ML21133A098

0 to Updated Final Safety Analysis Report, Chapter 9, Sections 1 Thru 2

ML21133A098
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Site:	Limerick
Issue date:	04/29/2021
From:	Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
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LGS UFSAR CHAPTER 9 - AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING The fuel storage and handling facilities provide safe and effective means of storage, cooling, transportation, and handling of nuclear fuel from the time it reaches the plant in a non-irradiated condition until it leaves the plant after post-irradiation cooling. These systems have no function related to the safe shutdown of the plant.

Additional spent fuel storage is provided on site and within the plant protected area boundary by an Independent Spent Fuel Storage Installation (ISFSI) meeting the requirements of 10CFR72.

Safety evaluations for the ISFSI are covered in a separate 10CFR72.212 document, and are therefore not detailed here. It is noted that existing UFSAR Section 9.1.4 and 9.1.5 references to a cask are actually referring to the shipping cask referred to above for the function to leaves the plant after post-irradiation cooling. Minor differences for the ISFSI transfer cask (TC) handling, such as cask positioning for preparations for loading and post-loading decontamination, lids placement and sealing, are not covered herein.

9.1.1 NEW FUEL STORAGE The new fuel is brought onsite prior to a scheduled refueling and is inspected and then stored in the spent fuel racks, which are located in the refueling area in the spent fuel pool (drawings M-122 and M-137).

9.1.1.1 Design Bases New fuel is stored in the spent fuel pool. The design bases are discussed in Section 9.1.2.1.

9.1.1.2 Facility Description The new fuel is stored in the spent fuel storage racks. The facility description is discussed in Section 9.1.2.2. Section 9.1.4 describes receipt inspection and handling of new fuel within the reactor enclosure.

9.1.1.3 Safety Evaluation New fuel is stored in the spent fuel pool. Storage of new fuel bundles in the storage vault is prohibited by existing fuel handling procedures that require new fuel to be stored in the spent fuel pool. The safety evaluation is discussed in Section 9.1.2.3.

9.1.2 SPENT FUEL STORAGE The spent fuel storage facility provides specially designed underwater storage space for the new and spent fuel assemblies. The facility is located in the refueling area and is shown in drawings M-122 and M-137. Prior to each refueling, new fuel is brought onsite, inspected, and stored in the spent fuel storage racks.

9.1.2.1 Design Bases CHAPTER 09 9.1-1 REV. 19, SEPTEMBER 2018

LGS UFSAR The following design bases for the spent fuel storage facilities are discussed in the indicated sections:

a. Seismic and quality group Section 3.2 classification

b. Protection from wind and Section 3.3 tornado effects

c. Flood design Section 3.4

d. Missile protection Section 3.5

e. Protection against dynamic Section 3.6 effects associated with postulated rupture of piping

f. Seismic design Section 3.7

g. Environmental design Section 3.11

h. Fire protection Section 9.5.1 In addition to the design bases listed above, the following design bases and codes and standards are applicable to the spent fuel storage facilities:

a. The spent and new fuel storage facility is designed to store fuel assemblies so that a keff less than or equal to 0.95 is maintained, assuming that the fuel storage racks are fully loaded with fuel of the highest anticipated enrichment and flooded with nonborated water, in compliance with GDC 62. This is achieved through use of a neutron poisoning material (Boral) sandwiched in the racks between adjoining fuel assemblies to ensure subcriticality by at least 5% k under all conditions. The presence of Boral is verified by receipt and inservice testing.

b. The spent fuel storage facility and all piping connections are designed to prevent a loss of cooling water from the spent fuel pool that could uncover stored fuel.

c. The spent fuel storage facility is designed to remain functional following an SSE.

d. Failures of systems or structures not designed to seismic Category I standards and located in the vicinity of the spent fuel storage facility do not cause a decrease in the subcriticality provided.

e. The design of the spent fuel storage racks is such that a fuel assembly can be inserted into any designated rack storage location.

f. The spent fuel storage facility is designed to prevent accidental criticality of stored fuel under adverse environmental and postulated fuel handling accident conditions.

CHAPTER 09 9.1-2 REV. 19, SEPTEMBER 2018

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g. The spent fuel pool is designed to withstand thermal- stresses resulting from the pool water boiling.

h. The spent fuel pool is designed to accommodate the storage of approximately 539% of the total number of fuel assemblies in the reactor core. (Placement of the storage racks which contain control rod blades and/or defective fuel assemblies in the fuel pool could limit fuel storage capability to approximately 513%.) Storage capacity is dependent on rack installation.

i. Shielding for the stored spent fuel assemblies is designed to protect plant personnel from exposure to direct radiation greater than that permitted for continuous occupational exposure during normal operations.

j. Applicable codes and standards are provided in Table 9.1-21.

9.1.2.2 Facility Description 9.1.2.2.1 General Description The spent fuel storage facility consists of the spent fuel pool, containing spent fuel storage racks, and serves as the storage area for irradiated fuel assemblies. The spent fuel pool and the adjacent reactor well, dryer/separator pool, and cask loading pit are located in the refueling area as shown in drawings M-122 and M-137. Figure 9.1-34 shows the initial layout of the Unit 1 fuel storage pool with a storage capacity of 3665 fuel assemblies (3714 fuel assemblies including the control rod blade/defective fuel storage rack). Figure 9.1-36 shows the initial layout of the Unit 2 fuel storage pool with a storage capacity of 3921 fuel assemblies (3970 fuel assemblies including the control rod blade/defective fuel storage rack). Figures 9.1-35 and 9.1-37 depict the potential maximum capacity layout for Unit 1 and Unit 2 (respectively) with a fuel storage capacity equal to the maximum licensed capacity of 4117 fuel assemblies. Cooling and cleanup of the spent fuel pool water is discussed in Section 9.1.3. Fuel handling systems are discussed in Section 9.1.4. The reactor enclosure crane is discussed in Section 9.1.5.

9.1.2.2.2 Component Description 9.1.2.2.2.1 Spent Fuel Pool The spent fuel pool is a post-tensioned, reinforced concrete structure that forms an integral part of the reactor enclosure. The pool has a volume of approximately 46,000 ft3 and is filled with demineralized water to a normal depth of 38'-3". This provides about 23 feet of water above the tops of the stored fuel assemblies.

The spent fuel pool is lined with stainless steel plate (Table 9.1-14) to minimize leakage and reduce corrosion product formation. A leakage collection system is provided to permit expedient detection of leaks through the stainless steel liner plate and to prevent the uncontrolled loss of pool water to areas below the pool. Drainage paths, designed to permit free gravity flow, are formed by welding channels behind the pool wall liner welded joints and by two trench monitoring systems embedded in the floor slab below the floor liner. The design of a typical drainage system is shown in Figure 9.1-40. Pool leakage is routed through a piping system, provided at the base of the pool wall, one of three dirty radwaste funnels as shown in drawing M-53. Leakage from each of seven segments of the leak collection system is routed through separate piping to enable identification of the area of the liner that is leaking.

CHAPTER 09 9.1-3 REV. 19, SEPTEMBER 2018

LGS UFSAR Leakage is detected by observation of water flowing out of the piping into the dirty radwaste funnel (drawing M-53) or by low level indication in the skimmer surge tank or the spent fuel pool (drawing M-53). Flow into the funnels is observed during periodic operator inspections. Skimmer surge tank low level alarms and trips are described in Section 9.1.3.5.

9.1.2.2.2.2 Spent Fuel Storage Racks The Unit 1 and Unit 2 spent fuel pools are licensed for a maximum fuel storage capacity of 4117 fuel assemblies each. Analyses were performed prior to the installation of maximum density fuel storage racks for the Unit 2 spent fuel pool. All analyses performed for the Unit 2 spent fuel pool are bounding for the Unit 1 spent fuel pool design and installation.

The high density fuel storage racks are installed in the Unit 1 spent fuel storage pool in such a manner as to ensure that there is a Boral plate between each adjoining fuel storage position. Each storage module is level with each other module at the top. There are 7.25 inches of clearance from the bottom of the module to the pool floor. This ensures adequate clearance for cooling water to enter each fuel cell and, through natural convection, keep each fuel assembly adequately cooled.

The maximum density spent fuel storage racks are modular, freestanding, top entry racks designed to maintain the spent (and new) fuel in a space geometry whereby each fuel assembly has a neutron poisoning material between it and any adjoining fuel assemblies. This precludes the possibility of criticality under normal and abnormal conditions. The only point of contact of the maximum density spent fuel storage rack is with the bottom liner plate.

The maximum density spent fuel storage racks (Figure 9.1-41) consists of five basic structural components: the storage box subassembly, the baseplate, the neutron poison material, the picture frame sheathing, and the support legs. The storage box subassemblies (Figure 9.1-41) are fabricated from two precision formed stainless steel channels by seam welding. Each storage box subassembly has two lateral holes punched near the bottom edge to provide an auxiliary flow path.

The stainless steel picture frame sheathing is attached to each side of a storage box subassembly by welds. The picture frame sheathing serves as the locator and retainer of the neutron poison material. A storage cell box subassembly, with picture frame sheathing and neutron poison material on all four sides is referred to as a composite box. The composite boxes are arranged in a checkerboard array and welded corner to corner to form an assemblage of storage cell locations (Figure 9.1-41). This assemblage is welded to the baseplate. The baseplate is 3/4 inches thick and provides a continuous horizontal surface for supporting the fuel assemblies. The baseplate has a concentric hole with a 45° taper in each storage cell location to provide a seating surface conforming to the fuel assembly. Adjustable support legs are welded to each corner of the baseplate. Each adjustment support leg is provided with a socket to enable remote leveling of the maximum density spent fuel storage rack after its placement in the fuel pool.

There are various sizes of maximum density spent fuel storage racks (Figures 9.1-35 through 9.1-37 and Table 9.1-23). The nominal center-to-center spacing between fuel assemblies in a maximum density spent fuel storage rack is 6.244 inches.

The neutron poison material used in the maximum density spent fuel storage racks is Boral. Boral is a thermal neutron poison material composed of boron carbide and 1100 alloy aluminum. The Boral is in plate form and is located and retained by the picture frame sheathing (Figure 9.1-41).

The sheathed volumes which contain the Boral plates are vented.

CHAPTER 09 9.1-4 REV. 19, SEPTEMBER 2018

LGS UFSAR The maximum density fuel storage racks are installed in the spent fuel storage pool in such a manner as to ensure that there is a Boral plate between each adjoining fuel storage position. Each maximum density spent fuel storage rack is level with each other rack at the top. There is a clearance of approximately 7.00 inches between the bottom of the rack and the fuel pool floor liner.

This ensures adequate clearance for cooling water to enter each fuel storage cell.

All materials used for construction are specified in accordance with the issue of the ASTM specifications. Traceability of major rack components to a heat lot are maintained. In addition, the suppliers' quality assurance/ quality control program was audited by the manufacturer and user to ensure that the Boral has the required minimum B4C density and uniform B4C distribution in each sheet. Boral traceability is maintained.

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.

The rack materials will have no significant degradation due to the total radiation doses expected in the spent fuel pool over the design life.

The racks are designed to withstand the loading under the following loading conditions: dead, live, jammed fuel assembly, dropped fuel assembly, thermal, OBE and design basis event seismic, SRV, and LOCA or chugging.

An ISI program will be in effect throughout the period in which spent fuel is stored in the spent fuel pool to ensure that the quality of the poisoned racks is maintained (Section 9.1.2.4).

9.1.2.2.2.3 Refueling Area Cavities As shown in drawing M-122 and M-137, adjacent to the spent fuel pool, but on opposite sides of it, are the cask loading pit and the reactor well. Adjacent to the reactor well is the dryer/separator pool. The cask loading pit is common to the two reactor units. Like the spent fuel pool, these cavities are lined with stainless steel plate and are provided with liner leakage collection systems.

The reactor well and the cask loading pit are connected to the spent fuel pool by fuel transfer canals approximately 4 feet wide. Each canal is provided with two gates to prevent loss of water from the spent fuel pool during periods when the adjacent cavity is not filled with water.

The cask loading pit is used for Low Level Radwaste (LLW) storage. In the cask loading pit free standing storage racks containing Storage Containers (SCs). The rack compartments hold LLW including control rod blades (CABs), fuel channels in sheaths, LPRM strings, velocity limiters, filters, stellite ball bearing and other small components. During LLW processing, processing equipment will be placed in the cask loading pit or on the Refuel Floor. The equipment can be removed when not in use. Equipment includes Activated Services Compactor (ASC) and Control Rod Blade Compactor (CABC), grapple, shear and punching tools, and supporting equipment. All LLW components are classified as non-safety related and Seismic II/I components. Because of the LLW stored in the cask loading pit, it must remain full at all times.

The cask loading pit is designed to permit the underwater loading of spent fuel assemblies into ISFSI transfer cask and/or spent fuel shipping casks. Decontamination of the spent fuel shipping and/or ISFSI transfer cask can be performed on the refueling floor adjacent to the cask loading pit. Cross connecting the Unit 1 and 2 spent fuel pools CHAPTER 09 9.1-5 REV. 19, SEPTEMBER 2018

LGS UFSAR (SFPs) through the cask loading pit allows the transfer of spent fuel and equipment from one Unit's Spent Fuel Pit to the other unit's SFP. Prior to transferring fuel or equipment, adequate spent fuel pool cooling capacity must be verified.

The reactor well is a basically circular cavity located directly above the primary containment.

Removal of the drywell head and reactor vessel head provides direct access from the reactor well to the inside of the reactor vessel. The reactor well is filled with water during transfer of fuel assemblies between the well and the spent fuel pool. Seals are provided at the bottom of the reactor well between the drywell and reactor well wall and a refueling bellows is provided between the reactor vessel and containment seal plate to prevent leakage of water out of the well.

The dryer/separator pool provides a place for storage of the steam dryer and steam separator when these two reactor components are removed from the reactor vessel. The dryer/separator pool is connected to the reactor well to permit underwater transfer of components between the two cavities. Seal plugs are provided to prevent loss of water during periods when the adjacent cavity is not filled with water.

9.1.2.2.2.4 Other Features The ventilation system for the refueling floor area is designed to minimize potential offsite exposures if there is a significant release of radioactivity from the spent fuel storage facility. The ventilation system for the refueling floor area is discussed in Section 9.4.2.

The area radiation monitoring equipment for the spent fuel storage facility is described in Section 12.3.4.

9.1.2.3 Safety Evaluation The subcriticality of the fuel assemblies stored in the spent fuel storage racks is the result of the geometrical arrangement of the fuel array and the presence of neutron-absorbing materials. The arrangement of the fuel assemblies in the spent fuel storage racks results in keff less than or equal to 0.95, assuming that the spent fuel storage racks are fully loaded with fuel of the highest anticipated enrichment. This is discussed in Section 9.1.2.3.1.

The design is in conformance with Regulatory Guide 1.13.

The bottoms of the fuel transfer canals between the spent fuel pool and the reactor well and between the spent fuel pool and cask loading pit are above the top of the stored spent fuel, thus ensuring that failure of the gates in these canals cannot result in the uncovering of the fuel.

To ensure that the spent fuel pool water level is not lowered by a malfunction of the fuel pool cooling and cleanup system, the system takes suction from the pool near the normal water level via the skimmer surge tanks. The system return lines enter the spent fuel pool from above the normal water level and are provided with siphon breaker holes near the normal water level to preclude the possibility of siphoning the pool.

The spent fuel storage facility is designed in accordance with seismic Category I requirements as specified in Section 3.2. The components (and supporting structures) of any system, equipment, or structure that is not seismic Category I and whose collapse could result in loss of a required CHAPTER 09 9.1-6 REV. 19, SEPTEMBER 2018

LGS UFSAR function of the spent fuel storage facility are analytically checked to determine that they will not collapse when subjected to seismic loading resulting from the SSE.

All spent fuel pool gates and liners are seismic Category I.

Liner leakage detection system piping, the fuel pool cooling and cleanup system piping, and the wave suppression scupper piping are all seismic Category IIA. The only other piping attached to or in the spent fuel pool is from the RHR and ESW systems for backup cooling and makeup. This piping is seismic Category I.

Loss of any of the seismic Category IIA piping would not affect the ability to maintain spent fuel cooling or to maintain adequate submergence of the fuel.

Accidental dropping of movable heavy objects into the fuel pool is precluded by the use of administrative procedures, electrical interlocks to limit the load 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 electrical interlocks and administrative procedures are described in Section 9.1.4. In addition, heavy load handling in the vicinity of the fuel pools is in compliance with NUREG-0612 guidelines such that the likelihood of a heavy load drop is precluded.

The design of the spent fuel pool provides enough reinforcement in the concrete to withstand the thermal-stresses resulting from pool boiling.

The maximum density spent fuel storate rack is capable of withstanding a force of 4000 pounds in any direction at the top of the rack without any permanent deformation of the fuel cell in the active fuel region.

The auxiliary hoist of the reactor enclosure crane has a maximum uplift force in excess of 4000 pounds. The auxiliary hoist does not handle new fuel above, or in the fuel pool, except to transfer new fuel to the fuel prep machines. Other fuel handling operations are performed by the hoists on the refueling platform which have an uplift (controlled) limit of 1200 pounds or less.

If there is a stuck fuel assembly, the lifting bail yields at an uplift force greater than 1500 pounds; therefore, possible damage to the racks is precluded even if the 1200 pound control limit is not considered.

9.1.2.3.1 Criticality Control 9.1.2.3.1.1 Basic Assumptions of Criticality Analysis The spent fuel storage facility is designed to prevent accidental criticality of stored fuel under adverse environmental and postulated fuel handling accident conditions. The geometry of the spent fuel storage array is such that Keff will be 0.95. Studies were performed which establish fuel storage compliance limits consistent with Keff 0.95 for the Maximum Density racks (Reference 9.1-9). All General Electric supplied fuel shall comply with the fuel storage reactivity criteria in Reference 9.1-7 (GESTAR-II), which conservatively bounds the rack designs analyzed in Reference 9.1-9. Additional studies for fuel designs not covered under GESTAR-II are listed as references in Section 9.1.6. To ensure that the design criteria are met, the following normal and abnormal spent fuel storage conditions were analyzed:

a. Normal positioning in the spent fuel storage array CHAPTER 09 9.1-7 REV. 19, SEPTEMBER 2018

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b. Pool water temperature increases to 212°F

c. Abnormal positioning in the spent fuel storage array

d. Dropped fuel bundle adjacent to storage area

e. Dropped fuel bundle on top of and through fuel storage racks.

Transport to Diffusion Correction = 0.917 - 0.898 - 0.001

= 0.018 This result should be applied to the adjusted PDQ keff

c. Summary of Results:

keff (PDQ) adjusted 0.915 Transport to Diffusion Correction, PDQ-KENO 0.018 Final keff (upper bound, 95% confidence) 0.933 Design Limit, keff 0.950 The final keff value (0.933) includes all the design specification tolerances, the postulation of a dropped fuel assembly, the model bias, and the 95% confidence interval from the KENO calculations. However, the negative reactivity effect (approximately -0.005 k) due to the presence of U-234 and the parasitic structure materials (i.e. spacer grids) in each assembly, and the positive reactivity effect (0.006 k) due to the possible absence of a Boral plate are not included.

9.1.2.3.1.2.1 Assumptions of Criticality Analysis - Maximum Density Spent Fuel Storage Racks To ensure the analysis followed a conservative approach and conformed to the general guidelines of criticality safety analysis, the calculations were performed using the following criteria:

1) The racks are assumed to contain the most reactive fuel authorized to be stored.

2) Moderator is pure, unborated water at a temperature corresponding to the highest reactivity (4°C, 39.2°F)

3) Criticality safety analyses are based upon the keff of an infinite array of storage cells, ie, no credit is taken for neutron leakage (except as necessary in the assessment of abnormal/accident conditions)

4) Neutron absorption in minor structural member is neglected, ie, spacer grids are replaced by water.

CHAPTER 09 9.1-8 REV. 19, SEPTEMBER 2018

LGS UFSAR 9.1.2.3.1.2.2 Calculational Models - Maximum Density Spent Fuel Storage Racks Criticality analyses of the maximum density spent fuel storage racks were performed using both the CASMO-3 code (a two-dimensional multi-group transport theory code) and the KENO-5a code (a Monte Carlo code), using the 27-group SCALE cross-section library with NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral treatment). SCALE is an acronym for the Standardized Computer Analysis for Licensing Evaluation, which is a standard cross-section set developed by the Oak Ridge National Laboratory for the USNRC.CASMO-3 was also used as a means of evaluating small reactivity increments associated with manufacturing tolerances.

Benchmark calculations indicated a bias of 0.0000 0.0024 for CASMO-3 and 0.0101 0.0018 (95% confidence) for NITAWL-KENO-5a.

In the geometric model used in the calculations, each fuel rod and its cladding was described explicitly and reflecting boundary conditions (zero neutron current) were used in the axial direction and at the centerline of the Boral and stainless steel plates between storage cells. These boundary conditions have the effect of creating an infinite array of storage cells in the X and Y directions.

Since the cross section library in NITAWL is not valid below 20°C, a small correction factor (0.0026 k, determined by CASMO-3) to 4°C was applied to the NITAWL-KENO-5a calculations.

9.1.2.3.1.2.3 Reference Case Calculations - Maximum Density Spent Fuel Storage Racks

a. Physical Parameters and Basic Storage Cell Geometry The design basis storage rack cell consists of an egg-crate structure (Figure 9.1-41) with fixed neutron absorber material (Boral) of 0.0200 g/cm2 boron-10 areal density (0.0185g/cm2 boron-10 minimum) positioned between the fuel assembly storage cells. The storage cell design for analysis is based upon a nominal center-to-center lattice spacing of 6.244 inches. The 0.075 inch thick stainless steel box which defines the fuel assembly storage cell has a nominal inside dimension of 6.05 inches. This allows adequate clearance for inserting or removing the fuel assemblies, with or without the Zircaloy channel. Boral plates are not needed or used on the exterior walls of racks facing non-fueled regions, ie, the fuel pool walls.

Manufacturing tolerances, used in evaluating uncertainties in reactivity, are listed in Table 9.1-26.

Along the outer periphery of racks facing another rack, the Boral is 4.70 inches wide rather than 5.0 inches wide. This small reduction in width is compensated by the water-gap between racks and by the thickness of the stainless steel.

Therefore, the reactivity consequences of the slightly reduced Boral width in these locations are negligible.

b. Results of the Design Basis Case Calculations Using the input data from Table 9.1-24 and Figure 9.1-42, the keff values of the design basis case at 4°C were calculated. The maximum reactivity calculated by the CASMO-3 code is 0.9444. The maximum reactivity calculated by the NITAWL-KENO-5a code is 0.9412.

CHAPTER 09 9.1-9 REV. 19, SEPTEMBER 2018

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c. Fuel Effect Because the racks must be able to store various fuel types additional fuel types were evaluated using the data from Table 9.1-24. The results are summarized in Table 9.1-27.

9.1.2.3.1.2.4 Sensitivity and Tolerance Reactivity Calculations - Maximum Density Spent Fuel Storage Racks

a. Temperature Effect Using the design basis storage rack geometry, the temperature of the fuel pool water was varied. The results are given in Table 9.1-25.

b. Void Effect The effect of boiling was studied by varying the voids from 0% to 20% at a temperature of 252F with the reference geometry. The results are given in Table 9.1-25. As indicated, keff decreases continuously as the void fraction increases.

Boiling at the submerged depth of the racks would occur at approximately 250F.

c. Effect of Boron Loading Variation The Boral Absorber plates used in the storage cells are nominally 0.075 inches thick. With a boron-10 areal density of 0.0200 g/cgm2. The manufacturing tolerance limit is 0.0015 g/cm2 in the boron-10 loading, which assures that the minimum boron-10 areal density will not be less than 0.0185 g/cm2. Differential CASMO-3 calculations indicate that this tolerance limit results in an incremental reactivity uncertainty of 0.0051 k (Table 9.1-26).

d. Effects of Boral Width Tolerance Variation The reference storage cell design uses a Boral plate width of 5.00 1/16 inches.

For a reduction in width of the maximum tolerance, the calculated positive reactivity increment is +0.0018 k (Table 9.1-26).

e. Effect of Storage Cell Lattice Pitch Variation The design storage cell lattice spacing between fuel assemblies is 6.244 inches.

For the manufacturing tolerance of 0.040 inches, the corresponding uncertainty in reactivity is 0.0027 k as determined by the CASMO-3 code (Table 9.1-26).

f. Effect of Stainless Steel Thickness Tolerance The nominal thickness of the stainless steel box and the stainless steel picture frame sheathing are 0.075 and 0.035 inches respectively. The maximum positive reactivity effect for a mean stainless steel thickness tolerance of 0.008 inches is 0.0004 k as determined by the CASMO-3 code (Table 9.1-26).

g. Effect of Fuel Enrichment and Density Variation CHAPTER 09 9.1-10 REV. 19, SEPTEMBER 2018

LGS UFSAR The nominal design enrichment is 3.50 wt% U-235. CASMO-3 calculations of the sensitivity to an enrichment variation of 0.05 wt% U-235 yielded an uncertainty in reactivity of 0.0033 k.

The design basis calculation assumed a UO2 stack density 95% theoretical density, corresponding to a density of 10.42 g/cm3 . Calculations were also made to determine the sensitivity to a conservative tolerance in UO2 fuel density of 0.20 g/cm3 in density or a maximum density of 10.62 g/cm3. These calculations indicate that the storage rack k is increased by 0.0026 k (Table 9.1-26). A lower fuel density would result in a correspondingly lower value of reactivity.

h. Effects of Zirconium Flow Channel The design basis calculations assumed the presence of a flow channel. Elimination of the zirconium flow channel results in a decrease in reactivity of approximately 0.008 k.

i. Effects of Water Gap Spacing Between Racks For normal storage conditions with a water gap between racks, the array keff of a rack would be less than the reference design k and Boral plates along the wall of the racks facing the water gap would not be necessary. However, as an additional and precautionary measure, rack design provides for Boral plates on storage cells on one of the two rack walls along the water gap. With this conservative configuration, the design assures that, even under abnormal or accident conditions.

the storage rack reactivity will remain less than the 0.95 keff regulatory limit.

9.1.2.3.1.2.5 Special Cases - Maximum Density Spent Fuel Storage Racks

a. Abnormal Location of a Fuel Assembly It is theoretically possible to suspend a fuel assembly of the highest allowable reactivity outside and adjacent to the spent fuel storage rack.

Neutron leakage, inherent along the rack edge, significantly reduces the reactivity consequences of an extraneous fuel assembly. Three dimensional KENO-5a calculations show that the keff, with an assembly outside of and adjacent to a rack, is less than the design basis k of an infinite radial array. Thus, the abnormal location of a fuel assembly will have a negligible reactivity effect.

b. Eccentric Fuel Assembly Positioning The fuel assembly is normally located in the center of the storage rack cell with the bottom fittings and spacers that mechanically restrict lateral movement of the fuel assemblies. Nevertheless, calculations with the fuel assembly moved into the corner of the storage rack cell (four-assembly cluster at the closest approach), resulted in a substantial negative reactivity effect. Thus, the nominal case, with the fuel assembly positioned in the center of the storage cell, yields the maximum reactivity.

c. Dropped Fuel Assembly CHAPTER 09 9.1-11 REV. 19, SEPTEMBER 2018

LGS UFSAR For a drop of a fuel assembly on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the fuel of more than 12 inches. A separation of 12 inches is sufficient to preclude neutron coupling (ie, an effectively infinite separation). Maximum expected deformation under seismic or accident conditions will not reduce the minimum spacing between fuel assemblies to less than 12 inches. Consequently, fuel assembly drop accidents will not result in a significant increase in reactivity

(<0.0001 k) due to the separation distance.

d. Fuel Rack Lateral Movement Normally, the individual racks in the spent fuel pool are separated by a water gap. Lateral motion of a rack, postulated as a consequence of the design basis earthquake, could reduce or eliminate this water gap. Since the exterior walls of one of the racks along the water gap contains a Boral plate, closing the gap would not result in any increase above the design basis reactivity.

9.1.2.3.1.2.6 Summary and Conclusion - Maximum Density spent Fuel Storage Racks For the design basis reactivity calculations with uniform enrichment of 3.50 wt%, the nominal storage cell infinite multiplication factor, k, is 0.9366 (CASMO-3 code). With a k of 0.0078 for all known uncertainties statistically combined, the maximum k in the fuel rack is 0.944 which is below the design basis limit of 0.95 for keff. Independent calculations with NITAWL-KENO-5a (1,250,000 histories in 2500 generations) gave a k (including all uncertainties) of 0.9335 +0.0077 (95% confidence, maximum k of 0.941) which is in good agreement with the CASMO-3 calculations (Table 9.1-28).

9.1.2.3.2 Structural Analyses 9.1.2.3.2.6 Bases For Analyses - Maximum Density Spent Fuel Storage Rack The maximum density spent fuel storage racks are seismic Category I equipment as defined in Regulatory Guide 1.13. These racks are designed to withstand the effects of a design basis earthquake and remain functional, in accordance with Regulatory Guide 1.29 and 10CFR50.

The physical properties for the materials used in the manufacture of the maximum density spent fuel storage racks are given in Table 9.1-29. The load combinations and allowable stresses for the maximum density spent fuel storage racks are given in Table 9.1-30.

The structure of the racks is designed to remain functional (although some permanent deformation may occur) and to maintain the required spacing between stored fuel assemblies in the event of impact of a "heavy fuel assembly (consolidated fuel assembly)" (1360 pounds) dropped on the rack from a height of 36 inches. In this case, local plastic deformation is allowed at the point of impact.

The structure of the racks is also analyzed for effects of the impact of the fuel bundle dropped through an empty storage cavity. Local failure of the baseplate is acceptable, a gross structural failure is not permissible. A comparative analysis with impact conditions, as stated above, is also CHAPTER 09 9.1-12 REV. 19, SEPTEMBER 2018

LGS UFSAR conducted on a rack due to an uplift force of 4000 pounds on a stuck fuel bundle. No permanent deformation of the fuel cell in the active fuel region is allowed in this case. A force of 4000 pounds is well in excess of the maximum (controlled) uplift force from the refueling platform main hoist of 1200 pounds.

9.1.2.3.2.7 Seismic Analyses - Maximum Density Spent Fuel Storage Rack A combination of time history and static seismic analysis was performed for the maximum density spent fuel storage racks by Holtec International. Rack dynamic simulations were performed using a 3-D single rack 22-degrees of freedom (DOF) model and a Whole Pool Multi-Rack (WPMR) model which considered all the racks in the pool.

a. 3-D 22-DOF Single Rack Model The fuel rack motion is captured by modeling the rack as a twelve degree-of-freedom structure (Figures 9.1-43 through 9.1-45). Movement of the rack cross-section at any height is described by six degrees-of-freedom of the rack base and six degrees-of-freedom at the rack top. Rattling fuel assemblies within the rack are modeled by five lumped masses. Each lumped mass has two horizontal displacement degrees-of-freedom. Vertical motion of the fuel assembly mass is assumed to be equal to the rack vertical motion at the baseplate level. The centroid of each fuel assembly mass can be located off center, relative to the rack structure centroid at that level, to simulate a partially loaded rack.

Seismic motion of a rack is characterized by random rattling of the fuel assemblies in their individual storage locations. All fuel assemblies are assumed to move in-phase within a rack. This exaggerates computed dynamic loading on the rack structure and therefore yields conservative results.

Fluid coupling between the rack and fuel assemblies, and between the rack and the fuel pool wall, is simulated by appropriate inertial coupling in the system kinetic energy. Fluid coupling terms for rack-to-rack coupling are based on opposed-phase motion of adjacent modules.

Fluid damping and form drag is conservatively neglected. Sloshing is negligible at the top of the rack and is neglected in the analysis of the rack.

Potential impacts between rack and fuel assemblies are accounted for by appropriate "compression only" gap elements between the masses involved. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at the top and bottom of the rack in two horizontal directions. Bottom elements are located at the baseplate elevation.

Pedestals are modeled by gap elements in the vertical direction and as "rigid links" for transferring horizontal stress. Each pedestal support is linked to the fuel pool liner by two friction springs. Local pedestal spring stiffness accounts for floor elasticity and for local rack elasticity just above the pedestal.

CHAPTER 09 9.1-13 REV. 19, SEPTEMBER 2018

LGS UFSAR Rattling of fuel assemblies inside the storage locations causes the gap between fuel assemblies and the cell wall to change from a maximum of twice the nominal gap to a theoretical zero gap. Fluid coupling coefficients are based on the nominal gap.

b. Whole Pool Multi-Rack (WPMR) Model The single rack 3-D (22-DOF) model outlined in the preceding subsection is used to evaluate structural integrity physical stability, and to initially assess kinematic compliance of the rack modules. Prescribing the motion of the racks adjacent to the module being analyzed is an assumption in the single rack simulation. For closely spaced racks, demonstration of kinematic compliance is further confirmed by modeling all modules in one comprehensive simulation using a Whole Pool Multi-Rack (WPMR) model. In the WPMR analysis, all racks are modeled, and their correct fluid interaction is included in the model.

The presence of fluid moving in the narrow gaps between the racks and between the racks and the fuel pool walls causes both near and far field fluid coupling effects. A single rack simulation can effectively include only hydrodynamic effects due to contiguous racks when a certain set of assumptions is used for the motion of the contiguous racks. In a WPMR analysis, far field fluid coupling effects of all the racks are accounted for by using the correct model of pool fluid mechanics. The external hydrodynamic mass due to the presence of walls or adjacent racks is computed in a manner consistent with fundamental fluid mechanics principles using conservative nominal fluid gaps in the fuel pool at the beginning of the seismic event.

The friction coefficient is ascribed to the support pedestal/pool bearing pad interface is consistent with experimental data. Friction coefficients, developed by a random number generator with Guaussian normal distribution characteristics, are imposed on each pedestal of each rack in the pool. The assigned values are then held constant during the entire simulation in order to obtain reproducible results. Thus, the WPMR analysis can simulate the effect of different coefficients of frictions at adjacent rack pedestals.

In the WPMR analysis, a reduced degree-of-freedom (RDOF) set is used to model each rack plus contained fuel. The rack structure is modeled by six degrees-of-freedom. A portion of contained fuel assemblies is assumed to rattle at the top of the rack, while the remainder of the contained fuel is assumed as a distributed mass attached to the rack. The rattling portion of the contained fuel is modeled by two horizontal degrees-of-freedom.

Thus, the WPMR model involves all the racks in the spent fuel pool with each individual rack modeled as an eight degree-of-freedom structure. The rattling portion of the fuel mass, within each rack, is chosen to ensure comparable results from displacement predictions from the single rack analysis using a 22-DOF model and the 8-DOF analysis under the same conditions. Unit 1 spent fuel pool configuration is bounded by the Unit 2 analyses.

9.1.2.3.2.8 Dropped Fuel Bundle Analysis - Maximum Density Spent Fuel Storage Rack Analysis were performed to evaluate the effects of a dropped fuel assembly. Two scenarios were evaluated, the deep drop and shallow drop.

In the deep drop scenario a "heavy fuel assembly (consolidated fuel assembly)" is postulated to drop from a height of 36 inches above the top of a storage location and impacts the baseplate of CHAPTER 09 9.1-14 REV. 19, SEPTEMBER 2018

LGS UFSAR the rack. While local failure of the baseplate is acceptable, a gross structural failure is not permissible. Also, the stored fuel array must remains subcritical. Results show that the spacing between cells is unaffected and the stored fuel array remains subcritical. Local baseplate deformation in the vicinity occurs, but the dropped fuel assembly does not impact the fuel pool liner. The impact results in a maximum baseplate movement toward the liner of less than 2.21 inches. The nominal baseplate height above the fuel pool liner is approximately 7.00 inches therefore, contact does not occur. If a fuel assembly drops through a cell located over a support leg, the impact load transmitted through the support leg to the fuel pool liner is less than the loads caused by a seismic event. Therefore, the concrete bearing pressures a calculated for the seismic event bound those due to a fuel drop accident.

In the shallow drop scenario a "heavy fuel assembly" is postulated to drop from a height of 36 inches above the top of the rack and impacts the top of the rack. Permanent deformation of the rack is acceptable, but it must be confined to the region above the top of the active fuel stored in the rack. The elevation of the top and bottom of active fuel must not be altered. Results show that the deformation caused by a dropped assembly will be less than or equal to 2.29 inches from the top of the rack. This region of deformation is less than the available cell length above the active fuel (over one foot). This load case bounds the scenario where the dropped fuel assembly rolls over after impact and impacts multiple locations.

9.1.2.3.2.9 Pool Interface Loads - Maximum Density Spent Fuel Storage Rack The maximum density spent fuel storage load acting on the fuel pool were determined using the results of the 3-D 22-DOF model and the WPMR model described in Section 9.1.2.3.2.7. The results were determined to be acceptable and are summarized in Table 9.1-31.

9.1.2.3.2.10 Summary and Conclusions - Maximum Density Spent Fuel Storage Rack The results of the structural analyses for the maximum density spent fuel storage racks are acceptable and summarized in Table 9.1-31.

9.1.2.3.3 Installation of New Maximum Density Racks Racks are lifted individually from the refueling floor lay-down area and lowered into position in the pools using the 125 ton reactor enclosure crane and additional lifting devices provided by the rack vendor.

Each rack is aligned and leveled as it is placed in its proper pool. Each rack is a freestanding unit that rests on the pool floor using bearing pads attached to corner leveling screws. Leveling screws are adjusted by remotely controlled tools.

The requirements of 10CFR50, Appendix B, as included in the LGS Units 1 and 2 Quality Assurance Plan, Volume I, are implemented for receipt, storage, and installation of the spent fuel racks. Work is performed in accordance with procedures that are reviewed and approved prior to use. Inspections are performed and documented by personnel other than those who performed the work. Site activities are subject to audit by Nuclear Oversite.

9.1.2.4 Tests and Inspections CHAPTER 09 9.1-15 REV. 19, SEPTEMBER 2018

LGS UFSAR The spent fuel storage racks require no special periodic testing or inspection for nuclear safety purposes. However, test coupons may be installed in the Unit 1 and Unit 2 Spent Fuel Storage Racks. Test coupon analysis will only be performed on a discretionary basis based upon specific suspected degradation of neutron poisoning material. Spent fuel storage rack assignment in the spent fuel pool is surveyed at the time of installation.

9.1.2.4.1 Deleted 9.1.2.4.2 Deleted 9.1.2.4.3 Test Coupon Description and Installation A typical test coupon is physically designed to simulate as nearly as possible, the actual in-service geometry, physical mounting, materials, and flow conditions of the neutron poison in the storage racks. The test coupons are mounted on a hanger assembly within the central eight feet of the fuel zone of the storage racks, where the gamma flux is expected to be reasonably uniform.

9.1.2.4.4 Test Coupon Inspection Test coupon inspection is intended to identify changes in physical properties of the neutron poisoning material by performing various non-destructive and destructive testing.

9.1.3 FUEL POOL COOLING AND CLEANUP SYSTEM The FPCC system is designed to remove the decay heat generated by the spent fuel assemblies stored in the fuel pool and to maintain the pool water at a clarity and purity suitable both for underwater operations and for the protection of personnel in the refueling area. The FPCC system has no function related to the safe shutdown of the plant.

9.1.3.1 Design Bases

a. The FPCC system piping is designed so that operator error or a loss of piping integrity cannot result in the draining of the spent fuel pool so that stored fuel would be uncovered.

b. The FPCC system is designed to provide a source of makeup water to ensure the maintenance of the fuel pool water level.

c. All piping and components of the FPCC system that form part of the flow path for makeup water from the ESW system, RHRSW system, and cooling water to and from the RHR system are designed to remain functional following an SSE.

d. The FPCC system is comprised of three pumps and three heat exchangers connected by a common header providing three approximately 50% capacity cooling trains. The Service Water System removes heat from the heat exchangers.

Following an accident, two of the three pumps may be powered by emergency on-site power and any heat exchanger may be cooled by the Reactor Enclosure Cooling Water System. A precoat-type filter/demineralizer package is provided, through which a portion of the system flow can be diverted. As similar backup filter/demineralizer package common to both units is also provided.

e. The FPCC system was originally designed to maintain the bulk water temperature in the spent fuel pool at or below 140°F under normal operating conditions with two CHAPTER 09 9.1-16 REV. 19, SEPTEMBER 2018

LGS UFSAR FPCC pumps and two FPCC heat exchangers in operation. This is based on the normal heat load discharge history shown in Table 9.1-2A.

The Unit 1 FPCC system is capable of removing 1.61 x 107 BTU/hr and Unit 2 FPCC system is capable of removing 1.805 x 107 BTU/hr with any one train unavailable. The FPCC systems have been evaluated for increased spent fuel storage capacity and faster fuel transfer rate (6.7 vs. 10 days) resulting in a heat load of 1.805 x 107 BTU/hr. The results show that on Unit 2 with any two FPCC pumps and any two FPCC heat exchangers in operation, the spent fuel pool bulk temperature will peak at 143oF. The spent fuel pool bulk temperature will exceed 140oF for approximately 2.5 days. This is based on the normal heat load discharge history shown in Table 9.1-2B (reference 9.1-15). This analysis is not applicable to Unit 1, since Unit 1 has a smaller capacity FPCC system and the ADHR would be used for heat loads exceeding 1.61 x 107 BTU/hr.

Each refueling offload scenario is cycle specific and is likely to vary from Table 9.1-2B. Therefore, administrative controls are maintained to assure that the spent fuel pool heat load does not exceed FPCC heat removal capacity during shutdown, refueling and power operating conditions such that the FPCC system is capable of maintaining the spent fuel pool cooling temperature at a maximum of 143oF, assuming a single active failure during normal operating conditions. If the anticipated heat load exceeds the FPCC capability, the Residual Heat removal (RHR) System is placed in service and FPCC is secured. When the spent fuel pool and reactor refueling cavity well are connected, the combined decay heat load of the in-core fuel and spent fuel pool inventory is typically removed by the RHR system, alone, operating as an alternate decay heat removal method (ADHR).

f. The FPCC system was originally designed to permit the RHR system to be used, through a cross-tie, to maintain the bulk water temperature in the spent fuel pool at or below 140°F, with a maximum anticipated decay heat load of 3.64x107 Btu/hr.

This is based on one full core offloaded from the reactor ten days after shutdown to fill the spent fuel pool, plus the previous normal refueling loads from 18 month refuelings as shown in Table 9.1-2A.

A bounding analysis of the FPCC system has been performed, which considered a maximum reactor power of 3527 MW, spent fuel pool increased storage capacity and a faster fuel transfer rate. This analysis results in a maximum anticipated heat load of 4.832 x 107 Btu/hr. This is based on one full core offloaded from the reactor vessel approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> after shutdown plus the previous normal refueling loads from 24 month refuelings as shown in Table 9.1-2B (references 9.1-16 and 9.1-17) . A one-time per unit refueling cycle length extension to approximately 27 months (from a nominal 24 months) for Unit 1 Cycle 7 and Unit 2 Cycle 5 was evaluated. It was determined that the increase in decay heat generation due to each affected core's total operating period of 6.25 years (vs. 6 years) has no significant effect on maximum calculated fuel pool heat load.

The FPCC system cross-tie to RHR is normally used for the RHR ADHR method to maintain both the spent fuel pool and reactor water temperatures below 140F during refueling. See Section 5.4.7 for further discussion of ADHR method.

Additionally, the FPCC system(s) or the RHR fuel pool assist mode can be used to remove decay heat from both the spent fuel pool(s) and the reactor vessel by cooling the spent fuel pool(s). Evaluations have shown that with the reactor cavity flooded up and the fuel pool gates removed, natural circulation will transfer decay and sensible heat from the reactor vessel to the fuel pool(s). The required cooling CHAPTER 09 9.1-17 REV. 19, SEPTEMBER 2018

LGS UFSAR capacity of the FPCC system(s) or the RHR fuel pool assist mode can be verified by calculation or demonstration to show that these systems are alternate methods of decay heat removal. Natural circulation will also induce reactor coolant circulation through the reactor core such that reactor core temperature can be accurately measured at an appropriate location.

g. The FPCC system is designed to maintain the optical clarity of the water in the spent fuel pool and other refueling area cavities (cask loading pit, reactor well, and dryer/separator pool) so that fuel handling and equipment handling operations are not hampered by limited visibility.

h. The FPCC system is designed to limit the fission product and activated corrosion product concentrations in the water of the spent fuel pool and other refueling area cavities to permit continuous occupancy of the refueling area by plant personnel.

The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the FPCC system are discussed in Section 3.2.

9.1.3.2 System Description 9.1.3.2.1 General Description The FPCC system consists primarily of the pool water collection equipment, including wave suppression scupper and skimmer surge tanks, a cooling train with two heat exchangers, two pumps, a cleanup loop, and the discharge diffusers in the spent fuel pool. A backup heat exchanger and a backup pump are also included in the system.

9.1.3.2.2 Component Description Design parameters for individual components of the FPCC system are listed in Table 9.1-1.

9.1.3.2.2.1 Skimmer Surge Tanks Two skimmer surge tanks collect overflow water from the spent fuel pool and reactor well through skimmer drain openings, with adjustable weirs at the pool surface elevation. The skimmer surge tanks also collect overflow water from the cask loading pit and the dryer/separator pool via skimmer drain headers. A wave suppression scupper along one side of the spent fuel pool is drained via the skimmer drain headers. The skimmer surge tanks provide a suction head for the fuel pool cooling pumps and RHR pumps and serve as a buffer volume during transient flows in the normally closed-loop FPCC system.

9.1.3.2.2.2 Fuel Pool Heat Exchangers Three fuel pool heat exchangers of the shell and straight tube-type are provided to transfer heat from the fuel pool water to the service water system (Section 9.2.1). Normally a maximum of two heat exchangers are in service and the third serves as a backup. The heat exchangers are arranged in parallel and are located in the reactor enclosure below the bottom of the skimmer surge tanks. Fuel pool water circulates through the shell side of the heat exchangers, and service water circulates through the tube side.

CHAPTER 09 9.1-18 REV. 19, SEPTEMBER 2018

LGS UFSAR The service water side of the heat exchangers is maintained at a higher pressure than the fuel pool water side to minimize the possibility of radioactive contamination of the service water system from a tube leak.

Interconnecting piping is provided so that the heat exchangers can also be cooled by the RECW system in case the service water system is not available.

9.1.3.2.2.3 Fuel Pool Cooling Pumps Three centrifugal pumps are provided to circulate water through the FPCC system. Normally a maximum of 2 pumps are in operation and the third serves as a backup. The pumps are arranged in parallel and take suction from the fuel pool heat exchangers through a common header. Two of the pumps are powered from Class 1E sources. The pumps can be operated from a panel located in the reactor enclosure and a panel located in the radwaste enclosure.

9.1.3.2.2.4 Fuel Pool Filter/Demineralizer Package The cleanup loop of the FPCC system includes a filter/demineralizer package located separately in shielded cells in the radwaste enclosure. A spare filter/demineralizer package common to the two reactor units is also provided.

The fuel pool filter/demineralizer is a precoat-type, using powdered cation-anion resins as the coating media on the external surface of the filter elements. The filter elements are cylindrical stainless steel mesh, mounted vertically in a tube sheet and replaceable as a unit. The upper head of the filter/demineralizer vessel is removable for installing and replacing the filter elements. The ion exchange resin is a mixture of finely ground cation and anion resins, in proportions determined by service; this resin is referred to as precoat.

The precoat is applied to the surface of the filter elements by a flowing process called precoating.

To maintain the filter medium on the filter elements during the interval between the precoating operation and the return-to-normal operation, or if there is an abnormal decrease in process flow, a centrifugal holding pump is provided with each filter/demineralizer to recirculate water through the unit.

The fuel pool filter/demineralizer system is designed to maintain water quality within the following limits:

a. Conductivity: 3 mho/cm at 25°C

b. Chlorides: 0.5 ppm

c. pH: 5.3 to 7.5 at 25C

d. Total Suspended Solids: 1 ppm A strainer is provided in the effluent stream of each filter/ demineralizer to limit the migration of filter medium particles that pass through the filter elements. The strainer is capable of withstanding a differential pressure greater than the shutoff head of the fuel pool cooling pumps.

9.1.3.2.2.5 Backwash and Precoat Equipment CHAPTER 09 9.1-19 REV. 19, SEPTEMBER 2018

LGS UFSAR The backwash and precoat subsystem is common to the two reactor units and serves all three filter/demineralizer packages. Included in the subsystem are a backwash air accumulator, precoat tank, and filter precoat pump. Compressed air from the backwash air accumulator is used in conjunction with condensate water to backwash the spent filter medium from the filter elements with a minimum volume of water. Backwash slurry is drained to the waste sludge tank of the solid radwaste system (Section 11.4).

New ion exchange resin is mixed in the precoat tank and transferred as a slurry by the filter-precoat pump to the filter/demineralizer, where it is deposited on the filter elements. An agitator is provided with the precoat tank for mixing. Piping and valves associated with the precoat subsystem allow the precoat water to be recirculated to the precoat tank or the suction side of the filter precoat pump. The precoat subsystem can also be used for cleaning and decontamination of the filter/demineralizers.

9.1.3.2.3 System Operation The FPCC system removes decay heat from fuel stored in the spent fuel pool and includes equipment to maintain the purity of the water in the system. The FPCC system heat removal capacity is shown in Table 9.1-2B. Water from the spent fuel pool flows through weirs and a wave suppression scupper at the pool surface into two skimmer surge tanks adjacent to the pool. Water in the skimmer surge tanks flows by gravity through the fuel pool heat exchangers to the suctions of the fuel pool cooling pumps. From the pumps, water is returned to the spent fuel pool through two return lines. A portion of the discharge flow of the pumps can be diverted through the cleanup loop before being returned to the pool. Heat is removed from the fuel pool heat exchangers by the service water system (drawing M-10).

The filter/demineralizer in the cleanup loop maintains pool water purity and clarity by a combination of filtration and ion exchange. Disposable ion exchange resins in the filter/demineralizer remove ionic fission product and corrosion product impurities and also serve as a filter for particulate matter. The ion exchange resins are replaced when the pressure drop across the filter/demineralizer is excessive (25 psid at rated flow) or when the resins are exhausted as indicated by discharge conductivity being above the limit for the spent fuel pool given in Section 9.1.3.2.2. The backwash and precoat operations are controlled from the radwaste enclosure.

The FPCC system is sampled monthly at the filter/demineralizer influent and effluent lines. These grab samples are analyzed by gamma isotopic analysis. The ratio of the influent activity to the effluent activity (decontamination factor) will be determined from this analysis. The fuel pool filter/

demineralizer will be regenerated on low decontamination factor.

The grab samples will also be analyzed for pH, chloride, conductivity, and turbidity. Turbidity is an indication of insolubles. The pH, conductivity, and chloride limits are given in Section 9.1.3.2.2.

During normal plant operation, the FPCC system serves only the spent fuel pool. During refueling operations, however, when the reactor well, dryer/separator pool, and/or cask loading pit are filled with water, the FPCC system can be aligned to recirculate and process the water in all these cavities. Water from the refueling water storage tank (Section 9.2.7) is used to fill the refueling area cavities. The refueling water pumps fill the cask loading pit through its drain line and fill the reactor well and the dryer/separator pool through diffusers in the reactor well. After refueling activities are completed, the refueling water pumps transfer water from the refueling area cavities back to the refueling water storage tank via a condensate filter/demineralizer if additional cleanup is CHAPTER 09 9.1-20 REV. 19, SEPTEMBER 2018

LGS UFSAR required. Gravity draining of the refueling water directly to the refueling water storage tank is also possible.

As the heat load on the spent fuel pool changes, the number of operating fuel pool cooling pumps and heat exchangers is adjusted to maintain the desired water temperature. The FPCC system has sufficient cooling capacity to maintain the spent fuel pool water at a temperature at or below 143°F, with a normal decay heat load of 1.61 x 107 Btu/hr on Unit 1 and 1.805 x 107 Btu/hr on Unit 2, with two pumps and two heat exchangers operating. The Unit 1 and 2 spent fuel pools can also be maintained cross connected as long as adequate spent fuel pool cooling capacity is verified.

If a heat load exceeding the capacity of the FPCC System is placed in the spent fuel pool, a cooling train of the RHR system (Section 5.4.7), consisting of an RHR pump and heat exchanger, can be substituted for the FPCC pumps and heat exchangers for cooling the pool water. A cross-connection between the drain line from the skimmer surge tanks and the RHR system allows one RHR pump to take suction from the skimmer surge tanks and pump fuel pool water through an RHR heat exchanger before returning it to the pool via return piping provided specifically for use with the RHR system Fuel Pool Cooling assist mode or via the normal Shutdown Cooling return path with RHR Alternate Decay Heat Removal method. The interconnecting piping between the RHR system and FPCC system is provided with either of two spool pieces, one with blind flanges for normal operation and one open spool piece for when the intertie is needed (drawing M-51).

Administrative controls prevent the use of the RHR system intertie unless the associated reactor is shut down and is in the refueling mode. With the fuel pools cross-tied with one unit in refueling mode and RHR aligned to fuel pool cooling SST, there is no impact on the ability of refueling unit RHR system to perform its safety related functions. Also, there is no impact on the ability to safely shutdown the operating unit.

The conditions under which cooling of the spent fuel pool water by the RHR system alone would be required include the unloading of a full core load of irradiated fuel into the pool or isolation of fuel pool cooling system to support RHR ADHR operation or RHR Fuel Pool Cooling assist operation.

The RHR system has sufficient heat removal capacity to maintain the spent fuel pool water at a temperature at or below 140°F, with a maximum anticipated decay heat load of 4.832 x 107 Btu/hr.

The RHR system may also be used for cooling if the fuel pool cooling system should be unavailable. Connection of the RHR system requires:

a. Installation of RHR system spool pieces

b. Manual closure of normally open valve 053-007

c. Manual opening of normally closed valves 053-006, 053-024A and 053-024B

d. Adjustment of fuel pool overflow weirs, if required

e. Adjustment of Fuel Pool/SST water level to ensure adequate suction head for RHR pumps, if required For Fuel Pool Cooling assist mode, the RHR supply and return lines and their associated valves are located in the fuel pool cooling pump and heat exchanger room at el 283' in the reactor enclosure.

The normal position of the above valves provides isolation of the safety-grade, seismic Category I RHR system from the FPCC system. The portion of the FPCC system downstream of valve 053-007 is designed to seismic Category I criteria (designated seismic Category IIA as discussed in Section 3.2.1) and would not be expected to fail due to a seismic event.

CHAPTER 09 9.1-21 REV. 19, SEPTEMBER 2018

LGS UFSAR If normal fuel pool cooling should be lost as a result of a pipe break in the seismic Category IIA portion of the system, the quantity of water released would be limited to the inventory of the pool and reactor cavity above the overflow weirs, the skimmer surge tanks, and the pump suction piping. The flood height and environmental conditions resulting from this break would not prevent personnel access, which is required to take the manual actions described above. The maximum temperature (150F) and pressure (31 psig) of the water in the line are not high enough to significantly affect the temperature, pressure, or humidity conditions in the room (the room is open at the top to the reactor enclosure access area at el 283'). The released fluid would not be highly radioactive. The water would drain out of the room at approximately the same rate as it flows from the break.

If there is a LOOP, the Class 1E buses are powered by the diesel generators, and the two FPCC pumps that receive Class 1E power can be restarted. Since normal service water is not available in this case, the FPCC heat exchangers can be cooled by the RECW system (which is in turn cooled by the ESW system) via interconnecting piping, after installation of normally removed spool pieces (drawings M-10, M-11 and M-13).

If there is a complete loss of capability to remove heat from the spent fuel pool using heat exchangers, heat can be removed by allowing the pool to boil and adding makeup water to maintain the pool water level. Makeup water requirements are shown in Table 9.1-2B. Makeup water is normally supplied to the skimmer surge tanks from the demineralized water makeup system (Section 9.2.5) by the actuation of a remote manually operated valve. If makeup water from this source is not available, makeup can be provided from the ultimate heat sink (spray pond) by either of two seismic Category I flow paths (drawings M-10, M-51 and M-53). The first of these backup makeup sources is a loop of the ESW system (Section 9.2.2) via a cross-connecting line to one of the RHR system return pipes in the spent fuel pool. The two ESW pumps in the ESW loop provide redundancy in motive power for this source of makeup supply. The manual valves that must be opened to initiate makeup from the ESW system are located in the control structure and are accessible after an accident. The second of these backup makeup sources is a loop of the RHRSW system (Section 9.2.3) via the piping of one RHR system loop and the cross-connecting piping leading to the RHR return piping in the spent fuel pool. The two RHRSW pumps in the RHRSW loop provide redundancy in motive power for this source of makeup supply. The manual valves that must be actuated and the spool piece that must be installed to establish this flow path are located in the reactor enclosure. These backup makeup water sources provide flow rates greater than the maximum makeup requirements shown in Table 9.1-2B. The spray pond design water volume includes its use as a source of makeup to the spent fuel pool for 30 days, without makeup to the pond during which time the cooling function of the FPCC system or RHR system can be established or an alternate makeup water supply can be established.

9.1.3.3 Safety Evaluation The cooling water return lines to the spent fuel pool associated with both the FPCC system and RHR system penetrate the walls of the spent fuel pool within 4 feet of the normal pool water level.

FPCC system cooling water return lines are provided with siphon breaker holes as close as possible to the normal water level to prevent the siphoning of pool water below the level of the holes. RHR system cooling water return lines terminate below the normal water level and above the minimum water level to prevent pool siphoning. The FPCC system takes suction from the spent fuel pool via the skimmer surge tanks through openings in the liner plate at the normal water level. There are no other piping penetrations in the pool liner plate. Therefore, there is no operator CHAPTER 09 9.1-22 REV. 19, SEPTEMBER 2018

LGS UFSAR error or FPCC system malfunction that could result in the draining of the spent fuel pool so that stored fuel would be uncovered.

The spent fuel pool is provided with redundant seismic Category I makeup capability to ensure an adequate supply of makeup water to the spent fuel pool under conditions of maximum anticipated evaporation associated with fuel pool boiling. The radiological consequences of a boiling spent fuel pool are discussed in Section 9.1.3.6. Makeup water is supplied from the spray pond using either the ESW system or RHRSW system. Redundant pumps in each loop of the ESW and RHRSW systems provide assurance of the availability of motive power to pump the makeup water.

The portions of the FPCC system that form part of the flow path during makeup from the ESW or RHRSW systems or spent fuel pool cooling by the RHR system, up to and including the boundary isolation valves, are designed in accordance with seismic Category I requirements as discussed in Section 3.2. The components (and supporting structures) of any system, equipment, or structure that is not seismic Category I and whose collapse could result in the loss of a required function of the FPCC system are analytically checked to determine that they will not collapse when subjected to seismic loading resulting from the SSE.

The design of the FPCC system with respect to the following areas is discussed in the indicated sections:

a. Protection from wind and tornado Section 3.3 effects

b. Flood design Section 3.4

c. Missile protection Section 3.5

d. Protection against dynamic effects associated with postulated rupture Section 3.6 of piping

e. Environmental design Section 3.11 A failure mode and effects analysis of the FPCC system is provided in Table 9.1-3.

The water level in the spent fuel storage pool is maintained at a height 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 FPCC system is normally minimal. Further details of radiological considerations are discussed in Chapter 12.

9.1.3.4 Inspection and Testing Requirements The FPCC system is preoperationally tested in accordance with the requirements of Chapter 14.

The safety-related systems that provide makeup water are periodically tested in accordance with the requirements of Chapter 16.

9.1.3.5 Instrumentation Applications CHAPTER 09 9.1-23 REV. 19, SEPTEMBER 2018

LGS UFSAR Instrumentation is provided to annunciate the high and the low spent fuel pool water levels in the control room. Skimmer surge tank level instrumentation provides level indication at the refueling floor control panel, both high and low level annunciation at the refueling floor control panel and the fuel pool cooling pump control panel, and by a common trouble alarm in the main control room panel. A low level alarm indicates that makeup to the FPCC system is needed.

The temperature of fuel pool cooling water at the common inlet and at the outlet of each fuel pool heat exchanger is recorded and high temperature is annunciated in the control room; high outlet temperature in the combined heat exchanger discharge is annunciated in the control room through the plant computer.

The discharge pressure of each fuel pool cooling pump is indicated locally. Low pump discharge and suction pressure is annunciated by a local alarm in the fuel pool cooling pump panel and by a common trouble alarm in the main control room. The pumps are tripped individually on low suction pressure.

High leakage rates through the refueling bellows or reactor well seals are annunciated at the refueling floor control panel and by a common trouble alarm in the main control room.

The equipment associated with the cleanup loop is controlled and monitored from panels in the radwaste enclosure. The parameters monitored include differential pressure across the filter/demineralizer, flow rate through the filter/demineralizer, and pressure differential across the strainer downstream of the filter/demineralizer.

9.1.3.6 Analysis for Nonseismic Fuel Pool Cooling and Cleanup System The FPCC system is not designed as a seismic Category I or engineered safeguards system except for interconnections with safety-related makeup water sources. The following analysis examines the consequences of a loss of spent fuel pool cooling and the use of the seismic Category I makeup water source.

It is assumed that loss of cooling to both spent fuel pools occurs. In addition, to maximize both the heat loads and the iodine inventories in the pools, sequential refuelings are postulated. The loss of cooling is assumed during the eleventh refueling, just after the refueling cavity water level is lowered and the RHR system is not available for cooling the cavity and spent fuel pool. The analysis involves an evaluation of the time to pool boiling, the makeup water requirements if the pool boils, and the thyroid dose consequences at the LPZ outer boundary due to iodine releases from the boiling pools. This analysis is still bounding for all spent fuel storage rack configurations up to and including maximum storage configuration of maximum density storage racks.

The assumptions used in this analysis were consistently chosen to be the worst case design basis assumptions, similar to those in regulatory guides for design basis accidents (e.g., Regulatory Guides 1.3, 1.25, etc). The combination of all of these design basis assumptions occurring at the same time is extremely unlikely, making this accident, as analyzed, one of very low probability.

Many of the assumptions are considered to be overly conservative; for example, operating experience with present BWR fuels (Reference 9.1-5) indicates that the assumption of 1% of the fuel with cladding failures is more conservative by at least a factor of 100 for 8x8 fuel bundles.

Further, while some iodine release rate spiking factors of over 50 have been observed during startups and shutdowns, there are presently no data for the less severe temperature transients that would be associated with a boiling spent fuel pool. The assumption of 10% of the activity in the CHAPTER 09 9.1-24 REV. 19, SEPTEMBER 2018

LGS UFSAR fuel gaps is at least 30 times the gap activity used for a realistic accident analysis in Chapter 15. A more realistic evaluation of this accident would result in releases of radioactivity, if any, many orders of magnitude below the calculated values. Further, the realistic releases would be well below the 10CFR50, Appendix I technical specifications, indicating that such an incident is of little or no consequence.

This conservatively assumes that it would take 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> to place the reactor in cold shutdown, realign valves, adjust the spent fuel pool weir, and install the interconnecting piping between the RHR system and the fuel pool cooling system (Section 5.4.7.1.1). The conservative results show that the pools would not boil until at least 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the loss of cooling at the average water heatup rate of 5.33oF/hr. If cooling is not restored before the pool boils, then makeup water from the seismic Category I ESW system can be added to the pool to keep the fuel covered at all times.

As shown in Table 9.1-4 the thyroid dose consequences of releasing the water that would evaporate from the boiling pool are well below the guideline values of 10CFR100.

The following assumptions were used to calculate the heat generation and boiling rates in the two spent fuel pools:

a. Each pool contains the maximum inventory of assemblies with the discharges shown in Table 9.1-2A. The last one-third cores are from the just-completed sequential refuelings. For Unit 1, the fuel has decayed for 47 days. For Unit 2, the decay time is 21 days, the length of time from shutdown until the water level in the refueling cavity has been lowered, and the RHR system would not be able to cool the refueling cavity and spent fuel pool. The minimum time to complete fuel unloading and loading for Unit 1 is 26 days. These decay times result in the maximum heat generation rates and maximum evaporation rates at times when the RHR system is not available.

b. The decay heat is calculated by using the decay heat equations presented in BTP APCSB 9-2. The decay heat generation rate for each pool is given in Table 9.1-5 for various times after the postulated loss of cooling.

c. All heat generated by the fuel is assumed to be absorbed by the water, to minimize the time to boiling. No heat is assumed to be lost to the surroundings by conduction through the concrete and steel or by evaporation. The temperature gradients from the fuel at the bottom of the pool to the cooler water at the top create convective water and heat currents that should thoroughly mix the water and promote an even distribution of heat, rather than localized points of surface boiling .

This does not preclude the performance of non-sequential outages less than 26 days in length.

The following assumptions were used to calculate the offsite doses for the loss of cooling to the spent fuel pools:

a. The saturation inventory of I-131 in the 3458 MWt core is 8.66x107 Ci.

CHAPTER 09 9.1-25 REV. 19, SEPTEMBER 2018

LGS UFSAR

b. During refueling, 276 fuel elements (one-third core) are transferred to the spent fuel pool. Iodine in fuel from past refuelings is negligible, due to the long times for decay.

c. It is assumed that 1% of the fuel rods in the core are defective and that this 1% is in the one-third core transferred to the spent fuel pool.

d. The iodine activity in the spent fuel pool water at the initiation of boiling is assumed to be negligible compared to the activity released from the fuel during pool boiling.

Activity in the core coolant or from a shutdown spike would have been cleaned up to acceptably low levels by the RWCU and spent fuel pool cleanup systems before fuel transfer begins.

e. The spent fuel pool cooling systems are assumed to fail 47 days after shutdown for the first reactor and 21 days after shutdown for the second reactor. The 26 day difference is the minimum time to refuel the first unit. The 21 days is the time, after a complete fuel transfer, during which the water level in the refueling cavity is lowered, and the RHR system would not be able to cool the refueling and spent fuel pools in case of an accident.

f. The gap activity, or 10% of the rod activity, is available for leakage from the defective 1% of the rods. The leakage rate is assumed to be 2.9x10-8 per sec, which corresponds to a release rate of 2.6x10-3 Ci/sec for I-131 or 3.5x10-1 Ci/sec at 30 minutes for noble gases. This is the full power design fuel leak rate. It should be noted that the available activity in the gaps of the defective fuel rods may have already been significantly depleted by the shutdown spike.

g. A constant spike factor of various magnitudes up to 50 was applied to the I-131 leakage rate from the fuel to account for the potential spiking effects during the temperature transient. The leakage rate returns to the normal full power unspiked rate of 2.9x10-8 per sec when boiling begins, since the fuel should now be close to its new steady-state temperature. Although analyses were performed for release rate spike factors of up to 50, there are presently no data to support the spiking phenomena in the spent fuel pool boiling situation. Spikes have been observed for the large, rapid temperature and pressure changes associated with shutdowns and startups, but such significant spiking effects would not be expected during the long, slow temperature change of less than 65°F that would be associated with a loss of cooling. A comparison with the Eickelpasch and Hock paper, "Fission Product Releases After Reactor Shutdown," IAEA Sm 178/19 (1974), shows that the measured I-131 release rate at 9 days after shutdown is approximately 0.2 to 0.3 of the release rate at power. Since the temperature of the fuel during boiling is expected to be well below operating temperature, the use of the power leakage rate is considered to be extremely conservative.

h. The activity released from the fuel is assumed to be uniformly mixed in the 40,600 ft3 (2.53x106 lbm) of water in each spent fuel pool. The temperature gradients created in the pool by the fuel maintains the mixing process.

CHAPTER 09 9.1-26 REV. 19, SEPTEMBER 2018

LGS UFSAR

i. The normal spent fuel pool design maximum temperature is 140oF. For conservatism this is assumed to be the pool temperature at the time of loss of cooling.

j. The activity release rate from the pool depends on the evaporation (boiling) rate.

No evaporation is assumed during the heatup period until the pool water reaches 212oF. All heat generated by the fuel is assumed to be absorbed by the water, and no losses through the concrete and steel are assumed. This results in the shortest time to boiling. The heat generation and evaporation rates after boiling starts are given as a function of time in Table 9.1-5.

k. The iodine partition factor at the pool surface is varied between 0.1 and 0.01 to determine the effects on the total release. Although the partition factor would vary as a function of time due to changing temperatures and iodine concentrations, a value greater than 0.1 should not be expected for this nonviolent evaporating process at the pool surface.

l. No credit is taken for iodine plateout on walls and equipment or for washout by condensing water vapor in the refueling area.

m. It is assumed that all activity in the steam released to the air in the refueling area is instantaneously released to the atmosphere without filtration or condensation in the ventilation system.

n. The atmosphere dispersion factors for dilution of the radioactive releases are the 5th percentile ground level X/Qs given below for the LPZ boundary distance. The time "0" is assumed to be the start of the accident when pool cooling is lost.

TIME X/Q (sec/m3) 0 - 8 hrs 4.0x10-5 8 - 24 hrs 2.9x10-5 1 - 4 days 1.4x10-5 4 - 30 days 5.4x10-6

o. The thyroid dose models and breathing rates given in Regulatory Guide 1.3 are used.

The following model was used to calculate the offsite thyroid doses from the release of radioiodine from the fuel:

a. The activity in the fuel available for leakage at the time of loss of cooling, S(0), is calculated using the reactor inventory equation from TID-14844, with the appropriate decay from shutdown until the loss of the cooling and the fractions of iodine available for release. During the pool heatup and boiling phases, the activity in the fuel gaps available for leakage, S(t), is adjusted for decay and losses by leakage to the pool.

CHAPTER 09 9.1-27 REV. 19, SEPTEMBER 2018

LGS UFSAR S (tn) = S (tn 1) e ( d 1) t (Ci)

+

(EQ. 9.1-1) where:

d = decay lambda (1/sec) 1 = leakage rate from the fuel (1/sec) t = tn - tn-1 = time increment (sec)

b. The activity in the spent fuel pool as a function of time, A(t), is given by the solution to the following:

A(t n )= A(t n1)e d t ev A(t n1)t (EQ. 9.1-2)

+1S (t n )1+(SF 1 )(1 (t n ))t / V where:

ev = evaporation lambda from the pool (1/sec) x PF PF = iodine partition factor at the pool surface SF = spiking factor

= step-function = = 0 for tn time to boil

= 1 for tn > time to boil V = pool volume (ft3)

A = activity in spent fuel pool (Ci/ft3)

Since spent fuel pool makeup water will be available, the evaporation lambda is found by dividing the evaporation rate (ft3/sec) by the constant pool water volume (ft3).

c. The activity released to the atmosphere at any incremental time tn is given by the following equation:

R(tn) = R(tn-1) + ev V A(tn)(1-f)t (EQ. 9.1-3) where:

R = activity released (Ci) f = SGTS filter efficiency fraction (assumed to be zero)

CHAPTER 09 9.1-28 REV. 19, SEPTEMBER 2018

LGS UFSAR The above equations are solved iteratively, using time steps of 450 seconds.

d. The thyroid dose at the LPZ is calculated using the equations and models from Regulatory Guide 1.3.

9.1.4 FUEL HANDLING SYSTEM 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 postirradiation cooling.

Safe handling of fuel includes design considerations for maintaining occupational radiation exposures as low as reasonably achievable during transportation and handling.

Design criteria for major fuel handling system equipment are provided in Tables 9.1-7 through 9.1-9, 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 are shown below and expanded further in Section 9.1.4.2. The equipment discussed in this section was also reviewed to verify compliance with the guidelines of ANS 57.1 and ANS 57.2 (1976).

There were no deviations identified except those discussed in Section 9.1.5.2 for the reactor enclosure crane and for the new fuel inspection/channeling stand and the upending platform.

The reactor enclosure crane main hoist or auxiliary hoist is used with a general purpose grapple to transfer new fuel from the new fuel inspection stand to the fuel prep machines in the fuel storage pool. From this point on, the fuel is handled by the telescoping grapple on the refueling platform.

The refueling platform is seismic Category I from a structural standpoint, in accordance with 10CFR50, Appendix A and 10CFR50, Appendix B. The allowable stress due to SSE loading is 120% of yield or 70% of ultimate, whichever is less. A dynamic analysis is performed on the structures using the response spectrum method, with load contributions resulting from each of three earthquakes being combined by the RMS procedure.

Working loads of the platform structures are in accordance with the AISC Manual of Steel Construction. All parts of the hoist systems are designed to have a safety factor of 5 based on the ultimate strength of the material. A redundant load path is incorporated in the main hoist so that no single component failure could result in a fuel bundle drop. Maximum deflection limitations are imposed on the main structures to maintain the relative stiffness of the platform. The welding of the platform is in accordance with AWS D14-1 or ASME Section IX. Gears and bearings meet AGMA Gear Classification Manual and ANSI B3.5 requirements. Materials used in the construction of load-bearing members meet ASTM specifications. For personnel safety, OSHA (29CFR1910.179) is applied. Electrical equipment and controls meet ANSI C-1, National Electric Code, and NEMA Publication No. ICl, MGl requirements.

The general purpose grapples and the main telescoping fuel grapples have redundant hooks. The main fuel grapple has an indicator that confirms positive grapple engagement.

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. At its normal up position, adequate shielding is still maintained over the fuel.

CHAPTER 09 9.1-29 REV. 19, SEPTEMBER 2018

LGS UFSAR Redundant electrical limit switches are a part of the fuel grapple to preclude the possibility of raising radioactive material out of the water. The cables on the auxiliary hoists incorporate an adjustable, removable stop that jams the hoist cable against the hoist structure to prevent hoisting when the free-end of the cable is at a preset distance below water level.

The provision of a separate cask loading pool, capable of being isolated from the fuel storage pool, eliminates the potential for accidental dropping of the cask and rupturing of the fuel storage pool.

See Chapter 15 for accident considerations.

9.1.4.2 System Description Table 9.1-6 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.

9.1.4.2.1 Spent Fuel Cask 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 cask can also be used for offsite shipment of irradiated reactor components such as control rod blades, incore monitors, etc.

The maximum loaded weight and hence the capacity of the cask is determined by the 125 ton lifting capacity of the reactor enclosure crane. The maximum loading height, ie, 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.

It 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 cask meets all applicable regulations of the DOT and 10CFR71.

No specific type of cask has been chosen. Over the lifetime of the plant, several different sizes and models that the fuel handling facilities can accommodate may be used.

Additionally, an Independent Spent Fuel Storage Installation Transfer Cask has been chosen, as described in the Updated Final Safety Analysis Report for the Standardized NUMOS Horizontal Modular Storage System for Irradiated Nuclear Fuel Rev 10, Transnuclear, Inc., NUH003.0103 and its NRC Certificate of Compliance No. 1004.

9.1.4.2.2 Cask Crane The reactor enclosure crane is discussed in Section 9.1.5.

9.1.4.2.3 Fuel Servicing Equipment The fuel servicing equipment described below is designed in accordance with the criteria listed in Table 9.1-7.

9.1.4.2.3.1 Fuel Prep Machine CHAPTER 09 9.1-30 REV. 19, SEPTEMBER 2018

LGS UFSAR The fuel preparation machine (Figure 9.1-5) is mounted on the wall of the fuel storage pool and can be used for stripping reusable channels from the spent fuel and for channeling the new 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.

The fuel preparation machine consists of a work platform, a frame, and a movable carriage. The frame and movable 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 uptravel stop to prevent raising irradiated fuel above the safe water shield level.

The uptravel stop can be adjusted to allow new fuel to be transferred from the reactor enclosure overhead crane to the fuel prep machine carriage without submerging the crane hook in the fuel pool water. The movable carriage is operated by a foot pedal controlled air hoist.

9.1.4.2.3.2 New Fuel Inspection/Channeling Stand The new fuel inspection/channeling stand serves as a support for a new fuel bundle undergoing receipt inspection and provides a working platform for technicians engaged in performing the inspection.

The new fuel inspection/channeling stand consists of a vertical column with a bearing seat, a lift unit to position the work platform at any desired level, and upper clamp to hold the fuel bundle in position. Bolted to the upper portion of the column is an articulating hoist mechanism consisting of two square tubes connected by a hinge pin. The articulating piece is normally lowered to avoid interference with the reactor enclosure crane. When channeling is being performed the articulating piece is raised by a double-acting air cylinder to its full height, and an electric hoist mounted on the piece that lowers the channel over the fuel bundle.

9.1.4.2.3.3 Channel Bolt Wrench The channel bolt wrench (Figure 9.1-7) is a manually operated device approximately 12 feet (3.6 meters) in overall length. The wrench is used for removing and installing the channel fastener assembly while the fuel assembly is held underwater in the fuel preparation machine.

The channel bolt wrench has a socket that mates and captures the channel fastener cap screw.

9.1.4.2.3.4 Channel Handling Tool The channel handling tool (Figure 9.1-8) 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, actuating knob, actuating shaft, angle guides, and jaws that engage the fuel channel. The jaws are actuated (extended or retracted) by manually rotating the actuating knob.

The channel handling tool is suspended by its bail from a spring balancer on the channel handling boom located on the fuel floor.

9.1.4.2.3.5 Fuel Pool Sipper CHAPTER 09 9.1-31 REV. 19, SEPTEMBER 2018

LGS UFSAR The fuel pool sipper (typical unit shown in Figure 9.1-9) provides a means of isolating a fuel assembly in demineralized water to concentrate fission products in relation to a controlled background.

The fuel pool sipper consists of a control panel assembly and a sipping container.

9.1.4.2.3.6 Fuel Inspection Fixture The fuel inspection fixture (typical unit shown in Figure 9.1-10) is used in conjunction with the fuel preparation machine to permit remote inspection of a fuel bundle or assembly. The fixture consists of two parts: a lower bearing assembly and a guide assembly at the upper end of the carriage. The fuel inspection fixture permits the rotation of the fuel bundle or assembly in the carriage and provides, in conjunction with the vertical movement of the carriage, complete access for inspection.

9.1.4.2.3.7 Channel Gauging Fixture The channel gauging fixture (Figure 9.1-11) is a go/no-go gauge which can be used to evaluate the condition of a new fuel channel before channeling.

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 the lower end of a usable fuel channel.

The channel gauging fixture is installed vertically between the two fuel preparation machines, and hangs from the fuel storage pool curb when in use. When not in use, the channel gauging fixture may be stored elsewhere in the fuel storage pool or on the fuel floor.

9.1.4.2.3.8 General Purpose Grapple The general purpose grapple (Figure 9.1-12) is a handling tool used generally with the fuel. The grapple can be attached to the reactor enclosure main hoist, auxiliary hoist, jib crane, and the auxiliary hoists on the refueling platforms. The general purpose grapple is used to place new fuel in the inspection stand and to transfer the new fuel to the fuel pool. It can be used to handle fuel during channeling.

9.1.4.2.3.9 New Fuel Upending Platform The upending platform is a structural steel frame that supports a 1 1/2 ton jib crane that has two hoists mounted on it. There is a 1-ton electric hoist and a 1/2 ton hand hoist. Only one hoist will be used at a time. The platform is a Seismic Category II structure.

The jib crane and electric hoist are used to raise the metal container into a vertical position and set it into the upending stand of the platform.

It is also possible to use the reactor enclosure crane to transfer the metal container into the unloading area.

The jib crane and the hand hoist are used to transfer the fuel bundles into one of the new fuel inspection/channeling stands which are located on either side of the platform.

9.1.4.2.3.10 Roller Conveyer CHAPTER 09 9.1-32 REV. 19, SEPTEMBER 2018

LGS UFSAR A gantry crane or the reactor enclosure crane will be used to transfer the metal container from a predesignated storage area to a roller conveyer. The roller conveyer is located adjacent to the upending platform and is used to position the metal container under the jib crane.

9.1.4.2.3.11 Gantry Crane The 2-ton gantry crane can be used to transfer the metal container from the predesignated storage area to the roller conveyer. The gantry crane is designed to meet the requirements of the Limerick heavy loads program.

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, independently of the platform. Local area underwater lights are small-diameter lights for additional illumination. Drop lights are used for illumination where needed.

A radiation-hardened portable underwater closed-circuit television camera is provided. The camera may be lowered into the reactor vessel and/or fuel storage pool to assist in the inspection and/or maintenance of these areas. The camera may also be equipped with a right-angle lens to allow viewing at 90°.

A general purpose plastic viewing aid is provided to float on the water surface to provide better visibility. The sides of the viewing aid are brightly colored to allow the operator to observe it if it fills with water and sinks. A portable, submersible, underwater vacuum cleaner is 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 fit into a standard shipping container for offsite burial. Fuel pool tool accessories are also provided to meet servicing requirements. A fuel sampler is provided to detect defective fuel assemblies during open vessel periods while the fuel is in the core. The fuel sampler 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 returning it to the primary coolant system. After a "soaking" period, the water sample is radiochemically analyzed to determine possible fuel bundle leakage.

9.1.4.2.5 Reactor Vessel Servicing Equipment The essentiality, the quality group, and the seismic category for this equipment are listed in Table 9.1-8. The following is a description of the equipment designs in reference to that table.

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.

9.1.4.2.5.2 Steam Line Plug The steam line plugs are used during reactor refueling or servicing. They are inserted in the steam outlet nozzles from inside the reactor vessel to prevent a flow of water from the reactor well into the main steam lines during servicing of SRVs, MSIVs, or other components of the main steam lines, while the reactor water level is raised to the refueling level.

CHAPTER 09 9.1-33 REV. 19, SEPTEMBER 2018

LGS UFSAR 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 noncorrosive materials.

Steam line plugs of various designs are utilized. Plug designs weighing more than a fuel bundle meet the requirements of NUREG-0612. Current steam line plugs including handling tool weigh less than 1200 pounds and do not constitute heavy loads.

9.1.4.2.5.3 Shroud Head Bolt Wrench This is a tool for the operation of the shroud head bolts. It is designed for a 40 year life, and is made of corrosion resistant materials for easy handling and corrosion resistance. 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 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 AISC Manual of Steel Construction.

9.1.4.2.5.5 Head Nut and Washer Rack The RPV head nut and washer rack maybe used for transporting and storing up to six nuts and washers. The rack is a box-shaped non-corrosive structure with dividers to provide individual compartments for each nut and washer. Each corner has a lug and shackle for attaching a four leg lifting sling.

The rack is designed with a safety factor of 5.

9.1.4.2.5.6 Head Stud Rack The head stud rack may be used for transporting and storing six reactor pressure vessel studs. It is suspended from the reactor enclosure crane auxiliary hoist for lifting studs from the reactor well to the operating floor.

The rack is made of a non-corrosive material.

9.1.4.2.5.7 Dryer and Separator Sling The dryer and separator sling 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 sling consists of a cruciform structure, which is suspended from a hook box with four wire ropes and turnbuckles. The hook box, with two hook pins, engages the reactor enclosure crane main hook extension or sister hook, depending on wet or dry transfer method. Synthetic slings are an approved equivalent alternate to the wire ropes. When synthetic slings are used they are CHAPTER 09 9.1-34 REV. 19, SEPTEMBER 2018

LGS UFSAR designed to be attached directly to the reactor enclosure crane sister hook. On the end of each arm of the cross is a socket with a pneumatically operated pin for engaging the four lifting eyes on the steam dryer or shroud head. The hook extension may be used when the dryer or shroud head is transferred underwater.

The sling is designed so that one hook pin and one main beam of the cross is capable of carrying the total load and so that no single component failure can cause the load to drop or swing uncontrollably out of an essentially level attitude.

The safety factor of the lifting members is 5 or greater for the ultimate breaking strength of the material. The structure is designed in accordance with AISC Manual of Steel Construction. The completed assembly is proof-tested at 125% or greater of rated load, and all structural welds are magnetic particle inspected after the load test.

9.1.4.2.5.8 Reactor Pressure Vessel Head Strongback/Carousel The Reactor Pressure Vessel (RPV) head strongback/carousel is used for lifting the vessel head.

The strongback/carousel is an integrated piece of equipment consisting of a cruciform shaped strongback, a circular monorail, and a circular storage tray.

The strongback is a box beam structure which has a hook box with three pins in the center for engagement with the reactor building crane main hoist hook. Each arm has a liftrod for engagement to the four lift lugs on the RPV head. The monorail is mounted on extensions of the strongback arms and four additional arms equally spaced between the strongback arms. The monorail circle matches the stud circle of the reactor vessel and serves to suspend the stud tensioners and the nut storage tray. Each tensioner has an air-operated hoist with individual controls.

The strongback carousel is considered to be single-failure proof, since failure of a single load arm or load pin will not release the load. It is designed in compliance with NUREG-0612, and the AISC manual of steel construction. The RPV strongback carousel is also designed in compliance with ANSI N14.6 with the exception of material testing of the hook pins, clevis pins, and clevis rods.

The lack of material testing does not affect the function or safety of the RPV strongback carousel as the assembly is load proof testing and undergoes pre-service NDE to ensure its continued integrity. The design load capacity for the equipment will be 135 tons.

9.1.4.2.5.9 Service Platform The text is this section described features of the reactor service platform. The reactor service platform is obsolete equipment that has been permanently removed from the facility. The information in this section was removed.

9.1.4.2.5.10 Service Platform Support The text is this section described features of the reactor service platform support. This equipment is obsolete equipment that has been permanently removed from the facility. The information in this section was removed.

9.1.4.2.5.11 Steam Line Plug Installation Tool The steam line plug installation tool is suspended from the reactor enclosure crane, auxiliary platform, or refueling platform for transporting and installing the steam line plugs in the steam line CHAPTER 09 9.1-35 REV. 19, SEPTEMBER 2018

LGS UFSAR nozzles of the reactor vessel. This tool is made of aluminum; it is designed for a safety factor of 5 and is in accordance with the Aluminum Construction Manual by the Aluminum Association.

9.1.4.2.5.12 Fuel Floor Auxiliary Platform (FFAP)

The FFAP or auxiliary platform provides worker access and component hoisting capability in the Unit 1 and 2 reactor cavities, fuel pools, and cask pit areas to perform maintenance activities during, but not limited to, refueling outages. The auxiliary platform is constructed of steel framing using materials in accordance with ASTM Standards. Welding is in accordance with AWS D1.1 structural welding. The auxiliary platform is not designed to perform movements of reactor fuel, irradiated components, and heavy loads.

9.1.4.2.5.13 Reactor Cavity Work Platform (RCWP)

The Reactor Cavity Work Platform (also known as the GE Scorpion RCWP) is approved for use at Limerick. The descriptive statements below apply to GE RCWP.

The RCWP provides workers access to in-vessel components for inspections and repairs concurrent with irradiated component handling activities. The RCWP is assembled on the refuel floor and then placed into the reactor cavity after flood-up and is removed from the cavity while it is still flooded up. Procedural controls will assure that cavity water level is maintained within the required range as the RCWP is inserted into and removed from the water.

The RCWP displaces some cavity water since the bottom of the GE RCWP work trough is located at approximately elevation 348 feet 10 inches, with hollow supporting beams extending approximately 10 inches below the bottom. This reduces shielding provided between irradiated components and the surface of the water (directly above the irradiated component). Radiation controls will be in place to ensure personnel are not inadvertently overexposed in the event of unexpected increases in dose rated on the RCWP. At the normal uptravel limit of the Refueling Platform hoists, the water shielding provided for the RCWP workers, in combination with radiation protection program controls and protective measures, is adequate.

The GE RCWP may be equipped with two jig cranes each de-rated to a load rating of 500 lbs and each weighing approximately 500 lbs. The RCWP is not designed to handle fuel, irradiated components or heavy loads. The GE RCWP jib cranes are permitted to extend into the boundary zone during fuel handling based on a determination in a GE analysis that a collision between a loaded fuel grapple and the GE RCWP would not result in dropping the fuel bundle or mast fuel grapple assembly onto the core; and therefore the original FHA remains bounding.

Additional administrative controls will be in place during handling of irradiated components in the reactor cavity to assure that irradiated components are not inadvertently raised in close proximity to personnel on the RCWP. Additional administrative controls will be in place during fuel handling to minimize the likelihood of a collision of the fuel grapple assembly with the GE RCWP jib crane.

Administrative controls similar to those implemented for fuel movement when the boundary zone limit system is not available will also be implemented for fuel movement when the RCWP jib crane is extended into the boundary zone.

The RCWP was constructed in accordance with AISC Manual of Steel Construction using materials in accordance with ASTM standards. Welding is in accordance with ANSI/AWS D1.6 ,

Structural Welding Code for Stainless Steel, and ANSI/AWS D1.1, Structural Welding Code for Carbon Steel. The RCWP meets Seismic IIA criteria and will not fall into the reactor vessel during an SSE. The GE RCWP does not rely on buoyancy to maintain its seismic qualification.

CHAPTER 09 9.1-36 REV. 19, SEPTEMBER 2018

LGS UFSAR 9.1.4.2.6 In-Vessel Servicing Equipment The instrument strongback attached to the reactor enclosure crane is used for servicing neutron monitoring incore detectors as they require replacement. The strongback initially supports the incore detectors into the reactor cavity. The incore detectors are then decoupled from the strongback and is guided into place while being supported by the instrument handling tool. The instrument handling tool is attached to a refueling platform auxiliary hoist and is used for removing and installing fixed incore detectors, as well as for handling neutron source holders and the source range monitor/intermediate range monitor dry tubes. The auxiliary platform may be used to install new local power range monitors and new source range monitor/intermediate range monitor dry tubes.

Each incore instrumentation guide tube is sealed by an O-ring on the flange. If the seal needs replacing, an incore 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 hoists on the refueling platform are used with appropriate grapples to handle control rod blades, incore detector dry tubes, sources, and other reactor vessel internals. Interlocks on both the grapple hoists and auxiliary hoist are used to provide safety; the refueling interlocks are described and evaluated in Section 7.7.

9.1.4.2.7 Refueling Equipment Fuel movement and reactor servicing operations are performed from a platform that spans the refueling, servicing, and storage cavities.

9.1.4.2.7.1 Refueling Platform The refueling platform is a gantry crane that is used to transport fuel and reactor components to and from the fuel storage pool and the reactor vessel. The platform spans the fuel storage pool and reactor cavity on rails bedded in the refueling floor. A telescoping mast and grapple suspended from a trolley is used for transporting and orienting fuel assemblies for placement into the core, storage rack, or shipping cask. The platform is controlled 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 interlocks to verify the grapple load, prevent unsafe operation over the vessel during control rod movements, and limit vertical travel of the grapple. The grapple in its normal up position provides 8 feet minimum water shielding over the top of active fuel (6 feet 6 inches shielding over the top of the fuel assembly) during transit. The fuel grapple hoist has a redundant load path so that no single component failure results in a fuel assembly drop.

A refuel platform is provided for each reactor unit. Due to the common refueling area for both units, the refuel platforms can be operated over either unit's spent fuel pool to move fuel assemblies.

Administrative controls have been established to assure that both platforms are not operated in close proximity when both platforms are transporting fuel.

Two 1000 pound capacity auxiliary hoists, administratively limited to 500 pounds for the handling of control rod blades within the reactor pressure vessel, one main trolley-mounted and one monorail mounted, are provided for servicing, such as LPRM replacement, fuel support piece replacement, CHAPTER 09 9.1-37 REV. 19, SEPTEMBER 2018

LGS UFSAR jet pump servicing, control rod blade replacement, main steam line plug installation and removal, and shroud head bolt latching and unlatching. The two auxiliary hoists mounted on the refueling platform are not intended for fuel handling. The main trolley auxiliary hoist and the monorail auxiliary hoist are each provided with a geared rotary limit switch that provides normal up and down limit stops. In addition, a stop block fastened to the hoist cable will operate a safety limit switch if the normal up limit should fail.

The auxiliary platform may be used to assist the refueling platform efforts, such as, installing new LPRMs, installing and removing main steam line plugs, and latching and unlatching shroud head bolts. The auxiliary platform is not designed to perform movements of reactor fuel, irradiated core components, and heavy loads.

If the motor is not stopped by either up limit, the stop block will jam against the hoist frame and trip the motor on (a) the load cell switch sensing of a jam condition, or (b) stalling of the hoist motor. In either event, there will be no resulting impact load on the cable because the block stops against the energy-absorbing portion of the hoist (i.e., spring-loaded plate or pivoted sheave arm).

Each cable is inspected by procedure prior to every refueling outage.

A Service Pole Caddy platform is attached on the rear side of the refueling platform at LGS. The platform provides an auxiliary work station for unlatching and latching the steam separator head bolts during refueling activities. The platform can also be utilized for other underwater servicing needs such as jet pump beam bolt untorquing and steam line plug installation. The platform is provided with high torque service poles and a motorized hoist to handle the poles.

9.1.4.2.8 Storage Equipment Specially designed equipment storage racks are provided. Additional storage equipment is listed on Table 9.1-6. For fuel storage rack description and fuel arrangement, see Sections 9.1.1 and 9.1.2.

Defective fuel assemblies may be placed in defective fuel storage containers, which in turn are stored in a suitable storage rack. Defective fuel assemblies are assemblies that are unable to contain the fuel pellets and cannot be handled with normal fuel handling equipment. These may be used to isolate leaking or defective fuel while in the fuel pool and during shipping. Channels can be removed from the fuel bundle while in a defective fuel storage container.

The storage racks which hold control rod blades and/or defective fuel assemblies are special storage racks designed to hold various core components that cannot fit in the spent fuel storage racks. Control rod blades, incore detectors and dry tubes may also be stored on fuel pool walls (see Section 11.4.2.1.4). The storage rack is designed to hold control rod blades, control rod guide tubes, sipping containers and spent local power range monitor (LPRM) containers in any of its 49 storage cavities and defective fuel containers (with spent fuel) in any of 5 specially designed cavities. The special storage rack is nominally 83 inches square and 159.25 inches high. The empty, dry weight of the rack is approximately 12,190 pounds. The storage rack is placed in the fuel pool as indicated in Figures 9.1-34 and 9.1-36.

The storage rack's cavities are formed by a checker board placement of canisters. The canisters, stainless steel plate material bent in a rectangular shape, are welded to the bottom grid assembly such that the distance between cavity centers (i.e., canister centerline to intermediate cavity centerline) is approximately 11.665 inches. Storage cavities along the rack periphery will have their CHAPTER 09 9.1-38 REV. 19, SEPTEMBER 2018

LGS UFSAR outer wall made from stainless steel plate welded to the outer walls of adjacent canisters. Along the top of the rack is a 0.187 inch thick 3.50 inch wide stainless steel perimeter bar. The perimeter bar allows lateral loads to be shared by all canisters and helps to maintain the specified center-to-center cavity spacing.

The bottom grid assembly is made from stiffened stainless steel plate. For each cavity, a 3.625 inch diameter chamfered hole in the bottom load support plate of the grid assembly acts to center rack stored items and accommodates cavity cooling water flow.

The five cavities designed to hold defective fuel containers have a bottom load support that is 3.75 inches lower in the bottom grid assembly than the other 44 rack cavities.

The rack is supported by 4 foot assemblies (foot pad and foot adjusting screw), one centered under each of the four corner cells. The foot assembly will accommodate slight variations in the pool floor liner surface and is adjustable over a 1-inch band to facilitate rack leveling. The upper surface of the bottom grid assembly will be 9.75 (+/- 0.50) inches above the pool floor. This will allow at least 22 feet 8 inches of water above a fuel assembly placed in any one of the rack cavities. Ample space will exist under the rack for the flow of water to each of the cells.

The rack is seismic Category I and qualified structurally by detailed dynamic and static analyses.

The rack is designed to withstand the effects of a Safe Shutdown Earthquake (SSE) and remain functional in accordance with NRC Regulatory Guide 1.29 and the Code of Federal Regulations Title 10, Part 50, Appendix A.

The dynamic evaluation for seismic loads used the non linear transient analysis computer program ANSYS. The transient boundary conditions (time histories) for the analysis were generated artificially utilizing the computer program SIMQKE with design response spectra characteristics provided by Limerick Specification 8031-G-19. The maximum calculated rack stress levels and displacements occurred during the Safe Shutdown Earthquake load analysis. The analysis determined that all computed stress levels were within appropriate limits for the rack structure and its individual constituents. Maximum potential displacements have been analyzed and determined to be acceptable.

The static evaluation determined rack suitability as a result of stress due to static load applications and fuel drop accident impact loads. All static load stresses are within appropriate limits for the rack structure and its individual constituents. The equivalent static load resulting from a dropped fuel assembly was obtained through an energy analysis. The impact damage resulting from the fuel drop accident is confined to a small area near the impact point. Stresses away from the impact point are low; therefore, the overall rack structure will not be adversely affected and no change in keff will be realized due to the rack structure realignment.

The fuel pool sipper may be used for out-of-core wet sipping at any time. It is used to detect a defective fuel bundle while circulating water through the fuel bundle in a closed system. 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 functions of the under-reactor vessel servicing equipment are as follows:

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a. Remove and install CRDs

b. Remove and install the thermal sleeve

c. Install and remove the neutron detectors. Table 9.1-9 lists the equipment and tools required for servicing. Of the equipment listed, the equipment handling platform and the CRD handling equipment are powered electrically and/or pneumatically.

The CRD handling equipment is used to perform a CRD exchange (i.e. remove and install) from the CRD housing. The CRD handling equipment is designed in accordance with the following requirements:

- American Institute for Steel Construction (AISC) - Steel Construction Manual, 8th Edition;

- Aluminum Association - Specifications for Aluminum Structures, 5th Edition;

- American Welding Society, Specifications D1.1 and D1.2;

- Federal Specification on MIL-W-83420D, Wire Rope, Flexible for Aircraft Control;

- American National Standard ANSI B30.9-1971 (Wire Ropes);

- AGMA - American Gear Manufacturer's Association - Services Factors and Lead Classifications;

- NEMA - National Electric Manufacturer's Association (standards).

The equipment used to perform a CRD exchange is capable of inserting a force of approximately 1300 pounds. Also, all of the lifting components are equipped with adequate brakes or gears to prevent uncontrolled movement on a loss of air or component failure. The equipment is capable of re-inserting a CRD if it is still coupled to its control rod.

The equipment handling platform 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 (Volume 37, No.

202, Part 191 ON), AISC, and ANSI C-1 (National Electric Code).

The thermal sleeve installation tool locks, unlocks, and lowers the thermal sleeve from the control rod guide tube.

The incore flange seal test plug can be used to determine the pressure integrity of the incore flange O-ring seal. It is constructed of noncorrosive material.

The key bender is designed to install and remove the antirotation key that is used on the thermal sleeve.

9.1.4.2.10 Description of Fuel Transfer 9.1.4.2.10.1 Arrival of Fuel Onsite CHAPTER 09 9.1-40 REV. 19, SEPTEMBER 2018

LGS UFSAR New fuel is delivered by truck (or by rail) and moved into the refueling hatch at grade level.

Secondary containment can be maintained while the new fuel is being hoisted to the refueling floor.

9.1.4.2.10.2 Refueling Procedure The general refueling floor layout is shown in drawings M-122 and M-137. Component drawings of the principal fuel handling equipment are shown in Figures 9.1-5 through 9.1-12.

The fuel handling process takes place primarily on the refueling floor above the reactor. The principal locations and equipment are shown in drawings M-122 and M-137. The reactor cavity, fuel pool, and cask storage pit are connected to each other by canals. The fuel transfer canal is open during reactor refueling, and the cask pit canal is open during spent fuel and ISFSI transfer cask handling in the Reactor Enclosure. The canals can be closed by redundant gates, which make watertight barriers. However, gates between the cask pit and the spent fuel pool will normally remain open during the transfer cask handling in the Reactor Enclosure.

New fuel, in shipping crates, is brought up to the refueling floor through the hatches, and spent fuel, in a shipping cask, is lowered through the hatches to a truck or rail car near grade level. The new fuel is placed in predesignated storage areas where it remains until it is processed and put in the spent fuel pool.

The method for transferring the fuel bundles between the crate, the new fuel inspection/channeling stand and spent fuel storage pool is accomplished by using a gantry crane or the reactor enclosure crane (125 ton main or 15 ton auxiliary hoist). The fuel is transferred from a predesignated storage area to a roller conveyer. The roller conveyer is located adjacent to the upending platform. At this point, the top is upended and placed into a vertical upending stand. After the fuel is unloaded, it is inspected and transferred to the fuel preparation machine.

The main or auxiliary hoist is used with a general purpose grapple to transfer a new fuel assembly to a fuel preparation machine. The fuel preparation machine lowers the new fuel into the pool, and henceforth, the fuel is positioned in a storage rack by the telescoping grapple on the refueling platform. It should be noted that secondary containment must be established when moving new fuel over recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

The storage racks in the fuel pool hold the fuel bundles or assemblies vertically, in an array that is subcritical under all possible conditions.

The new fuel inspection/channeling stand holds one bundle vertically. The inspector(s) rides up and down on a platform, and the bundle is manually rotated on its axis. Thus the inspectors can see all visible surfaces.

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, adequate water shielding is maintained over the fuel being handled.

The refueling platform moves over the fuel pool, lowers the grapple on the telescoping mast, and engages the bail on a new fuel assembly that is in the fuel storage rack. The assembly is lifted clear of the rack and moved through the transfer canal and over the appropriate empty fuel location in the core. The mast then lowers the assembly into the location, and the grapple releases the bail.

CHAPTER 09 9.1-41 REV. 19, SEPTEMBER 2018

LGS UFSAR The operator can then move the platform until the grapple is over a spent fuel assembly that is to be discharged from the core. The assembly is grappled, lifted, and moved through the transfer canal to the fuel pool. Here it can be placed in a spent fuel storage rack, fuel prep machine or a fuel bundle sipping container.

If channeling is to be performed, an operator working at the fuel prep machine using the channel bolt wrench, removes the channel fastener 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 the refueling platform grapple carries the bundle and places it in a storage rack. The channel handling boom hoist moves the channel to storage, if appropriate.

In actual practice, channeling and dechanneling may be performed in many sequences, depending on whether a new channel is to be used or a used channel is to be installed on a new bundle and returned to the core. A channel storage rack is conveniently located near the fuel prep machines for temporary storage of channels that are to be reused.

To preclude the possibility of raising radioactive material out of the water, redundant electrical limit switches are incorporated in the main hoist and interlocked to prevent hoisting above the preset limit. In addition, the cables on the auxiliary hoists incorporate adjustable stops that jam the hoist cable against the hoist structure, which prevents hoisting if the limit switch interlock system fails.

When spent fuel is to be shipped, it is placed in a cask. The refueling platform grapples a fuel bundle from the storage rack in the fuel pool, lifts it, carries it through the cask pit canal into the cask storage pit, and lowers it into the cask. When the cask is loaded, the crane sets the cask cover on the cask and then lifts the cask out of the pool. The cask is then decontaminated, sealed and lowered through the open hatchway to the truck or rail car at near grade level.

The provision of a separate cask storage pit, capable of being isolated from the spent fuel pool, eliminates the potential for accidental dropping of the cask and rupturing of the fuel storage pool.

Additional detailed information is provided below.

a. New fuel preparation

1. Receipt and inspection of new fuel The incoming new fuel is delivered to the station where it is unloaded in a designated receiving area. The new fuel is shipped inside of a dual container system. The system consists of an inner metal container and an outer metal container. Each set of containers holds two fuel bundles. The shipping weight of each unit is approximately three-thousand (3000) pounds. After the fuel is unloaded, the outer metal containers are first examined for damage received during shipment. The containers are transferred to the refueling hoist-way. The main or auxiliary hoist of the reactor enclosure crane will hoist the containers to the refueling floor where they are placed in a predesignated area.

A gantry crane or the reactor enclosure crane will transfer the inner metal containers to a roller conveyer. The roller conveyer is located adjacent to the upending platform. At this time, the top and front of the metal container are removed. Using a jib crane and an electric hoist, mounted on the CHAPTER 09 9.1-42 REV. 19, SEPTEMBER 2018

LGS UFSAR upending platform, the metal container is upended and placed into the unloading stand. Using the hand hoist, which is also attached to the jib crane, the fuel is removed from the metal container and placed into the new fuel inspection/channeling stand (one of the two).

2. Channeling Channeling of the fuel bundle takes place in the new fuel inspection/channeling stand. The channel is first examined for damage or defects. The new channel is raised over the fuel bundle and slowly lowered onto the fuel bundle. Once the channel is around the fuel, a channel fastener is used to clamp the channel to the fuel bundle, creating a fuel assembly. The complete fuel assembly is transferred to the fuel prep machine using the main or auxiliary hoist of the reactor enclosure overhead crane with a general purpose grapple. Once in the fuel prep machine, it is then lowered into the fuel pool and transferred to the fuel pool racks using the refueling platform.

3. Equipment preparation Before the plant shutdown for refueling, all necessary equipment must be placed in readiness. All necessary tools, grapples, slings, strongbacks, stud tensioners, etc are given a thorough check, and any defective (or well worn) parts are replaced. Air hoses on grapples are routinely leak tested. Crane cables are routinely inspected. All necessary maintenance and interlock checks are performed to ensure that there is no extended outage due to equipment failure.

The incore flux monitors, in their shipping container, are on the refueling floor. The channeled new fuel and the replacement control rod blades are ready in the fuel pool.

b. Reactor shutdown The reactor is shut down according to a prescribed procedure. During cooldown the RPV is vented and filled to above flange level to equalize cooling. The reactor well shield plugs can be removed. The upper layer of shield plugs may be removed once reactor power is lowered to less than or equal to 100%. The lower layer of shield plugs may be removed after the reactor has achieved hot shutdown. Prior to the removal of shield plugs in other than cold shutdown, there are specific conditions required to be satisfied per the general plant procedures. This is accomplished with the reactor enclosure crane and the lifting strongback. There are 12 pieces to handle.

This operation also includes the removal of the canal plugs and the slot plugs. A total of 16 separate plugs must be removed and placed on the refueling floor.

Removal of these plugs may be performed out of sequential order as conditions permit. The outer fuel pool gate may also be removed at this time. A sling is attached to the gate lifting lugs, and the reactor enclosure crane lifts the gate and places it on the fuel pool storage lugs.

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1. Drywell head removal Immediately after removal of the lower layer of reactor well shield plugs, the work to unbolt the drywell head can begin. The drywell head is attached by removable studs. The studs are unscrewed from the captive nut on the drywell and supported on the lip of the head.

The unbolted drywell head can be lifted by the overhead reactor enclosure crane attached by a hook box and transferred to its designated storage space on the refueling floor. The drywell seal surface protector can then be installed, if required.

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 watertight. Vessel head piping and head insulation must be removed and transported to storage on the refueling floor.

The water level in the vessel is now brought to flange level in preparation for head removal.

c. Reactor vessel opening

1. Vessel head removal The stud tensioners, as part of the strongback/ carousel assembly, are transported by the reactor enclosure crane and positioned on the reactor vessel head. Each stud is tensioned and its nut is loosened sequentially.

When the nuts are loose, they are backed off using a nut runner until only a few threads engage. The nut is then rotated free from the stud. The nuts and washers are placed in the racks provided for them and can be transported to the refueling floor for storage. Alternately, the nuts and washers may be placed in the carousel storage compartments and stored with the carousel. With the nuts and washers removed, the vessel stud protectors and vessel head guide caps are installed.

The head strongback/carousel, transported by the reactor enclosure crane, is attached to the vessel head, and the head is transported to the head holding pedestals on the refueling floor. The head holding pedestals keep the vessel head elevated to facilitate inspection and O-ring replacement.

The studs in line with the fuel transfer canal are removed from the vessel flange and can be placed in the rack provided. The loaded rack can be transported to the refueling floor for storage. Alternatively, the studs can be transported with the vessel head. The studs opposite the cattle chute are also removed to maintain balance when the studs are transported with the vessel head.

2. Dryer removal CHAPTER 09 9.1-44 REV. 19, SEPTEMBER 2018

LGS UFSAR The dryer/separator sling is lowered by the reactor enclosure crane and attached to the dryer lifting lugs. The dryer is lifted from the reactor vessel and transported to its storage location in the dryer/separator storage pool adjacent to the reactor well. Since the dryer is anticipated to be highly contaminated, the reactor well and storage pool is flooded and a wet transfer effected. An in-air transfer may be used if the dryer is kept wet to reduce the potential for airborne contamination.

3. Separator removal From the auxiliary platform, or refueling platform work area, the four main steam lines are plugged from inside the vessel using the furnished plugs for this duty. The servicing of the main steam line isolation or safety relief valves can thus be accomplished without adding to the critical refueling path time. The separator is unbolted using shroud head bolt wrenches. This is accomplished by working from the auxiliary platform, or refueling platform.

The dryer/separator sling is lowered into the vessel and attached to the separator lifting lugs. If not already done, the water in the reactor well and in the dryer/separator storage pool is equalized with the fuel pool water level.

This may be accomplished by raising the skimmer surge tank water levels, which communicates with both the spent fuel pool and reactor cavity. The separator is transferred underwater to its allotted storage place in the adjacent pool.

4. Fuel bundle sampling During reactor operation, the core offgas radiation level is monitored. If a rise in offgas activity has been noted, fuel assemblies can be sampled during shutdown to locate any leaking fuel. Fuel sipping or sampling may be performed in the reactor vessel or the fuel storage pool in accordance with appropriate procedures. If a defective bundle is found, it can be stored in a special defective fuel storage container to prevent the spread of contamination in the fuel storage pool.

d. Refueling and reactor servicing The remaining gate(s) isolating the fuel pool from the reactor well can be removed after the water level in the reactor well and dryer/separator storage pool is raised to the fuel pool water level, thereby interconnecting the fuel pool, the reactor well, and the dryer/separator storage pool. If required, the fuel pools can be cross-connected, at any time, through the cask storage pit to support refueling and reactor service.

The actual refueling of the reactor can now begin.

1. Refueling During a normal equilibrium outage, approximately one-third of the fuel is removed from the reactor vessel, two-thirds of the fuel is shuffled in the core (generally from peripheral to center locations), and one-third new fuel is installed. A full core off-load and reload may also be performed. The actual CHAPTER 09 9.1-45 REV. 19, SEPTEMBER 2018

LGS UFSAR 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 cavity. In addition to the fuel grapple, the refueling platform is equipped with two auxiliary hoists that can be used with various grapples to service other reactor internals.

To move fuel, the fuel grapple is aligned over the fuel assembly, lowered, and grappled to the fuel assembly bail. The fuel assembly is manually raised out of the core, then may be automatically moved through the fuel transfer canal to the fuel pool, positioned over the storage rack, and lowered to its new storage location. Fuel is shuffled, and new fuel is manually grappled then moved (automatically or manually) from the storage pool to the reactor vessel in the same manner. The fuel movements outside of fuel entry zones may be accomplished in either the manual or automatic modes of operation for the refueling platform. Fuel movements inside the fuel entry zone may only be completed in the manual mode of operation.

A portable refueling shield is provided to reduce radiation dose rates in the drywell that are due to the transfer of spent fuel assemblies from the reactor vessel to the spent fuel pool. During refueling, the lead and steel shield is located in the reactor well, between the reactor vessel and the spent fuel pool, which permits continuous personnel occupancy of the drywell.

e. Vessel closure The following steps, when performed, return the reactor to operating condition. In general, the procedures are the reverse of those described in the preceding sections. Many steps may be performed out of sequential order as conditions permit.

1. Install fuel pool gates and install cask pit gates as required. Restore operating unit skimmer surge tank as required.

2. Core verification: the core position of each fuel assembly must be verified to ensure that the desired core configuration has been attained.

3. CRD tests: the CRD timing, friction, and scram tests are performed.

4. Remove main steam line plugs.

5. Replace separator and latch.

6. Replace steam dryer.

7. Drain dryer/separator storage pool and reactor well and restore the outage unit skimmer surge tanks to normal operating levels as required.

8. Decontaminate reactor well.

9. Remove and store portable refueling shield.

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10. Deleted.

11. Remove drywell seal surface covering, if used.

12. Open drywell vents.

13. Decontaminate dryer/separator storage pool.

14. Replace vessel studs.

15. Replace slot plugs.

16. Install reactor vessel head.

17. Install vessel head piping and insulation.

18. Replace dryer/separator canal plugs.

19. Hydro test vessel, if necessary.

20. Install drywell head.

21. Install reactor well shield plugs. Installation of the lower layer of the reactor cavity shield plugs must be completed prior to the reactor exceeding 1%

reactor power. The upper layer of shield plugs and associated sealant material around the upper plugs must be installed prior to exceeding 25%

reactor power.

22. 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 incore flux monitors are calibrated.

9.1.4.2.10.3 Departure of Fuel from Site The spent fuel shipping cask arrives by rail car or truck in the refueling hatch at grade level. It is lifted from there by the 125 ton hook of a reactor enclosure crane to the refueling floor and placed in the shipping cask washdown area between the fuel pools of Units 1 and 2.

The cask outside is decontaminated of road dirt, the lid is removed by the reactor enclosure crane, and the cask is placed in the shipping cask pit. The 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.

Following underwater replacement of the cask lid, the cask is lifted for decontamination. The reactor enclosure crane then transfers the cask from the storage pit onto the shipping vehicle. At no time will the cask be lifted or carried over spent fuel.

9.1.4.3 Safety Evaluation Safety aspects (evaluation) of the fuel servicing equipment are discussed in Section 9.1.4.2.3, and safety aspects of the refueling equipment are discussed throughout Section 9.1.4.2.7. A description CHAPTER 09 9.1-47 REV. 19, SEPTEMBER 2018

LGS UFSAR of fuel transfer, including appropriate safety features, is provided in Section 9.1.4.2.10. In addition, the following summary safety evaluation of the fuel handling system is provided below.

The fuel prep machine removes and installs channels with all parts remaining underwater.

Mechanical stops prevent the carriage from lifting the 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 that prevent the loading of fuel bundles into this rack.

During the course of the controlled fuel inspection irradiated fuel components, no fuel assemblies or bundles, may be raised such that there is less than 7 feet of water shielding above individual fuel pins (or rods).

While an entire irradiated fuel bundle will not be raised above the minimum 7 feet shielding level, fuel inspection activities may require individual fuel rods to be raised to a height maintaining no less than 6 feet water coverage.

A pair of rails which guide an instrument holder may be installed between the fuel preparation machine handrail and the lower fuel inspection fixture to permit remote inspection of fuel bundles.

There are no nuclear safety problems associated with the handling of a single new fuel bundle or a pair of new fuel bundles. Equipment and procedures prevent an accumulation of more than two bundles in any location.

The refueling platform is designed to prevent it from toppling into the fuel pool or core during an SSE. Redundant interlocks are provided to prevent the following:

a. Travel of the refueling platforms over the reactor core while in the startup mode,

b. Lifting a fuel bundle over the reactor core with one or more control rods withdrawn from core cells containing fuel assemblies, and in the refuel mode, and

c. Withdrawing one or more control rods with the refueling platform carrying a fuel bundle over the reactor core while in the refuel mode, unless all four fuel assemblies have been removed from the core cell(s) with rod(s) withdrawn.

A boundary zone limit system is normally available to assist the refueling platform operator, who is responsible to prevent collision of the fuel grapple with the spent fuel pool or reactor cavity walls during movements of spent fuel. In its normal up position, the grapple is minimal 6 feet 6 inches below the water surface; therefore, a fuel assembly will not be raised to a point where it will not be adequately shielded by water during normal refueling conditions. The grapple is hoisted by redundant cables inside the mast and lowered by gravity. A digital readout is displayed to the operator, showing the exact coordinates of the grapple over the core.

The mast is suspended and gimballed from the trolley, near its top, so that the mast can be swung about the axis of platform travel, to remove the grapple from the water for servicing and storage.

The grapple has two independent hooks, operated by an air cylinder. Engagement is indicated to the operator.

CHAPTER 09 9.1-48 REV. 19, SEPTEMBER 2018

LGS UFSAR 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 two auxiliary hoists have load cells with interlocks that prevent the hoists from moving anything classified as a heavy load.

The two auxiliary hoists have electrical interlocks that prevent the lifting of their loads higher than 6 feet 6 inches underwater at the normal up position. Adjustable mechanical jam stops and redundant electrical interlocks on the cables back up these interlocks.

In summary, the fuel handling system complies with GDC 2, GDC 3, GDC 4, GDC 5, GDC 61, GDC 62, and GDC 63, and other applicable portions of 10CFR50.

A system level, qualitative-type FMEA relative to this system is discussed in Section 15.9.

The safety evaluation of the new and spent fuel storage is presented in Sections 9.1.1.3 and 9.1.2.3.

The safety considerations of the Control Rod Blade and Defective Fuel Storage Rack are presented in Section 9.1.4.2.

Regulatory Guide 1.13 (Rev 1) is applicable to the refueling platform within GE's scope of supply for this plant. The refueling platform is designed to prevent it from toppling into the fuel pool or core during an SSE.

9.1.4.4 Inspection and Testing Requirements 9.1.4.4.1 Inspection Refueling and servicing equipment is subject to the strict controls of quality assurance, incorporating the requirements of 10CFR50, Appendix B. Components defined as essential to safety, such as the fuel storage racks and fuel prep machine, have an additional set of engineering specified "quality requirements" that identify safety-related features that require specific QA verification of compliance to drawing requirements.

Before shipment, every safety-related component inspection item is reviewed by QA supervisory personnel and combined into a summary product quality checklist. By issuance of the product quality checklist, verification is made that all quality requirements have been confirmed and are on record in the product's historical file.

9.1.4.4.2 Testing Qualification testing is performed on refueling and servicing equipment before multiunit production.

Test specifications are defined by the responsible design engineer and may include the sequence of operations, load capacity, and life cycles tests. These test activities are performed by an independent test engineering group, and in many cases a full design review of the product is conducted before and after the qualification testing cycle. Any design changes affecting function that are made after the completion of qualification testing are requalified by test, calculation, or design review.

CHAPTER 09 9.1-49 REV. 19, SEPTEMBER 2018

LGS UFSAR Functional tests are performed in the shop before the shipment of production units and generally include electrical tests, leak tests, and sequence of operations tests.

When the unit is received at the site, it is inspected to ensure that no damage has occurred during transit or storage. Prior to use and at periodic intervals, each piece of equipment is again tested to ensure that the electrical and/or mechanical functions are operational.

Passive units, such as the fuel storage racks, are visually inspected before use.

9.1.4.5 Instrumentation Requirements Most of the refueling and servicing equipment is manually operated and controlled by the operator's visual observations. This type of operation does not necessitate a dynamic instrumentation system.

However, there are several components that do have instrumentation and control systems.

9.1.4.5.1 Refueling Platform The refueling platform has a nonsafety-related X-Y-Z position indicator system that informs the operator which core fuel cell the fuel grapple is accessing. The refueling platform controls in conjunction with the X-Y-Z position indicator system can automatically position the fuel grapple over any valid fuel cell location. An interlock status display is provided to the refueling platform operator. See Section 7.7 for a discussion of refueling interlocks.

Additionally, there is a series of mechanically and electronically activated switches and relays that provides monitor indications on the operator's console for grapple limits, hoist and cable load conditions, and confirmation that the grapple's hook is engaged.

A series of load cells is installed to provide automatic shutdown whenever threshold limits are exceeded on either the fuel grapple or the auxiliary hoist units.

9.1.4.5.2 Fuel Support Piece Grapple Although the fuel support piece grapple is not essential to safety, it has an instrumentation system consisting of mechanical switches and indicator lights. This system can provide the operator with a positive indication that the grapple is properly aligned and oriented and that the grappling mechanism is either extended or retracted. The operator may elect to perform manual manipulations to determine grapple engagement.

9.1.4.5.3 Other Equipment See Table 9.1-6 for additional refueling and servicing equipment not requiring instrumentation.

9.1.4.5.4 Radiation Monitoring The radiation monitoring equipment for the refueling and servicing equipment is evaluated in Section 7.6.

9.1.5 REACTOR ENCLOSURE CRANE CHAPTER 09 9.1-50 REV. 19, SEPTEMBER 2018

LGS UFSAR The reactor enclosure crane is a bridge crane mounted on runway rails that are supported by the secondary containment superstructure. The crane serves both units and is designed to be used normally during maintenance and refueling operations.

9.1.5.1 Design Bases

a. The reactor enclosure crane is designed to move loads from one location to another in the refueling area and also between plant grade level and the refueling area level.

b. The reactor enclosure crane is designed to handle loads with a maximum weight of 125 tons while maintaining a minimum safety factor of 5.

c. The main hoisting system of the reactor enclosure crane is designed so that the failure of any single component does not result in a sudden displacement or dropping of the load.

d. The reactor enclosure crane is designed to prevent movement of the crane over the new fuel and spent fuel storage areas in the absence of specific action by the crane operator to allow such movement.

e. The reactor enclosure crane is designed to maintain its structural integrity in the event of an SSE.

9.1.5.2 Equipment Design The reactor enclosure crane is a bridge crane mounted on runway rails that are supported by the secondary containment superstructure. The bridge consists of two welded box girders held together with structural end beams. These two end beams are supported by wheeled trucks that travel on top of the runway rails. Two trucks, each consisting of two wheels, are located at each end of the bridge.

The structural frame support for the reactor enclosure crane hoisting machinery is the trolley, which moves by tractive power on trucks over rails secured to the top of the two crane girders. Two hoists are provided, a main hoist with a design capacity of 125 tons and an auxiliary hoist with a design capacity of 15 tons (design rated load). The maximum critical load (MCL) for the main hoist is 125 tons and for the auxiliary hoist is 6.75 tons. The electric-powered hoists raise and lower loads by wire rope reeving through upper and lower sheaves, the lower sheaves being an integral part of the load block. Each hoist is equipped with a hook attached to its respective load block.

Design parameters for the reactor enclosure crane are listed in Table 9.1-10.

The fuel handling function of the reactor enclosure crane is to transport new fuel assemblies to the fuel preparation machines and to handle the spent fuel shipping casks. The crane was purchased prior to the issuance of ANS 57.1, and thus the requirements of this standard were not explicitly followed in the design and manufacture of the crane. The crane has been reviewed against the applicable requirements of ANS 57.1; these requirements and the results of the review are summarized in Table 9.1-22. ANS 57.1 guidelines not listed in Table 9.1-22 are either not CHAPTER 09 9.1-51 REV. 19, SEPTEMBER 2018

LGS UFSAR applicable to the reactor enclosure crane during fuel handling or are recommendations rather than requirements.

All of the applicable guidelines of ANS 57.2 (1976) are met for the cask handling function of the reactor enclosure crane, except that the crane bridge is not prevented from passing over the spent fuel pool. The hoists, however, are prevented from lifting loads heavier than a fuel assembly over stored spent fuel by interlocks and an administrative control system. Restrictions on heavy loads carried near the spent fuel pool are discussed in Reference 9.1-1.

The design of the reactor enclosure crane includes the following features:

a. Structural components All the structural components and machinery of the reactor enclosure crane are designed for a full capacity of 125 tons on the main hoist or 15 tons on the auxiliary hoist (design rated load). The maximum critical load (MCL) for the main hoist is 125 tons and for the auxiliary hoist is 6.75 tons, with a minimum safety factor of 5 against ultimate failure for the load-carrying parts and the machinery. The structural components, except for the bridge girders, have a design safety factor of 2.5 against yield. For the hoisting mechanism, all load-carrying parts except structural members are designed with a minimum safety factor of 5 against ultimate failure.

The calculated stresses of all load-carrying parts are in accordance with the requirements of Crane Manufacturer's Association of America Specification 70.

The structural members of the reactor enclosure crane are designed for a fatigue loading of 20,000-100,000 cycles, with each completed lift representing one cycle.

The rotating machinery is designed for a fatigue life expectancy of 2,000,000 cycles, with each rotating component cycle represented by one revolution. Any load below 50% of the crane rated capacity does not reduce the life expectancy of the crane.

b. Mechanical components The crane is of a single-trolley, indoor, electric overhead, bridge crane design. The trolley layout is shown in Figure 9.1-18.

The main hoist consists of two balanced, eight-part reeving systems to provide redundancy. The arrangement consists of two separate and redundant wire cables reeved side-by-side through the upper and lower sheaves, as shown in Figure 9.1-19. Each cable passes through a paired equalizer unit that adjusts for unequal cable length and is used as a load transfer safety system. This energy-absorbing device eliminates sudden load- displacement and shock to the crane system in the unlikely event of a cable break. The factor of safety (static) is halved when a cable breaks, but no swinging action occurs because each cable is reeved to both sides of the upper and lower sheaves.

The auxiliary hoist is provided with a two-part reeving system as shown in Figure 9.1-20.

CHAPTER 09 9.1-52 REV. 19, SEPTEMBER 2018

LGS UFSAR The main functions of the main hoist equalizer system are to continually adjust the hook load so that any load under normal operation is shared equally by the redundant reeving systems and to transfer the shock of a cable break in an acceptably safe dynamic fashion to the remaining cable. The equalizer assembly is shown in Figure 9.1-21. The main hoist uses a redundant equalizer shaft that consists of a solid rod within a hollow tube; either shaft can support the full load if there is a failure of the other. If there is an exaggerated displacement of the equalizer assembly caused by a cable break, either of two proximity limit switches would be activated. The equalizer system of the main hoist uses vane-type limit switches that stop the hoisting motion if the hoisting rope length needs adjustment.

The hoisting motion also stops if one set of reeving fails, so that the broken cable can be removed before it becomes entangled with the other reeving system. This equipment protection mechanism stops the hoisting motion by cutting power to the hoist motor and setting the hoist brakes. Before making a series of lifts the equalizer bar can be visually inspected, and adjusted if necessary, so that an unnecessary power shutoff does not occur. If the equalizer bar needs to be adjusted during a lift, the load can be lowered and the adjustment made at the cable drum anchors. If the equalizer bar reaches the limits of its travel, which should occur only if one of the cables had already failed, the load can be safely lowered with the remaining cable so that a new cable can be installed.

The main hook is a two-pronged sister hook with safety latches and a cored bail hole. Redundancy is provided in the main hook by incorporating a coaxial "hook within a hook" design. The safety latches are required when handling loads using flexible rigging (such as slings) in order to prevent rigging from coming off the main hook when the rigging is slack. The shaft of the outer hook is bored out to accommodate the inner hook shaft as shown in Figure 9.1-22. Each hook is independently supported by its respective crosshead and antifriction bearings that are in turn supported by the load block, as shown in Figure 9.1-23. The bail hole and each set of prongs have a design rated capacity of 125 tons, with a conservative safety factor to ultimate strength greater than 5.

The main hoist mechanism is equipped with one load brake and three redundant holding brakes. The load brake is a dc-actuated, eddy current, control-type brake and is used to regulate load lowering and raising speed. The holding brakes are dc magnet brakes that employ rectifiers to permit the use of ac power supply. Two of the holding brakes are applied immediately when power is interrupted to the main and creep motors. The third holding break is applied after a short time-delay. All three holding breaks are released when power is supplied to either the main or creep motor with no time delays. The torque rating of each of the four hoist brakes is at least 150% of rated full load hoist motor torque.

c. Crane controls An operator's cab is provided at the south end of the reactor enclosure crane. The crane can be controlled from the cab or from a radio controller.

Movements of the bridge, trolley, main hoist, and auxiliary hoist can be controlled from either the cab or the radio controller. Both the radio controller and cab controls include a main power control switch that will interrupt all power to the crane except CHAPTER 09 9.1-53 REV. 19, SEPTEMBER 2018

LGS UFSAR for the utility lights and heaters. Motion control push buttons in the cab are of the momentary contact-type that return to the OFF position when released.

The radio controller uses levers for bridge, trolley, main hoist and auxiliary hoist motion. They are spring return to off position when released.

d. Hoist limit switches The extent of travel for both the main hoist and the auxiliary hoist is limited for both the raising and lowering directions by a combination of limit switches. Redundant limit switches are provided for both hoists in the raising direction and for only the main hoist in the lowering direction. The primary protection for both hoists for both directions are geared limit switches coupled to the hoist drum shafts and interrupt power to the hoist motors via the control circuitry. The secondary protection for the auxiliary hoist in the raising direction is a weighted limit switch which also interrupts power to the auxiliary hoist motor via the control circuitry. The secondary protection for the main hoist in the raising direction for both the main and creep drive motors are weighted limit switches which interrupt power to each drive motor directly. The secondary protection for the main hoist in the lowering direction is a geared limit switch coupled to the gear case output shaft which interrupts all power to the crane except for the utility lights and heaters. When the power to any of the three hoist motors is interrupted, the breaks for that hoist will automatically set. The primary and upper redundant limit switches are wired so that the motor can be energized in the reverse direction after a limit switch has been tripped.

Both the main hoist and the auxiliary hoist are equipped with a centrifugal-type limit switch, located on the drum shaft, to provide automatic shutdown protection (hoist motor trip and setting of the holding brakes) against any control or motor malfunction that might result in a runaway condition of the load. The trip setting is at 120% or less of hoist motor synchronous speed.

A load-sensing system is provided for the main hoist, using a load cell mounted under the equalizer shaft. Digital readout of the load is provided in the operator's cab. Overload protection is provided by automatic shutdown (hoist motor trip and setting of the holding brakes), when the overload setpoint (100% to 115% of rated load) is exceeded.

The response time to stop the main hoist motion if there is actuation of a hoist overspeed, overload, or overtravel limit switch is such that the load block movement is less than 3 inches following actuation of one of the hoist limit switches.

e. Bridge and trolley drives and controls The bridge and trolley have static stepless control on travel speed from minimum to maximum speed. Both the bridge and trolley have braking systems that must be energized to release and that automatically set if there is a power loss. The bridge is equipped with two electric brakes and two hydraulic brakes, each of which has a torque rating equal to 125% of maximum torque of one bridge drive motor. The trolley is equipped with two magnetic brakes, each of which has a torque rating equal to 50% of the maximum torque of the trolley drive motor.

CHAPTER 09 9.1-54 REV. 19, SEPTEMBER 2018

LGS UFSAR The programmable limit system will provide protection to both the bridge and trolley near the ends of their respective rails to stop bridge or trolley movement before bumper contact occurs. In addition, the crane is equipped with zone travel programmable limit switches to prevent passage of the load over or near the spent fuel pool or the new fuel storage vault. These protected areas are shown in Figure 9.1-24. The tripping of the travel limit switches or motor overloads cuts off power to the bridge and trolley drive motors and sets the brakes. A key-locked bypass switch is provided in the crane controls so that the operator may consciously override the zone travel limitations.

f. Thermal overload protection Thermal overload protection is provided for motors on the crane to prevent continuation of motor-stalling torque. In addition, a thermal overload warning indication for the main hoist motor is provided in the operator's cab.

9.1.5.3 Loads Refueling operations and spent fuel shipping cask handling are discussed in Section 9.1.4. The following is a listing of specific items, and their approximate weights, that are handled by the reactor enclosure crane during and after each reactor refueling.

Drywell head 104 tons Reactor vessel head 88 tons Steam separator assembly 82 tons Steam dryer 39 tons Reactor well shield plugs No. 1 & 2 61 tons each No. 3 & 4 69 tons each No. 5 & 6 85 tons each No. 7 & 8 82 tons each No. 9 & 10 85 tons each No. 11 & 12 12 tons each Reactor refueling shield 22 tons Dryer/separator canal plugs 2 of No. 3 59 tons each No. 14 59 tons Refueling slot plugs No. 15 38 tons No. 16 38 tons Spent fuel shipping cask 100 tons Reactor Cavity Work Platform 38 tons CHAPTER 09 9.1-55 REV. 19, SEPTEMBER 2018

LGS UFSAR Independent Spent Fuel Storage Installation Transfer Cask 107.6* tons

• (115 tons assumed for conservatism) 9.1.5.4 Safety Evaluation Complete redundancy is provided for the main hoist load-bearing and load-holding equipment of the reactor enclosure crane, including sheaves, ropes, equalizer assembly, reducing gears, and holding brakes, so that no single component failure results in an uncontrolled lowering or dropping of a load within the rated capacity of the crane. The equalizer assembly is an energy-absorbing device that eliminates sudden load-displacement and shock to the crane system in the unlikely event of a rope break. This single failure proof design conforms to NUREG-0554 and NUREG-0612, as further discussed in Reference 9.1-1.

Load-bearing members and main hoist equipment of the reactor enclosure crane are designed in accordance with seismic Category I criteria so that the crane can structurally withstand the SSE and maintain the fully rated load in a static position during and following an SSE. Antiderail devices are installed to preclude derailment of the bridge or trolley under seismic loading.

The crane bridge and trolley are equipped with zone travel programmable limit switches to prevent passage of the load over or near the spent fuel pool and the new fuel storage vault. The travel limitations can be bypassed only by conscious operator action. Figure 9.1-24 illustrates the area on the refueling floor over which crane movement is prevented without such deliberate bypass.

The spent fuel shipping cask is described in Section 9.1.4. A spent fuel shipping cask drop accident analysis is discussed in Section 15.7.5.

The cask storage pit is separated from the spent fuel pool to minimize the proximity of the spent fuel shipping cask to the fuel storage racks during cask handling operations. The layout of the refueling area is such that the cask is moved in a single straight line between the cask storage pit and the refueling hatch. Both the cask storage pit and the refueling hatch are located on the plant centerline, which allows cask handling to be performed without moving the bridge.

A FMEA for the reactor enclosure crane is presented in Table 9.1-11. Although it has been withdrawn, for purposes of information presentation, conformance with Regulatory Guide 1.104 is discussed in Table 9.1-12.

9.1.5.5 Inspection and Testing Magnetic particle inspection of reactor enclosure crane components during fabrication is performed in accordance with ASTM E 109 and ASTM E 138 for dry and wet processes, respectively.

Ultrasonic inspection is performed in accordance with ASTM A 388. NDT procedures for testing plate, shafts, and for groove welds and radiographic examination of groove welds are based on applicable ASTM and AWS D1.1 specifications.

9.1.5.5.1 Structural Members The following welds are 100% examined by the magnetic particle method:

a. Girder top and bottom cover plate splice butt welds CHAPTER 09 9.1-56 REV. 19, SEPTEMBER 2018

LGS UFSAR

b. Girder web plate splice butt welds

c. Girder top and bottom cover plate to web plate fillet welds

d. Girder notch shelf plate to web plate fillet welds

e. End truck top and bottom cover plate to web plate fillet welds

f. Trolley side, top and bottom cover plate to web plate fillet welds

g. Trolley side to girt plate fillet welds

h. Load block frame fillet welds as accessible

i. End tie top plate to web plate fillet welds 9.1.5.5.2 Load-Bearing Components All gears, pinions, swivels, hoist shafts, and hook trunnions in both the main and auxiliary hoist assemblies are 100% nondestructively examined by the magnetic particle method, in addition to 100% ultrasonic examination of swivels.

9.1.5.5.3 Hooks Following forging, both the main and auxiliary hooks are examined by both the ultrasonic and magnetic particle methods. Each hook is then given a load test by the crane manufacturer according to the following procedure: A dimensional check of the hook is performed and the dimensions recorded. The hook is then proof-tested at 200% of its rated capacity. After the test, the hook is given a magnetic particle examination, and its dimensions are rechecked and recorded.

The hook is accepted if no cracks, permanent deformation, or other defects are produced by the testing.

9.1.5.5.4 Ropes Samples are taken from each wire rope to be used in the crane and are subjected to destructive breaking strength tests to verify that the rope exceeds the manufacturer's published value for breaking strength. At least two such tests are performed on each rope.

9.1.5.5.5 Performance and Acceptance Tests After erection of the crane in the reactor enclosure, extensive performance and acceptance testing is carried out, including:

a. Detailed checking of all mechanical and electrical components of the crane to verify proper assembly and operation

b. Running-in tests at no-load. This test includes all speeds and motions for which the crane is designed plus verification of the proper operation of all limit switches.

CHAPTER 09 9.1-57 REV. 19, SEPTEMBER 2018

LGS UFSAR

c. Load testing of the main hoist at 125% of its rated capacity and of the auxiliary hoist at 120% or greater of its rated capacity. These tests include the full range of movement for hoist raising and lowering and bridge and trolley travel.

d. Performance testing at 100% of the rated capacity. This test includes all speeds and motions for which the crane is designed, plus verification of the proper operation of all limit switches. However, performance testing at less than 100% of rated capacity will be conducted on the geared upper limit switch, the programmable limit switches that limit movement over the spent fuel pool, and prevent movement to the extreme ends of the bridge and trolley rails. A performance test at 100% of the rated capacity will be conducted on the weighted upper limit switch. However, the load will be raised from a position just off the floor at maximum speed, and the upper limit switch will be manually tripped (Table 9.1-12, footnote 18).

9.

1.6 REFERENCES

9.1-1 Limerick Generating Station, Limerick Units 1 and 2, Overhead Handling Systems Review Final Report, Bechtel Revision 6 (PECO Revision 1), August 1989.

9.1-2 R.J. Fritz, "The Effects of Liquids on the Dynamic Motions of Emmersed Solids",

Journal of Engineering for Industry, (February 1972).

9.1-3 "Simulated Rack Minimum Coefficient of Friction" by PaR.

9.1-4 Professor Ernest Rabinowicz, "Friction Coefficients of Water-Lubricated Stainless Steels for a Spent Fuel Rack Facility", Massachusetts Institute of Technology, performed for Boston Edison Company.

9.1-5 "Current Trends in Fuel Performance," presented at the ANS Topical Meeting on Water Reactor Fuel Performance, St. Charles, Illinois (May 9-11, 1977).

9.1-6 "Licensing Report for Spent Fuel Storage Capacity Expansion, Limerick Generating Station", HOLTEC International.

9.1-7 General Electric Standard Application for Reactor Fuel (GESTAR-II), including the United States Supplement, NEDE-24011-P-A and NEDE-24011-P-A-US, (latest approved revision).

9.1-8 Limerick Generating Station Spent Fuel Storage K-infinity Conversion Analysis, GENE-155-93032, DRF A00-05449, June 1993.

9.1-9 Evaluation of Limerick (Holtec) High Density Storage Rack K-infinity Criterion, GENE-155-94-075, DRF J11-02394, October 14, 1994.

9.1-10 SVEA-96 Lead Fuel Assemblies for Limerick 2, Supplemental Lead Fuel Licensing Report, ABB Atom Report BR 91-042, January 1991.

9.1-11 GE14 Compliance with Amendment 22 of NEDE-24011-P-A (GESTAR II),

NEDC-32868P, (latest approved revision).

CHAPTER 09 9.1-58 REV. 19, SEPTEMBER 2018

LGS UFSAR 9.1-12 GE14 Fuel Design Cycle-Independent Analyses for Limerick Generating Station, Units 1 and 2, GE-NE-L12-00884-00-01P, March 2001.

9.1-13 GE14 Spent Fuel Storage Rack Criticality Analysis for the Limerick Generating Station, Unit 2 Global Nuclear Fuel Document No. J11-03898-01-SFP, March 2001.

9.1-14 GE14 Spent Fuel Storage Rack Criticality Analysis for the Limerick Generating Station, Unit 1 Global Nuclear Fuel Document No. J11-03932-00-SFP, May 2001.

9.1-15 "Limerick Report for Spent Fuel Storage Capacity Expansion," Holtec International Report No. HI-931012, Revision 2, approved October 1, 1993 (Technical Specifications Amendments 82 and 43 for Unit 1 and Unit 2, respectively).

9.1-16 "LGS Unit 1 and 2 Thermal Hydraulic Analysis for Reducing the In-Core Decay Time from 125 Hour to 40 Hours," approved January 9, 1995.

9.1-17 "Thermal Hydraulic Evaluation of Limerick Unit 1 Fuel Pool Cooling," Holtec International Report No. HI-2094497 approved February 3, 2010, Exelon Doc No.

LEAM-MUR-0067.

9.1-18 "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II),

Global Nuclear Fuels Document NEDC-33270P, (latest approved revision).

9.1-19 "Criticality Safety Evaluation for GNF2 Fuel in the SFP at Limerick, Holtec International Report HI-2104779, Revision 1.

CHAPTER 09 9.1-59 REV. 19, SEPTEMBER 2018

LGS UFSAR Table 9.1-1 FUEL POOL COOLING AND CLEANUP SYSTEM DESIGN PARAMETERS FUEL POOL COOLING PUMP Type Vertical, centrifugal Quantity 2 plus 1 backup Flow, each 600 gpm Total discharge head 200 ft Rated power, each 50 hp Design pressure 127 psig Design temperature 150F FUEL POOL F/D HOLDING PUMP Type Horizontal, centrifugal Quantity 1 per filter/ demineralizer unit Flow, each 27 gpm Total discharge head 60 ft Rated power, each 3 hp Design pressure 175 psig Design temperature 155F FUEL POOL SKIMMER SURGE TANK Type Vertical, cylindrical Quantity 2 Capacity, each 4690 gal FUEL POOL FILTER/DEMINERALIZER Type Vertical, cylindrical Quantity 1 per reactor unit plus one spare Flow, each 550 gpm Design pressure 175 psig Design temperature 235F FUEL POOL HEAT EXCHANGER

• Type Shell and straight tube Quantity 2 plus 1 backup

• Duty, each 4.0x106 Btu/hr Tube Side Shell Side Flow, each 1100 gpm 600 gpm Design pressure 125 psig 75 psig Design temperature 200F 200F

• For a normal refueling discharge heat load of 16.1 MBTU/hr, each Unit 1 heat exchanger is approxinmatley 50% capacity. For a normal refueling discharge heat load of 18.05 MBTU/hr, each Unit 2 heat exchanger is approximately 50% capacity. Administrative controls assure that FPCC is not relied on for fuel pool cooling, unless the heat load is equal to or less than 16.1 MBTU/hr on Unit 1 and 18.05 MBTU/hr on Unit 2.

CHAPTER 09 9.1-60 REV. 15, SEPTEMBER 2010

LGS UFSAR Table 9.1-2 Table Information Deleted CHAPTER 09 9.1-61 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 9.1-2A FUEL POOL COOLING SYSTEM HEAT REMOVAL CAPACITY AND MAKEUP REQUIREMENTS (ORIGINAL DESIGN)

PARAMETER VALUE AT NORMAL HEAT LOAD VALUE AT MAXIMUM HEAT LOAD (3)(1)

Quantity of fuel 33% core: 4.5 yr exposure 33% core: 4.5 yr exposure(2)(1) 10 day decay 10 day decay 33% core: 4.5 yr exposure(3) 33% core 4.5 yr exposure(2) 1.5 yr decay + 1 mo 10 day decay 33% core: 4.5 yr exposure(3) 33% core: 4.5 yr exposure(2) 3.0 yr decay + 2 mo 10 day decay 33% core: 4.5 yr exposure(3) 33% core: 4.5 yr exposure(3) 4.5 yr decay + 3 mo 1.5 yr decay + 1 mo 33% core: 4.5 yr exposure(3) 33% core: 4.5 yr exposure(3) 6.0 yr decay + 4 mo 3.0 yr decay + 2 mo 33% core: 4.5 yr exposure(3) 33% core: 4.5 yr exposure(3) 7.5 yr decay + 5 mo 4.5 yr decay + 3 mo 33% core: 4.5 yr exposure(3) 33% core: 4.5 yr exposure(3) 9.0 yr decay + 6 mo 6.0 yr decay + 4 mo 33% core: 4.5 yr exposure(3) 33% core: 4.5 yr exposure(3) 10.5 yr decay + 7 mo 7.5 yr decay + 5 mo 33% core: 4.5 yr exposure(3) 33% core: 4.5 yr exposure(3) 12.0 yr decay + 8 mo 9.0 yr decay + 6 mo 33% core: 4.5 yr exposure(3) 33% core: 4.5 yr exposure(3) 13.5 yr decay + 9 mo 10.5 yr decay + 7 mo 13% core: 4.5 yr exposure(4)(6) 22% core: 4.5 yr exposure(5)(6) 15.0 yr decay + 9 mo 12.0 yr decay + 8 mo Initial heat load 1.63x107 Btu/hr 3.64x107 Btu/hr Number of pumps and heat exchangers 2 RHR system required Water makeup requirements due to 61.4 gpm maximum -

boiling losses CHAPTER 09 9.1-62 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 9.1-2A (Cont'd)

(1)

Exposure of fuel based on 39,420 MWD/MT for 4.5 yr exposure.

(2)

This makes a full core discharge. The 4.5 yr exposure for the entire full core discharge is conservative.

(3) 33% core = 276 assemblies for conservatism.

(4) 13% core = 102 assemblies. This fills all fuel rack storage locations.

(5) 22% core = 166 assemblies. This fills all fuel rack storage locations.

(6)

Assumes maximum spent fuel storage rack inventory, 23 racks (i.e. the Control Rod Blade and Defective Fuel Storage Rack is not in the pool).

CHAPTER 09 9.1-63 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 9.1-2B FUEL POOL COOLING SYSTEM HEAT REMOVAL CAPACITY AND MAKEUP REQUIREMENTS - INCREASED STORAGE CAPACITY (Reference 9.1-15)

UNIT 1 (6)

PARAMETER VALUE AT NORMAL HEAT LOAD(1) (2) (5) VALUE AT MAXIMUM HEAT LOAD(1) (2) (5)

Quantity of fuel 36% core: 6 yr exposure 100% core: 6 yr exposure 6.7 day decay (3) 9.1 day decay (3) 36% core: 6 yr exposure 36% core 6 yr exposure 2 yr decay 2 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 4 yr decay 4 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 6 yr decay 6 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 8 yr decay 8 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 10 yr decay 10 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 12 yr decay 12 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 14 yr decay 14 yr decay 34% core: 6 yr exposure 34% core: 6 yr exposure 16 yr decay 16 yr decay 37% core: 6 yr exposure 37% core: 6 yr exposure 18 yr decay 18 yr decay 15% core: 6 yr exposure 15% core: 6 yr exposure 20 yr decay 20 yr decay 103% core: 6 yr exposure 103% core: 6 yr exposure 22 yr decay 22 yr decay CHAPTER 09 9.1-64 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 9.1-2B (Contd)

FUEL POOL COOLING SYSTEM HEAT REMOVAL CAPACITY AND MAKEUP REQUIREMENTS - INCREASED STORAGE CAPACITY (Reference 9.1-15)

UNIT 2 PARAMETER VALUE AT NORMAL HEAT LOAD(1) (2) (5) VALUE AT MAXIMUM HEAT LOAD(1) (2) (5)

Quantity of fuel 36% core: 6 yr exposure 100% core: 6 yr exposure 6.7 day decay (3) 9.1 day decay (3) 36% core: 6 yr exposure 36% core 6 yr exposure 2 yr decay 2 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 4 yr decay 4 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 6 yr decay 6 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 8 yr decay 8 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 10 yr decay 10 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 12 yr decay 12 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 14 yr decay 14 yr decay 36% core: 6 yr exposure 34% core: 6 yr exposure 16 yr decay 16 yr decay 37% core: 6 yr exposure 36% core: 6 yr exposure 18 yr decay 18 yr decay 15% core: 6 yr exposure 36% core: 6 yr exposure 20 yr decay 20 yr decay 36% core: 6 yr exposure 36% core: 6 yr exposure 22 yr decay 22 yr decay 37% core: 6 yr exposure 37% core: 6 yr exposure 24 yr decay 24 yr decay 29% core: 6 yr exposure 29% core: 6 yr exposure 26 yr decay 26 yr decay CHAPTER 09 9.1-65 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 9.1-2B (Contd)

FUEL POOL COOLING SYSTEM HEAT REMOVAL CAPACITY AND MAKEUP REQUIREMENTS - INCREASED STORAGE CAPACITY (Reference 9.1-15)

PARAMETER VALUE AT NORMAL HEAT LOAD (3) (6) (7) VALUE AT MAXIMUM HEAT LOAD (3) (4)

Initial Heat Load 1.805 x 107 Btu/hr (3) (6) (7) 3.76 x 107 Btu/hr (3) (4)

Number of Pumps and 2 RHR system Heat Exchangers Required Water Makeup Requirements 39.7 gpm (7) 81 gpm (4) due to boiling losses (1)

Percentage of core is based on 764 fuel assemblies.

(2)

Fuel assemblies in the spent fuel pool prior to analysis.

(3) 125 hour0.00145 days <br />0.0347 hours <br />2.066799e-4 weeks <br />4.75625e-5 months <br /> in-core decay time prior to movement of fuel assemblies.

(4)

For a reactor power of 3527 MW, an in-core decay time of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and a transfer rate of 10 fuel assemblies per hour (approximately 2.8 day decay), the initial heat load is 4.801 x 107 Btu/hr with a water makeup requirement of 102.02 gpm (Reference 9.1-17).

(5)

A one time per unit refueling cycle length extension to approximately 27 months (from a nominal 24 months) for Unit 1 Cycle 7 and Unit 2 Cycle 5 was evaluated. The increase in decay heat generation due to each core's total operating period of 6.25 years (vs. 6 years) has been determined to have no significant effect on maximum calculated fuel pool heat load.

(6)

Administrative controls assure that FPCC is not relied on for fuel pool cooling unless the heat load is equal to or less than 1.61 x 107 BTU/hr due to the smaller heat removal capacity of the backup Unit 1 FPCC heat exchanger.

(7)

For a reactor power of 3527 MW, an in-core decay time of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and a transfer rate of 10 fuel assemblies per hour (approximately 2.8 day decay), the initial heat load is 2.477 x 107 Btu/hr with a water makeup requirement of 51.29 gpm (Reference 9.1-17).

CHAPTER 09 9.1-66 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 9.1-3 FUEL POOL COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS PLANT OPERATING COMPONENT FAILURE EFFECT OF FAILURE FAILURE MODE EFFECT OF FAILURE MODE SYSTEM COMPONENT MODE ON THE SYSTEM DETECTION ON PLANT OPERATION Normal Fuel pool cooling Failure of one pump Decreased or loss Low pressure The spare pump is started water pumps with two pumps of system flow alarm on local by the operator to restore operating or one control panel design flow.

pump when one and common pump is operating trouble alarm in main control room Normal Service water Loss of service Decreased system Low pressure alarm The spare pump is pumps water pump flow in main control started by the operator room to restore design flow .

Normal Fuel pool service Loss of service Decreased system Pump Trip alarm The spare pump is water booster water booster pump flow in main control started by the operator pumps room to restore design flow .

Normal Makeup from demin Any failure or Loss of normal Visual inspection The pool can be water tank failures that makeup of during filling filled by temporary means or result in loss of demineralized operation from one of the seismic water supply water Category I makeup supplies.

Accident RHR pump Failure of pump Loss of RHR cooling Low pressure alarm The other RHR pump associated with full core supplementation in main control with the RHR heat exchanger unloaded into room that services the fuel pool is spent fuel pool started to restore cooling flow to the spent fuel pool.

Accident RHR heat exchanger Loss of cooling Loss of RHR cooling RHRSW pump discharge The other RHRSW pump in the water to heat supplementation low pressure alarm RHRSW loop is started to exchanger with full in main control to restore cooling water core unloaded into room flow to the RHR heat exchangers.

spent fuel pool CHAPTER 09 9.1-67 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-3 (Cont'd)

PLANT OPERATING COMPONENT FAILURE EFFECT OF FAILURE FAILURE MODE EFFECT OF FAILURE MODE SYSTEM COMPONENT MODE ON THE SYSTEM DETECTION ON PLANT OPERATION Accident RHR system Failure of a single Loss of RHR cooling Valve indication If the RHR system cannot be valves power-operated valve supplementation in main control realigned or the failed with full core room valve manually operated to unloaded into spent restore cooling, boiling fuel pool may occur. In this case sufficient water from seismic Category I makeup sources can be supplied to offset water losses due to boiling.

Accident Power supply Loss of offsite Loss of normal Alarm in main Two of the fuel pool cooling power power to system control room pumps can be powered from the diesel generators.

Additionally, the RECW system, in conjunction with the ESW system, can be used to provide cooling water to the fuel pool heat exchangers.

These systems are powered from the diesel generators.

Accident FPCC system Loss of system Loss of cooling spent fuel pool The spent fuel pool will start function due to to spent fuel low level alarm to boil. The seismic Category seismically pool in the main I makeup sources are used to induced damage control room supply sufficient water to offset losses due to boiling.

Accident FPCC System Loss of system Loss of cooling Alarms in the If the FPCC pump controls or function due to to spent fuel main control RECW intertie piping is not LOOP coincident pool room accessible post-LOCA, the with a design seismic Category I ESW basis LOCA makeup source is used to supply sufficient water to offset losses due to boiling.

CHAPTER 09 9.1-68 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-3 (Cont'd)

PLANT OPERATING COMPONENT FAILURE EFFECT OF FAILURE FAILURE MODE EFFECT OF FAILURE MODE SYSTEM COMPONENT MODE ON THE SYSTEM DETECTION ON PLANT OPERATION Accident Seismic Category I Failure (including Loss of seismic Low pressure and The other ESW or RHRSW makeup supply loss of power) of 1 Category I makeup low level alarm pump in the ESW or RHRSW loop ESW or RHRSW pump water alarm in main is started to provide control room sufficient makeup water.

These pumps are powered from separate electrical divisions.

Accident RHRSW Failure of a single Loss of seismic Low level alarm The ESW system can be used Seismic Category I power-operated valve Category I makeup in main control to supply makeup water. (There makeup supply water. room. Valve are no power-operated valves indication in the ESW supply path whose single failure can cause a loss of makeup water.)

CHAPTER 09 9.1-69 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-4 RESULTS OF SPENT FUEL POOL (SFP) BOILING ANALYSIS Unit 1: Time from loss of cooling to boiling - 17.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />(1)

Water heatup rate following loss of cooling - 4.11F/hr(1)

Unit 2: Time from loss of cooling to boiling - 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />(1)

Water heatup rate following loss of cooling - 5.33F/hr(1)

LPZ thyroid doses (rem) for the 30 day duration of the accident:

For a Spent Fuel Pool PF(2) of 0.1 UNIT 1 UNIT 2 BOTH UNITS Spike - 50 3.82x10-2 3.75x10-1 4.13x10-1 Spike - 20 1.86x10-2 1.88x10-1 2.07x10-1 Spike - 5 8.77x10-3 9.47x10-2 1.03x10-1 For a Spent Fuel Pool PF(2) of 0.01 Spike - 50 4.02x10-3 3.98x10-2 4.38x10-2 Spike - 20 1.96x10-3 2.0x10-2 2.2x10-2 Spike - 5 9.27x10-4 1.01x10-2 1.1x10-2 (1)

Assuming loss of spent fuel pool cooling as described in Section 9.1.3.6 (2)

PF = Iodine partition factor at the pool surface CHAPTER 09 9.1-70 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-5 DECAY HEAT AND EVAPORATION RATES FOR LOSS OF SPENT FUEL POOL COOLING TIME AFTER LOSS OF COOLING (hrs) 13.5 17.5 24 96 360 480 600 720 UNIT 1 Decay heat in 10.4 10.4 10.4 10.2 9.6 9.37 9.18 9.0 pool (MBtu/hr)

Evaporation rate 0.0 171.4 171.1 168.16 158.28 154.58 151.36 148.47 (ft3/hr)

UNIT 2 Decay heat in 13.4 13.3 13.3 12.9 11.4 10.9 10.5 10.2 pool (MBtu/hr)

Evaporation rate 220.75 219.85 218.96 212.82 187.49 180.35 173.29 168.07 (ft3/hr)

CHAPTER 09 9.1-71 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 9.1-6 TOOLS AND SERVICING EQUIPMENT FUEL SERVICING EQUIPMENT Fuel preparation machines Channel bolt wrenches Channel handling tool Fuel pool sipper Channel gauging fixture General purpose grapples Fuel inspection fixture Jib crane Upending platform/jib crane Roller conveyer Gantry crane New fuel inspection/channeling stand SERVICING AIDS Pool tool accessories Actuating poles General area underwater lights Local area underwater lights Drop lights Underwater TV monitoring system Underwater vacuum cleaner Viewing aids Light support brackets Incore detector cutter Incore manipulator REACTOR VESSEL SERVICING EQUIPMENT Reactor vessel servicing tools Steam line plugs Shroud head bolt wrenches Head holding pedestals Head stud rack Dryer/separator sling Head strongback Steam line plug/installation tool Vessel nut handling tool Head nut and washer storage racks Reactor enclosure crane main hook extension Service Pole Caddy Fuel Floor Auxiliary Platform Reactor Cavity Work Platform Assembly CHAPTER 09 9.1-72 REV. 15, SEPTEMBER 2010

LGS UFSAR Table 9.1-6 (Cont'd)

IN-VESSEL SERVICING EQUIPMENT Instrument strongback Control rod grapple Control rod guide tube grapple Fuel support piece grapple Grid guide Jet pump servicing tools Control rod latch tool Instrument handling tool Control rod guide tube seal Incore guide tube seals Blade guides Fuel bundle sampler Peripheral orifice grapple Orifice holder Combined grapple, CRB/FSP REFUELING EQUIPMENT Refueling equipment servicing tools Refueling platform STORAGE EQUIPMENT Spent fuel storage racks Channel storage racks Control rod blade storage racks Defective fuel storage container UNDER-REACTOR VESSEL SERVICING EQUIPMENT CRD servicing tools CRD hydraulic system tools NMS servicing tool CRD handling equipment Equipment handling platform Thermal sleeve installation tool Incore flange seal test plug Key bender CHAPTER 09 9.1-73 REV. 15, SEPTEMBER 2010

LGS UFSAR Table 9.1-7 FUEL SERVICING EQUIPMENT ESSENTIAL QUALITY SEISMIC COMPONENT CLASSIFICATION(1) GROUP(2) CATEGORY(3)

Fuel prep machine PE E I New fuel inspection/channeling Stand NE E -

Channel bolt wrench NE E -

Upending stand/jib crane NE E -

Channel handling tool NE E -

Roller conveyer NE E -

Fuel pool sipper NE E -

Gantry crane NE E -

Fuel inspection fixture NE E -

Channel gauging fixture NE E -

General purpose grapple PE E -

Jib crane PE E -

(1)

NE - nonessential PE - passive essential (2)

E - industrial code applies (3)

(-) - no seismic requirements CHAPTER 09 9.1-74 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-8 REACTOR VESSEL SERVICING EQUIPMENT ESSENTIAL QUALITY SEISMIC COMPONENT CLASSIFICATION(1) GROUP(2) CATEGORY(3)

Reactor vessel servicing tools NE E -

Steam line plug PE E -

Shroud head bolt wrench NE E -

Vessel nut handling tool NE E -

Head holding pedestal NE E -

Head nut and washer rack NE E -

Head stud rack NE E -

Dryer and separator sling PE E -(4)

Reactor enclosure crane main hook extension PE E -(4)

Head strongback/carousel E E -(4)

Steam line plug installation tool NE E -

Service Pole Caddy NE E IIA Fuel Floor Auxiliary Platform NE E -

Reactor Cavity Work Platform NE E IIA (1)

NE - nonessential PE - passive essential (2)

E - industrial code applies (3)

(-) - no seismic requirements

(+) - seismic IIA requirements (4)

Dynamic analysis methods for seismic loading are not applicable, because this equipment is supported by the reactor service crane. Lifting devices are designed with a minimum safety factor of 5 and undergo proof testing.

CHAPTER 09 9.1-75 REV. 15, SEPTEMBER 2010

LGS UFSAR Table 9.1-9 UNDER-REACTOR VESSEL SERVICING EQUIPMENT AND TOOLS SEISMIC EQUIPMENT/TOOL CLASSIFICATION CATEGORY CRD handling Nonessential -

equipment Equipment handling Nonessential -

platform Spring reel Nonessential -

Thermal sleeve Nonessential -

installation tool Incore flange Nonessential -

seal test plug Key bender Nonessential -

CRD servicing Nonessential -

tools CRD hydraulic Nonessential -

system tools NMS servicing Nonessential -

tool

(-) - no seismic requirements CHAPTER 09 9.1-76 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-10 REACTOR ENCLOSURE CRANE DESIGN PARAMETERS BRIDGE Runway length 323'-0" Bridge span 129'-2" Bridge weight 410,000 lb Number of wheels 8 Type of wheels Parallel tread Wheel size 27 in Number of drive motors 2 Drive motor power 7.5 hp @ 720 rpm Maximum travel speed 40 fpm Minimum travel speed 4 fpm Minimum incremental movement 1/4 in Number and type of brakes (2) magnetic shoe (2) hydraulic shoe Type of bumpers Spring Type of control 562 static stepless TROLLEY Length of trolley travel 129'-2" Trolley span 26'-6" Trolley weight 124,000 lb Number of wheels 4 Type of wheels Parallel tread Wheel size 24 in Drive motor power 7.5 hp @ 720 rpm Maximum speed 26.8 fpm Minimum speed 3 fpm Minimum incremental movement 1/8 in Number and type of brakes (1) magnetic shoe, (1) magnetic disc Type of bumpers Spring Type of control 562 ac static stepless CHAPTER 09 9.1-77 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-10 (Cont'd)

HOISTS Main Auxiliary Rated lifting capacity 125 tons 15 tons Drum size (pitch diameter) 64.25 in 30.25 in Upper sheave size (pitch diameter) 30 in None Lower sheave sizes (pitch diameter) 33.75 in, 30 in None Equalizer sheave size (pitch diameter) 17.75 in None Rope-type 6x37 6x37 preformed, preformed, SEIP, SS 304 IWRC IWRC Rope diameter 1.25 in 1.125 in Reeving-type 8 part 2 part Number of reeving systems 2 1 Hoist motor power 50 hp 25 hp Maximum hook speed 5 fpm 23 fpm Minimum hook speed 0.5 fpm 2.3 fpm Minimum incremental hook movement 1/32 in 1/8 in Maximum travel of hook 164'-6" 165'-5" Number and type of load brakes (1) eddy (1) eddy Current current Number and type of holding brakes (3) magnetic (2) magnetic Shoe shoe Type of control 563 static 563 static Stepless stepless CHAPTER 09 9.1-78 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-11 REACTOR ENCLOSURE CRANE FAILURE MODES AND EFFECTS ANALYSIS COMPONENT COMPONENT OR OR SUBSYSTEM EFFECT OF FAILURE FAILURE MODE SUBSYSTEM FAILURE MODE ON THE SYSTEM DETECTION REMARKS Power supply LOOP All crane movements Crane operator -

stopped by setting crane holding brakes and tripping all drive motors Main hoist hooks Failure of one hook None. The redundant Periodic inspection, -

hook supports the load. if not identified during crane use)

Main hoist wire Failure of one rope Spurious, dynamic, load Crane operator Two vane switches, Ropes 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 failure and cut off crane movements. The load power to the crane.

is supported by the remaining rope at a minimum static factor of safety of 5.

Main hoist drum Failure of drum Possible load stalling, Crane operator The load can be Shaft or noise and irregular positioned over its hoist operation. The storage or lay-down area crane operator and then lowered by stops hoist operation, manual operation of the resulting in setting of hoist holding brakes.

the holding brakes and the safe suspension of the load.

Main hoist holding Failure of one brake None. Two additional Periodic inspection Three holding brakes are Brakes in the open position holding brakes stop provided, all rated at the main hoist movement 150% of the hoist full and hold the load. load motor torque at the point of application.

CHAPTER 09 9.1-79 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-11 (Cont'd)

COMPONENT COMPONENT OR OR SUBSYSTEM EFFECT OF FAILURE FAILURE MODE SUBSYSTEM FAILURE MODE ON THE SYSTEM DETECTION REMARKS Trolley holding Failure of one brake None. When the trolley Crane operator Two holding brakes are Brakes in the open position drive motor is stopped, provided, both rated power to the holding at 50% of the trolley brakes is also cut off, drive motor torque at which sets the the point of application.

redundant brake.

Bridge holding Failure of one brake None. When the bridge Crane operator One holding brake and one Brake in the open position drive motors are stopped, foot-operated hydraulic power to the holding brake are provided for brakes is also cut off, each of the two bridge which sets the drive motors. The remaining brake. holding brake is rated at 100% of the bridge drive motor torque.

Main hoist drive Failure of one gear Two gear cases (drive gear cases case, resulting in and idler case) are gear disengagement: provided for the main hoist.

a) Drive gear case Spurious drum Crane operator revolving in the load lowering direction.

The overspeed switch activated by revolving drum sets the hoist holding brakes and stops the load.

b) Idler gear case None. The drive gear Crane operator case with the hoist motor and the holding brakes maintains control of the load.

CHAPTER 09 9.1-80 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-11 (Cont'd)

COMPONENT COMPONENT OR OR SUBSYSTEM EFFECT OF FAILURE FAILURE MODE SUBSYSTEM FAILURE MODE ON THE SYSTEM DETECTION REMARKS Main hoist Failure of switch: -

upward movement limit switch a) Open Immediate power cutoff Crane operator or (geared-type) to the main hoist. As a periodic testing result, the hoist is stopped through action of the hoist holding brakes.

b) Closed None. If the hoist Periodic testing continues its upward travel it is stopped by action of the redundant upper limit switch.

Main hoist Failure of switch: Tripped by physical upward movement contact with moving limit switch load block as it moves (contact-type) upward.

a) Open Immediate power cutoff 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. Hoist upward Periodic testing movement is limited by the redundant upper limit switch, before it reaches this limit switch.

CHAPTER 09 9.1-81 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-11 (Cont'd)

COMPONENT COMPONENT OR OR SUBSYSTEM EFFECT OF FAILURE FAILURE MODE SUBSYSTEM FAILURE MODE ON THE SYSTEM DETECTION REMARKS Main hoist Failure of one downward movement limit switch: -

limit switches (geared-type) a) Open Immediate power cutoff 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.Hoist movement downward is terminated by the redundant lower limit switch.

Bridge and Failure of one trolley movement switch associated programmable limit with a given bridge -

switches or trolley position:

a) Open Immediate cutoff 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.

CHAPTER 09 9.1-82 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-12 COMPARISON WITH NRC REGULATORY GUIDE 1.104 (FEBRUARY 1976) -

OVERHEAD CRANE HANDLING SYSTEMS FOR NUCLEAR POWER PLANTS NON-REGULATORY POSITION CONFORMANCE CONFORMANCE NOTES C.1 Performance Specification Design Criteria (1)

a. Separate performance X specification

b. Environmental operating conditions

i. Electrical motors X ii. Balance of crane (2)

1. Box girders X vent and draining (3)

2. NDTT X (16)

3. Cold proof test X (4)

4. Low alloy steel cold proof test

c. Seismic Category I X

d. NDE - lamellar tearing X (5)

e. Fatigue analysis X

f. Preheat - postheat X (6) welding C.2 Safety Features (7)

a. Controls, devices, X safe holding position

b. Auxiliary system, X dual component, immobile position CHAPTER 09 9.1-83 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-12 (Cont'd)

NON-REGULATORY POSITION CONFORMANCE CONFORMANCE NOTES (8)

c. Means for repairing X

d. Design for safety X while repairing C.3 Equipment Selection

a. Dual load attachment X points (9)

b. Lifting devices - X redundant design

c. Dual hoisting, 5 fpm X limitation

d. Head block and X load block

e. Dual reeving system - X rope standard (10)

f. Fleet angles X (11)

g. 200% static X design test

h. Sensing of overspeed, X overloading, etc.

(12)

i. Control system, X motors, torque (13)

j. Two-blocking X precautions, etc.

k. Drum protection X

l. Excessive breakdown X torque (14)

m. Hoisting brakes, X holding brakes

n. Dynamic and static X alignment

o. Increment drives X CHAPTER 09 9.1-84 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-12 (Cont'd)

NON-REGULATORY POSITION CONFORMANC CONFORMANCE NOTES

p. Trolley and bridge X motors

q. Cab located controls X

r. Safety devices, limit X devices

s. Operating manuals - MWL X

t. Change from construction X to operating

u. Installation instructions X C.4 Mechanical Check, Testing and Preventive Maintenance

a. Mechanical check X

b. 125% static test (2-block)

(19)

i. 125% static test X (18) ii. 100% full X performances test (15) iii. 2-block X

c. Preventive maintenance X (16)

d. Cold proof test X (17)

C.5 Quality Assurance X (1)

Position C.1.a. The load lifts during construction were not greater than those for plant operation, therefore no separate specifications were prepared.

(2)

Position C.1.b(1). Box girders are not of a closed design.

(3)

Position C.1.b(3). The crane manufacturer did not perform impact testing on any structural members.

(4)

Position C.1.b(4). Not applicable - no ASTM A514 is used.

CHAPTER 09 9.1-85 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-12 (Cont'd)

(5)

Position C.1.e. The structural members of the crane are designed for a fatigue loading of 20,000-100,000 cycles, which far exceeds the conservatively estimated 4000 cycles in the life of the crane. Additional fatigue analysis is therefore deemed unnecessary.

(6)

Position C.1.f. The preheat and postheat treatment of the welds is in accordance with AWS D1.1.

(7)

Position C.2.a. Inadvertent operation must also be a function of administrative control, not just of design. The present design is considered to comply with the regulatory position.

(8)

Position C.2.c. Provisions are made for manual operation of the main hoist holding brakes for lowering the load. No special provisions are made for manually moving the immobilized bridge or trolley; however, there are options available for moving the bridge or trolley if the electric power cannot be restored.

(9)

Position C.3.b. The loads that have been determined to be critical loads are the spent fuel shipping cask and the portable refueling shield. A lifting yoke which is single failure proof in accordance with NUREG-0612 is provided for use in lifting the shipping cask with the reactor enclosure crane. Although the lifting rig provided for use with the portable refueling shield is not redundant in design, it is designed to support a weight equivalent to three times the weight of the portable refueling shield. This high degree of conservatism protects against the possibility of failure of the refueling shield's lifting rig.

(10)

Position C.3.f. The fleet angle from drum to lead sheave is 31/4 degrees for LGS, which complies with the regulatory position. The regulatory position also recommends limiting the fleet angles between individual sheaves to 11/2 degrees, whereas this parameter is 31/2 degrees for LGS. This difference is justified by the following considerations:

a. The 31/2 degree limitation has been proven to be a reliable parameter for rope leads off of drums, which are more critical than rope leads from sheaves, the latter being more deeply grooved.

b. With redundant reeving, sheave spacings are double the normal spacings. Thus, to maintain a 11/2 degree fleet angle, the distance from the hook to the top of the crane would have to be needlessly and excessively increased to such a degree that it would be inconsistent with a good crane design.

The design ratio of running sheave pitch diameters to the rope diameter is 24:1 instead of the 30:1 or 26:1 recommended by Regulatory Guide 1.104. The 24:1 ratio is justified because, due to the large diameter of the wire rope used, 30:1 and 26:1 diameter ratios, sheave blanks are not readily available, and because the 24:1 ratio is recommended by the ASME Standard Committee on the Design of Overhead and Gantry Handling Systems for Critical Loads at Nuclear Power Plants in their comments on Regulatory Guide 1.104, dated March 18, 1976, and is consistent with the recommendations of CMAA Specification #70.

(11)

Position C.3.g. ANSI B30.2 allows for a 125% load test of these components, and the crane is tested to this value.

CHAPTER 09 9.1-86 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-12 (Cont'd)

(12)

Position C.3.i. The electric controls for the subject crane are set to limit the motor torque to 150% of rated motor torque.

(13)

Position C.3.j. The crane is provided with dual upper limit switches to preclude the possibility of a "two- block" occurrence and an overload switch combined with overcurrent and rate of current rise cutouts to automatically terminate hoisting if there is a load hang-up.

(14)

Position C.3.m. The crane design meets the requirements of this regulatory position, except that holding brake heat dissipation is accomplished by alternating the lowering and holding to provide time for cooling the brake mechanism. Also, administrative control is used to limit the lowering speed to less than 3.5 fpm if there is manual brake operation during emergency lowering.

(15)

Position C.4.b.iii. No testing of these conditions is performed, since the reeving system is not designed for two-blocking or load hang-up as indicated in footnote 13 above.

(16)

Position C.4.d. A cold proof test is performed with 125% of rated load below the minimum operating temperature for the crane during normal plant operation, which is 60oF.

However, in no case is the test load increased above 125% of the rated load. Crane testing in excess of 125% of the rated load may adversely affect the safety of the crane, since any such tests may propagate undetectable material defects and thus increase the probability of crane component failures. Furthermore, such testing violates the ANSI B30.2 and 29CFR1910.179(k).

A NDE following the cold proof test is performed on 10% of each critical weld. Cold proof testing every 40 months, followed by NDE, will not be performed since the crane operating temperature will not be less than 60oF.

(17)

According to the implementation section D.2 of the guide, the quality assurance discussion of paragraph C.5 is not applied to cranes ordered before September 1976. Although the LGS crane was ordered in August 1973, a quality assurance program was required of the vendor. However, this program could not and did not address the recommendations of paragraphs C.1 - C.4 as suggested in paragraph C.5.b of the guide.

(18)

The geared upper limit switch will be tested using a 10% load to demonstrate the adequacy of the limit switch. The geared upper limit switch is dependent only on the drum position, and not on the crane load.

The programmable limit switches that limit movement over the spent fuel pool will be tested using a 0% load to demonstrate the adequacy of the limit switches. The programmable limit switches are dependent only on the position of the bridge and trolley, and not on the crane load.

CHAPTER 09 9.1-87 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-12 (Cont'd)

The travel programmable limit switches on the bridge and trolley that prevent movement to the extreme ends of the bridge and trolley rails will be tested using a 0% load to demonstrate the adequacy of the limit switch. The travel limits on the bridge and trolley are dependent only on the position of the bridge and trolley, and not on the crane load.

The weighted upper limit switch will be tested using 100% of the rated capacity by raising the load from a position just off the floor at a maximum speed and manually tripping the weighted upper limit switch before the hook attains enough height to trip the limit switch.

Manual testing of the weighted upper limit switch will verify that the main hoist will stop raising and that all brakes will set with 100% of the rated capacity on the main hoist.

Keeping the load at a minimum height off the floor will enable the weighted upper limit switch to be tested with the load in a safer position.

(19)

The main hoist was tested at 125% of its rated load by raising the load off the floor and holding via the holding brakes then set back down. Not all positions of hoisting and bridge and trolley travel were tested with the 125% load. The auxiliary hoist was tested with at least 120% of its rated load by raising and lowering the load from minimum to maximum speed and holding the load via the holding brakes.

CHAPTER 09 9.1-88 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-13 UNIT 1 SPENT FUEL STORAGE RACK MODULES Module Number of Cells Module Envelope Size Shipping Number I.D. (Note # 1) Weight Of Cells N-S E-W N-S E-W (Pounds) Per Rack Direction Direction (Inches) (Inches) Note 2 A1 18 14 112.73 87.76 23300 252 A2 18 14 112.73 87.76 23300 252 A3 17 14 106.49 87.76 22400 238 B1 18 17 112.73 106.49 27100 306 B2 18 17 112.73 106.49 27200 306 B3 17 17 106.49 106.49 26500 289 C1 18 18 112.73 112.73 28300 324 C2 17 18 106.49 112.73 27600 306 D 17 17 106.49 106.49 25700 289 E1 17 14 106.49 87.76 22100 238 E2 12 17 75.27 106.49 19200 204 F 12 18 75.27 112.73 17400 191 G 17 13 106.49 85.51 20700 221 H 18 13 112.73 112.73 25100 249 Notes 1 See Figure 9.1-34 for rack locations.

2 Weight is dry weight of rack only, without fuel.

CHAPTER 09 9.1-89 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-14 SPENT FUEL POOL MATERIALS Material Description Specification Alloy Finish Fuel Pool Liner Plates ASTM A240 Type 304 Stainless steel Bars and shapes ASTM A479 Type 304 Stainless steel Control Rod Blade and Defective Fuel Storage Rack Perimeter bars ASTM A240 Type 304L Stainless Steel Canisters ASTM A240 Type 304L Stainless Steel Bottom grid assembly ASTM A240 Type 304L Stainless Steel Foot assemblies ASTM A564 Type 630 Stainless Steel Spent Fuel Storage Racks Sheet Metal Stock ASME SA240 Type 304L Stainless Steel Support Leg ASME SA240 Type 304L Stainless Steel (internally threaded)

Support Leg ASME SA564 Type 630 Stainless Steel (externally threaded)

CHAPTER 09 9.1-90 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-15 through Table 9.1-20 have been deleted CHAPTER 09 9.1-91 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-21 SPENT FUEL RACKS APPLICABLE CODES AND STANDARDS NUCLEAR REGULATORY COMMISSION Regulatory Guide 1.29 Seismic Design Classification (Rev 3), September 1979 Regulatory Guide 1.92 Combining Modal Responses and Spatial Components in Seismic Response Analysis (Rev 1), February 1976 Regulatory Guide 1.70 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (Rev 3), November 1978 SRP 3.8.4 Seismic Category I Structures, 1981 SRP 9.1.2 Spent Fuel Storage Review Responsibility, 1981 SRP 9.2.5 Ultimate Heat Sink, Pages 9.2.5-8 through 9.2.5-14, 1975 Regulatory Guide 1.38 Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants (Rev 2), 1977 Regulatory Guide 1.13 Spent Fuel Storage Facility Design Basis (Rev 1), December 1975 Regulatory Guide 1.61 Damping Values for Seismic Design of Nuclear Power Plants (Rev 0), October 1973 INDUSTRY CODES AND STANDARDS ASME Boiler and Pressure Vessel Code Sections III, V, IX, and Subsection NA Appendix I and XVII, 1977 Edition up to and including the winter 1978 Addendum AISC Steel Construction Manual AISC (7th Edition), June 1973, Including supplements AA Aluminum Construction Manual, Third Edition, April 1976, Specifications for Aluminum Structures AA Aluminum Standards and Data 5th Edition, January 1976 ASTM ASTM Standards: A240, A276, A312, B209, B26, B211, B221 (Latest revision in effect at time of purchase order unless otherwise specified.)

ANSI N45.2 Quality Assurance Requirements of Nuclear Power Plants, 1971 CHAPTER 09 9.1-92 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-21 (Contd)

ANSI N45.2.2 Packaging and Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants, 1972, except Paragraphs 2.4 and 2.6 ANSI N16.9 Validation of Calculation Methods for Nuclear Criticality Safety, 1975 ASNT-TC-1A American Society for Nondestructive Testing, 1975 Edition AWS D1.1 American Welding Society, Structural Welding Code (Rev 2), 1979 10CFR50 Code of Federal Regulations, Title 10, Part 50 (Appendix A & B)

ACI 318 (1971) Building Code Requirements for Reinforced Concrete ANSI/ANS 57.2 Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants ANSI/ANS 57.3 Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants MILITARY SPECIFICATIONS Mil-R-24243 Aluminum Rivet Specification Mil-A-8625C Anodizing CHAPTER 09 9.1-93 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-22 COMPLIANCE WITH ANS 57.1 FOR REACTOR ENCLOSURE CRANE Extent of Deviation and/or ANS 57.1 Guidelines(1) Compliance Justification 6.1 Safety Classification and Design Standards 6.1.e (Minimum industry Partial The specification for the crane was standard requirements) issued for purchase in 1973, prior to the issuance of the design standards in 6.1.e. The crane was designed to earlier versions of applicable ASME and CMAA standards than those listed in ANS 57.1.

Section 9.1.5 provides further information on design standards used for the crane.

6.2 System Design Requirements 6.2.1 Safety Requirements 6.2.1.1 (Interlock protection) Partial No underload indication or overload interlock is provided for the auxiliary hoist. The worst possible load drop from this hoist would be a ne fuel assembly drop into the spent fuel pool. The effects of this would be less severe than those resulting from a spent fuel handling accident (Section 15.7.4).

6.2.1.3 (Physical safety features) Full -

6.2.1.4 (Bridge travel Partial A warning gong can be manually annunciator) actuated from the operator's cab and from the radio controller.

6.2.1.5 (Fail-safe on full loss Full -

of power)

CHAPTER 09 9.1-94 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-22 (Cont'd)

Extent of Deviation and/or ANS 57.1 Guidelines(1) Compliance Justification 6.2.1.6 (Manual disconnect) Partial Radio and cab controls include a main power switch that can cut all power to the crane.

6.2.2 Analysis Requirements 6.2.2.1 (Design verification) Full -

6.2.2.2 (Seismic design) Partial The crane is designed as seismic Category I but is not required to function following an SSE.

All features required to prevent dropping the load are designed not to fail (Section 3.2).

6.2.3 Design Features 6.2.3.1 (Retaining devices) Partial Retaining devices are generally not provided because parts are never removed when the crane trolley is near the reactor vessel or spent fuel pool. Bead chains are not used.

6.2.3.2 (Fastener locking devices) Partial The suitable fastener devices listed are not used. The potential for parts falling into a fuel assembly is small because the crane trolley is restricted from traveling near the spent fuel pool and from carrying loads over the open reactor vessel (Section 9.1.5.2 and Reference 9.1-1).

6.2.3.3 (Surface finish) Full -

6.2.3.6 (Inertial loads) Full -

6.2.3.7 (Design for repair) Full -

6.2.3.9 (Manual movement) Partial Loads can be manually lowered after loss of power.

CHAPTER 09 9.1-95 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-22 (Cont'd)

Extent of Deviation and/or ANS 57.1 Guidelines(1) Compliance Justification 6.2.3.10 (Leakage collection) Partial All lubricated parts that can be positioned above the pool are provided with drip pans or other means to prevent lubricant from falling into the pool. Collection systems are not sized for 1.1 times the lubricant volume. However, the crane trolley is normally prohibited from traveling near the pool (Section 9.1.5.2).

6.2.3.12 (Load drop interlocks) Partial See ANS 57.1 Guideline 6.2.1.1 6.2.3.14 (Loudspeaker system) Partial A 6 channel public address system is provided in the drywell and refueling floor from the main control room. A telephone system is provided between the main control room and refueling floor. Both systems are powered by the plant emergency auxiliary system, which is backed up by emergency diesel generators. Circuits for the public address system and the telephone system are run in different conduits.

6.2.3.18 (Operator controls) Full -

6.2.3.20 (Corrosion) Full -

6.2.3.21 (Immersed components) Full -

6.2.3.22 (Materials properties) Full -

6.2.3.23 (Lamellar tearing) Partial Lamellar tearing is not specifically addressed in the crane design documents. Crane welding is in accordance with AWS D1.1-75 and AWS D14.1.

6.2.3.24 (Materials compatibility) Full -

CHAPTER 09 9.1-96 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-22 (Cont'd)

Extent of Deviation and/or ANS 57.1 Guidelines(1) Compliance Justification 6.2.3.25 (Lubricants and coatings) Full -

6.2.4 Specific Components Requirements 6.2.4.1 Handling Equipment 6.2.4.1.2 (Personnel clearances) Full -

6.2.4.1.5 (Area radiation monitors) Full -

6.2.4.1.9 (Automatically de-energize Full -

drive systems) 6.2.4.1.10 (High humidity) Full -

6.3 Testing and Maintenance Provisions 6.3.1 Testing 6.3.1.1 Electrical Test Full -

Capabilities 6.3.1.2 Mechanical Test Full -

Capabilities 6.3.2 Maintenance (standard Full -

parts)

CHAPTER 09 9.1-97 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-22 (Cont'd)

Extent of Deviation and/or ANS 57.1 Guidelines(1) Compliance Justification 6.3.2.1 Handling Equipment 6.3.2.1.3 Auxiliary Fuel Handling Full -

Crane (easy access) 6.3.2.1.9 General Requirements Partial The features listed are generally (safe disconnection of not used. The crane trolley is electrical hardware) normally prohibited from traveling near the spent fuel pool (Section 9.1.5.2).

(1)

Guidelines not listed are either not applicable to the fuel handling function of the reactor enclosure crane or are recommendations, not requirements, of ANS 57.1.

CHAPTER 09 9.1-98 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-23 SIZES AND WEIGHTS OF MAXIMUM DESNITY SPENT FUEL STORGE RACKS NUMBER OF CELLS MODULE ENVELOPE SIZE (See Note #1)

NUMBER OF MODULE N-S E-W SHIPPING CELLS I.D. Direction Direction N-S E-W WEIGHT RACK A1 18 14 112.77" 87.71" 22700# 252 A2 18 14 112.77" 87.71" 22700 252 A3 18 14 112.77" 87.71" 22700# 252 B1 18 17 112.77" 106.51" 27600# 306 B2 18 17 112.77" 106.51" 27600# 306 B3 18 17 112.77" 106.51" 27600# 306 C1 18 18 112.77" 112.77" 29200# 324 C2 18 18 112.77" 112.77" 29200# 324 D 17 17 106.51" 106.51" 26100# 289 E1 17 14 106.51" 87.71" 21500# 238 F 14 18 87.71 112.77" 22400# 248 G 17 18 106.51 112.77" 25800# 286 H 18 18 112.77" 112.77" 27000# 49 J 7 7 83" 83" 12200# 49

((Existing)

J 14 14 87.71" 87.71" 17700# 196 (Alternate)

- See Figure 9.1-35 through 9.1-37 for locations CHAPTER 09 9.1-99 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-24 REFERENCE CASE MAXIMUM DENSITY SPENT FUEL STORAGE RACK INPUT PARAMETERS Standard GE-9 GE-11 Siemens GE-14 GNF2 1 8x8 8x8 9x9 9x9 10x10 10x10 FUEL ROD DATA Clad outside diameter, in. 0.483 0.483 0.440 0.489 0.404 0.404 Clad inside diameter, in. 0.419 0.419 0.384 0.373 0.352 0.3567 Clad material Zr-2 Zr-2 Zr-2 Zr-2 Zr-2 Zr-2 UO2 density, g/cc UO2 10.503 10.566 10.42 10.42 10.42 10.42 (Tolerance +0.20)

Pellet diameter, in. 0.410 0.411 0.376 0.3665 0.345 0.3496 Enrichment, wt% U-235 4.90 4.90 3.50 3.50 3.50 3.50 (Tolerance +0.05)

WATER ROD DATA Number of water rods 2 1 2 1* 2 2 Inside diameter, in. 0.531 1.26 0.920 1.458 0.920 0.920 Outside diameter, in. 0.591 1.34 0.980 1.516 0.980 0.980 FUEL ASSEMBLY DATA Fuel rod array 8x8 8x8 9x9 9x9 10x10 10x10 Number of fuel rods 62 60 74 72 92 92 Fuel rod pitch, in. 0.640 0.640 0.566 0.569 0.510 0.510 Fuel channel, material Inside dimension, in. 5.278 5.278 5.278 5.278 5.278 5.283 Outside dimension, in. 5.478 5.438 5.478 5.438 5.438 5.433

• Square central channel 1

See Reference 9.1-18 CHAPTER 09 9.1-100 REV. 16, SEPTEMBER 2012

LGS UFSAR TABLE 9.1-25 MAXIMUM DENSITY SPENT FUEL STORAGE RACK EFFECT OF TEMPERATURE AND VOID ON REACTIVITY CASE INCREMENTAL REACTIVITY CHANGE, K 4C (39F) Reference 20C (68F) -0.003 80C (176F) -0.016 122C (252F) -0.028 122C, 20% Void -0.039 CHAPTER 09 9.1-101 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-26 MAXIMUM DENSITY SPENT FUEL STORAGE RACKS REACTIVITY UNCERTAINTIES DUE TO MANUFACTURING TOLERANCES QUANTITY NOMINAL VALUE TOLERANCE K Boron loading 0.0200 g/cm2 +0.0015 g/cm2 +0.0051 Boral width 5.00 inches +1/16 inches +0.0018 Lattice spacing 6.244 inches +0.040 inches +0.0027 SS thickness 0.075 and +0.008 inches +0.0004 0.035 inches mean Fuel enrichment 3.50% U-235 +0.05% U-235 +0.0033 Fuel density 10.42 g/cm3 +0.20 g/cm3 +0.0026 Statistical +0.0074 combination of uncertainties CHAPTER 09 9.1-102 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-27 MAXIMUM DENSITY SPENT FUEL STORAGE RACKS

SUMMARY

OF K FOR VARIOUS FUEL TYPES FUEL MAXIMUM K Standard (8x8) 0.944 GE 8x8 0.942 GE 9x9 0.941 Siemens - 9x9 0.934 GE 10X10 0.939 GNF2 - 10x101 0.9375 1

See Reference 9.1-19 CHAPTER 09 9.1-103 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-28 MAXIMUM DENSITY SPENT FUEL STORAGE RACKS

SUMMARY

OF CRITICALITY SAFETY ANALYSES CASMO KENO Temperature of analysis 4C 4C Gadolinia content None None Fuel enrichment, wt% U-235 3.50 3.50 Reference k 0.9366 0.92341 Calculational bias 0.0000 0.0101 Uncertainties Calculational +0.0024 +0.0018 KENO statistics NA +0.0011 Tolerances2 +0.0074 +0.0074 Removal of flow channel negative negative Eccentric position negative negative Statistical combination3 +0.0078 +0.0077 of uncertainties TOTAL 0.9366 + 0.0078 0.9335 + 0.0077 Maximum reactivity 0.9444 0.9412 1 Includes k correction of +0.0026 to 4C as determined by CASMO calculations.

2 See Table 9.1-26.

3 Square root of sum of square of all independent uncertainties.

CHAPTER 09 9.1-104 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-29 MAXIMUM DENSITY SPENT FUEL STORAGE RACK MATERIAL DATA RACK MATERIAL DATA (200F)

(ASME - Section II,Part D)

Material Young's Modulus E Yield Strength Sy Ultimate (psi) (psi) Strength Su (psi)

ASTM 240 27.6 x 106 25000(1) 71000(1) 304L S.S.

SUPPORT MATERIAL DATA (200°F)

1. SA-240, 27.6 x 106 25000(1) 71000(1)

Type 304L (upper part of support feet)

2. SA-564-630 27.9 x 106 106,300 140,000 (lower part of supportfeet age hardened at 1100F)

(1) The material purchase order requires that the minimum limits for mechanicals be those of SA240-304.

CHAPTER 09 9.1-105 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-30 LOAD COMBINATIONS AND ALLOWABLE STRESSES -

MAXIMUM DENSITY SPENT FUEL STORAGE RACKS There are two sets of criteria to be satisfied by the rack modules:

a. Kinematic Criteria In order to be qualified as a physically stable structure it is necessary to demonstrate that an isolated rack in water would not overturn when an event of magnitude 1.1 times the governing faulted seismic loading conditions is applied

b. Stress Limit Criteria Stress limits must not be exceeded under the postulated load combinations.

The following loading combinations ar applicable and are the bounding load combinations.

Loading Combinations Service Level D+L Level A D + L + To D + L + To + E D + L + To + SRV D + L + T2 + E Level B D + L + To + Pf D + L + T2 + E9 Level D D + L + To + Fd D + L + T2 + E9 + SRV + LOCA The functional capability of the fuel racks should be demonstrated.

CHAPTER 09 9.1-106 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-30 (Contd)

D = Dead weight-induced internal moments (including fuel assembly weight)

L = Live Load (not applicable for the fuel rack, since there are no moving objects in the rack load path)

Fd = Force caused by the accidental drop of the heaviest load from the maximum possible height specified in the PECO Specification Pf = Upward force on the racks caused by postulated stuck fuel assembly E = Operating Basis Earthquake (OBE)

E9 = Safe Shutdown Earthquake (SSE)

To = Differential temperature induced loads (normal operating or shutdown condition based on the most critical transient or steady state condition)

T2 = Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions)

SRV = Safety Relief Valve Loading LOCA = Loss-of-Cooling Accident Loads T2 and To produce local thermal stresses. the worst thermal stress field in a fuel rack is obtained when an isolated storage location has a fuel assembly generating heat at maximum postulated rate and surrounding storage locations contain no fuel. Heated water makes unobstructed contact with the inside of the storage walls,thereby producing maximum possible temperature difference between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is support pedestals do not experience secondary (thermal) stresses.

CHAPTER 09 9.1-107 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-30 (Contd)

Stress limits for Various Conditions Stress limits are derived form the ASME Code,Section III, Subsection NF Parameters and terminology are in accordance wit the ASME Code.

Normal and Upset Conditions (Level A or Level B)

a. Allowable stress in tension on a net section is:

Ft = 0.6 Sy (Sy = yield stress at temperature)

(Ft is equivalent to primary membrane stress)

b. Allowable stress in shear on a net section is:

Fv = .4 Sy

c. Allowable stress in compression on a net section (k )2 1 - /2C2c Sy Fa =

5 k k

{() + [3() /8Cc] - [()3/8Cc3]}

3 r r where:

(2n2E)

Cc = []1/2 Sy

= unsupported length of component k= length coefficient which gives influence of boundary conditions; e.g.

k= 1(simple support both ends)

= 2 (cantilever beam)

= 1/2 (clamped at both ends)

E= Young's Modulus CHAPTER 09 9.1-108 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-30 (Contd) r= radius of gyration of component kl/r for the main rack body is based on the full height and cross section of the honeycomb region.

d. Maximum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry is :

Fb = 0.60 Sy (equivalent to primary bending)

e. Combined bending and compression on a net section satisfies:

fa Cmxfbx Cmyfby

+ + <1 Fa DxFbx DyFby where:

fa = Direct compressive stress in the section ffx = Maximum bending stress along y-axis fby = Maximum bending stress along x-axis Cmx= Cmy = 0.85 fa Dx = 1 Fax fa Dy = 1 Fey 12 n2 E Fax,ey = 2 k

23 ()x,y r

and subscripts x,y reflect the particular bending plane.

CHAPTER 09 9.1-109 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-30 (Contd)

f. Combined flexure and compression (or tension) on a net section:

fa fbx fby

+ + <1.0 0.6Sy Fbx Fby The above requirements are to be met for both direct tension or compression.

Level D Service Limits Section F-1334 (ASME Section III, Appendix F), states that limits for the Level D conditions are the minimum of 1.2 (Sy/Ft) or (0.7Su/Ft) times the corresponding limits for the Level A condition. Su is ultimate tensile stress at the specified rack design temperature. For example, if the material is such that 1.2Dy is less than 0.7Su9 then the multiplier on the Level A limits, to obtain Level D limits, is 2.0. For the Limerick plant, however, a more conservative multiplier of 1.6 set in the UFSAR. All level D evaluations are based on this more conservative limit.

CHAPTER 09 9.1-110 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-31

SUMMARY

OF WPMR ANALYSIS RESULTS FOR 700# REGULAR FUEL LOADINGS AND COMPARISON WITH BOUNDING RESULTS OF SINGLE RACK ANALYSES Maximum Pedestal Maximum Displacements of Vertical CONDITIONS OF THE RUN Holtec Racks, in. Load, lbf Rack Top Baseplate SEISMIC x y x y 1

WPMR2-R; SSE + SRV 1.768 (R3) 3.491 (R3) 1.605 (R3) 3.443 (R3) 210,000 (R1-F4)1 14 Holtec racks and 1 +

existing 7x7 rack; 700# LOCA channeled fuel fully loaded SSE 3.079 (R3) 4.874 (R3) 2.869 (R3) 4.634 (R3) 205,900(R1-F4)

OBE 1.316 (R3) 2.623 (R3) 1.264 (R3) 2.622 (R3) 150,600 (R10-F3)

WPMR1RFR: SSE + SRV 1.091 (R7) .9466 (R3) 1.042 (R7) .6953 (R3) 208,900 (R10-F4) 15 Holtec racks; 700# +

channeled fuel fully LOCA loaded SSE 1.121 (R7) .8924 (R1) 1.072 (R7) .6558 (R3) 196,400 (R13-F4)

OBE .4880 (R7) .4488 (R7) .4938 (R7) .3510 (R3) 133,100 (R7-F2)

Bounding results from single rack 1.0482 1.0149 .9445 .6037 266,191 analysis with 700# channeled fuel loadings for (SSE+SRV + LOCA), SSE and OBE Seismic loadings CHAPTER 09 9.1-111 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-31 (Contd)

SUMMARY

OF WPMR ANALYSIS RESULTS FOR 700# REGULAR FUEL LOADINGS AND COMPARISON WITH BOUNDING RESULTS OF SINGLE RACK ANALYSES CONDITIONS OF THE RUN Maximum Maximum Maximum Fuel-Cell Rack-to-Wall Rack-to-Wall Maximum Stress Maximum Impact Impact Load, lbf Impact Load, lbf Factor of Rack Peak Dynamic Load Pedestals Water Pressure Per Cell on Wall, psi lbf.

SEISMIC Rack Top Baseplate Rack Top Baseplate Bounding results 1040 0 0 113.9 1650.9 .650 N/A from single rack analysis with 700# channeled fuel loadings for (SSE+SRV +

LOCA), SSE and OBE seismic loadings CHAPTER 09 9.1-112 REV. 13, SEPTEMBER 2006

LGS UFSAR TABLE 9.1-31 (Contd)

SUMMARY

OF WPMR ANALYSIS RESULTS FOR 700# REGULAR FUEL LOADINGS AND COMPARISON WITH BOUNDING RESULTS OF SINGLE RACK ANALYSES CONDITIONS OF THE RUN Maximum Maximum Maximum Fuel-Cell Rack-to-Wall Rack-to-Wall Maximum Stress Maximum Impact Impact Load, lbf Impact Load, lbf Factor of Rack Peak Dynamic Load Pedestals Water Pressure Per Cell on Wall, psi lbf.

SEISMIC Rack Top Baseplate Rack Top Baseplate WPMR2-R; SSE + SRV 1412.0 (R3) 0 0 4185 6570 .384 22.8 14 Holtec racks + No. 1381 No. 250 (R1-F4)2 and 1 LOCA (8) (44) existing 7x7 rack; 700#

channelled fuel fully loaded SSE 1165.1 (R3) 0 5599 4318 3903 .351 23.5 No. 212 No. 138 No. 214 (R1-F4)

(6) (8) (8)

OBE 851.8 (R3) 0 0 1421 1696 .258 17.8 No. 173 No. 222 (R7-F2)

(43) (16)

WPMR1RFR: SSE + SRV 690.8 (R3) 0 0 5271 4656 .371 24.6 15 Holtec racks; + No. 138 No. 214 (R10-F4) 700# LOCA (8) (8) channeled fuel fully loaded SSE 656.3 (R7) 0 0 3851 4818 .352 23.2 No. 138 No. 247 (R13-F4)

(8) (41)

OBE 542.9 (R3) 0 0 1560 2059 .265 17.1 No. 147 No. 249 (R7-F2)

(17) (43)

1) In parenthesis are rack number and foot number with the maximum value, see Figure 9.1-47.

2) Impact spring number indicated in tables; in the parenthesis is the impact spring number indicated on layout Figure 9.1-47.

CHAPTER 09 9.1-113 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.1-32 has been deleted CHAPTER 09 9.1-114 REV. 13, SEPTEMBER 2006

LGS UFSAR 9.2 WATER SYSTEMS 9.2.1 SERVICE WATER SYSTEM The service water system supplies cooling water required for normal plant operation. The SWS has no safety-related function.

9.2.1.1 Design Bases

a. The SWS is designed to remove heat from heat exchangers in the turbine, reactor, and radwaste enclosures, and to transfer this heat to the cooling towers where it is dissipated.

b. The SWS is designed to operate during normal plant operation and plant shutdown with offsite power available.

c. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the SWS are discussed in Section 3.2.

9.2.1.2 System Description The SWS, shown schematically on drawing M-10, is a single-loop cooling system utilizing three 50% capacity centrifugal pumps operating in parallel (one pump is on standby status). The pumps are located in the circulating water pump structure and circulate cooling water from the cooling tower, through the heat exchangers and back to the cooling tower. Each of the two generating units is provided with a separate SWS and cooling tower, although interconnections are provided so that either SWS can cool equipment common to both units. Design parameters for the major components of the SWS are listed in Table 9.2-1.

The SWS takes suction from the cooling tower basin. Water treatment to prevent long-term corrosion and organic fouling is discussed in Section 10.4.5.1.2.

During periods of low service water flow demand, the system is operated utilizing a single service water pump. System reliability is improved by operating a single service water pump closer to the rated capacity. To maintain back-up capability, an automatic pump start feature is available to be placed into service to automatically start the back-up pump in the event of a failure of the operating pump. This auto start feature is typically used during power operations when pump redundancy is desirable.

The service water flow to the heat exchangers is regulated by hand valves, temperature control valves, or solenoid-operated valves, as shown in drawing M-10.

The service water side (tube side) of the fuel pool heat exchangers is kept at a higher pressure than the shell side to prevent potential radioactive contamination of the service water, in case of a tube leak in a fuel pool heat exchanger. Pressure switches, located in the suction line to the fuel pool service water booster pumps, trip the pumps on low service water pressure.

During normal operation, the SWS supplies cooling water to various heat exchangers and coolers associated with the ESW system as shown in drawing M-11. In the event of a LOOP or LOCA, cooling for these components is automatically supplied by ESW. Switch-over to ESW cooling and isolation of the SWS from these components is discussed in Section 9.2.2.

The RECW and TECW heat exchangers are designed for nonessential service. However, they can be cross-connected to the ESW system (Section 9.2.2) so that in case of LOOP, the plant CHAPTER 09 9.2-1 REV. 20, SEPTEMBER 2020

LGS UFSAR operator can provide these heat exchangers with ESW. Isolation between the SWS and the ESW system for these heat exchangers is provided by two normally closed valves connected in series.

These valves are key-locked and remote manually operated. Isolation of return lines is provided by check valves and one normally closed valve connected in series.

Interties are provided between the Unit 1 and Unit 2 SWS suction and return headers to provide for continued operation of a unit's SWS during plant shutdown in which the associated cooling tower may be out-of-service for maintenance. In this mode, an SWS pump of the shutdown unit takes suction from the opposite unit's SWS suction header and return flow is routed to the opposite unit's return header to its cooling tower. Heat loads on the SWS are reduced when the unit is shut down, and nonoperating or unnecessary heat exchangers are valved out and throttling may be performed to maintain proper pump and flow conditions.

During a refuel outage, the service water system supports decay heat removal during certain periods. In the event of loss of all service water pumps on the refuel unit, the operating unit can supply a limited quantity of service water through the common service water header to support decay heat removal. The service water is returned to the operating unit via the return header intertie described above.

9.2.1.3 Safety Evaluation The SWS has no safety-related function. Failure of the system does not compromise any safety-related system or component, nor does it prevent a safe shutdown of the plant.

9.2.1.4 Tests and Inspections System operability is demonstrated by use during normal plant operation. The service water system on each unit is equipped with three pumps in parallel, each of which is capable of providing somewhat greater than 50% of the system requirements at design conditions. This capability provides sufficient margin for component operational degradation and permits the removal of one pump from service for maintenance without affecting the normal operation of the plant.

Operational degradation of service water system components would be detected by observation of inadequate cooling or increased pressure drop. Most of the heat exchangers required for normal plant operation are supplied with redundant equipment that may be valved into service. The potential for plant disruption due to service water system component trouble is therefore minimized.

9.2.1.5 Instrumentation Applications Local instrumentation is provided at the pumps and at the heat exchangers for maintenance, testing, and performance evaluation. The service water pumps discharge header is equipped with a pressure switch which sounds an alarm in the control room on low discharge pressure.

Radiation monitors in the service water return header to the cooling towers alarm in the control room on high or low radiation. The low alarm indicates monitor malfunction. See Section 11.5 for additional information on process radiation monitoring.

9.2.2 EMERGENCY SERVICE WATER SYSTEM The ESW system is designed to supply cooling water to selected equipment during a LOOP condition or LOCA. The ESW system is safety-related.

CHAPTER 09 9.2-2 REV. 20, SEPTEMBER 2020

LGS UFSAR 9.2.2.1 Design Bases

a. The ESW system is designed to supply cooling water to essential equipment during a LOOP or LOCA.

b. The ESW system is designed to withstand the most severe natural phenomenon or site-related event (e.g., earthquake, tornado, hurricane, flood, or transportation accident) without impairing its function.

c. The ESW system is designed to prevent inadvertent leakage of radioactivity to the environment.

d. The ESW system is designed with sufficient capacity and redundancy so that a single active failure cannot impair the capability of the system to perform its safety-related functions.

e. The ESW system is designed to include the capability for full operational testing.

f. The ESW system is designed with connections to provide cooling water to certain nonessential equipment during a LOOP or LOCA as directed by Emergency Operating Procedures.

g. The ESW system is designed to include the capability of operating one loop from the remote shutdown panel.

h. The ESW system is designed to seismic Category I requirements. The quality group classification and corresponding codes and standards that apply to the ESW system are discussed in Section 3.2.

9.2.2.2 System Description The ESW system is shown schematically in drawing M-11. Major equipment design parameters are listed in Table 9.2-2. It should be noted that the design parameters listed in Table 9.2.2 do not constitute performance requirements. ESW pump performance requirements are based on its ability to support the heat transfer rates listed in table 9.2-3. The system is common to Units 1 and 2, and consists of two independent loops (A and B), with two 50% system capacity (100% loop capacity) pumps per loop. Required heat removal rates for various heat exchangers and coolers served by ESW are listed in Table 9.2-3.

The ESW system is designed to supply cooling water to the following safety-related equipment:

a. RHR motor oil coolers

b. RHR pump compartment unit coolers

c. Core spray pump compartment unit coolers

d. Control room chillers

e. Standby diesel generator heat exchangers

f. RCIC pump compartment unit coolers CHAPTER 09 9.2-3 REV. 20, SEPTEMBER 2020

LGS UFSAR

g. HPCI pump compartment unit coolers

h. Spent fuel pools (makeup water)

In addition to the above equipment, emergency procedures direct providing ESW to the following nonsafety-related equipment during a LOOP:

a. RECW heat exchangers

b. TECW heat exchangers The reactor recirculation pump seal and motor coolers can be manually supplied with ESW as described in Section 9.2.8.2.

During normal plant operation all of the above equipment, with the exception of the diesel generators, is provided with cooling by the service water system.

Essential heat loads normally cooled by the service water system are automatically transferred to the ESW system under LOOP and LOCA accident conditions. This transfer is accomplished by the following automatic valve realignments:

a. The normally open check valve in each service water supply line closes as a result of the trip of the service water pumps, on a loss of offsite power, and the start of the ESW pumps.

b. The normally closed check valve in each ESW supply line opens as a result of the start of the ESW pumps.

c. The normally open isolation valves installed in series in each service water discharge line close on an ESW pump start signal.

d. The normally closed isolation valves installed in parallel in each ESW discharge line open on an ESW pump start signal.

The ESW pumps start automatically on diesel generator operation (e.g., diesel 1A or 2A cause ESW pump A to start) after speed, voltage, and bus breaker conditions are met, and after a load sequencing delay. ESW pump operation causes automatic valve and sluice gate realignments to:

a. Take pump suction from the spray pond.

b. Provide ESW to safety-related equipment.

c. Return the ESW to the spray pond via the RHRSW system and the spray networks.

The spray pond is described in Section 9.2.6. The operator may elect to realign the system to utilize the cooling towers, if they are available.

The components listed on Table 9.2-3 are provided with cooling water from either ESW loop A or B. Design heat removal rates are also shown in Table 9.2-3. Each diesel generator can be supplied with cooling water from ESW loop A or loop B. Normal system alignment, however, is CHAPTER 09 9.2-4 REV. 20, SEPTEMBER 2020

LGS UFSAR such that loop A supplies cooling water to the A and C diesel generators, and loop B supplies the B and D diesel generators.

Each loop is designed such that ESW flow to one unit can be isolated without adversely affecting flow to the other unit.

The valves on an ESW branch line serving a diesel generator are fed from power supplies of the same channel as the diesel generator. All other safety-related components in the ESW system are powered by channels A or C if they are in ESW loop A, or by channels B or D if they are in ESW loop B.

The ESW pumps can be manually started from the control room. The A ESW pump and associated A loop valves can be operated from the remote shutdown panel. The B and C ESW pumps can also be operated from the B and C pumps motor circuit breaker cubicles, respectively.

The ESW pumps are located in the spray pond pump structure. The A and B loop pumps are separated by physical barriers. The pump structure is described in Section 3.8.4. The spray pond pump structure is designed so that sufficient NPSH and adequate hydraulic conditions for proper pump operation are maintained down to a pond level below that associated with the design basis water loss. To verify adequate hydraulic conditions in the pump structure wet pits for proper long-term pump operation, it has been determined from model studies that acceptable hydraulic conditions exist down to elev. 243-6. Drawing M-389 shows the elevations of the spray pond water levels, suction piping, and the ESW pump.

ESW loop A and B piping is physically separated or protected so that no single postulated event can impair the systems capability to perform its required safety functions. The ESW is combined with the RHRSW before it is returned to the spray pond or cooling tower. The return from each ESW loop is connected to both the A and B RHRSW loops.

Each ESW loop is also provided with two crossties to the RHRSW System and associated isolation valves. One crosstie is installed in a valve pit near the Spray Pond Pumphouse and the second crosstie is installed in the Reactor Building Pipe Tunnel. The crosstie lines and associated isolation valves allow the underground portion of the ESW or RHRSW supply piping of a given loop to be isolated for inspection or repairs without interrupting the supply of cooling water to the ESW or RHRSW System. Operating restrictions, however, apply to certain crosstied configurations and the use of these lines is administratively restricted.

The ESW supply and return yard piping is made of carbon steel, with approximately 200 mils corrosion allowance. All piping outside of the pump structure, main plant, and spray pond is coated for corrosion protection. All heat exchangers, except those in the RHR pump motor coolers have 90/10 Cu-Ni tubes. The 2A-P202 pump motor has type 316 stainless steel tubes in the oil coolers.

The remaining RHR pump motor oil coolers have 90/10 Cu-Ni tubes. The heat transfer rates for essential components are verified by testing to ensure that the design basis heat transfer rate is being maintained. The fouling factor for each component is typically 0.002 hr-ft2-°F/Btu. Water treatment provisions to prevent fouling are discussed in Section 9.2.6.

Radiation detectors are located on the combined ESW/RHRSW system return lines to detect inleakage of radioactive water. Only reactor recirculation pump seal coolers have a potential for radioactive inleakage to the ESW system. However, this line contains locked valves which further reduces the potential for inleakage into the ESW system. Control room alarms are provided on CHAPTER 09 9.2-5 REV. 20, SEPTEMBER 2020

LGS UFSAR high radiation. Radioactive leakage from the ESW system is controlled by closing the ESW return valves for the loop where the leak has been identified and realigning the system for operation from the alternate loop. These actions are initiated manually from the control room.

A significant difference between inlet and outlet flows in each ESW loop is indicated for net system water loss or inleakage and annunciated in the control room for net water loss only. Flooding detectors annunciate high water level in the RHR, RCIC, HPCI, and core spray pump compartments and the diesel generator compartments.

Credible pump leakage rates would be of no consequence because the design flow rate of each pump is approximately 6400 gpm. As stated in Section 9.2.2.2, a significant difference between the common pump discharge flow and return flow from each ESW loop is indicated and annunciated in the control room. This differential flow is indicative of ESW loop leakage, other than pump leakage, and provides the operator with indication of system degradation. The instrumentation is shown in drawing M-11.

If the lead cooler in any ECCS/RCIC pump compartment degrades to the point that room temperatures exceed desired levels, the redundant/standby cooler will automatically start. In the event that cooling capability degrades to the point that the compartments cannot be effectively cooled, high compartment temperature will be annunciated in the control room. Operator corrective actions could entail cooler repair, switching to the use of pumps located in another compartment, diverting more ESW flow to the compartment from other nonessential cooling loads, and/or opening compartment doors to facilitate cooling by natural circulation. After a DBA LOCA, use of ECCS equipment in another compartment is the most likely operator response due to high radiation conditions in the Reactor Enclosure. No credit has been taken for manual operator actions in the Reactor Enclosure in this scenario. One room unit cooler per ECCS/RCIC loop may be removed from service at the same time under administrative controls. Failure of both unit coolers in a HPCI, RCIC or Core Spray compartment is no worse than failure of the supported HPCI, RCIC or Core Spray equipment in the compartment because each compartment contains equipment associated with only one ECCS/RCIC loop. However, each RHR compartment contains LPCI equipment and two LPCI loops. Thus, a failure in one LPCI loop or a failure in one LPCI loop support system (such as the air cooling system or a power supply) could impact the other LPCI loop's equipment in the room. If both LPCI pumps' lead and standby unit coolers are in service prior to the DBA LOCA, no single active failure will result in less than the necessary number of coolers remaining functional to limit the RHR room temperature to less than 140 Degrees F.

Analysis has demonstrated that any two unit coolers can adequately cool the RHR room to support both LPCI loops. If one or both LPCI pump standby unit coolers is/are out of service prior to the DBA LOCA (with the associated lead unit cooler in service), a single active failure could result in only one functional unit cooler in one RHR room, in which case that RHR room's temperature may exceed 140 Degrees F. Administrative controls are in place to assess and limit the risk of such maintenance configuration for the ECCS/RCIC room unit coolers. These coolers are a non-Tech Spec support system.

9.2.2.3 Safety Evaluation All safety-related components (including supporting structures) of the ESW system are designed to seismic Category I requirements, as defined in Section 3.7. The piping is designed, fabricated, inspected, and tested in accordance with the requirements of ASME Section III, Class 3. The ESW system, with the exception of the buried piping and the piping in the spray pond, is housed within either the reactor enclosure or the spray pond pump structure, both of which are designed to the CHAPTER 09 9.2-6 REV. 20, SEPTEMBER 2020

LGS UFSAR seismic Category I requirements discussed in Section 3.8.4. Evaluation of the ESW system with respect to the following areas is discussed in separate sections as indicated:

a. Protection from wind and Section 3.3 tornado effects

b. Flood design Section 3.4

c. Missile protection Section 3.5

d. Protection against dynamic Section 3.6 effects associated with the postulated rupture of piping

e. Environmental design Section 3.11

f. Fire protection Section 9.5.1

g. Icing and freezing Section 9.2.6 Each of the two independent ESW system loops is isolated from the other by barriers, separate trenches, or distance to ensure that simultaneous loss of both loops cannot occur. Power is supplied from four independent divisions. Failure of either a MOV, diesel generator, electrical division, or pump does not prevent the system from removing the full heat load. This arrangement ensures that the full heat removal capacity required is available after a postulated single active failure. See Table 9.2-4 for a failure mode and effects analysis of the ESW system.

Although not specifically included in the ESW system design basis, the cooling provided by a single ESW loop is sufficient to safely shut down both units at all times with the exception of the first 10 minutes of a postulated DBA LOCA. During this time, three RHR pumps may be required on the LOCA unit for flooding (LPCI mode). The RHR pumps are arranged so that only two can be cooled by each ESW loop. See section 6.3.2.5 for a discussion of single failure modes and effects on ECCS reliability.

Connections between the ESW system and the normal service water system, which supplies the equipment other than the diesel generators during normal plant operation, are provided with automatic AOVs and check valves. These isolate the service water system from the ESW system, so that a failure in the normal service water system does not impair ESW system operation. The check valves used at the interface of the essential (ESW) from nonessential (service water) piping is verified to be operable and to have less than 10 gpm leakage as part of the LGS ISI Program.

Leakage of this magnitude will not compromise the safety function of the ESW system. Thus, even with this leakage at each interface, the ESW system could withstand a single active failure and still perform its safety function. In addition, seismic Category I manual gate valves are provided upstream of each interface check valve. These valves can be closed for more effective blocking during long-term operation of the ESW system (drawing M-11).

The crosstie piping between the ESW and RHRSW supply loops is provided with manual isolation valves which are locked in the safe position to prevent inadvertent mispositioning of these valves from disabling these systems.

Each ESW loop return is intertied to both ESW/RHRSW combined return headers to the spray pond, so that loss of a combined return header does not cause the loss of an ESW loop.

CHAPTER 09 9.2-7 REV. 20, SEPTEMBER 2020

LGS UFSAR During certain maintenance configurations, one ESW loop may be lost due to the loss of a single ESW/RHRSW combined return header. However, sufficient redundancy and heat removal capacity remains in the other ESW loop such that ESW can perform its safety function.

Each ESW pump motor is powered from a separate Class 1E bus. Upon LOOP, the standby diesel generators automatically start to supply power to their respective buses, and thus power to the safety-related components of the ESW system. See Section 8.3 for a discussion of the onsite power system.

9.2.2.4 Tests and Inspections The ESW system is preoperationally tested in accordance with the requirements of Chapter 14, and periodically tested in accordance with the requirements of Chapter 16. Inservice inspection is in accordance with the ASME Section XI for Section III, Class 3 components, or 10 CFR 50.69 Alternative Treatment, when applicable. Refer to Section 6.6.1 for further information.

9.2.2.5 Instrumentation Applications The ESW system is designed for remote operation from the control room. In addition, one pump and the valves associated with its loop can be operated from the remote shutdown panel. A second pump in loop A can be operated using local controls at the C pump motor circuit breaker cubicle. The loop B pump can be operated using local controls at the B pump motor circuit breaker cubicle.

Local and remote indication is provided to monitor process parameters of the system. The following alarms are annunciated in the control room: pump discharge low pressure; flow differential high between influent and effluent showing net water loss. Temperature elements on pump motor provide input to plant computer.

Radiation alarms from the combined ESW/RHRSW return headers are annunciated in the control room.

9.2.3 RHR SERVICE WATER SYSTEM The RHRSW system is a safety-related system, designed to supply cooling water to the RHR heat exchangers of both units.

9.2.3.1 Design Bases

a. The RHRSW system is designed to provide a reliable source of cooling water for all operating modes of the RHR system, including heat removal under postaccident conditions. It also provides water to flood the reactor core, or to spray the primary containment after an accident, if necessary.

b. The RHRSW system is designed to withstand the most severe natural phenomenon or site-related event (e.g., earthquake, icing conditions, tornado, hurricane, flood, or transportation accident) without impairing its function.

CHAPTER 09 9.2-8 REV. 20, SEPTEMBER 2020

LGS UFSAR

c. The RHRSW system is designed with sufficient capacity and redundancy so that a single failure of any active component, assuming the loss of offsite power, cannot impair the capability of the system to perform its safety-related functions.

d. The RHRSW system is designed to include the capability for testing through the full operational sequence that brings the system into operation for reactor shutdown and for LOCA.

e. The RHRSW system is designed to limit the possibility of any radioactive material release to the environment.

f. The RHRSW system is designed to include the capability of operating one loop from the remote shutdown panel.

g. The RHRSW system is designed to seismic Category I requirements. The quality group classification and corresponding codes and standards that apply to the design of the RHRSW system are discussed in Section 3.2.

9.2.3.2 System Description The RHRSW system is shown schematically in drawing M-12. Major equipment design parameters are given in Table 9.2-5. It should be noted that the design parameters listed in Table 9.2-5 do not constitute performance requirements. RHRSW pump performance requirements are based on the pumps ability to support heat transfer regardless of discharge head. The system is common to the two reactor units, and consists of two loops. Each loop services one RHR heat exchanger in each unit, and provides sufficient cooling for safe shutdown, cooling, and accident mitigation of both units. The two RHRSW system return loops are cross-connected for flexibility.

Two valves in series are provided on the cross-connect, so failure in one loop cannot affect the operation of the other. Each loop has two pumps located in the spray pond pump structure. One pump supplies 100% flow to one RHR heat exchanger. During two-unit operation, there are two heat exchangers (one per unit), and therefore, two of the four pumps are required for safe shutdown and accident mitigation. Description of the pump structure is found in Section 3.8.4. The RHR heat exchangers are described in detail in Section 5.4.7. The spray pond piping network is described in Section 9.2.6.

Spray pond pump structure wet pits are designed so that sufficient NPSH and adequate hydraulic conditions for proper pump operation are maintained down to a pond level below that associated with the design basis water loss. To verify adequate hydraulic conditions in the pump structure wet pits for proper long-term pump operation, it has been determined from model studies that acceptable hydraulic conditions exist down to elev. 243-6. Drawing M-389 shows the elevations of the spray pond water levels, suction piping and the RHRSW pump.

The RHRSW flows through the tube side of the RHR heat exchangers; the tubes of which are made of corrosion-resistant AL-6XN (Unit 1 and Unit 2B) and 304L stainless steel (Unit 2A). To reduce corrosion, the tube side of the heat exchangers can be laid up with demineralized water when the RHRSW system is not in operation.

It should be noted that a corrosion monitoring system was installed when 304SS tubes were in the Unit 1 RHR heat exchangers. The Unit 1 heat exchanger tubes have been changed to AL6XN material and there is no longer a need to actively monitor the corrosion of the original 304SS tube bundles. Accordingly the monitoring equipment has been placed in a de-energized and isolated CHAPTER 09 9.2-9 REV. 20, SEPTEMBER 2020

LGS UFSAR inactive status. The design and minimum heat transfer rates are the same and are shown on drawing E11-1020-G-002 for normal and accident modes of RHR system operation. These heat transfer rates are based on the assumption that 5% of the RHR heat exchanger tubes are plugged and the remainder are fouled, RHR service water temperature is at a maximum, and RHR service water flow is at a minimum. Actual heat transfer rates will be greater than the minimum values based on actual tube and service water conditions plus the manufacturer's design conservatisms.

Credible pump leakage rates will be of no consequence. Excessive leakage in the RHRSW system would cause a high temperature alarm in the RHR system, a flooding alarm in the RHR room, or a low flow indication for RHRSW in the RHR heat exchanger inlet. The redundant RHRSW loop can be used while the leakage problem is being corrected.

The RHRSW system supply and return yard piping is made of carbon steel with approximately 200 mils corrosion allowance. A portion of the RHRSW return piping within the pipe tunnel is stainless steel. All piping outside the pump structure, main plant, and spray pond is externally coated for corrosion protection. Exposed spray pond piping is painted. Water treatment provisions to prevent fouling are discussed in Section 9.2.6. The spray networks can be drained after use to minimize corrosion and provide protection from freezing.

The RHRSW system is available for normal shutdown or emergencies, and does not operate during normal power generation, except that the RHRSW system can be used in conjunction with the RHR system suppression pool cooling mode to maintain the suppression pool below specified temperature limits, and RHRSW can be operated in the spray mode for spray pond cooling and chemistry control purposes.

For two-unit operation, during a plant operating mode in which one unit is in an accident shutdown (LOCA) and the other unit is in a normal shutdown (including LOOP occurring during normal shutdown), with one RHRSW loop in service (two RHRSW pumps in operation, and one ESW pump in operation using one spray pond return header and two spray networks), the unit undergoing normal shutdown will be supplied with 5570 to 8000 gpm of RHRSW flow to the RHR heat exchanger. The RHRSW cooling water flow to the unit under normal shutdown is sufficient to remove the heat load on its RHR heat exchanger. The time to cool the RPV down to 200F will increase but will remain within the shutdown requirements of the Technical Specifications. The unit in an accident shutdown will still be supplied with 8000 GPM of RHRSW flow to the RHR heat exchanger. The required RHRSW flows to each RHR heat exchanger can be attained by adjusting HV-51-2F068A/B or HV-51-1F068A/B. RHRSW flow to the RHR heat exchangers can be monitored through FI-51-1R602A/B and FI-51-2R602A/B.

The RHRSW pump motors obtain their power from separate Class 1E buses; the A and B pumps from Unit 1 buses A and B, respectively, and the C and D pumps from the Unit 2 buses A and B, respectively. If LOOP occurs, the diesel generators start automatically, providing emergency power to the buses. The pumps are started manually. It is not necessary to secure any of the ECCS pumps before starting the RHRSW pumps. The RHRSW valves obtain their power from the safeguard buses, with loop A valves receiving power from Division 1 and 3, and loop B valves from Division 2 and 4. The RHR heat exchanger inlet valves are powered by different electrical sources.

Specifically, HV51-1F014A and RHRSW pump OAP506 are powered by Division I, Unit 1 while HV51-1F014B and RHRSW pump OBP506 are powered by Division II, Unit 1. In the event of a power supply failure, an RHR heat exchanger is removed from service by manually closing the heat exchanger service water outlet valve.

CHAPTER 09 9.2-10 REV. 20, SEPTEMBER 2020

LGS UFSAR The RHRSW return and the return from both ESW loops share a common return header to the spray pond. Loss of one RHRSW/ESW return loop does not affect the capability of the second return loop to safely shut down either or both units during emergency conditions.

Under certain maintenance configurations, flow from the two return headers may be combined in one line for a limited period of time. Any active valves which could fail and disable this line will be blocked in the safe position. Passive failures which could cause the total failure of this line during this limited duration have been evaluated and are not considered credible.

Upon standby emergency diesel generator or ESW pump start, the RHR service water system automatically aligns itself to the spray pond mode, if it is not already in that mode. If the cooling tower mode is available, the system can be manually aligned to it. Bypass lines are provided to discharge water directly to the pond, rather than the spray networks, during periods when the pond is frozen.

Double remotely operated isolation valves are provided on the cross-tie lines between the RHRSW system and the RHR pump discharge, for flooding the containment or reactor core, if such action is necessary and no other source of water is available.

Radiation monitors are provided in the RHRSW system return headers to the spray pond. High activity in the return header causes an alarm and the pumps associated with that loop to stop.

Overrides are provided to restart the pumps.

For detection of possible RHRSW system pipe breaks, flow indicators on each RHR heat exchanger inlet, and RHR room flooding detectors that annunciate on high water level, are provided.

9.2.3.3 Safety Evaluation Safety-related components (including supporting structures) of the RHRSW system are designed to seismic Category I requirements, as defined in Section 3.7. The piping is designed, fabricated, inspected, and tested in accordance with the requirements of ASME Section III, Class 3 except as noted in Table 3.2-7. The RHRSW system, with the exception of the buried piping and the piping in the spray pond, is housed within either the reactor enclosure or spray pond pump structure, both of which are designed to seismic Category I requirements as discussed in Section 3.8.

Evaluation of the RHRSW system with respect to the following areas is discussed in separate sections as indicated:

a. Protection from wind and Section 3.3 tornado effects

b. Flood design Section 3.4

c. Missile protection Section 3.5

d. Protection against dynamic Section 3.6 effects associated with the postulated rupture of piping CHAPTER 09 9.2-11 REV. 20, SEPTEMBER 2020

LGS UFSAR

e. Environmental design Section 3.11

f. Fire protection Section 9.5.1

g. Icing and freezing Section 9.2.6 There are two RHRSW loops, each loop serving one RHR heat exchanger in each unit, to supply cooling water for plant shutdown. Each loop is isolated from the other by barriers, separate trenches, or distance to ensure that simultaneous loss of both loops cannot occur. Failure of either a MOV, a diesel generator, or pump does not prevent the system from performing its safety function. This arrangement ensures that the full heat removal capacity required is available after the postulated active failure of a single component. See Table 9.2-6 for a failure mode and effects analysis of the RHRSW system.

Motors of the four RHRSW system pumps are connected to separate Class 1E buses.

9.2.3.4 Tests and Inspections The RHRSW system is preoperationally tested in accordance with the requirements of Chapter 14, and periodically tested in accordance with the requirements of Chapter 16. Inservice inspection is in accordance with ASME Section XI for Section III, Class 3 components, or 10 CFR 50.69 Alternative Treatment, when applicable. Refer to Section 6.6.1 for further information.

9.2.3.5 Instrumentation Applications The RHRSW system is designed for remote operation from the control room. Each pump has a discharge pressure switch which causes an alarm in the control room upon low discharge pressure. Each loop has a pressure transmitter in the common supply line for pressure indication in the control room.

The RHRSW supply line flow to each heat exchanger is indicated in the control room. Pressure and temperature in the return lines from each heat exchanger are indicated in the control room.

High radiation alarms and radiation recorders are provided in the control room for each common return line. Temperature elements on the pump motors provide input to the plant computer.

Loop A pumps and valves can also be operated from the remote shutdown panel and pressure indication for the common supply line is provided. Loop B pump can be operated using local controls at the B pump motor circuit breaker cubicle.

9.2.3.6 Regulatory Commitments The following table provides a description of the regulatory commitments identified in Exelon letter dated October 29, 2010 (Ref. 9.2-29), Attachment 4, as updated by supplemental letters dated December 3, 2010 (Ref. 9.2-30), and March 23, 2011 (Ref. 9.2-31), and is included in the Updated Final Safety Analysis Report as required by Amendment Nos. 203 to Facility Operating License No. NPF-39 and 165 to Facility Operating License No. NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively, dated July 29, 2011 (Ref. 9.2-32).

The amendments consist of changes to the LGS Technical Specifications (TS) of each unit extending the allowed outage time (AOT) from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to seven (7) days for the Suppression Pool Cooling mode of the Residual Heat Removal (RHR) system, the Residual Heat Removal CHAPTER 09 9.2-12 REV. 20, SEPTEMBER 2020

LGS UFSAR Service Water (RHRSW) system, the Emergency Service Water (ESW) system, and the

[Alternating Current] A.C. Sources - Operating (Emergency Diesel Generators [EDGs]).

The specific purpose of the amendments is to provide a 7-day TS AOT window to allow for repairs of the common RHRSW system piping. The AOT extension is only allowed once every other calendar year, for each unit, while one unit is operating and the opposite unit is shut down, with the reactor vessel head removed and rector cavity flooded.

The piping repairs are intended to repair certain known areas of degraded RHRSW return piping, and are expected to take several refueling outages on each unit to complete the full scope of repairs. However, the TS changes may be used for other future RHRSW piping maintenance activities, as necessary, as long as the appropriate TS restrictions and the regulatory commitments identified below are observed.

These regulatory commitments will be controlled as part of a special procedure developed specifically to govern plant operation while in the extended TS AOTs to perform RHRSW system piping repairs.

Technical Specification 3.7.1.1 Action a.3.a and a.3.b were changed to allow the use of the Risk Informed Completion Time (RICT) Program per Amendment Nos. 240 to Renewed Facility Operating License No. NPF-39 and 203 to Renewed Facility Operating License No. NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively, dated 2/28/2020. The amendments will allow the AOT to be increased to a maximum of 30 days based on the associated plant risk to the operating Unit.

Regulatory Commitments

1. The following action will be taken prior to entry into the proposed configuration:

• Proper standby alignment of the operable RHRSW subsystem will be ensured by local verification of boundary valve and power supply position in accordance with a check-off list contained in the special procedure developed specifically to govern plant operations in the extended AOTs.

2. Also, the following actions will be taken prior to entry into the proposed configuration:

• When the 'A' RHRSW subsystem is inoperable to allow for repairs of the RHRSW A subsystem piping with Limerick Generating Station Unit 2 shutdown, reactor vessel head removed and reactor cavity flooded, the following equipment will be verified as available and protected as defined in procedure OP-AA-108-117:

o ESW loop A o Unit 1 RHR subsystems A and C o D11, D13, and D23 4kV buses and emergency diesel generators o Unit 1 Division 1 and Division 3 Safeguard DC

• When the 'A' RHRSW subsystem is inoperable to allow for repairs of the RHRSW A subsystem piping with Limerick Generating Station Unit 1 shutdown, reactor vessel head removed and reactor cavity flooded, the following equipment will be verified as available and protected as defined in procedure OP-AA-108-117:

CHAPTER 09 9.2-13 REV. 20, SEPTEMBER 2020

LGS UFSAR o ESW loop A o Unit 2 RHR subsystems A and C o D11, D21, and D23 4kV buses and emergency diesel generators o Unit 2 Division 1 and Division 3 Safeguard DC

• When the 'B' RHRSW subsystem is inoperable to allow for repairs of the RHRSW B subsystem piping with Limerick Generating Station Unit 2 shutdown, reactor vessel head removed and reactor cavity flooded, the following equipment will be verified as available and protected as defined in procedure OP-AA-108-117:

o ESW loop B o Unit 1 RHR subsystems B and D o D12, D14, and D24 4kV buses and emergency diesel generators o Unit 1 Division 2 and Division 4 Safeguard DC

• When the 'B' RHRSW subsystem is inoperable to allow for repairs of the RHRSW B subsystem piping with Limerick Generating Station Unit 1 shutdown, reactor vessel head removed and reactor cavity flooded, the following equipment will be verified as available and protected as defined in procedure OP-AA-108-117:

o ESW loop B o Unit 2 RHR subsystems B and D o D12, D22, and D24 4kV buses and emergency diesel generators o Unit 2 Division 2 and Division 4 Safeguard DC

3. Activities in the switchyard that adversely affect risk exposure are those that have the potential to cause a loss of offsite power, such as testing and maintenance activities.

Therefore, testing and discretionary maintenance will be prohibited during the RHRSW subsystem piping repairs in the at-power unit switchyard and on equipment in the outage unit switchyard supporting operability of its offsite source. Accordingly, during the RHRSW subsystem piping repairs, the at-power unit switchyard will be protected in its entirety using either a lock and/or chain different than that used for normal access to the switchyard, or a physical barrier placed in front of the gate used for normal access to the switchyard. In addition, equipment in the outage unit switchyard supporting operability of its offsite source will be protected during the RHRSW subsystem piping repairs using protected equipment signs and physical barriers, such as barrier rope, physical devices, tape, etc., to prevent access to the equipment. This will be controlled through applicable corporate and station procedures for equipment protection, and through the special procedure developed specifically to govern plant operation while in the extended AOTs.

4. Operational Risk Activities (ORAs), as defined in procedure WC-AA-104, involve activities on risk significant systems that have the potential to derate the plant, i.e., cause a loss of planned generation. Typical ORAs involve: an activity that could cause equipment actuations that could cause loss of planned generation; instrument, fuse, or circuit board removal/installation; an activity that will cause a 1/2 scram or 1/2 trip; pressurization of common instrument sensing lines; placing of jumpers or lifting energized leads; an activity that could cause vibration or impact near operational risk sensitive equipment, etc. Such activities will be prohibited on the online unit during the RHRSW piping repairs. Exceptions to this must be approved by the senior plant management.

This will be controlled as part of the special procedure developed specifically to govern plant operation while in the extended AOTs.

CHAPTER 09 9.2-14 REV. 20, SEPTEMBER 2020

LGS UFSAR

5. The extended weather forecast will be examined to ensure severe weather conditions that would threaten the loss of offsite power are not predicted prior to entry into the AOT.

In the event of an unforeseen severe weather condition due to rapidly changing conditions, such as severe high winds, a briefing with crew operators will be performed to reinforce operator actions and responses in the event of a loss of offsite power (E-1 0/20). This will be controlled via the special procedure developed specifically to govern plant operation while in the extended AOTs.

6. Shift briefs will be performed to reinforce other potentially important operator actions associated with the performance of the extended AOT (i.e., operator actions to refill the condensate storage tank (CST), operator actions to vent containment, operator actions to maximize control rod drive (CRD) injection to the vessel, and operator actions to support continued use of feedwater and condensate post-trip as necessary and if available).

Additionally, during the 'A' RHRSW subsystem outage, a shift brief on alternate remote shutdown operations will be performed since some of the normally operated equipment from the remote shutdown panel will not be available. This will be controlled via the special procedure developed specifically to govern plant operation while in the extended AOTs.

7. Unattended transient combustibles and hot work will be prohibited in the areas listed below during the extended AOT. This will be controlled via the special procedure developed specifically to govern plant operation while in the extended AOTs.

For an 'A' RHRSW subsystem outage window:

• Fire Area 15, Unit 1 Division 2 (012) safeguard 4kV switchgear room

• Fire Area 17, Unit 2 Division 2 (022) safeguard 4kV Switchgear room

• Fire Area 24, Main Control Room (ECCS B panel 10-C601 (Bay A, B))

• Fire Area 24, Main Control Room (ECCS B panel 20-C601 (Bay A, B))

• Fire Area 25, Auxiliary Equipment Room For an 'B' RHRSW subsystem outage window:

• Fire Area 13, Unit 1 Division 1 (011) safeguard 4kV Switchgear room

• Fire Area 19, Unit 2 Division 1 (021) safeguard 4kV Switchgear room

• Fire Area 24, Main Control Room (ECCS A panel 10-C601 (Bay C, D, E, F))

• Fire Area 24, Main Control Room (ECCS A panel 20-C601 (Bay C, D, F))

• Fire Area 25, Auxiliary Equipment Room

• Fire Area 26, Remote Shutdown Panel 9.2.4 CLARIFIED WATER SYSTEM The plant clarified water system provides filtered, clarified river water for use as lubricating water, and as the input stream for the makeup demineralizer system.

The plant clarified water system, shown in drawing M-17 is not safety-related, and does not convey radioactive material.

9.2.4.1 Design Bases CHAPTER 09 9.2-15 REV. 20, SEPTEMBER 2020

LGS UFSAR

a. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the clarified water system is discussed in Section 3.2.

b. Raw water is clarified and filtered to an effluent turbidity not exceeding 1.0 NTU and is stored in the clarified water storage tank.

c. The clarified water system is not connected to any systems having the potential for containing radioactive material.

d. The clarified water system is designed to meet maximum anticipated plant water demands.

9.2.4.2 System Description The clarified water system is shown in drawing M-17.

Influent for the clarified water system is supplied by the Perkiomen Creek or Schuylkill River. The river water turbidity is reduced in the raw water clarifier by chemical addition. The clarified water from the clarifier outlet flows by gravity to the clearwell. The flow through the clarifier is controlled by a level control valve on the clarifier inlet.

The clarifier is designed to operate on the upflow sludge blanket principle. Chemical addition is in proportion to the inlet flow to the clarifier. Backflushing and sludge blowdown from the clarifier can be manually controlled or automatically controlled in proportion to inlet flow. The sludge is directed to the normal waste for disposal.

The clarified water service pump takes suction from the clearwell, and discharges to the clarified water storage tank via three anthracite bed pressure filters. The clarified water standby pump serves as a backup to the clarified water service pump. The anthracite filters remove suspended solids from the clarifier effluent to maintain proper water clarity within the clarified water storage tank. Each filter has a design flow rate of 100 gpm. Backwashing of the anthracite filters is either manual or automatic, with normal operation being automatic. Under automatic operation, the filters are sequentially backwashed by the filter backwash pump, and are returned to service at the conclusion of the backwash cycle. The clarified water standby pump also serves as a backup to the filter backwash pump. The clarified water service pump, filter backwash pump, and clarified water standby pump are provided with minimum flow recirculation loops for pump protection during low flow demand.

Flow to the clarified water storage tank is controlled by a level control valve on the tank inlet. A controller, normally in manual mode, throttles the inlet valve to the clarified water storage tank level.

The tank is located in the yard next to the water treatment enclosure, and is furnished with steam heating to prevent freezing.

The clarified water storage tank furnishes water for the following systems during normal operation:

a. Demineralized water makeup system

b. Lube water system CHAPTER 09 9.2-16 REV. 20, SEPTEMBER 2020

LGS UFSAR The makeup demineralizer feed-pumps provide water to the makeup demineralizer system as needed to maintain a pre-established level in the demineralized water storage tank. Operation of the pumps is manually initiated by local or remote hand switches.

A valve in the makeup demineralizer feed pump discharge header is automatically closed when neither feed pump is running, thereby preventing inadvertent drainage of the clarified water storage tank during pump shutdown. If the valve fails to open when either feed pump is started, or if the valve fails closed, the makeup demineralizer feed pumps are tripped by a position switch. The pumps are also tripped on low-low level signal from the clarified water storage tank.

With the exception of the clarified water storage tank, which is located outside, the equipment in the clarified water system is located in the water treatment enclosure.

9.2.4.3 Safety Evaluation The clarified water system has no safety-related function. Failure of the system does not compromise any safety-related system or component, nor does it prevent a safe shutdown of the plant.

9.2.4.4 Testing and Inspection Requirements System operability is demonstrated by use during normal plant operation.

9.2.4.5 Instrumentation Requirements The clarified water system is furnished with control panels, located in the water treatment enclosure, which are designed for remote manual and automatic control of the processes. Clarified water flow, and turbidity are monitored to verify system performance, and to alarm when abnormal conditions exist.

The clarified water storage tank is equipped with level switches to actuate high and low level alarms; to trip the clarified water service pump, the filter backwash pump, and the makeup demineralizer pumps .

Automatic operation of any of the lube water pumps due to low pressure in pump discharge line is annunciated in the control room. Low-low pressure in the lube water pump discharge is alarmed locally.

Local pressure indicators are provided at the discharge of all pumps in the systems for pump head indication.

9.2.5 DEMINERALIZED WATER MAKEUP SYSTEM The demineralized water makeup system provides a supply of treated water suitable as makeup for the plant and reactor systems, and for other demineralized water requirements. The demineralized water makeup system has no safety-related function.

9.2.5.1 Design Bases

a. The demineralized water makeup system is designed to provide 80 gpm supply of demineralized water, to meet maximum anticipated plant operating requirements.

CHAPTER 09 9.2-17 REV. 20, SEPTEMBER 2020

LGS UFSAR

b. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the demineralized water makeup system are discussed in Section 3.2.

c. The demineralized water makeup system is designed so that the treated water composition does not exceed the limits given below:

Conductivity, micromhos/cm 0.2 Total dissolved solids, ppm 0.1 Silica, ppm SiO2 0.010 9.2.5.2 System Description The demineralized water makeup system is shown in drawing M-18. Design parameters for the major components of the system are given in Table 9.2-8.

Clarified and filtered river water is supplied to a vendor supplied water purification system by the makeup demineralizer feed pumps (Section 9.2.4). Chlorine residual and organics are removed by the activated carbon filter which precedes the vendor supplied water treatment system. The carbon filter may be bypassed if not required as determined by the water treatment vendor. Inlet water pressure to the carbon filter is regulated to a pre-established value by a pressure-reducing valve.

The original demineralizer system is no longer in service. All demineralized water makeup needs are met by a vendor supplied water purification system.

Discharge from the vendor supplied water purification system is continuously monitored by conductivity measuring devices which initiate alarms on a local panel, and divert the flow to chemical waste when maximum allowable levels are reached. Monitoring of silica shall be accomplished through the chemistry sampling program.

Demineralized water is stored in the 50,000 gallon capacity demineralized water storage tank. The tank is outside, and is furnished with steam heating to prevent freezing. This water supply is used to fill the CSTs and refueling water storage tanks, prior to unit operation. Demineralized water is also used prior to unit operation for plant systems flushing and filling. During normal operation, demineralized water is used for, but not limited to the following services:

a. Condensate and refueling water storage tanks

b. Fuel pool skimmer surge tanks

c. RECW system

d. TECW system

e. Auxiliary boilers

f. SLCS

g. Chemistry laboratory CHAPTER 09 9.2-18 REV. 20, SEPTEMBER 2020

LGS UFSAR

h. Radwaste system

i. Chilled water system

j. Plant demineralized water hose stations

k. Zinc Injection Passivation system (GEZIP)

The demineralized water supply is provided by the demineralized water transfer jockey pump, or the demineralized water transfer pumps which are operated as needed to meet plant demineralized water demand.

9.2.5.3 Safety Evaluation The demineralized water makeup system has no safety-related function. Failure of the system does not compromise any safety-related system or component, nor does it prevent a safe shutdown of the plant.

9.2.5.4 Testing and Inspection Requirements System operability is demonstrated by use during normal plant operation.

9.2.5.5 Instrumentation Requirements Local and remote indicators, alarms, and pressure relief valves are provided to monitor the system process and to protect system components. Pressure differential, level, flow, temperature, and alarms are provided for each applicable point in the demineralizer system. High and low level alarms are installed in the demineralized water storage tank.

9.2.6 ULTIMATE HEAT SINK The UHS, a spray pond, serves the safety-related functions of providing cooling water, and acting as a heat sink for the ESW system and the RHRSW system during accident conditions.

9.2.6.1 Design Bases

a. The UHS is designed to provide sufficient cooling water to the ESW and RHRSW systems, permitting simultaneous safe shutdown and cooldown of both units, and maintaining them in a safe shutdown condition.

b. In the event of an accident in one unit, the UHS is designed to provide sufficient cooling water to the ESW and RHRSW systems to dissipate the heat for that accident safely, to permit the concurrent safe shutdown and cooldown of the other unit, and to maintain both units in a safe shutdown condition.

c. The UHS is sized so that makeup water is not required for at least 30 days, and design basis temperature and chemistry limits for safety-related equipment are not exceeded.

CHAPTER 09 9.2-19 REV. 20, SEPTEMBER 2020

LGS UFSAR

d. The UHS is designed to perform its safety function during periods of adverse meteorological conditions, resulting in maximum water consumption and minimum cooling capability.

e. The UHS is designed to withstand the most severe natural phenomenon or site-related event (e.g., SSE, tornado, hurricane, flood, freezing, or transportation accident), and reasonably probable combinations of less severe phenomena and/or events, without impairing its safety function.

Note: For the case of tornadoes, the cooling towers and Schuylkill River makeup pumps supplement the Spray Pond as required, to mitigate the storm damage and provide required cooling.

f. The UHS is designed to withstand any credible single failure of manmade structural features without impairing its safety function.

g. The UHS is designed to automatically supply sufficient cooling water to the ESW and RHRSW systems, as required.

h. The UHS is designed to include the capability for full operational testing.

i. The UHS is designed so that the cooling towers may be used as the heat sink for normal shutdown operations, and for postaccident operations if they are available.

Codes and standards applicable to the UHS are listed in Table 3.2-1.

9.2.6.2 System Description 9.2.6.2.1 General Description The UHS for both units is a highly reliable, seismic Category I spray pond, that ensures that an adequate source of cooling water is available at all times for reactor shutdown and cooldown, and for accident mitigation. The RHRSW and ESW systems (see Sections 9.2.2 and 9.2.3) receive cooling water via the spray pond pump structure located on the pond perimeter, and return the water to the spray pond via the spray networks, or winter bypass lines .

9.2.6.2.2 Spray Pond Description The spray pond is of seismic Category I design, excavated below-grade, and sized for a water volume adequate for 30 days of cooling under design basis conditions. The general arrangement of the spray pond, spray networks, and spray pond pump structure is shown in Figure 9.2-6.

Details of the spray pond excavation and finished grading are shown in Figures 3.8-55, 3.8-56, and 3.8-57. A summary of pond design data is shown in Table 9.2-9.

The pond is lined to minimize seepage. The pond bottom and soil cut slopes are covered with a soil-bentonite lining, while the rock-cut slopes are lined with shotcrete. See Sections 3.8.4.1 and 2.5 for further details of pond geology, geometry, and linings.

The pond is provided with a seismic Category I overflow weir to accommodate normal water level fluctuations, and an emergency spillway to limit the maximum water level in the pond during maximum precipitation conditions.

CHAPTER 09 9.2-20 REV. 20, SEPTEMBER 2020

LGS UFSAR Four spray networks, which are arranged symmetrically in the pond, provide cooling for the ESW and RHRSW return water. The layout of the spray networks is shown in drawing M-384. The networks and their supply piping are suspended above water on reinforced concrete columns.

9.2.6.2.3 Spray Pond Pump Structure The spray pond pump structure houses the RHRSW and ESW pumps, and associated piping and valves. See drawings M-388, M-390 and M-389 for the general arrangement of equipment within the structure, and drawings M-11 and M-12 for piping and instrumentation diagrams of the ESW and RHRSW systems.

The pump structure is located on the edge of the spray pond. Openings are provided in front of the structure to allow pond water to flow into the wet pits where the pump suctions are located. Closure of the sluice gates on these openings, and realignment of system valves allows the ESW and RHRSW systems to shift from the spray pond mode to the cooling tower mode.

The wet pit area is divided into two sections, corresponding to the A and B loops of the RHRSW and ESW systems. The two areas are separated by a wall, with a sluice gate that can be closed to isolate the two trains. Each pump is installed in its own bay. A removable screen is placed at the entrance of each bay.

Provision of adequate net positive suction head for the pumps is discussed in Sections 9.2.2 and 9.2.3.

HVAC equipment maintains necessary ambient conditions for proper operation of the equipment (Section 9.4-7).

9.2.6.2.4 System Components Piping and valves associated with the spray pond system are carbon steel. Safety-related portions are designed to seismic Category I requirements as defined in Section 3.7 and are installed, inspected, and tested in accordance with the requirements of ASME Section III, Class 3.

Four spray networks are provided: two redundant networks for each ESW/RHRSW combined return loop. The networks and network supply piping are completely above water, and are provided with a corrosion allowance of 250 mils. Network header piping is sized for proper flow rates to all nozzles in the network. Piping is sloped to allow complete drainage of the networks and network supply piping, to minimize corrosion and prevent freezing. The exterior surface of the piping is painted to provide additional corrosion protection.

The spray nozzles are precision cast stainless steel, of a design that provides good thermal performance, while minimizing drift loss. All nozzles are located at el 258', 7 feet above the normal pond water level, allowing for proper droplet formation and flight time. Procedural guidance assures proper flow rates so that the pressure drop across the nozzles necessary for proper spray performance (3.4 psi, minimum) is achieved for all anticipated modes of RHRSW and ESW operation. The nozzles have no internal parts susceptible to clogging. A cleanout connection is provided at the end of each of the twenty-eight spray pond supply header pipes, in order to periodically flush corrosion products from these pipes and thus ensure proper functioning of the spray nozzles.

CHAPTER 09 9.2-21 REV. 20, SEPTEMBER 2020

LGS UFSAR A winter bypass line is provided for each ESW/RHRSW combined return line to allow bypassing the spray networks and returning the heated water directly to the pond volume.

Makeup water to the pond is supplied via a 6 inch branch line from the Schuylkill River makeup line to the cooling towers. The line enters the pond below normal pond water level; a single normally closed gate valve in the line prevent siphoning from the pond in the event of loss of pressure in the makeup line. Makeup to the spray pond is controlled manually. A level detector in the pond which alarms in the Main Control Room assures that proper level is maintained.

An 8 inch blowdown line and associated weir are provided on the western pond perimeter. The line is used for overflowing excess water during rain conditions and for water quality control.

Two drain sumps are provided in the bottom of the pump structure. The sump pumps discharge to the pond above the water surface to avoid siphoning of the pond to the pump structure.

9.2.6.3 System Operation The spray pond is designed to automatically supply cooling water to the ESW and RHRSW systems when required, and to continue this function with a minimum of operator attention.

9.2.6.3.1 Normal Operation The spray pond is normally in a standby mode, and, except for periodic testing or RHRSW operation for spray pond cooling and chemistry control, is used only for cooldown and shutdown operations, or during emergency or accident situations. Starting the ESW pumps causes sluice gates and system valves to automatically align for spray pond operation, if not already in that position. The operator may subsequently stop and start pumps, and remove or add spray networks from service as necessary to maintain proper flows and diesel loadings.

During standby, the pond level is manually maintained above the Tech. Spec. Minimum level of 250-10 (9-10 pond depth) by the makeup and blowdown lines. Design basis minimum operating level corresponds to el. 250-10 (9-10 pond depth). During long-term operation, without makeup and blowdown, the concentration of scale-forming constituents, which can impair heat exchanger performance, increases due to evaporation. Provisions are available for the manual addition of chemicals to inhibit scale formation from calcium carbonate. Sufficient spray pond inventory is provided such that other scale-producing agents, such as calcium sulfate, do not reach concentrations that might cause scaling during the 30 day postaccident period, when no makeup or blowdown is assumed.

9.2.6.3.2 Winter Operation The spray pond is designed to perform its safety functions with an initial ice layer on the pond surface.

During icing conditions, return flow to the pond is initially directed to the winter bypasses, which inject the warm return water directly to the pond volume. The bypasses are directed toward the ends of the pond to allow the return water to circulate and mix with the pond volume, and avoid hydraulic short-circuiting. The increasingly warmer pond water causes any ice layer present on the pond surface to melt. Once a hole is formed in the ice layer, a return path for spray water is available, and the spray networks may be used as water temperature dictates.

9.2.6.3.3 Cooling Tower Operation CHAPTER 09 9.2-22 REV. 20, SEPTEMBER 2020

LGS UFSAR During normal shutdown and cooldown operations, the RHRSW system uses the spray pond or the cooling tower as the source of cooling water, and as the heat sink. The system may be aligned to use the cooling towers as the heat sink and water source by shutting the spray pond sluice gates, and diverting the return flow to the towers. Cooled water from the cooling tower basins flows back to the wet pits.

Should an emergency or accident condition arise while in the cooling tower mode, ESW pump start signals cause the system to automatically realign to the spray pond mode. Subsequently, and after assessment of the cooling tower's availability, the system may be placed in the cooling tower mode, if so desired.

9.2.6.4 Spray Pond Thermal Performance 9.2.6.4.1 Design Meteorology Conservative design meteorological conditions were determined for two cases: maximum water consumption and minimum heat transfer. Since no long-term meteorological data from the LGS site exist, long-term data (40 years) from the NWS at Philadelphia International Airport for the period January 1, 1941 through December 31, 1980, were selected as the most representative available long-term record.

To demonstrate the applicability of Philadelphia data, a coefficient of thermal performance comparison was made between LGS and Philadelphia for the same 5 year period (January 1, 1972 through December 31, 1976) using 3 hourly averages of wind speed, dry-bulb temperature, and relative humidity from each site. Wind speed was taken from the LGS aerovane 30 feet above grade on Tower No. 1. The dry-bulb temperature from LGS Weather Station No. 1 (about 5 feet above grade) was used. Relative humidity was determined through an average of eight stations at and surrounding the LGS site. Location of towers and stations is discussed in Section 2.3.2.

A comparison of the frequency distributions of dry-bulb temperatures by 10% relative humidity groups shows that the Philadelphia data are similar to the LGS data for the 21/2 year period from January 1, 1972 through June 30, 1974. At the higher humidities, where the minimum heat transfer case occurs, the distributions are essentially identical. At lower humidities, the Philadelphia dry-bulb temperatures are slightly higher. Representative comparisons are shown in Figures 9.2-8 and 9.2-9.

A comparison of the frequency distribution of daily average relative humidities, without regard to temperature, is listed in Table 9.2-10. This table also shows very similar frequency distribution over the 21/2 year period, with Philadelphia exhibiting a slightly higher frequency of low humidities, and a slightly lower frequency of high humidities.

A comparison of the distribution of daily average wind speeds from Philadelphia Airport data with that from the LGS site, also listed in Table 9.2-10, show rather poor agreement. The airport exhibits considerably higher daily average wind speeds, especially during daytime hours since 1964. This is believed to be due to the location of the wind sensor near the runways, which results in the measurement of manmade wind from the aircraft. Since Philadelphia wind speeds are not considered representative of the LGS site, a natural wind speed of zero was assumed in the analysis of meteorological data for the spray pond design, in the minimum heat transfer case. This is a conservative assumption.

CHAPTER 09 9.2-23 REV. 20, SEPTEMBER 2020

LGS UFSAR Since these comparisons of meteorological data from LGS with similar data from the Philadelphia Airport indicate that the Philadelphia Airport relative humidity and temperature are quite representative of the LGS site, and because the Philadelphia record is of sufficiently long duration to establish worst case conditions, the 40 years of Philadelphia data for humidity and temperature were used as the basis for the design of the spray pond.

The design meteorology was selected using the method of Reference 9.2-1, as discussed below.

For maximum water consumption, the worst 30 day period was determined using a coefficient of water loss which combines losses due to evaporation and drift. The evaporation loss was calculated from the results of the work of Ranz and Marshall (Reference 9.2-2) on evaporation from water drops in air. The drift loss was calculated as a function of wind speeds, using a computer model (Section 9.2.6.4.1.1) whose predictions are based on field test results (Reference 9.2-3) of identical nozzles. A coefficient of water loss value (Section 9.2.6.4.1.1), which is proportional to water loss rate, was calculated for each day in the 40 year period of record. Running averages of the coefficient values were used to select the worst 30 day period for water loss. The worst 30 days for water loss began May 30, 1958. The average conditions for that period are presented in Table 9.2-11. Since wind speeds at Philadelphia are generally higher than at the site, the use of Philadelphia data yields conservative results for maximum water consumption.

The minimum heat transfer meteorological conditions were determined using a similar technique.

The Philadelphia meteorological data were modified for the minimum heat transfer case, however, by assigning a zero natural wind speed value to all data. This is conservative, in that it minimizes heat transfer. The average drop velocity was added to the natural wind speed to calculate evaporation. A coefficient of thermal performance (Section 9.2.6.4.1.1) was calculated for each day, using average daily data. Running averages were taken to determine the worst 30 days for heat transfer. The worst day was June 30, 1945, and the worst 30 day period began August 12, 1959. The resulting design meteorological conditions are given in Table 9.2-11.

Using the coefficient of performance technique described above, the worst meteorological conditions were also determined for the 5 year period from January 1, 1972 through December 31, 1976, with both Philadelphia and LGS site 3 hourly average data. Actual reported natural wind speeds were used in this comparison for the maximum water loss case; Philadelphia wind speeds were converted to a 30 foot elevation for comparison. Natural wind speed added to the drop velocity was used in selecting worst case conditions for the minimum heat transfer case.

Results, listed in Table 9.2-12, show that the worst conditions are very similar at both locations, with the Philadelphia conditions being more conservative in all cases, except the 30 day minimum heat transfer case. The use of zero natural wind speed in the actual thermal analysis of the spray pond makes the use of the Philadelphia data conservative in the minimum heat transfer case.

This comparison confirms that use of the Philadelphia meteorology data, as modified, is appropriate as a conservative basis for the design of the spray pond.

9.2.6.4.1.1 Meteorology Analysis Methods Drift Loss Model An independent model has been developed to predict drift losses. A review of the literature reveals no efforts directly applicable to calculation of drift losses from a spray pond. Due to basic system CHAPTER 09 9.2-24 REV. 20, SEPTEMBER 2020

LGS UFSAR differences, cooling tower drift measurements cannot be applied directly to spray ponds; therefore, a model was developed from principles of analytical mechanics. The following parameters are included in the model:

a. Drop size spectrum

b. Wind speed and direction

c. Elevation necessary for loss of a drop from the pond

d. Distance of each nozzle from the perimeter of the pond in the direction of drift

e. Pressure drop across the nozzle

f. Angle at which water leaves the nozzle

g. Vertical air entrainment of droplets Drift is caused by the horizontal drag force exerted on small drops as they move relative to the air.

A water drop leaves the nozzle with a certain initial velocity, and from that time its motion is determined by drag and gravitational forces. By solving the equations of motion, the position of each drop is determined as a function of time. When all initial velocities are considered, the positions of drops of the same size that left the nozzle at the same time, trace out a locus in the horizontal plane. When drops of similar size are grouped together, a locus results for each drop size group. A schematic drawing of typical loci is shown in Figure 9.2-10, for one wind speed at one elevation. The loci are concentric circles for a wind velocity of zero, and are somewhat distorted and translated in the wind direction for nonzero wind speeds.

Once the loci have been determined, for a given wind speed, the fraction of flow lost by drift for each drop size group is the ratio of the length of the locus outside the pond perimeter to the total locus length. Since a locus represents the position of drops of a given group, no drops from that group are off the locus at the elevation; consequently, the length of the locus is used to calculate loss fraction rather than the enclosed area. The percentage of flow lost by drift is the sum over the drop size groups of the product of drift loss fraction and flow fraction.

n P= fi Bi (EQ. 9.2-1) i=1 where:

P = percentage of flow lost by drift fi = fraction of flow in drop size group (i) that is lost Bi = fraction of total flow in drop size group (i) n = number for drop size groups CHAPTER 09 9.2-25 REV. 20, SEPTEMBER 2020

LGS UFSAR In order to facilitate evaluation of the drag coefficient for each drop size group, the drops are assumed to be spherical. High speed photographs show that drops deviate very little from being spherical, especially for smaller diameters, which are most important in drift loss considerations.

Since the drag force on a sphere is proportional to the relative velocity raised to a power between 1 and 2 (depending on the Reynold's number), the resulting equations of motion are nonlinear. An approximation is made to allow a solution in closed-form, in which the drag force is assumed to vary linearly over a certain range of velocities. Two velocity ranges are used, and for all velocities the approximation equals or exceeds the actual drag force, thus preserving conservatism by maximizing drift losses. This approximation technique is depicted in Figure 9.2-11.

The linear drag force approximation, in combination with Newton's Second Law, is used to determine the acceleration of a drop. The acceleration is integrated to determine the position of the drop as a function of time. This is done for both the X and Z directions, shown in Figure 9.2-12, to determine the coordinates, Xi (t,)' Zi (t,)', of drop position for the ith drop size group, as a function of time and initial direction. The motion in the Y direction is used only for calculating the drop exposure time.

In order to find the locus of a given drop size group at the elevation necessary for loss from the pond, the time of flight or exposure time, must be calculated. The motion of a drop in the Y direction, shown in Figure 9.2-12, is used to calculate the exposure time. Since the water leaves the nozzle in a conical pattern, no drag is applied for the first few feet of travel in the vertical direction to allow maximizing the time in the air, which maximizes drift losses. Drag is applied immediately in the calculation of X and Z coordinates, in order to maximize drift losses. The vertical position, Y(t), is determined as a function of time; subsequently, the elevation necessary for loss from the pond is substituted for the position. The resulting implicit equation is solved for exposure time. There is a different exposure time for each drop size group, due to the dependence of drag force on drop diameter.

With the exposure time determined, the locus for each drop size group can be generated by considering all initial velocity directions. A computer program has been written to supply the coordinates of points on each locus. The locus in the X, Z plane for each drop size group is integrated numerically over its length to determine the fraction of the locus, hence the drop size group, that is beyond the perimeter of the pond. The losses for the different drop size groups are summed to determine the total drift loss percentage from the pond.

The computer model generates a drift loss versus perimeter distance curve (Figure 9.2-13), which is used to generate a drift loss vs wind speed curve that is unique for each pond. A typical example is shown in Figure 9.2-14. Drift loss vs wind speed is used as input to the coefficient of water loss, discussed later, and in the system model.

The percentage of flow lost due to drift is an input parameter for the system model. The system model uses it as a loss term in determining the water remaining in the spray pond at any time after the start of operation of the sprays. The drift loss for the LGS spray pond is determined as a function of wind speed. This information is entered as a table, wind speed versus drift loss, in the system program. The drift loss is determined from the table at each time step in the calculation of system parameters.

Coefficient of Thermal Performance CHAPTER 09 9.2-26 REV. 20, SEPTEMBER 2020

LGS UFSAR The coefficient of performance for heat loss was calculated from the following equation:

C= (1/MdCp) [ -KDp (2 + 0.6 Nre1/2Npr1/3) (Td-Ta) - (DvDphg)

• ) - (P /R T * ))]

(2 + 0.6 Nre1/2Nsc1/3) ((Pai/RvT d ao v a (EQ. 9.2-2) where:

C = coefficient of performance, F/sec Md = drop mass, lb Cp = specific heat of water, Btu/lb-F (assumed = 1)

K = thermal conductivity of moist air, Btu/sec-ft-F Dp = drop diameter, ft (0.008 ft)

Nre = Reynold's number, moist air (dimensionless)

Npr = Prandtl number (dimensionless)

Ta = air dry-bulb temperature, F Td = drop temperature, F (pond temperature)

Dv = diffusion coefficient of water vapor in air, ft2/sec hg = enthalpy or water vapor, Btu/lb Nsc = Schmidt number (dimensionless)

Pai = vapor pressure at Ta, lb/ft2 Pao = partial pressure of vapor in air, lb/ft2 Td* = Td + 460, R Ta* = Td + 460, R w = density of water, lbm/ft3 (62.4)

Pao = (RH) (C+C2Ta+C3Ta2+C4Ta3+C5Ta4+C6Ta5) 70.5 lb/ft2 Pai = (C1+C2Td+C3Td2+C4Td3+C5Td4+C6Td5) 70.5 lb/ft2 CHAPTER 09 9.2-27 REV. 20, SEPTEMBER 2020

LGS UFSAR C1 = 3.76515x10-2 inch Hg C2 = 2.5463x10-3 inch Hg/F C3 = 2.96588x10-5 inch Hg/F2 C4 = 8.40577x10-7 inch Hg/F3 C5 = 7.-1246x10-10 inch Hg/F4 C6 = 4.33688x10-11 inch Hg/F5 RH = relative humidity Nsc = /Dv hg = 1090 Btu/lbm Dv = 4.06 x 10-8 (T a )2.5/(Ta + 901)ft2/sec Npr = Cp/k Nre = VDp/

V = Wind velocity ft/sec Dp = Drop diameter, 0.008 ft v = Kinematic viscosity of air

= 1.45x10-4 + (Ta - 32) (5.15x10-7) ft2/sec K = 3.89x10-6 + (3.89x10-7 (Ta - 32))/68 Btu/sec-ft-°F Md = wV = wd3/6 lbm w = 62.4 lb/ft3 Rv = 85.5 (gas constant, vapor)

Coefficient of Water Loss The water loss coefficient of performance is calculated from the following equation:

Cw = 1.25x104 DvDp (2 + 0.6 Nre1/2Nsc1/3) (Pai/RvT d*)

CHAPTER 09 9.2-28 REV. 20, SEPTEMBER 2020

LGS UFSAR

• ) + fd (Pao/RvT a (EQ. 9.2-3) where:

fd = drift loss fraction (from drift loss curve for corresponding wind speed typical curve shown in Figure 9.2-14)

Cw = water loss coefficient The remainder of terms are as defined in Equation 9.2-2.

The equations used to represent the typical curve of Figure 9.2-14 are:

a. For wind speed less than or equal to 2.5 miles per hour fd = C x 10-4 WS (EQ. 9.2-4) where:

WS = wind speed, mph C = a constant for the specific pond geometry

b. For wind speed greater than 2.5 miles per hour fd = Ae(B(WS)) (EQ. 9.2-5) where:

A and B = constants for the specific pond geometry 9.2.6.4.2 Spray Pond Water Requirements This section discusses the water requirements that were considered in determination of the spray pond design volume and used in the pond thermal performance analysis.

9.2.6.4.2.1 Evaporation Due to Plant Heat Load Evaporation due to the plant heat load is determined from the total heat dissipated by the spray pond in 30 days, due to decay heat, sensible heat, and auxiliary system heat loads. The total decay heat dissipated in 30 days is 5.47x1010 Btu for both units, based on Reference 9.2-4 for finite reactor operation. This includes contributions from fission products and heavy elements.

The LOCA/SSD auxiliary system total heat generation for the first 30 days is 2.27x1010 Btu. The auxiliary system heat load includes loads due to operation of engineered safeguards equipment, including the standby diesel generators, which would be in service following the loss of offsite power. The LOCA/SSD auxiliary heat generation rate averages 31.2x106 Btu/hr for the first 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, and 31.5x106 Btu/hr thereafter, total for both units. The two-unit SSD auxiliary heat generation rate averages 64.2x106 Btu/hr for the first 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and 65.2x106 Btu/hr thereafter, total CHAPTER 09 9.2-29 REV. 20, SEPTEMBER 2020

LGS UFSAR for both units. The total sensible heat load of the reactor pressure vessel, its internals, and the suppression pool is 1.93x108 Btu, referenced to a final temperature of 121.4oF. The resultant evaporation is 13.58x106 gallons in 30 days.

9.2.6.4.2.2 Natural Evaporation Natural evaporation of the spray pond was based on the design meteorology of Section 9.2.6.4.1 and a constant pond surface area of 9.6 acres. The water loss due to natural evaporation is shown in Table 9.2-13. This is conservative because the average surface area over the 30 day accident scenario is less than 9.6 acres and the effect of the sprays suppressing evaporation is not taken into account.

The evaporation calculation was based on the methodology of Reference 9.2-5.

9.2.6.4.2.3 Drift Loss Drift loss due to entrainment of spray droplets during periods of moderate and high winds has been estimated using data from spray pond tests (Reference 9.2-3) and a drift loss computer model (Section 9.2.6.4.1.1). The 30 day period of highest water loss was determined by considering both evaporative and drift losses, as described in Section 9.2.6.4.2. For the 40 year period of meteorological record (January 1, 1941 to December 31, 1980), the worst 30 day period for water loss was found to be May 30 through June 28, 1958. The average wind speed for this 30 day period is 10.43 mph. Drift loss is minimized by placement of over 90% of the nozzles a minimum of 100 feet away from the perimeter of the pond. The remaining nozzles, located near the corners of the pond, are at least 60 feet away from the pond perimeter.

The operation of four RHRSW pumps and three ESW pumps after a two-unit SSD, results in the maximum loss of 1.44x105 gallons in 30 days, based on a spray flow rate of 9000 gpm per RHRSW pump and 5666 gpm per ESW pump.

This drift loss value is based on a pond surface area of 9.9 acres and nozzle height of 7.17 feet above the pond surface.

9.2.6.4.2.4 Seepage The design seepage rate of 1.825x106 gallons in a 30 day period was used in the thermal analysis.

A seepage test was conducted to verify that the allowable design seepage rate for the lined pond would not be exceeded. This test, which was conducted during the period December 13, 1982 to May 21, 1983, determined that the actual seepage loss was between 2.1x105 gallons and 3.8x105 gallons in 30 days, depending on how evaporation losses are measured. This rate is between 11%

and 21% of the allowable design rate. Considering possible error contributions from all measurements, the upper limit of the seepage loss during the test is estimated to be approximately 40% of the allowable design rate, and the lower limit is effectively zero. Details of the spray pond seepage measurements and testing are discussed in Section 2.5.

9.2.6.4.2.5 Sedimentation An allowance of 2.75x106 gallons of water is provided to ensure that frequent pond cleaning is unnecessary. This corresponds to a sediment layer of approximately 12 inches on the pond bottom. Accumulation of sediment on the spray pond pumphouse intake area concrete slab will be periodically checked as part of the inspection program required by Regulatory Guide 1.127. Based CHAPTER 09 9.2-30 REV. 20, SEPTEMBER 2020

LGS UFSAR on the results of these inspections, a suitable program of reinspection and cleaning will be established. Any excessive accumulation of sediment will be removed by suction dredging equipment to prevent excessive particulate concentrations from occurring near the intake structure.

Water removed from the pond with the sediment will be returned to the pond such that intake water quality is not adversely affected or limited to the capacity of the clean water makeup line.

9.2.6.4.2.6 Fuel Pool Cooling and Makeup In the event of a LOOP, the spray pond is used as a source of cooling water for the spent fuel pool via the RECW system (Section 9.2.8). The heat dissipation for this service is included as an auxiliary system heat load in Section 9.2.6.4.2.1.

Provisions are also made to supply pond water directly to the fuel pools via the ESW or RHRSW system, in the event of unavailability of the fuel pool cooling system. Since cooling via the RECW system and direct makeup would not occur simultaneously, the makeup function does not represent an additional water loss.

9.2.6.4.2.7 Water Quality The spray pond is treated with a mineral scale inhibiting chemical. The addition of chemical treatment allows the background chemical concentrations of the spray pond to increase due to normal water losses from drift and evaporation during non-accident operation. During an accident, the spray pond will normally increase in concentration, assuming no makeup. The mineral scale inhibiting chemical concentration will also increase in the same manner, preventing deposits on heat transfer surfaces. The spray pond will be analyzed periodically and adjustments made to scale inhibitor concentrations, as needed.

Provisions are also made to reduce the fouling effects associated with silting and biologics.

Chemical treatment also includes the use of biocides and dispersants to maintain water quality.

9.2.6.4.2.8 Minimum Water Level for Operation To verify adequate hydraulic conditions in the pump structure wet pits for proper long-term pump operation, it has been determined from model studies that acceptable hydraulic conditions exist down to el. 243'-6". This operational limit establishes the dead storage volume of the spray pond as 6.95 million gallons. This dead storage is utilized to fulfill the sedimentation volume requirements (Section 9.2.6.4.2.5) and used in the analysis of spray pond temperatures.

As discussed in the following section, the pond water level will not drop below el. 243'-6" at the end of the design basis 30 day period.

9.2.6.4.2.9 Total Volume Allocation The design water allowances are shown in Table 9.2-13. Substantial margin (4.74x106 gallons) is shown to exist for the maximum water loss case 9.2.6.4.3 Evaluation of UHS Performance The spray pond performance has been analyzed to assure that the design spray pond volume is adequate for 30 days of cooling without makeup or blowdown, and that the cooling water temperature does not exceed the design limit for the design basis heat input and meteorological conditions. The heat loads for the following conditions were examined:

a. One-unit LOCA, one-unit SSD CHAPTER 09 9.2-31 REV. 20, SEPTEMBER 2020

LGS UFSAR

b. Two-unit SSD The heat loads for the one-unit LOCA, one-unit SSD were modeled as a two-unit SSD with cooldown at the maximum rate of 100F per hour and four RHR heat exchangers operable for the first 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of the event. This model maximizes the heat rejection to the pond during the time when the peak pond temperature is reached. This results in the bounding pond temperature and maximum water loss. Therefore the two-unit SSD model with maximum heat rejection was used for the pond thermal and water loss analysis.

Other assumptions used in performing the transient analysis for performance evaluation include:

a. No makeup water is supplied to the spray pond during the 30 day period.

b. The decay heat generation rate is in accordance with NUREG 0800 Branch Technical Position ASB 9-2. (Reference 9.4-4)

c. Coincident with the postulated LOCA/SSD, the water level in the spray pond is at design minimum (corresponding to 29.07x106 gallons less the sedimentation allowance) .

d. Uniform mixing is assumed in the spray pond. Short circuiting was not observed in tests (Reference 9.2-3).

The spray efficiency values are determined using the spray cooling model described below.

Comparisons of model predictions with test data show the model to be conservative (Section 9.2.6.4.3.4). The spray efficiency used in these transient analyses is presented in Figure 9.2-17 as a function of hot water temperature. The spray efficiency is determined for the minimum spray network flow rate of 8500 gpm. This flow corresponds to the minimun flow of two RHRSW pumps and one ESW pump divided between two spray networks. The minimum flow was combined with the heat loads defined above to create a model which envelopes all spray pond operating conditions.

Other pertinent data used in the analysis is presented in Table 9.2-14. The heat rejection rate to the spray pond is given in Figure 9.2-18 and Table 9.2-15 for decay heat, station auxiliary heat and sensible heat. The integrated decay heat, station auxiliary heat sensible heat and total heat are given in Figure 9.2-19 and Table 9.2-16. The initial spray pond temperature assumed, 88oF, corresponds to extreme monthly average atmospheric conditions that maximize initial temperature (Reference 9.2-8).

9.2.6.4.3.1 Analytical Techniques and Input Parameters To perform the required transient analyses that demonstrate adequate performance of the ultimate heat sink, as measured against criteria specified by the U. S. Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.27,9 a spray pond model utilizing a new ultimate-heat-sink performance computer program has been developed. This computer program consists of various theoretical models that simulate the thermal performance of the spray pond subject to an imposed heat load and meteorological conditions, both of which vary with time. These meteorological conditions include the environmental dry-bulb temperature, relative humidity, and prevailing wind speed, if appropriate. In addition, the effects of solar radiation directly to the pond, infrared radiation from the atmosphere to the pond, and radiation from the pond to the environment are included along with heat transfer via convection and conduction between the spray pond and environment.

CHAPTER 09 9.2-32 REV. 20, SEPTEMBER 2020

LGS UFSAR These models comprise the several modules (subprograms) that are described below. The computer program is written in the FORTRAN77 programming language, also known as FORTRAN X3.8 (1978), using the Microsoft FORTRAN v5.1 compiler specifically for a microcomputer equipped with a 80386 central processing unit (CPU) and 80387 co-processor or 80486 CPU.

9.2.6.4.3.1.1 Spray-Pond Model The spray-pond model (module) is used by the Limerick Spray Pond model to calculate the mass and temperature of the spray pond from conservation principles. Both the water inventory (pond mass) and temperature are expected to also change with time. The spray pond is assumed to be thermodynamically described by a single bulk temperature. Heat transfer and water loss from the pond surface can be considered a secondary effect with regard to the heat and mass transfer action of the spray network.10 Thermal stratification is diminished because the pond volumes are relatively small compared to the amount of water being circulated.11 The rate of change of (internal) energy for the spray pond is determined from a simple energy balance dEp(t)

= Hdp + Hsn + Han Hbr Hev Hcn Hps (EQ. 9.2-6) dt where:

Hdp - is the energy added via sprayed water, in the form of drops, falling on the spray pond surface; Hsn - is the solar insolation (direct solar radiation to the pond);

Han - is the radiation from the atmosphere to the pond; Hbr - is the infrared radiation from the pond to the atmosphere; Hev - is the energy loss due to evaporation from the pond surface; Hcn - is the conductive energy loss from the pond to the atmosphere; and Hps - is the energy lost from the pond due to seepage through the basin liner.

Heat lost via conduction through the pond basin liner is neglected. In writing the above expression, it is explicitly assumed that the water in the spray pond can be represented by a single temperature that is independent of location.

In an analogous fashion, the rate of change of pond mass is just dMp(t)

= wdp Wpe Wsw Wps (EQ. 9.2-7) dt where:

wdp - is the mass flow rate to the pond, in the form of drops falling to the pond surface, from the spray; wpe - is the evaporation rate from the pond surface; wsw - is return flow to the service water systems; and CHAPTER 09 9.2-33 REV. 20, SEPTEMBER 2020

LGS UFSAR wps - is the seepage losses through the pond basin liner.

Generally, apart from the direct solar radiation to the spray pond, the mass and energy flow rates are functions of time by way of the pond temperature which varies with time.

The total mass and energy of the spray pond can be obtained via numerical integration of a first-order Taylor's series expansion:

dM(t)

M(t + t) M(t) + t dt (EQ. 9.2-8) dE(t)

( + t) Et Et ()+ t dt The temperature of the spray pond is determined from knowledge of the pond total internal energy Ep and mass Mp. Although the specific energy of water is a function of both temperature and pressure, it is only weakly dependent upon the pressure, and to an excellent approximation (about one part in ten thousand or 0.01%) values along the saturation line may be used. For temperatures below 373.15 K (100oC), the internal specific energy ef is given by the product of the specific heat at constant volume for water, that is independent of temperature, and the temperature in Celsius. Hence, the spray pond specific energy is given by Ep e(p, Tp) ef(Tp) cvw(Tp To) (EQ. 9.2-9)

Mp where cvw = 4189.8 +/- 5.1 J/kg-oC and To = 273.15 K. Approximate determination of this spray pond temperature in this manner will produce an error in the temperature of about 0.01oC. In addition, note that the internal specific energy has been, in compliance with custom, referenced to 0oC, that is, at Tp = 273.15oK, the specific energy is zero.

9.2.6.4.3.1.1.1 Heat Load As a matter of definition, the heat load rejected to the ultimate-heat sink QHx is that energy added to the water from the spray-pond that is used as cooling for the sundry heat exchangers. The energy at the spray nozzles is the sum of the enthalpy of the cooling flow plus this heat load. Part of this energy is transferred to the atmosphere via evaporation or is lost by action of a prevailing wind carrying the sprayed water in the form of drops beyond the pond boundary. The energy added to the spray pond is equal to the heat load QHx, plus the product of the spray mass flow rate and the pond specific enthalpy wswhf(Tp), minus the energy lost via evaporation from the sprays and driftloss, and is described below in Section 9.2.6.4.3.1.1.5. Figure 9.2-18 shows the spray-pond heat load as a function of time for a postulated design basis accident.

9.2.6.4.3.1.1.2 Solar Insolation The average daily solar radiation absorbed by a body of water varies according to site latitude, time of year, and time of day. Various tabulations, graphs, and formulas are available which give the instantaneous solar radiation as a function of these parameters with an accuracy of about 5 CHAPTER 09 9.2-34 REV. 20, SEPTEMBER 2020

LGS UFSAR percent. Because the average spray pond temperature is required for a period of one month, values for the instantaneous solar radiation will require time increments on the order of one hour.

Such a formulation is cumbersome and requires long microcomputer running times and a formulation for average daily solar radiation is more appropriate. Thus, the average daily solar insolation (radiation that has been received) can be expressed in the form12 Han = (1 0.0071C 2)HoSp (EQ. 9.2-10) where Ho is the absorbed solar radiation for a clear sky in units of watts per meter squared, C represents the cloud in tenths of sky, and Sp is the pond surface area in units of square meters.

Because of the rotation of the earth, Ho changes significantly throughout the daylight hours and, of course, is zero during the night. To facilitate computation, Ho is taken to be an average over a given time increment. However, if the increment is longer than about three or four hours, the accuracy of the absorbed solar radiation is unacceptable. Because the characteristic time required for significant changes in the ultimate-heat-sink system is on the order of one day, the absorbed solar radiation Ho is taken to be the average daily absorbed solar radiation in units of watts per square meter of pond surface for a cloudless sky, and can be parameterized as a function of latitude and time (day) of year:13 2N Ho = Asn() + Bsn()sin + Csn() (EQ. 9.2-11) 366 where the parameters Asn(), Bsn(), and Csn() are functions of latitude between 26 N and 46 N. The standard error associated with equation [EQ 9.2-11] is less than 8.7 W/m2.

9.2.6.4.3.1.1.3 Atmospheric Radiation Infrared radiation from the atmosphere is a function of many variables, including the distribution of temperature, moisture, carbon dioxide, ozone and other atmospheric constituents. However, since not all these parameters are normally known, Anderson14 proposed an empirical relationship of the form:15 Hra = Sp Ara + BraPv (Ta) (1 ) T4a (EQ. 9.2-12) where = 56.687 pW/m2-K4 is the Stefan-Boltzmann constant, pv the saturation pressure in units of Pa, is the pond surface reflectivity, and Ta the atmospheric temperature in units of K. The parameters Ara and Bra are empirical parameters and functions of the cloud cover parameter C. As before, Sp is the surface area of the spray pond in units of meters squared.

9.2.6.4.3.1.1.4 Spray-Pond Radiation The energy flux radiated from the spray pond in units of watts per square meter of pond surface area is given simply by Hbr Sp T4p Hbr = Sp T4p

= (EQ.9.2-13)

CHAPTER 09 9.2-35 REV. 20, SEPTEMBER 2020

LGS UFSAR where is the emissivity of the pond taken to be 0.97,16 is the Stefan-Boltzmann constant, Sp is the spray pond surface area in units of meters squared, and Tp is the temperature of the pond surface in units of C. Because the flow rate out from the pond to the sundry heat exchangers is in excess of 87% of the pond volume per day, it can be assumed that the pond is well mixed and the pond surface temperature is well-represented by the bulk average temperature Tp.

9.2.6.4.3.1.1.5 Spray-Pond Evaporation Evaporation from the spray pond due to the plant heat load and heat input by the environment (natural evaporation) is calculated as part of the pond thermal performance module in the ultimate-heat-sink computer program. Evaporation from the surface of a body of water is a complicated phenomenon involving many factors and much discussed in the literature. The evaporative energy flux in units of watts per square meter of pond surface is proportional to the difference in steam (water vapor) concentration between the pond surface and the atmosphere. Hence, an empirical relation giving the energy loss rate in units of watts can be written in the form:

Hpe=C(U)Pv(Tp)-aPv(Ta)Sp (EQ. 9.2-14) where C is a constant to convert from units of Btu/(day ft2) to W/m2, (U) is the "wind function" and Sp is the surface area of the spray pond. The term pv(Tp) is the saturation pressure evaluated at the pond temperature and thus is a measure of the steam concentration at the spray pond surface, while the term a pv(Ta) is the partial pressure of steam (water vapor) in the atmosphere and so represents the concentration there. Hence, the difference between these terms is the concentration difference which drives the net evaporation.

The "wind function" (U) describes the effect of a prevailing wind as it blows across the spray pond surface. One of the most accurate, but slightly non-conservative, empirical relations for the wind function is that given by Ryan and Harleman:17 (U)=22.4(Tv)1/3+14U (EQ. 9.2-15) where Tv is the "virtual" temperature difference between the spray pond surface water and the air above the pond and U is the prevailing wind speed in miles per hour. The virtual temperature has units of F and is given by 5 patm Tp patm Ta Tv = - (EQ. 9.2-16) 9 patm - 0.378 pv (Tp) patm - 0.378 pv (Ta) where Tp and Ta are respectively the spray pond and atmosphere temperatures in K, pv(Tp) and pv(Ta) are respectively the vapor pressures at the pond surface and atmosphere in units of Pa, and patm is the total atmospheric pressure also in units of Pa.

Other wind speed functions commonly used can be selected by the user. The wind speed function proposed by Brady et al18 which is derived from large-lake data is given by CHAPTER 09 9.2-36 REV. 20, SEPTEMBER 2020

LGS UFSAR (U)=70 + 0.7U 2 (EQ. 9.2-17) which has units Btu/(ft2 day)/mm Hg. The prevailing wind speed U has units of miles per hour and is measured at ten feet above the surface of the pond. This model of the wind speed is generally more conservative than that proposed by Ryan and Harleman.

Comparison of predictions using each model with experimental data shows that the Brady formulation gives pond temperatures that are 5 - 11 C higher than the Ryan and Harleman formulation.19 The wind speed function used by Thackston and Parker20 is (U)=C2U (EQ. 9.2-18) where C2 is a conversion constant to convert from Btu/(hr ft2)/in Hg to Btu/(day ft2)/mm Hg and is an empirical constant which depends on the size, shape and exposure of the water body, and on the location of the wind measurement. In the referenced work of Thackston and Parker, a value of 13.9 is used which was approximately the value obtained from their Lake Hefner studies.

9.2.6.4.3.1.1.6 Spray-Pond Conduction and Convection The conduction and convection heat losses from the ultimate-heat-sink spray pond is driven by the temperature difference between the spray-pond and the atmosphere, and is dependent upon the prevailing wind. An empirical relation for this rate of energy loss is21 9

Hcn = CB(U)(Tp - Ta)Sp (EQ. 9.2-19) 5 where (U) is the wind-speed function selected for the evaporation losses in Section 9.2.6.4.3.1.1.5, and C is a coefficient that converts from Btu/(day ft2) to W/m2. The parameter B =

0.26 mm Hg/F is Bowen's Coefficient, and the temperatures have units K.

9.2.6.4.3.1.2 Spray-Region Model A spray field or region above the surface of the spray pond is created by operation of the spray network. Each nozzle creates a spray which forms a cone of water with an average angle to the horizontal. In calm wind conditions, the sprayed water forms an "umbrella" which reaches some maximum height H above the pond surface and range R from the nozzle. All nozzles in concert generate a spray field or spray region above the surface of the spray pond in which the density of sprayed drops is great. As they pass through this region, the drops lose mass and energy via evaporation before entering the spray pond. As a result of this evaporative mass and energy loss, the temperature and relative humidity of the surrounding air is altered. Hence, consideration must be given to the modification of the environment in this region in order to adequately simulate the performance of the spray network.

CHAPTER 09 9.2-37 REV. 20, SEPTEMBER 2020

LGS UFSAR Because the temperature and relative humidity vary appreciably throughout the spray field, this region is divided into segments. Two different methods for division are necessary depending upon whether or not there is a prevailing wind. As the sprayed water is hot, relative to the environment, it will heat the surrounding air and thereby reduce its density compared to the air outside the spray region. This difference in density will generate or induce a wind. Depending upon the temperature of the sprayed water and meteorological conditions, this buoyancy-induced wind can have a wind speed greater than that of a prevailing wind of low speed. However, as the prevailing wind speed increases, the effects of buoyancy become unimportant. Thus, to capture the physics of the spray region thermal performance when buoyancy is important, it is necessary to simulate the spray-region performance differently depending upon whether or not there exists a prevailing wind.

Hence, both a low-wind-speed model and a high-wind-speed model are required. Which of these models dominate is determined by which model has the greater spray-region efficiency which is described in more detail below in Section 9.2.6.4.3.1.2.4. Since this is not obvious, a priori, the Limerick Spray Pond model calculates the spray-region performance for each model separately and automatically selects those results (inputs to the spray pond) from that model with the greatest calculated efficiency.

9.2.6.4.3.1.2.1 Spray-Region Geometry The spray nozzles are assumed to be arranged in a rectangular pattern with the long dimension parallel to the x-axis and the short dimension parallel to the y-axis. Hence, the spray region will also have a rectangular geometrical pattern. The height of the spray region H is just the height reached by the spray above the surface of the pond. The length of the spray region is just the length of the nozzle geometrical pattern (in the x-direction) plus twice the range of drops from a nozzle; similarly, the width of the spray region is the geometrical width (in the y-direction) plus twice this range. This geometry is subdivided differently depending upon whether a low-wind-speed model or a high-wind-speed model is being used. However, for both models, the height and overall extent of the spray region are the same.

A ballistic drop trajectory that neglects drag forces is used to determine the height H and range R for a given initial (nozzle) drop velocity uo. The nozzle spray angle , effective nozzle flow area Anz, and number of nozzles Nnz are assumed known. From these, the initial drop velocity at the nozzle is given by v f w nz uo = (EQ. 9.2-20)

A nz N nz where vf is the liquid specific volume for water evaluated at the spray temperature at the nozzle.

Once the initial drop velocity, which is a vector quantity, has been determined, vertical and horizontal components are easily obtained. Hence, the height is just uo2 H sin2 + y o - y p (EQ. 9.2-21) 2g where yo is the height of the spray nozzle above the water surface and yp < 0 is the elevation of the pond surface with respect to the initial surface elevation (yp,init = 0). The horizontal range of a drop is the product of the horizontal velocity and the sum of the time required for the drop to reach the apex of its trajectory and the time required to free fall to the pond surface:

CHAPTER 09 9.2-38 REV. 20, SEPTEMBER 2020

LGS UFSAR uo2 2g(y o - y p)

R 1 + 1 + sin(2) (EQ. 9.2-22) 2g uo2 sin2 9.2.6.4.3.1.2.2 High-Wind-Speed Model For the high wind speed submodel, the spray region is divided into N rectangular segments normally oriented such that the wind velocity is parallel to the long axis of the region as depicted in Figure 9.2-20. Such an orientation is considered conservative because were the wind assumed parallel to the short region axis there would be a larger volume with lower relative humidity and hence more evaporative loss from the sprayed water, hence a larger efficiency. However, the amount of water carried outside the spray pond boundary by the wind is minimum for the wind parallel to the long axis of the spray network and maximum for the wind parallel to the short axis.

Therefore, spray pond analyses performed using the Limerick Spray Pond model are either for maximum pond temperature or maximum water loss depending upon whether the long or short axis is input as the "length" of the spray network.

Since the total segment mass is explicitly assumed to be conserved, dM i

= wsp,i + ws,i-1 + w a,i ws,i - w a,i - w dp,i = 0 (EQ. 9.2-23) dt where:

wsp,i - is the mass flow rate from all spray nozzles to the ith spray region segment and is just the product of total mass flow rate to all nozzles in the spray network and the ratio of the segment area (in the x-y plane) to the spray region area; ws,i is the water vapor (steam) mass flow rate from the (i-1)th segment to the ith segment; wa,i is the dry air mass flow rate from the (i-1)th segment to the ith segment; ws,i - is the water vapor (steam) mass flow rate leaving the ith segment; wa,i - is the dry air mass flow rate leaving the ith segment; and wdp,i - is the dry water drop mass flow rate for all diameter groups leaving the ith segment for the spray pond.

As the segment mass is explicitly assumed to be conserved, so then are the individual air and steam-water (vapor-liquid) masses also conserved. Note that because of a change in phase, only the combination of steam plus water is conserved, steam and water are not conserved separately.

In addition, it is also assumed that there is no flow out the top of the spray region, Hence, it follows that CHAPTER 09 9.2-39 REV. 20, SEPTEMBER 2020

LGS UFSAR (EQ. 9.2-24) w sp,i = w dp,i + w ev,i w a,i = w a,i-1 w s,i = w s,i-1 + w ev,i where wev,i is the total evaporative rate from the drop of all diameter groups. As the mass flow rate entering the spray pond from the ith segment wdp,i is known, being determined in the drop dynamic module described below, the total rate of evaporation from all drops in the ith segment, wev,i, and the vapor (steam) mass flow rate to the (i+1)th segment, ws,i, is also known.

9.2.6.4.3.1.2.3 Low-Wind-Speed Model For the low-wind-speed submodel, the spray field is divided into a set of concentric rectangular segments as shown in Figure 9.2-21. The number of these segments is the same as the number in the high-wind speed model. Due to buoyancy, the warm humid air in each segment will rise and thereby draw cooler air from the outer segment. As a result, cool humid air (air-steam mixture) flows into each segment through its outer surface. Part of the flow exits the segment through the segment top surface and the remainder flows to the next inner segment. All the air entering the center segment leaves through the top of that segment. Although there is no prevailing wind, there is an induced wind as a result of buoyancy. As before, each segment is assumed to be in thermodynamic equilibrium (single segment temperature). The hot water from the spray enters through the segment top as in the case of the high-wind speed model. Heat energy lost by spray is gained by the moist air. An energy balance similar to that in the high-wind speed model is constructed which, in this case however, includes the air leaving through the top of the segment.

This flow out the top is calculated considering the buoyancy-driven motion Consider the free-convection for a segment of the spray region in which the moist air temperature varies about some mean value. If all the moist air in this region were at this mean temperature and if the fluid were not moving, the pressure gradient in the segment would be given by the equation of motion with u = 0:

p = <> g (EQ. 9.2-25))

in which the average density <> is at the average temperature <T>. If the velocity gradients result entirely from temperature inequalities, the moist air motion is usually slow and equation [EQ. 9.2-25] may be assumed to be a reasonably good approximation of the pressure gradient. Making this assumption, the time-independent equation of motion in the vertical (z) direction can be expressed in the form:

1 v 2i <>i dGi

= 1 - g + (EQ. 9.2-26))

2 z dV where dGi/dV is the rate per unit volume momentum is transferred to the moist air (air-steam mixture) in the segment via the falling drops. Integration over the segment volume and utilizing the Gauss-Ostrogradskii theorem,22 the above expression becomes yields the vertical velocity at the CHAPTER 09 9.2-40 REV. 20, SEPTEMBER 2020

LGS UFSAR top of the ith segment and is used to determine the induced wind velocity at the outer boundary of the spray region:

Gi vi = 21 gH + 2 (EQ. 9.2-27))

Si The parameter Si represents the flow area at the top of the ith segment. The rate of momentum transfer to the moist air (air-steam mixture) via the falling drops in the segment G i is given by Newton's Third Law and can be written as 1 Nk tF dy k dy G = - Rk C D Dk l 2

- V i l k - V i dt (EQ. 9.2-28))

2 k =1 4 t dt dt where Vi is the velocity in the vertical direction (along the y-axis) in the ith segment, Rk is the rate at which drops with an initial diameter Dk are created, and tF is the "flight time" of that drop. The flight time is defined as the elapsed time required for a drop to travel from the nozzle to the surface of the spray pond drop and depends upon the initial drop diameter.

As before, the total mass in the ith segment is postulated to be conserved. Hence, d Mi

= w sp,i + w s,i- 1 + w a,i w s,i - s,i A i v i - w a,i - a,i A i v i - w dp,i 0 (EQ. 9.2-29))

dt where Ai is the area of the sides of the ith segment and a,i is the air density of that segment.

Because of phase change (evaporation), only the combination of steam plus water is conserved in addition to air, steam and water are not conserved separately. Therefore, Wsp,i=Wdp,i+Wev,i Wa,i=Wa,i a,iAiVi (EQ. 9.2-30))

Ws,i=Ws,i-1+Wev,i-s,iAiVi Starting at the last (inner most) segment where all air flows through the top of the segment, the above equation can be applied in a bootstrap fashion until the first segment is reached and the exerting dry air mass flow rate, and hence the induced inlet velocity, can be calculated. Once this induced velocity is known, the temperature in each segment is calculated exactly as for the high-wind speed model, except for the flow of humid air (air-steam mixture) through the segment top.

Because the induced velocity is a function of the temperature difference, and thus the density difference, between the environment and a particular segment, this calculation proceeds iteratively until a convergence condition on the average drop temperature is satisfied.

CHAPTER 09 9.2-41 REV. 20, SEPTEMBER 2020

LGS UFSAR 9.2.6.4.3.1.2.4 Spray-Region Thermal Performance The spray water in the form of drops passes through this region, losing mass and energy along the way by evaporation, and enters the spray pond. It is crucial that the spray pond temperature remain below a certain value in order for the service water systems to remove the necessary amount of heat. This depends upon the evaporative mass and energy transfer from the spray drops to the air-steam mixture constituting the spray region atmosphere. This region or volume is characterized by the temperature and the relative humidity and is explicitly assumed to be at atmospheric pressure, defined as 1.01325 bar = 101.325 MPa.

The spray water is considered to enter the spray field from the top at the initial temperature and exit through the bottom to the spray pond. The temperature of the drop (final drop temperature) as it enters the pond depends upon the amount of evaporative cooling that has occurred which, in turn, depends upon the temperature and relative humidity of the spray region.

The spray region for both the high- and low-wind speed submodels are divided into N small segments of equal volume as shown in Figure 9.2-22. Under the segment indexing convention used in the computer program, the 0th section is considered to be the environment, or the atmosphere, adjacent to the spray region. The parameters describing the environment are the (dry bulb) temperature, wet bulb temperature, relative humidity, and prevailing wind speed. Except for the wind speed, these parameters are transient and specified by the user as input data. (Either the wet bulb temperature or the relative humidity, but not both, are specified and the unspecified parameter is calculated by the computer program.) The last segment is labeled as the (N+1)th segment. Each segment is considered to be in quasi-static thermodynamic equilibrium. Hence, the rate of energy change in the ith segment is zero and the energy gained by the air-steam (water) mixture is equal to that lost by the spray in that segment. Spray water enters the ith region from the nozzles with a flow rate wnz and a liquid specific enthalpy hf parameterized by the water temperature at the spray nozzle -- the spray temperature Tsp. The specific enthalpy at the nozzle is determined from the spray energy which is equal to the energy leaving the spray pond plus the heat load QHx QHx hf (Tsp) = hf (T p) + (EQ. 9.2-31))

w sp which is a defining equation for the spray temperature Tsp. Thus, the energy flow rate to the ith spray-region segment via sprayed water is Esp,i w sp,i hf (Tsp) (EQ. 9.2-32))

Drops shortly after leaving the nozzle with a discrete spectrum of sizes described by the parameter k which is defined as the volume fraction of all drops which form with an initial diameter D k. The drop group index k ranges from unity to a maximum value Nk which is the number of drop groups constituting the spectrum. Drops with initial diameter Dk are formed at a rate equal to 6 k v f Rk = w sp,i (EQ. 9.2-33)

D3k where vf is the specific volume of the liquid water evaluated at the spray initial temperature and N is the number of spray-region segments.

CHAPTER 09 9.2-42 REV. 20, SEPTEMBER 2020

LGS UFSAR Each drop, regardless of size, has an initial velocity uo oriented with angle to the horizontal, and moves ballistically subject to gravitation and drag, losing mass and energy along its trajectory until entering the spray pond. A drop which originally had a diameter Dk enters the pond with a final mass md,k which is smaller than the initial drop mass and a final drop temperature T d,k. The drop group index parameter k is affixed to indicate the particular group to which the drop belongs.

Equation [EQ 9.2-33] is also the rate at which drops that initially had a diameter Dk enter the spray pond, albeit with a smaller diameter. Drops which completely evaporate along their flight trajectory are considered to enter the spray pond with zero mass. The rate at which the spray in the form of drops enters the pond from the ith segment of the spray region is equal to the product of the rate of drop formation for drops with diameter Dk and the final mass md,k of a drop which had that initial diameter, summed over the drop-size spectrum:

Nk w dp,i = R k=1 k md,k (EQ. 9.2-34))

For the kth drop-diameter group, the energy loss rate from the ith spray-region segment is just the product of the drop specific enthalpy cpwTd,k, the final mass of the drop md,k and the rate at which drops in this diameter group are created Rk. The total rate at which energy in the form of drops leaves the ith spray region is just the sum of the rates for each drop-diameter group:

Nk Edp,i R k=1 k md,k c pw Td,k (EQ. 9.2-35))

Therefore, the rate at which the energy of the ith spray region changes is just the evaporative energy losses from the spray water, plus the energy of the steam (water vapor) and air entering the ith segment via the prevailing or induced wind, minus the rate of energy leaving the segment. By hypothesis, the segment is quasi-statically in equilibrium and thus:

dEi

= Esp,i - Edp,i + w a,i-1 c pa Ti-1 + ws,i-1 hs (Ti-1) - ws,i hs (Ti) - w a,i c pa Ti = 0 (EQ. 9.2-36))

dt where hs is the specific enthalpy of steam and Ti is the temperature of the ith spray-region segment. Between 273.15 K (0 C) and 373.15 K (100 C), the vapor (steam) specific enthalpy hs can be expressed as a linear function of the Celsius temperature plus an additive constant at the referent temperature of 0 C. Such an expression is accurate to within -0.13% and can be readily derived from steam tables. Hence, over this temperature range:

hg(T) ~

Cps (T - T0) + hso where cps = 1748.0 J/kg-C, hso = 2501.3 kJ/kg, and To = 273.15 K. It is similarly assumed that mass contained in the spray region does not change with time. Hence, in the ith segment, the rate of mass change is zero. In addition, the air and steam-water (vapor-liquid) masses are separately CHAPTER 09 9.2-43 REV. 20, SEPTEMBER 2020

LGS UFSAR conserved. It is here that differences arise between the high-wind speed model and the low-wind speed model. These differences entail how the steam and air mass flow rates between the ith and (i+1)th sections are calculated, and will be discussed below in the appropriate sections.

Using these expressions the energy balance equation for the ith segment (equation EQ. 9.2-36),

the equilibrium temperature is given by Ei + (w a,i- 1 cpa + w s,i- 1 c ps)Ti w ev,i (hso - cps To)

Ti =

w a,i c pa + w s,i c ps (EQ. 9.2-38))

Ei - w ev,i (c ps Ti- 1 + hs1)

= Ti- 1 +

w a,i c pa + w s,i cps where (EQ. 9.2-39))

Er,i Esp,i - Edp,i hs1 hso - cps To and wa,i-1 is the air (dry) mass flow rate entering the ith segment while wa,i is the flow rate leaving the segment.

The average spray-region thermal efficiency is a parameter useful for describing the performance of the spray region and is defined as Tsp - d

= (EQ. 9.2-40))

Tsp - Twb where Tsp is the temperature of the sprayed water at the nozzle, T wb is the meteorological wet-bulb temperature, and d is the average drop temperature which is given by N

w d,i c pw Td,i d i= 1 N (EQ. 9.2-41))

c pw w i= 1 d,i 9.2.6.4.3.1.3 Drop Model Between the time a drop leaves the nozzle and until it enters the spray pond, a drop moves along a ballistic trajectory exchanging both mass and energy with the spray region. The mass and energy CHAPTER 09 9.2-44 REV. 20, SEPTEMBER 2020

LGS UFSAR exchange rates are determined by the drop thermodynamics while the trajectory is determined by the drop dynamics. For a given spray-region segment, each drop diameter group of the drop-size spectrum is considered separately and the results of the total spectrum combined to determine the mass flow rate to the spray pond surface wdp,i, the energy flow rate to the pond surface dEdp,i/dt, and the average temperature of the spectrum Tdp,i.

9.2.6.4.3.1.3.1 Ballistic Motion A spherical drop of diameter Dk leaves the spray nozzle with an initial velocity uo which makes angle to the horizontal. The equations of motion for this drop are d 1 d mk mk u = - i Cd D2k lu - Vl(u - V) - mk g - u (EQ. 9.2-42))

dt 2 4 dt where u is the instantaneous drop velocity vector and V is wind velocity vector which is presumed to lie in the horizontal plane, and Cd is the drag coefficient which is a function of the Reynolds number. The last term on the right side of equation [EQ. 9.2-42] is the rate of change of the momentum of the evaporating drop mass and is required for momentum conservation.23 An approximation of Cd as a function of Reynolds number Re is given in the literature as24 24 (Re 2)

Re Cd ~ 185 (2 Re 500) (EQ. 9.2-43)

Re0.6 0.44 (Re 500) where the Reynolds number is given by i

Re = lu - Vl D k (EQ. 9.2-44))

i in which i and i are the average density and viscosity of the ith segment and Dk instantaneous drop diameter. Equation [EQ. 9.2-42] can be resolved along basis directions to yield d 2y k dy k mk = - i C D D 2 k lu - Vl - mk g dt2 8 dt d 2xk dxk mk = - C D 2 lu - Vl - V dt 2 8

i D k dt CHAPTER 09 9.2-45 REV. 20, SEPTEMBER 2020

LGS UFSAR Numerical solution of equation [EQ. 9.2-45] via Euler's method gives the trajectory of the drop expressed in terms of drop elevation yk and range xk which are functions of time. When the elevation of the drop yk is equal to the elevation of the spray pond surface yp, the integration ceases. This procedure is repeated for each of the Nk drop groups. The final drop mass of the kth group is md,k, i.e., the mass of the drop when its elevation is equal to that of the spray pond surface.

The rate at which drops in this group reach the pond surface is assumed to equal the rate at which they are created at the spray nozzle Rk. This is not strictly true as some drops will be carried by a prevailing wind beyond the spray pond boundary and thus become lost from the system. However, such an assumption is conservative as it deposits more energy to the spray pond than would occur in reality. Therefore, it is assumed that the rate for drops of the kth diameter group to enter the spray pond is equal to the rate at which they were created. Thus, the mass flow rate to the pond is given by equation [EQ. 9.2-34].

9.2.6.4.3.1.3.2 Mass and Energy Transfer As the spray nozzles generate a spectrum of drop sizes and thus diameters, the mass and energy lost from the drops must be integrated over all drop diameters. A default drop size spectrum for the Sprayco 1751A nozzle is contained in the computer program; however, a different spectrum may be entered. A drop from the spray nozzle has an initial temperature equal to that of the water flowing to the nozzle. The internal energy of a drop of diameter Dk is given by 3

Ed,k = md,k cvw (T d,k - T o) = D k cvw (T d,k - T o) (EQ. 9.2-46))

6 to a good approximation. The instantaneous rate at which the drop internal energy is changing is just the sum of the convective heat transfer from the drop to the ith spray-region segment plus the evaporative energy loss, the latter of which is the product of the evaporative mass loss and the vapor specific enthalpy evaluated at the drop temperature. Hence, the instantaneous rate of energy change is dEd,k Ed,k = - hx,k D 2k (Td,k - Ta,i) - me,k hg (Td,k) (EQ. 9.2-47))

dt where the last term on the right side of equation [EQ. 9.2-47] is the rate of evaporation loss. The convective heat transfer coefficient hx,k for a spherical drop of diameter Dk is given in the literature25 by ka hx,k = (2 + 0.6 Re1/ 2 Pr1/ 3) (EQ. 9.2-48))

Dk where ka is the thermal conductivity of air which depends upon temperature. The instantaneous rate of evaporation from a spherical drop can be expressed in terms of the difference in vapor pressure at the drop surface and in the region far from the drop:26 CHAPTER 09 9.2-46 REV. 20, SEPTEMBER 2020

LGS UFSAR dmd,k 2 xso - xs 2 psat (Td,k) - psat (Ta,i) me,k - = km,k D k = km,k D k (EQ. 9.2-49))

dt 1 - xso p - psat (Ta,i) where km,k is the mass-transfer coefficient and has units of kg/s-m2, xso is the vapor (steam) mass fraction at the drop surface, and xs the mass fraction of the spray region segment. The mass fractions of the vapor (steam) are determined assuming both air and the water vapor (steam) can be described by the equation of state for an ideal gas and is thus equal to the ratio of the partial pressure to the total pressure.

During evaporation, both heat and mass transfer are occurring. A brute force solution requires simultaneous solution of the momentum, heat, and mass transfer field equations. However, in some cases, the heat and mass transfer boundary-layer profiles can be made to collapse on each other by proper definition of variables. This allows the solution of many mass-transfer problems at low mass-transfer rates by analogy with corresponding problems in heat transfer. Using one of the Chilton-Colburn analogies,27 the mass-transfer coefficient for a drop of diameter Dk is given in the literature17 Dk 1/ 2 a,i 1/ 3

( )

a,i a,i u k a,i a,i a,i k m,k = 2 + 0.60 = 2 +

1/ 2 1/ 3 0.60 Rea,i Sc a,i (EQ. 9.2-50))

Dk a,i a,i a,i Dk modified to change units and the dimensionless number groups forming the Reynolds and Schmidt numbers will be immediately recognized. Note that the convective and evaporative energy losses are functions of the drop diameter, velocity and temperature and will necessarily change along the drop's trajectory. The drop internal energy rate of change can be expressed in terms of the rate of change of drop mass and temperature as follows (EQ.9.2-51) d Ed,k d md,k d Td,k 3

= cvw (Td,k - To) + md,k cvw = - hx,k D k (Td,k - To) - md,k hg (Td,k) dt dt dt from which the rate of change in drop temperature can be obtained to be 2

(EQ. 9.2-52) d Td,k hx,k D k (Td,k - Ta,i) - me,k [hfg (Td,k) + psat (Td,k) v f (Td,k)]

= -

dt 3 D k cvw 6

6 vf psat (Td,k) - psat (Ta,i)

= - hx,k (Td,k - T a,i) + k m,k ( hfg + psat v v )

D k cvw p - psat (T d,k)

As can be seen from equation [EQ. 9.2-52], the drop temperature rate of change consists of two terms representing convective and evaporative energy loss. Since evaporative mass and energy loss depend upon the drop diameter, each drop with diameter Dk will have a temperature Tk. This expression is numerically integrated from the time the drop is generated to the time when the drop enters the spray pond. During this integration, the properties of the air-steam (water) mixture in the CHAPTER 09 9.2-47 REV. 20, SEPTEMBER 2020

LGS UFSAR spray region are considered constant. It is necessary to correct the convective heat transfer coefficient used in equation [EQ. 9.2-52] and the mass transfer coefficient used in equation [EQ.

9.2-49] because these factors depend, to some degree, upon the rates of heat and mass transfer.

For low transfer rates, a multiplicative correction factor is adequate.

The rate of energy transfer to the spray pond by the drops from the ith segment is the product of the rate of drop formation, the final drop mass, and the liquid specific enthalpy, evaluated at the final drop temperature Td,k, also summed over the drop-size spectrum:

Nk Edp,i = R k=1 k md,k hf (T d,k) (EQ. 9.2-53))

Hence, the total evaporation rate and final drop temperature for the ith segment is just the difference between the rates at which mass enters the segment minus the rate at which mass leaves:

Nk w ev,i = w sp,i - R k =1 k md,k (EQ. 9.2-54))

The final drop mass and temperature are obtained by numerically integrating the equations [EQ.

9.2-49] and [EQ. 9.2-52]. Final drop mass and temperature are defined as mass and temperature of the drop just prior to entering the spray pond. This integration need not be performed in synchrony with the mass and energy flow rates to the spray region and pond because the characteristic time for changes in drop mass and temperature is much shorter than the characteristic time for changes in spray region and pond parameters.

9.2.6.4.3.1.4 Drift Loss Model The determination of water loss from the sprays due to the action of a prevailing wind, commonly referred to as "driftloss," has been incorporated as an integral part of the new Limerick Spray Pond model. Ideally, the driftloss mass flow rate would be calculated in conjunction with the drop dynamics of the ultimate-heat-sink computer program. However, as a practical matter, driftloss calculations are done in a separate module in order that an ultimate-heat-sink analysis requires a reasonable computer runtime. By computing the driftloss separately from the drop dynamics, certain approximations are necessary and are crafted such that the driftloss mass flow rate is conservative.

Mass loss from the spray region takes place when elevation of the drop yk is equal to the elevation of the spray pond surface and the drop horizontal position, or range, Rk is beyond the pond boundary. To calculate the drift loss, the ballistic trajectory of drops from each drop diameter group is determined using equation and expressed in terms of drop elevation yk and range xk as functions of time. However, the drop mass is assumed constant and equal to the final drop mass calculated in the drop dynamics module. A constant-mass approximation is conservative since the drop inertia along its trajectory is less than would normally be the case; hence, the drag forces due to the prevailing wind will increase the drop range. Because driftloss constitutes a small fraction of the total spray-pond mass loss, such an approximation is not severe.

CHAPTER 09 9.2-48 REV. 20, SEPTEMBER 2020

LGS UFSAR A water drop, assumed to be spherical with a constant diameter Dk, leaves the spray nozzle with an initial velocity uo at an angle to the horizontal. The equation of motion for this drop is du 1 2 mk = - i Cd Dk lu - Vl(u - V) - mk g (EQ. 9.2-55))

dt 2 4 where u is the instantaneous drop velocity vector and V is wind velocity vector which is presumed to lie in the horizontal plane, and Cd is the drag coefficient which is a function of the Reynolds number. The last term on the right side of equation [EQ. 9.2-55] is the rate of change of the momentum of the evaporating drop mass and is required for momentum conservation.28 An approximation of Cd as a function of Reynolds number Re is given above by [EQ. 9.2-43].

Equation 9.2-55 can be resolved along basis directions to yield (EQ. 9.2-56))

d 2y k dy mk 2

= - i C D D 2k lu - Vl k - mk g dt 8 dt 2

d xk dxk mk = - C D D 2 k lu - Vl - V dt 2 8

i dt Numerical solution of equation [EQ. 9.2-56] gives the trajectory of the drop expressed in terms of drop elevation yk and range xk which are functions of time. When the elevation of the drop yk is equal to the elevation of the spray pond surface yp, the integration ceases. This procedure is repeated for each of the Nk drop groups. The final drop mass of the kth group is md,k, i.e., the mass of the drop when its elevation is equal to that of the spray pond surface as determined in the drop dynamics module.

If, during the driftloss ballistic trajectory calculation, the range of the drop, when it reaches the elevation of the spray-pond surface, is greater than the distance to the spray-pond boundary, the drop is presumed lost. In addition, all drops of this diameter that originate from a location closer to the edge of the spray network than the difference between the drop range and distance from the network to the pond boundary are also lost. If the mass of the lost drop m*d,k is set is equal to the final drop mass md,k, the driftloss rate is given by w dl,i =

k k Rk m*d,k (EQ. 9.2-57))

where Rk is rate at which drops of the kth drop-diameter group are produced and k is fraction of all drops sprayed in this group that are lost and given by xk ly k =0 - B k = (EQ. 9.2-58))

Lsn CHAPTER 09 9.2-49 REV. 20, SEPTEMBER 2020

LGS UFSAR Clearly, the parameter k is bounded by unity and zero; that is, if it found to be less than zero, there is no drift loss is set equal to zero. The corresponding rate at which energy leaves the spray region is Edl =

k k Rk m*d,k c pw Td,k (EQ. 9.2-59))

These driftloss mass and energy flow rates are subtracted from the rate of change of spray-pond mass and energy during the spray-pond thermal performance determination.

9.2.6.4.3.2 Spray Cooling Thermal Performance Model The computer model developed for the analysis includes the effects of the following parameters:

a. Drop mean diameter

b. Wind speed and direction

c. Air dry-bulb temperature

d. Air wet-bulb temperature or relative humidity

e. Height of nozzles above water level

f. Pressure drop through the nozzle or height attained by the spray

g. Dimensions of the spray volume

h. Water flow rate in spray volume The spray pond is modeled as shown in Figure 9.2-20 for high wind speeds (above approximately 3 mph), and as shown in Figure 9.2-21 for low wind speeds, when cooling is assumed to be by natural convection only. The individual spray patterns are lumped together to form the spray volume, which is divided into a number of increments in the direction of the air movement. The temperature and vapor content of the air in each increment is assumed to be uniform within the increment, and is numerically the same as that exiting the preceding increment. The sprayed water temperature, air temperature, and air moisture content for each increment are calculated, and the results are combined to yield an average sprayed water temperature for the spray volume.

A critical aspect of the calculation is the determination of the evaporation rate within the increment.

The empirical work of Ranz and Marshall (Reference 9.2-2) on droplet heat and mass transfer is used as the basis for the evaporation rate and air temperature calculations. In their experiments, Ranz and Marshall suspended a drop from a capillary tube, supplied a known air flow over the drop surface, and measured the drop temperature, air temperature, drop diameter (held constant with water flow through the capillary tube from a microburet), and makeup flow rate from the microburet.

In this way the heat transfer and evaporation rate were measured. Heat and mass transfer coefficients were derived by correlation with the data.

CHAPTER 09 9.2-50 REV. 20, SEPTEMBER 2020

LGS UFSAR The increment mass and energy balance used in the calculation of spray cooling efficiency is shown schematically in Figure 9.2-22. Water enters the increment through the spray nozzles (flow rate (mws) at temperature (Ts)) and exits the increment after undergoing mass and energy transfer (flow rate (mwp) at temperature (Tdi)). The amount of mass and energy transferred is calculated from heat and mass transfer coefficients derived empirically:

Nnv = hcD = 2.0 + 0.60 Npr1/3Nre1/2 (EQ. 9.2-60)

K NSH = hdD = 2.0 + 0.60 Nsc1/3Nre1/2 (EQ. 9.2-61)

D where:

Nnv = Nusselt number NSH = Sherwood number hc = heat transfer coefficient for conduction and convection hd = mass transfer coefficient D = drop diameter K = thermal conductivity of air/vapor mixture Dv = diffusivity of vapor in air Npr = Prandtl number Nre = Reynolds number Nsc = Schmidt number The energy transfer rate is the sum of the contributions from conduction, convection, and evaporation. The lifetime of a drop in the increment, calculated from the pond geometry and other parameters affecting the drop trajectory, is used with the energy transfer rate to determine the temperature of the cooled water leaving the increment. The moisture content of the air leaving the increment is determined from the mass transfer (evaporation) rate and the air flow rate (residence time of the air in the increment). The temperature of air exiting the increment is calculated from an energy balance on the increment. The exit air temperature and moisture content for increment (i) is used in increment (i + 1) to determine the heat and mass transfer rate in that increment. This process is repeated until all the increments have been treated.

At low wind speeds (less than approximately 3 mph), air enters the spray volume from all sides rather than one; therefore, a new increment definition is used for low wind speeds. The definition is shown schematically in Figure 9.2-21. The air velocity entering each increment is determined from the density difference between the air/vapor mixture in the increment and the ambient.

CHAPTER 09 9.2-51 REV. 20, SEPTEMBER 2020

LGS UFSAR The spray cooling efficiency for zero wind speed is used as the lower limit for the spray cooling efficiency calculated for nonzero wind speeds (no natural convection). That is, if the spray cooling efficiency calculated by neglecting natural convection is less than that calculated with natural convection only, then the efficiency is taken to be that determined for natural convection (zero wind speed) only. This procedure shows good agreement with the test results, and avoids excessive conservatism.

The result of the calculation described above is a set of cooled water temperatures, one for each increment. Since the air temperature and moisture content for each increment is different, the cooled water temperatures are different. The incremental flow rate weighted average cooled water temperature, T , is calculated.

n T = FiTi/n (EQ. 9.2-62) i=1 where:

Fi = spray water flow rate in increment (i)

Ti = cooled water temperature, increment (i) n = number of increments in the spray volume The thermal efficiency, Eth, is calculated from the ambient air wet-bulb temperature, Twb, and the water temperature before spraying, Ts.

Eth = Ts - _T (EQ. 9.2-63)

Ts - Twb The thermal performance prediction model gives more conservative results than other prediction methods available. The primary conservatism in the model is the lack of convective air motion into the spray volume (for all but very low wind speeds), which results in lower calculated efficiencies.

The convective air motion is most important at low wind speeds; consequently, the degree of conservatism increases as wind speed decreases. Since thermal performance at low wind speeds is most important, this is a desirable effect as long as the degree of conservatism is not unrealistic. Data taken at an existing spray pond have been used to demonstrate the degree of conservatism of the model, as discussed in Section 9.2.6.4.3.4.

9.2.6.4.3.3 Results of the UHS Performance Evaluation The results of the performance evaluation of the spray pond system demonstrate that the spray pond meets the requirements for the maximum cooling water temperature and the 30-day water supply. The cooling water temperature for the design basis accident heat load is presented in Figure 9.2-23 and Table 9.2-17. The maximum calculated cooling water temperature, for the design (accident) heat load, is 93.48F. This temperature is below the spray pond thermal CHAPTER 09 9.2-52 REV. 20, SEPTEMBER 2020

LGS UFSAR performance design limit of 95F. The minimum remaining water volume in the spray pond after 30 days is approximately 4.74 million gallons for the maximum water-loss analysis.

Furthermore, adequate means are provided in the design of the spray pond system to ensure continued cooling capability beyond 30 days. Water required to maintain the functional capability of the spray pond can be supplied by truck or other means.

9.2.6.4.3.4 Comparison of Spray Pond Thermal Performance Results In order to verify the conservatism of the spray pond thermal performance model, the model has been applied to a the Rancho Seco spray pond system which has well-documented performance (Reference 9.2-3). At the request of the Atomic Energy Commission, SMUD arranged to have the Rancho Seco spray ponds tested to verify the ability of the ponds to meet the design criteria.

SMUD asked the University of California, Berkeley, to perform an evaluation of the performance of the ponds. Of particular interest from a performance standpoint were the thermal efficiency of the nozzles and drift loss versus wind speed.

A comparison of the relevant parameters between the Rancho Seco and LGS ponds is given in Table 9.2-18.

The result of the comparison of the Rancho Seco performance test and model predictions is given in Table 9.2-19. It can be seen that the model predictions for the performance are more conservative than the measured efficiencies.

9.2.6.5 Safety Evaluation 9.2.6.5.1 Thermal Performance As demonstrated in Section 9.2.6.4, the full spray pond is capable of providing cooling water within the design temperature limit for at least 30 days for the design basis event using conservative meteorology and assumptions.

9.2.6.5.2 Effects of Severe Natural Events or Site-Related Events The UHS is capable of fulfilling its safety function concurrent with any of the following events: SSE; tornado; flood; drought; transportation accident; or fire.

The UHS is designed as seismic Category I to ensure that it remains functional following an SSE.

Pond and pump structure designs are discussed in Section 3.8.4. The effects of the SSE with respect to wave generation and wave loadings on the network support columns are discussed in Section 2.4.8.

Functional integrity is maintained during a tornado because of the redundancy and separation of the networks, and the tornado protection of the spray pond pumphouse, and contingency procedures as discussed in Section 3.5.1.4.

The PMF level for the Schuylkill River is at el 181', whereas the bottom of the spray pond is at el 241', 60 feet above the PMF level. Therefore, the spray pond is affected only by flooding due to precipitation on the pond, and run-off from the adjacent drainage area. This is discussed in Section 2.4.8. The provision of an emergency spillway ensures that the pond cannot reach a level that CHAPTER 09 9.2-53 REV. 20, SEPTEMBER 2020

LGS UFSAR could impair performance. Further, calculations show that the volume and velocity of water after PMF will not cause erosion.

Drought does not impair the cooling capability of the spray pond system. The spray pond is designed so that makeup water is not required for at least 30 days.

Potential damage to the spray pond from the design transportation accident (railroad explosion) is prevented by the physical separation between the spray pond and the tracks of the railway. The design transportation accident is discussed in Section 2.2.3.

Fire protection for the spray pond pumphouse is discussed in Section 9.5.1.

The spray pond system is physically removed from the balance of the station, except for interconnecting piping. The effect of failure of other station components cannot impair the capability of the spray pond to fulfill its safety objective.

9.2.6.5.3 Freezing Considerations The only safety-related systems with piping that could be exposed to freezing conditions are the ESW and RHRSW systems. The majority of the piping for these systems is located inside heated buildings or buried below the frost line and therefore are not exposed to freezing conditions. The spray pond network piping is drained after each use and therefore is protected from freezing. Any piping that is not within buildings, buried, or drained is electrically heat traced to protect against freezing. The electrical supply for the heat tracing is not supplied from a Class 1E source because, in the event of auxiliary power loss, water would be flowing in the piping. This design is consistent with IEEE 622 (1979), "Recommended Practice for the Design and Installation of Electric Pipe Heating Systems for Nuclear Power Generating Stations."

The spray pond is permitted to freeze during cold weather; ice formation may occur at the pond surface. The spray pond system is designed to perform its safety function during periods of maximum freezing. The design basis freezing event is one producing a maximum of 15 inches of ice on the pond surface. Maximum ice thickness was determined by calculation of ice thickness for 33 winters from 1941 to 1974, using the daily average Philadelphia meteorology (Section 9.2.6.4.1). Heat and mass transfer analyses established mathematical models of freezing of the pond surface. These models are compared with measured reservoir ice thickness data to verify a conservative design. For each freezing period, a daily incremental thickness was calculated. The maximum thickness calculated for the 33 year period was 13.4 inches in February 1961. A design value of 15 inches has been selected to provide additional conservatism. Measurements taken on quiescent pools near the site during the winter of 1977, considered to be the most severe on record for this area, were all less than the design thickness.

Two spray network bypass lines are provided in the design to allow spray pond operation while ice exists at the pond surface, as discussed in Section 9.2.6.3.2.

An analysis demonstrates that hole formation occurs well before the system design maximum temperature of 95F is reached at the pump suction. This analysis assumes hydraulic short circuiting between the discharge and suction piping and a simultaneous one-unit LOCA and shutdown of the other unit.

CHAPTER 09 9.2-54 REV. 20, SEPTEMBER 2020

LGS UFSAR Frazil ice formation does not interfere with operation of the spray pond system. The phenomenon of frazil ice formation is associated with supercooled and generally swiftly flowing water. Since the spray pond is a nonflowing body of water, supercooling of the water cannot occur during either standby or operation of the spray pond system. A small amount of frazil ice formation at the surface of the spray pond is possible during periods of high winds. However, any mixing of the surface water deeper into the pond results in melting of the frazil ice. Once an ice layer forms on the pond, the effects of wind on water velocities and turbulence are eliminated, surface heat transfer will be greatly reduced, and frazil ice does not form.

The formation of anchor ice in the spray pond may occur under high wind conditions, but would be restricted to the uppermost part of the support columns. Ice accumulation on the support columns does not interfere with the operation of the spray pond system. As with frazil ice, formation of an ice layer inhibits anchor ice formation.

An accumulation of ice due to freezing rain sufficient to cause complete blockage of the spray nozzle was considered, but is not considered credible because of the large 1-9/64 inch opening and the shape of each nozzle, which has a raised vertical lip and sloped contour.

The design of concrete columns and footings supporting the pipe network in the spray pond considers the loads due to ice formation and ice expansion. These safety-related components have been adequately designed by combining these loads with other concurrent loads according to Table 3.8-10, including an SSE coincident with maximum ice thickness.

The interior of the spray pond pump structure is heated to a minimum of 65 oF to prevent icing of the pumps and other components, or icing at the surface of the water in the wet pits. Ice from the spray pond is prevented from entering the wet pits by locating the tops of the sluice gates below normal pond water level.

9.2.6.5.4 Other Considerations The ability of the UHS to perform its safety function assuming a single failure of an active component is demonstrated in discussions of the ESW and RHRSW systems, in Sections 9.2.2 and 9.2.3, respectively.

Transient analyses have been performed to determine the potential effects of water hammer on the spray networks. These analyses indicated that water hammer loads will not be significant and are adequately provided for in the design of the spray networks.

There are no credible failures of manmade structural features associated with the UHS. As discussed previously in this section, the spray pond is constructed completely in excavation.

Structures associated with the pond (the pumphouse, spray network support piers, and the overflow weir) are designed to seismic Category I requirements. The system is designed so that the effects of flooding, missiles, and fire cannot affect the spray pond's safety function, as discussed in Sections 3.4, 3.5, and 9.5.1, respectively.

9.2.6.6 Conformance to Regulatory Guide 1.27 Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Power Plants," contains guidance on the design and performance analysis for ultimate heat sinks.

CHAPTER 09 9.2-55 REV. 20, SEPTEMBER 2020

LGS UFSAR The LGS spray pond UHS was designed taking Revision 1 of this regulatory guide into consideration. The LGS UHS and performance analysis is in accordance with this revision of the guide, as demonstrated in this section, with the following exceptions and clarifications by regulatory guide position paragraph:

a. Paragraph C.1.a: The LGS performance analysis complies with the intent of this paragraph, in that a more conservative "coefficient of water loss" combining dry-bulb temperature, wet-bulb temperature, and wind speeds is used to determine the weather data periods resulting in the maximum amount of evaporation and drift loss.

b. Paragraph C.1.b: LGS complies partially with this paragraph,in that average temperatures for the 1 day and 30 day periods are used instead of diurnal variations in temperature. The coefficient of thermal performance was calculated for each day, using daily average data (Section 9.2.6.4.1).

Revision 2 of this regulatory guide is presently in effect. With respect to this revision, the following exceptions apply:

a. Paragraph C.1: Same as response to Paragraphs C.1.a and C.1.b of Revision 1 above.

b. Paragraph C.1.a: The critical time periods approach is not used as described in the LGS analysis. The worst 1 day followed by the worst 30 day meteorological conditions were used in accordance with Revision 1 of the regulatory guide. This approach, along with the other conservatisms in the analysis, is adequate to verify pond performance under design conditions.

c. Paragraph C.1.b: Same as Paragraph C.1 and C.1.a above.

d. Paragraph C.1.c: This paragraph is not applicable to the LGS design. The coefficient of thermal performance method used in the LGS design selects conservative design basis meteorology, as discussed in Section 9.2.6.4.1.

With respect to Paragraph C.2 of both revisions of the guide, LGS meets the criteria as discussed in Sections 3.4, 3.5, 9.2.6.5.4, and 9.5.1.

With respect to Paragraph C.3 of both revisions of the guide, LGS employs the option of one source of water for the UHS, for which, as discussed in Section 9.2.6.5.4 and the referenced sections therein, there is an extremely low probability of loss of capability.

9.2.6.7 Instrumentation and Alarms As shown in drawing M-12, spray pond low level and high temperature are alarmed in the control room. Spray pond surface temperature (within the upper 2 feet of the surface) is provided in the control room. During operation of the RHRSW system, spray pond temperature can also be obtained via the inlet to the RHR heat exchanger in the control room.

Pond makeup and blowdown volumes are indicated by flow totalizers located in the makeup and blowdown lines.

CHAPTER 09 9.2-56 REV. 20, SEPTEMBER 2020

LGS UFSAR 9.2.6.8 Tests and Inspections The UHS is preoperationally tested in accordance with the requirements of Chapter 14.

A seepage test was performed to verify that the values used in this analysis are conservative (Section 2.5.4.6.1).

Preoperational tests are performed to verify that design flows and pressures to the networks and within the networks are obtained.

Analyses have indicated that water hammer loads will not be significant, and therefore, no special testing for water hammer loads will be performed.

The verification of proper system flow and distribution, and the results of the SSES thermal performance tests (Docket No. 50-387) verify the LGS design, and therefore a thermal performance test for LGS is not considered necessary.

The spray pond system is tested periodically during normal plant operation in accordance with the requirements of Chapter 16.

9.2.7 CONDENSATE AND REFUELING WATER STORAGE FACILITIES The condensate and refueling water storage facilities provide storage of condensate water for use in normal plant operations and refueling operations. The facilities have no safety-related function, except for the section of piping located in the reactor enclosure that supplies condensate to the HPCI pump.

9.2.7.1 Design Bases The condensate and refueling water storage facilities are designed to perform the following functions:

a. Supply water to fill the reactor well, the dryer/separator storage pool, and the spent fuel cask storage pit of one unit during refueling operations, and provide storage for this water when refueling is completed.

b. Supply condensate for various processes in the radwaste system and makeup for the plant systems, including the condenser hotwells.

c. Supply condensate to the suctions of the HPCI, RCIC, and core spray pumps of both units.

d. Provide a minimum storage capacity of 135,000 gallons for the RCIC and HPCI pumps associated with each unit.

e. Provide the capability to filter the water in the refueling water storage tank by pumping it through the condensate demineralizers and returning it to the storage tank.

CHAPTER 09 9.2-57 REV. 20, SEPTEMBER 2020

LGS UFSAR

f. Provide storage for condensate rejected from the cycle.

g. Provide storage for condensate from the radwaste system.

h. Provide the capability to drain the reactor well through the condensate filter demineralizer and back to the storage tank.

i. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the condensate and refueling water storage facilities are discussed in Section 3.2.

j. Postulated radioactive releases due to liquid radwaste tank failures are discussed in Section 2.4.12.

9.2.7.2 System Description The condensate and refueling water storage system is shown in drawing M-08. Design parameters for the major components of the system are given in Table 9.2-20.

9.2.7.2.1 Condensate Storage Tanks (Units 1 and 2)

The CSTs are the normal source of water for the HPCI and RCIC pumps for both operational use and testing. In addition, the tanks supply water to the core spray system for testing. For a discussion of the HPCI and core spray systems, refer to Section 6.3. For a discussion of the RCIC system, refer to Section 5.4.6.

The condensate transfer pumps and condensate transfer jockey pump take their suction from either of these tanks to provide water for various services in the radwaste enclosure, the reactor enclosure, and for the fuel pool filter/demineralizer backwash.

The lead condensate transfer pump, as selected by the hand selector switch, runs continuously and the standby condensate transfer pump will start automatically on low discharge header pressure. The condensate transfer jockey pump will operate only as a manual backup. The condensate transfer pumps and condensate transfer jockey pump are automatically tripped by a level switch upon receipt of a CST low level signal, and a discharge header low flow signal automatically trips the standby condensate transfer pump.

Makeup to the CSTs is supplied by the demineralized water transfer pumps. The tanks also act as surge tanks for the condensate system by receiving any rejected condensate from and making up any deficiency in the heat cycle.

9.2.7.2.2 Refueling Water Storage Tank The refueling water storage tank stores the water that is used to fill the reactor well, the dryer/separator storage pool, and the fuel cask storage pit of both Units 1 and 2.

During refueling operations the water is pumped from the storage tank to the respective reactor well and dryer/separator pool via the manually operated refueling water pumps.

CHAPTER 09 9.2-58 REV. 20, SEPTEMBER 2020

LGS UFSAR When refueling is complete, the water in the reactor well and dryer/separator pool is pumped by the refueling water pumps to the storage tank through one of the condensate filter/demineralizers.

Makeup for the refueling water storage tank is supplied by the demineralized water transfer pumps, taking suction from the demineralized water storage tank.

The refueling water storage tank also provides water to fill the spent fuel cask storage pit. This water can be returned to the tank by the refueling water pumps through one of the condensate filter/demineralizers.

The Unit 1 and the Unit 2 CSTs and the refueling water storage tank are located outdoors, and are provided with freeze protection. The area occupied by the Unit 1 CST and refueling water storage tank is surrounded by a common dike capable of holding the combined contents of the tanks in the event of tank rupture or overflow. The Unit 2 CST is surrounded by a dike capable of holding the tank contents in the event of tank rupture or overflow. Tank overflows are routed directly to the liquid radwaste system. Drains from the dike areas can be selectively routed to either the normal waste or radwaste systems.

9.2.7.3 Safety Evaluation The condensate and refueling water storage facilities have no safety-related function except for the section of piping located in the reactor enclosure that supplies condensate to the HPCI, RCIC, and core spray pumps (Section 6.3). Failure of the nonsafety-related portions of the system does not compromise any safety-related system or component, nor does it prevent a safe shutdown of the plant.

9.2.7.4 Tests and Inspections The condensate and refueling water storage facilities are preoperationally tested in accordance with Chapter 14.

9.2.7.5 Instrumentation Applications 9.2.7.5.1 Condensate Storage Tanks These tanks are each provided with a level transmitter that operates a pen recorder located in the control room. In addition to the level transmitters, each tank has high and low level switches that alarm in the control room, and a low switch that trips the condensate transfer pumps and the condensate transfer jockey pump when the tank level reaches low level. The tank has a high level switch which actuates on CST high level to automatically close the condenser reject valves.

9.2.7.5.2 Refueling Water Storage Tank This tank is provided with a level transmitter that operates pens on Unit 1 and Unit 2 pen recorders located in the control room. In addition, the tank has high and low level switches which alarm in the control room, and a low level switch that trips the refueling water pumps on tank low level.

9.2.8 REACTOR ENCLOSURE COOLING WATER SYSTEM The RECW system is a closed-loop system that provides cooling water for miscellaneous reactor auxiliary plant equipment. The RECW system is not safety-related, except for the containment CHAPTER 09 9.2-59 REV. 20, SEPTEMBER 2020

LGS UFSAR penetrations and isolation valves associated with the water supply to the reactor recirculation pump seal and motor oil coolers.

9.2.8.1 Design Bases

a. The RECW system is designed to remove the maximum anticipated heat loads developed by the components served by the system over the full range of the normal plant operating conditions and ambient temperature conditions.

b. The RECW system is designed to operate during normal operation and on LOOP without occurrence of a LOCA.

c. The system is designed to permit the use of corrosion inhibitors to prevent long-term corrosion and organic fouling of the water passages in the system.

d. The RECW system is designed to serve as a barrier between potentially radioactive systems and the plant service water system.

e. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the RECW system are discussed in Section 3.2.

f. A thermal relief valve was installed in Unit 1 and 2 primary containments on non-safety related RECW lines to enhance the design of the RECW systems to limit thermally induced pressurization concerns that were described in Generic Letter 96-06.

9.2.8.2 System Description The RECW system consists of two 100% capacity cooling water pumps, two 100% heat exchangers, one head tank, one chemical addition tank, associated valves, piping, and controls as shown in drawing M-13. Major equipment design parameters are summarized in Table 9.2-21.

The RECW system provides demineralized cooling water to nonessential equipment located in the reactor and radwaste enclosures which have the potential to carry radioactive fluids, or which require a clean water supply to minimize long-term corrosion.

During normal operation, one RECW pump and one or two heat exchangers is in service. The second pump automatically starts on low pressure in the supply header. During normal plant operation, the RECW system furnishes cooling water to the following components:

a. Cleanup nonregenerative heat exchangers

b. Cleanup recirculation pump seal coolers/motor coolers

c. Reactor recirculation pump seal and motor oil coolers

d. Reactor enclosure equipment drain sump cooler

e. Sample station coolers CHAPTER 09 9.2-60 REV. 20, SEPTEMBER 2020

LGS UFSAR

f. PCIG compressors and aftercoolers The demineralized cooling water is circulated throughout the closed-loop by the RECW pumps.

The pump motors are connected to Class 1E busses, and on a LOOP without occurrence of a LOCA, both pumps automatically restart. In this mode, the RECW system may be used to furnish cooling water to the reactor enclosure equipment drain sump cooler, and the reactor recirculation pump seal and motor oil coolers. However, the cooling capacity of RECW will be limited by the unavailability of the Service Water System. The rest of the system can be isolated by a valve which is remote manually operated from the control room. Cooling water to the RECW heat exchanger can be provided by the ESW System after a LOOP by manual realignment of the ESW and Service Water systems.

On a LOCA, both RECW pumps are tripped when their respective load centers are shed from the Class 1E bus, and both pumps are inhibited from automatically starting; however, they can be manually started as allowed by diesel generator loading. Coolant for the reactor recirculation pump seal and motor oil coolers can also be supplied by the ESW system if the RECW supply is not available. This switch-over can be accomplished by manual operation of the connecting valves.

The connecting valves with motor operators have their power supplies disconnected, as shown in drawing M-13. The primary containment isolation valves associated with the RECW system are discussed in Section 6.2.4.

The RECW system has also the capability to supply cooling water to the fuel pool heat exchangers via a removable spool piece. The RECW system also has the capability during shutdown to supply drywell chilled water to the RWCU non-regenerative heat exchanger(s) and the 1A RWCU pump motor cooler in order to increase the cooling capacity of the RWCU alternate method of reactor decay heat removal. This line-up may only be established when containment isolation valves are not required to be operable or when the affected containment penetration is isolated as dictated by the Technical Specifications and Technical Requirements Manual.

Makeup water is supplied to the RECW head tank from the demineralized water system. The head tank provides necessary makeup water to the RECW system, as required. Chemicals are added to the system through the chemical addition tank for corrosion prevention.

The RECW pumps, heat exchangers, chemical addition tank, and head tank are all located in the reactor enclosure.

9.2.8.3 Safety Evaluation The RECW system has no safety-related function, and is not required to be operable following a LOCA. Failure of the system does not compromise any safety-related system or component, nor does it prevent a safe shutdown of the plant.

RECW system piping and valves are discussed in Section 3.7, 6.2, and Table 3.2-1..

The reactor recirculation pump seals are cooled following a shutdown of the pump caused by LOOP for economic, and not safety reasons.

9.2.8.4 Tests and Inspections The RECW system is preoperationally tested in accordance with the requirements of Chapter 14.

CHAPTER 09 9.2-61 REV. 20, SEPTEMBER 2020

LGS UFSAR 9.2.8.5 Instrumentation Applications Local and remote indicators, alarms, and pressure relief valves are provided to monitor the system process and to protect system components. Low pressure and high temperature in the RECW system supply header are alarmed in the control room. Demineralized water supply to the RECW head tank is controlled by a level switch in the tank. High and low levels in the tank are also alarmed in the control room.

Continuous radiation monitors are installed in the pump suction header of the RECW system. This instrumentation records and alarms radioactivity leakage into the system in the control room.

9.2.9 TURBINE ENCLOSURE COOLING WATER SYSTEM The TECW system is a closed-loop cooling system that provides cooling water for miscellaneous turbine plant components. The TECW system is not safety-related.

9.2.9.1 Design Bases

a. The TECW system is designed to remove the maximum anticipated heat loads developed by the components served by the system, over the full range of the normal plant operating conditions and ambient temperature conditions.

b. The TECW system is designed to operate during normal plant operation and on LOOP without occurrence of a LOCA.

c. The system is designed to permit use of corrosion inhibitors to prevent long-term corrosion and organic fouling of the water passages in the system.

d. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the TECW system are discussed in Section 3.2.

9.2.9.2 System Description The TECW system consists of two 100% capacity cooling water pumps, two 100% capacity heat exchangers, one head tank, one chemical addition tank, and associated valves, piping, and controls as shown on drawing M-14. Major equipment design parameters are summarized in Table 9.2-22.

During normal plant operation, one TECW pump and one or two heat exchanger are in service.

The second pump starts automatically on low pressure in the supply header. During normal plant operation, the TECW system furnishes demineralized cooling water to the following turbine plant components:

a. Condensate pump motor bearing oil coolers

b. Instrument air compressors and aftercoolers

c. Service air compressor and aftercooler CHAPTER 09 9.2-62 REV. 20, SEPTEMBER 2020

LGS UFSAR

d. Sample station coolers Demineralized cooling water is circulated throughout the closed-loop by the TECW pumps. The pump motors are connected to Class 1E busses, and on a LOOP both pumps automatically start.

On a LOCA, both pumps trip when their respective load centers are shed from the Class 1E bus.

Both pumps are inhibited from restarting; however, the operator can manually reconnect the pumps to the Class 1E bus if diesel loadings permit.

Makeup water is supplied to the TECW head tank from the demineralized water system. The head tank provides necessary makeup water to the TECW system, as required. When required, chemicals are added to the system through the chemical addition tank for corrosion prevention.

Valves on the cooling water supply to the instrument and service air compressor aftercoolers, and water jackets are interlocked with the compressor motors so that cooling water is provided while the compressors are running. Temperature control valves modulate the cooling water flow to maintain a set compressor temperature. The flow rate of cooling water to all other coolers is manually set by individual valves on the cooling water piping for each unit.

The TECW system pumps, heat exchangers, chemical addition tank, and head tank are all located in the turbine enclosure.

9.2.9.3 Safety Evaluation The TECW system has no safety-related function, and is not required to be operable following a LOCA. Failure of the system does not compromise any safety-related system or components, nor does it prevent a safe shutdown of the plant.

9.2.9.4 Tests and Inspections The TECW system is preoperationally tested in accordance with the requirements of Chapter 14.

9.2.9.5 Instrumentation Applications Local and remote indicators, alarms, and pressure relief valves are provided to monitor the system process and protect system components. Low pressure and high temperature in the TECW supply header are alarmed in the control room. The demineralized water supply to the TECW head tank is controlled by a level switch in the tank. High and low levels in the tank are also alarmed in the control room.

9.2.10 CHILLED WATER SYSTEMS 9.2.10.1 Drywell Chilled Water System The DCWS provides chilled water for cooling the air supply to the following areas:

a. Drywell

b. Reactor enclosure

c. Refueling area CHAPTER 09 9.2-63 REV. 20, SEPTEMBER 2020

LGS UFSAR

d. Turbine enclosure

e. Radwaste enclosure In addition, the system provides chilled water to the following equipment:

a. Recirculation pump motor air cooler

b. Drywell equipment drain sump cooling coil

c. Sample coolers

d. Mechanical vacuum pump seal cooler

e. RWCU non-regenerative heat exchangers

f. 1A RWCU pump motor cooler 9.2.10.1.1 Design Bases

a. The DCWS is not safety-related, except for the containment penetrations and the containment isolation valves. These are designed to meet seismic Category I requirements and are discussed in Section 6.2.4.

b. The chilled water piping inside the drywell is designed to ensure that it has no adverse effects on adjacent safety-related equipment in the event of a SSE.

c. During normal operation, the DCWS is designed to provide chilled water at 50 F to components serviced by the system.

d. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the DCWS are discussed in Section 3.2.

9.2.10.1.2 System Description 9.2.10.1.2.1 General Description The DCWS is shown in drawing M-87. Design parameters for major components of the DCWS are listed in Table 9.2-23. Major components of the system include the following:

a. Two centrifugal water chillers

b. Two chilled water circulating pumps

c. One head tank (expansion tank)

d. One chemical feed tank The water chiller unit includes a centrifugal compressor, condenser, evaporator, pump-out unit, oil pump, oil heater, all refrigerant piping, instrumentation, and controls. Each chiller unit and circulating pump is sized for 100% capacity.

CHAPTER 09 9.2-64 REV. 20, SEPTEMBER 2020

LGS UFSAR Cooling water for the chiller condensers is provided by the service water system. Makeup water for the DCWS is provided by the demineralized water system through an automatic makeup water valve.

The chemical addition subsystem provides control against corrosion. Chemical addition to the system is manually initiated.

The head tank is vented to the atmosphere, and is located at the highest elevation in the system.

9.2.10.1.2.2 System Operation During normal operation, one chilled water circulating pump and one chiller unit operate. However, to provide additional drywell cooling during normal operation, the system may be aligned with 2 chilled water circulating pumps, 2 containment piping systems, and one chiller unit operating.

Reactor enclosure supply air cooling coils are not supplied with chilled water under normal conditions. The second chiller and pump can be started in the event of high temperature in the drywell, or reactor enclosure. The standby chiller is manually started if there is a failure of the operating chiller. The standby pump starts automatically or can be started manually if there is a failure of the operating pump.

The closed-loop refrigerant system in the water chiller unit extracts heat from water in the evaporator, and rejects heat to service water in the condenser. A temperature sensor, in the chilled water outlet, positions the inlet guide vanes of the compressor to maintain a constant outlet water temperature.

Inside containment, two independent piping systems connected to two sets of coolers are provided.

Changeover from one to the other system is accomplished by switching the containment isolation valves from the control room. A containment isolation signal closes the containment isolation valves. The valves can be reopened from the control room, if desired, to resume chilled water flow.

The RECW system may provide backup to the portion of DCWS that serves the drywell only when containment isolation valves are not required to be operable. In addition, the associated containment isolation valves for this service are administratively controlled as locked closed valves as described in UFSAR Section 6.2.4.3.1.3.2.11.

The RECW system also has the capability during shutdown to supply drywell chilled water to the RWCU non-regenerative heat exchanger(s) and the 1A RWCU pump motor cooler in order to increase the cooling capacity of the RWCU alternate method of reactor decay heat removal. This line-up may only be established when containment isolation valves are not required to be operable or when the affected containment penetration is isolated as dictated by the Technical Specifications and Technical Requirements Manual.

Chilled water flow through the cooling coils and unit coolers is controlled by temperature-actuated valves, as shown in drawing M-87.

Chilled water can be supplied to the refueling floor air supply cooling coil, to provide comfortable conditions for plant personnel.

9.2.10.1.3 Safety Evaluation CHAPTER 09 9.2-65 REV. 20, SEPTEMBER 2020

LGS UFSAR The DCWS has no safety-related function, except for the containment penetrations and containment isolation valves which are described in Section 6.2.4.

9.2.10.1.4 Tests and Inspections The DCWS is preoperationally tested in accordance with the requirements of Chapter 14.

9.2.10.1.5 Instrumentation Applications Chillers and pumps are operated from the control room. Failure of the lead chiller or pump causes an alarm in the control room. Pressures at pump suction, pump discharge, and chiller discharge are displayed locally. Chilled water temperatures and flows are indicated in the control room.

Level in the head tank is indicated locally and alarmed through a computer in the control room at low or high level.

9.2.10.2 Control Structure Chilled Water System The CSCWS provides chilled water to maintain stipulated ambient air temperature (Section 3.11) in the following areas:

a. Control room

b. Auxiliary equipment room including the computer room

c. Emergency switchgear compartment

d. Battery room

e. SGTS compartment and access area 9.2.10.2.1 Design Bases

a. The CSCWS is designed to remain functional following an SSE.

b. The CSCWS is designed to supply adequate chilled water at 44oF to maintain the stipulated ambient air temperatures.

c. The CSCWS is designed so that a single failure of any active component, assuming LOOP, cannot result in the loss of chilled water supply.

d. The seismic category, quality group classification, and corresponding codes and standards that apply to the design of the CSCWS are discussed in Section 3.2.

e. The CSCWS supports operation of the Control Room HVAC System by providing chilled water during normal and design bases accident conditions.

9.2.10.2.2 System Description 9.2.10.2.2.1 General Description CHAPTER 09 9.2-66 REV. 20, SEPTEMBER 2020

LGS UFSAR The CSCWS is shown in drawing M-90. The system is common to Units 1 and 2, and consists of two independent, 100% capacity, chilled water systems. Piping interties, with double isolation valves, are provided between the two systems to provide flexibility. Design parameters for major components of the CSCWS are listed in Table 9.2-24. Major components of each system include the following:

CHAPTER 09 9.2-67 REV. 20, SEPTEMBER 2020

LGS UFSAR

a. One centrifugal water chiller

b. One chilled water circulating pump

c. One head tank (expansion tank)

d. One chemical feed tank The chiller unit includes a centrifugal compressor, condenser, evaporator, pump-out unit, oil pump, oil heater, all refrigerant piping, and controls.

Cooling water for the chiller condensers is provided by the service water system during normal operation. On LOOP, the cooling water supply automatically switches over to the ESW system.

Makeup water for the CSCWS is provided by the demineralized water system through an automatic makeup water valve.

The chemical addition subsystem provides control against corrosion. Chemical addition to the system is manually initiated.

The head tank is vented to the atmosphere, and is located at the highest elevation in the system.

9.2.10.2.2.2 System Operation One of the two CSCWS is designed to be in operation during all modes of plant operation. For normal operation, hand switches for both chillers are placed in the normal after start or normal after stop position. When the lead chilled water pump is started from the control room, the corresponding chiller is normally started after all the safety requirements on the chiller unit are satisfied, and the chilled water flow is established .

The closed-loop refrigerant system in the chiller unit extracts heat from water in the evaporator, and rejects heat to service water in the condenser. A temperature sensor in the chilled water outlet positions the inlet guide vanes of the compressor to maintain a constant outlet water temperature.

The CSCWS is designed to perform its safety function assuming a single active failure. If there is a failure of any of the operating cooling coils or unit coolers except the main control room HVAC supply units, the standby unit starts and causes the standby chiller to start without interrupting the operation of the lead chiller. If the operating cooling coil for the main control room HVAC supply fails, the standby unit is manually started when required. If the lead chiller fails, the standby chiller is automatically activated by the standby fan cabinet when the lead fan cabinet fails to maintain the required temperature. The operating and standby chilled water loops can be interconnected by operating the gate valves in their interconnecting lines. The lead chiller continues to operate whenever the standby chiller is started automatically.

Chilled water flow through cooling coils and unit coolers is controlled by temperature-actuated valves as shown in drawing M-90.

9.2.10.2.3 Safety Evaluation CHAPTER 09 9.2-68 REV. 20, SEPTEMBER 2020

LGS UFSAR All safety-related components of the CSCWS are designed to seismic Category I requirements, as defined in Section 3.7. The safety-related components of the CSCWS are located in the seismic Category I designed control structure as discussed in Section 3.8. Power is supplied to these seismic Category I components from the Class 1E power sources as discussed in Section 8.3.

Piping, valve, or equipment failures (including supporting structures) that could result in the generation of a missile, in flooding, or in a pipe whip that could cause damage to safety-related equipment are discussed in Sections 3.5 and 3.6.

Two independent 100% capacity systems provide complete mechanical redundancy. Coupled with the redundancy of electrical design, a failure of any single active component cannot result in a complete loss of both CSCWS, thus assuring a safe shutdown condition. For a FMEA of the CSCWS, refer to Table 9.2-25.

9.2.10.2.4 Tests and Inspections The CSCWS is preoperationally tested in accordance with the requirements of Chapter 14, and periodically tested in accordance with the requirements of Chapter 16.

9.2.10.2.5 Instrumentation Applications Chillers and pumps are operated from the control room. Failure of the lead chiller or pump causes an alarm in the control room. Pressures at pump suction, pump discharge, and chiller discharge are displayed locally. Chilled water temperatures and flows, and motor electrical currents, are indicated in the control room.

Level in the head tank is indicated locally and alarmed through a computer in the control room at low level.

9.2.11 REFERENCES 9.2-1 E. Rabin and D. Meyers, "Selection of Design Meteorology for Safety-Related Spray Pond Systems," Transactions of American Nuclear Society, 22, p. 508 (November 1975).

9.2-2 W.E. Ranz and W.R. Marshall, "Evaporation from Drops," Chemical Engineering Progress, 48 (3 and 4) (March 1952 and April 1952).

9.2-3 V.E. Schrock and G.J. Trezek, "Rancho Seco Nuclear Service Spray Ponds Performance Evaluation", Report WHM-4, University of California, Berkeley, for Sacramento Municipal Utility District (July 1, 1973).

9.2-4 NRC, BTP APCSB 9-2, "Residual Decay Energy for Light- Water Reactors for Long-Term Cooling," Attachment to NRC Standard Review Plan for Ultimate Heat Sink, (April 1975).

9.2-5 P.J. Ryan and D.R.F. Harleman, "An Analytical and Experimental Study of Transient Cooling Pond Behavior," Ralph M. Parsons Laboratory for Water Resources and Hydrodynamics, Report 161, MIT Press, January 1973.

CHAPTER 09 9.2-69 REV. 20, SEPTEMBER 2020

LGS UFSAR 9.2-6 W.M. Kays and A.L. London, "Compact Heat Exchangers", 2nd ed., McGraw-Hill, New York (1964).

9.2-7 F. Kreith, "Principles of Heat Transfer", 2nd ed., International Textbook Company (1958).

9.2-8 Vanderbilt University, "Effect of Geographical Location on Cooling Pond Requirements and Performance", EPA Water Quality Office, Project No. 16130 FDQ (March 1971).

9.2-9 United States Nuclear Regulatory Commission, "Ultimate Heat Sink for Nuclear Power Plants," Regulatory Guide 1.27.

9.2-10 R. Codell, "Analysis of Ultimate-Heat-Sink Spray Ponds," NUREG-0733, Division of Engineering, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C., August 1981, p. 4-1.

9.2-11 G. H. Jirka, G. Abraham, and D. R. F. Harleman, "An Assessment of Techniques of Hydrothermal Prediction," Technical Report No. 203, R. M. Parsons Laboratory for Water Resources and Hydrodynamics, Department of Civil Engineering, Massachusetts Institute of Technology, Cambridge, MA, July 1975; (NUREG-0044, March 1976) .

9.2-12 Edward L. Thackston and Frank L. Parker, "Effect of Geographical Location of Cooling Pond Requirements and Performance," Department of Environmental and Water Resources Engineering, Vanderbilt University, Nashville, TN, Project No.

16130 FDQ, U. S. Environmental Protection Agency, Water Quality Office, March 1971, pp 7-10.

9.2-13 Ibid.

9.2-14 E. R. Anderson, "Energy-Budget Studies, Water Loss Investigations: Lake Hefner Studies," Professional Paper 269, U.S. Geological Survey, Washington D. C., 1954.

9.2-15 Thackston and Parker, loc. cit., pp. 11-12.

9.2-16 Thackston and Parker, loc. cit., p. 14.

9.2-17 Patrick J. Ryan and Donald R. F. Harleman, "An Analytical and Experimental Study of Transient Cooling Pond Behavior," DSR 73304, DSR 80004, DSR 80317, Ralph H. Parsons Laboratory for Water Resources and Hydrodynamics, Dept. of Civil Engr., Massachusetts Institute of Technology, Report No. 161, Prepared under the support of Hydrologic Engineering Center, U. S. Corps of Engineers (Contract No.

DACM05-71-0113), Engineering Energetics Progrwn, National Science Foundation (Grant No. GK-32472), and Duke Power Company, January 1973, p. 59, Eq (2.36).

9.2-18 D. K. Brady, W. L. Graves, and J. C. Geyer, "Surface Heat Exchange at Power Plant Cooling Lakes," Cooling Water Discharge Project Report No. 5, Prepared for Edison Electric Institute Research Project No 49, Department of Geography and Environmental Engineering, John Hopkins University, EEI Publication 69-401, New York, NY, November 1969.

CHAPTER 09 9.2-70 REV. 20, SEPTEMBER 2020

LGS UFSAR 9.2-19 R. Codell, Comparison Between Field Data and Ultimate Heat Sink Cooling Pond and Spray Pond Models," NUREG-0858, U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington D.C., September 1982.

9.2-20 Thackston and Parker, loc. cit., p. 14.

9.2-21 R. Codell and W. K. Nuttle, "Analysis of Ultimate Heat Sink Cooling Ponds,"

NUREG-0693, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission, Washington D. C., November 1980, p. 7.

9.2-22 R. B. Bird, W. E. Stewart, and E. N. Lightfoot, "Transport Phenomena," John Wiley

& Sons, inc., New York, 1960, p. 731, et seq.

9.2-23 Grant R. Fowles, "Analytical Dynamics," Holt, Rinehart, and Winston, Inc., New York, 1962, p. 151, et seq.

9.2-24 Bird, Stewart, and Lightfoot, loc. cit., pp. 190-196.

9.2-25 R. B. Bird, W. E. Stewart, and E. N. Lightfoot, "Transport Phenomena," John Wiley

& Sons, inc., New York, 1960, p. 647.

9.2-26 Bird, Stewart, and Lightfoot, loc. cit., pp. 648-649.

9.2-27 T. H. Chilton and A. P. Colburn, "Ind. Eng. Chem.," 26, 1183 (1934).

9.2-28 Grant R. Fowles, "Analytical Dynamics," Holt, Rinehart, and Winston, Inc., New York, 1962, p. 151, et seq.

9.2-29 Letter from Pamela B. Cowan (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request, Proposed Technical Specification Allowed Outage Time Extensions to Support Residual Heat Removal Service Water Maintenance," dated October 29, 2010.

9.2-30 Letter from Pamela B. Cowan (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "License Amendment Request Response to Additional Questions, Proposed Technical Specification Allowed Outage Time Extensions to Support Residual Heat Removal Service Water Maintenance," dated December 3, 2010.

9.2-31 Letter from David P. Helker (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request, Proposed Technical Specification Allowed Outage Time Extensions to Support Residual Heat Removal Service Water Maintenance," dated March 23, 2011.

9.2-32 Letter from Peter Bamford, U.S. Nuclear Regulatory Commission, to Michael J.

Pacilio, Exelon Nuclear, "Limerick Generating Station, Units 1 and 2 - Issuance of Amendments RE: Allowed Outage Time Extensions to Support Residual Heat Removal Service Water Maintenance (TAC Nos. ME3551 and ME3552)," dated July 29, 2011.

CHAPTER 09 9.2-71 REV. 20, SEPTEMBER 2020

LGS UFSAR Table 9.2-1 SERVICE WATER SYSTEM DESIGN PARAMETERS SERVICE WATER PUMPS Quantity 3 (50% capacity each)

Type Horizontal, centrifugal Capacity 18,000 gpm Head 250 feet Motor power rating 1500 hp FUEL POOL SERVICE WATER BOOSTER PUMPS Quantity 3 (33-1/3% capacity each)

Type Horizontal, centrifugal, single-stage Capacity 1300 gpm Head 60 feet Motor power rating 25 hp CHAPTER 09 9.2-72 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-2 EMERGENCY SERVICE WATER SYSTEM DESIGN PARAMETERS PUMPS Quantity 4 (50% capacity each)

Fluid Spray pond water Type Wet pit turbine Design flow, each 6400(1) gpm Total dynamic head 240(1) feet Material Casing Carbon steel Shaft Stainless steel Impeller Aluminum bronze MOTORS Rated power 500 hp Voltage/phase/cycle 4000/3/60 Speed 1200 rpm Service factor 1.0 (1) Design parameters listed in Table 9.2.2 do not constitute performance requirements. ESW pump performance requirements are based on its ability to support the heat transfer rates listed in Table 9.2-3 regardless of discharge head.

The pump design flow and total dynamic head values in this table were the basis for selecting the pumps; however, there is no licensing basis requirement for these particular design parameter values. Startup Testing and Surveillance Testing has demonstrated that the ESW System fulfills its required safety functions with the pumps delivering less flow and developing less total dynamic head. The pump design flow/head values in this table provide considerable margin from the system licensing design basis performance requirement to deliver the flow rates in Table 9.2-3.

CHAPTER 09 9.2-73 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-3 EMERGENCY SERVICE WATER SYSTEM DESIGN FLOWS AND HEAT TRANSFER RATES COMPONENT FLOW RATE in GPM (3) HEAT TRANSFER RATE (Quantity per ESW Loop) Loop A Loop B in 106 Btu/hr Standby Diesel Generator Heat 1800 1800 32.232 for four D/Gs Exchangers (4)

RHR Pump Motor Oil Cooler (4) 20.4 20.4 0.040 per pump RHR Pump Room Unit Coolers 184 (5) 184 (5) 0.548 for A/C room (8 total, 4 each room, 1 room each Unit) 0.520 for B/D room Core Spray Pump Room Unit Coolers 216 (6) 216 (6) 0.226 per room (8 total, 2 each room, 2 rooms each Unit)

HPCI Pump Room Unit Coolers (4 total, N/A 400 (4) 0.726 per room (4) 2 each room, 1 room each Unit)

RCIC Pump Room Unit Coolers (4 total, 160 (4) N/A 0.303 per room (4) 2 each room, 1 room each Unit)

Main Control Room Chiller (1) 450 450 2.873 per ESW Loop RECW Heat Exchanger (1) 1166 (1) 1166 (1) 6.41 (1) for one HX TECW Heat Exchanger (1) 243 (1) 243 (1) 0.3402 (1) for one HX Reactor Recirculation Pump seal cooler 260 (1) 260 (1) 0.628 (1) for two pumps and motor oil cooler (2 sets total, 1 set each Unit/Pump)

Makeup to fuel pools 60 60 N/A TOTAL 4399.4 4399.4 NOTES:

(1) Non-essential service (2) Footnote 2 deleted (3) Flow rates required for each heat exchanger and room cooler to remove design heat loads will vary with heat exchanger/room cooler fouling. Acceptance criteria are administratively controlled by plant surveillance procedures.

(4) The minimum required flow for these coolers is zero. However, the coolers are still normally connected to the ESW system. Therefore all the essential components must receive their minimum flow rates with these coolers connected to ESW and receiving some flow. The flow and heat transfer rates shown are those used to calculate the maximum heat that could be removed from the HPCI and RCIC pump rooms to be used when analyzing the Ultimate heat sink. The Ultimate heat sink must be capable of dissipating the heat that will be removed from these rooms without adversely impacting its required function to support other essential components.

(5) RHR room - 23 gpm per cooler; the flow rates for both ESW loops are based on the higher A/C RHR room heat load and assumes two coolers per room in service (one blocked out of service and the other one fails - ESW valves does not open). From an ESW flow balancing standpoint, this scenario is the design basis case. From a Spray Pond performance standpoint, a different scenario is the design basis case. For the Spray Pond design basis case, the total heat load from the accident unit RHR rooms is 1.955 MBtu/hr.

(6) Core Spray room - 27 gpm per cooler CHAPTER 09 9.2-74 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 9.2-4 EMERGENCY SERVICE WATER SYSTEM FAILURE MODES AND EFFECTS ANALYSIS PLANT OPERATING SYSTEM COMPONENT FAILURE EFFECT OF FAILURE FAILURE MODE EFFECT OF FAILURE MODE COMPONENT MODE ON SYSTEM DETECTION ON PLANT OPERATION Accident ESW pump Loss of ESW 'A' pump ESW loop 'A' flow reduced Pump discharge low ESW pump 'C' running; (typical for 'B', 'C', & 'D' pressure alarm control adequate cooling flow pumps) room maintained to ESW loop 'A' diesels and components(1)

Accident Instrument air supply Loss of air pressure to air System air operated valves Alarm in control room AOVs fail to safe position; pressure to air operated become inoperable cooler inlet valves fail open; valves isolation valves fail shut; ESW supply valves fail open. System cooling to essential components not impaired.

Accident Power Supply LOOP Normal power supply lost to Alarm in control room Standby diesel generators start system pumps and valves automatically, supplying power to all system components.

ESW pumps to start automatically.

Accident Power Supply Failure of Dll Class 1E bus Loss of ESW 'A' pump Alarm in control room (typical for D12, D23, & D24) system valves powered from for D13 & D14 Unit 1 D11 (Section 8.3) operation/ Unit 2 Same as for loss of ESW 'A' construction) pump above.

Accident Power Supply Failure of D13 Class 1E bus Loss of power to system Alarm in control room MOVs fail as-is. AOVs fail to (typical for D14, D21 & D22) valves powered from D13 safe condition as described for (Section 8.3) loss of instrument air above.

System components not impaired.

CHAPTER 09 9.2-75 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-4 (Cont'd)

PLANT OPERATING SYSTEM COMPONENT FAILURE EFFECT OF FAILURE FAILURE MODE EFFECT OF FAILURE MODE COMPONENT MODE ON SYSTEM DETECTION ON PLANT OPERATION Accident Controls Short-circuit in controls Valve fails in unsafe Valve position indication. Case 1: Cooler inlet valve causes valve to fail in unsafe position Coolers & chillers - trouble closure - redundant cooler condition (pump failure alarm. Diesel generators - automatically starts operation discussed above) high temperature alarm. when standby fan unit(s) start.

ESW return line-differential No effect on plant operation.

flow alarm. When the standby Core Spray, HPCI or RCIC cooler is not in service, an eventual loss of the associated pump will occur.

When a standby RHR cooler is not in service, eventual loss of both RHR pumps in the room may occur. No loss of safety function will occur since the ECCS pumps in separate rooms are redundant and HPCI/RCIC are needed for only 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or less.

Case 2: Diesel generator inlet/outlet valves closure - loss of diesel generator - see the effects above.

Case 3: Loop return valve closure - valve to opposite combined RHRSW/ESW loop remains open, no loss of flow.

Case 4: Intertie to nonessential system (service water and TECW) opens - redundant motor-operated or check valve remains shut, no loss of system integrity. Isolation valves on connections to RECW are remote manual (Section 6.2.4).

Case 5: Chiller control valve or outlet valve closure-redundant chiller operates or redundant outlet valve remains open. No effect on plant operation.

(1)

During the first ten minutes of a postulated LOCA, 3 RHR pumps, and therefore 3 diesels and their associated Class 1E busses may be required for the affected unit (Table 8.3-2). Both ESW pumps in a loop automatically start and would be operating during this time.

CHAPTER 09 9.2-76 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-5 RESIDUAL HEAT REMOVAL SERVICE WATER SYSTEM DESIGN PARAMETERS PUMPS Quantity 4(1)

Fluid Spray pond water Type Wet pit turbine Design flow, each 9000(2) gpm Total dynamic head 240(2) feet Material Casing Carbon steel Shaft Stainless steel Impeller Aluminum bronze MOTORS Power rating 700 hp Voltage/phase/cycle 4000/3/60 Speed 1200 rpm Service factor 1.0 (1) One pump supplies 100% flow to one RHR heat exchanger. During two-unit operation, two heat exchangers (one per unit), and therefore two of the four pumps, are required for safe shutdown and accident mitigation. For two-unit and one ESW/RHRSW loop operation, the RHR heat exchanger of the unit under normal shutdown will receive 67% to 100% RHRSW flow and the unit with the LOCA will receive 100% RHRSW flow.

(2) Design parameters listed in Table 9.2-5 do not constitute performance requirements. RHRSW pump performance requirements are based on the pumps ability to support heat transfer regardless of discharge head. The pump design flow and total dynamic head values in this table were the basis for selecting the pumps; however, there is no licensing basis requirement for these particular design parameter values. Startup Testing and Surveillance Testing has demonstrated that the RHRSW System fulfills its required safety functions with the pumps delivering less flow and developing less total dynamic head. The pump design flow/head values in this table provide considerable margin from the system licensing design basis performance requirement to deliver at least 5570 gpm to an RHR heat exchanger on the unit undergoing emergency shutdown and at least 8000 gpm to an RHR heat exchanger on the unit experiencing an accident (see Section 9.2.3.2).

CHAPTER 09 9.2-77 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 9.2-6 RHR SERVICE WATER SYSTEM FAILURE MODES AND EFFECTS ANALYSIS PLANT OPERATING SYSTEM COMPONENT FAILURE EFFECT OF FAILURE FAILURE MODE EFFECT OF FAILURE MODE COMPONENT MODE ON SYSTEM DETECTION ON PLANT OPERATION Normal/Accident RHRSW pump Failure of one RHRSW Loss of cooling water to one Pump discharge low pressure alarm Unavailability of one RHR heat pump heat exchanger RHR in control room RHRSW pumps supply sufficient exchanger. Three remaining cooling to the other RHR heat exchangers of both units for safe shutdown, cooldown, and accident mitigation.

Accident Power supply LOOP supply Loss of normal power to all Alarm in control room Plant is tripped. Standby diesel generators system components automatically start to supply power to all system components.

Accident Power supply Failure of D11 safeguard Loss of power to RHRSW 'A' Alarm in control room Case I: System initially aligned to spray bus (typical for D12) pump and system valves pond mode; loss of cooling to one RHR powered from D11 (refer to heat exchanger, remaining RHRSW Section 8.3) pumps provide sufficient cooling water to other RHR heat exchangers. Valves remain in original position, failing as-is, thereby having no effect on system operation.

Case II: System initially aligned to cooling tower mode; Loss of cooling to one RHR heat exchanger loss of 'A' spray network and loop 'A' winter bypass. Cooling towers isolated by redundant valves when system is automatically aligned to spray pond mode.

CHAPTER 09 9.2-78 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-6 (Cont'd)

PLANT OPERATING SYSTEM COMPONENT FAILURE EFFECT OF FAILURE FAILURE MODE EFFECT OF FAILURE MODE COMPONENT MODE ON SYSTEM DETECTION ON PLANT OPERATION Remaining spray networks or winter bypass, and RHRSW pumps provide adequate cooling for shutdown, cooldown, and accident mitigation.

Accident Power supply Failure of D23 safeguard Loss of power to system Alarm in control room Case I: System initially aligned to spray bus (typical for D13 during valves powered from D23 pond mode; valves remain in original Unit 1 only operation) (Section 8.3) position, failing as-is thereby having no effect on system operation.

Case II: System initially aligned to cooling tower mode; loss of 'C' spray network and loop 'B' winter bypass. Unit 1 cooling tower isolated by redundant valves when system aligned to spray pond mode.

Remaining networks and winter bypass provide sufficient cooling.

Accident Power supply Failure of D24 safeguard Loss of power to system Alarm in control room Case I: System initially aligned to spray bus (typical for D14 during valves powered from D24 pond mode; valves remain in original Unit 1 only operation) (Section 8.3) position, failing as-is, thereby having no effect on system operation.

Case II: System initially aligned to cooling tower mode; loss of 'D' spray network and loop 'B' winter bypass. Unit 2 cooling tower isolated by redundant valves when system aligned to spray pond mode.

Remaining networks and winter bypass provide sufficient cooling.

CHAPTER 09 9.2-79 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-6 (Cont'd)

PLANT OPERATING SYSTEM COMPONENT FAILURE EFFECT OF FAILURE FAILURE MODE EFFECT OF FAILURE MODE COMPONENT MODE ON SYSTEM DETECTION ON PLANT OPERATION Accident Power supply Loss of D11, D13, D21, and Loss of RHRSW pumps 'A' Alarm in control room Loss of cooling to one RHR heat D23 safeguard buses and 'C' and system valves exchanger in each unit. Valves fail as-is (typical for loss of D12, D14, powered from D11, D13, D21 and thereby do not affect system D22, and D24 safeguard and D23 operation. Two RHRSW pumps are buses) (Section 8.3) sufficient for safe shutdown, cooldown, and accident mitigation.

Accident Controls Short-circuit in controls Valve fails in unsafe Valve position indication, flow Case 1: RHR Heat exchanger inlet/outlet causes valve to fail in unsafe condition temperature, and pressure valve closure - loss of heat exchanger.

condition (pump failure indication for RHR heat exchanger Redundant heat exchanger provides discussed above) valves sufficient cooling.

Case 2: Spray network return valve closure - loss of one spray network.

Remaining 3 networks provide sufficient cooling.

Case 3: Winter bypass valve closure -

loss of one RHRSW loop (until spray operation begins). Remaining RHRSW loop supplies sufficient cooling to both units. ESW system unaffected since it returns to both RHRSW return loops.

Case 4: Spray pond sluice gate closure -

remaining sluice gate and wet pit intertie provide flow path into wet pit. No effect on operation.

CHAPTER 09 9.2-80 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-6 (Cont'd)

PLANT OPERATING SYSTEM COMPONENT FAILURE EFFECT OF FAILURE FAILURE MODE EFFECT OF FAILURE MODE COMPONENT MODE ON SYSTEM DETECTION ON PLANT OPERATION Case 5: Cooling tower supply and return or TECW return valve opens - redundant MOV or check valve prevents loss of flow.

No effect on operation.

Normal/Accident Radiation monitors Radiation monitor fails low Loss of radiation detection in High radiation alarm in control Monitoring of Loop return flows continues RHRSW return loop. room by taking grab samples on a periodic RHRSW pumps basis in accordance with the off-site dose automatically tripped Calculation manual. Pumps can be returned to service with manual override.

Normal/Accident Radiation monitors Radiation monitor fails high Loss of radiation detection in High radiation alarm in control Monitoring of Loop return flows continues RHRSW return loop room by taking grab samples on a periodic RHRSW pumps basis in accordance with the off-site dose automatically tripped. Calculation manual. Pumps can be returned to service with manual override.

CHAPTER 09 9.2-81 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-7 CLARIFIED AND DOMESTIC WATER SYSTEMS DESIGN PARAMETERS RAW WATER CLARIFIER Quantity 1 Type Circular, upflow, solid contact, with internal variable-speed agitator Capacity 310 gpm Motor power rating 3/4 hp CLEARWELL Quantity 1 Type Vertical, cylindrical Capacity 5000 gallons CLARIFIED WATER SERVICE PUMP CLARIFIED WATER STANDBY PUMP FILTER BACKWASH PUMP Quantity, each 1 Type Horizontal, centrifugal, single-phase Capacity, each 300 gpm Head 90 feet Motor power rating 15 hp PRESSURE FILTERS Quantity 3 (33-1/3% capacity each)

Type Vertical, pressure, anthracite bed Capacity 100 gpm Design pressure 100 psig CHAPTER 09 9.2-82 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 9.2-7 (Contd)

CLARIFIED WATER STORAGE TANK Quantity 1 Type Vertical, cylindrical Capacity 200,000 gallons MAKEUP DEMINERALIZER FEED PUMPS Quantity 2 (100% capacity each)

Type Horizontal, centrifugal, single-stage Capacity 90 gpm Head 230 feet Motor power rating 15 hp LUBE WATER PUMPS Quantity 2 (100% capacity each)

Type Horizontal, centrifugal, single-stage Capacity 120 gpm Head 300 feet Motor power rating 25 hp CHAPTER 09 9.2-83 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 9.2-8 DEMINERALIZED WATER MAKEUP SYSTEM DESIGN PARAMETERS CARBON FILTER Quantity 1 Type Vertical, pressurized, activated carbon bed Flow rate 120 gpm Design pressure 125 psig DEMINERALIZED WATER STORAGE TANK Quantity 1 Type Vertical, cylindrical Capacity 50,000 gallons DEMINERALIZED WATER TRANSFER PUMPS Quantity 2 (100% capacity each)

Type Horizontal, centrifugal, double suction, single stage Capacity 150 gpm Head 350 feet Motor power rating 40 hp DEMINERALIZED WATER TRANSFER JOCKEY PUMP Quantity 1 Type Horizontal, centrifugal, single stage Capacity 60 gpm Head 350 feet Motor power rating 20 hp CHAPTER 09 9.2-84 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-9 SPRAY POND DESIGN DATA SPRAY NOZZLES Number 240 per spray network (4 networks)

Type Hollow-cone spray pattern (Spray Engineering Co. ramp bottom model 1751A)

Capacity, each 51.0 - 64.6 gpm SPRAY POND Dimensions at pond bottom Length 1000 feet Width 400 feet Depth 9'10" (at TECH SPEC minimum water level of 250'10" MSL)

Surface area 9.9 acres Storage volume 29.06x106 gallons Lining Soil-Bentonite mixture Network flow rate 12,250 - 15,500 gpm CHAPTER 09 9.2-85 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-10 PERCENT FREQUENCY DISTRIBUTION OF DAILY AVERAGE RELATIVE HUMIDITY AND WIND SPEED LGS PHILADELPHIA PHILADELPHIA 21/2 YEARS 21/2 YEARS 34 YEARS Relative Humidity

(%)90-100 12.3 7.9 6.3 80-89 17.7 17.3 15.7 70-79 29.4 22.9 24.7 60-69 20.1 23.7 26.2 50-59 14.7 17.5 18.5

<50 5.8 10.7 8.6 Wind Speed (mph) 0-2 10.8 0.0 0.2 3-4 31.1 0.2 3.4 5-6 25.3 10.1 16.0 7-8 14.5 25.7 25.3 9-10 8.7 24.6 21.3 11-15 7.5 31.2 26.6 16-20 1.8 7.4 6.0

>20 0.4 0.9 1.1 CHAPTER 09 9.2-86 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-11 SPRAY POND SYSTEM DESIGN METEOROLOGY WORST WORST WORST 30 DAYS FOR DAY FOR HEAT 30 DAYS FOR WATER LOSS TRANSFER HEAT TRANSFER Average dry-bulb temperature, oF 67.5 85.6 78.1 Average wet-bulb temperature, oF 59.2 78.6 73.4 Average natural wind speed, mph 10.43 0.0 0.0 Period 05/30/58 to 6/30/45 8/12/59 to 06/28/58 09/10/59 CHAPTER 09 9.2-87 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-12 COEFFICIENT OF THERMAL PERFORMANCE COMPARISON OF LGS AND PHILADELPHIA WORST CASE METEOROLOGICAL CONDITIONS (JANUARY 1, 1972 - DECEMBER 31, 1976)

WORST 30 DAYS WORST DAY FOR WORST 30 DAYS FOR WATER LOSS HEAT TRANSFER FOR HEAT TRANSFER LGS PHILADELPHIA LGS PHILADELPHIA LGS PHILADELPHIA Average dry-bulb 72.0 75.1 80.7 84.2 75.3 78.1 temperature (F)

Average wet-bulb 64.8 65.8 74.6 78.3 70.4 70.9 temperature (F)

Average natural wind 5.6 9.7 2.3 5.7 4.1 7.1 speed at 30 ft elevation (mph)

Period 07/22/72 to 06/24 to 07/19/72 07/19/72 07/07 to 07/07 to 08/20/72 07/23/74 08/05/75 08/05/75 Coefficient 0.0135 0.0147 -12.22 -11.92 -14.28 -15.62 CHAPTER 09 9.2-88 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-13

SUMMARY

OF SPRAY POND WATER ALLOWANCES VOLUME (millions of gallons)

Allowances Final Design Water Losses (max heat transfer case )

Evaporation, plant heat 13.58 load Evaporation, natural 1.82 Drift loss .14 Seepage 1.83 Total Water Loss 17.37 Nonwater Loss Sedimentation 2.75 Excess for minimum 4.20 operating level (el. 243-6) _____

Total Nonwater Loss 6.95 Total Required 24.33 Reserve for Water Loss 4.74 Total Water Pond Capacity 29.07 CHAPTER 09 9.2-89 REV. 19, SEPTEMBER 2018

LGS UFSAR Table 9.2-14 DATA USED IN PERFORMANCE ANALYSES A. INITIAL CONDITIONS (10 minutes after LOCA)

LOCA UNIT SHUTDOWN UNIT Reactor Vessel Temperature, oF 250 550 Water mass, lbm 550,000 700,000 RPV heat capacity, metal, 3.53x105 3.53x105 Btu/oF Water on Drywell Floor Temperature, oF 180 -

Water mass, lbm 500,000 -

Suppression Pool Temperature, oF 150 104.6 Water mass, lbm 7.01x106 7.36x106 Spray Pond Temperature, oF 88 B. PHYSICAL PARAMETERS Mass median drop size 3000 microns Included spray angle 76 degrees Pressure drop through nozzle 7 psi @ 51 gpm Height of spray above nozzle 7 ft @ 51 gpm Min. heat transfer Max. water loss Flow/nozzle 35.4 gpm 64.6 gpm Nozzle height above pond 6.0 ft 7.17 ft CHAPTER 09 9.2-90 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 9.2-14 (Cont'd)

C. FLOW RATES (gpm)(3)

SHUTDOWN TIME PARAMETER LOCA UNIT UNIT COMMON (1) (1) 10-30 min Containment spray 0 (1) (1)

Core spray 7,430 (1) (2) (1)

Relief valves & RCIC (1)

RHR HX (shell side) 10,000 0 (1)

RHR HX (tube side) 9,000 0 (1) (1)

ESW 13,000 (1) (1)30-180 min Core spray 7,430 RHR HX (shell side)

(1) (1)

LPCI mode 10,000 (1)

RHR HX (tube side) 9,000 18,000 (1) (1)

ESW 13,000 (1) (1) 3-30 hr Core spray 7,430 (2) (1)

RHR HX (shell side) 10,000 (1)

RHR HX (tube side) 9,000 18,000 (1) (1)

ESW 13,000 (1) (1)30-720 hr Core spray 7,430 (1)

RHR HX (shell side) 10,000 10,000 (1)

RHR HX (tube side) 9,000 9,000 (1) (1)

ESW 13,000 (1)

Not applicable (2)

Flow rate adjusted to limit cooldown rate to 100oF/hr (3)

These flow rates are used to compute heat loads to the spray pond. The spray pond performance analysis is based on bounding flow rates of 17,000 gpm (Min. Heat Transfer) and 53,000 gpm (max. Water Loss Case).

CHAPTER 09 9.2-91 REV. 17, SEPTEMBER 2014

LGS UFSAR Table 9.2-15 HEAT REJECTION TO THE SPRAY POND (ONE-UNIT LOCA, ONE-UNIT SSD)

CHAPTER 09 9.2-92 REV. 19, SEPTEMBER 2018

LGS UFSAR Table 9.2-15 (cont.)

HEAT REJECTION TO THE SPRAY POND (ONE-UNIT LOCA, ONE-UNIT SSD)

CHAPTER 09 9.2-93 REV. 19, SEPTEMBER 2018

LGS UFSAR Table 9.2-16 INTEGRATED HEAT REJECTION TO THE SPRAY POND CHAPTER 09 9.2-94 REV. 19, SEPTEMBER 2018

LGS UFSAR Table 9.2-16 (cont.)

INTEGRATED HEAT REJECTION TO THE SPRAY POND CHAPTER 09 9.2-95 REV. 19, SEPTEMBER 2018

LGS UFSAR Table 9.2-17 SPRAY POND TEMPERATURE TRANSIENT (ONE-UNIT LOCA, ONE-UNIT SSD)

CHAPTER 09 9.2-96 REV. 19, SEPTEMBER 2018

LGS UFSAR Table 9.2-18 SIZE COMPARISON OF RANCHO SECO AND LGS SPRAY PONDS LGS RANCHO SECO Water flow rate (gpm) 49,000 (max) 16,500 Nozzles 960 @ 51.0 gpm 304 @ 53.6 gpm Water pressure at nozzle (psig) 7 7 Pond length at bottom (ft) 1000 330 Width at bottom (ft) 400 165 Depth (ft) 10 5 Volume (gal) 29.6x106 5.7x106 Area (acres) 9.9 1.3 CHAPTER 09 9.2-97 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-19 PERFORMANCE COMPARISON OF RANCHO SECO TEST RESULTS AND MODEL RESULTS Ambient Conditions  % Efficiency Twb Tdb Wind Speed Spray Rancho Model (oF) (oF) (mph) Temperature Seco 60.98 81.50 13.0 79.88 41.7 35.72 61.52 80.96 12.5 80.06 47.5 34.89 55.04 51.08 5.3 77.36 32.5 31.84 48.56 51.98 1.0 77.36 28.8 31.00 64.94 56.48 6.0 77.54 30.9 31.6 71.06 57.56 6.5 78.62 35.5 22.10 72.32 95.00 7.0 80.06 38.9 31.91 69.62 93.02 6.6 81.14 34.3 28.85 66.56 85.64 8.4 80.78 45.8 30.02 60.98 72.32 3.8 80.24 34.5 32.96 60.26 69.26 3.8 79.7 28.5 34.61 54.20 57.92 1.0 101.4 35.2 39.16 53.06 57.02 1.6 100.0 36.2 38.23 51.98 55.94 1.3 97.88 34.4 37.24 48.92 53.06 1.0 101.66 34.5 36.70 48.02 51.08 0.4 97.34 34.6 35.62 CHAPTER 09 9.2-98 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-20 CONDENSATE AND REFUELING WATER STORAGE FACILITIES DESIGN PARAMETERS CONDENSATE STORAGE TANK Quantity 1 Type Vertical, cylindrical, dome roof, with heating coil Capacity 200,000 gallons REFUELING WATER STORAGE TANK Quantity 1 Type Vertical, cylindrical, dome roof, with heating coil Capacity 550,000 gallons CONDENSATE TRANSFER PUMPS Quantity 2 (100% capacity each)

Type Horizontal, centrifugal, single-stage Capacity, each 600 gpm Head 250 feet Motor power rating 60 hp CONDENSATE TRANSFER JOCKEY PUMP Quantity 1 Type Horizontal, centrifugal, two-stage Capacity 70 gpm Head 250 feet Motor power rating 15 hp REFUELING WATER PUMPS Quantity 2 (100% capacity each)

Type Horizontal, centrifugal, single-stage Capacity, each 1500 gpm Head 180 feet Motor power rating 100 hp CHAPTER 09 9.2-99 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-21 REACTOR ENCLOSURE COOLING WATER SYSTEM DESIGN PARAMETERS RECW PUMPS Quantity 2 (100% capacity each)

Type Horizontal, centrifugal, single-stage Capacity 1500 gpm Head 170 feet Motor power rating 100 hp RECW HEAT EXCHANGERS Quantity 2 (100% capacity each)

Type Horizontal, 2 pass, 1 shell Duty 27x106 Btu/hr Shell design Fluid Demineralized water Flow rate 700,000 lb/hr Design pressure 175 psig Design temperature 200oF Tube design Fluid Service water Flow rate 1,500,000 lb/hr Design pressure 150 psig Design temperature 200oF RECW HEAD TANK Quantity 1 Type Vertical, cylindrical Capacity 790 gallons REACTOR ENCLOSURE CHEMICAL ADDITION TANK Quantity 1 Type Vertical, cylindrical Capacity 5 gallons CHAPTER 09 9.2-100 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-22 TURBINE ENCLOSURE COOLING WATER SYSTEM DESIGN PARAMETERS TECW PUMPS Quantity 2 (100% capacity each)

Type Horizontal, centrifugal, single-stage Capacity 225 gpm Head 140 feet Design temperature 100oF Motor power rating 15 hp TECW HEAT EXCHANGERS Quantity 2 (100% capacity each)

Type Horizontal, 2 pass, 1 shell Duty 2.5x106 Btu/hr Shell design Fluid Demineralized water Flow rate 110,000 lb/hr Design pressure 100 psig Design temperature 200oF Tube design Fluid Service water Flow rate 250,000 lb/hr Design pressure 150 psig Design temperature 200oF TECW HEAD TANK Quantity 1 Type Vertical, cylindrical Capacity 210 gallons TURBINE ENCLOSURE CHEMICAL ADDITION TANK Quantity 1 Type Vertical, cylindrical Capacity 5 gallons CHAPTER 09 9.2-101 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-22 (Cont'd)

CONDENSATE PUMP MOTOR UPPER BEARING COOLERS Quantity 3 (100% capacity each)

Type Coil Duty 76,350 Btu/hr Tube design Fluid Demineralized water Flow rate 6000 lb/hr Design pressure 100 psig Design temperature 100oF CONDENSATE PUMP MOTOR LOWER BEARING COOLERS Quantity 3 (100% capacity each)

Type Coil Duty 12,725 Btu/hr Tube design Fluid Demineralized water Flow rate 6000 lb/hr Design pressure 100 psig Design temperature 100oF CHAPTER 09 9.2-102 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-23 DRYWELL CHILLED WATER SYSTEM DESIGN PARAMETERS WATER CHILLERS Type Centrifugal Capacity 1500 tons Power rating 1303 kW Entering chilled water temperature 65oF Leaving chilled water temperature 50oF CHILLED WATER CIRCULATING PUMPS Type Centrifugal, horizontal, split case Flow rate 2400 gpm Head 110 feet Motor power rating 100 hp HEAD TANK Type Vertical, cylindrical Volume 300 gallons Pressure Atmospheric CHEMICAL FEED TANK Type Vertical, cylindrical Volume 5 gallons CHAPTER 09 9.2-103 REV. 13, SEPTEMBER 2006

LGS UFSAR Table 9.2-24 CONTROL STRUCTURE CHILLED WATER SYSTEM DESIGN PARAMETERS WATER CHILLERS (1)

Type Centrifugal Capacity 192 tons Power rating 289 kW Entering chilled water temperature 50.5oF Leaving chilled water temperature 44.0oF CHILLED WATER CIRCULATING PUMPS Type Centrifugal, vertical, in-line Flow rate 705 gpm Head 80 feet Motor power rating 25 hp HEAD TANK Type Vertical, cylindrical Volume 34 gallons Pressure Atmospheric CHEMICAL FEED TANK Type Vertical, cylindrical Volume 5 gallons (1) Values specified resulted from changes made to the condenser flow rate documented by ECR 12-00110. Originally, supplied water chiller was rated for 250 tons, 329 kW power rating and an entering chilled water temperature of 52.5 oF.

CHAPTER 09 9.2-104 REV. 16, SEPTEMBER 2012

LGS UFSAR Table 9.2-25 CONTROL STRUCTURE CHILLED WATER SYSTEM FAILURE MODES AND EFFECTS ANALYSIS COMPONENT EFFECT OF FAILURE FAILURE EFFECT OF FAILURE PLANT OPERATING MODE SYSTEM COMPONENT FAILURE MODE ON THE SYSTEM MODE DETECTION ON PLANT OPERATION

1. Normal Normal power Total LOOP All system components Alarm in the No loss of safety supply are automatically control room function supplied with power from diesel generators.

2. Accident Class 1E power Loss of Class 1E None. The two Alarm in the No loss of safety supply power supply to redundant systems control room function one loop are powered from separate Class 1E busses. Cooling water to the chiller condensers is supplied from redundant ESW loops.

3. Accident Water chillers Chiller failure The standby chilled Alarm in the No loss of safety (LOCA or LOCA & LOOP) water system is control room function automatically started from the control room. (Note 1 )

4. Accident Chilled water Pump failure The standby chilled Alarm in the No loss of safety (LOCA or LOCA & LOOP) circulating pump water pump is control room function automatically started from the control room.

5. Normal or accident System pressure Failure of Loss of one chilled Alarm in control No loss of safety boundary pressure boundary water loop. Standby room due to low function in one loop chiller and pump are level in head started manually tank from the control room. A completely redundant piping system is provided for the standby system.

Note 1 - The control structure chiller has several design feature protective trips to prevent chiller damage. If such a chiller trip occurs, the trip must be reset at the local panel before the chiller can be restarted from the control room. The local chiller control panel is located on elevation 200 in the control structure which is readily accessible from the main control room during accident conditions.

CHAPTER 09 9.2-105 REV. 13, SEPTEMBER 2006