ML20097E188

Proposed Tech Specs for Increase in Spent Fuel Pool Heat Loads

ML20097E188
Person / Time
Site:	South Texas
Issue date:	02/08/1996
From:	HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20097E183	List: ML20097E188 ST-HL-AE-5233, Application for Amends to Licenses NPF-76 & NPF-80,revising Spent Fuel Pool Heat Licensing Basis to Provide Greater Flexibility for Normal Refueling Practices
References
NUDOCS 9602130349
Download: ML20097E188 (60)

Text

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. ATTACHMENT 3 COVER SHEET The following Updated Final Safety Analysis Report pages are provided in support of this amendment. Proposed revisions are indicated as appropriate.

1 3.1-32* l 3.1-33 TC 9-1 (Table of Contents) .

TC 9-12 (List of Tables) 9.1-1* 9.2-10*

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• Pages with no changes shown are provided to support review of the proposed License Amendment.

l-9602130349 DR 960208 ADOCK 05000498 PDR MISC-96\$233.w

. STPEGS UFSAR intent of these regulations is to ensure that the levels of any radioactive material effluents in unrestricted areas is as low as reasonably achievable.

The Liquid Waste Processing System (LWPS) is designed to recycle as much process waste as can be accommodated within the plant water balance. All releases are monitored and controlled and the system has been designed to prevent accidental discharges. The principal source of gaseous effluents from the plant during normal operation is the. hydrogen continuously vented from the volume control tank. This gas is exhausted through an ambient temperature treatment system, including charcoal adsorbers, which removes radioiodines and particulates, to the plant main exhaust duct.

Solid wastes, including spent resins, filter sludges, filter cartridges, evaporator bottoms, and contaminated tools, equipment, and clothing, are collected, packaged and shipped offsite in approved shipping containers.

3.1.2.6.2 Criterion 61 - Fuel Storare and Handlinz and Radioactivity: The Fuel Storage and Handling System, Radioactive Waste System, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant ,

inventory under accident conditions.  ;

I 3.1.2.6.2.1 New Fuel Storaze - Evaluation Acainst Criterion 61 - New fuel is placed in dry storage in the new fuel storage vault which is located inside the Fuel-Handling Building (FHB). The storage vault within the FHB provides adequate shielding for radiation protection. Storage racks preclude accidental criticality (see " Evaluation Against Criterion 62"). The new fuel storage racks do not require any special inspection and testing for nuclear safety purposes.

3.1.2.6.2.2 Spent Fuel Handline and Storare - Evaluation Azainst Criterion 61 - Irradiated fuel is stored underwater in spent fuel storage racks located at the bottom of the spent fuel pool. Spent fuel pool water is circulated through the Spent Fuel Pool Cooling and Cleanup System (SFPCCS) to ,

maintain fuel pool water temperature, purity, water clarity, and water level. l The spent fuel storage racks preclude accidental criticality (see " Evaluation Against Criterion 62").

Reliable decay heat removal is provided by the closed-loop SFPCCS, which consists of two cooling trains, two purification trains, a surface skimmer loop, and required piping, valves and instrumentation. Water is drawn from the spent fuel pool by the spent fuel pool pumps, is pumped through the tube side of the heat exchangers and is returned. Each suction line, which is protected by a strainer, is located at an elevation 4 ft below the normal water level, while the return line terminates at an elevation 6 ft above the top of the fuel assemblies and contains an antisiphon hole near the surface of the water to prevent gravity drainage. The SFPCCS is designed to remove the amount of decay heat produced by the number of spent fuel assemblies that are stored following refueling. Each train is capable of removing 100 percent of Misc 96\5Diw 3.1-32 Revision 0

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. STPEGS UFSAR the normal maximum design heat load and 50 percent of the abnormal maximum design heat load. When the spent fuel assemblies resulting from a refueling are stored, the system can maintain the spent fuel cooling water temperature as stated in Table 9.1-1 when the heat exchangers are supplied with component cooling water at the design flow and temperature. If it is necessary to remove a complete core l from the reactor while the spent fuel assemblies from the previous refueling l still remain in the spent fuel pool, the system can maintain the spent fuel cooling water as stated in Table 9.1-1.

System piping is arranged so that failure of any pipeline cannot drain the spent fuel pool or the in-Containment temporary storage area below a depth of l approximately 23 ft of water over the top of the stored spent fuel assemblies. A minimum depth of approximately 10 ft of water over the top of the stored spent fuel assemblies is required to limit direct radiation to 2.5 mR/hr.

High- and low-level alarms in the control room are actuated upon pool water level changes. Fission product concentration in the pool water is minimized by use of the filters and demineralizers. This minimizes the fission product releases from the pool to the FHB environment.

No special tests are required because at least one pump, heat exchanger, filter, and demineralizer are continuously in operation while fuel is stored in the pool.

Duplicate units are operated periodically to handle abnormal heat loads or to replace a unit for servicing. Routine visual inspection of the system components, instrumentation and trouble alarms are adequate to verify system operability, t 3.1.2.6.2.3 Radioactive Waste Systems - Evaluation Azainst Criterion 61 -

The Radioactive Waste Systems provide all equipment necessary to collect, process and prepare for disposal all radioactive liquids, gases and solid waste produced as a result of operation.

The LWPS is divided into two sections: one section treats reactor-grade liquid which is recyclable after processing, and the other section treats non-reactor-grade water from inputs such as floor drains which is released from the plant after processing. Processing may include filtration, ion exchange, analysis, and evaporation. Spent resins are de-watered for disposal as solid radwaste. If conditions require, evaporator bottoms are processed for disposal as solid radwaste. Dry solid radwastes are packaged in steel drums or fiber drums, cartons or boxes. Gaseous radwastes are monitored, processed, recorded, and restricted so that radiation doses to members of the public in unrestricted areas are below those allowed by applicable regulations.

Routinely accessible portions of the FHB and Mechanical-Electrical Auxiliaries Building (MEAB) have sufficient shielding to maintain dose rates ALARA. See USAR, Section 12.3, for shielding design criteria. The MEAB and its associated systems are designed to preclude accidental release of radioactive materials to the environs.

l The Radwaste Systems are used on a routine basis and do not require specific l testing to assure operability. Performance is monitored by radiation monitors during operation.

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The Fuel Storage and Handling and Radioactive Waste Systems are designed to assure adequate safety under normal and postulated accident conditions.

husc-96u233w 3.1-33 Revision 4

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• STPEGS UFSAR TABLE OF CONTENTS CHAPTER 9 AUXILIARY SYSTEMS Section Title Eagg 9.1 FUEL STORAGE AND HANDLING 9.1-1 9.1.1 New Fuel Storage 9.1-1 9.1.1.1 Design Bases 9.1-1 9.1.1.2 Facilities Description 9.1-1 9.1.1.3 Safety Evaluation 9.1-2 9.1.2 Spent Fuel Storage 9.1-2 9.1.2.1 Design Bases 9.1-2 9.1.2.2 Facilities Description 9.1-4 9.1.2.3 Safety Evaluation 9.1-5 9.1.3 Spent Fuel Pool Cooling and Cleanup System 9.1-5 9.1.3.1 Design Bases 9.1-5 9.1.3.1.1 Spent Fuel Cooling 9.1-6 9.1.3.1.2 Dewatering Protection 9.1-6 9.1.3.1.3 Water Purification 9.1-6 9.1.3.2 System Description 9.1-6 9.1.3.2.1 Component Description 9.1-8 pj))2]Egl$$$$[{Jpp{Js@]fooM@MMi$liglSjf@ lim 10pp4Elosihj[ $91139 9.1.3.3 Safety Evaluation 9.1-10 9.1.3.3.1 Availability and Reliability 9.1-10 9.1.3.3.2 Spent Fuel Storage Area Dewatering 9.1-10 9.1.3.3.3 Water Quality 9.1-10 9.1.3.3.4 Spent Fuel Pool Boiling Dose Analysis 9.1-11 9.1.3.4 Instrumentation Application 9.1-11 9.1.3.4.1 Temperature 9.1-11 9.1.3.4.2 Pressure 9.1-11 9.1.3.4.3 Flow 9.1-11 9.1.3.4.4 Level 9.1-11 9.1.3.5 Tests and Inspections 9.1-12 9.1.4 Fuel Handling System (FHS) 9.1-12 9.1.4.1 Design Bases 9.1-12 9.1.4.2 System Description 9.1-12 9.1.4.2.1 New Fuel Handling 9.1-12 9.1.4.2.2 Refueling Procedure 9.1-13 9.1.4.2.2.1 Phase I - Preparation 9.1-14 9.1.4.2.2.2 Phase II - Reactor Disassembly 9.1-14 9.1.4.2.2.3 Phase III - Fuel Handling 9.1-15 9.1.4.2.2.4 Phase IV - Reactor Assembly 9.1-16 9.1.4.2.3 Spent Fuel Shipment 9.1-16 9.1.4.2.4 Component Description 9.1-17 9.1.4.2.4.1 Refueling Machine 9.1-17 9.1.4.2.4.2 Fuel Handling Machine 9.1-18 9.1.4.2.4.3 New Fuel Elevator 9.1-19 9.1.4.2.4.4 Fuel Transfer System 9.1-19 9.1.4.2.4.5 Rod Cluster Control Change Fixture 9.1-19 9.1.4.2.4.6 Spent Fuel Assembly Handling Tool 9.1 20 9.1.4.2.4.7 New Fuel Assembly Handling Tool 9.1-20 Misc-950233.w TC 9-1 Revision 3

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 .                                  STPEGS UFSAR LIST OF TABLES CHAPTER 9 Table       Title                                           fage 1

9.1-1 Spent Fuel Pool C ling :nd C1::nup Sy:t::, j D ign P:::::ter %tepjinalfsiidforQ81Montti Rsipg@ld 9.1-30 9.1-2 Spent Fuel Pool Cooling and Cleanup System, i Component Design Parameters 9.1-31 9.1-3 FHB 15/2-Ton Crane - Compliance with Regulatory Guide 1.104 9.1-35 9.1-4 NSSS Vendor Recommended Specifications and j Guidelines for Spent Fuel Pool Water Purity 9.1-37 ] 9.1-5 Spent Fuel Pool Cooling and Clean-Up System i Failure Modes and Effects Analysis 9.1-38 l 9.1-6 Spent Fuel Pool Boiling 9.1-39 ) 9.2.1 Essential Cooling Water System Equipment Data 9.2-36 9.2.1-2 Essential Cooling Water System, Failure Modes and Effects Analysis 9.2-38 9.2.1-3 Evaluation of Corrosive / Scale-Forming Tendencies of the Essential Cooling Pond 9.2-43 , 9.2.1-4 ECW Relief Valves to which ASME ' B&PV Code Case N-242-1 Applies 9.2-44 I 9.2.2-1 Component Cooling Water System Design l Parameters 9.2-45 9.2.2-2 CCW Water Chemistry, Normal Plant Operation 9.2-48 9.2.2-3 Component Cooling Water System, Failure Modes and Effects Analysis 9.2-49 ) 9.2.2-4 Component Cooling Water Requirements 9.2-63  ! 9.2.2-5 CCWS Relief Valves to Which ASME B&PV Code Case N-242-1 Applies 9.2-65 ) 9.2.5-1 Maximum Heat Rejected to the Essential Cooling Water System for the Safe Shutdown Unit 9.2-67 9.2.5-2 Maximum Heat Rejected to the Essential Cooling Water System for the LOCA Unit 9.2-68 9.2.5-3 Maximum Consumptive Use of Water in the Essential Cooling Pond for a Safe Shutdown in One Unit and LOCA in the other Unit 9.2-69 9.2.5-3.1 Maximum Consumptive Use of Water in the

Essential Cooling Pond for the Safe Shutdown l of Two Units 9.2-70 Combinations of Phenomena Considered 9.2.5-4 in the Design of the Ultimate Heat Sink 9.2-71 9.2.5-5 Summary of ECP Performance ,

l During Minimum Cooling 9.2-72 9.2.7-1 Reactor Makeup Water System Component Design Parameters 9.2-73 9.2.7.2 Reactor Makeup Water System Failure Modes and Effects Analysis 9.2-74 9.3-1 Gas Vessels - Design Data 9.3-64 Misc-ms233* TC 9-12 Revie on 3 l

STPEGS UFSAR l 9.1 FUEL STORAGE AND HANDLING l Facilities for the receipt and storage of new fuel and the storage and I A transfer of spent fuel are housed in the Fuel Handling Building (FHB). separate and independent FHB is provided for each unit of the South Texas Project Electric Generating Station (STPEGS). Each FHB is designed as a i l controlled-leakage seismic Category I structure. The design of the FHB Heating, Ventilating and Air-Conditioning (NVAC) System is discussed in j Section 9.4.2. The structural design considerations are described in Section 3.8.4. 9.1.1 New Fuel Storage 9.1.1.1 Desien Bases. The new fuel storage pit is a reinforced concrete pit and an integral part of each seismic Category I FHB. This pit provides temporary dry storage for approximately one-third of a core (66 fuel assemblies) of new fuel. The fuel is stored in racks (Figure 9.1.1-1) composed of individual vertical cells fastened together to form three 2 x 11 modules which may be bolted to anchors in the floor and walls of the new fuel storage pit. The new fuel racks are classified as seismic Category I components, as defined by Regulatory Guide (RG) 1.29, and American Nuclear Society (ANS) safety class (SC) 3 (Section 3.2). The new fuel racks are designed with a center-to center spacing of 21 inches. This spacing provides a minimum of 12 in. between adjacent fuel assemblies. This separation is sufficient to maintain a suberitical array assuming optimum moderation. Space between storage positions is blocked to prevent insertion of fuel. All rack surfaces that come into contact with the fuel assemblies are made of annealed authentic stainless steel, and the support structure is painted carbon steel. The racks are designed to withstand normal operating loads, as well as to remain functional with the occurrence of a Safe Shutdown Earthquake (SSE).

            'The new fuel racks are designed to withstand a maximum uplift force of 5,000 pounds and to meet the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section III, Appendix XVII.

The new fuel storage pit access hatch is a three section cover. This cover will minimize the introduction of dust and debris into the pit. The cover is designed to withstand the impact force of a new fuel assembly dropped from the maximum elevation allowed by the 2-ton hoist of the FHB overhead crane. In addition, space is provided for the storage of fuel during refueling inside the Reactor Containment Building (RCB). See Section 9.1.2.1 for a description of the racks. 9.1.1,2 Facilities Descriotion. The FHB abuts the south side of the RCB and is adjacent to the west side of the Mechanical-Electrical Auxiliaries Building (MEAB) of each unit. The locations of the two FHBs are shown in the station plot plan on Figure 1.2 3. For general arrangement of the new fuel storage facilities, refer to Figures 1.2-39 through 1.2 48. New fuel assemblies are received in the receiving area of each FHB and temporarily stored in the shipping containers in the new fuel handling area. 9.1 1 Revision 0

1 I 1 STPEGS UFSAR In the new fuel handling area, each new fuel assembly is removed from its shipping container and inspected visually to confirm the assembly has not been

    -damaged during shipment. The new fuel assemblies are transported from the inspection area to the new fuel storage pit or to the new fuel elevator by the 15/2-ton, dual-service FHB crane. The 2-ton hoist of this crane is designed          ;

to handle new fuel assemblies. New fuel handling is discussed in detail in 1 Section 9.1.4. Use of the 2-ton hoist of the 15/2-ton crane or of the fuel-handling machine to handle new fuel ensures that the design uplift of the racks will not be exceeded. The new fuel storage pit is situated in the approximate center of each FHB, The floor of the new fuel storage pit is at El. 50 ft-3 inches. The new fuel storage pit access hatch is provided with a three section protective cover at El. 68 ft. The fuel assemblies are loaded into the new fuel storage racks through the top and stored vertically. 9.1.1.3 Safety Evaluation. Units 1 and 2 of the STPECS are each  ! provided with separate and independent fuel handling facilities. l Flood protection of each FHB is discussed in Section 3.4.1. Flooding of the new fuel storage pic from fluid sources inside either FHB is not considered credible since all fluid systems components are located well below the , elevation of the new fuel storage pit access hatch. A floor drain is provided  ; in the new fuel storage pit to'ainimize collection of water. , The applicable design codes and the ability of the FHB to withstand various external loads and forces are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7. Missile protection of the FHBs is discussed in Section 3.5. Failure of nonseismic systems or structures will not decrease the degree of suberiticality provided in the new fuel storage pit. In accordance with American National Standards Institute (ANSI) N18.2, the design of the normally dry new fuel storage racks is such that the effective multiplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place, assuming optimum moderation (under dry or fogged conditions). For the unborated flooded condition, assuming new fuel of the highest anticipated enrichment in place, the effective multiplication factor does not exceed 0.95. Credit may be taken for the inherent neutron-absorbing effect of the materials of construction. The new fuel assemblies are stored dry, the 21-in spacing ensuring a safe geometric array. Under these conditions, a criticality accident during refueling and storage is not considered credible. Consideration of criticali-ty safety analysis is discussed in Section 4.3. Design of the facility in accordance with RG 1.13 ensures adequate safety under both normal and postulated accident conditions. The new fuel storage rachs also meet the requirements of General Design Criterion (CDC) 62. 9.1.2 Spent Fuel Storage 9.1.2.1 Desien Bases. The spent fuel pool (SFP) is a stainless steel-lined reinforced concrete pool and is an integral part of each FHB. All spent 9.1 2 Revision 0

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1 STPECS UFSAR fuel racks are classified as seismic Category I, as defined by RC 1.29, and as ANS SC 3. 3 The spent fuel storage facility provides storage capacity for 1,969 high Two density poison spent fuel racks in a honeycomb Therearray in each are 288 unit.cells in storage storage regions are provided in the SFP. 3 Region 1 type racks and 1,681 storage cells in Region 2 type racks. Figure 9.1.2 2 shows the pool layout for both Units 1 and 2. The six Region 1 rack modules are located in the northwest corner of the spent fuel pool. The Region 1 racks have 10.95-in. nominal center-to-center spacing with removable poison assemblies between the cells. This region is conservatively designed to accommodate unirradiated fuel at enrichments to 4.0 weight l3 percent. Region 1 storage cells are each bounded on four sides by a water box except on the periphery of the pool. The neutron poison material (Boraflex") is located in these water boxes. This is accomplished by capturing two sheets of the poison material on two outside opposite faces of a thin-walled rectan-gular box. The poison sheets are captured under thin stainless steel sheets which are intermittently welded all around to the thin-walled rectangular box. This configuration allows any irradiation gas formation in the poison to escape. A locking device engages the structure under the lead-in guide to hold the assembly in place. Special tools are provided for unlocking, removing,. reinstalling, and locking this poison assembly. The axial location of the poison with respect to the active fuel region is provided and main-tained by this welded assembly structure. Figure 9.1.2-3 shows a typical Region 1 spent fuel rack. The react *.vity characteristics of fuel assemblies which are to be placed in the spent fuel storage racks are determined and the assemblies are categorized by reactivity. Alternately, if necessary, all assemblies may be treated as if each assembly is of the highest reactivity class until the actual assembly reactivity classification is determined. Section 5.6 of the Technical Specifications provides the definitions of the reactivity classifications and the allowed storage patterns. Fuel assemblies are loaded into the racks in a geometrically safe configuration to ensure rack suberiticality. 3 Fuel assembly reactivity requirements for close packed storage and checker-board storage are specified in the Technical Specifications. The boron concentration of the water in the spent fuel pool is maintained at or above is less than or equal to the minimum value needed to ensure that the rack K 0.95intheeventofmisplacedassembliesinthecl,o,sepackedstorageareasor , in checkerboard storage areas. Consideration of criticality safety is r discussed in Section 4.3. l l The Region 2 racks have a 9.15-in, nominal center to center spacing with fixed )3 l poison material surrounding each cell. A sheet of neutron poison material is I captured between the side walls of all adjacent boxes. To provide space for the poison sheet between boxes, a double row of matchin5 flat round raised areas are coined into the side walls of all boxes. The raised dimension of these locally formed areas on each box vall is half the thickness of the poison sheet. The boxes are fusion welded together at all these local areas. The poison sheets are scalloped along their edges to clear these areas. Figure 9.1.2-4 shows a typical Region 2 spent fuel rack. 9.1-3 Revision 3 l

STPEGS UFSAR The axial location of the poison with respect to the active fuel region is provided and maintained by the structure of each box. At the outside

      . periphery of each rack, a sheet of poison material is captured under thin stainless sheets which are intermittently welded all around to the box.

All rack surfaces that come into contact with fuel assemblies are made of annealed austenitic stainless steel. These materials are resistant to corrosion during normal and emergency water quality conditions. The racks are designed to withstand normal operating loads as well as to remain functional with the occurrence of an SSE. The racks are designed with adequate energy absorption capabilities to withstand the impact of a dropped spent fuel assembly from the maximum lift height of the spent fuel pit bridge hoist. The racks are designed to withstand a maximum uplift force equal to the uplift force of the bridge hoist. The 14-in, and 16-in. racks also meet the require-ments of ASME Code, Section III, Appendix XVII. The high-density spent fuel racks meet the criteria of Appendix D to Standard Review Plan (SRP) 3.8.4 Shielding for the SFP is adequate to protect plant personnel from exposure to radiation in excess of published guideline values as stated in Section 12.1. A depth of approximately 10 ft of water over the top of the spent fuel assemblies will limit direct radiation to 2.5 mR/hr (surface dose rate). The FHB Ventilation Exhaust System is designed to limit the offsite dose in the' event of a significant release of radioactivity from the fuel, as discussed in Sections 12.3.3, 15.7.4, and 9.4.2. The FHB is designed to prevent missiles from contacting the fuel. A more detailed discussion on missile protection is given in Section 3.5'. In addition, space is provided for storage of fuel during refueling inside the RCB for 64 fuel assemblies in four 4 x 4 modules having 16-in. center-to-center spacing (Figure 9.1.2-1A). These modules are firmly bolted in the floor. 9.1.2.2 Facilities Descriotion. The FHB abuts the south side of the RCB and 'is edjacent to the west side of the MEAB of each unit. The locations of the two FHBs _are shown in the station plot plan on Figure 1.2 3. For general arrangement of the spent fuel storage facilities, refer to Figures 1.2-39 through 1.2 48. The spent fuel storage facilities are de:;1gneo for the underwater storage of opent fuel assemblies and control rods after their removal from the reactor vessel. The spent fuel is transferred to the FHB and handled and stored in

 -        the spent fuel pool underwater. 'The fuel is stored to permit some decay, then transferred offsite. For a detailed discussion of spent fuel handling, see Section 9.1.4.
        -The SFP is located in the northwest quadrant of each FHB. The floor of the pool is at El. 21 ft-11 in., with normal water level at E1. 66 ft-6 inches.

The top of a fuel assembly in a storage rack does not extend above the top of the stotage rack which is El. 39 ft-10 in, maximum. The fuel assemblies are loaded into the spent fuel racks through the top and are stored vertically. I 9.1-4 Revision 3 a

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9.1.2.3 Safety Evaluation. Units 1 and 2 of the STP are each provided with separate and independent fuel handling facilities. Flood protection of each FHB is discussed in Section 3.4.1. A detailed discussion of missile protection is provided in Section 3.5. The applicable design codes and the various external loads and forces considered in the design of the FHB are discussed in Section 3.8.4. Details I of the seismic design and testing are presented in Section 3.7. I Design of this storage facility in accordance with GDC 62 and RG 1.13 ensures a safe condition under normal and postulated accident conditions. The K.,, of the-spent fuel storage racks is maintained less than or equal to 0.95, even if unborated water is used to fill the spent fuel storage pool, by both the designs of the fuel assemblies and the storage rack and the use of administra-tive procedures to control the placement of burned and fresh fuel. Under accident conditions, the K.,, is maintained well below 0.95 assuming 700 ppm borated water. The boron concentration of the water in the spent fuel pool is maintained at or above the minimum value needed to ensure that the rack F% ,, is less than or equal to 0.95 in the event of misplaced assemblies in the close packed storage areas or in checkerboard storage areas. Con-sideration of criticality safety is discussed in Section 4.3. The SFP is designed to maintain leaktight integrity. To ensure such integrity, the pool is lined with stainless steel plate, and plate welds are backed with channels to detect and locate leakage. Leakage entering these channels is directed to the Liquid Waste Processing System (LWPS) via the FHB sump. Should a leak be detected, either by a low-level alarm (setpoint: 6 in, below normal water level) or by the fuel pool liner channel leak detection method, the operator would initiate makeup to the spent fuel pool. Makeup capability is provided by permanently installed connections to: (1) the Demineralized Water System (DWS), (2) the Reactor Makeup Water System (RMWS), and (3) the refueling water storage tank (RWST) in the Emergency Core Cooling i System (ECCS).  !

     ^ :::plet: 1::: f SFP ::: ling i: :t ::ncid:::d : :::dible cuent sin : the

p: :nt: irv:1ved cr: d::ign:d te SC 3 ::!::i C:teg:ry I ::quirc:: t: :nd

uld b: p: :r:d frc: r:dund::t Engin::::d S:f:ty F::ture: (ESP) p: ::

cupplic:. Furth:r, th: cyst::: pr viding ::: ling cre r:dund:nt. Ther:f:re, n cingle failur: c:uld result ir : :::plet 1::: :f fuel p::1 ::aling. Fer : mece-detailed di:: ::ler :f SFP ::: ling, refer t: S::ti:r 9.1.3 S[Riplad Q ichilssirc @ ] 9.1.3 Spent Fuel Pool Cooling and Cleanup System The Spent Fuel Pool Cooling and Cleanup System (SFPCCS) is designed to remove the decay heat generated by spent fuel assemblies stored in the SFP and/or the in-Containment storage area. A second function of the system is to maintain visual clarity and purity of the spent fuel cooling water and the refueling water. 9.1.3.1 Dgsien Bases. The SFPCCS design heat loads are given in Table 9.1-1. System capabilities to withstand natural phenomena and piping rupture are a(dressed in Chapter 3. The spent fuel pool cooling portions of the SFPCCS are designed to seismic Category I requirements, and are located in the FHB, a seismic Category I building. The spent fuel pool water purification Misc-%\s233 w 9.1-5 Revision 0 s

INSERT #1 - UFSAR Section 9.1.2.3 (page 9.1-5) A complete loss of Spent Fuel Pool water inventory is not considered a credible event since the components involved are designed to Safety Class 3, Seismic Category I requirements. System piping is arranged so that a loss of piping i integrity, for all credible accidents, does not result in draining of the Spent Fuel Pool below a minimum depth of 23 feet above the top of the fuel. Details of the system design are provided in Section 9.1.3.2. Loss of Spent Fuel Pool heat removal is described in Section 9.1.3.3. 1 L i l I MISC-96\$233 w  ! l

STPEGS UFSAR portions of the SFPCCS are not required for safety functions and are not designed to seismic Category I requirements. 9.1.3.1.1 Soent Fuel Cooline: The SFPCCS is designed to remove the amount of decay heat produced by the number of spent fuel assemblies that are stored following refueling. The system design incorporates two trains of equipment. Each train is capable of removing 100 percent of the normal maximum design heat load and 50 percent of the abnormal maximum design heat load. The system can maintain the spent fuel cooling water temperature at or below the maximum allowable temperatures specified by Table 9.1-1. JAdd Inserty#2) This temperature is based on the heat exchangers (HXs) being supplied with component cooling water (CCW) at the design flow and temperature. The flow through the spent fuel storage areas provides sufficient mixing to maintain uniform water conditions. If it ic nececcary te rc:c::  :  :::plet : re frc: the recate Q 7arfajf61Rcore 6ff16ad, the system can maintain the spent fuel cooling water below the maximum allowable tempercture specified by Table 9.1-1. Makeup water requirements will be provided by either reactor makeup water, demineralized water, or refueling water. The makeup flowpath from the reactor makeup water storage tank (RMWST) is seismic Category I. The flowpaths from the demineralized water storage tank (DUST) and from the RWST are non-seismic Category I. 9.1.3.1.2 Dewaterine Protection: A depth of approximately 10 ft of water over the top of the stored spent fuel assemblies will limit direct radiation to 2.5 mR/hr. System piping is arranged so that failure of any pipeline cannot drain the spent fuel pool or the in-Containment temporary storage area below a depth of approximately 23 ft of water over the top of the stored spent fuel assemblies. JAddjInsertjy3) Additionally, means are provided to detect component or system leakage. Refer to Section 9.3.3 for the detailed description of leak detection via the floor drains. In addition, l the water level instrumentation provides a means of leakage detection. 9.1.3.1.3 Water Purification: The system's demineralizers and filters are designed to provide adequate purification to permit unrestricted access for plant personnel to spent fuel storage areas and to maintain optical clarity of the spent fuel cooling water and the refueling water. The optical clarity of the spent fuel pool surface is maintained by use of the system's skimmer pump, skimmer / strainer assemblies, and skimmer filter. The optical clarity of the refueling cavity water is maintained by the reactor cavity filtration system. The Nuclear Steam Supply System (NSSS) vendor-recommended specifications and guidelines for the spent fuel pool water purity are provided in Table 9.1-4 and the monitoring frequency is provided in Table 9.3-3. 9.1.3.2 System Description. The SFPCCS, shown on Figures 9.1.3-1 and 9.1.3-2 (piping and instrumentation diagrams (P& ids)), consists of two cooling trains Withjs[cbamon?dischargsiheader, two purification trains, a surface skimmer loop and a reactor cavity filtration system. The SFPCCS removes decay heat produced by spent fuel after it is removed from the reactor. Spent fuel is removed from the reactor core during the refueling sequence and placed in the SFP, where it is stored until it is shipped offsite for reprocessing or permanent storage. If, for some reason, it is desirable Misc 96\5233 w 9.1-6 Revision 4

INSERT #2 - UFSAR Section 9.1.3.1.1 (page 9.1-6) , Table 9.1 1 incorporates various Spent Fuel Pool loading scenarios for 18-month fuel reload cycles. ! INSERT #3 - UFSAR Section 9.1.3.1.2 (page 9.1-6) The return line contains an anti-siphon hole just below the low water level to prevent gravity drainage due to an open drain valve or due to credible design basis nine breaks. i l I' MISC-%\5233 w

STPECS UFSAR l or necessary to delay the transfer of the spent fuel to the SFP, the in-Containment storage area can be used for temporary storage of up to one third of a core. The system normally handles the heat load from one core region freshly discharged from the reactor. Heat is transferred from the SFPCCS through the HXs to the Component Cooling Water System (CCWS). i When the SFPCCS is in operation, water drawn from the SFP (and/or from the in-Containment storage area) by the SFP pumps is pumped through the tube side of the HXs, and then is returned to the spent fuel pool (and/or the in-Containment storage area). Each suction connection, which is provided with a strainer, is located at an elevation 4 ft below the normal water level (approximately 23 ft above the top of the fuel assemblies). The return line contains an antisiphon hole near the surface of the water to prevent gravity drainage. To maintain spent fuel cooling water purity, a bypass circuit composed of a demineralizer and a filter is connected to each cooling train. The demineral-izers are charged with either a mixed resin (cation and anion resin) or cation resin only, dependant on the type of contamination indicated by the required chemical analyses. While the heat removal operation is in process, a portion of the spent fuel cooling water is diverted upstrean of each HX and passed  ; through the purification circuit, returning downstream of the HXs. The  : demineralizers remove ionic corrosion impurities and fission products. ' Filters are provided to remove any additional particulates and to prevent any resin fines from entering the system from the demineralizer discharge. Transfer canal water may be circulated through the same purification circuits by removing the gate between the canal and the spent fuel pool. These ] purification loops are sufficient for removing fission products and other contaminants which may be introduced into the spent fuel cooling water. One purification loop may be isolated from the heat removal portion of the SFPCCS. By so doing, the isolated equipment may be used in conjunction with either the reactor coolant drain tank pumps or the refueling water purifica-tion pump to clean and purify the refueling water whileConnections spent fuel cooling and are provided spent fuel cooling water cleanup operations proceed. such that the refueling water may be pumped from either the RUST or the refueling cavity through the demineralizer and filter, and discharged to either the refueling cavity or the RUST. Samples are periodically taken to determine the need for purification of the water as well as the purification efficiency. To further assist in maintaining spent fuel cooling water clarity, the spent fuel pool is cleaned by a skimmer loop. Water is removed from the surfaces via two skimmer / strainer assemblies located in the SFF. Water is pumped through a filter by a skimmer pump and returned to the pool surface at a single location remote from the skimmer / strainer assemblies. Piping for future addition of a third skimmer / strainer for cleaning the surface of the fuel transfer canal water is also provided. l The SFP is initially filled with water having the same boron concentration as that in the RWST. Borated water may be supplied from the RWST via the SFPCCS return header, or by running a temporary line from the boric acid blending tee, located in the Chemical and Volume Control System (CVCS), directly into to the pool. Demineralized water can also be added for makeup purposes (i.e., replace evaporative losses) through a connection in the SFPCCS return header. 9.1-7 Revision 3

STPEGS UFSAR The water in the spent fuel pool may be separated from the water in the l transfer canal by a gate. The gate is installed so that the transfer canal may be drained to allow maintenance of the fuel transfer equipment. The water in the transfer canal is first pumped, via a portable pump, into the spent fuel pool and then is transferred to the recycle holdup tanks in the Boron Recycle System (BRS) by a SFP pump. When maintenance on the fuel transfer equipment is completed, the water is returned directly to the SFPCCS by the recycle evaporator feed pumps (BRS). A portable pump is again used to return water to the transfer canal. When spent fuel assemblies are stored in the in-Containment storage area, l either of the cooling trains may be utilized to remove the decay heat. The in-Containment storage area is sized to temporarily store one-third of a core. The in-Containment storage area is directly connected to the refueling cavity and is filled with refueling water whenever the refueling cavity is filled. Thus, during refueling, the in-Containment storage area is always ready for use. During refueling outages, the clarity of the water in the reactor cavity is maintained by the Reactor Cavity Filtration System. Water is removed from the reactor cavity pool through a submerged strainer located one foot above the reactor cavity floor, pumped through four cartridge-type filters, and returned to the reacter cavity pool. 9.1.3.2.1 Component

Description:

The design codes and classifications of the components are given in Section 3.2. Equipment design parameters are given in Table 9.1-2. Soent Fuel Pool Pumos The pumps are horizontal, centrifugal units, with all wetted surfaces being stainless steel. The pumps draw water from the spent fuel pool (and/or the in-Containment storage area) and deliver it to the HXs for cooling and to the purification trains for cleanup. Spent Fuel Pool Skimmer Pumo This horizontal, centrifugal pump takes suction from the SFP via adjustable surface skimmer / strainer assemblies and from the fuel transfer canal and circulates the water through a filter and returns it to the SFP and the fuel transfer canal. All wetted surfaces of the pump are austenitic stainless steel. Refueline Water Purification Pumo This centrifugal pump is used to circulate water from the RWST through a SFP demineralizer and filter, j Suent Fuel Pool Heat Exchanzers The HXs are the shell and U-tube type. Spent fuel cooling water circulates through the tubes while CCW circulates through the shell. Each HX is sized for 50 percent of the heat load, n e design h: t lead cf the cFP M: le-based l

- th: d:: y heat gen: ret:d by en: third f : ::re placed ir the SFP chertly fter ::::ter chutd:r- during : refueling peratier uith one third of a ::re fren e r preficu refueling :Ir :dy ir the p::1.]RepladsIyithlinssrtQ#4f]

MISC-%\52h 9.1-8 Revision 0 l

1 l

 -                                                                                 l INSERT #4 - UFSAR'Section 9.1.3.2.1 (page 9.1-8)

The heat exchangers have been evaluated to accommodate the Abnormal Maximum heat loads and temperatures given in Table 9.1-1. J 4 a 1 l l t MISC-96\$233.w

 ,                                        STPEGS UFSAR                                 j Soent Fuel' Pool Demineralizers The two flushable demineralizers are designed to provide adequate spent fuel cooling water purity for unrestricted access of plant personnel to the spent fuel storage areas.

Soent Fuel Pool Filters A filter is located in each purification train, downstream of the demineralizer, to collect possible particulates and resin fines passed by the 2 demineralizer. The filter assembly utilizes a disposable cartridge filter and is readily accessible for filter change. Soent Fuel Pool Skimmer Filter The SFP skimmer filter is used to remove particles swept from the spent fuel pool surface which are not removed by the skimmer / strainer assembly. The filter assembly utilizes a disposable cartridge filter and is readily acces-sible for filter change. Soent Fuel Pool Strainers A strainer is located in each SFP pump suction line from the SFP to prevent introduction of relatively large particles that might clog the spent fuel demineralizers or damage the SFP pumps. Soent Fuel Pool Skimmer / Strainer Assemblies Two assemblies are provided. These assemblies make it possible to take l suction from the pool surface and remove debris from the skimmer process flow. i Fuel Transfer Canal Skimmer / Strainer Assembly (Future Expansion) Piping is provided for future addition of one assembly which would take suction from the transfer canal surface. Debris would be removed via the skimmer filter. In-Containment Storace Area Strainer A strainer is located in the SFP pump suction line from the in-Containment storage area to prevent introduction of relatively large particles that might , clog the SFP demineralizers or damage the SFP pumps. I Reactor Cavity Filtration System The Reactor Cavity Filtration System is a skid-mounted package system including a horizontal, centrifugal pump with an electric motor driver, four filter housings with cartridge-type filters, suction screen, and the necessary valves, instrumentation, and piping. (AddjInsertil#5)) Misc.96633 w 9.1-9 Revision 0

INSERT #5 - Add New UFSAR Section 9.1.3.2.2 After Section 9.1.3.2.1 (page 9.1-9)

             -9.1.3.2.2    Spent Fuel Pool Cooling During Refueline Operations Under the guidance provided in Standard Review Plan Section 9.1.3, during a full-core offload, the abnormal maximum criteria apply and a single active failure need not be considered. For this condition, the Standard Review Plan requires the temperature of the Spent Fuel Pool water to be kept below boiline with the available normal systems in operation. Since a full-core offload is the normal refueling practice at the South Texas Project, this case has been analyzed )

considering a single active failure (e.g., loss of one Spent Fuel Pool cooling train). Table 9.1-1 shows that the "no-boiling" criterion is met for the South  ! Texas Project Abnormal Maximum case. 1 l Table 9.1-1 gives the Spent Fuel Pool temperature limits for various fuel load and Spent Fuel Pool cooling configurations. Cycle specific calculations will be performed, if necessary, to ensure that these Spent Fuel Pool temperature limits are met. Cycle specific calculations are not required if the typical South Texas Project fuel load conditions, specified in Table 9.1-1, represent the intended refueling operation. During full-core offload conditions at South Texas Project, two Spent Fuel Pool cooling trains are administratively required to be available. At least one Spent Fuel Pool cooling train will be available at all times backed by an on-site power source. The second cooling train will at least be functional, backed by either an on-site power source or an offsite power source available yia the switchyard. l I MISC-96\5233 w

 .s                                                                                        l l

< STPEGS UFSAR I l l i 9.1.3.3 Safety Evaluation. l 1 1 9.1.3.3.1 Availability and Reliability: The SFPCCS has no emergency l function during an accident except to provide adequate cooling to the SFP. l Since it is not necessary for automatic initiation post-accident, the SFP I pumps are manually placed on the emergency power bus after completion of , automatic load sequencing. In the event of failure of a SFP pump or loss of cooling to a SFP HX, the second cooling train would provide continued cooling of the stored spent fuel with the spent fuel cooling water at a higher equilibrium temperature. Mdd3dsdrt%} A failure modes and effects analysis for the SFPCCS is given in Table 9.1-5. 9.1.3.3.2 Soent Fuel Storace Area Dewatering: The most serious failure of this system would be complete loss of water in one of the storage areas. To protect against that possibility, the SFP pump suction connections enter near the normal water level so that the storage areas cannot be siphoned. The eccling unter returr line te cce' cterage crec centairr en enticipher hele te prevent the percibility cf ciphening {Adg1rsarQM] These design features assure that neither the SFP nor the in-Containment storage area can be drained more than 4 ft below the normal water level (normal water level is approximately 27 ft above the top of the stored spent fuel). If a seismic event results in the failure of a non-seismic pipe in the reactor makeup water tank compartment, both reactor makeup pumps may be lost because of flooding in this compartment. As a result, makeup for the SFP from the RMWS would be lost. However, the following means may still be available to provide makeup to the SFP.

1. Water from the RWST (seismic Category I) through the non-seismic Category I refueling water purification pump 1A and SFP demineralizer lA to the SFP cooling return line.

2. Demineralized water through a 2-in, line connecting the demineralized water system to the SFP cooling return line.

3. Fire water from either one or both of the hose reel and hose cabinet located near the SFP.

In the unlikely event that all of these backup sources were not available. a seismically qualified makeup water source would be provided by connecting temporary hoses to the vent and drain valves Incated on the low head safety injection (LHSI) pump discharge piping. The LHSI pumps are located in the FHB at the lowest level. The hoses will be routed through building stairways and equipment hatches to the SFP. A: uming ec=plete less of SFP cccling and a heat Iced : deceribed ir 9.1. 2. 3. ^ belcu, the he tup ti=c te bciling 1 appremi= tely S heur hered uper .::ter ler: equivalent te ^ f t cf pec1 Icvel. [Ryp})cp1Gith]I6aerp[sS[]This allows sufficient time to route the hoses through the building. 9.1.3.3.3 Water Ouality: Whenever a fuel assembly with defective cladding is removed from the reactor core, a small quantity of fission products may enter the spent fuel cooling water. The purification loops i provide a means of removing fission products and other contaminants from the ' water. By maintaining radioactivity concentrations in the spent fuel cooling water at 5 x 10-2 Ci/cc (# and y) or less, the dose at the water surface is 2.5 mR/hr or less. Misc-9asmw 9.1-10 Revision 0

( INSERT #6 - UFSAR Section 9.1.3.3.1 (page 9.1-10) l. Table 9.1-1 shows various Spent Fuel Pool loading scenarios for 18-month fuel l reload cycles. Table 9.1-1 incorporates the routine refheling practice of full-core offload,18-month cycles. INSERT #7 - UFSAR Section 9.1.3.3.2 (page 9.1-10) The return line contains an anti-siphon hole just below the low water level to prevent gravity drainage due to an open drain valve or due to credible design basis p_ine breaks. ( INSEhr #8 - UFSAR Section 9.1.3.3.2 (page 9.1-10) For a complete loss of Spent Fuel Pool cooling, the maximum heat load occurs for the South Texas Project Abnormal Maximum case listed in , Table 9.1-1. This analysis assumes a Spent Fuel Pool water level 4 feet below the normal water level, and that both Spent Fuel Pool cooling trains simultaneously fail just when the Spent Fuel Pool reaches its , maximum temperature. This worst case scenario time-to-boil is calculated to be M hours. 1 1 1 l MISC 06\5231w

O STPECS UFSAR 9.1.3.3.4 Soent Fuel Pool Boiline Dose Analysis: In the event of a fire or moderate energy line crack in the FHB that disables both trains of SFP l cooling, the SFP temperature would begin to rise and, assuming no corrective l action, would eventually boil. The following analysis examines the dose l consequences of a loss of the SFPCCS and the use of the seismic Category 1  ; makeup water source (PJNS). I l It is assumed that a loss of the SFPCCS occurs after a refueling where aThe full core has been removed and placed into the SFP 120 hours after shutdown. heat loads supplied to the pool are comprised of the following sources: 1) the full core removed prior to the event; 2) 92 assemblies which have decayed 36 i i days after shutdown; and 3) spent fuel from the previous 20 refueling off- t loads. The last full core offload fills.the SFP to the maximum capacity of l 1969 assemblies. For the purpose of this calculation, the pool is conserva-This 3 tively assumed to boil instantaneously after the loss of the SFPCCS. Throughout loss of SFPCCS is assumed to occur at 120 hours after shutdown.  ; the event, theleakag)erateforiodineisassumedtobethenormalfullpower rate (1.3 x 10-e 3,e The iodine available for release is based upon the l i gap activity containing 10 percent of the rod inventory and the leakage occurs  ; from the defective 1 percent of the rods. The activity of the refueling water prior to initiation of the event is assumed to be negligible. Using these assumptions and those found in Table 9.1-6, the thyroid dose consequences of releasing the iodine as a result of SFP boiling are well below the dose requirements of 10CFR, Part 100. 9.1.3.4 Instrumentation Aeolication. The instrumentation provided for the SFPCCS is discussed below. Alarms and indications are provided as noted. 9.1.3.4.1 Temocrature: Instrumentation is provided to measure the temperature of the water in the SFP and in the in-Containment storage area and to give local indication as well as annunciation at the main control board when normal temperatures are exceeded. Instrumentation is also provided to give local indication of the temperature of the spent fuel cooling water as it leaves each HX. 9.1.3.4.2 Pressure: Instrumentation is provided to measure and give local indication of the pressures in the suction line of the SFP skimmer pump and in the suction and discharge lines of the refueling water purification pump and of each SFP pump. Instrumentation is also provided at locations upstream and downstream of each SFP filter, each SFP skimmer filter, and each of the SFP demineralizer so that the pressure differential across these l filters can be determined. 1 f 9.1.3.4.3 Upw: Instrumentation is provided to measure and give local l indication of the flow in the outlet line of each SFP filter. Instrumentation j is also provided to measure discharge flow from the refueling water purifica-tion pump and to provide a low flow alarm on the main control board in l l l addition to providing local flow indication. l 9.1.3.4.4 Level: Instrumentation is provided to give an alarm in the i control room when the water level in the SFP or in the in-Containment storage  ; I area reaches either the high or low-level setpoints (6 in above or below l l normal water level). l 9.1-11 Revision 3

i STPECS UFSAR 9.1.3.5 Tests and Insoections. Active components of the SFPCCSThe are in either continuous or intermittent use during normal system operation. SFFCCS is included in the inservice inspection requirements described in Section 6.6 and the inservice testing requirements described in Section 3.9.6. 9.1.4 Fuel Handling System 9.1.4.1 Desien Bases. The FHS consists of equipment and structures utilized in the transporting and handling of the fuel from the time it reaches the station until it leaves the station. The following design bases apply to the FHS:

1. Fuel-handling devices have provisions to avoid dropping or jamming of fuel assemblies during transfer operation.

2. Fuel lifting and handling equipment will not fail in such a manner as to damage seismic Category I equipment in the event of an SSE. l l

The Fuel Transfer System (FTS), where it penetrates the Containment, has 1 3. provisions to preserve the integrity of the Containment pressure boundary, including a means to test for leak tightness.

                   ~

4 Each machine used to lift spent fuel has a limited maximum lift height so that the minimum required depth of water shielding is maintained. l

5. The cask handling components and the FHB layout limit vertical lift above the floor of the cask to less than 30 f t above the floor during ,

any moving sequence. l

6. The Spent Fuel Cask Handling System (SFCHS) utilizes a wet handling technique.

7. The pool gates are designed to maintain their integrity in the event of an SSE.

9.1.4.2 System Descrintion. The equipment in the FHS is comprised of 1Lfting equipment, handling equipment, an FTS, and the SFCHS. The structures associated with the FHS are the refueling cavity, refueling canal, and in-Containment fuel storage area inside the RCB; and the fuel transfer canal, SFP, new fuel storage pit and inspection area, and cask loading pool and decontamination platform in the FHB. The equipment is located in seismic Category I buildings. 9.1.4.2.1 New Fuel Handline: The new fuel arrives onsite either by rail or truck. The rail track and truck receiving areas are located on the ground floor of the FHB. When new fuel is delivered to the receiving area within the FHB, the shipping containers are unloaded from the transport vehicle and examined for shipment damage. The shipping containers are then lifted to the operating floor by the FHB overhead crane. The shipping containers can be placed directly in the new fuel inspection laydown area on the operating floor or lowered through the equipment hatch in the operating floor to the new fuel handling area below by means of the THB overhead crane. The shipping containers may be unloaded in either area. The shipping con-tainers are placed horizontally on the floor. The containers are then stacked 9.1-12 Revision 0

• J STPEGS UFSAR

                                                                                                               \

using the overhead crane servicing this area. One by one, these shipping containers are unstacked, their covers removed, and the pivotal, fuel support j structure within the shipping container is elevated from the horizontal to the i vertical' position. In the new fuel handling area, this is accomplished using i the new fuel handling area overhead crane. In the new fuel inspection laydown j area on the operating floor, this is accomplished using the FHB overhead j crane. The various clamping devices securing the fuel assembly to the support ' structure are then removed. The fuel assembly is lifted from the shipping container support structure. Inspection activities may now be conducted in either areas. Alternatively, inspection activities may be conducted at a later time following transfer of the fuel to the new fuel storage pit. Following inspection, unacceptable new fuel assemblies are set aside for dispositionin5 Acceptable new fuel assemblies are engaged by the new fuel handling tool, which is in turn attached to the hook of the FHB overhead crane. If the fuel was unloaded in the new fuel handling area, the new fuel assembly must be secured beneath the equipment hatch, released from the new fuel handling area overhead crane, and then engaged by the FHB overhead crane and lifted through the equipment hatch to the operating floor. Fuel , assemblies are either inserted into the new fuel storage racks in the new fuel storage pit, placed in the new fuel elevator which is located in the fuel  ; transfer canal, or placed directly into the SFP (Cycle 1 fuel only). Those assemblies placed in the new fuel elevator are lowered to the bottom of the fuel transfer canal. They are then engaged by the spent fuel handling tool, which is in turn suspended from the fuel handling machine. The fuel handling  !

        -machine either transfers the assembly to the spent fuel storage racks (before-                      ,

initial refueling) or to the FTS upender for transfer to the RCB for refueling operations. The upender pivots the assembly to the horizontal position and . the FTS fuel container carries it through the fuel transfer tube to an upender

• inside the Containment.

i 9.1.4.2.2 Refueline Procedure: The refueling operation follows a. - detailed procedure to ensure a safe, efficient refueling operation. Prior to initiating refueling, shutdown conditions are as specified in the Technical  : Specifications. Criticality protection for the refueling operation,' including a requirement for checks .of boron concentration, is specified in the Technical l Specifications, l Protection against uncontrolled rod cluster control assembly (RCCA) bank , withdrawal from a suberitical condition is described in Section 15.4.1 and I includes source-range, intermediate-range, and power-range high neutron flux i trips. The transient is assumed to be terminated by the power-range high neutron flux (low setting) reactor trip. Protection against uncontrolled

        .borori dilution is described in Section 15.4.6.                                                      {

l The following significant points are assured by the refueling procedure. , l

1. The refueling water and the reactor coolant contain approximately 2,800 l l ppm boron. This concentration is sufficient to keep the core approximately 5 percent ak/k suberitical during the refueling operations with all control rods removed and the core refueled to provide sufficient excess reactivity for operation to the next refueling outage.

9.1 13 Revision 4

q 1 o.-  ; 1 STPEGS UFSAR l l l l .2. The water level in the refueling cavity is high enough to keep the. 1 l

                 . radiation levels within acceptable limits when the fuel assemblies are l                  being removed from the core.

l The refueling operation is divided into four major phases: (1) preparation, l (2) reactor disassembly, (3) fuel-handling, and (4) reactor assembly. A general description of a typical refueling operation through the four phases i is given below. This description applies to rapid (unrodded) refueling which will normally be used during a refueling shutdown to maximize plant availability. The  ; description also points out the different steps that would be included in a nonrapid (rodded) refueling operation, which is typically used for extended I !' shutdowns involving nonroutine maintenance. l- 9.1.4.2.2.1 Phase I - Precaration - The reactor is shut down .M ds in) I and then simultaneously borated and cooled down to refusling shutdown l condition. Following initiation of normal purge and a radiation survey of the l Containment Building, refueling operations may proceed. The fuel transfer j [ equipment and refueling machines are checked for proper operation. Each new , fuel assembly is brought from dry storage 'in the FHB as described in Section j 9.1.4.2.1. After transfer through the fuel transfer tube, the FTS fdel l l assembly container is pivoted to the vertical position by the in-Containment  ! upender. ;The refueling machine transfers the new fuel into the in-Containment  ! l storage racks (shown in Figure 9.1.2.1.a). Refer to Figures 1.2-14 through l-1.2-18 for general arrangement of the in-Containment fuel storage area, 1 t When the Reactor Coolant System (RCS) has been cooled to 150*F, RCS draining is started. The RCC assemblies (control rods) are withdrawn to their full-out position, and each control rod's holdout device is activated to ensure that j the, rod is held in its withdrawn position inside its upper internals. guide tube and reactor head pressure housing. As soon as RCS draining has-lowered  ; reactor coolant level to the reactor vessel nozzle centerline, degassing-  ! operations using the RCS Vacuum Degassing System (Setion 11.3) may be used, if l necessary, to remove radioactive gases prior to head removal. i For nonrapid refueling, the prepar.ation for refueling is similar except that l !~ the control rods are not withdrawn from the core and the RCS temperature is reduced to 140*F before draining, 9.1.4.2.2.2 Phase II - Reactor Disassembiv - The seismic tie rods attached to the missile shield are disconnected and stored. The insulation is removed from the vessel head flange area, and the Roto-Lok studs are i detensioned and removed from the vessel flange. A stud hole plug is installed in each hole after the stud is removed to prevent entry of water. In addition, all flux mappin5 detectors and thimbles are retracted through the bottom of the reactor vessel. The refueling cavity is prepared for flooding l by removing all. tools, closing the refueling canal drain holes and by j installing underwater lights. The reactor cavity-is then flooded to 12 in. , below the top of the head flange. The upper head package (i.e., head, ' missile, cable bridge, upper internals, control rods, and rod drives) is lifted by the polar crane until the cdosure head guide pins are clear. Water from the RUST is pumped into the RCS by the LHSI pumps, causing the water to j overflow into the refueling cavity. The vessel head is lifted in conjunction l l with the water level in the refueling )

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9.1-14 Revision 4 l

O STPEGS UFSAR cavity. When the refueling cavity is full, the upper package is moved to storage at the end of the refueling cavity opposite the refueling canal. If a radiation survey indicates the need, additional water shielding may be provided by pulling a vacuum through the reactor vessel head vent connection. For nonrapid refueling, reactor disassembly includes additional steps starting with disconnecting the cables that run across the cable bridge. This allows the bridge to be lifted by the polar crane separately from the reactor head. The control rod drive shafts are disconnected leaving the rods in the cors. The upper internals assembly is then disconnected from the reactor vessel head. After the cable bridge is lifted clear of the reactor head, the head is lifted about 24 in. to permit a visual inspection of the RCC assembly drive shafts. This ensures that they are free in the control rod drive mechanism (CRDM) housings and were not raised with the reactor head. The refueling cavity is then flooded to a level just below the closure head. Simultaneous lifting of the reactor head and filling of the refueling cavity proceeds in the same manner as described for rapid refueling. The' vessel head is placed on a dry storage pedestal in a roped off area of the operating floor at the north end of the containment. RCC drive shafts are unlatched from their respective RCC assemblies. Finally, the upper internals are removed from the vessel by using the reactor internals lifting device , suspended from the polar crane. The internals package is vet-stored on a stand in the north end of the refueling cavity. . 9.1.4.2.2.3 Phase III - Fuel Handline - Fuel assemblies are removed from and inserted in to the reactor core by the refueling machine. The spent fuel assemblies ere removed from the core in a sequence which is planned before each refueling. The positions of partially spent fuel assemblies are shuffled, and new fuel assemblies are added to the core. The general fuel handling sequence is:

1. The refueling machine is positioned over a spent fuel assembly in the most depleted region of the core.

2. The spent fuel assembly is lifted by the refueling machine to a predetermined height sufficient to clear the reactor vessel and still leave sufficient water covering the spent fuel assembly to eliminate any radiation hazard to the operating personnel.

3. The fuel transfer car is moved into the refueling canal from the fuel transfer canal.

4 The FTS fuel assembly container is pivoted to the vertical position by the upender..

5. The refueling machine is moved from over the core to line up the spent fuel assembly with the fuel assembly container.

6. The refueling machine loads the spent fuel assembly into the fuel assembly container of the FIS transfer car.

7. The container is pivoted to the horizontal position by the upender.

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9. The fuel assembly container is pivoted to the vertical position by the i upender. The spent fuel assembly is unloaded by the spent fuel handling l tool, which is suspended from the fuel handling machine hoist.

10. The spent fuel assembly is placed in the spent fuel storage rack after being transferred through the gate between the fuel transfer canal and the SFP. 1 l

11. Partially spent fuel assemblies are moved is}]:hhQ@e}@pigtR@HPM

[p M g @ lggjg}pigp @ pg $ g to new positions in the reactor core, and new fuel assemblies are moved from the in-Containment storage racks  ! to the core to replace the spent fuel assemblies that were removed l during the preceding fuel handling steps. l l

12. This procedure is continued until refueling is completed.

During nonrapid refueling, some of the spent fuel assemblies that are removed from the reactor core will contain a rod cluster control (RCC) (control rod) I element. Such assemblies are placed in the RCC change fixture by the l refueling machine. The RCC change fixture is located adjacent to the in- l Containment storage racks. Here the RCC element is removed from the spent  ; fuel assembly and deposited in a partially spent or new fuel assembly l previously placed in the RCC change fixture. Another step generally performed during either rapid or nonrapid refueling is l the removal of an irradiated specimen from the reactor core for examination. Also, remote boroscope and television camera inspections of the core and

   .          reactor vessel are performed using equipment suspended from the refueling machine.

9.1.4.2.2.4 Phase IV - Reactor Assembly - Reactor assembly, following refueling, is essentially achieved by reversing the operations given in " Phase II - Reactor Disassembly". 9.1.4.2.3 Soent Puel Shioment: Spent fuel is shipped offsite either by rail or truck in heavily shielded spent fuel shipping casks licensed for use by the Department of Transportation (DOT). General arrangements of the cask handling area are provided in Section 1.2. Upon receipt in the FHB, the spent fuel cask shipping vehicle (rail or truck) is braked and blocked in position for removal of the spent fuel cask. The cask is then inspected for shipment damage and to ascertain the degree of additional cleaning that will be required to remove road dirt or radioactive contamination from the cask surface. The gate between the cask loading pool and the decontamination area is removed using the 15/2-ton FHB overhead crane. The upper and lower spent fuel cask impact structures are removed, and spent fuel cask appurtenances are disconnected. The spent fuel cask yoke is removed from its storage area and attached to the spent fuel cask trunnions using the 150-ton, overhead cask-handling crane. As the spent fuel cask is upended, the yoke is maintained in a vertical position by lateral movement of the overhead spent fuel cask handling crane trolley. (The reverse procedure is performed when the cask is loaded onto the shipping vehicle.) The spent fuel cask is then lifted clear of the shipping vehicle, moved over to the access bay, and lowered to the decontamination area, where any additional cleaning of road dirt or surface contamination is performed. Misc-ms233 w 9.1-16 Revision 0

. _ __._ _ _ _ _ . _ _ __ _ _ ___ ~ _ _ _ _ . ._ __ _ O I i STPECS UFSAR Using the 150 ton, overhead cask handling crane, which is still connected to the cask, the spent fuel cask is lifted, moved horizontally, and lowered into the cask loading pool. The lifting yoke is removed and placed in its storage  ; location. The spent fuel cask head is unbolted and moved, using the FHB 1 overhead crane, to a temporary storage location. The gate between the cask ' loading pool and the decontamination area is replaced. The cask loading pool and channel are filled with borated water to the same level s's the SFP 1 (nominally El. 66 ft-6 in.). The two gates between the cask loading pool and l the SFP are removed, by the 15/2-ton (15 ton main hook and a 2-ton auxiliary hook) FHB overhead crane, and fuel transfer is initiated. Ioading of the spent fuel assemblies is accomplished in the following manner: the spent fuel assembly in the SFP is engaged by the long spent fuel handling tool, which is , in turn attached to the fuel handling machine; the fuel assembly is removed i from the storage rack and transferred through the gate area into the cask I loading pool and then lowered into the spent fuel cask. The height to which  ! the spent fuel assembly can be lifted is restricted by the length of the spent l fuel handling tool and the fuel handling machine design in order to provide a minimum of 10 ft of shielding water above the spent fuel assembly. This spent fuel handling procedure is repeated until the spent fuel cask is full. The gates sealing the cask loading pool from the SFP are then replaced using the 15/2 ton FHB overhead crane. The spent fuel cask head is lifted from its storage area using the 15/2-ton overhead FHB crane and placed on the spent fuel cask. The cask head bolts that secure the cask head to the cask are installed and tightened " finger l tight" using long handled tooling _ operated from the fuel-handling machine. The cask loading pool level is lowered to at least 6 in..below the bottom of

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the gate between the cask loading pool and the decontamination area. This gate is removed using the 15/2 ton overhead crane. Using the 150-ton cask-handling crane, the cask lifting yoke is attached to the cask-lifting trunnions and the cask is transferred from the cask loading pool to the cask decontamination area. A radiological survey is made of the cask to determine the extent of cask decontamination necessary. The head bolts are torqued to the proper value, and various preshipment testing requirements characteristic of the spent fuel cask being used are completed. These tests may include a cask leak test and monitoring of cask coolant activity. The contaminated outside surface areas are manually decontaminated using high pressure sprayers or by using scrub brushes with detergent, rinses, and wipes. Once the cask is decontaminated and has passed the preshipment testing requirements, it is lifted back through the access bay and returned to its horizontal position on the shipping vehicle. The cask holddown mechanisms are secured, the cask top and bottom impact limiters are replaced, and all appropriate appurtenances are reconnected. Shipping papers are completed and the cask is removed from the FHB and released for shipment. 9.1.4.2.4 comoonent Descrintion: 9.1.4.2.4.1 Refueline Machine - The refueling machine (Figure 9.1.4-1) l is a rectilinear bridge and trolley crane with a vertical mast extending down into the refueling cavity. The bridge spans the refueling cavity and runs on rails set into the edge of the refueling cavity. The bridge and trolley motions are used to position the vertical mast over a fuel assembly in the core. A long tube, with a pneumatic gripper on the end, is lowered down out of the mast to grip a fuel assembly. The gripper tube is long enough so that 9.1 17 Revision 0

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O STPECS UFSAR the upper end is still contained in the mast when the gripper end contacts the fuel. A winch mounted on the trolley raises the gripper tube and fuel assembly up into the mast tube. The fuel is transported while inside the mast tube to its new position. The refueling machine gripper tube also includes a secondary gripper mechanism (located above the primary fuel assembly gripper) to remove and insert a thimble plug into a new fuel assembly. The refueling machine drive uses a direct current control device that gives stepless, variable speeds from zero to ful1 speed. All controls for the refueling machine are mounted in a console on the trolley. The bridge and trolley are positioned by a servo system in relation to an X-Y coordinate grid pattern referenced to the reactor core. Bridge and trolley position is indicated by an electric position-repeat-back system. Readsut dials are read directly by the operator at the console. The drives for the bridge, trolley, and winch are variable speed and include a separate inching control on the bridge and trolley. The maximum speed is 60 ft/ min-for the bridge and 20 ft/ min for the trolley and hoist. An auxiliary monorail hoist on the refueling machine uses a two step magnetic controller to give hoisting speeds of approximately 7 ft/ min and 20 ft/ min for use in handling accessory equipment. Electrical interlocks and limit switches on the bridge and trolley drives prevent damage to the fuel assemblies. The wine.h is also provided with redundant limit switches plus a mechanical stop to prevent a fuel assembly from being raised above a safe shielding depth should the limit switch fail. i In an emergency, the bridge, trolley, and winch can be operated manually using a handwheel on the motor shaft. The refueling machine is provided with a television system which permits viewing of all fuel assembly positions and fuel movements to within 6 in of the top of the core. This system includes a telescoping boom, a monitor, and a videotape recorder. The refueling machine is designed to permit failed fuel detection by means of a sip test. A fuel assembly will be withdrawn inside the mast and held stationary above the core for a period of time. Doors at the bottom of the mast are shut to enclose the fuel assembly completely. The water surrounding the fuel assembly will then be circulated through detection equipment located on the crane trolley. Any activity increase will be evaluated to determine the severity of cladding failure. 9.1.4.2.4.2 Fuel Handline Machine - The fuel handling machine consists of an electric monorail hoist carried on a wheel mounted bridge (Figure 9.1.4-2), which spans the SFP, fuel transfer canal, and cask loading pool. The fuel handling machine is used exclusively for handling fuel assemblies and core components by means of handling tools suspended from the hoist. The hoist travel and tool ler.gths are designed to limit the maximum lift of a fuel assembly or core component to a safe shielding depth. The fuel handling machine has a two-step magnetic controller for the bridge and hoist. The bridge speeds are 10 ft/ min and 30 ft/ min, and the hoist speeds are 7 ft/ min and 20 ft/ min. A hydraulic coupling is used in the bridge 9.1-18 Revision 0

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• i STPEGS UFSAR drive to limit-starting acceleration. The hoist trolley.is manually positioned along the monorail by a chainfall.

9.1.4.2.4.3 New Fuel Elevator - The new fuel elevator (Figure 9.1.4-3) ' consists of a box-shaped elevator assembly with its top and open and sized to < house one fuel assembly. The new fuel elevator is normally used to lower a new fuel assembly into the g fuel transfer canal, where the fuel handling machine can transport it to the FTS equipment for transfer into the Containment. The new fuel elevator can also be used to raise's new or spent fuel assembly provided administrative ' controls and procedures are utilized. 9.1.4.2.4.4 Fuel Transfer System - The FTS (Figure 9.1.4-4) includes an underwater, electric-motor-driven transfer car that runs on tracks extending - from the refueling canal in the RCB through the fuel transfer tube and into j the fuel transfer canal in the FHB, and a hydraulically actuated lifting arm (upender) at each end of the transfer tube. In the refueling canal the , fuel container mounted on the transfer car receives a fuel assembly in the vertical position from the refueling machine. The upender then lowers the fuel assembly to a horizontal position for passage through the transfer tube. After passing through the tube, the fuel container is raised to a vertical position by the other upender for removal of the fuel assembly. The fuel handling machine lifts the fuel assembly out of the fuel container and moves it to the desired storage position in the spent fuel storage racks. The transfer car is driven by a pusher arm connected;to two contin >ous roller i chains. The roller chains are driven by an electric motor mounted near the operating floor of the FHB next to the fuel transfer canal. They are connected to the chain-drive sprockets by a vertical drive shaft. The fuel container is center-pivoted, and the pivot structure serves to attach the container to the transfer car so they move together as a single unit. There - is a mechanical stop at each end of the rail on which the car travels. Also,  ; there is another mechanical stop for the vertical limit of travel when the fuel container is upended. Each of the upender If.fting arms has its own hydraulic power unit and control console located on the operating floor of its-respective side of the Containment wall. - During reactor operation, the transfer car is. stored in the fuel transfer canal. ' A blind flange is bolted on the refueling canal end of the transfer tube to seal the Reactor Containment. The end of the tube in the FHB'is closed by a gate valve. 9.1.4.2.4.5 Rod Cluster Control Channe Fixture'- The RCC change fixture is located in the in-Containment fuel storage pit. The change fixture is used i for periodic RCC element inspections and for transfer of the RCC elements from l one fuel assembly to another, or transfer of secondary source assemblies from one fuel assembly to another (Figure 9.1.4 5). During rapid refueling i operations, the RCC change fixture is not required for transfer of RCC elements since the RCC elements are removed from the reactor core along with - the upper head package. The major subassemblies which constitute the change i fixture are the frame and track structure, the carriage, the guide tube,'the gripper, and the drive mechanism. The carriage is a moveable container 1 supported by the frame and track structure. The tracks provide a guide for  ; the four flanged carriage wheels and allow horizontal movement of the carriage  ! during the changing operation. The positioning stops on both the carriage and frame locate each of the three carriage compartments directly below the guide 1 9.1-19 Revision 4 1 I ei _ __ _ _ _ . _ _ _ _ _ _ _ _m .,. , - - ._ _

a STPEGS UFSAR i tube. Two of these compartments are designed to hold individual fuel assemblies while the third is made to support a single core component (e.g., RCCA or secondary source assembly). Situated above the carriage and mounted - on the in-Containment fuel storage pit wall is the guide tube. The guide tube provides proper orientation of the gripper and core component as.they are , being raised and lowered. The gripper is a pneumatically actuated mechanism which engages the core component. It has two flexure fingers which can be 7 inserted into the top of the core component when air pressure is applied to  : the gripper piston. Normally, the fingers are locked in a radially extended  !' position. Mounted on the operating deck is the drive mechanism assembly, which is composed of the manual carriage drive mechanism, the revolving stop operating handle, the pneumatic selector valve for actuating the gripper , piston, and the electric hoist for elevating the gripper. . The pneumatic gripper and winch life the core component out of a spent fuel assembly and up into the guide tube. Then by repositioning the carriage, a new fuel assembly is brought under the guide tube and the gripper lowers and releases the core component into the new fuel assembly. The refueling machine loads and unloads the fuel assemblies into and from the carriage. 9.1.4.2.4.6 Soent Fuel Assembiv Handline Tool - The spent fuel assembly handling tool (Figure 9.1.4-6) is used to handle new and spent fuel assemblies in the spent fuel pit. It is a manually actuated tool, suspended from the fuel handling machine. The gripping portion of the tool consists of a mechanical finger which engages the fuel assembly top nozzle when actuated. The operating handle to actuate the fingers is located at the top of the tool. When the fingers are latched, a pin is inserted into the operating handle, to prevent the fingers from being accidentally unlatched during fuel handling .. operations. 9.1.4.2.4.7 New Fuel Assembiv Fandline Tool - The new fuel assembly

                 - handling tool (Figure 9.1.4-7) is used to lift and transfer fuel assemblies
                 . from the new fuel shipping containers to the new fuel storage racks, to transfer fuel assemblies from the new fuel storage racks to the new fuel-elevator, or to directly place new fuel in the ~ SFP (Cycle 1 only). It is a manually actuated tool, suspended from the FHB overhead crane, which uses four cam actuated latching fingers to grip the underside of the fuel assembly top nozzle. The operating handles which actuate the fingers are located on the side of the _ tool. When the fingers are latched, a safety screw is turned in to prevent the accidental unlatching of the fingers.

9.1.4.2.4.8 Reactor Vessel Head and Unoer Internals Liftine Device -

l. The reactor vessel head and upper internals lifting device consists of a f

welded and bolted structural steel frame with suitable rigging to enable the crane operator. to lift the head or the head and upper internals for storage during refueling operations. The lifting device normally remains attached to the reactor vessel head during plant operation. The missile shield and the control rod drive mechanism cooling shroud are attached to the head lifting device. L 9.1.4.2.4.9 Reactor Internals Liftine Device - For rapid refueling, the upper internals are normally lifted in one lift as part of the upper head l package. For nonrapid refueling, the head may be separated from the upper

internals and lifted separately. In this case, the reactor internals lifting rig (Figure 9.1.4-8) is used to remove the upper internals, 9.1-20 Revision 0

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STPECS UFSAR 9.1.4.2.4.10 Feactor VesJe1 Stud Tensioner - The stud tensioners (Figure 9.1.4 9) are employed to secure or release the head closure joint at every refueling. The stud tensioner is a hydraulically operated device that uses oil as the working fluid. The device permits preloading and unloading of the reactor vessel closure studs at cold shutdown conditions. Stud tensioners minimize the time required for the tensioning or unloading operation. Three tensioners are provided and are applied simultaneously to three studs located 120 degrees apart. A single hydraulic pumping unit operates the tensioners, which are hydraulically connected in series. The studs are tensioned to their operational load in two steps to prevent high stresses on the reactor vessel flange region and unequal loading of the studs. Relief valves on each tensioner prevent overtensioning of the studs due to excessive pressure. 9.1.4.2.4.11 Seent Fuel Cask - The design of cask storage and cask handling facilities is based on a design cask wei hing 5 up to 150 tons and measuring approximately 21 ft long by 10 ft in diameter. The shipping vehicle (rail car or truck) will take the cask to and from the FHB and will be equipped with a cask-cooling system and a storage area for the cask yoke. Thu FHB arrangement and cask handling equipment are designed to preclude the occurrence of any accident to a loaded spent fuel shipping cask beyond th: regulatory specified design accident conditions for the cask. Overland offsite transportation of the cask will conform to transportation rules and re6ulations, 49CFR173.

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9.1.4.2.4.12 Fuel Handline Buildine Overhead Crane - This 15/2-ton overhead crane runs over the entire FHB area. Travel of this crane is shown on Figure 9.1.4 13. The design services of this crane include the following:

1. Transfer of new fuel assembly shippin6 containers from the shipping vehicle (truck or rail car) to the new fuel handling area.

2. Transfer of new fuel assemblies from new fuel shipping containers to new fuel elevator, new fuel storage pit. or SFP (cycle 1 only).

3. Transfer of spent fuel shipping cask head from the cask to temporary storage on the FHB operating floor and then back onto the shipping cask when the fuel loading is complete.

4. Replacement of safety injection and Containment spray pumps and SFP cooling HXs.

5. Removal and replacement of pool gates for fuel transfer operations.

9.1.4.2.4.13 New Fuel Handline Area Overhead Crane - This 5 ton overhead crane is used exclusively to handle new fuel assemblies and their shipping containers in the new fuel handling area. , 9.1.4.2.5 Industrial Codes and standards: The following industrial ' codes and standards are used in the design of fuel handling equipment.

1. Cranes: Crane Hanufacturers Association of America (CMAA) specification l No. 70, Class A-1.

2. Structural: ASME Code, Section III, Appendix XVII.

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O STPEGS UFSAR

3. Electrical: Applicable standards and requirements of the National Electric Code, National Fire Protection Association No. 70, and National Electrical Manufacturers Association standards MCI and ICS are used in the design, installation, and manufacturing of all electrical equipment.

4 Materials: Materials conformed to the specifications of the American Society for Testing Materials standard.

5. Safety: The design meets the applicable requirements of Section 1910.179 of Subpart N of the Occupational Safety and Health Act Code.

6. Others: American Institute of Steel Construction; American National Standards Institute; American Society of Testing Materials; Institute of Electrical & Electronic Engineers; National Electric Manufacturers Association; Occupational Safety & Health Administration; American Welding Society; Expansion Joint Manufacturers Association; ASME B&PV Code Sections VIII and XI; American Concrete Institute; Hydraulic i Institute Standards. ]

9.1.4.3 Safety Evaluation. Design of the FHS in accordance with RC 1.13 and with CDCs 2, 5, 61, and 62 ensures a safe condition under normal and postulated accident conditions. l 9.1.4.3.1 Safe Handline:

              .            9.1.4.3.1.1 Desien Criteria for the Refueline Machine and the Fuel-           )

Handline Machine - The primary design requirement of the machine is reliability. A conservative design approach is used for all load-bearing parts. Where practicable, components are used that have a proven record of  ! reliable service. Throughout the design, consideration is given to the fact I that the machine will spend long idle periods stored in an atmosphere of 80'F and high humidity. In general, the crane structure is considered in the Class A1, Standby Service, as defined by CHAA Specification No. 70. l 1 All components critical to the operation of the machine and parts which could fall into the reactor are positively restrained from loosening. 9.1.4.3.1.2 Refueline Machine - The refueling machine design includes the following provisions to ensure safe handling of fuel assemblies:

1. Electrical Interlocks The electrical interlocks which ensure safe operation of the crane are designed to meet single-failure criteria.

2. Bridge, Trolley, and Hoist Drive Mutual Interlocks I l

Bridge, trolley, and winch drives are mutually interlocked, using redundant interlocks to prevent simultaneous operation of any two drives, and therefore can withstand a single failure. l i l . 9.1-22 Revision 0

i STPECS UFSAR  !

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3. Bridge Trolley Drive Cripper Tube Up Bridge and trolley drive operation is prevented except when th;

• gripper ]

tube up" position switches are actuated. The interlock is redundant and can withstand a single failure.

4. Cripper Interlock An interlock is supplied which prevents the opening of a solenoid valve in the air line to the gripper except when there is less than 600 pounds of suspended weight indicated on a load cell. As backup protection for this interlock, the mechanical weight actuated lock in the gripper prevents opening of the gripper under load even if air pressure is applied to the operating cylinder. This interlock is redundant and can withstand a single failure.

5. Excessive Susper.W .eight An excessive suspended weight load cell limit switch and a backup i deflection actuated limit switch prohibit raising the guide tube if the  !

load suspended from it significantly exceeds the weight of the gripper, a fuel assembly, and an RCCA. The interlock is intended to prevent ~ inadvertent damage to a fuel assembly or adjacent components if the assembly becomes stuck during its removal.

6. Hoist-Gripper Position Interlock An interlock in the hoist drive circuit in the up direction permits the hoist to be operated only when either the open or closed indicating switch on the gripper is actuated. The hoist gripper position interlock i 1

consists of two separate circuits that work in parallel such that one l circuit must be closed for the hoist to operate. If one or both interlocking circuits fail in the closed position, an audible and visual alarm on the console is actuated. The interlock, therefore, is not redundant but can withstand a single failure since both an interlocking circuit and the monitoring circuit must fail to cause a hazardous condition. j

7. Bridge and Trolley Hold Down Devices i Both the refueling machine bridge and trolley are horizontally restrained on the rails by two pairs of guide rollers, one pair at each wheel location on one truck only. The rollers are attached to the bridge truck and contact the vertical faces on either side of the rail to prevent horizontal movement. Vertical restraint is accomplished by antirotation bars located at each of the four wheels for both the bridge and trolley. The antirotation bars are bolted to the trucks and extended under the rail flange for the bridge restraint; the trolley restraints extend beneath the top flange of the bridge girder which supports the trolley rail. Both horizontal and vertical restraints are adequately designed to withstand the forces and overturning moments resulting from an SSE.

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8. Design Load The desi5 n load for structural components is their deadweight plus 5,700 pounds (three times the weight of a fuel assembly and RCCA).

9. Main Hoist Braking System The main hoists are equipped with two independent braking systems. A solenoid-release, spring-set electric holding brake is mounted on the motor shaft. The brake operates normally to release upon application of current to the motor and to set when current is interrupted. The second brake is a mechanically actuated load control brake internal to the hoist gear box; it sets if the load starts to overhaul the hoist. It is necessary to apply torque from the motor to raise or lower the load. In raising the motor cams, the brake opens; in lowering, the motor slips the brake, allowing the load to lower. The brake actuates upon loss of torque from the motor for any reason and is not dependent upon any electrical circuits. On the main hoist, the motor brake is rated at 350 percent operating load and the mechanical brake at 300 percent.

The Main Hoist Braking System is supplied with redundant paths of load support such that failure of any one component will not result in free fall of the fuel assembly. Two wire ropes are anchored to the winch drum and carried over independent sheaves to a load equalizing mechanism on the top of the gripper tube. In addition, supports for the sheaves and equalizing mechanism are backed up by passive restraints to pick up the load in the event of failure of this primary support. Each cable system is designed to support 13,750 pounds or 27,500 pounds acting  ; together. , 10. Hoist Down Limit A geared limit switch on the main hoist prevents lowering the gripper 4 tube significantly below the position that would normally engage fuel assembly in the reactor core. The working load of fuel assembly, RCCA, and gripper is approximately 2,-850 pounds. The gripper itself has four fingers gripping the fuel, any two of which will support the fuel assembly weight. The gripper mechanism contains a spring actuated mechanical lock which prevents the gripper from opening unless the gripper is under a compressive load. The refueling machine gripper and hoist system are routinely load tested prior to refueling operations in accordance with the surveillance requirements of the Technical Specifications. 9.1-24 Revision 0

1 STPEGS UFSAR l 9.1.4.3.1.3 Fuel-Handline Machine - The fuel handling machine includes j the following safety features:

1. The bridge and hoist controls are interlocked to prevent simultaneous operation of both the bridge drive and the hoist. The interlocks are redundant and can withstand a single failure.

2. / redundant overload protection device is included on the hoist to limit the uplift force which could be applied to the spent fuel storage racks.

The protection device limits the hoist load to 125 percent (5,000 pounds) of the rated 2-ton hoist capacity. This device can withstand a single failure. A load monitoring device is provided between the hoist and spent fuel handling tool. By monitoring drag load when raising or lowering assemblies, it can be determined immediately if the assembly is hanging or experiencing unusual loads.

3. The design load on the hoist is the weight of one fuel assembly and RCCA (1,900 pounds), one failed fuel container (1,700 pounds), and the tool (400 pounds), which gives it a total weight of approximately 4,000 pounds.

4. Restraining bars are provided on each truck to prevent the bridge from overturning.

9.1.4.3.1.4 Fuel Transfer System - The following safety features are f provided for in the FTS: , 1

1. Transfer Car Permissive Switch l 1

The transfer car controls are located in the spent fuel pit area; therefore, conditions in the Containment are not visible to the i operator. The transfer car permissive switch allows a second operator in the Containment to exercise some control over car movement if visible conditions warrant it. Transfer car operation is possible only when both upenders are in the down position as indicated by the limit switches. The permissive switch is a backup for the transfer car upender interlock. Assuming the fuel container is in the upright position in the Containment and the upender interlock circuit fails in its permissive condition, the operator in the spent fuel pit area still cannot operate the car because of the transfer car permissive switch interlock. The interlock, therefore, can withstand a single failure.

2. 1.ifting Arm Transfer Car Position Two redundant interlocks allow lifting arm operation only when the transfer car is at the respective end of its travel and therefore can withstand a single failure.

Of the two redundant interlocks that allow lifting arm operation only when the transfer car is at the end of its travel, one interlock is a position limit switch in the control circuit. The backup interlock is a mechanical latch device on the lifting arm that is opened by the car moving into position. 9.1-25 Revision 0

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3. . Transfer Car Valve Open Two redundant interlocks on the transfer tube valve permit operation of the transfer car only when the transfer tube valve position switch indicates the valve is fully open and, therefore, can withstand a single failure.

4. Transfer Car Upender The transfer car upender interlock is primarily designed to protect the equipment from overload and possible damage if an attempt is made to move the car when the fuel container is in the vertical position. This interlock is redundant and can withstand a single failure. The basic interlock is a position limit switch in the control circuit. The backup interlock is a mechanical latch device that is opened by the weight of the fuel container when in the horizontal position.

5. Upender Refueling Machine The refueling canal upender is interlocked with the refueling machine.

Whenever the transfer car is located in the refueling canal, the upender cannot be operated unless the refueling machine mast is in the fully retracted position or the machine is over the core. 9.1.4.3.1.5 Fuel Handline Tools and Eculement - All fuel handling tools and equipment which are used over the open reactor vessel are designed to prevent inadvertent decoupling from crane hooks; i.e., lifting rigs are pinned to the crane hook and safety latches are provided on hooks supporting tools. Tools required for handling internal reactor components are designed with fail-safe features that prevent disengagement of the component in the event of operating mechanism malfunction. These safety features apply to the following tools:

1. Control Rod Drive Shaft Unlatching Tool The air cylinders actuating the gripper mechanism are equipped with backup springs which close the gripper in the event of loss of air to the cylinder. Air valves are equipped with safety locking rings to prevent inadvertent actuation.

2. Spent Fuel Handling Tool When the fingers are latched, a pin is inserted into the operating handle to prevent inadvertent actuation. The tool weighs approximately 400 pounds and is preoperationally tested at 125 percent of the weight of one fuel assembly and RCCA (1,900 pounds).

3. New Fuel Assembly Handling Tool When the fingers are latched, a safety screw is inserted to prevent inadvertent actuation. The tool weighs approximately 100 pounds and is preoperationally tested at 125 percent of the weight of one fuel assembly and RCCA (1,900 pounds).

9.1-26 Revision 0

l

        ,o                                                                                             -

o STPEGS UFSAR 9.1.4.3.1.6 overhead Cranes - Overhead cranes used in refueling and fuel handling operations include the polar crane (417/15-ton, Unit I and 500/15 ton, Unit 2), the 150-ton cask handling crane, the 15/2 ton FHB crane, and the 5-ton new fuel handling area crane. These cranes are classified as non nuclear safety (NNS) Class since they neither provide nor support any safety system function. However, during and after a seismic event, the cranes and their supports are designed to retain structural integrity and prevent collapse and damage to safety-related equipment and structures. Operability need not be retained. A report under separate cover has been submitted to the Nuclear Regulatory Commission (NRC) concerning control of heavy loads. This report contains details of crane / load combinations and the safeguards that prevent damage to spent fuel, the reactor core, equipment required for safe shutdown, and decay heat removal. l l

1. Polar Crane The polar crane is used for general handling operations in the containment during refueling. These operations include:

a. Removal of the upper package

b. Removal of pumps, pump motors, and heat exchangers

c. Handling of pool gate

d. Handling of inservice inspection (ISI) rig

e. Movement of hatch covers.

A head drop analysis is discussed in Letter NS CE-1101 (June 11,1976) and received NRC approval by letter on November 30, 1976. This crane is provided with seismic restraints to prevent derailment in the event of an SSE.

2. Cask Handling Overhead Crane .

This 150 ton crane is provided for handling the spent fuel shipping cask. Crane design and building arrangement preclude travel of this crane over the SFP: consequently, the shipping cask cannot be lifted or dropped over the spent fuel racks. This crane is designed to maintain its structural integrity and hold its load under the dynamic loading conditions of the SSE. Building arrangement and lifting rig design prevents this crane from lifting the cask higher than 30 f t above the floor. The spent fuel cask drop accident is discussed in Section 15.7.5. 9.1 27 Revision 0 = _. _ ___

0 STPECS UFSAR

3. Fuel Handling Building Overhead Crane The 15/2-ton capacity crane is.co be used for general handling operations in the FHB. These operations include:

a. Movement of new fuel assemblies

b. Removal of pumps and heat exchangers

c. Handling of the pool gates

d. Movement of the cask head

e. Movement of hatch covers This crane is designed to maintain its structural integrity under the dynamic loading of the SSE. The crane will retain its load under such dynamic loadings.

This crane main hoist is also provided with a redundant reeving system. With this redundancy, the crane can withstand a single failure without dropping its load and therefore meets the intent of RG 1.104 A more detailed description of compliance of the 15/2-ton FHB crane with RC 1.104 is given in Table 9.1-3. 4 New Fuel Handling Area Overhead Crane The 5 ton naw fuel handling area overhead crane is used for movement of new fuel assemblies within the new fuel handling area. Dropping of new fuel assemblies due to SSE induced dynamic loading of the crane will'not result in an offsite radiological hazard. The crane travels over no

               . safety related equipment.

9.1.4.3.2 seismic considerations: The safety classifications for all fuel handling and storage equipment are listed in Table 3.2.B 2. SC 1, 2, and 3 equipment is. designed to withstand the effects of an SSE without loss of capability to perform its safety function. Further, the combined normal and SSE stresses are limited to the allowable stresses as defined by ASME Code, Section III, Appendix XVII-2110. SC 1 and 2 equipment is designed to withstand the forces of an Operating Basis Earthquake-(OBE), with the combined normal and OBE stresses being limited to the allowable stresses, as defined by ASME Code, Section III, Appendix XVII. For SC 3 equipment, consideration is given to the CBE only insofar as failure of the SC 3 equipment might adversely affect SC 1 or 2 equipment. i For NNS equipment, design for the SSE is considered if failure might adversely affect a SC 1, 2, or 3 component. Design for OBE is considered if failure of the NNS component might adversely affect an SC 1 or 2 component. 9.1.4.3.3 containment Pressure Boundary Inteerity: The fuel transfer tube, which connects the refueling canal (inside the RCB) and the SFP (outside the Containment), is closed on the refueling canal side by a blind flange at all times except during refueling operations. Further discussiot. on the fuel transfer tube can be found in Section 3.E.2.1.3.3. 9.1 28 Revision 0

l l . \ STPECS UFSAR j l i 9.1.4.3.4 Radiation Shieldine: During all phases of spent fuel transfer, the gamma dose rate at the surface of the water is 2.5 mR/hr or less. This is accomplished by maintaining approximately 10 ft of water above the top of the fuel assembly during all handling operations, i l The two machines used to lift spent fuel assemblies are the refueling machine and the fuel handling machine. The refueling machine contains positive stops , that prevent the top of a fuel assembly from being raised to within i

                                                                                                     )

The approximately 10 ft of the normal water level in the refueling cavity. hoist on the fuel handling machine moves spent fuel assemblies with a lory. 3 handled tool. Hoist travel and tool length likewise limit the maximum lift of a fuel aseembly to within approximately 10 ft of the normal water level in the SFP. 9.1.4.4 Tests and Insoections. As part of normal plant operations, the fuel handling equipment is inspected for operability before each refueling operation. During the operational testing of this equipment, procedures are followed that will verify the correct performance of the FHS interlocks. 9.1.4.5 Instrumentation Recuirements. A description of the instrumentation and controls is provided in Section 9.1.4.3 for the refueling machine, the fuel handling machine, the FTS, and the SFCHS. I 4 4 4 j 't 9.1 29 Revision 0 4 W w

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TABLE 9.1-1 g-5- SPENT FUEL POOL HEATUP ANALYSIS RESULTS FOR 18-MONTH RELOAD CYCLES 9 e

    -*                                                                                                                                       ~

MXIfRSI POOL ~ MXI9EJM POOL" MAXIMUM , h w SPENT FUEL FUEL LOAD TYFICAL MXIMUM ALLOWABLE POOL TEpr. TEpFERATURE POOL MODE (SRP 9.1.3 Criterie) STP FUEL LOAD TEDFERATURE HEAT LOAD WITM ! (SRP 9.1.3 Criterie) (1 COOLING TRAIN) (2 COOLING TRAINS) 2 COOLING TRAINS

                                                                                                                                                                                                                                                                               - (10' Stu/hr)t:)

SRP 1 LOAD a 150 MRS

           , REQUIREMENT         1 LOAD a 1 YR                                         N/A                                                  140*F                                           142*F                                    124*F                                            23.6 NORML           1 LOAD E 400 DAYS MAXIMUM          (1 LOAD = 88 ASSEMBLIES) j                 STPEGS                                              88 ASSEMBLIES 8150 MS
               . NORMAL                       N/A                    88 ASSEMBLIES S 1 YEAR                                                 140*F                                           155'F                                    131*F                                            31.4 MAXIMUM                                              88 ASSE95 LIES + PREVIOUSLY DISCMAGED a 2% - 28 YRS"3 STPEGS                                              1 FULL CORE a 100 MR$t u RAPID                        N/A                   88 ASSEMBLIES a 16 MONTHS                                               N/A                                            200*F                                    154'F                                            61.2-REFUELING                                              88 ASSEf5 LIES + PREVIOUSLY DISCMRGED 8 2n - 30 YEARS"3                                                                                                                                                                                                                       y m

SRP 1 FULL-CORE a 150 MRS tn 9 REQUIREMENT 1 LOAD a 36 DAYS N/A NO BOLLING tui 190'F 149'F 53.6 O E ABNORML 1 LOAD a 400 DAYS l MMIMUM (1 LOAO = 88 ASSEMBLIES) c

                                                                                                                                                                                                                                                                                                   *n ta                                                                                                                                                                           en STPEGS                                             1 FULL-CORE a 150 INts )

ABNORML N/A 88 ASSEMBLIES S 36 DAYS NO BOILING W3 200*Ft n 153*F 61.8 b i MAXIMUM 88 ASSEMBLIES 8 16 MONTHS 88 ASSEMBLIES

• PREVIOUSLY DISCMAGED e 2 30 YRS"3 i

FOOTNOTES:

1. ABBREVIATIONS: SRP = STANDARD REVIEW PLAN (NUREG-0000), EFPM = EFFECTIVE FULL POER MOURS OF FUEL BURNUP.

2. THE MAXIMUM MEAT LOAD IS BASED ON FUEL EXPOSURE OF 40000 EFPM AND SRP ASB 9-2 DECAY MEAT FORDRALATIONS. SEE FOOTNOTES 3 & 4 FOR EXCEPTIONS. f

! 3. FOR THE RAPID REFUELING ABNORMAL MAXIMUM CASE, THE FULL CORE BURNUP ASSUMED FOR THE 88 NEW ASSEMBLIES IS 16 MONTMS AND FOR THE OTHER 105 ASSEMBLIES IT IS 40000 EFPH.

4. FOR THE STPEGS ABNORML MAXIMUM CASE, THE FULL CORE BURNUP ASSUMED FOR TME 88 NEW ASSEMBLIES IS 36 DAYS AND FOR THE OTHER 105 ASSEMBLIES IT IS 40000 EFPH. .

5. THIS CONFIGURATION MAINTAINS FULL CORE DISCMARGE CAPABILITY,' f.e.,1776 FUEL ASSEMBLIES IN THE SPENT. FUEL POOL AND 193 LOCATIONS AVAILABLE.

i (FOOTNOTES CONTINUED ON NEXT PAGE) h

TABLE 9.1-1 (CONTINUED) . E 8 h h 6. ALL FUEL STORAGE LOCATIONS ARE FILLED WITH SPENT FUEL, i.e., 1969 FUEL ASSEMBLIES ARE IN THE SPENT FUEL POOL.

7. fu THE EVENT OF FIRE OR MODERATE ENERGY LINE BREAK IN THE FUEL HANDLING BUILDING THAT DISABLES BOTH TRAINS OF SPENT FUEL POOL COOLING, THE SPENT FUEL POOL MAY EVENTUALLY BOIL. MAKEUP WATER CAN BE PROVIDED VIA THE REACTOR MAKEUP PUMPS. IN ADDITION, SPENT FUEL POOL MAKEUP WATER CAN ALSO BE SUPPLIED USING LOCAL HOSE STATIONS IN THE FUEL HANDLING 90lLDING. DETAILS PROVIDED IN SECTION 3.3 OF THE FIRE HAZARDS ANALYSIS REPORT.

8. IF BOTH REACTOR MAEUP WATER Pt.MPS ARE UNAVAILABLE BECAUSE OF FLOODING IN THE MECHANICAL AUXILIARY BUILDING, THE SEISMIC CATEGORY - 1 LOW HEAD SAFETY INJECTION PUMP WOULD BE AVAILABLE FOR REFILLING THE SPEN 7 FUEL POOL SY CONNECTING TEMPORARY MOSES TO THE VENT AND DRAIN VALVES LOCATED ON LOW HEAD SAFETY INJECTION PUMP DISCHARGE PIPING SO THAT REFUELING WATER COULD BE DELIVERED TO THE SPENT FUEL POOL.

9. THE 200* F MAXINUM SPENT FUEL POOL BULK WATER TEMPERATURE IS BASED ON THE FOLLOWING TWO CASES:

CASE #1. FULL-CN E DISCHARGE (AS IN NOTE 4) 150 HOURS AFTER SHUTDOWN WITH TWO COMPONENT COOLING WATER TRAINS SUPPLYING COOLING WATER TO ONE SPENT FUEL POOL HEAT EXCHANGER. CASE #2. FULL-CORE DISCHARGE (AS IN NOTE 4) 175 HOURS AFTER SHUTDOWN WITH ONE COMPONENT COOLING WATER TRAIN SUPPLYING COOLING WATER TO ONE SPENT FUEL POOL HEAT EXCHANGER.

                                                                                                                                                                                                   ;a e                                                                                                                                                                                                 e e

a e E E A L

1

                                                                                            \

STPECS UFSAR l TABLE 9.1-2 4 SPENT FUEL POOL COOLING AND CLEANUP SYSTEM COMPONENT DESIGN PARAMETERS 1 i Spent Fuel Pool Cooling Pump l 2 l Number 150 l Design pressure, psig l 200 Design temperature, 'F 2,500 Design flow, gal / min l Design head, ft 200 Material Stainless steel l Spent Fuel Fool Skimmer Pump l 1 1 Ndmber I 150 i Design pressure, psig

                                     *F                                200                  J Design temperature 100 Design flow, gal / min 50 Design head, ft Haterial                                                 Stainless steel Refueling Unter Purification Pump 1

Number 150 Design Pressure, psig 200 Design temperature, *F 200 Design flow, gal / min 200  ; Design head, ft Material Stainless steel 9.1-31 Revision 0

t i i i e STPECS UFSAR TABLE 9.1-2 (Continued) SPENT FUEL POOL COOLING AND CLEANUP SYSTEM l COMPONENT DESIGN PARAMETEE1 i

          ' Spent Fuel Pool Heat Exchanger Number                                                  2 l                                                                                                          ,

Design heat transfer, Beu/hr 9.1 x 10 8 Shell Iub.s Design pressure, psig 150 150 Design temperature, *F 200 200 Design flow, Ib/hr 1.5 x 10' 1.4 x 10' l: Inlet temperature. *F 105 120 Outlet temperature, 'T 111.1 113.6 Fluid circulated Component Spent fuel cooling cooling - water water Material Carbon steel Stainless steel Spent Fuel Pool Demineralizer i Number 2 l Design pressure, psig 300 . Design temperature, 'T 250 Design flow, gal / min 250 Resin volume, fts 75 Material Austenitic j stainless steel l l l l l I i l 9.1 12 Revision 0 i me6

        ,     , , . , ,       ,                                                            .,n..,

STPECS UFSAR j l TABLE 9.1-2 (Continued) SPENT FUEL POOL COOLING AND CLEANUP SYSTEM  ; COMPONENT DESIGN PARAMETERS I l l \ i Spent Fuel Fool Filter 2 Number l 300  ; Desi5n pressure, psig I 250 l Design temperature, 'F l 250 Design flow, gal / min Filtration requirement 984 retention of particles above 5 l microns l Material, vessel Austenitic l stainless steel l Spent Fuel Pool Skimmer Filter 1 Number 300  ; Design pressure, psig 250 Design temperature. 'F l 250 i Design flow, gal / min ) Filtration requirement 984 retention of j particles above 5 i microns Material, vessel Austenitic stainicss steel Spent Fuel Pool Strainer 2 Number Design pressure, psig Not applicable 200 Design temperature, 'F Design flow, gal / min 5.000 Material Austenitic stainless steel 9.1-33 Revision 0 i

     . .o o

STPEGS UFSAR TABLE 9.1-2 (Continued) SPENT TUEL POOL COOLING AND CLEANUP SYSTEM COMPONENT DESIGN PARMETERS Spent Fuel Pool Skimmer / Strainer Assembly i Number 2 Design pressure, psig 50 Design temperature. *F 200 Design flow, gal / min 50 Material Austenitic stainless steel In Containment Storage Area Strainer Number 1 Design prersure, psig Not applicable Design temperature. *F 200 , l Design flow, gal / min 2,500 Material Austenitic stainless steel Reactor Cavity Filtration System . 2 Number Design Pressure, psig 150 Design temperature 'F 200 Design flow, gal / min 250 Design head of pump, ft 48 Filtration requirement of filters 10 micron Material Austenitic stainless steel 9.1-34 Revision 0 1

s* e STPECS UFSAR TABLE 9.1-3 THB 15/2-TON CRANE - COMPLIANCE WITH REGULATORY GUIDE 1.104 STPECS Meets the STPEGS Intent of Rev. O STPEGS Not $pplicable by Complies and Complies with Takes Either Rev. O or Re5ulatory with a Proposed Revision Exception the Proposed Revision Dated January 1978 to Rev. O Dated January 1978 Position Rev. O X C.1.a 1.b(1) X X (2) X (3) X (4) 1.c X 1.d X 1.e X 1.f X C.2.a X 2.b X 2.c X 2.d X C.3.a X 3.b

    .                         X                                                                        !

3.c X 3.d X 3.e X " 3.f X 3.g X 3.h X 3.1 X 3.j X .

                                                                                    ~

3.k X 3.1 X 3.m X 3.n X 3.o Note 1 3.p X 3.q X 3.r X 3.s Note 2 X 3.t 3.u X

1. Controlled plugging measures shall be provided se that if the operator  !

reverses a drive while it is in motion, the torque during reverse shall l Se automatically controlled to a predetermined torque limit during deceleration.

2. The crane-is designed to CMAA standards (i.e, a 5:1 minimum factor of j safety for each component) for 15-ton lifts. l 1

9.1 35 Revision O

                                                                                                 ~

I w .. _ , - -

                                                                                                   -- )

i STPEGS UFSAR TABLE 9.1-3 (Continued) FHB 15/2. TON CRANE - COMPLI ANCE VITH REGiflATORY CUIDE 1.104 STPEGS Heets the STPEGS Intent of Rev. O STPEGS Noi Applicable by Complies and Complies with Takes Either Rev. O or Regulatory with a Proposed Revision Exception the Proposed Revision Dated January 1978 to Rev. O Dated January 1978 Position Rev. O C.4.a X 4.b X 4.c X X 4.d C.5.a X 5.b X

                                                                           ~

i l l l l l 9.1 36 Revision 0

e s. STPECS UFSAR TABLE 9.1-4 , NSSS VENDOR RECOMMENDED SPECIFICATIONS AND GUIDELINES i POR SPENT FUEL POOL VATER PURITY 1 Soecification Parameters Boric Acid, ppm B =2800 l Chloride, ppb s150  ;

             . Fluoride, ppb                                         s150 l

1 Guideline Parameters pH @ 77*F 4.0 - 4.7 i l Aluminum, ppb s500 Calcium + Magnesium, ppb s500 Magnesium s250 l 1 I l

                                                                                                      }

l l f 9.1-37 Revision 4

9 F l 5 9 ,

• TAgtE 9.1-5 0 SPENT FlEL P%L COOLING AIS CLEAN-UP $YSTEM M FAlluRei M(BES Als EFFECTS AllALYSIS*

Plant Method Failure Effect Description Safety Operating Failure of Failure on System Safety of Component Function Mode ** Mode (s) Detection Famction Capability General Remarks l SFP Circulate the 1-6 One ptmp Status None - A redundant i Cooling water to the falls to monitoring cooling train is Ptsups spent fuel provide avaltable which will

(Typicat) poot . adequate Temperature provide adequate cooling flow indication of SFP of the SFP l

water Class 1E Provide 1E 1-6 Loss one Bus under- None - A red adant power AC Power power to the train of- voltage alarms train exists to power the (Typical) pumps power to redtedent ptmp its ESF status m associated monitoring of d N E'"F" $.

      ?                                                                                                                                                                                                                                                                                      m 7                                                                                                                  Ptap status g                                                                                                                  lights                                                                                                                                                              c m

ESF monitoring on " Channel Provide DC 1-6 Loss of DC None - Redundant trains III DC control power. power UPS failure, DC provide system capability Power. trouble alarm (Train B) ESF monitoring for ptmp (not rtsining, rn cor. trol power) Chamel IV Provide DC 1-6 Loss of DC ESF monitoring on None - Re4Jndmat trains DC Power control power power UPS fatture DC - provide system capability (Train C) trothie alarm ESF monitoring for pamp (not running, no

                                                                                       ,                                 control power) w a

WWY

      $ ** Single failure only applies to normal modes of SFPCCS operation.

m

• P1 ant Modes
• 1. Power operation 3. Not Standry 5. Cold Shutdown 0 2. Starttp 4. Hot Shutdown 6. Refueling o

k m ___ __.___...._______._m_._ _ _ _ _ _ _ _ _ _ ___.___ ____ .- m _ _____ _ _ _ _ -. - - - _ - --

UFSAR - INSERT #9 E 8 5 a t:: y 8 DESCRFRON OF PLANT MEUf0D OF FAHERE EFFECT ON COMPONENT SAFETY FUNCHON OPERAUNO PARERE MODE (S) FAILURE DETECHON SYSTEM SAFETY GENERAL REMARKS MODE *

• FUNCHON CAPABRITY MOVs 0447 & Circulate CCW I-6 MOV fai!s Loss of CCW Flow Alarm NONE. Frequency of occurrence is minimal 0032 water to SFP closed. (Fails to at Main Control Board De valve can be since the valves are normally open and (Normally Open) Cooling Ileat (Full Power Reopen (non-Class IE). opened either close only on an ESF signal After Exchangers A & B. to Refueling following an remoklY or Ioenlly receiving an ESF signal, Emergency with ESF Signal) High SFP Temperature by an operator. Operating Procedure instructs the Note: full-core Alarm at Main Control operator to open the MOVs within an  !

The FMEA of offload) Board (non-Class IE). allotted time frame. MOVs 0447 & @ 0032 is applicable If the valve fails to reopen remotely y , to the SFPCCS. after receiving an ESF signal, the valve @ g This does not can be reopened locally. e address the ESF c: $ function of the ne consequences ofloss of SFP $ , valves. cooling are mitigated by $ administrative!y providing makeup water through: (1) Reactor makeup water system, (2) Water from RWST (3) Demineralized water, or (4) Fire water. 1 hn O

4

     ,+

8 l

• STFEGS UFSAR Table 9.1-6 f SPENT RJEL POOL BOILING l

i Parameters 3,800 Core Thermal power, MWt  ! Decay assumed for the last full core offload 120 hours l3 i Hours of reactor exposure per as'sembly 40,000 hours SRP 9,1.3 j Decay heat generator methodology 1 Volume of water in spent fuel pool, ft 8 47,105 (water level 3 ft below normal water level with 10% of the pool volume assumed to be , 3 stainless steel) Dispersion Factors (y/Q) Table 15.B-1 l Dose Consecuences Exclusion Zone Boundary (0-2 hours) 1.3 x 10-8 Thyroid, rems ~ 3 Low Population Zone (0-30 days) 7.1 Thyroid, rems l I 9.1-39 Revision 3

     ?

STPEGS UFSAR 9.2.2 Component Cooling Water System 9.2.2.1 Desien Bases. The CCWS meets the requirements of 10CFR50, Appendix A, General Design Criteria (CDC) 1, 2, 3, 4, 5, 44, 45, 46, 54, 56 and

57. STPEGS Units 1 and 2 have. separate but identical CCWSs.

The CCWS is designed to:

1. Provide. cooling water to various nuclear plant components during all modes  !

of plant operation. This includes plant equipment required for safe shutdown and ESF equipment required after a postulated DBA.

2. Provide an intermediate fluid barrier between potentially radioactive systems and the ECWS to reduce the possibility of leakage of radioactive contamination to the outside environment.

3. Perform its cooling function following a DBA with offsite or standby power sources, automatically and without operator action, assuming a single l active or passive failure.  ;

4. Provide cooling water at 60*F to 105*F temperature during normal operation.

The' maximum temperature during DBA is 120.5'F (refer to Table 9.2.5-5 for temperature for the individual scenarios).

5. Conform to seismic category I requirements and safety classifications as indicated on Figures 9.2.2-1 through 9.2.2-5 and in Table 3.2.A-1. l 1

       ~

6. Permit periodic inspection of important components and periodic and  ;

functional testing to assure the integrity and operability of the system.  ! See Sections 3.9.6 and 6.6. In addition, the CCWS is protected from the effects of tornado loadings, missiles, flooding, pipe whip, and jet forces frca pipe breaks. See Sections 3.3.2, 3.4.1, 3.5, and 3.6. 9.2.2.2 system Descriotion. 1 9.2.2.2.1 Descriotion: The CCWS consists of three separate redundant  ; ' trains, each with a pump, HX, associated piping, and valves, that service two RCFCs, Residual Heat Removal (RHR) HX, and RHR pump, as shown on Figures 9.2.2-1 through 9.2.2-3. The three trains are connected to a common header Which services other equipment as shown in Figures 9.2.2-4 and 9.2.2-5. In addition, a compartmentalized surge tank is used to accommodate the water thermal expansion and contraction, and a chemical addition tank is used to balance the water l chemistry (Table 9.2.2-2) of the system. A CCW HX bypass line is provided to mait.tain 60*F minimum CCWS temperature. This line is only used when the ECW temperature is very low. For heat removal following a DBA, all three CCWS trains will operate if available, but two trains are capable of performing the heat removal function. Except for the seal aster HX, reactor coolant pump (RCP) lube oil coclers and thermal barrier, RCP motor air coolers, RHR pump seal coolers, centrifugal charging pump (CCP) supplementary coolers, CCP lube oil coolers, and positive displacement pump -supplementary cooler, the remaining equipment is isolated by valves which close on an SI signal. Flow to the RCP lube oil coolers, thermal barrier, and motor. air coolers is automatically isolated upon reaching the Containment pressure HI-3 setpoint. An SI signal opens the pneumatic valve (closed during normal operation) to provide cooling water to each RHR HX. Also, an SI signal shifts the cooling water supply to the RCFCs from the chilled water system to the CCWS by closing the chilled water and opening the CCVS motor-operated supply and return valves. 9.2-10 Revision 4

                      . . _ _ .         . . _ _ - _ , . - . . , , __ _       ..m._ ., ____ , _

, STPEGS UFSAR Cooling water to the spent fuel pool (SFP) HXs is manually restored within

      ^1-ht her: MEyMylfpMfpepijsip3HpljpMEppg@MIOpHfi(@lPQMpjMS following the SI-induced isolation.

Upon a LOOP signal, the RCB chilled water supply and return valves are closed automatically. The CCWS motor-operated valves (MOVs-0057, 0069, 0148, 0136, 0210, 0197) to and from the RCFC remain closed. The CCWS motor-operated valves (MOVs-0235, 0236, 0393, 0297) to the letdown and excess letdown heat exchangers remain open. Operator action from the main control room is required to restore flow to the RCFCs by closing MOVs-0235, 0236, 0393, 0297 and opening MOVs-0057, 0069, 0148, 0136, 0210, 0197 within 30 minutes after LOOP. The following components are cooled by the CCWS:

1. ESF Loads

a. RHR HXs

b. RCFCs

2. Non-ESF Loads

a. RCPs - lube oil coolers and thermal barrier

b. RCP motor air coolers

c. RHR pumps - seal coolers

d. Centrifugal charging pumps (CCPs) - lube oil coolers

e. Excess letdown HX

f. Reactor coolant drain tank (RCDT) HX

g. Seal water HX

h. Boron Recycle System (BRS) recycle evaporator package

1. Boron Thermal Regeneration System (BTRS) chiller unit J. Post-accident sampling system (PASS) sample coolers, primary sampling coolers, boric acid sample coolers, and radiation monitor sample coolers

k. Liquid Waste Processing System (LWPS) evaporator package

1. SFP HXs

m. Letdown HX

n. CCP supplementary coolers

o. Positive displacement pump supplementary cooler.

Misc.9eJ233.w 9.2-11 Revision 2

e' I STPECS UFSAR The maximum heat loads and required flow rates for each component during the different modes of operation are listed in Table 9.1.2-4 Each CCW HX is designed to meet the normal operational heat loads. The design maximum CCU outlet temperature from the CCW HX is 105*F. During normal operation, one CCW pump, one compartment of the CCW surge tank, one ECUS pump, and one CCW HX are in use to provide a source of cooling water. The system has sufficient capacity to meet the required heat removal rates for the remaining operating conditions such as startup, shutdown, and recirculation. The CCWS is required to remove residual and sensible heat from the Reactor Coolant System (RCS) through the RHR HXs from approximately 4 hours after normal shutdown (when the RHR HXs are placed in service) until the reactor coolant temperature reaches 150*F. Operation of the system is then continued in order to maintain cold shutdown. The CCW flow from each RHR HX is indi-cated on the main control board and displayed through the QDPS. The position of each CCW block valve in the outlet line of each RHR HX is nonitored by open/close indicating lights on the main control board and by the ESF Status Monitoring System. Since the RHR HX does not need to function until after recirculation switchover is completed, following a design basis LOCA, the operator would have sufficient time to open the CCW block valve by using the manual switch in the control room, if the SI signal failed to actuate the valve. If this attempt to open the block valve fails, the corresponding SI train should be shut down and the core cooling requirement will be provided from the two remaining SI trains. The CCW traina should also be shut down because of flow considerations (minimum pump flow).. The normal cooldown heat load is removed by all three CCW pumps and HXs. If one of the three trains is inoperative, the cooldown function is still effective, although the cooldown period is extended beyond the three train cooldown time of 12 hours. However, this does not affect the safe shutdown and cooldown of the reactor. Pumps, safety-related valves, and instrumentation required to operate each CCW train are provided with power from their respective standby power sources if normal and offsite power fail. Two of the three cooling trains are available within 60 seconds for a safe shutdown and cooldown of the reactor in case of a LOOP and a simultaneous failure of one standby power source. , Each CCW pump is connected to a separate ESF bus which is powered from either  ! offsite or standby power sources. The pump centrol logic provides for automatic starting of the standby pump (s) in the event of low system pressure or low ECW pressure during normal operation. The CCW surge tank is partitioned into three equal volumes by internal  ; baffles. Each compartment is connected to the inlet piping of one of the CCW pumps. The surge tank is sized to accommodate the water thermal expansion and contraction and the inleakage or outleakage in the system until the leaking component is isolated. The internal baffles are designed to provide separation between redundant CCW trains, so excess leakage from a pipe break in one train does not affect the operability of the other trains. The surge tank is located at the highest point in the system to ensure the required NPSH for proper operation of the CCW pumps and to provide a path for collecting entrapped gases. A radiation monitor activates an alarm if the CCW radioactivity increases beyond a preset level (see Section 11.5) due to inleakage from systems being cooled. Hakeup to the CCWS is added automatically from the Demineralized Water System (DUS). Low surge tank level 9.2-12 Revision 2

e 8

STPECS UFSAR initiates makeup; high surge tank level terminates makeup. As a backup by manual valve alignment, makeup water can be obtained from the Reactor Makeup Water System (RNWS) which is a safety-related and seismic Category 1 makeup

               , source.         A high-high surge tan'k level alarms in the control room. An alarm is also indicated in the control room if the level in the surge tank continues to decrease due to inadequate makeup. Further decrease in the surge tank level closes the valves on the lines to some non-ESF equipment (BTRS chiller unit,                     l LNPS evaporator package, sample coolers, radiation monitor sample coolers, letdown HX, excess letdown HX, BRS evaporator package, and RCDT HX).                             i If the level in the surge tank continues to decrease, each CCW train is                        ;

i automatically isolated from the other trains by closing the motor-operated  ! supply and return valves located upstream and downstream of the common supply l and the return headers, respectively. Simultaneously, the pneumatic cross- i connect. valves (FV 4656 and FV-4657) in the CCP and positive displacement pump i coolers supply and r'eturn headers (CCW train A and CCW train B from CCW train l C) and motor-operated valves MOV 0768 and MOV-0772 are closed to prevent loss ) of more than one CCW train at a time. As soon as the CCW common header is isolated, the low pressure switch located in the common header will automatic- I ally start the standby CCW pump (s). I l CCW supply to the RCPs is terminated as a result of the isolation of the CCW common header. However, the seal water to the RCPs from the CCPs "s main-tained since CCW supply to the CCP lube oil coolers is not isolated. Should the water level fall below the top of the surge tank dividing baffles, an alarm is indicated in the control room. Low-low surge tank compartment level initiates another alarm in the control room. The operator may trip the pump to respond to these alarms. Self-actuated, spring-loaded relief valves are provided for lines and com-ponents which could be pressurized beyond their design pressure due to improper operation or failure of the physical barrier separating the CCW from other high pressure systems. The system, except for NNS portions, is designed to seismic Category I requirements and is located inside seismic Category I structures. A failure l l in the NHS portion would not affect the operation of the remainder of the l system since low (makeup inadequate) surge tank level isolates the' NNS section. The system equipment and piping. are classified as SC 2 and 3 and NNS, as shown on Figures 9.2.2-1 through 9.2.2 5. The codes and standards to l i l which the system is designed are listed in Table 3.2.A-1. Table 9.2.2-2 specifies the water chemistry of the system. The system is j initially filled with water from the domineralized water storage tank (DWST) through the surge tank. The corrosion inhibitor is added during the system filling via the surge tank manhole. During normal operation the corrosion inhibitor concentration is maintained via the chemical addition tank utilizing the CCW pump discharge pressure for injection to the surge makeup header. 9.2.2.2.2 Comoonents: 4

1. Heat Exchangers 4

The CCW HXs are of the shell and straight tube type. The ECW circulates through the tubes while CCW circulates through the shell. Titanium tubes are 9.2-13 Revision 2 f

          - - . -                  -    . _ .  . .  . . _ .             -   . - . = .        . _ _    -       _    ..

1 STPEGS UFSAR a The shell used to provide maximum corrosion resistance to the brackish ECW. is carbon steel. The shell-side (CCW) pressure is maintained higher than the Design tube-side pressure to prevent leakage of ECW into ets system. parameters are listed in Table 9.2.2-1.

2. Pumps The CCW pumps are motor-driven horizontal, double-suction, dual-volute l centrifugal pumps with antifriction bearings. Design parameters are listed in l Table 9.2.2-1.

3. Surge Tank The CCW surge tank is constructed of carbon steel. The tank is partitioned to fore three separate volumes in the lower portion of the tank. The upper portion of the tank is open to each compartment. Design parameters are listed in Table 9.2.2-1.

4. Chemical Addition Tank The chemical addition tank is a vertical cylindrical tank constructed of carbon steel. Its design parameters a a given in Table 9.2.2-1.

5. Valves Self-actuated, The valves used in the CCUS are constructed of carbon eteel.

spring-loaded relief valves are provided for lines and components which might be pressurized beyond their design pressure by improper operation or malfunc-tion. Table 9.2.2-5 lists those relief valves in the CCVS for which Code Case N 242-1 is permitted. This Code Case provides alternate rules which may be used for the acceptance of metallic materials which were not manufacturad or supplied in complete conformance with the rules of NCA-3800 (or NA-3700) and which are used in the construction of items for which the applicable code is Winter 1973 Addendum or later. The valves are listed by tag number.

6. Piping ,

All CCW piping is carbon eteel with welded joint endIn connections, except at this case, flanged components which might be removed for maintenance. connections are used. 9.2.2.3 Safeev Evaluation. 9.2.2.3.1 Availability and Reliability: The normal operation of each CCW pump and HX train is rotated to monitor operational capability and balance operating time. l The operability of safety-related valves and equipment, as required during various modes of operation of the system, is tested by simulated signals corresponding to each mode on a routine basis. The CCVS performs vital cooling functions during and after a postulated DBA, Sufficient and therefore is designed to meet the single failure criterion. cooling capacity is provided to fulfill all system requirements under accident conditions, assuming a single failure in the CCUS. 9.2-14 Revisicn 2

_ . _ _ . _____ . . ,_ __m _ . _ _ _ _ _ _ _ . _ . _ . _ _ __ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ l h

 't STPEGS UFSAR TomeettheESFcoolingrequirements,theCdWpumpsareautomatically sequenced on standby power in the event of a LOOP. The electric power supplies to the pumps, valves, instrumentation, and control cabling for each cooling train are physically separated. Power is supplied to each pump from an independent ESF bus.

The CCW pumps, HXs, SC 2 and 3 valves, and piping are designed to seismic Category I requirements and are located in seismic Category I structures. They are protected against internally generated missiles, pipe whip, jet forces, and flooding due to pipe rupture. The CCWS equipment is located in areas shielded from radiation and is accessible for maintenance or inspection during power operation. A failure mode and effect analysis (FEMA) for the CCWS is given in Table 9.2.2-3. The CCWS design incorporates automatic isolation of cooling water to the RCPs by the Containment isolation phase B signal. Loss of cooling water to the RCPs due to a spurious isolation signal or the operator closing a single isolation valve is avoided by parallel Containment isolation valves, as shown on Figure 9.2.2-5. The control logic for the valves is designed so that a spurious isolation signal in any one actuation train does not isolate flow; however, isolation is assumed on a penetration basis even assuming the single failure of one of the three actuation trains. j 9.2.2.3.2 Lankare Minimization. Detection. and Isolation: To minimize the possibility of leakage, welded construction is used throughout the CCWS where practicable. ! Makeup to the system is automatic. The surge tank level is monitored by the main plant computer and the QDPS. Opening of the valve providing tank makeup is alarmed by the computer to give an indication of system leakage. If the normal source of makeup (the DWS) fails or is inadequate, the surge tank level instrumentation provides alarms and actuations as discussed in Section 9.2.2.2.1. Should a large inleakage into the CCWS develop, the level in the surge tank will rise, and the high-high-level condition will be annunciated. If a leaking fluid is radioactive, then a high radiation level will be alarmed. Refer to Section 11.5 for further information on the process radiation aceitor. If the level in the surge rank continues to increase, the CCW will discharge to the CCW sump through the open vent. The vent is designed to accommodate the maximum inleakage due to the rupture of one RCP thermal barrier, which is estimated at 275 gal / min. The increased pressure due to a rupture of the RCP thermal barrier will close two active self-actuated pressure regulated valves located at the CCW outlet from' the RCP thermal barrier and isolate the RCS inleakage. In addition, a high flow or high temperature in the return line from the thermal barrier, indicating failure of L the RCP thermal barrier pressure-retaining boundary, will close the CCW motor-operated valve downstream of the two pressure regulated valves; an alarm in the control room alerts the operator as well. The portion isolated is designed for RCS pressure and temperature. . The operator, by checking flow and temperature readings against normal values, can locate the affected portion of the system and isolate this portion by 9.2-15 Revision 0

1 STPECS UFSAR i 1 closing the appropriate remotely operated or manual valves. Very small leaks will be' detected by periodic inspection of the system piping and valves. j The relief valves on the CCW lines to the various HXs are sized to relieve the volumetric expansion occurring if the shell side of the HX is isolated and high-temperature process fluid flows through the HX tubes. 9.2.2.4 Tests and Insnections. During preoperational testing, the following will be checked: 1 J

1. Calibration of all instrumentation l

2. Actuation of automatic controls at their proper setpoints

3. Actuation of alarms at the'setpoints 4 Operation of power-operated valves

5. Operation of CCW pumps and checking of flow and discharge pressure

6. Checking and adjusting required flow to each component serviced by CCWS l

7. System water chemistry

                -All the above functions are checked periodically during the life of the plant.                                                   ]
                                                                                                                                                  }

Inservice inspections of ASME B&rV Code, Section III, Class 2 and 3 components

          .       are performed in accordance with ASME B&PV Code, Section XI.

9.2.2.5 Instrumentation Aeolication. The Engineered' Safety Features Actuation System (ESFAS) for the CCWS is discussed in Section 7.3.1. Controls for remote manual operation of each CCW pump and selection of standby CCW pumps are provided in the control room. Remote manual control of pneumatically operated valves under normal operating conditions is provided in the control room. Pneumatically operated valves are not required for control

                -purposes and fail in the safe position on LOOP. Remote manual control of all motor-operated and solenoid operated valves necessary for post-lOCA cooling, for surge tank makeup, for maintenance of the CCWS trains, and for Containment                                                 l isolation is provided in the control room. The valves that are power locked out will require that the motor control center (NCC) breaker be closed to enable remote operability.- CCW pump operating status and valve position                                                       !

indicating lights are provided in the control room. For valves that are power locked out the indicating lights are not illuminated and control room indica- I i tion is given via bypass / inoperable for valves not in the power locked out position. Provisions are made for local indication and control in the ' switchgear room for each CCW pump and selected valves in the CCWS. The temperature in each CCW main loop is monitored by the QDPS. The tempera-ture at each CCW pump discharge is also monitored by the plant computer. The CCW return temperatures from all equipment are monitored through the plaat computer except for the sample coolers, boron recycle evaporator, positive displacement pump supplementary cooler, CCP lube oil coolers, LWPS evaporator package, and CCP supplementary coolers. ' Temperatures for this equipment, with the exception of the sample coolers, are monitored locally. 9.2-16 Revision 0 w., , .- - - _ . . - , - _ . - , - - - , - - - , ,. . . - . . . - , . . , . , . , - -

_ m _-_ _ _ . . _ _ _ _ _ _ _ . .. _ _ _ . . _ ._ _ _ . - _ _ _ _ _ $ , STPECS UFSAR I ' j Flows through the CCW main loops, RHR KXs, and RCFCs are displayed in the , control room on panel indicators and logged by the QDPS while the remaining equipment, with the exception of the letdown HX and sample coolers, has local flow indications or test connections. The temperature difference in conjune-4 tion with the flow data from individual components is used to monitor the particular component's performance. Each CCW main loop' outlet pressure is indicated in the control room through indicators and the QDPS. The radiation 1 - level in the CCWS is available in the control room through the Radiation Monitoring System (RMS) display. ] Local pressure gauges are provided on the suction and discharge lines of each CCW pump. Local level gauges are provided for each compartment of the surge l tank. Surge tank level indication is displayed in the control room through - indicators and the QDPS. High and low levels in the surge tank are alarmed in , the control room. Measurements taken downstream of each CCW HX, which indicate system malfunction, are annunciated in the control room. Local flow  ! indication is provided on each RCP lube oil cooler and motor air cooler out- I let. Low flow condition at each RCP lube oil cooler and motor air cooler outlet is annunciated in the control room. 9.2.3 Makeup Domineralized Water System 9.2.3.1 Desien Bases. The Makeup Demineralized Water System (MDWS) uses ozonated and filtered service water as influent and removes the ionic impurities of the raw water to provide high-purity demineralized water suitable for use in the primary and secondary cycles of the plant. Sodium [ hypochlorite is used as a backup in the event the ozone generator fails. The purity of the water produced is suitable for preoperational tests, wet layup, startup, and normal operation of the plant primary and secondary systems. The system is designed for a specific nominal capacity, and transient demands for abnormal operating conditions are allowed for by the buffer effect of the storage volumes of the Demineralized Water Distribution System. For a more complete description of the Demineralized Water Distribution System and supply interfaces with other plant systems, refer to Sections 9.2.6 and 9.2.7 and Figures 9.2.6-1, 9.2.6-2, and 9.2.7-1. The capability of the MDWS to store, handle, and dispense any chemicals used in 4 the demineralization and regeneration process is described in Section 3.6 of the Environmental Report. i Piping, ion exchangers, and associated equipment of the MDWS are constructed l of corrosion-resistant material or lined carbon steel to prevent contamination of the water by corrosion products. The equipment is designed in accordance with ASKE B&PV Code, Section VIII, Division 1. The MDUS is designed to produce demineralized water to meet NSSS vendor recommended guidelines as follows: Specific Conductivity"' Less than 0.1 micro mhos/cm at 25*C 0xygen less than 100 ppb Suspended Solids"' Less than 50 ppb Total Silica, S iO, Less than 50 ppb 9.2-17 Revision 0 a , , ., ,c, . , - , - -}}