O E NA (Supplied w/Nuclear System) 3b Shroud Head Bolt Wrench (Wetlift) NE O E NA 3c Shroud Head Bolt Wrench(Scientech) 3d Shroud Head Bolt Wrench Type NE O E NA HTC (Scientech) t-------+------------------+--------i 4 Vessel Nut Handling Tool NE O E NA I
5 Head Holding Pedestal NE O E NA 6 Head Nut and Washer Rack NE O E NA 7 Head Stud Rack NE O E NA 8a Deleted 8b Deleted 8c Dryer and Separator Strongback PE O E NA*
9 Head Strongback/Carousel PE O E NA*
10 Service Platform (e) NE O E NA 11 Service Platform Support (e) NE O E NA 12a Steam Line Plug Inst. Tool NE O E NA (Integral with REM*Light Model) 12b MSL Plugs I/R Tool (Wetlift) PE O E NA 13 Rigid Pole Handling System PE O E NA*
(Wetlift) 14 Refuel Floor Auxiliary Platform NE O E **
(RFAP) 15 Jet Pump Plugs NE O E I 16 360 Degree Refuel Work Platform NE O E **
Table Notes (a) NE - Nonessential PE - Passive Essential (b) O - Other (c) B - ASME Code Section III Class-2 D - ANSI B31.1 E - Industrial Code Applies I - Electrical Codes Apply (d) NA - No Seismic Requirements I - Seismic Category I (e) The Service Platform and Service Platform Support are not used and have been eliminated.
- Dynamic analysis methods for seismic loading are not applicable, as this equipment is supported by the reactor service crane.
- Seismic Analysis was performed to ensure that this item is not a Safety Impact Item.
FSAR REV. 69 Page 1 of 1
SSES-FSAR TABLE 9.1-4 UNDER-REACTOR VESSEL SERVICING EQUIPMENT AND TOOLS CLASSIFICATION Seismic
-E_quip_tnerit(fool
.. . . Clas-sificatj(;Sn Safety Class
. . Category
- 1. CRD Handling Equipment - Non-Essential "Othe( NA
- 2. Equipment Handling Platform Non-Essential 11 0ther" NA
- 3. Thermal Sleeve Removal Tool Non-Essential 11 0ther 11 NA
- 4. In-Coore Flange Seal Test Plug Non-Essential "Other 11 NA
- 5. Key Bender Non-Essential 11 0ther11 NA Table Notes NA - No Seismic Requirements i Rev. 54, 10/99 Page 1 of 1
SSES-FSAR Table Rev. 63 TABLE 9.1-5 TOOLS AND SERVICES EQUIPMENT Fuel Servicing Equipment In-Vessel Servicing Equipment Fuel Preparation Machines Multiple LPRM Strongback New Fuel Inspection Stand Instrument Strongback Channel Bolt Wrenches Control Rod Grapple Channel Handling Tool Control Rod Guide Tube Grapple Fuel Pool Sipper Fuel Support Grapple Channel Gauging Fixture Grid Guide General Purpose Grapples Control Rod Latch Tool (Standard Handle, Extended Fuel Inspection Fixture Handle, Flag)
Fuel Transfer Stand Instrument Handling Tool New Fuel Channel Up Ender Control Rod Guide Tube Seal New Fuel Up Ending Stand In-Core Guide Tube Seals Blade Guides Servicing Aids Fuel Bundle Sampler Pool Tool Accessories Peripheral Orifice Grapple Actuating Poles Orifice Holder General Area Underwater Lights Peripheral Fuel Support Plug Local Area Underwater Lights Fuel Bail Cleaner Drop Lights Control Rod/Fuel Support Piece Combination Grapple Underwater TV Monitoring System Underwater Vacuum Cleaner Viewing Aids Light Support Brackets In-Core Detector Cutter In-Core Manipulator Refueling Equipment Refueling Equipment Servicing Tools Refueling Platforms Reactor Vessel Servicing Equipment Storage Equipment Reactor Vessel Servicing tools Multi-Purpose Storage Canister Steam Line Plugs (REM*Light Model) Spent Fuel Storage Racks Main Steam Line (MSL) Plugs Channel Storage Racks (Spring Disk Model) [Wetlift]
MSL Plugs Restraint Ring [Wetlift] Control Rod Storage Racks Shroud Head Bolt Wrench In-Vessel Racks
[Supplied with Nuclear System] New Fuel Storage Rack Shroud Head Bolt Wrench [Wetlift] Control Rod Guide Tube Storage Rack Shroud Head Bolt Wrench Channel Bolt Storage Fixture
[Supplied Scientech]
Fuel Rod Storage Basket Shroud Head Bolt Wrench Type HTC
[Supplied Scientech]
Head Holding Pedestals Under-Reactor Vessel Servicing Equipment Head Stud Rack Dryer-Separator Strongback Control Rod Drive Servicing Tools Head Strongback /Carousel CRD Hydraulic System Tools Steam Line Plug/Installation Tool (REM*Light)
Control Rod Drive Handling Equipment
[Integral with REM*Light Plug]
MSL Plugs I/R Tool [Wetlift] Equipment Handling Platform Vessel Nut Handling Tool Thermal Sleeve Installation Tool Head Nut and Washer Storage Racks In-Core Flange Seal Test Plug Rigid Pole Handling System [Wetlift] Key Bender Refuel Pool Auxiliary Platform (RFAP)
Jet Pump Plugs 360 Degree Refuel Work Platform FSAR Rev. 69 Page 1 of 1
SSES-FSAR Table Rev. 55 TABLE 9.1~6a UNIT 1 REACTOR BUILDING CRANE FAILURE MODES AND EFFECT ANALYSIS Component or Component or Subsystem Effect ~f Failure Failure Mode Subsystem Failure Mode on the System Detection
- Remarks Power supply Loss of offsite power All crane movements Crane operator stopped by setting crane holding brakes and tripping all drive motors.
Main hoist hooks Failure of one hook Non, the redundant hook Periodic supports the load. inspection, jf not identified during the crane use Main hoist wire ropes Failure *of one rope Spurious, dynamic, load Crane operator Two vane switches.
transfer to the redundant mounted on the equalizer rope followed by setting frame, are provided to of crane holding brakes detect the wire rope and cessation of all crane failure and cut off power movements. The to the hoist.
dynamic load transfer will not cause the redundant rope load to exceed one-third of the rope breaking strength.
Main hoist drum Failure of drum shaft Possible load stalling, or Crane operator Then, the load can be noise and irregular hoist pos,tioned over its operation. Crane storage or taydown area operator to stop hoist and lowered by manual operation; this will result operation of the hoist in setting of the holding holding brakes.
brakes and the safe load suspension.
FSAR Rev. 56 Page 1 of 5
SSES--FSAR Table Rev. 55 TABLE 9.1-6a UNIT 1 REACTOR BUILDING CRANE FAILURE MODES AND EFFECT ANALYSIS Component or Component or Subsystem Effect of Failure Failure Mode Subsystem Failure Mode on the System Detection Remarks Main hoist holding Failure in open position of None, two additional Periodic inspection Three holding brakes are brakes one brake, when main holding brakes stop the provided, all rated at hoist is in operation and main hoist movement and 150% of the hoist motor power to the main hoist hold the load. torque, at the point of holdinQ brakes is cut off. aoolication.
Trolley holding brakes Failure, in open position. None, the power to the Crane operator Two holding brakes are of one brake when trolley trolley drive motor is cut provided, one rated at is in operation and power off at the same time, and 100% and the other rated to holding brakes is cut the trolley is stopped by 150% of the trolley drive off. the redundant holding motor torque at the point brake. of application.
Bridge holding brake Failure of the bridge None*, the power to the The holding brake is rated holding brake when bridge drive motors is cut 100% of the bridge drive bridge in operation and off at the same iime, and motor torque for each of the power to the brake is the bridge is stopped by the two bridge drive cut off. the holding brake on the motors.
other side of the bridge.
FSAR Rev. 56 Page 2 of 5
SSES*FSAR Table Rev. 55 TABLE 9.1--6a UNIT 1 REACTOR BUILDING CRANE FAILURE MODES AND EFFECT ANALYSIS Component or Component" or Subs*ystem Effect of Fai Iure Failure Mode Subsystem Failure Mode on the System Detectio'n Remarks Main hoist drive gear Failure of one gear case, Two gear cases (drive
- cases resulting in gear and idler case) are disengagement provided for the main hoist.
a) drive gear case Spurious drum revolving in the load lowering direction. The overspeed switch activated by revolving drum will set the hoist holding brakes and stop the load.
b) idler gear case None,.the drive gear case Crane operator with the hoist motor and the holding brakes maintain control of the load.
Main hoist upward Failure of switch movement, geared (lower) limit switch a) open Immediate power cut off Crane operator or to the-main hoist. As a periodic testing result, the hoist is stopped through action of the hoist holding brakes.
b) closed None, if hoist continues Periodic testing its upward travel it.will be stopped by action of the backup upper limit switch.
FSAR Rev. 56 Page 3 of 5
SSES~FSAR Table Rev. 55 TABLE 9.1-6a UNIT 1 REACTOR BUILDING CRANE FAILURE MODES ANO EFFECT ANALYSIS Component or Component or Subsystem Effect of Failure Failure Mode Subsystem Failure Mode on the System Detection Remarks Main hoist upward Failure of switch Tripped by physical movement (upper) contact with moving limit switch a) open Immediate power cut off Crane operator or upward lower load block.
to the main hoist. As a periodic testing result! the hoist is stopped through action of the hoist holding brakes.
b) closed None, hoist upward Periodic testing movement will be limited by the backup lower geared limit switch, before it reaches the upper limit switch.
Main hoist downward Failure of one limit switch movement geared limit switches a) open Immediate power cut off Crane operator or to the main hoist. As a periodic testing result, the hoist is stopped through action of the hoist holding brakes.
b) closed None, if hoist moves Periodic testing downward beyond the limit it will be stopped by the other (backup) limit switch.
FSAR Rev. 56 Page 4 of 5
SSES-FSAR Table Rev. 55 TABLE 9.1 .. 6a UNIT 1 REACTOR BUILDING CRANE FAILURE MODES AND EFFECT ANALYSIS Component or Component or Subsystem Effect of Failure Failure Mode Subsystem Failure Mode on the System Detection Remarks Main hoist overload Failure of the overload switch *switch a) open* Immediate power cut off Crane operator to the main hoist. As a result, the hoist is stopped through action of the hoist holding brakes.
b) closed None, the hoist motor overcurrent and Clcurrent rate of rise" protection backs up the failed load switch.\
Bridge and trolley Failure of one switch movement limit associated with a given switches bridge or trolley position b) oper. Immediate cut off of Crane operator power to respective drive motor(s) and holding brake(s) and stopping of all crane movements.
b) closed The load may enter the Periodic testing restricted area. unless prevented by crane operator action.
FSAR Rev. 56 Page 5 of 5
SSES-FSAR Table Rev. 55 TABLE 9.1-6b UNIT 2 REACTOR BUILDING CRANE FAILURE MODES AND EFFECT ANALYSIS Component or Component or Subsystem Effect of Failure Failure Subsvstem Failure Mode on the System Mode Detection Remarks Power supply Loss of offsite power A.II crane movements stopped by setting crane Crane operator holding brakes and tripping all drive motors Main hoist hook or wire Failure of hook or wire Load drop Crane operator The results of this accident are addressed in rope rope the response to NUREG 0612.
Main hoist drum Failure of drum shaft Possible load stalling, or noise and irregular hoist Crane operator Then, the load can be positioned over its operation. Crane operator to stop hoist storage or laydown area and lowered by operation; this will result in setting of the holding manual operation of the hoist holding brakes and the safe load suspension. brakes.
Bridge holding brakes Failure of the bridge None, the power to the bridge drive motors is cut One holding brake, [provided tor each of the holding brake when bridge off at the same time, and the bridge is stopped by two bridge drive motors. The holding brake is in operation and power the holding brake on the other side of the bridge. is rated at 100% of the bridge drive motor to the brake is cut off. toraue Trolley holding brakes Failure of the trolley None, the power to the trolley drive motor is cut Crane operator The trolley is supplied with two brakes, one holding brake when trolley off at the same time, and the trolley is stopped by is in operation and power the holding brake.
rated at 100% and the second [rated]at 50%
of the full load motor torque.
I to the brake is cut off.
Main hoist holding brake Failure in open position of None, an additional holding brake stays the main Periodic Inspection Two holding brakes are provided, both rated one brake, when main hoist movement and holds the load. at 150% of the hoist motor torque, at the hoist in operation and point of application.
power to the main hoist holding brakes is cut off.
Main hoist gear cas~ Failure [of]gearcase, Uncontrolled lowering Crane operator Results of this accident are evaluated in resulting in gear response to NU REG 0612.
disengagement FSAR Rev. 61 Page 1 of 2
SSES-FSAR Table Rev. 55 TABLE 9.1-Gb UNIT 2 REACTOR BUILDING CRANE FAILURE MODES AND EFFECT ANALYSIS Component or Component or Subsystem Effect of Failure Failure Subsvstem Failure Mode on the Svstem Mode Detection Remarks dain hoist upward Failure of switch Immediate power cut off to the main hoist. As a Crane operator or novement, geared a) open result, the hoist is stopped through action of the periodic testing lower) limit switch hoist holdina brake.
b) closed None, if the hoist continues its upward travel, it Periodic testing will be stopped by action of the backup upper limit switch.
dain hoist upward Failure of switch Immediate power cut off to the main hoist. As a Crane operator or novement (upper) limit a) open result, the hoist is stopped through action of the periodic testing
- witch hoist holding brake.
b) closed None, hoist upward movement will be limited by Periodic testing the backup lower geared limit switch, before it reaches the upper limit switch.
dain hoist downward Failure of switch Immediate power cut off to the main hoist. As a Crane operator or novement switch a) open result, the hoist is stopped through action of the periodic testing hoist holdinc:i brakes.
b) closed Potential exists for reverse winding of the drum Crane operator or 1\11 major heavy loads are lifted in areas and damaging the wire rope and other periodic testing "here it is not possible to lower the hoist to components. 3 point that the switch would be challenged.
3ridge and trolley Failure of one switch Immediate cut off of power to respective drive Crane operator novement limit associated with a motors(s) and holding brake(s) and stopping of
- witches given bridge or trolley all crane movements.
position a) open b) closed The load may enter the restricted area, unless Periodic testing prevented by crane operator action.
FSAR Rev. 61 Page 2 of 2
SSES-FSAR TABLE 9.1-7a LOAD COMBINATIONS & FACTORED ALLOWABLE STRESS LIMITS
- The following load combinati0f1S shall be satisfied:
a} Normal Operating Conditions Stress Limits i) D+L Fs ii) D+L+P Fs iii) D+L+H Fs iv) D+L+T Fs v) D+L+T+P Fs vi) D+L+T+H Fs vii) D+L+T+E 1.25 Fs viii) P+H Fs ix) D+L+E Fs x) D+L+SRV Fs (Note 2) xi) D+L+T+SRV Fs (Note 2) xii) D+L + T +E+SRV 1.5 Fs (Note 2) xiii) D+L+ T +F+SRV 1.5 Fs (Note 2) b) Design Accident and Extreme Environmental Conditions Stress Limits i) D+L+T+E' (See Note 1) ii) D+L+T'+E' (See Note 1) iii) D+L+T+I 1.25 Fs iv) D+L+T+I 1.33 Fs v) D+L+T' 1.33 Fs vi) D+L+I 1.25 Fs vii) D+L +T +SRV+LOCA or CHUGGING (See Note 1,2) viii) O+L +T +E+SRV+LOCA or CHUGGING (See Note 1,2) ix) D+L+T+E'+SRV+LOCA or CHUGGING (See Note 1,2)
NOTE:
- 1. In no case shall the allowable stress exceed 0.9Fy in bending, 0.85Fy in axial tension or compression and 0.5Fy in shear. Where Fs is governed by requirements of stability (local or lateral buckling), fs shall not exceed 1.5Fs.
Rev. 53, 04/99 Page 1 of 1
SSES-FSAR TABLE 9.l-7b LOAD DEFINITIONS D Dead load of racks including the support framing.
L = Live load due to the weight of fuel assemblies considered as varying from zero to full load, and loadings corresponding to varying placement of the fuel assemblies in the rack considered so that the roost critical loads are obtained.
T = Thermal effects, loads moments and forces based on the most critical transient or steady state condition during normal operation and shut down conditions.
p = iifting force of 4000 pounds applied to the top of rack at any fuel bundle location. (This is necessary in the event that the fuel assembly or grappling device binds during normal removal.)
H = Horizontal force of 1000 pounds applied to the top of rack at any fuel bundle location and at a varying angle from 0° to 45° from the horizontal.
E = Loads and resulting forces and moments generated by .the Operating Basis Earthquake, (OBE) resulting from ground surface horizontal acceleration of O.OSg and vertical ground surface acceleration of 0.033g, acting simultaneously.
E' = Loads and resulting forces and moments generated by the Design Basis Earthquake (DBE) resulting from ground surface horizontal acceleration of O.lOg, and vertical ground surface acceleration of 0.067g, acting simultaneously.
SRV = Safety Relief Valve Loads T' = Thermal effects, loads forces and moments which may occur during a design accident.
I Impact loads resulting from the following as a result of a dropped fuel bundle impacting the racks from an elevation of 18 inches above the rack. The height of the fuel bundle above the racks is limited by the fuel handling equipment. The racks are analyzed for a bundle dropping thru an empty cavity. The . racks will remain functional for this case.
LOCA = Loads Associated wtih Loss of Coolant Accident CHUGGING= Chugging Loads Rev. 35, 07/84
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT NEW FUEL STORAGE FIGURE 9.1-1, Rev 49 AutoCAD: Figure Fsar 9_1_1.dwg
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT NEW FUEL VAULT COVER DETAILS FIGURE 9.12
6.625".....,......,_~....
C/CTYPE '-.
FUELASS'Y SIDE PLATE CORNER ANGLE POISON CAN ASSEMBLY ---1.,,____
,.,t (FSAR FIGURE 9.1-5)
BOTTOM GRID ADJUSTABLE FOOT-FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT SPENT FUEL RACK ISOMETRIC FIGURE 9.1-3, Rev 54 AutoCAD: Figure Fsar 9_1_3.dwg
FSAR REV.65 I
SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT SPENT FUEL RACK ARRANGEMENT FIGURE 9.1-4, Rev 49 I AutoCAD: Figure Fsar 9_1_4.dwg
Poison Can Sorat (TVP)
Section A-A (partial view of top grid)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT SPENT FUEL RACK DETAIL FIGURE 9.1-5, Rev 49 AutoCAD: Figure Fsar 9_1_5.dwg
6 156" I 5.215" I Zr Chan nel 0.12 o" -
Al Ca n 0.125 " -
VOID
~-----~ ---- -----...
0.047 " -
- Water hole s 0
~)
Water ga p 0.3505" -r-Wate r gap 0.047 5" Bora 1 ~
I 5.12" I
1 ..
0.125" X 5.250"
-1 6.625" FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT REFERENCE CASE FUEL STORAGE POISON CAN FIGURE 9.1-6, Rev 49 AutoCAD: Figure Fsar 9_1_6.dwg
FIGURE 9.1-7-1 REPLACED BY DWG. M-153, SH. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT FIGURE 9.1-7-1 REPLACED BY DWG. M-153, SH. 1 FIGURE 9.1-7-1, Rev. 55 AutoCAD Figure 9_1_7_1.doc
FIGURE 9.1-7-2 REPLACED BY DWG. M-153, SH. 2 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT FIGURE 9.1-7-2 REPLACED BY DWG. M-153, SH. 2 FIGURE 9.1-7-2, Rev. 55 AutoCAD Figure 9_1_7_2.doc
FIGURE 9.1-8 REPLACED BY DWG. M-154, SH. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT FIGURE 9.1-8 REPLACED BY DWG. M-154, SH. 1 FIGURE 9.1-8, Rev. 49 AutoCAD Figure 9_1_8.doc
-/ / :"
UPPER FRAME LOWER FRAME CARRIAGE FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT FUEL PREPARATION MACHINE FIGURE 9.1-9, Rev 55 AutoCAD: Figure Fsar 9_1_9.dwg
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT NEW FUEL INSPECTION STAND FIGURE 9.1-10, Rev 50 AutoCAD: Figure Fsar 9_1_10.dwg
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT CHANNEL BOLT WRENCH FIGURE 9.1-11, Rev 54 AutoCAD: Figure Fsar 9_1_11.dwg
7 FSAR REV.65 I
SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT I
CHANNEL HANDLING TOOL FIGURE 9.1-12, Rev 54 AutoCAD: Figure Fsar 9_1_12.dwg
3WAYVALVE FLEXIBLE PLASTIC TUBING 1.12 O.D. X R-t10.13MAX.DIA.
.75 I.D. X 40 FT. LG.
WATER REGUALTOR APPROX.
I183.37 2 HOSES I* I FUEL BUNDLE SIPPING CONTAINER 8.25 +/-0.12 DEMINIERALIZED T
WATER INLET FROM REFUELING FLOOR 5 gpm AT 30 psi MINIMUM MALE CONNECTOR FOR 0.75 I.D. HOSE FSAR REV.65 i FUEL POOL 15 fl. RAILING APPR I t SUSQUEHANNA STEAM ELECTRIC STATION 33.00 +/-0.12 FIGURE 9.1-13, Rev 54 AutoCAD: Figure Fsar 9_1_13.dwg PUMP-' DRAIN TO FUEL POOL L3w1RECAP MOTOR STARTER SWITCH 0.75 O.D. TUBING UNITS 1 & 2 NEMA 17.00+/-0.12 ... , 14-----15.81 +/-0.12 I FUEL POOL SIPPER STANDARD L5-15 CAT. NO.
GL-4726, VENDOR GESCO FINAL SAFETY ANALYSIS REPORT
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT FUEL INSPECTION FIXTURE FIGURE 9.114
1-FRAME 3-SCREW 6-LOCKWASHER --4-SHIM 8-FLATWASHER 9-NUT (TYPICAL OF 8)
- - - GAUGING BLOCK BLOCK MOUNTING HWD 7-BOLT
~::!...--------<;{ 6-LOCKWASHER 8-FLATWASHER (TYPICAL OF 4)
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT CHANNEL GAUGING FIXTURE FIGURE 9.1-15, Rev 49 AutoCAD: Figure Fsar 9_1_15.dwg
STUD RECEPTACLE (7 /16-14UNC)
VIEW PORT TORSION BAR HOUSING SET SCREW GRAPPLE HOOKS FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT GENERAL PURPOSE GRAPPLE FIGURE 9.1-16, Rev 49 AutoCAD: Figure Fsar 9_1_16.dwg
FIGURE REPLACED BY PPL DRAWING C-1807, SH. 1 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT FIGURE REPLACED BY PPL DRAWING C-1807, SH. 1 FIGURE 9.1-18, Rev. 56 AutoCAD Figure 9_1_18.doc
FIGURE REPLACED BY PPL DRAWING C-1807, SH. 2 FSAR REV. 65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT FIGURE REPLACED BY PPL DRAWING C-1807, SH. 2 FIGURE 9.1-19, Rev. 56 AutoCAD Figure 9_1_19.doc
5-TOP AUXILLARY HOIST C
OF REACTOR BUILDING CRANE SHIPPING CONTAINER NEW FUEL INSPECTION STAND FUEL D
NEW FUEL rit&li 5*
STORAGE VAULT 11 NEW ~UEL E
1t;i;;.~'it;;t~;;~:;;~1,:;;tJ~\t~li~l 11 11 11 .,:: -::;, r-~:;-J STORAGE RACKS';;.:->.*,
11 II u
~~,~:_::**
_,~._f:
- .~:~_*
- .i::,
"' ;>>ti
~~~*
FUEL POOL F
FUEL STORAGE RACKS
~~;;r=~:
- f{2;?2,~;
- ?;~,;;**5-;:*:~~.;-.-.~;;LB)I FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT SIMPLIFIED SECTION OF NEW FUEL HANDLING FACILITIES (SECTION X-X, FIGURE 9.1-18)
FIGURE 9.1-20, Rev 55 AutoCAD: Figure Fsar 9_1_20.dwg
,-, n I~....., \f I L I I \ t;. I I I I I
\\I
\ \!
G CHANNEL HANDING BOOM
'e I I f\. \
11 I I \ \\ REFUELING I ,1 : l, \ PlATFDRM r-1.,.,,.'t I I ~I
\,xJ
....................................1I..................................
1....-..-.--...-.-..---~
H I T~LESCDPING H TELESCOPING I I MAST MAST I I LJ I I I I FUEL POOL 11 I I A
I I K LI II II FUEL I PREPARATION 11 MACHINE I I REACTOR V WELL II II F
I I II FUEL STORAGE II RACKS A
II 11 II 11 J 11 I I II I I CORE II u
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT SIMPLIFIED SECTION OF REFUELING FACILITIES (SECTION Y-Y, FIGURE 9.1-18)
FIGURE 9.1-21, Rev 49 AutoCAD: Figure Fsar 9_1_21.dwg
00 BUILDING CRANE HI HI Ill Ill Ill Ill REFUELING Ill PIATFORM Ill Ill Ill OPEN Ill HATCHWAY Ill Ill B
I F I FUEL STORAGE RACKS FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT SIMPLIFIED SECTION OF FUEL SHIPPING FACILITIES (SECTION Z-Z, FIGURE 9.1-19)
FIGURE 9.1-22, Rev 55 AutoCAD: Figure Fsar 9_1_22.dwg
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT UNIT 2 REFUELING MAST AND GRAPPLE OUTLINE FIGURE 9.123
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT UNIT 1 REFUELING MAST AND GRAPPLE OUTLINE FIGURE 9.1231
SecurityRelated Information Figure Withheld Under 10 CFR 2.390 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT FUEL TRANSFER STAND FIGURE 9.124
TERMINAL CONNECTION VIEW PORT SOCKET SOCKET TOP PLATE SIDE PLATE CYLINDER ARM---+..--l....:
PINS~::!'"'---+--~-+
BLOCK CURVED PLATE AIR OPERATED GENERAL PURPOSE GRAPPLE FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT AIR OPERATED GENERAL PURPOSE GRAPPLE BWR 6 FIGURE 9.1-25, Rev 55 AutoCAD: Figure Fsar 9_1_25.dwg
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT NEW FUEL CHANNEL UP ENDER FIGURE 9.1-26, Rev 1 AutoCAD: Figure Fsar 9_1_26.dwg
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT NEW FUEL UP ENDING STAND FIGURE 9.1-27, Rev 1 AutoCAD: Figure Fsar 9_1_27.dwg
FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT NEW FUEL INSPECTION EQUIPMENT GENERAL ARRANGEMENT FIGURE 9.1-28, Rev 1 AutoCAD: Figure Fsar 9_1_28.dwg