ML20049H283
| ML20049H283 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 02/12/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 8202230058 | |
| Download: ML20049H283 (600) | |
Text
{{#Wiki_filter:. . . - - _. _ - - _ _ _ .- -.-- . --. . - - . . - - . .. - _ __ - -- . - -. - . r GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i , 4 i CHAPTER 9 i AUXILIARY SYSTEMS 1 l i i l i i l 8202230058 820212 i PDR ADOCK 05000447 K PDR
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 SECTION 9.1 [~ CONTENTS Section Title Page 9.1 FUEL STORAGE AND HANDLING 9.1-1 9.1.1 New Fuel Storage (High Density) 9.1-1 9.1.1.1 Design Basis 9.1-1 9.1.1.1.1 Safety Design Bases 9.1-1 9.1.1.1.1.1 Safety Design Bases - Structural 9.1-1 9.1.1.1.1.2 Safety Design Bases - Nuclear 9.1-1 9.1.1.1.2 Storage Design Bases 9.1-2 9.1.1.2 Facilities Description 9.1-2 9.1.1.3 Safety Evaluation 9.1-3 9.1.1.3.1 Criticality Control 9.1-3 9.1.1.3.2 Structural Design 9.1-4 9.1.1.3.3 Protection Features of the New Fuel Storage Facilities 9.1-5
\ 9.1.2 Spent Fuel Storage (High Density) 9.1-7
[C 9.1.2.1 Design Basis 9.1-7 9.1.2.1.1 Safety Design Bases 9.1-7 9.1.2.1.1.1 Safety Design Bases - Structural 9.1-7 9.1.2.1.1.2 Safety Design Bases - Nuclear (High Density) 9.1-8 9.1.2.1.2 Storage Design Bases 9.1-9 9.1.2.2 Facilities Description 9.1-9 9.1.2.3 Safety Evaluation 9.1-10 9.1.2.3.1 Criticality Control 9.1-10 9.1.2.3.2 Structural Design 9.1-11 9.1.2.3.3 Protective Features of Spent Fuel Storage Facilities 9.1-13 9.1.2.4 Testing Inspection 9.1-17 9.1.2.5 Summary of Radiological Considerations 9.1-17 9.1.3 Fuel Pool Cooling and Cleanup System 9.1-17 9.1.3.1 Design Bases 9.1-17 9.1.3.1.1 Safety Design Bases 9.1-17 k,_) 9.1.3.1.2 Power Generation Design Bases 9.1-18 9.1.3.2 System Description 9.1-19 9.1-i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued) O Section Title Page 9.1.3.3 Safety Evaluation 9.1-23 9.1.3.4 Inspection and Testing Requirements 9.1-25 9.1.3.5 Radiological Considerations 9.1-25 9.1.4 Fuel-Handling System 9.1-26 9.1.4.1 Design Bases 9.1-26 9.1.4.2 System Description 9.1-28 9.1.4.2.1 Spent Fuel Cask 9.1-28 9.1.4.2.2 Cask Crane 9.1-28 9.1.4.2.3 Fuel Servicing Equipment 9.1-28 9.1.4.2.3.1 Fuel Prep Machine 9.1-29 9.1.4.2.3.2 New Fuel Inspection Stand 9.1-29 9.1.4.2.3.3 Channel Bolt Wrench 9.1-29 9.1.4.2.3.4 Channel-Handling Tool 9.1-30 9.1.4.2.3.5 Fuel Pool Sipper 9.1-30 9.1.4.2.3.6 Channel Gauging Fixture 9.1-30 9.1.4.2.3.7 General-Purpose Grapple 9.1-31 9.1.4.2.3.8 Jib Crane 9.1-31 9.1.4.2.3.9 Fuel-Handling Platform 9.1-32 9.1.4.2.3.10 Channel-Handling Boom 9.1-32 9.1.4.2.3.11 Fuel Transfer System 9.1-32 9.1.4.2.4 Servicing Aids 9.1-38 9.1.4.2.5 Reactor Vessel Servicing Equipment 9.1-39 9.1.4.2.5.1 Reactor Vessel Service Tools 9.1-39 9.1.4.2.5.2 Steamline Plug 9.1-40 9.1.4.2.5.3 Shroud Head Bolt Wrench 9.1-40 9.1.4.2.5.4 Head Holding Pedestal 9.1-40 9.1.4.2.5.5 Head Stud Rack 9.1-41 9.1.4.2.5.6 Dryer and Separator Strongback 9.1-41 9.1.4.2.5.7 Head Strongback/ Carousel 9.1-42 In-Vessel Servicing Equipment 9.1- ;4 9.1.4.2.6 9.1.4.2.7 Refueling Equipment 9.1-45
' .1.4.2.7.1 Refueling Platform 9.1-45 9.1-ii
I GESSAR II .22A7007 I 238 NUCLEAR ISLAND Rev. O i CONTENTS (Continued) Section Title Page 9.1.4.2.7.2 Auxiliary and Vessel Platforms 9.1-46 9.1.4.2.7.3 Fuel-Handling Platform 9.1-47 l 9.1.4.2.8 Storage Equipment 9.1-47 9.1.4.2.9 Under-Reactor Vessel Servicing Equipment 9.1-48 9.1.4.2.10 Description of Fuel Transfer 9.1-49
; 9.1.4.2.10.1 Arrival of Fuel on Site 9.1-49 9.1.4.2.10.2 Refueling Procedure 9.1-50
- 9.1.4.2.10.2.1 New Fuel Preparation 9.1-53 9.1.4.2.10.2.1.1 Receipt and Inspection of New l 9.1-53 Fuel 9.1.4.2.10.2.1.2 Channeling New Fuel 9.1-53 1
9.1.4.2.10.2.1.3 Equipment Preparation 9.1-54
- 9.1.4.2.10.2.2 Reactor Shutdown 9.1-54 9.1.4.2.10.2.2.1 Drywell Head Removal 9.1-54 l
^ 9.1.4.2.10.2.2.2 Reactor Well Servicing 9.1-55 9.1.4.2.10.2.3 Reactor Vessel Opening 9.1-55
! 9.1.4.2.10.2.3.1 Vessel Head Removal 9.1-55 9.1.4.2.10.2.3.2 Dryer Removal 9.1-56 9.1.4.2.10.2.3.3 Separator Removal 9.1-56 9.1.4.2.10.2.3.4 Fuel Bundle Sampling 9.1-56. 9.1.4.2.10.2.4 Refueling and Reactor Servicing 9.1-57 9.1.4.2.10.2.4.1 Refueling 9.1-57 9.1.4.2.10.2.5 Vessel Closure 9.1-59 9.1.4.2.10.3 Departure of Fuel From Site 9.1-60 3 9.1.4.3 Safety Evaluation of 'ruel-Handling System 9.1-63 9.1.4.4 Inspection and Testing j Requirements 9.1-65 9.1.4.4.1 Inspection 9.1-65
- 9.1.4.4.2 Testing 9.1-66 9.1.4.5 Instrumentation Requirements 9.1-67 9.1.4.5.1 Refueling Platform 9.1-67
] ' ())
\_ ,
f 1
- 9.1-iii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued) Section Title Page 9.1.4.5.2 Fuel Support Grapple 9.1-67 9.1.4.5.3 Inclined Fuel Transfer Tube 9.1-68 9.1.4.5.4 Other 9.1-68 9.1.4.5.5 Radiation Monitoring 9.1-68 I
\
i l l i i 1 i 9.1-iv
GES'SAR II' 22A7007 238 NUCLEAR ISLAND Rev. 0 l SECTION 9.1 TABLES Table Title Page 9.1-1 Tools and Servicing Equipment 9.1-69 9.1-2 Fuel Servicing Equipment 9.1-70 9.1-2a Fuel Transfer System Components 9.1-71 9.1-3 Reactor Vessel Service Equipment 9.1-73 9.1-4 Under-Reactor Vessel Servicing Equipment and Tools 9.1-74 i t G 9.1-v/9.1-vi
GESSAR II 22A7007 238 NUCLEAR ISLANL Rev. 0 ,f. SECTION 9.1 () _ ILLUSTRATIONS Figure Title Page 9.1-1 High Density Fuel Storage Module 9.1-75 9.1-2 Eccentric Fuel Positioning 9.1-76 9.1-3 Fuel Preparation Machine Shown Installed in Facsimile Fuel Pool 9.1-77 9.1-4 New Fuel Inspection Stand 9.1-78 9.1-5 Channel Bolt Wrench 9.1-79 9.1-6 Channel-Handling Tool 9.1-80 9.1-7 Fuel Pool Sipper 9.1-81 9.1-8 General-Purpose Grapple 9.1-82 9.1-9 Jib Crane 9.1-83 9.1-10 Channel-Handling Boom 9.1-84 9.1-11 Transfer Tube 9.1-85 9.1-12 Plant Refueling and Servicing Sequence 9.1-86 s 9.1-13a Inclined Fuel Transfer System 9.1-87 () 9.1-13b Inclined Fuel Transfer System 9.1-88 9.1-14 Simplified Section of New Fuel Handling Facilities (Fuel Building) 9.1-89 9.1-15 Simplified Section of Refueling Facilities (Reactor Building) 9.1-90 9.1-16 Drywell Head Removal Sequence 9.1-91 9.1-17 Reactor Vessel Head Removal Sequence 9.1-92 9.1-18 Steam Dryer Remcval Sequence 9.1-93 9.1-19 Separator Removal Sequence 9.1-94 9.1-20 Fuel Bundle Transfer Sequence 9.1-95 9.1-21 Fuel Bundle Laydown Areas 9.1-96 9.1-22 Containment Building Laydown Areas 9.1-97 9.1-23a Fuel Pool Cooling and Cleanup System P&I Diagram 9.1-99 9.1-23b Fuel Pool Cooling and Cleanup System P&I Diagram 9.1-100 9.1-23c Fuel Pool Cooling and Cleanup System P&I Diagram 9.1-101 {g 9.1-24a Filter /Demineralizer System P&I Diagram 9.1-103
'N- 9.1-24b Filter /Demineralizer System P&I Diagram 9.1-104 9.1-vii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ILLUSTRATIONS (Continued) Fiqure Title Pace 9.1-24c Fuel Pool Cooling and Cleanup System P&I Flow Diagram 9.1-105 9.1-25a Filter /Demineralizer Sys Fuel Pool Cooling
& Cleanup P&I 9.1-107 i
9.1-25b Filter /Demineralizer Sys Fuel Pool Cooling
& Cleanup P&I 9.1-108 O
O 9.1-viii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
- 9. AUXILIARY SYSTEMS 7-s t I 9.1 FUEL STORAGE AND HANDLING 9.l.1 New Fuel Storage (High Density) 9.1.1.1 Design Basis 9.1.1.1.1 Safety Design Bases 9.1.1.1.1.1 Safety Design Bases - Structural (Figure 9.1-1)
(1) The new fuel storage racks contain storage space for 30% of one full core of fuel assemblies (with channels) or bundles (without channels). They art designed to with-stand all credible static and seismic loadings. (2) The racks are designed to protect the fuel assemblies and (O)
%j bundles from excessive physical damage which may cause the release of radioactive materials in excess of 10dFR20 and 10CFR100 requirements, under normal and abnormal conditions caused by impacting from either fuel assem-blies, bundles or other equipment.
(3) The racks are constructed in accordance with the Quality Assurance Requirements of 10CFR50, Appendix B. (4) The racks are categorized as Safety Class 2 and Seismic Category I. 9.1.1.1.1.2 Safety Design Bases - Nuclear (1) A full array of loaded new fuel racks is designed to be subcritical, by at least 5% tik . Neutron-absorbing material, as an integral part of the design, is employed () to assure that the calculated k eff, including biases and uncertainties, will not exceed 0.95 under all normal and abnormal conditions, or 0.98 under optimum moderation. 9.1-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.1.1.1.2 Safety Design Bases - Nuclear (Continued) (a) Monte Carlo techniques are employed in the calcu-lations performed to assure that k does not eff exceed 0.95 under all normal and abnormal conditions. (b) The assumption is made that the storage array is infinite in all directions. Since no credit is taken for neutron leakage, the values reported as effective neutron multiplication factors are, in reality, infinite neutron multiplication factors. (c) The biases between the calculated results and experimental results, as well as the uncertainty involved in the calculations, are taken into account as part of the calculational procedure to assure that the specific k limit is met. eff 9.1.1.1.2 Storage Design Bases (1) The new fuel storage racks provided in the new fuel storage vault provide storage for 30% of one full core fuel load. (2) The new fuel modules are designed and arranged so that fuel assemblies and bundles can be handled efficiently during refueling operations. 9.1.1.2 Facilities Description (1) The location of the new fuel storage facility within the complex is shown in Section 1.2. O 9.1-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.1.2 Facilities Description (Continued) (~'J} (2) The new fuel storage racks are top entry racks designed to maintain the new fuel while precluding the possibility of criticality under normal and abnormal conditions. The upper tieplate of the fuel element rests against the module to provide lateral support. The lower tieplate sits in the bottom of the rack, which supports the weight of the fuel. (3) The rack arrangement is designed to prevent accidental insertion of fuel assemblies or bundles between adjacent racks. The storage rack is designed to provide accessi-bility to the fuel bail for grappling purposes. Nominal fuel spacing from center to center is 6.56 inches by 6.56 inches. [ ] (4) The floor of the new fuel storage vault is sloped to a drain located at the low point. This drain removes any water that may be accidentally and unknowingly introduced into the vault. The drain is part of the floor drain subsystem of the liquid radwaste system. (5) The radiatien monitoring equipment for the new fuel storage area is described in Subsection 7.1.1.6.2. 9.1.1.3 Safety Evaluation 9.1.1.3.1 Criticality Control The design of the new fuel storage racks, which includes neutron-absorbing materials, provides for an effective multiplication factor (keff) f r both normal and abnormal storage conditions equal to or less than 0.95. To ensure design criteria are met, n the following normal and abnormal new fuel storage conditions were (%.j) analyzed: 9.1-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.1.3.1 Criticality Control (Continued) (1) normal positioning in the new fuel array, and (2) eccentric positioning in the new fuel array (Figure 9.1-2) . The new fuel storage area will accommodate fuel (k gg i 1.35 at 20 C in standard core geometry) from a multi-unit BWR facility with no safety implications. 9.1.1.3.2 Structural Design (1) The new fuel vault contains one 13x17 fuel storage rack, which provides storage for a maximum of 221 fuel assemblies or bundles. (2) The new fuel storage racks are designed to be supported above the vault floor by a support structure. Since the racks are freestanding (i.e., no supports above the base), the support structure also provides the required stability. (3) The racks include individual solid tube storage compart-ments which provide lateral restraints over the entire length of the fuel assembly. (4) The weight of the fuel assembly or bundle is supported axially by the rack lower support. (5) The racks are fabricated from stainless steel. Materials used for construction are specified in accordance with the latest issue of applicable ASTM specifications at the time of equipment order. O 9.1-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ("') 9.1.1.3.2 Structural Design (Continued) L) (6) The nominal center-to-center spacing for the fuel assemblies or bundles between rows is 6.56 inches. The maximum spacing between racks is 2.0 inches. (7) Lead-in guides at the top of the storage spaces provide guidance of the fuel during insertion. (8) The racks are designed to withstand, while maintaining the nuclear safety design basis, the impact force generated by the vertical free-fall drop of a fuel assembly from a height of 6 ft. (9) The rack is designed to withstand a pullup force of 4000 lb and a horizontal force of 1000 lb. There are no readily definable horizontal forces in excess of 1000 lb {- ' and, in the event a fuel assembly should jam, the maxi-mum lif ting force of the fuel-handling platform grapple (assumes limit switches fail) is 3000 lb. (10) The new fuel storage racks require no periodic special testing or inspection for nuclear safety purposes. 9.1.1.3.3 Protection Features of the New Fuel Storage Facilities The new fuel storage vault is housed in the Fuel Building (Sub-section 3.8.4). The vault and Fuel Building are Seismic Category I structures. The Fuel Building provides protection from severe natural phenomena such as tornadoes, tornado missiles, floods and high winds. Fire protection features are described in Subsection 9.5.1 and Appendix 9A. The storage rack structure is designed to withstand the impact ("'T resulting from a falling weight. Tests using a simulated fuel V bundle of the correct weight and size have been conducted to 9.1-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.1.3.3 Protection Features of the New Fuel Storage Facilities (Continued) verify that the rack casting can withstand the impact from a bundle dropped from a maximum allowable height above the array. The rack casting failed when a drop of 6.17 ft (4314 ft-lb) was made at midspan. Procedural fuel-handling requirements and equipment design dictate that no more than one bundle at a time can be handled over the storage racks and at a maximum height of 2 ft above the upper rack. Therefore, the racks cannot be displaced in a manner causing critical spacing as a result of impact from a falling object. The five-ton general-purpose building crane can traverse the full length of the fuel building. A corridor is provided along the shield building side (not over) of the pools and vault; roof hatches are provided in the vicinity of the FPPCU equipment. This permits removal of major equipment by way of the hatch, thus elim-inating the need to move these components along or over the pools and vault. The shipping cask cannot be lifted or moved above the new fuel vault because of inadequate clearance. Should it become necessary to move major loads along or over the pools, administrative controls will require that the load be moved over the empty portion of the spent fuel pool and to avoid the area of the new fuel storage vault. New fuel is aarried to the new fuel vault and placed in the storage rack using the fuel-handling platform. During positioning of new fuel into the new fuel racks, the grapple is always above the upper fuel rack casting, and the grapple interfaces only with the
- fuel bundle bail and could not engage the fuel rack. Thus, the transfer devices used for new fuel handling to the new fuel vault cannot impose uplift loads on the rack castings.
O 9.1-6 1
< GESSAR II 22A7007 j
i 238 NUCLEAR ISLAND Rev. O l 9.1.1.3.3 Protection Features of the New Fuel Storage Facilities (Continued) 1 j The new fuel racks are designed to be restrained by holddown bolts to assure that rack spacing does not vary during the SSE. 4 i The storage rack structure is so designed that the height of the i rack is less than the length of the fuel bundle. Therefore, the , 1
- upper tieplate of the bundle cannot pass below the top cross mem-
! ber of the rack, In addition, the new fuel vault is provided with a removable cover to ensure that water, debris and dust do not fall into the vault. This cover will also prevent small parts and components from fall- , ! ing on the new fuel. 9.1.2 Spent Fuel Storage (Iligh Density) 1 9.1.2.1 Design Basis 9.1.2.1.1 Safety Design Bases 9.1.2.1.1.1 Safety Design Bases - Structural (Figure 9.1-1) (1) The spent fuel storage racks in the Fuel Building and Containment contain storage space for 394% of one full core of fuel assemblies (with channels) or bundles (without chanpels). They are designed to withstand all
- credible static and seismic loadings.
i - ! (2) The racks are designed to protect the fuel assemblies and bundles from excessive physical damage which may cause the release of radioactive materials in excess of f I LOCFR20 and 10CFR100 requirements, under normal and abnormal conditions caused by impacting from either fuel l 9 assemblies, bundles or other equipment. ). 9.1-7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.2.1.1.1 Safety Design Bases - Structural (Continued) (3) The racks are constructed in accordance with the Quality Assurance Requirements of 10CFR50, Appendix B. (4) The racks are categorized as Safety Class 2 and Seismic Category I. (5) The pool level is maintained by structural concrete walls with a stainless steel liner. The bottoms of the pool gates are sufficiently high to maintain the water level over the spent fuel storage racks for adequate shielding and cooling. All pool fill and drain lines enter the pool above the safe shielding water level. Redundant anti-siphon vacuum breakers are located at the high point of the pool circulation lines to preclude a pipe break from siphoning the water from the pool and jeopardizing the safe water level. 9.1.2.1.1.2 Safety Design Bases - Nuclear (High Density) (1) A full array in the loaded spent fuel rack is designed to be subcritical, by at least 5% Ak. Neutron-absorbing material, as an integral part of the design, is employed to assure that the calculated k eff, including biases and uncertainties, will not exceed 0.95 under all normal and abnormal conditions. (a) Monte Carlo techniques are employed in the calcu-lations performed to assure that k does not eff exceed 0.95 under all normal and abnormal conditions. (b) The assumption is made that the storage array is infinite in all directions. Since no credit is 9.1-8
! GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i 1 J O 9.1.2.1.1.2 Safety Design Bases - Nuclear (High Density) (Continued) i taken for neutron leakage, the values reported as ) effective neutron multiplication factors are, in reality, infinite neutron multiplication factors. (c) The biases between the calculated results and i experimental results, as well as the uncertainty l involved in the calculations, are taken into account as part of the calculational procedure to assure that the specific k limit is met. eff 9.1.2.1.2 Storage Design Bases (1) The fuel storage racks provided in the spent fuel storage pool provide storage for 326% of one full core fuel load. C* ! (2) The fuel storage racks provided in the containment pool provide storage for 68% of one full core fuel load. (3) The spent fuel racks are designed and arranged so that fuel assemblies and bundles can be handled efficiently during refueling operations. 9.1.2.2 Facilities Description (1) The spent fuel storage racks provide storage in the containment and spent fuel pools for spent fuel received from the reactor vessel during the refueling operation. The spent fuel storage racks are top entry racks designed to maintain the spent fuel while precluding the possibil-ity of criticality under normal and abnormal conditions. I The upper tieplate of the fuel element rests against the rack to provide lateral support. The lower tieplate sits
, 9.1-9 d
i
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0 9.1.2.2 Facilities Description (Continued) in the bottom of the rack, which supports the weight of the fuel. (2) The rack arrangement is designed to prevent accidental insertion of fuel assemblies or bundles between adjacent modules. The storage rack is designed to provide accessibility to the fuel bail for grappling purposes. Nominal fuel spacing from center to center is 6.56 inches by 6.56 inches. (3) The location of the spent fuel pool and the containment pool within the complex is shown in Section 1.2. 9.1.2.3 Safety Evaluation 9.1.2.3.1 Criticality Control The design of the spent fuel racks, which includes neutron-absorbing materials, provides for an effective multiplication factor (keff) f r both normal and abnormal storage conditions equal to or less than 0.95. Normal conditions exist when the fuel storage modules are located in the pool and are covered with approximately 25 ft of water for radiation shielding, with the maximum number of fuel assemblies or bundles in their storage position. An abnormal condition may result from accidental drop-ping of fuel or damage caused by the horizontal movement of fuel or equipment. To ensure that design criteria are met, the following normal and abnormal spent fuel storage conditions were analyzed: (1) normal positioning in the spent fuel array; (2) eccentric positioning in the spent fuel array (Figure 9.1-2); 9.1-10
.. . - - . -_ - . . _ - ___ . - = - .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O l () 9.1.2.3.1 Criticality Control (Continued) (3) pool water temperature increased to 100 C; and i ! (4) normal storage array of ruptured fuel.
! The spent fuel storage area will accommodate fuel (kinf 1 1.35 at 20 C in standard core geometry) from a multi-unit BWR facility with i no safety implications, t
9.1.2.3.2 Structural Design i l (1) The spent fuel pool contains 12 racks, four each of 13x13 racks and eight each of 13x17 racks, which provides stor-age for a maximum of 2444 fuel assemblies or bundles. ) ] (2) The containment pool contains three 13x13 racks, which
- provides storages for a maximum 507 fuel assemblies or
() bundles. 1' (3) The fuel storage racks are designed to be supported
- above the pool floor by a support structure. The support j structure allows sufficient pool water flow for natural
{ convection cooling of the stored fuel. Since the modules are freestanding (i.e., no supports above the base), the l support structure also provides the required dynamic
- stability.
(4) The racks include individual solid tube storage compart- ! monts, which provide lateral restraints over the entire 4 length of the fuel assembly or bundle. (5) The weight of the fuel assembly or bundle is supported axially by the rack fuel support. O 9.1-11 1---_-_.-.... . . . - . . - . . , _ . . . _ - - - , . , , , _ . . . . . _ . - . . . - - - , . . . - _ . _ _ , _ .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 0 9.1.2.3.2 Structural Design (Continued) (6) The racks are fabricated from stainless steel. Materials used for construction are specified in accordance with the latest issue of applicable ASTM specifications at the time of equipment order. (7) The nominal center-to-center spacing for the fuel assemblies or bundles between rows is 6.56 inches. The maximum spacing between racks is 2.0 inches. (8) Lead-in guides at the top of the storage spaces provide guidance of the fuel during insertion. (9) The racks are designed to withstand, while maintaining the nuclear safety design basis, the impact force gen-erated by the vertical free-fall drop of a fuel assembly from a height of 6 ft. ll (10) The rack is designed to withstand a pullup force of 4000 lb and a horizontal force of 1000 lb. There are no readily definable horizontal forces in excess of 1000 lb and in the event a fuel assembly should jam, the maximum lifting force of the fuel-handling platform grapple (assumes limit switches fail) is 3000 lb. (11) The fuel storage racks are designed to handle irradiated fuel assemblies. The expected radiation levels are well below the design levels. The fuel storage facilities will be designed to Seismic Category I requirements to prevent earthquake damage to the stored fuel. From the foregoing analyses, it is concluded that the spent fuel storage arrangement and design meet the safety design bases. 9.1-12
l l GESSAR II 22A7007 1 238 NUCLEAR ISLAND Rev. 0 l I i (' 1 9.1.2.3.2 Structural Design (Continued) G' The fuel storage pools have adequate water shielding for the stored spent fuel. Adequate shielding for transporting the fuel is also provided. Liquid level sensors are installed to detect a low pool water level, and adequate makeup water is available to assure that the fuel will not be uncovered should a leak occur. Since the fuel storz.ge racks are made of noncombustible material and are stored under water, there is no potential fire hazard. The large water volume also protects the spent fuel storage racks from potential pipe breaks and associated jet impingement loads. The spent fuel storage racks require no periodic special testing or inspection for nuclear safety purposes. Regulatory Guide Compliance - Regulatory Guide 1.13 O V For commitment and revision number, see regulatory guide commit-ment matrix in Section 1.8. This regulatory guide is applicable to spent fuel storage facilities. The building containing the fuel storage facilities, including the storage racks and pool, is designed to protect the fuel from damage caused by: (1) natural events such as earthquake, high winds and flooding, and (2) mechanical damage caused by dropping of fuel assemblies bundles, or other objects onto stored fuel. 9.1.2.3.3 Protective Features of Spent Fuel Storage Facilities The cask crane is physically restricted from traversing the spent fuel storage pool. The five-ton general-purpose crane will (n) traverse the full length of the fuel building. In addition, roof 9.1-13 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.2.3.3 Protective Features of Spent Fuel Storage Facilities 4 (Continued) hatches are provided over the operating floor in the vicinity of the FPCCU system. These hatches permit removal of these items from the building and eliminate the need to move major loads along the aisle over the spent fuel pool and new fuel storage vault. Under normal operating conditions, a maximum of approximately 30 to 35% of the storage rack positions would be filled with spent fuel bundles leaving approximately 65 to 70% of the fuel storage pool empty. Administrative controls will specify the fuel storage pool area to be filled with spent fuel bundles for decay storage, in order to assure that, if required, the fuel storage pool could be traversed by the loaded five-ton general-purpose crane without moving a load over stored spent fuel. Administrative controls are to be used to prevent the polar crane from handling large loads over the upper pools in the containment. Transfer of fuel assemblies between the new fuel vault and the water-filled storage pool and also within the storage pool, trans-fer pool and cask vault is performed with the fuel-handling plat-form. Either the fuel grapple or the auxiliary fuel hoist is used, depending on the transfer operation. The grapple and hoist, pro-vided with load sensing and limiting devices, are designed to the following load limits: Fuel Grapple Auxiliary lioi st (lb) (lb) __ load-limiting switch 1200 1000 load-sensing switch 485 485 stall torque of hoist system 3000 3000 The load-limiting features of the platform fuel grapple and auxiliary hoist will prevent damage to the fuel racks during fuel transfer operations. These load limits provide a redundant safety feature, since the fuel-handling grapples are not lowered below 9.1-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.2.3.3 Protective Features of Spent Fuel Storage Facilities
'(\-)) (Continued) the upper fuel rack and they will interface only with the fuel briil .
Guard rails around the spent fuel pool prevent the falling of fuel-handling arca machinery into the pool. Other objects that could conceivably fall into the fuel shall not transfer energy amounts exceeding the specified limits of the fuel racks. The preclusion of accidental dropping of the spent fuel cask on the spent fuel racks is accomplished by structural barriers at the end of the cask crane rail which prevent cask movement over the stored fuel. The spent fuel cask is transferred through the canal in the floor with the center c' oravity of the cask below floor level.
Normal motion of the cask crane is to move the cask to the center of the cask vault under operator control. Motion past this point would move the cask crane past electrical power cutoffs and into mechanical barriers at the end of the crane rail. Accidental release of the cask would not result in the cask dropping into the fuel storage pool. Layout of the cask-handling equipment, as shown in the general arrangement, provides a dry canal for transfer of the cask between the transport vehicle and the cask l
pool. The canal arrangement contributes to cask-handling safety by reducing the overall cask lift height and allowing transfer s l with the major portion of the cask below the fuel building oper-ations floor. The elevation of the cask canal floor between the cask decontamin-l ation vault and the cask pool is set to permit moving the cask into the cask pool with the bottom of the cask no higher than 30 ft above the floor of the cask pool. The gate between the cask pool ("N
\' and the spent fuel storage pool is recessed into the wall between l
9.1-15
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0
/
9.1.2.3.3 Protective Features of Spent Fuel Storage Facilities (Continued) the pools. This arrangement prevents the cask from contacting the pool gate if it should swing against the canal entranco during handling and transport of the cask, thus insuring the integrity of the gate. However, in the event of gate failure, the loss of pool water f rorn the storage pool into the cask pool would not uncover the stored spent fuel. The Fuel Building and neactot Building fuel storage facilities are designed to Seismic Category I Icquirements to prevent earthquake damage to the stored fuel. The_er.terior wall and roof of the fuel building superstructure are designed as a low leakage barrier to confine potential airborne radiation contamination within the fuel building and the erhaust air-treatment systems, The Fuel Building is part of secondary containment. The shield building construction is designed to serve primarily as a radiation shie'd during accident conditions, and the exterior siding and roof of the fuel building superstructure are designed to prevent damage to the stored fuel by tornado borne missiles. The design of the reinforced concrete and roof utilized is based on analysis of tornado conditions and missile protection design bases presented in Section 3.5. Offsite exposure to release of radioactive products from damaged or failed fuel in the Fuel Building is dependent on three systems. The functions of these systems, as described in Subsection 9.4.2, are: (1) the ventilation exhaust radiation monitoring system detects radioactivity in the Fuel Building atmosphere; (2) the Standby Gas Treatment System minimizes the release of contaminated air to the environment; and (3) the Fuel Building isolation control system automatically closes isolation dampers to block potential leakage of contaminated air to the environment. O 9.1-16
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O a
~
I (N 9.1.2.3.3 Protective Features of Spent Fuel Storage Facilities
\s-) (Continued)
The FPCCU system described in Subsection 9.1.3 provides adequate and continuous cooling for the spent fuel. From the foregoing analyses, it is concluded that the spent fuel storage arrangement and design meet the safety design bases and satisfy the intent of Regulatory Guide 1.13. 9.1.2.4 Testing Inspection The spent fuel storage racks require no periodic special testing or inspection for nuclear safety purposes. 9.1.2.5 Summary of Radiological Considerations 7-~g By adequate design and careful operational procedures, the safety (_s/ design bases of the spent fuel storage arrangement are satisfied. Thus, the exposure of plant personnel to radiation is maintained well below publiehed guideline values. Further details of radio-logical considerations, including those for the spent fuel storage arrangement, are presented in Chapter 12. 9.1.3 Fuel Pool Cooling and Cleanup System 9.1.3.1 Design Bases 9.1.3.1.1 Safety Design Bases The Fuel Pool Cooling and Cleanup (FPCCU) System shall be designed to remove the decay heat from the fuel assemblies, maintain pool water level and remove radioactive materials from the pool and thus minimize the release of radioactive elements stored in the contain-rN ment upper pool and the pools in the fuel building. 9.1-17
GESSAn 11 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.3.1.2 Power Generation Design Bases The FPCCU System shall: (1) minimize corrosion product buildup and shall control water clarity, so that the fuel assemblies can be efficiently handled underwater; (2) minimize fission product concentration in the water which could be released from the pool to the refueling building environment; (3) monitor fuel pool water level and maintain a water level above the fuel sufficient to provide shielding for normal building occupancy; (4) maintain the pool water temperature below 125 F under normal operating conditions. The temperature limits of 125 F is set to establish a minimum acceptable environ-ment for personnel working in the vicinity of the fuel pool. A maximum normal heat load from spent fuel stored in the pool is the sum of the decay heat released by the average spent fuel batch discharged from the 18-month equilibrium fuel cycle at the earliest refueling time, plus the heat being released by the batch discharged at the previous refueling having been in the pool for one year. The heat sources are based on full power operation for four years prior to removal of fuel assemblies from the reactor and a batch size of 37% of the core. Also, the system shall remove the heat transferred from the drywell to the containment pool through the drywell head. The RIIR syscem will be used to supplement the FPCCU Sys-tem under the maximum load condition as defined in Subsection 9.1.3.3. (5) maximize water purity for visual purposes. 9.1-18
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/"'1 9.1.3.2 System Description The PPCCU System (Figures 9.1-24a,b,c and 9.1-25a,b) maintains the coatainment pool, the spent fuel and cask storage pool and the fuel transfer pool below a desired temperature, at an acceptable radiation level and at a degree of clarity necessary to transfer and service the fuel bundles. It also maintains the containment pool temperature, radiation level and clarity necessary to transfer and service the reactor internals and fuel bundles.
The FPCCU System cools the fuel storage pool by transferring the spent fuel decay heat through two 8.8 x 10 6 Btu /hr heat exchangers to the essential service water system. Each of the two heat exchangers is designed to transfer one half the system design heat load. The system utilizes two parallel 1100 gpm pumps to provide a system design flow of 2200 gpm. Each pump is suitable for con-tinuous duty operation. The major portion of the equipment is () located in the Fuel Building except for the valves, piping and instrumentation associated with the containment pool. This equip-ment is located in the reactor building. The system pool water temperature is maintained at or below 125 F. The decay heat released from the stored fuel is transferred to the essential service water system. The Residual lleat Removal (RHR) System supplements the FPCCU to remove abnormal heat loads such as when a larger batch than normal is removed from the core. Fuel storage pool water is circulated by means of overflow through skimmers around the periphery of the pool and a scupper at the end of the transfer pool. The overflow is collected in the fuel pool drain tank and the flow passes through the heat exchanger and filter-domineralizer and back to the pool through the diffusers. Likewise, the containment pool water overflow is collected in the fuel pool drain tank, where against a closed-loop circulation is () attained. The fuel pool drain tank is sized to contain the water 9.1-19 l
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0 9.1.3.2 System Description (Continued) drained from the inclined transfer tube during downward fuel trans-fer, as well as the volume of water above the skimmer weirs, which drains from the pools following a temporary loss of circulation. Clarity and purity of the pool water are maintained by a combina-tion of filtering and ion exchange. The filter-demineralizers maintain total dissolved heavy element content (Cu, 'fi , Fe, lig , etc.) at 0.1 ppm or less with a pil range of 6.0 to 7.5 for com-patibility with aluminum fuel storage racks and other equipment. Each tilter unit in the filter-domineralizer subsystem has adequate capacity to maintain the desired purity level of the pools under normal operating conditions. The flow rate is designed to be approximately that required for two complete water changes per day for the fuel transfer and storage pools. The maximum system flow rate is twice that needed to maintain the specified water quality. Water may be returned to condensate storage after being filtered and demineralized. The FPCCU System is designed to remove suspended or dissolved impur-ities from the following sources: (1) dust or other airborne particles; (2) surface dirt dislodged from equipment immersed in the l pool; i (3) crud and fission products emanating from the reactor l during refueling; i (4) debris from inspection or disposal operations; and 1 (5) residual cleaning chemicals or flush water. O 9.1-20
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 7_ ) t 9.1.3.2 System Description (Continued) The filter-demineralizer vessel is constructed of phenolic resin-coated carbon steel. A post-strainer in the effluent stream of the filter-demineralized limits the migration of filter material. The filter-holding element can withstand a differential pressure greater than the developed pump head for the system. The filter-demineralizer units are located separately in shielded cells with enough clearance to permit removing filter elements from the vessels. Each cell contains only the filter-demineralizer and piping. All valves (inlet, outlet, recycle, vent, drain, etc.) are located on the outside of one shielding wall of the room, together with necessary piping and headers, instrument elements and controls. Penetrations through shielding walls are located so as not to () compromise radiation shielding requirements. The filter-demineralizers are controlled from a local panel. A differential pressure and conductivity instrument provided for each filter-demineralizer unit indicates when backwash is required. Suitable alarns, differential pressure indicators and flow indi-cators monitor the condition of the filter-demineralizers. System instrumentation is provided for both automatic and remote-manual operations. A low-low level switch stops the circulating pumps when the fuel pool drain tank reserve capacity is reduced to the volume that can be pumped in approximately one minute with one pump at rated capacity (1100 gpm). Level switches are also located in both containment and fuel storage pools. Whenever the water levels are too high or too low, an alarm and indicator light are activated. Also monitored is the water level in the drain collection tank. A low level signal initiates closure or contain-() nien t isolation valves. Temperature elements are provided to 9.1-21
GESSAR II 22A7007 238 NUCLEA7 ISLAND Rev. 0
') . l . 3 . 2 System Description (Continued) record pool temperature in the main control room. (Spent Pael Pool Cooling and Cleanup System Instrumentation and Controls are described in Section 7.6.)
The circulating pumps are controlled from the control room and a local panel. Pump low suction pressure automatically turns off the pumps. A pump low discharge pressure alarm is indicated in the control room and on the local panel. The circulating pump motors are powered frem their corresponding unit shutdown board. These boards receive power from the diesel-generators if normal power is not available. Circulating pump motor loads are con-sidered nonessential loads and will be operated as required under accident conditions. The water level in the spent fuel storage pool is maintained at a height which is sufficient to provide shielding for normal build-ing occupancy. Radioactive particulates removed from the fuel pool are stored in filter-demineralizer units which are located in shielded cells. For these reasons, the exposure of plant personnel to radiation from the FPCCU System is minimal. Further details of radiological considerations for this system are described in Chapter 12. l The circulation patterns within the containment upper pool and spent fuel storage pool are established by placing the diffusers and skimmers so that particles dislodged during refueling opera-tions are swept away from the work area and out of the pools. The return lines to the pool are prevented from siphoning the pool in the event of a pipe rupture by redundant vacuum breakers at the high point of the lines. O 9.1-22
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 "T 9.1.3.2 System Description (Continued) (b Heat from pool evaporation is handled by the building ventilation system. Makeup water is provided through a remote-operated valve. Irradiated fuel shall not be stored in the upper containment stor-age pool during reactor operation. 9.1.3.3 Safety Evaluation The maximum possible heat load is the decay heat of the full core load of fuel at the end of the fuel cycle plus the remaining decay heat of the spent fuel discharged at previous refuelings. The temperature of the fuel pool water may be permitted to rise to approximately 150 F under these conditions. During shutdown con-ditions, if it appears that the fuel pool temperature will exceed 125 P, the operator connects the FPCCU System to the RHR System. / Combining the capacities enables the two systems to keep the water temperature below 125 F. The RHR System will only be used to supplement the fuel pool cooling when the reactor is shut down. The reactor will not be started up whenever portions of the RHR systems are needed to cool the fuel pool. The connecting piping from the fuel storage pool to the RHR system is designed Seismic Category I and is completely independent of the fuel pool system piping. These connections may also be utilized during emergency conditions to assure cooling of the spent fuel regardless of the availability of the fuel pool cooling system. The volume of water in the storage pool is such that there is enough heat absorption capability to allow sufficient time for switching over to the RHR system for emergency cooling. The 150 F temperature limit is set to assure that the fuel build-ing environment does not exceed equipment environmental limits. O 9.1-23
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.3.3 Safety Evaluation (Continued) The spent fuel storage pool is designed so that no single failure of structures or equipment will cause inability to: (1) maintain irradiated fuel submerged in water; (2) re-establish normal fuel pool water level; or (3) remove decay heat from the pool. In order to limit the possibility of pool leakage around pool pene-trations, the pool is lined with stainless steel. In addition to providing a high degree of integrity, the lining is designed to withstand abuse that might occur when equipment is moved about. No inlets, outlets or drains are provided that might permit the pool to be drained below a safe shielding level. Lines extending below this level are equipped with syphon breakers, checkvalves, or other suitable devices to prevent inadvertent pool drainage. Interconnected drainage paths are provided behind the liner wolds. These paths are designed to: (1) prevent pressure buildup behind the liner plate; (2) prevent the uncontrolled loss of contaminated pool water to other relatively cleaner locations within the con-tainment or fuel-handling area; and (3) provide expedient liner leak detection and measurement. These drainage paths are formed by welding channels behind the liner weld joints and are designed to permit free gravity drainage or pumping to the equipment drain tank. A makeup water system and pool water level instrumentation are provided to replace evaporative and leakage losses. flakeup water during normal operation will be supplied from condensate. Redundant loops of the essential service water system (which are both Seismic Category I) can be used as a source of makeup water in case of failure of the normal makeup water system. The cooling portion of the fuel pool system is designed to Seismic Category I up to and including the isolation valves for the filter demineral-izer. Also, a Seismic Category I bypass is provided around the filter demineralizer. This will assure continued performance of the heat removal function if the filter-demineralizer portion is 9.1-24
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 9.1.3.3 Safety Evaluation (Continued) damaged by a seismic event. The heat absorption capacity of the storage pools provides several hours to activate the valves manually to bypass the filter-demineralizer portion and provide cooling of the spent fuel storage pools. From the foregoing analysis, it is concluded that the FPCCU System meets its design bases. Regulatory Guide 1.13 This regulatory guide is applicable, in part, to fuel pool cooling and cleanup systems and requires adequate filtering, cooling and makeup of spent fuel storage pool water. For commitment revision and number, see regulatory guide commitment matrix in Section 1.8. () 9.1.3.4 Inspection and Testing Requirements No special tests are required because, normally, one pump, one heat exchanger and one filter-demineralizer are operating while fuel is stored in the pool. The spare unit is operated period-ically to handle abnormal heat loads or to replace a unit for servicing. Routine visual inspection of the system components, instrumentation and trouble alarms is adequate to verify system operability. 9.1.3.5 Radiological Considerations The water level in the spent fuel storage pool is maintained at a height which is sufficient to provide shielding for normal building occupany. Radioactive particulates removed from the fuel pool are collected in filter-demineralizer units which are located in shielded cells. For these reasons, the exposure of plant personnel s_- 9.1-25
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.3.5 Radiological Considerations (Continued) to radiation from the FPCCU System is minimal. Further details of radiological considerations for this and other systems are described in Chapters 11, 12 and 15. 9.1.4 Fuel-Handling Syster 9.1.4.1 Design Bases The Fuel-Handling System is designed to provide a safe and offec-tive means for transporting and handling fuel from the time it reaches the plant until it leaves the plant after post-irradiation cooling. Safe handling of fuel includes design considerations for maintaining occupational radiation exposures as low as practicable during transportation and handling. Design criteria for major Fuel-Handling System equipment are pro-vided in Tables 9.1-2 through 9.1-4, which list the safety class, quality group and seismic category. Where applicable, the appropriate ASME, ANSI, Industrial and Electrical Codes are identified. Additional design criteria are shown below and expanded further in Subsection 9.1.4.2 (System Dc3cription). The transfer of new fuel assemblies between the uncrating area and the new fuel inspection stand and/or the ne, "nel storage vault is accomplished using the 5-ton fuel builtin, crane i equipped with a suitable grapple. l l The 1,000-lb auxiliary hoist on the fuel-handling platform is used with an auxiliary fuel grapple to transfer new fuel from the new fuel vault to the fuel storage pool. From this point on, the fuel will either be handled by the telescoping grapples on the fuel-handling platform or on the refueling platform, and will be trans-ported between the Reactor and Fuel Buildings by the Fuel Transfer System. l 9.1-26
GESSAR II 22A7007 238 UUCLEAR ISLAND Rev. 0 9.1.4.1 Design Bases (Continued) These platforms are Safety Class 2 and Seismic Category I from a structural standpoint in accordance with 10CFR50, Appendices A and B. Allowable stress due to safe shutdown earthquake (SSE) loading is 120% of yield or 70% of ultimate, whichever is least. A dynamic analysis is performed on the structures using the response spectrum method with load contributions resulting from each of three earthquakes being combined by the RMS procedure. Working loads of the platform structures are in accordance with the AISC Manual of Steel Construction. All parts of the hoist systems are designed to have a safety factor of five, based on the ultimate strength of the material. A redundant load path is incorporated in the fuel hoists so that no single component failure could result in a fuel bundle drop. Maximum deflection limitations are imposed on the main structures to maintain relative stiffness of the plat-form. Welding of the platforms is in accordance with AWS D14-1 or ASME Boiler and Pressure Vessel Code Section IX. Gears and bear-ings meet AGMA Gear Classification Manual and ANSI B3.5. Materials used in construction of load bearing members are to ASTM specifi-cations. For personnel safety, OS!!A Part 1910-179 is applied. Electrical equipment and controls meet ANSI CI, National Electric Code, and NEMA Publication No. ICl, MGl. The auxiliary fuel grapple and the main telescoping fuel grapples have redundant lifting features and an indicator which confirms positive grapple engagement. . The fuel grapple is used for lifting and transporting fuel bundles. It is designed as a telescoping grapple that can extend to the proper work level and, in its fully retracted state, still maintain adequate water shielding over fuel. I In addition to redundant electrical interlocks to preclude the possibility of raising radioactive material out of the water, the l 9.1-27 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.1 Design Bases (Continued) cables on the auxiliary hoists incorporate an adjustable, removal stop that will jam the hoist cable against some part of the plat-form structure to prevent hoisting when the free end of the cable is at a preset distance below water level. Provision of a separate cask loading pool, capable of being isolated from the fuel storage pool, will eliminate the potential accident of dropping the cask and rupturing the fuel storage pool. Furthermore, limitation of the travel of the crane handling the cask will preclude transporting the cask over any fuel storage pool. (See Chapter 15 for accident considerations.) 9.1.4.2 System Description Table 9.1-1 is a listing of typical tools and servicing equipment supplied with the nuclear system. The following paragraphs describe the use of some of the major tools and servicing equip-ment and address safety aspects of the design where applicable. 9.1.4.2.1 Spent Fuel Cask (Applicant to supply) 9.1.4.2.2 Cask Crane (Applicant to supply) 9.1.4.2.3 Fuel Servicing Equipment The fuel servicing equipment described below has been designed in accordance with the criteria listed in Table 9.1-2. O 9.1-28
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ('s 9.1.4.2.3.1 Fuel Prep Machine b The fuel preparation machine (Figure 9.1-3) is counted on the wall of the fuel storage pool and is used for stripping reusable channels from the spent fuel and for rechanneling of the new fuel. The machine is also used with the fuel inspection fixture to pro-vide an underwater inspection capability. The fuel preparation machine consists of a work platform, a frame and a moveable carriage. The frame and moveable carriage are located below the normal water level in the fuel storage pool, thus providing a water shield for the fuel assemblies being handled. The fuel preparation machine carriage has a permanently installed up-travel-stop to prevent raising fuel above the safe water shield level. The moveable carriage is operated by a foot pedal controlled air hoist. 9.1.4.2.3.2 New Fuel Inspection Stand The New Fuel Inspection Stand (Figure 9.1-4) serves as a support for the new fuel bundles undergoing receiving inspection and pro-vides a working platform for technicians engaged in performing the inspection. The New Fuel Inspection Stand consists of a vertical guide column, a lift unit to position the work platform at any desired level, bearing seats and upper clamps to hold the fuel bundles in position. 9.1.4.2.3.3 Channel Bolt Wrench The channel bolt wrench (Figure 9.1-5) is a manually operated device approximately 12 ft (3.6m) in overall length. The wrench is used for removing and installing the channel fastener assembly
/~h while the fuel assembly is held in the fuel preparation machine.
9.1-29
GESCAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.3.3 Channel Bolt Wrench (Continued) ll The channel bolt wrench has a secket which mates and captures the channel fastener capscrew. 9.1.4.2.3.4 Channel-Ilandling Tool The channel-handling tool (Figure 9.1-6) is used in conjunction with the fuel preparation machine to remove, install and transport fuel channels in the fuel storage pool. The tool is composed of a handling bail, a lock / release knob, extension shaft, angle guides and clamp arms which engage the fuel channel. The clamps are actuated (extended or retracted) by manually rotating lock / release knob. The channel-handling tool is suspended by its bail from a spring balancer on the channel-handling boom located on the fuel pool periphery. 9.1.4.2.3.5 Fuel Pool Sipper The fuel pool sipper (Figure 9.1-7) provides a means of isolating a fuel assembly in domineralized water in order to concentrate fission products in relation to a controlled background. The fuel pool cipper consists of a control panel assembly and a sipping container cover. 9.1.4.2.3.6 General-Purpose Grapple The general-purpose (Figure 9.1-8) is a handling tool used gen-erally with the fuel. The grapple can be attached to the jib crane to handle fuel during channeling. O 9.1-30
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.3.7 Jib Crane The jib trane (Figure 9.1-9) consists of a motor-driven swing boom monorail and a motor-driven trolley with an electric hoist. The jib crane is mounted along the edge of the fuel building fuel storage pool to be used during refueling operations. Use of the jib crane leaves the refueling platform or fuel-handling platform 1, free to perform general fuel shuffling operations and still permit uninterrupted fuel preparation in the work area. The hoist has two full-capacity brakes and in-series adjustable up-travel limit switches. Upon hoisting, the first two independently adjustable limit switches automatically stop the hoist cable terminal approx-imately 8 ft below the jib crane base. Continued hoisting is possible by depressing a momentary contact, up-travel override pushbutton on the pendant together with the normal hoisting push button. Two additional independent switches automatically cut hoist power at the maximum safe uptravel limit. When the jib crane is esed in the handling of hazardous radioactive materials O) ( that must be kept below a specific water level, a fixed tool length stop is ir. stalled on the hoist cable to prevent further hoisting when that level is reached. The jib crane is normally located adjacent to the fuel storage pool and connected to the service outlet provided. 9.1.4.2.3.8 Fuel-Handling Platform Refer to Subsection 9.1.4.2.7 for a description of the fuel-handling platform. 9.1.4.2.3.9 Channel-Handling Boom A channel-handling boom (Figure 9.1-10) with a spring-loaded balance reel is used to assist the operator in supporting a portion of the weight of the channel as it is removed from the fuel assem-bly. The boom is set between the fuel preparation machines. With () the channel-handling tool attached to the reel, the channel may be conveniently moved between the fuel preparation machines. l 9.1-31
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.3.10 Fuel Transfer System The Inclined Fuel Transfer System (Figure 9.1-11) is used to transfer fuel, control rods, defective fuel storage containers and other smal] items between the containment and the fuel building pools by means of a carriage traveling in a transfer tube (a 23-in. I.D. stainless steel pipe). In the containment upper pool, the transfer tube connects to pool penetration and to a sheave box. Connected to the sheave box is a 24-in. flap valve, a vent pipe, cable enclosures and a fill valve. In the fuel building pool, the transfer tube connects to a 24-in. gate valve. A bellows connects the building penetration to the valve and transfer tube to prevent water entrapment between the tube and penetration. A 4-in. Weldolet located on the transfer tube approximately 2 ft above the fuel building pool water level and a motor-operated valve are provided for connections to a drain pipe for water level control in the transfer tube. A containment isolation assembly containing a blind flange and a bellows, which connects the containment isolation assembly to the containment penetration, is provided to make containment isolation. A hand-operated 24-in. gate valve is provided to isolate the reactor building pool water from the transfer tube so that the blind flange can be installed. A hydraulically actuated upender is provided in each pool for rotating part of the carriage (tilt tubn) to the vertical position for loading and unloading and to the inclined position for trans-fer. The carriage consists of the tilt tube and a follower con-nected with a pivot pin which allows upending of the tilt tube while maintaining the follower in the inclined position. The carriage has rollers and v; heels which ride on tracks within the transfer tube and upenders to assure low friction, correct carriage orientation and smooth transition across valves and between other components. The tilt tube is designed to accept two different inserts - a fuel bundle insert with a two-bundle capacity and a control rod insert for control rods, defective fuel storage container, and other small items. 9.1-32
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
- 9.1.4.2.3.10 Fuel Transfer System (Continued)
A winch, located on the containment refueling floor, uses two cables attached to the lower end of the follower for pulling the carriage from the fuel building to the containment and for con-trolling the carriaga descent velocity. A slow winch speed is provided for starting and stopping the carriage to limit the acceleration on the fuel assemblies. Cable underload and overload protection is provided by a load cell. Carriage position readout in provided. Cable enclosures, attached to the sheave box and projecting above the containment upper pool water level, provide the means for cable exit from the transfer tube while isolating the pool water from the tube. A vent pipe, with a fluid stop connected to the containment venti-lation system, isolates the displaced air in the tube during filling from the Reactor Building atmosphere and confines the water surge to the pool water. A hydraulic power unit is provided in each building to actuate the cylindars attached to the upenders, fill valve, flap valve and Fuel Building gate valve. In both buildings, the pool area in which the transfer system com-ponents are located is physically separated from the fuel storage area by a concrete wall which serves as a positive barrier to prevent fuel in the storage area from being uncovered in the event of loss of pool water through the transfer system. In addition, these walls are provided with gates to allow drainage of the trans-for pool areas for maintenance and/or removal of the transfer tube , and components. Control panels are provided in close proximity to each transfer pool area and are connected for voice and interlock communication. Each panel has control buttons for actuating the upender, a button for initiating the transfer sequence to the other building and a l 9.1-33
GESSAR ZI 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.3.10 Fuel Transfer System (Continued) The transfer operation functions on an automatic O stop button. basis with provision made for manual override. Automatic sequenc-ing is accomplished by use of an electronic controller located in the fuel building, which utilizes sensors for confirming the successful completion of each step before initiating the next step. The completion of a transfer sequence is signaled at the control panels. Interlocks assure the correct sequencing of the transfer system components and fuel-handling equipment during automatic or manual override operation. Interlocks prevent the refueling platform from moving into the Reactor Building transfer area unless the gate valve (12) is closed and the upender (3) is in the vertical posi-tion, and prevent upender movement if the platform is in the transfer area. Interlocks prevent the fuel-handling platform from moving into the Fuel Building transfer area unless the upender (13) is in the vertical position, and prevent movement of the upender if the platform is in the transfer area. The refueling interlocks instrumentation and control and other control safety aspects of the refueling system are described and evaluated in Section 7.6. The operational sequence for the Fuel Transfer System (Figure 9.1-
- 13) is described as follows. As a starting point, assume the carriage is in the containment transfer pool with the tilt tube (10B) supported by the upender (3) in the inclined position. In this position, the sheave box cover and gate valve (7) are open with the fill valve (4), gate valve (12) and drain valve (9) closed.
The operational sequence is as follows: i (1) The hydraulic cylinder is actuated to push the upender , and tilt tube (3 and 10B) to the vertical position. 9.1-34 l I l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.3.10 Fuel Transfer System (Continued) fs\ v (2) Load and unload fuel, control rods or other items into and from the tilt tube. (3) The hydraulic cylinder is actuated to pull the tilt tube into the inclined position for transfer. (4) The automatic operation is started by depressing the transfer button on the containment control panel. This starts the winch (3) unwinding the cables to lower the carriage. (5) The carriage is stopped approximately 2 feet above the gate valve (7), (6) The sheave box cover is closed. (7) The drain valve (9) is opened and water is drained to () V the level of drain pipe attachment to the transfer pipe. (8) The gate valve (12) is opened. (9) The winch lowers the carriage until it is stopped and supported by the pivot arm framing (14). (10) The hydraulic cylinder is actuated to push the upender (13) and tilt tube (10B) to the vertical position. (11) Unload and load cargo. (12) The hydraulic cylinder is actuated to lower the tilt tube and upender to the inclined position.
-(13 ) The winch is actuated by depressing the fuel building control panel's transfer button and pulls the carriage to a position approximately 2 feet above the gate valve (12), where it is automatically stopped, s ,/ (14) The gate valve (12) and drain valve (9) are closed.
9.1-35
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.3.10 Fuel Transfer System (Continued) (15) The fill valve (4) is opened. (16) The sheave box cover is opened when sensors indicate that the transfer tube, sheave box, vent pipe and cable enclosures are filled with water. (17) The carriage is pulled to the containment transfer pool (starting point). After transfer operations are completed, the carriage will be stored in the containment transfer pool on the upender (3). Containment isolation is then made as follows: (1) Close the manual gate valve. (2) Remove bolts from the containment isolation assembly as required to allow insertion of the blind flange. (Not all bolts will have to be removed.) Loosen remain-ing bolts to allow approximately 1.25 in. movement of the transfer tube flange. (3) Lower transfer tube with the hydraulic cylinders. (4) Insert the blind flange and install bolts. (5) Pull the transfer tube up with the cylinders. (6) Tighten the bolts and relieve pressure on the cylinders. Containment is made by the containment isolation essembly and blind flange, containment bellows (6) and the steel containment penetration. Special gaskets and double-ply bellows are provided for leak checking to assure containment isolation. Refer to Table 9.1-2a for component identification essential classifications, safety classifications, quality groups and seismic categories.
- 9.1-36
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (~) 9.1.4.2.4 Servicing Aids V General area underwater lights are provided with a suitable reflector for illumination. Suitable light support brackets are furnished to support the lights in the reactor vessel to allow the light to be positioned over the area being serviced independent of the platform. Local area underwater lights are small diameter lights for additional illumination. Drop lights are used for illumination where needed. A radiation-hardened designed portable underwater closed circuit television camera is provided. The camera may be lowered into the reactor vessel and/or fuel storage pool to assist in the inspection and/or maintenance of these areas. The camera lens is capable of pitching ninety degrees, which allows infinite scanning of three hundred and sixty degrees, solid angle. A general-purpose, plastic viewing aid is provided to float on the water surface to provide better visibility. The sides of the viewing aid are brightly colored to allow the operator to observe it in the event of filling with water and sinking. A portable, submersible-type, underwater vacuum cleaner is provided to assist in removing crud and miscellaneous particulate matter from the pool floors or reactor vessel. The pump and the filter unit are completely submersible for extended periods. The filter " package" is capable of being remotely changed, and the filters will fit into a standard shipping container for offsite burial. Fuel pool tool accessories are also provided to meet servicing requirements. A fuel sample is provided. This is to be used to detect defective fuel assemblies during open vessel periods while the fuel is in the core. The fuel sample head isolates individual fuel assemblies by sealing the top of the fuel channel and pumping water from the bottom of the fuel assembly, through the fuel channel, to a sampling station, and returning the water to the primary coolant system. After a " soaking" period, a water sample is obtained and is radio-chemically analyzed. 9.1-37
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.5 Reactor Vessel Servicing Equipment The essentiality and safety classifications, the quality group, and the seismic category for this equipment are listed in Table 9.1-3. Following is a description of the equipment designs in reference to that table. 9.1.4.2.5.1 Reactor vessel Service Tools These tools are used when the reactor is shut down and the reactor vessel head is being removed or reinstalled. Tools in this group are: Stud Handling Tool Stud Wrench Nut Runner Stud Thread Protector Thread Protector Mandrel Bushing Wrench Seal Surface Protector Stud Elongation Measuring Rod Dial Indicator Elongation Measuring Device Head Guide Cap These tools are designed for a 40-yr life in the specified environ-ment. Lifting tools are designed for a safety factor of 5 or better with respect to the ultimate strength of the material used. When carbon steel is used, it is either hard chrome plated, O 9.1-38
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 9.1.4.2.5.1 Reactor Vessel Service Tools (Continued) parkerized, or coated with an approved paint per Regulatory Guide 1.54. 9.1.4.2.5.2 Steamline Plug The steamline plugs are used during reactor refueling or servicing; they are inserted in the steam outlet nozzles from inside of the reactor vessel to prevent a flow of water from the reactor well into the main steamline during servicing of safety relief valves, main isolation valves, or other components of the main steamlines, while the reactor water level is at the refueling level. The steamline plug design provides two seals of different types. Each one is independently capable of holding full head pressure. The equipment is constructed of noncorrosive materials. All calculated safety factors are 5 or better. The plug body is designed in , () accordance with the " Aluminum Construction Manual" by the Aluminum Association. 9.1.4.2.5.3 Shroud llead Bolt Wrench This is a hand-held tool for operation of the shroud head bolts. It is designed for a 40-yr life and is made of aluminum for easy handling and to resist corrosion. Testing has been performed I to confirm the design. l 9.1.4.2.5.4 Head Holding Pedestal Three pedestals are provided for mounting on the refueling floor for supporting the reactor vessel head and strongback/ carousel during periods of reactor service. The pedestals have studs which engage three evenly spaced stud holes in the head flange. The flange surface rests on replaceable wear pads made of aluminum. O V i 9.1-39
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.5.4 Head Holding Pedestal (Continued) When. resting on the pedestals, the head flange is approximately 3 ft above the floor to allow access to the seal surface for inspection and 0-ring replacement. The pedestal structure is a carbon steel weldment coated with an approved paint. It has a base with bolt holes for mounting it to the concrete floor. A seismic analysis was made to determine the seismic forces imposed onto the pedestals, floor an-hors, using the floor response spectrum method. The structure is designed to withstand these calculated forces and meet the requirements of AISC. 9.1.4.2.5.5 Head Stud Rack The head stud rack is used for transporting and storage of eight reactor pressure vessel studs. It is suspended from the Reactor Building polar crane hook when lifting studs from the reactor well to the operating floor. The rack is made of aluminum to resist corrosion and is designed for a safety factor of 5 with respect to the ultimate strength of the material. The structure is designed in accordance with the " Aluminum Con-struction Manual" by the Aluminum Association. 9.1.4.2.5.6 Dryer and Separator Strongback The Dryer and Separator Strongback is a lifting device used for transporting the steam dryer or the shroud head with the steam separators between the reactor vessel and the storage pools. The O 9.1-40
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev, 0 9.1.4.2.5.6 Dryer and Separator Strongback (Continued) ( strongback is a cruciform-shaped I-beam structure, which has a hook box with two hook pins in the center for engagement with the reactor building polar crane sister hook. The strongback has a socket with a pneumatically operated pin on the end of each arm for engaging it to the four lift eyes on the steam dryer or shroud head. The strongback has been designed such that one hook pin and one main beam of the cruciform will be capable of carrying the total load and so that no single component failure will emise the load to drop or swing uncontrollably out of an essentially level atti-tude. The safety factor of all lifting members is 5 or better in reference to the ultimate breaking strength of the material. The structure is designed in accordance with "The Manual of Steel () Construction" by AISC. The completed assembly is proof-tested at 125% of rated load, and all structural welds are magnetic-particle inspected after load test. 9.1.4.2.5.7 Head Strongback/ Carousel Th3 RPV Head Strongback/ Carousel is an integrated piece of equip-ment consisting of a cruciform-shaped strongback, a circular mono-rail and a circular storage tray. The strongback is a box-beam structure which has a hook box with two hook pins in the conter for engagement with the reactor ser-vice crane sister hook. Each arm has a lift rod for engagement to the four lift lugs on the RPV head. The monorail is mounted on extensions of the strongback arms and four additional arms equally spaced between the strongback arms. The monorail circle matches the stud circle of the reactor vessel and it serves to suspend stud tensioners and nut-handling devices. The storage tray is ( 9.1-41
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.5.7 Head Strongback/ Carousel (Continued) suspended from the ends of the same eight arms and surrounds the RPV flange. A manifold is mounted underneath the hook box for distributing hydraulic and pneumatic pressures to equipment travel-ing on the monorail. The head strongback/ carousel serves the following functions: (1) Lifting of Vessel Head: The strongback, when suspended from the Reactor Building polar crane main hook, will transport RPV head plus the carousel with all its attach-ments between the reactor vessel and storage on the pedestals. (2) Tensioning of Vessel Head Closure: The carousel, when supported on the RPV head on the vessel, will carry eight tensioners, its own weight, the strongback, storage of nuts, washers, thread protectors, and associ-ll ated tools and equipment. The eight tensioners are suspended equally spaced from a monorail above the vessel stud circle. Each tensioner has an air-operated hoist with individual controls. (3) Storage with RPV Head: The carousel, when stored with the RPV head on the head holding pedestals, carries the same load for (2) above. When in storage position, it accommodates nut cleaning and inspection. (4) Storage without RPV Head: During reactor operation, the carousel is stored on the refueling floor, straddling tne three pedestals. The strongback, with its lifting components, is designed to meet the Crane Manufacturers Association of America, Specification No. 70. The design provides a 15% impact 9.1-42
1 GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 [~ 3 9.1.4.2.5.7 Ilead Strongback/ Carousel (Continued) Ng allowance and a safety factor of 5 in reference to the ultimate strength of the' material used. After comple-tion of welding and before painting, the lifting assembly is proof load tested and all load-affected welds and lift pins are magnetic-particle inspected. The steel structure is designed in accordance with "The Manual of Steel Construction" by AISC. Aluminum
, structures are designed in accordance with the " Aluminum Construction' Manual" by the Aluminum Association.
The strongback is designed in accordance with ASME, American National Standard for overhead hoists ANSI B30. 16-1973, Paragraph 16-1.2.2.2 and such that one hook pin and one main beam of the structure is capable of carry-(\ x,) ing the total lead, and so that no single component failure will cause the load to drop or swing uncontroll-ably out of an essentially level attitude. Regulatory Guide 1.54 General compliance or alternate assessment for Regulatory Guide 1.54, which provides design criteria for protective coatings, may be found in Subsection 6.1.2. 9.1.4.2.6 In-Vessel Serv.cing Equipment The instrument handling tool is attached to the refueling platform auxiliary hoist and is used for removing and installing neutron source holders. Each in-core instrumentation guide tube is sealed by an 0-ring on the flange, and, in the event that the seal needs replacing, an in-core guide tube sealing tool is provided. The f~h O 9.1-43
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.6 In-Vessel Servicing Equipment (Continued) tool is inserted into an empty guide tube and sits on the beveled guide tube entry in the vessel. When the drain on the Water Seal Cap is opened, hydrostatic pressure seats the tool. The flange can then be removed for seal replacement. The auxiliary hoist on the refueling platform is used with appro-priate grapples to handle control rods, flux monitor dry tubes, sources and other internals of the reactor. Interlocks on both the grapple hoists and auxiliary hoist are provided for safety purposes; the refueling interlocks are described and evaluated in Section 7.6. 9 1.4.2.7 Refueling Equipment Fuel movement and reactor servicing operations are performed from platforms which span the refueling, servicing and storage cavities. The containment is supplied with a refueling platform for fuel movement and servicing, an auxiliary platform for servicing operations from the refueling floor level and a vessel platform for reactor servicing from the vessel flange level. The fuel building is suppliod with a fuel-handling platform for fuel movement and servicing. 9.1.4.2.7.1 Refueling Platform The refueling platform is a gantry crane, which is used to trans-port fuel and reactor components to and from pool storage and the reactor vessel. The platform spans the fuel storage and vessel pools on bedded tracks in the refueling floor. A telescoping mast and grapple suspended from a trolley system is used to lift and orient fuel bundles for core, atorage rack or upender placement. Control of the platform is from an operator station on the main O 9.1-44
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3 9.1.4.2.7.1 Refueling Platform (Continued) trolley. A position-indicating system and travel limit computer is provided to locate the grapple over the vessel core and prevent collisions with pool obstacles. Two 1000-lb capacity auxiliary hoists, one main and one auxiliary monorail trolley-mounted, are provided for in-core servicing such as detector module replacement, fuel support replacement, jet pump servicing and con-trol rod blade replacement. The grapple in its fully retracted position provides 8 ft 6 in. minimum water shielding over the active fuel during transit. The fuel grapple hoist has a redundant load path so that no single component failure will result in a fuel bundle drop. Interlocks on the platform: (1) prevent unsafe operation over the vessel during control rod movements; (2) prevent collision with the auxiliary platform; (3) avoid unsafe operation in the transfer tube upender zone; (4) limit travel of the fuel grapple; and (5) interlock grapple hook engagement with hoist load and hoist up power. 9.1.4.2.7.2 Auxiliary and Vessel platforms An auxiliary platform is provided to allow versatility of opera-tions. This platform operates over the Reactor Building pool and provides an additional work area for reactor servicing. A 500-lb capacity hoist is provided for reactor servicing tasks. Part of the auxiliary platform is used as the vessel flange level service platform. The reactor level servicing platform provides a reactor flange level working surface for in-vessel inspection and reactor internals servicing, and permits servicing access for the full vessel diameter. Typical operations to be performed are in-service inspection and jet pump servicing. No hoisting equipment is provided with this plat-form, as this function can be performed from the refueling platform or auxiliary platform. The platform operates on tracks at the 9.1-45
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.7.2 Auxiliary and Vessel Platforms (Continued) reactor vessel flange level and is lowered into position by the reactor building crane using the dryer / separator strongback. The platform weighs approximately 4,000 lb and features 5-ft-wide work areas and motorized travel. The platform power is supplied by a cable from the refueling floor elevation, 9.1.4.2.7.3 Fuel-Handling Platform The fuel-handling platform is a gantry crane, which is used to transport fuel within the fuel building storage pool. The platform spans the fuel storage and transfer tube upender pools on tracks bedded in the fuel building floor. A telescoping mast and grapple is used to lift and orient fuel bundles for storage rack or upender placement. Control of the platform is from an operator station on the main trolley. A vertical position-indicating system is pro-vided for the grapple. Limit switches are located on the end trucks to interlock the platform from running into pool obstacles. A 1000-lb capacity auxiliary hoist is mounted on the auxiliary monorail trolley and is used for moving new fuel from the new fuel vault to the storage pool and control rod blade transport. Both main fuel hoist and monorail auxiliary hoist have redundant load paths such that no single component failure will result in a fuel bundle drop. During transfer of fuel, the grapple in its fully retracted position provides 8 ft 6 in. ninimum water shielding over the active fuel. 9.1.4.2.8 Storage Equipment Specially designed equipment storr7e racks are provided. Additional storage equipment is listed on Table 9.1-1. For fuel storage racks description and fuel arrangement, see Subsections 9.1.1 and 9.1.2. Defective fuel assemblies are placed in special fuel storage con-tainers, which are stored in the equipment storage rack, both of 9.1-46
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.8 Storage Equipment (Continued) ( which are designed for the defective fuel. These may be used to isolate leaking or defective fuel while in the fuel pool and during anipping. Channels can also be removed from the fuel bundle while in a defective fuel storage container. The fuel pool sipper may be used for out-of-core wet sipping at any time. It is used to detect a defective fuel bundle while circu-lating water through the fuel bundle in a closed system. The con-tainers cannot be used for transporting a fuel bundle. The bail on the container head is designed not to fit into the fuel grapple. 9.1.4.2.9 Under-Reactor Vessel Servicing Equipment The primary functions of the under-reactor vessel servicing equip-ment are to: (1) remove and install control rod drives; (2) ser-() vice thermal sleeves; and (3) install and remove the neutron detectors. Table 9.1-4 lists the equipment and tools required for. servicing. Of t.he equipment listed, the equipment-handling plat-form and the CBD handling equipment are powered pneumatically. The CRD handling equipment is designed for the removal and instal-lation of the control rod drives from their housings. This equip-ment is used in conjunction with the equipment-handling platform. It is designed in accordance with OSHA-1910.179 and American Institute of Steel Construction (AISC). The equipment-handling platform provides a working surface for equipment and personnel performing work in the under-vessel area. It is a polar platform capable of rotating 360 . This equipment l is designed in accordance with the applicable requirements of OSHA (Vol. 37, No. 202, Part 1910N), AISC, and ANSI-C-1 (National Electric Code). l 4 V ! 9.1-47 i l
m GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.9 Under-Reactor Vessel Servicing Equipment (Ccntinued) The water seal cap is designed to prevent leakage of primary cool-ant from in-core detector housings during detector replacement. It is designed to industrial codes, manufactured from noncorrosive material. The thermal sleeve installation tool locks, unlocks and lowers the thermal sleeve from the CRD guide tube. The in-core flange seal test plug is used to determine the pressure integrity of the in-core flange 0-ring seal. It is constructed of noncorrosive material. The key bender is designed to install and remove the antirotation key that is used on the thermal sleeve. 9.1.4.2.10 Description of Fuel Transfer The Fuel-Handling and Transfer System provides a safe and effective me,ns for transporting and handling fuel from the time it reaches the plant until it leaves the plant after post-irradiation cooling. Subsection 9.1.4.2.9 described the equipment and methods utilized in fuel handling. The following subsections describe the inte-grated fuel transfer system which ensures that the design bases of the fuel-handling system and the requirements of Regulatory Guide 1.13 are satisfied. 9.1.4.2.10.1 Arrival of Fuel on Site The new fuel is delivered to the plant on flatbed truck or railcar. The new fuel is delivered to the receiving stations in the Fuel Building through the rail and truck entry door. There, the in-coming new fuel is unloaded, inspected and prepared for use as described in Subsection 9.1.4.2.10.2.1. 9.1-48
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.10.2 Refueling Procedure The plant _ refueling and servicing sequence diagram is shown in Figure 9.1-12. Fuel-handling procedures are shown in Figures 9.1-13 through 9.1-20 and described below. Typical Fuel Building and Containment Building layouts are shown in Figures 9.1-21 and 9.1-22, and component drawings of the principal fuel-handling ~ equipment are shown in Figures 9.1-3 through 9.1-11. With the reactor shut down, the containment isolation between the fuel and reactor building can be opened. At this time, all channeled new bundles may be transferred to the containment pool where the refueling platform places them into containment pool-storage racks. If there are no channeled new fuel bundles, an unchanneled bundle is placed in the fuel preparation machine to await an irradiated spent fuel channel. When the reactor is suffi-ciently cooled, the gates separating the upper containment pool from the reactor well are closed (Figure 9.1-22) . A pipe connection [} at the bottom of the reactor well allows the reactor well water to drain by gravity to the main condenser. After the reactor well has been drained, the drywell head and vessel head are removed by the polar crane and placed in their respective storage areas (Fig-ure 9.1-22). The polar crane and cruciform-shaped strongback will be used to handle the 121-ton load of RPV head and attachments. The strongback is designed so that no single component failure will cause the load to drop or swing uncontrollably out of an essentially horizontal attitude. I The strongback attaches to the crane sister hook by means of an integral hook box and two hook pins. Each pin is capable of carrying the rated load. Each main beam of the crucifurm is capable of carrying the rated load. t
- On both ends of each leg are adjustable lifting rods, suspended vertically to attach the lifting legs to the RPV head. These are
[} 9.1-49 I
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.10.2 Refueling Procedure (Continued) for adjustment for even four-point load distribution and allow for some flexibility in diametrical location of the lifting lugs on the head. The maximum potential drop height is at the point where the head gets lifted vertically from the vessel and before moving it hori-zontally to the head storage pedestals. The elevation difference from vessel flange to storage elevation is approximately 30 ft. The shroud head load of 68 tons and the steam 3ryer load of 51 tons will both be lifted with the dryer / separator strongback. This strongback is a cruciform shape with box-shaped sockets at the four ends. Each socket box has two compartments to accommodate the two different lug spacings on the dryer and on the shroud head. Pneumatically operated lifting pins will penetrate the sockets to engage the lifting lugs. Each of the above strongbacks are load tested at 125% rated load. At this test, measurements are taken before test load, under test load and after releasing load, to verify that deflections are within acceptable limits. A magnetic particle test of structural welds is performed after the load test to assure structural integrity. A seal exists around the vessel opening to seal the drywell from the upper containment pool. In the meantime, additional water is pumped into the upper containment pool. Once the upper contain- . ment pool is filled, the dryer and separator are removed and transferred to their storage areas within the upper containment pool using the dryer / separator strongback. The tools used in these and subsequent reactor servicing operations are listed in Table 9.1-1. Once access to the core is possible, the refueling 9.1-50
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O m Iv) 9 1.4.2.10.2 Refueling Procedure (Continued) platform can relocate assemblies to and from the containment pool storage racks. Simultaneously, CRD hydraulic system and the neu-tron monitoring system may be serviced from beneath the vessel. During refueling, the refueling platform transfers the spent fuel from the core to the containment pool upender, and fuel transfer is made to the Fuel Building upender. The fuel-handling platform places the spent fuel assembly in the fuel preparation machine, where its channel is removed and fitted to the new fuel bundle previously placed in the machine. During channeling, the spent fuel bundle is placed in the storage racks by the fuel-handling platform. The platform then transfers the new fuel assembly to the upender. The fuel-handling platform then places another new fuel bundle in the fuel preparation machine for channeling. The critical path fuel-handling operations in the fuel-handling area
) will be performed by the fuel equipment. Therefore, the operation of the fuel-handling platform will be administratively coordinated with the operations of the refueling platform and transfer system to assure a safe, continuous fuel-handling process.
When refueling and servicing are completed, the steam separator assembly is replaced in the vessel, the steamline plugs removed and the steam dryer returned to the vessel. At this point, the gates are installed, isolating the reactor well from the upper pools. The reactor well is then drained to the main condenser. With the reactor well empty, the vessel and drywell heads are replaced. The reactor well is then filled. When all transfers from the upper containment pool to the fuel building have been completed, the containment is isolated by installing the blank flange in the transfer tube, and start-up
-, operations can begin.
V 9.1-51
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.10.2.1 New Puel Preparation 9.1.4.2.10.2.1.1 Receipt and Inspection of New Puel The incoming new fuel will be delivered to a receiving station within the fuel building. The crates are unloaded from the trans-port vehicle and examined for damage during shipment. The crate dimensions are approximately 32 x 32 x 216 inches. Each crate contains two fuel bundles supported by an inner metal container. Shipping weight of each unit is approximately 3000 lb. The receiving station shall include a separate area where the crate cover and the inner metal container can be removed from the crate. Both inner and outer shipping containers are reusable. IIandling during uncrating is accomplished by use of the fuel building crane or the separate receiving room crane. The inner container is tilted to a position which is almost vertical, while the fuel bundles are unstrapped and removed from the container with the fuel building crane. They are then transported to storage in the new fuel storage racks located in the new fuel storage vault or to the new fuel inspection stand located in the receiving area. The actual inspection of the new fuel is normally deferred until all the reusable containers are emptied and the area around the new fuel vault cleared. At that time, the individual fuel bundles are removed from the vault, inserted in the new fuel inspection stand, dimensionally and visually inspected, and returned to the storage vault to await assembly with channels. The new fuel inspection stand accommodates two fuel assemblies at one time. 9.1.4.2.10.2.1.2 Channeling New Puel New fuel is unloaded from the new fuel vault and transported to the fuel racks in the fuel pool. Usually, channeling new fuel is done concurrently with dechanneling spent fuel. Two fuel prepara-tion machines are located in the fuel pool, one used for 9.1-52
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ( 9.1.4.2.10.2.1.2 Channeling New Fuel (Continued) dechanneling spent fuel and the other to channel new fuel. The procedure-is as follows: Using the fuel-handling platform in the fuel building, a spent fuel bundle is transported to the fuel prep machine. The channel is untolted from the bundle using the channel bolt wrench. The channel-handling tool is fastened to the top of the channel and the fuel prep machine carriage is lowered removing the fuel from the channel. The channel is then positioned over a new fuel bundle located in fuel prep machine No. 2 and the process reversed. The channeled new fuel is stored in the pool storage racks ready for insertion into the reactor. 9.1.4.2.10.2.1.3 Equipment Preparation Another ingredient in a successful refueling outage is equipment and new fuel readiness. Equipment long lying dormant must be brought to life. All tools, grapples, slings, strongbacks, stud (} tensioners, etc., will be given a thorough inspection and opera-tional check, and any defective (or well worn) parts will be replaced. Air hoses on grapples will be checked. Crane cables will be routinely inspected. All necessary maintenance will be performed to preclude outage extension due to equipment failure. 9.1.4.2.10.2.2 Reactor Shutdown The reactor is shut down according to a prescribed procedure
, detailed in Chapter 13. During cooldown, the reactor pressure vessel is vented and filled to above flange level to promote cooling. The drywell head cavity is deflooded during this time in preparation for drywell and vessel head removal.
9.1.4.2.10.2.2.1 Drywell llead Removal () Immediately after cooldown the dryer pool gate and the separator pool gate are installed and the reactor well is deflooded. The work to 9.1-53
GESSAR II' 22R7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.10.2.2.1 Drywell Head Removal (Continued) remove the drywell head can begin. The drywell head will be attached by a quick disconnect mechanism. To remove the head, the quick disconnect pins are withdrawn and stored separately for reinsertion when the head is replaced. The drywell head is lifted by the overhead building crane to its appointed storage space on the refueling floor. The drywell seal surface protector is installed before any other activity proceeds in the reactor well area. 9.1.4.2.10.2.2.2 Reactor Well Servicing When the drywell head has been removed, several pipo lines are exposed. These lines penetrate thc reactor well through two open-ings. The piping must be removed and the openings sealed. There are also six vent openings which must be made watertight. Water level in the vessel is now lowered to flange level in preparation for head removal. 9.1.4.2.10.2.3 Reactor Vessel Opening 9.1.4.2.10.2.3.1 Vessel Head Removal The combination head strongback and carousel stud tensioner is transported by the Reactor Building crane and positioned on the reactor vessel head. Each stud is tensioned and its nut loosened in a series of two to ' three passes. Finally, when the nuts are loose, they are backed off using a nut runner until only a few threads engage. The nut is hand rotated free from the stud and the nuts and washers are placed in Ja racks provided for them on the carousel. When all the nuts and washers are removed, the vessel stud protectors and vessel head guide caps are installed. 9.1-54
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
)/
9.1.4.2.10.2.3.1 Vessel Head Removal (Continued) Next, the head, strongback and carousel are transported by the Reactor Building crane to the head holding pedestals on the refuel-ing floor. The head holding pedestals keep the vessel head ele-vated to facilitate inspection and 0-ring replacement. The six studs in line with the fuel transfer canal are removed from the vessel and placed in the rack provided for them. The loaded rack is transported to the refueling floor for storage. Removal of these studs provides a path for fuel movement. 9.1.4.2.10.2.3.2 Dryer Removal The reactor well is filled with water from the main condenser and the upper pool gates removed and stowed. The dryer-separator strongback is lowered by the reactor building crane and attached [J to the dryer lifting lugs. The dryer is lifted from the reactor vessel and transported underwater to its storage location in the dryer storage pool adjacent to the reactor well. 9.1.4.2.10.2.3.3 Separator Removal In preparation for the separator removal, the steamline plugs are installed in the four main steam nozzles. The separator is then unbolted from the shroud using the four shroud head bolt wrenches furnished for this purpose. When the unbolting is accomplished, the separator is lowered into the vessel and attached to the separator lifting lugs. The separator is lifted from the reactor vessel and transported underwater to the storage location in the separator pool adjacent to the reactor well. 9.1.4.2.10.2.3.4 Fuel Bundle Sampling During reactor operation, the core offgas radiation level is () monitored. If a rise in offgas activity has been noted, the 9.1-55
1 GESSAR 11 22A7007 238 MUCLEAR I S LAND Rev. 0 9.1.4.2.10.2.3.4 Fuel Bundle Sampling (Continued) reactor core may be sampled during shutdown to locate any leaking fuel assemblies. The fuel sampler isolates up to a 16-bundle array in the core. This stops water circulation through the bundles and allows fission products to concentrate if a bundle is defective. After 10 minutes, i water sample is taken for fission product analysis. If a defective bundle is found, it is transferred to the Fuel Building storage pool and stored in a special defective fuel storage container to minimize background activity in the storage pool. 9.1.4.2.10.2.4 Refueling and Reactor Servicing The gate isolating the containment pool from the reactor well is removed, thereby interconnecting the containment pool, the reactor well and the fuel transfer area. The refueling of the reactor can now begin. 9.1.4.2.10.2.4.1 Refueling During an annual equilibrium outage, approximately 25% of the fuel is removed from the reactor vessel, 25% of the fuel is shuffled in the core (generally f rom peripheral to center locations) and l 25% new fuel is installed. The actual fuel handling is done with the refueling platform. It is used as the principal means of l transporting fuel assemblies between the reactor well and the con-tainment pool; it also serves as a hoist and transport device. It provides an operator with work surface for almost all the other servicing operations. The platform travels on track extending along each side of the reactor well and pool and supports the refueling grapple and auxiliary hoists. The platform design per-mits travel over safety railings placed around the poolm. The l l grapple is suspended from a trolley that can traverse the width of the platform. Platform movement is controlled from an operator station on the trolley. Railing should be provided to keep all 9.1-56 1
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.10.2.4.1 Refueling (Continued) b(~N unauthorized personnel from entering the platform area or the inside of the refueling platform track area. The refueling platform has two 1/2-ton auxiliary hoists. One auxiliary hoist is mounted on the reactor side of the refueling platform and projects approximately 2 ft from the frame. This hoist normally can be used with appropriate grapples to handle control rods, guide tubes, fuel support pieces, sources and other internals of the core. The auxiliary hoist can also serve as a means of handling other equipment within the pool. The second auxiliary hoist is mounted on the platform trolley. The platform control system permits variable-speed, simultaneous operation of all three platform motions. Maximum speeds are: (1) bridge - 50 fpm U(~N (2) trolley - 30 fpm (3) grapple hoist - 40 fpm A single operator can control all the motipns of the platform required to handle the fuel assemblies during refueling. Inter-locks on both the grapple hoist and auxiliary hoists prevent hoist-ing of a fuel assembly over the core with a control rod withdrawn; interlocks also prevent withdrawal of a blade with a fuel assembly over the core attached to either the fuel grapple or auxiliary hoists. Interlocks block travel over the reactor in the startup mode. The refueling platform contains a system that indicates position of the fuel grapple over the core. The readout, in the operator's cab, matches the core arrangement cell identification numbers. The
/3 f i position indicator is accurate within 1/4 inches, relative to G
9.1-57
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0 9.1.4.2.10.2.4.1 Refueling (Continued) actual position, and minimizes jogging required to correctly place the grapple over the core. To move fuel, the fuel grapple is aligned over the fuel assembly, lowered and attached to the fuel bundle bail. The fuel bundle is raised out of the core, moved through the refueling slot to the containment pool, positioned over the storage rack and lowered into the rack. Fuel is shuffled and new fuel is moved from the containment pool to the reactor vessel in the same manner. 9.1.4.2.10.2.5 Vessel Closure The following steps, when performed, will return the reactor to operating condition. The procedures are the reverse of those described in the preceding sections. Many steps are performed in parallel and not as listed. (1) Core verification - the core position of each fuel assembly must be verified to assure the desired core configuration has been attained. Underwater TV with a video tape is utilized. (2) Control rod drive tests - the control rod drive timing, friction and scram tests are performed as required. (3) Replace separator. (4) Bolt separator and remove four steamline plugs. (S) Replace steam dryer. (6) Install pool gates. (7) Drain reactor well. 9.1-58
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O w 9.1.4.2.10.2.5 Vessel Closure (Continued)
/ )
(8) Remove drywell seal surface covering; open drywell vents. (9) Replace vessel studs. (10) Install reactor vessel head. (11) Install vessel head piping and insulation. (12) Hydro test vessel if required. (13) Install drywell head; leak check. (14) Flood reactor well. (15) Stow pool gates. (N \- (16) Startup tests - the reactor is returned to full power operation. Power is increased gradually in a series of steps until the reactor is operating at rated power. At specific steps during the approach to power, the in-core flux monitors are calibrated. 9.1.4.2.10.3 Departure of Fuel From Site The empty cask arrives at the plant on the special flatbed railcar. The personnel shipping barrier and transfer impact structure are removed from the larger casks and stored outside the rail entry door. Health physics personnel check the cask exterior to deter-mine if decontamination is necessary. Decontamination, if required, and washdown to remove road dirt, is performed before removal of the cask from the transport vehicle. The fuel facility door is opened and the railcar and cask moved into the building. The rail-car is blocked in position and the engine moved out of the building. Cs]'I 9.1-59
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.2.10.3 Departure of Fuel From Site (Continued) The rail entry door is closed and the cask and railcar are inspected for shipping damage. The cask Cooling System of the transport vehicle is disconnected. The cask yoke is removed from its storage position on the flat-bed and attached to the cask trunnions. The yoke engagement, car brakes and wheel blocks and clearances for cask tilt and lift are checked. The cask is tilted to the vertical position with combined main hoist lift and trolley movement With the cask in a vertical position, the cask is lifted approxi..2.cly 5 ft off the railcar skid mounting trunnions to clear the upper coolant duct. The cask is moved horizontally into the cask decontamination vault and slowly lowered to the floor of the vault. Closure head lifting cables on the yoke are attached to the head and secured and the closure nuts are disengaged. The cask is next raised approximately 13 ft for transfer into the cask pool. In this position, the base of the cask is approximately 13 ft below the fuel building operations floor (f13or elevation +11'-0") with approximately 1-ft clearance above the floor of the canal between the cask decontamination vault and the cask pool. The cask is moved through the canal to a position over the center of the cask pool and slowly lowered into the cask pool until it rests on the cask pool floor (floor elevation -32'-0"). The cask and yoke displacement volume raise the cask pool level to approxi-mately 1 ft below the floor of the canal to the decontamination vault. The cask lifting yoke is lowered until disengaged from the cask trunnions and the closure head lifted off the cask. Closure head and yoke are moved into the cask decontamination vault for storage. The gates between the cask decontamination vault are replaced and the cask pool level raised to the level of the fuel storage pool. With the pool levels equalized, the canal gates between the cask 9.1-60
- _ . . - _ - - - - . - _ - _ - - . - = . . - - - . . . . . .--
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 s 9.1.4.2.10.3 Departure of Fuel From Site (Continued) pool and the fuel storage pool are removed and spent fuel transfer j from the storage racks to the cask is started. 1 Spent fuel is transferred underwater from storage in the fuel pool l to the cask using the telescoping fuel grapple mounted on the fuel- [ handling platform. When the cask is filled with spent fuel, the i
- pool canal gate between the cask storage pool and the fuel storage i
is replaced. The cask pool level is below the canal to the cask decontamination vault and the gate is removed. The closure head is replaced on the cask and the lift yoke engaged with the cask trunnions. The loaded i cask is raised to the transfer position (base of cask approximately 13 ft below the operations floor), transferred to the cask decon-tamination vault, and slowly lowered approximately 13 ft to the vault floor. 3
- The cask is checked by health physics personnel and decontamination performed in the decontamination vault with high pressure water sprays, chemicals and hand scrubbing as required to clean the cask to the level required for transport. Cooling connections are available in the decontamination vault in the event cooling is j required during decontamination activities. The remaining closure l nuts are replaced and tightened. Smear tests are performed to verify cleaning to offsite transportation requirements.
The cleaned cask is replaced on the cask skid mountings with the l cask crane. The cask cooling system of the transport vehicle is connected to the cask and the cask internal pressure and tempera-ture are monitored. When they are at equilibrium conditions, the cask is ready for shipment to a fuel processing plant. The personnel barrier and impact structure are replaced (on the larger i casks, this is done after the railcar is removed from the building). O The Fuel Building rail facility doors are opened and the railcar
] moved out of the Fuel Building.
9.1-61 i 5
,-,,_,m.--.. ._ _ _ - , . _ - , _ _ - , _ , - , _ - - - , . - - _ , _ , , .., _ _ . . _ , . . . _ , . - - - . . . , - - . _ . , , _ _ _ , _ , , -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.3 Safety Evaluation of Fuel-Handling System Safety aspects (evaluation) of the fuel servicing equipment are discussed in Subsection 9.1.4.2.3 and safety aspects of the refueling equipment are discussed throughout Subsection 9.1.4.2.7. A description of fuel transfer, including appropriate safety features, is provided in Subsection 9.1.4.2.10. In addition, the f,llowing summary safety evaluation of the Fuel-Handling System is provided below. The fuel prep machine removes and installs channels with all parts remaining underwater, Mechanical stops prevent the carriage from lifting the fuel bundle or assembly to a height where water shielding is less than 8 ft. Irradiated channels, as well as small parts such as bolts and springs, are stored underwater. The spaces in the channel storage rack have center posts which prevent the loading of fuel bundles into this rack. There are no nuclear safety problems associated with the handling O of new fuel bundles, singly or in pairs. Equipment and procedures prevent an accumulation of more than two bundles in any location. The refueling platform is designed to prevent it from toppling into the pools during a SSE. Redundant safety interlocks, as well as limit switches, are provided to prevent accidentally running the grapple into the pool walls. The grapple utilized for fuel movement is on the end of a telescoping mast. At full retraction of the mast, the grapple is 8 ft below water surface, so there is no chance of raising a fuel assembly to the point where it is inadequately shielded by water. The grapple is hoisted by redundant cables inside of the mast, and is lowered by gravity. A digital readout is displayed to the operator, showing him the exact coordinates of the grapple over the core. The mast is suspended and gimbaled from the trolley, near its top, so that the mast can be swung about the axis of platform travel, 9.1-62
GESSAR II .22A7007 238 NUCLEAR ISLAND Rev. 0 () 9.1.4.3 Safety Evaluation of Fuel-Handling System (Continued) in order to remove the grapple from the water for servicing and 1_ for storage. l 1 The grapple has two independent hooks, each operated by an air cylinder. Engagement is indicated to the operator. Interlocks
; prevent grapple disengagement until a " slack cable" signal from l the lifting cables indicates that the fuel assembly is seated.
i The slack cable indication is also used to determine if a fuel I bundle is lodged in a position other than its normal, seated position in the core. 1 In addition to the main hoist on the trolley, there is an auxiliary hoist on the trolley and another hoist on its own monorail. These three hoists are precluded froca operating simultaneously because control power is available to only one of them at a time. The two auxiliary hoists have load cells with interlocks which prevent the hoists from moving anything as heavy as a fuel bundle. The two auxiliary hoists have electrical interlocks which prevent i the lifting of their loads higher than 8 ft underwater. Adjustable , mechanical jam-stops on the cables back up these interlocks. l In summary, the fuel-handling system complies with General Design Criteria 2, 3, 4, 5, 61, 62 and 63, and applicable portions of ! 10CPR50. i I j A system-level, qualitative-type failure mode and effects analysis relative to this system is discussed in Subsection 15A.6.5. i i The safety evaluation of the new and spent fuel storage is pre- ! sented in Subsections 9.1.1.3 and 9.1.2.3. i 9.1-63
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.3 Safety Evaluation of Fuel-Handling System (Continued) Regulatory Guide 1.13 This regulatory guide is applicable to the refueling platform within the General Electric scope of supply for this plant. For commitment and revision number, see regulatory guide commitment matrix in Section 1.8. The refueling platform is designed to prevent toppling into the pool during an SSE. Redundant safety interlocks and limit switches are provided to prevent accidentally running the fuel grapple into the pool walls. 9.1.4.4 Inspection and Testing Requirements 9.1.4.4.1 Inspection Refueling and servicing equipment is subject to the strict controls of quaiity assurance, incorporating the requirements of federal regulation 10CFR50, Appendix B. Components defined as essential to safety (e.g., the fuel storage racks, refueling platform, and fuel transfer syst.m containment isolation assembly) have an additional set of engineering specified " quality requirements" that identify safety-related features which require specific QA verifi-cation of compliance to drawing requirements. For components classifierl as American Society of Mechanical Engineers (ASME) Section III, the shop operation must secure and maintain an ASME "N" stamp, which requires the submittal of an acceptable ASME quality plan and a corresponding procedural manual. Additionally, the shop operation must submit to frequent ASME audits and component inspections by resident state code inspectors. 9.1-64
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O J
, 9.1.4.4.1 Inspection (Continued)
Prior to shipment, every compontat inspection item is reviewed by QA supervisory personnel and combined into a summary product ! quality checklist (PQL). By issuance of the PQL, verification is made that all quality requirements have been confirmed and are
- on record in the product's historical file.
I 9.1.4.4.2 Testing + 1 Qualification testing is performed on refueling and servicing ! equipment prior to multi-unit production. Test specifications
- are defined by the responsible design engineer and may include sequence of operations, load capacity and life cycles tests.
l These test activities are performed by an independent test. engineer-I ing group and, in many cases, a full design review of the product is conducted before and after the qualification testing cycle. Any design changes affecting function, that are made after the 4 N completion of qualification testing, are requalified by test or calculation. i Functionel tests are performed in the shop prior to the shipment of production units and generally include electrical tests, leak j tests and sequence of operations tests. When the unit is received at the site, it is inspected to ensure no damage has occurred during transit or storage. Prior to use and at periodic intervals, each piece of equipment is again tested to ensure the electrical and/or mechanical functions are operational. . Passive units, such as the fuel storage racks, are visually l inspected prior to use. 1 j Fuel-handling and vessel servicing equipment preoperational tests 1 are described in Subsection 14.2.12.1.12. 4 I 9.1-65
- _ _ _ _ . _ . . _ _ . , , _ . . _ . . . . _ . _ , . , _ _ . . _ _ . . - , . . _ . _ _ . , . _ _ _ _ , _ . _ ._._.____,_m.._,_. . . . . . _ . . _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.1.4.5 Instrumentation Requirements The majority of the refueling and servicing equipment is manually operated and controlled by the operator's visual observations. This type of operation does not necessitate the need for a dynamic instrumentation system. Ilowever, there are several components, that are essential to pru-dont operation, that do have instrumentation and control systems. 9.1.4.5.1 Refueling Platform The refueling platform has a nonsafety-related X-Y-Z position indi-cator system that informs the operator which core fuel cell the fuel grapple is accessing. Interlocks and control room monitor are provided to prevent the fuel grapple from operating in a fuel cell where the control rod is not in the proper orientation for refueling. (See Subsection 7. 6.1. 3 for discussion of refueling interlocks.) Additionally, there is a series of mechanically activated switches and relays that provides monitor indications on the operator's console for grapple limits, hoist and cable load conditions, and confirmation that the grapple's hook is either engaged or released. A series of load cells is installed to provide automatic shutdown whenever threshold limits are exceeded on either the fuel grapple or the auxiliary hoist units. 9.1.4.5.2 Fuel Support Grapple Although the fuel support grapple is not essential to safety, it has an instrumentation system consisting of mechanical switches O 9.1-66
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 9.1.4.5.2 Fuel Support Grapple (Continued) and indicator lights. This system provides the operator with a positive indication that the grapple is properly aligned and oriented and that the grappling mechanism is either extended or retracted. 9.1.4.5.3 Inclined Fuel Transfer Tube The instrumentation sensors for this system provide the inputs to a programmable controller that automatically sequences the opening and closing of valves, the inclination and vertical upending of the fuel carriage, water levels, and the carriage traversing speeds. The microprocessor control and proximity-type sensors also pro-vide monitor and status conditions of the fuel transfer operation () on each of the two operator's consoles, one. located in the Fuel Building and the other on the RPV refueling floor. Monitor indicators and interlocks are provided in the reactor control room to indicate whenever personnel have accessed radiation hazardous areas along the transfer tube's route. 9.1.4.5.4 Other Refer to Table 9.1-1 for additional refueling and servicing equip-ment not requiring instrumentation. 9.1.4.5.5 Radiation Monitoring The radiation monitoring equipment for the refueling and servicing equipment is evaluated in Subsection 7.6.1. O 9.1-67/9.1-68
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O t () TOOLS AND SERVICING EQUIPMENT Table 9.1-1 Fuel Servicing Equipment In-Vessel Servicing Equipment , Channel Handling Boom Instrument Strongback Fuel Preparation Machines Control Rod Grapple New Fuel Inspection Stand Control Rod Guide Tube Grapple Channel Bolt Wrenches Fuel Support Grapple Channel Handling Tool Grid Guide Fuel Pool Sipper Control Rod Latch Tool Jib Crane Instrument Handling Tool General Purpose Grapples Control Rod Guide Tube Seal-Fuel-Handling Platform In-Core Guide Tube Seals Fuel Transfer System Blade Guides Fuel Bundle Sampler Peripheral Orifice Grapple Servicing Aids Orifice Holder Peripheral Fuel Support Plug Pool Tool Accessories Actuating Poles General Area Underwater Lights Refueling Equipment Local Area Underwater Lights Drop Lights Refueling Platform Underwater TV Monitoring System Auxiliary Platform O Underwater Vacuum Cleaner Viewing Aids Storage Equipment Light Support Brackets Underwater Viewing Tube Fuel Storage Racks Channel Storage Racks Reactor Vessel Servicing Equipment Defective Fuel Storage Containers Reactor Vessel Servicing Tools In-Vessel Racks Steam Line Plugs and Installation Equipment Storage Racks Tools Shroud Head Bolt Wrenches Head Holding Pedestals Head Stud Rack Under-Reactor Vessel Servicing , Dryer-Separator Strongback Equipment ! Head Strongback/ Carousel ! (including Stud Tensioners) Control Rod Drive Servicing , Tools ! CRD Hydraulic System Tools ! Water Seal Cap ! Control Rod Drive Handling l Equipment Equipment Handling Platform i Thermal Sleeve Inntallation Tool In-Core Flange Seal Test Plug i () Key Bender , e i ! 9.1-69 I _ , . _ . _ - . - . _ . - . _ . , ._ . - - ~ , - . . _ _ _ _ . , . - , _ - , , _ _ . - - _ - _ . , . _ _ . _ _ . _ _ . , - - _ . _ , _ . _ , . . - , _ . _ - - . . - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 9.1-2 ' FUEL SERVICING EQUIPMENT Essential Safety Component Classifi- Classifi- Quality Seismic No. Identification cation cation Group Category (a) (b) (c) (d) 1 Fuel Prep Machine NE O E NA 2 New Fuel Inspection PE 2 E I Stand 3 Channel Bolt Wrench NE O E NA 4 Channel-llandling Tool NE O E NA 5 Fuel Pool Sipper NE 0 E NA 6 Gcaeral-Purpose NE 0 E NA Grapple 7 Jib Crane PE 2 E I 8 Fuel-Ilandling Platform PE 2 E I 9 Channel-Handling Boom NE O E NA O Notes: (a) NE = Nonescential PE = Passive Essential (b) 0 = Other (c) B = ASME Code Section III Class 2 D = ANSI B31.1 E = Industrial Code Applies I = Electrical Codes Apply (d) NA = No Seismic Requirements I = Class I O 9.1-70
GESSAR II 22A7007 238 NUCLEAR ISLAND Rov. O O Table 9.1-2a FUEL TRANSFER SYSTEM COMPONENTS Essential Safety , Component Classifi- Classifi- Quality Seismic l No. Identification cation cation Group Category (a) (b) (c) (d) i 1 Winch NE O E NA 2 liydraulic Power Supply NE O E NA 1 3 Fluid Stop NE O E NA
] ) 4 Vent Pipe NE O D NA 5 Cable Enclosures NE O D NA 6 Top IIorizontal Guide NE 0 E NA Arms 7 Upper Pool Upender NE O E NA i 8 Trunnion Box NE O D NA 9 liydraulic Cylinder NE O E NA 10 Upper Pool Framing NE O E NA 11 Sheave Box Cover NE O D NA
} 12 Ilydraulic Cylinder NE O E NA 13 Fill Valve NE O D NA 14 Sheave Box NE 0 D NA 15 Sheave Pipe NE O D I 16 Ilydraulic Cylinder NE O E NA l 17 Manual Gate Valve NE O D I I
- 18 Containment Isolation PE 2 B I 19 Containment Bellows PE 2 B I 20 Transfer Tube NE O D I 21 Ilydraulic Power Supply NE O E NA 9.1-71 I
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Tc.ble 9.1-2a FUEL TRANSFER SYSTEM COMPONENTS (Continued) Essential Safety Component Classifi- Classifi- Quality Seismic No. Identification cation cation Group Category (a) (b) (c) (d) 22 Mid-Support NE O D I 23 Wire Rope (Cables) NE O E NA 24 Carriage NE O E NA 24A Tilt Tube NE 0 E NA 24B Follower NE 0 E NA 25 Gate Valve NE O D I 26 Bellows NE 0 D NA 27 Drain Valve NE 0 D NA 28 IIorizontal Guide Arms NE 0 E NA 29 Valve Support Structure NE O D I 30 Lower Pool Framing NE 0 E NA 31 Lower Pool Upender NE 0 E NA 32 Pivot Arm Framing NE O E NA Control System Notes: (a) NE = Nonessential PE = Passive Essential l (b) 0 = Other (c) B = ASME Code Section III Class 2 D = ANSI B41.1 E = Industrial Code Applies I = Electrical Codes Apply (d) NA = No Seismic Requirements I = Class I i 9.1-72
GESSAR II 22A7007 238 NUCLEAR ISLAND Rnv. 0 [ ~
) Table 9.1-3 REACTOR VESSEL SERVICE EQUIPMENT Essential Safety Component Classifi- Classifi Quality Seismic No. Identification cation cation Group Category (a) (b) (c) (d) 1 Reactor Vessel Service NE 0 E NA Tools 2 Steamline Plug NE O E NA 3 Shroud Head Bolt Wrench NE 0 E NA 4 Head Holding Pedestal PE 0 E I 7 Head Stud Rack NE O E NA 6 Dryer and Separator PE O E NA*
Strongback 7 Head Strongback Carousel NE O E NA A Notes: J (a) NE = Nonessantial , PE = Passive Essential (b) 0 = Other (c) B = ASME Code Section III Class 2 D = ANSI B31.1 E = Industrial Code Applies I = Electr' cal Codes Apply (d) NA = No Seismic Requirements
- Dynamic analysis methods for seismic loading are not applicable, as this equipment is supported by the reactor service crane.
Lifting devices have been designed with a minimum safety factor of 5 and undergo proof testing. (m
.~s 9.1-73
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 9.1-4 UNDER-REACTOR VESSEL SERVICING EQUIPMENT AND TOOLS Safety Seismic Equipment / Tool Claraification Class Category
- 1. CRD Handling Ncnessential "Other" NA Equipment
- 2. Equipment Handling Nonessential "Other" NA Platfo rm
- 3. Water Seal Cap Nonessential "Other" NA
- 4. Thermal Sleeve Nonessential "Other" NA Removal Tool
- 5. In-Core Flange Nonessential "Other" NA Seal Test Plug
- 6. Key Bender Nonessential "Other" NA Notes:
NA = No Seismic' Requirements j lll f 9.1-74 l l
238 N L ISLAND O 4l > c.
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GESSAR II 22A7007 238 IJUCLEAR I SIJsN D Rev. O O E B B MM E B B E a rd a am
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GESSAR 11 ^ 238 NUCLEAR ISLAND Rev. 0
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ()~^, \ l h FOR BOLT UNCOUPL NG
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O A 1 2 REACTOR BUILDING , I..
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O REF UE LING O P LAT F OR M r-- g -
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 0 270 g s .~ l l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O gg o
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O 1800 v 9 -
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GESSAR II 22h7007 238 NUCLEAR ISLAND Rev. 0
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 SECTION 9.2 3 CONTENTS Section Title Page 9.2 WATER SYSTEMS 9.2-1 9.2.1 Essential Service Water (ESW) 9.2-1 System 9.2.1.1 Design Bases 9.2-1 9.2.1.1.1 Safety Design Bases 9.2-1 9.2.1.1.2 Power Generation Design Bases 9.2-2 9.2.1.2 System Description 9.2-2 9.2.1.3 Safety Evaluation 9.2-4 9.2.1.3.1 Failure Analysis 9.2-4 9.2.1.3.2 Safety Evaluation of Equipment 9.2-5 9.2.1.4 Testing and Inspection Requirements 9.2-8 9.2.1.5 Instrumentation and Control Requirements 9.2-9 p- 9.2.2 Closed Cooling Water System 9.2-10 ( ) 9.2.2.1 Design Bases 9.2-10 9.2.2.1.1 Power Generation Design Bases 9.2-10 9.2.2.2 System Description 9.2-11 9.2.2.3 Safety Evaluation 9.2-12 9.2.2.4 Tests and Inspections 9.2-13 9.2.2.5 Instrumentation Application 9.2-13 9.2.3 Demineralized Water Makeup System 9.2-14 9.2.3.1 Design Bases 9.2-14 9.2.3.2 System Description 9.2-16 9.2.3.3 Safety Evaluation 9.2-17 9.2.3.4 Tests and Inspections 9.2-17 9.2.3.5 Instrumentation Application 9.2-18 9.2.4 Potable and Sanitary Water Systems 9.2-18 9.2.4.1 Design Bases 9.2-18 9.2.4.2 System Description 9.2-19 9.2.4.3 Safety Evaluation 9.2-20 9.2.4.4 Tests and Inspections 9.2-20 g j 9.2.4.5 Instrumentation Application 9.2-21 9.2-i t
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued) O Section Title Page 9.2.5 Ultimate Heat Sink 9.2-21 9.2.6 Condensate Storage Facilities and Distribution System 9.2-21 9.2.6.1 Design Bases 9.2-22 9.2.6.1.1 Safety Design Bases 9.2-22 9.2.6.1.2 Power Generation Design Bases 9.2-22 9.2.6.2 System Description 9.2-22 9.2.6.3 Safety Evaluation 9.2-25 9.2.6.4 Tests and Inspections 9.2-25 9.2.6.5 Instrumentation Application 9.2-26 9.2.7 Plant Chilled Water Systems 9.2-27 9.2.7.1 Design Bases 9.2-27 9.2.7.1.1 Power Generation Design Bases 9.2-27 9.2.7.1..'.1 Drywell Chilled Water System 9.2-27 9.2,7.1.1.2 Control Building Chilled Water System 9.2-27 9.2.7.1.1.3 Reactor Island Chilled Water System 9.2-27 9.2.7.1.2 Safety Design Bases 9.2-28 9.2.7.2 System Description 9.2-28 9.2.7.2.1 Drywell Chilled Water System 9.2-28 9.2.7.2.2 Control Building Chilled Water System 9.2-30 9.2.7.2.3 Reactor Island Chilled Water System 9.2-31 9.2.7.3 Safety Evaluation 9.2-31 9.2.7.3.1 Drywell Chilled Water System 9.2-31 9.2.7.3.2 Control Building Chilled Water System 9.2-32 9.2.7.3.3 Reactor Island Chilled Water System 9.2-33 9.2.7.4 Tests and Inspections 9.2-34 9.2.7.4.1 Drywell 9.2-34 9.2.7.4.2 Control Building 9.2-34 9.2.7.4.3 Reactor Island 9.2-35 9.2.7.5 Instrumentation Application 9.2-35 9.2.7.5.1 Drywell Chilled Water System 9.2-35 9.2-ii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued) C) Section Title Page 9.2.7.5.1.1 Instrument Application 9.2-35 9.2.7.5.2 Control Building Chilled Water System 9.2-36 9.2.7.5.3 Reactor Island Chilled Water System 9.2-37 9.2.8 Heated Water Systems 9.2-38 9.2.8.1 Design Bases 9.2-38 9.2.8.2 System Description 9.2-38 9.2.8.3 Safety Evaluation 9.2-39 9.2.8.4 Tests and Inspections 9.2-40 9.2.8.5 Instrumentation Application 9.2-40 9.2.9 Nuclear Island / BOP Interfaces 9.2-41 9.2.9.1 Essential Service Water System 9.2-41 9.2.9.1.1 Design Criteria 9.2-42 9.2.9.1.2 Interfaces 9.2-42 9.2.9.2 Ultimate Heat Sink 9.2-43 %/ 9.2.9.3 Nuclear Island Condensate Distribution System - BOP Condensate Storage Facilities Interface 9.2-43 9.2.9.3.1 Design Criteria 9.2-43 9.2.9.3.2 Specific Interfaces 9.2-44 9.2-iii/9.2-iv
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O s s SECTION 9.2
)
C_/ TABLES Table Title Page 9.2-1 Essential Service Water 9.2-45 9.2-2 Division 3 HPCS Service Water System 9.2-57 9.2-3 Essential Service Water System Active Failure Analysis 9.2-58 9.2-4 Closed Cooling Water Systera Component 9.2-59 9.2-5 Demineralized Water Distribution System User Requirements 9.2-61 9.2-6 Condensate Distribution System Flow Requirements 9.2-65 9.2-7 Drywell Chilled Water System Component Description 9.2-69 9.2-8 Control Building Chilled Water System Component Description 9.2-70 9.2-9 Reactor Island Nonessential Chilled Water System Component Description 9.2-71 9.2-10 Heated Water System Heat Load and Flow Requirements 9.2-73
)
(D G 9.2-v/9.2-vi
GESSAR II 22A7707 238 NUCLEAR ISLAND Rev. 0 SECTION 9.2 ILLUSTRATIONS Figure Title Page 9.2-la Essential Service Water System (Div. 1) P&I Diagram 9.2-77 t 9.2-lb Essential Service Water System (Div. 2) P&I Diagram 9.2-78 9.2-2 HPCS Service Water System (Div. 3) P&I Diagram 9.2-79 3 9.2-3a Closed Cooling Water System P&I Diagram 9.2-81 9.2-3b Closed Cooling Water System P&I Diagram 9.2-82
! 9.2-4a Demineralized Water and Condensate Distribution Flow Diagram 9.2-83 9.2-4b Demineralized Water and Condensate Distribution Flow Diagram 9.2-84 9.2-4c Demineralized Water and Condensate Distribution Flow Diagram 9.2-85 9.2-5 Potable Water System 9.2-87 j
() 9.2-6 9.2-7a Sanitary Water System Drywell Chilled Water System P&I Diagram 9.2-89 9.2-91 9.2-7b Drywell Chilled Water System P&I Diagram 9.2-92 9.2-8a Control Building Chilled Water System P&I Diagram 9.2-93 9.2-8b Control Building Chilled Water System P&I Diagram 9.2-94 9.2-9a Reactor Island Nonessential Chilled Water System P&I Diagram 9.2-95 9.2-9b Reactor Island Nonessential Chilled Water System P&I Diagram 9.2-96 9.2-9c Reactor Island Nonessential Chilled Water System P&I Diagram 9.2-97 9.2-10a Heated Water Distribution System P&I Diagram 9.2-99 9.2-10b Heated Water Distribution System P&I Diagram 9.2-100 i O 1 9.2-vii/9.2-viii i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2 WATER SYSTEMS 9.2.1 Essential Service Water (ESW) System 9.2.1.1 Design Bases 9.2.1.1.1 Safety Design Bases (1) The Essential Service Water (ESW) System shall be - designed to remove heat from plant auxiliaries which are required for a safe reactor shutdown, as well as those auxiliaries whose operation is desired following a LOCA, but not essential to safe shutdown. (2) The safety-related portion of the ESW System shall provide a means of flooding the vessel, drywell and containment through the RHR System during the post-LOCA period. (UD (3) The ESW System shall be designed to perform its required cooling function following a LOCA, assuming a single active or passive failure. (4) The safety-related portions and valves isolating the nonsafety-related portions of the ESW system shall be designed to Seismic Category I and ASME Code , Sec-tion III, Class 3, Quality Assurance B, Quality Group C, IEEE-279 and IEEE-308 requirements. (5) The ESW System shall be designed to limit leakage to the environment of radioactive contamination that may enter the ESW from the RHR System. (6) Safety-related portions of the ESW System shall be pro-tected from flooding, spraying, steam impingement, pipe whip, jet forces, missiles, fire and the ef fect of (} 9.2-1 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Revo 0 9.2.1.1.1 Safety Design Bases (Continued) failure of any non-Seismic Category I equipment, as required. (7) The safety-related portion of the ESW System shall be designed to meet the foregoing design bases during an Lol> P . 9.2.1.1.2 Power Generation Design Bases ( 1) The ESW System shall be designed to cool various plant auxiliaries as required during: (a) normal operation; (b) emergency shutdown; (c) normal shutdown; and (d) testing. (2) Other requirements are to be described by Applicant. 9.2.1.2 System Description The ESW System distributes cooling water during various operating modes, during shutdown and during post-LOCA operations. The system removes heat from plant auxiliaries and transfers it to the ultimate hea t sink . Figures 9.2-la and b (K-121A and B) and 9.2-2 ( K- 12 2 ) show the piping and instrumentation diagrams. l l The ESW System serves the auxiliary equipment listed in l Table 9.2-1. The equipment is identified by equipment number, and its location in Figure 9.2-la or 9.2-lb is indicated by a coordi-Figure 9.2-la is classified as Safety Class 3, ! nate system. l Division 1. Figure 9.2-lb is classified as Safety Class 3, Division 2. l The ESW System is designed to perform its required cooling func-l tion following a postulated LOCA, assuming a single active failure in any mechani'al or electrical syste:a. In order to meet this 9.2-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.1.2 System Description (Continued) (} requirement, the ESW System provides two complete trains, which are mechanically and electrically separated. In case of a failure which disables either Division 1 or Division 2, the other division operating in conjunction with the HPCS ESW System meets plant safe shutdown requirements, including a LOCA or a loss of offsite power, or both. ESW Division 1 equipment is supplied from Division 1 ESF electric power and control, and ESW Division 2 from Division 2 ESF power and control. . The HPCS ESW System (Division 3) has been completely separated f rom Division 1 and 2. Division 3 provides cooling requirements for conditions which require IIPCS or HPCS diesel operation. The ' ilPCS service water pump is required to operate whenever the HPCS DG or IIPCS pump room cooler starts. These cooling requirements are met under postulated LOCA conditions, or during a loss of () offset power, or both concurrently. The IIPCS service water equipment is supplied with Division 3 electric power and con-trols. The equipment served by this system with its required cooling capacities is listed in Table 9.2-2. All equipment in Division 3 is Safety Class 3. During normal operation, service water flows through all the Division 1 and 2 equipment except the RHR heat exchangers. Division 3 may operate during any plant operating modes, such as testing of the Division 3 system. . Division 3 is not normally required except during a LOCA and LOPP. For LOPP only, the HPCS service water pump may be terminated af ter the first 6 hr. During all plant operating modes, Division 1 and 2 service water pumps are normally operating. There fore , if a LOCA occurs, the ESW systems required to shut down the plant safely are already in operation. R {v\ 9.2-3
GESSAR II 22R7007 238 NUCLEAR ISLAND Rev. 0 9.2.1.2 System Description (Continued) The nonsafety-related parts of the ESW System are not required for safe shutdown and, hence, are not safety systems. Isolation valves separate the ESW System f rom the nonsafety-related subsys-tem during a LOCA, in order to assure the integrity and safety functions of the safety-related parts of the system. Nonsafety-related parts of the ESW System should be operated during all other modes , including the emergency shutdown f ollowing an LOPP . Instrumentation is provided to detect significant leakage in the nonsafety-related subsystem. The water flow is measured in both en trance and exit pipes . Any significant leakage shows up as a dif ference between the two flow measurements . A dif ferential flow switch detects leakage and isolates this subsystem, thus assuring continued operability of the safety-related services. The Applicant is to provide description of the service water sys-tem outside of the Nuclear Island. The ESW flow rates establish interf ace flow requirements f rom the BOP. Description of the ESW pumps is included in Subsection 3.2.5 l (Ultimate lleat Sink) . This description is within the scope of the 1 l Applicant. l 9.2.1.3 Safety Evaluation 9.2.1.1.1 Failure Analysis A system failure analysis of passive and active components of the ESW System is presented in Table 9.2-3. Any of the assumed fail-ures of the ESU System are detected in the control room by varia-tions of and/or alarms from the various system instruments and also from the Leak Detection System sensing leakage in the ECCS pump and heat exchanger areas . 9.2-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (~N 9.2.1.3.2 Safety Evaluation of Equipment Equipment served by the ESW System is listed in Tables 9.2-1 and 9.2-2. The tables contain five operating modes: (1) normal operation; (2) initial shutdown (20 hr); (3) extended shutdown (plus 20 hr) ; (4) post-LOCA and LOPP; and (5) nonaccident LOPP. The flow rate, and heat loads are given for each equipment in each operating mode . In the event of a LOCA, the RI nonessential service water system, which is a subsystem of the ESW System, is isolated by proper isolation valves. The nonsafety-related portion of the system is (}
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automatically isolated in the event of a rupture in the nonsafety-related subsystem. The flow to and from the nonsafety-related subsystem is measured. A differential flow switch is activated by a significant leak, sending an isolation signal to close two valves. One valve on the supply line and one valve on the dis-charge line are used, with suitable power and controls from divi-sional sources to assure isolation in the event of any single active failure. Single isolation valves are used on the basis that an active failure of one isolation valve disables only that system of which it was a part. The ESW System is designed to withstand a single active failure without losing its capability to participate in the safe shutdown of the reactor following a LOCA or DBA. Table 9.2-3 gives the result of a system failure analysis of active and passive components. 9.2-5
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0 9.2.1.3.2 Safety Evaluation of Equipment (Continued) Redundant trains of the ESW System are separated and protected to the extent necessary to assure that sufficient equipment remain in operating to permit shutdown of the unit in the event of any of the following (Separation is applied to electrical equipment and instrumentation and controls as well as to mechanical equipment and piping.): (1) flooding, spraying or steam release due to pipe rupture or equipment failure; (2) pipe whip and jet forces resulting from postulated pipe rupture of nearby high energy pipes; (3) missiles which may result from equipment failure ; (4) fire; and g (5) failures of any non-Category I equipment (pertains to Seismic Category I equipment) . Liquid radiation monitors are provided to sample the discharge ESW from the RHR heat exchangers , both Division 1 and Division 2. Upon detection of radiation leakage in one of the systems, that system is isolated by operator action from the Control Room, and the total cooling load can be met by the redundant, 100% system. Cons eq ue ntly , radioactive contamination released by the ESW System to the environment does not exceed allowable limits defined by 10CFR100. The safety-related parts of the ESW System and the HPCS service water system are designed to Seismic Category I and ASME Code, Section III, Class 3, Quality Assurance B and Quality Group L requirements. The design also meets IEEE-279 and IEEE-308 9.2-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 /~'s 9.2.1.3.2 Safety Evaluation of Equipment (Continued) U requirements . Isolation valves for nonsafety-related service water systems also meet the above requirements. The Reactor. Island nonessential service water system, a subsystem of the ESW System, is designed to the ANSI B31.1 Power Piping Code and the requirements of Quality Group D. The design pressure and temperature of the ESW System and HPCS service water piping system are 150 psig and 150*F, respectively. The supply temperature of the service water is set at 100*F maxi-mum, with the fouling f actor for equipment at 0.002 inch. System low point drains and high point vents are provided as required. ("N The HPCS service water system, Division 3, is maintained full of
-' water when not in service.
System components and piping materials are selected to be compati-ble with the available site cooling water in order to minimize corrosion. Adequate corrosion safety factors are used to assure the integrity of the system during the life of the plant. During all plant operating modes, both Division 1 and Division 2 have at least one service water pump operating. The re fore , if a LOCA occurs, the service water system requi red to shut down the plant safely is already in operation. If a loss of offsite power occurs during a LOCA, the pumps momentarily stop until transfer to standby diesel-generator power is completed. The pumps are restarted automatically according to the diesel loading sequence. If a LOCA occurs, all nonsafety-related components are automati-cally isolated from the ESW System. Consequently, no operator O 9.2-7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.1.3.2 Safety Evaluation of Equipment (Continued) action is required, following a LOCA, to start the ESW system in its LOCA operating mode. (To be confirmed by the Applicant, who s upp lies these pumps . ) It is a system requirement that the chloride (as chloride ion) content in the ESW System and the HPCS service water supply shall not normally exceed 34 ppm. The analysis and control of the chloride content is to be done outside of the Nuclear Island. The Applicant is to furnish a description of chloride control, as well as for other parts of the ESW System located outside of the Nuclear Island. 9.2.1.4 Testing and Inspection Requirements The ESW System and llPCS service water subsystem are designed to permit periodic in-service inspection of all system components to assure the integrity and capability of the system. The ESW System is designe' fcr periodic pressure and functional testing to assure: ( 1) the structural and leaktight integrity by visible inspection of the components; (2) the operability and the performance of the active components of the system; and (3) the operability of the system as a whole. The tests shall assure, under conditions as close to design as p rac ti cal , the performance of the full operational sequence that brings the system into operation for reactor shutdown and for LOCA, including operating of applicable portions of the Reactor Protection System and the trans fer between normal and standby power sources. O 9 . 2- 8
- GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.1.4 Testing and Inspection Requirements (Continued)
The ESW System is designed to conform with the foregoing require-ments. Initial tests shall be made as described in Subsec-tion 14.2.12.1.38. 9.2.1.5 Instrumentation and Control Requirements All equipment is provided with either globe or butterfly valves to give the capability for manual control. These valves are accessible downstream of the equipment for regulation of flow through the equipment or for balancing the circuits. The isola-tion valves to the nonessential service water system are auto-matically and remote-manually operated. Pressure taps or indicators at equipment are provided to enable the operator to adjust the differential pressure across each heat () exchanger or cooler and also to allow leak checking. Maximum pressure drop for individual pieces of equipment is specified in the plant requirements documents. Locally mounted temperature indicators or test wells are furnished on the equipment cooling water discharge lines to enable verifica-tion of specified heat removal during plant operation. The required heat removal and flow rates are shown in Tables 9.2-1 and 9.2-2. The combination of pressure taps (or indicators) and temperature indicators allows correct system balancing with or without a system heat load. For purposes of system balancing, provisions for flow measurement are provided at the following locations : (1) total Fuel Building discharge; (2) CCW exchangers discharge; (3) Control Building chiller ccndensers; (4) nonessential total flows; (5) RHR exchangers; (6) diesel-generator exchangers; and
) (7) Radwaste Building.
J 9.2-9
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.1.5 Instrumentation and Control Requirements (Con tinued) Connections to the Process Radiation Monitoring System are pro-vided on the RHR exchangers discharge and Division 1 and 2 return headers to detect radioactive contamination resulting from a tube leak in one of the RHR exchangers , CCW exchangers or fuel pool exchangers. Isolation valves for RHR heat exchangers, and nonessential service water subsystems on Divisions 1 and 2 are provided with remote manual switches and indication on the remote shutdown panel. 9.2.2 Closed Cooling Water System 9.2.2.1 Design Bases 9.2.2.1.1 Power Generation Design Bases (1) The Closed Cooling Water (CCW) System is designed to cool aelected auxiliary plant equipment over the full range of reactor operation during normal and upset conditiens. (2) The CCW System is designed to provide a closed cooling water loop between systems which are potentially radio-active and the service water used for cooling. This provides an additional barrier between the possibly contaminated systems and the service water, which may be discharged to the environment. (3) The CCW system is designed to take into account and j minimize long-term corrcaion which may degrade system performance. O 9.2-10 i
GESSAR II 22A7007 238 NUCLEAR ISLNiiD _ Rev. 0 k 9.2.2.2 System Description l' ' [O s The CCW System is a closed loop system which 'provides' parallel-flow cooling to auxiliary equipment in the containment, drywe ll, + auxiliary building, fuel building and radwaste building'.,See s
" ~
Table 9.2-4 for a listing of the equipr.cnt served and the details on heat load and flow rate. dighres 9.2-3(a) and (b) (K-120 A&B) show in schematic form the CCW System and the' equipment served. '
. + ,
The CCW System consists of two circulating punips, tdo shell and ' tube heat exchangers , an expansion tank, and the issocia'ted piping, , valves, and instrumentation hs sh,own on Figur62E 9.2-3a and 36.. Each of the two pumps and two heat exchangeru li siE[d. for 100% <
- u. -. ,
capacity based on normal coolingsrequirements. Th'uJ7'only one pump and one heat exchanger need'be in operation tosperform the _ normal cooling function. Cross cUnnedtions are provi9ed so that either exchanger can be us[d
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iith'citlfer phmpt - - (j ,
, 9 Both heat exchangers are of the shell and tube typ'e, desig,ned and
~
constructed in accordance wiCh ASME Code Section VIII, Division 1. The expansion tank , to accommodate expansion and surge in ' the system, is also designed and constructed in accordance with
, , -.~w ASME 3 Code, Section VIII, Division 1. CCW sy' stem cdnt'ainment and dry-well penetrations, and ieSlation valves are des [g'ned to Seismic
~
Category I, ASME Code , Section III, Clas.c 2,sQualfE7 Group B and (_ Quality Assurance B requireme n ts . Th'e remainder of the system is designed in accordance with ANSI B31 1, Powe1 \ Piping Code and i the requirements of Quality Group D. Tne CCW System is capable of beingisolatedunderLOCAconditiynsofhighdrywellpressureand low RPV water level. ^ ~
- N *
. m The expansion tank is sized f r 3Sb-ga] . capaci,ty to prevent lobb%
/
of pump suction during. conditions such as reffilling'<ani empty sys-l
- s s -
tem component. A nitrogen blanket is provided in the exphnsion I [ ) tank to prevent air from contacting the system water. Mafeup ( \~/ l water, from the plant demineralized water system, is injected into l 9.2-11 ' l
's
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.2.2 System Description (Continued) the system by way of the expansion tank. The CCW System is filled with demineralized water. The water used to fill the cooling system will not be chemically treated to limit corrosion. The system components are designed with an appropriate corrosion allowance added to all piping and equipment. Contact of the water with the atmosphere is minimized to prevent the addition of oxygen to the systems. The expansion tank and any other open parts of the system are sealed against the a tmosph ere . During normal operation, the two CCW pumps are powered from sepa-rate electrical nondivisional buses. In the case of loss of pre-ferred AC power, the pumps are powered from the standby diesel-generators. In the event of a LOCA accompanied by a loss of pre-fe rred AC power, no power is supplied. .- 9.2.2.3 Safety Evaluation The operation of the closed cooling water system is nonsafety-re l a ted , since it is not required to assure any of the .following condi tions : ( 1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safety shutdown condition; and (3) the ability to prevent mitigate the consequences of accidents which could result in potential offsite expo- . sures comparable to the guideline exposures of 10CFR100. The CCW System provides a barrier between potentially contaminated systems and the service water which may be discharged to the 9.2-12
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () 9.2.2.3 Safety Evaluation (Continued) environment. A radiation monitor is provided in the CCd System to indicate inleakage to this system from a contaminated system. Additional radiation monitors are provided in the service water system discharge to detect radioactive contamination which may enter this system from the CCW System. Should a break occur in the 1 CCW System, cooling water leakage will be collected in the floor drains and sumps and piped to the radwaste system for processing. Leakage is detected by pressure instrumentation in the system supply headers , by monitoring local sump flow (utilizing the leak detection system) and by monitoring the expansion tank level. Portions of the CCW System which penetrate the containment are provided with redundant, divisional containment isolation valves which are automatically activated by a LOCA signal, or may be man-ually actuated by the operator in the control room. ( A single failure analysis has not been provided for the closed cooling water system, since this system is not required to perform a safety function. 9.2.2.4 Tests and Inspections Pumps in the CCW System are proven operable by their continuous use during normal plant operation. Motor-operated isolation valves can be tested to assure that they are capable of opening and closing by operating manual switches in the control room and observing the position lights. Routine visual inspection of the system components, instrumentation and trouble alarms are adequate to verify system operability. 9.2.2.5 Instrumen tation Application The CCW System is a balanced constant-flow system. Local pressure and temperature gauges and pressure test points are provided i 9.2-13
GESSAR II 22A7007 238 NUCLEAR ISLAND Revo 0 9.2.2.5 Instrumentation Application (Continued) throughout the system to measure pressure drop and temperature rise. A temperature sensor downsteam of the heat exchangers transmits a temperature signal that actuated an alarm in the control room on high water temperature. This signal is also received by a temperature controller that modulates a three-way va lve , controlling the amount of CCW passing through the CCW heat exchangers. The standby pump is brought on line if the other pump cannot maintain pressure in the system. The standby pump is started and an alarm actuated in the control room by means of a pressure switch in the cooling water header.. System water level is maintained from the plant domineralized water system by a level controller in the system surge ta nk . Connections to and from the process radiation monitoring system are provided in the CCW System return header to detect and indi-cate the presence of radioactive fluid in-leakage to the system. 9.2.3 Domineralized Water Makeup System 9.2.3.1 Design Bases (1) Raw water treatment shall be described by the Applicant. (2) Domineralized water, as received from the BOP, shall be distributed within the Nuclear Island. (3) The Demineralized Water Makeup (DWM) System shall pro-vide reactor quality water for preoperational tests, startup and for normal operational uses in systems within the Nuclear Island. O 9.2-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (v ') 9.2.3.1 Design Bases (Continued) (4) The DWM System shall maintain a required makeup water quality as follows: Conductivity (pmho/cm) <3 at 25 C Chlorides, as C1 (ppm) <0.05 pH 5.3 to 7.5 at 25 C Apply conductivity and pH limits after correction for dissolved CO 2' It is the responsibility of the Applicant to monitor, and make necessary corrections for, the above properties. v (5) The DWM System shall supply water for the equipment shown in Figures 9.2-4a, b, and c (K-124A, B and C) and in Table 9.2-5, which lists flow rates. (6) The DWM System is not safety-related, except for those ESF portions which are part of primary or secondary containment. (7) All piping, except those portions which are part of primary and secondary containment, shall be designed to ANSI B31.1, Power Piping Code, per the requirements of Quality Group D of Regulatory Guide 1.26. (8) The portion of the demineralized water piping which penetrates containment shall be designed to Seimsic N/ 9.2-15
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.3.1 Design Bases (Continued) Category 1 and ASME Code Section III, Class 2, Quality Group B and Assurance B requirements. (9) Isolation valves in series shall be provided for the piping pene tration through containment. The valves are automatically actuated by a LOCA isolation signal. A remote manual switch and open/ closed position lights are provided in the Control Room for verification of proper valve operation. (10) Piping which penetrates secondary containment walls shall be designed to Seismic Category I, ASME Code, Section III Class 3, Quality Group C and Quality Assurance B requirements. (11) All tanks, connected piping and other equipment are made of corrosion-resistant material. 9.2.3.2 System Description Raw water treatment and demineralization of the feed to the DWM System are provided by the Applicant. Distribution of the deminer-alized water is not a safety-related function. Therefore, only primary and secondary containment isolation valves as well as certain, connected portions of piping are safety-related. Demineralized water is distributed throughout the Nuclear Island to the equipment and systems listed in Table 9.2-5. The flow rates shown in the table are not average, but rather the maximum, instantaneous flow that may occasionally be needed. Interface requirements from the BOP are , for water of the quality listed in Subsection 9.2.3.1, 160 gpm a t 85 psig at the Auxiliary 9.2-16
. V
..' s GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 9.2.3.2 System Description (Continued) {"'}l w Building and 124 gpm at 55 psig at the Radwaste Building interface. Minimum water temperature is 40 F with a maximum of 100 F. 9.2.3.3 Safety Evaluation Operation of the DWM System is not required to assure any of the following conditions : (1) integrity of the reactor coolant pressure boundary; (2) capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) ability to prevent or mitigate the consequences of events which could result in potential offsite exposures.
/]
L ,' The DWM System is not safety-related. However, the systems incor-porate features that assure reliable operation over the full range of normal plant operations. 9.2.3.4 Tests and Inspections The Domineralized Water Distribution System is proved operable by its use during normal plant operation. Portions of the system normally closed to flow can be tested to ensure operability and integrity of the system. The air-operated isolation valves are capable of being tested to assure their operating integrity by manual actuation of a switch located in the control room and by observation of associated position indication lights. fT v 9.2-17
GESSAR II 22A7007 238 NUCLEAR ISLAND Revo 0 9.2.3.5 Instrumentation Application h While many of the recipients of water from the DWM System require intermittent flows, the overall system is a balanced, constant flow system during normal plant operation. Local pressure and temperature gages are provided through the system to measure pressure and tempe r ature . Flow to the various systems is balanced by means of manual valves at the individual takeoff points. Divisional isolation valves are installed at the primary containment boundaries. Discharge pres-sure from the Nuclear Island domineralized water booster pump to the CCW surge tank is regulated by means of a pressure control valve. 9.2.4 Potable and Sanitary Water Systems 9.2.4.1 Design Bases (1) The Potable Water System shall be designed to distribute potable water, ob tained from the BOP , to the Reactor Island in suf ficient quantity for normal plant operation and for shutdown periods. (2) The Sanitary Water System shall be designed to effec-tively remove Potable Water System waste water from the Reactor Island. (3) The Potable Water System shall not in any way compromise the quality of water supplied for potable use by the owne r- app licant . (4) There shall be no cross connections with water supply or drain systems that are in any way susceptible to radio-logical contamination. 9.2 18
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.4.2 System Dascription See Figure 9.2-5 (K-126) for a graphic description of the Potable Water System and Figure 9.2-6 (K-0 51) for the Sanicary Water System. The Potable and Sanitary Water Systems are completely independent of, and not connected to, any systems that contain, or can con-ceivably contain, radioactive materials. These are essentially closed systems which do not have any connections with radioactive drains or radioactive waste disposal systems. Incoming potable water, from the BOP, enters the Control Building through a pipe line designed to ANSI B31.1 requirements. This qualification prevails up to and including the first block valve inside the Con trol Building. Minimum requirements are 75 gpm and 18 psig. I N_'/ Cold water is distributed to all Control Building sanitary fix-tures, sinks , wash basins, drinking fountains and hose bibs. A branch line tapped of f the cold water line route s water through an electric water heater located within the Control Building for use in sinks, showers and wash basins. The heater is sized to provide a 100-gph recovery rate and stores 120 gallons. Ele ctrical rating is 30 kW (480V, 39, 60 Hz). A pressure / temperature relief valve is provided which discharges to the floor drain. Maximum system pressure is governed by a preset pressure control valve. Pressure surges, which could result in destructive water hammer, are controlled by five strategically located water hammer accumulators. (Applicant to describe supply system.) O 9.2- 19
GESSAR II 22A7007 238 NUCLEAR ISL1 Rev. 0 9.2.4.2 System Description (Continued) The Sanitary Water System, which is the drain system for potable water waste, is a closed system with no exposure to other drains carrying potenti al radiological contamination. Water in the Sani-tary Water System is grab-sampled at regular intervals for radio-logical conta:aination, if such contamination is detected, the waste water can ba directed to the Radwaste System for cleanup. (Appli-cant to describe waste disposal system.) 9.2.4.3 Safety Evaluation The Potable and Sanitary Water Systems are not necessary to assure: (1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor; or (3) the capability to prevent or mitigate the consequences of events which co'ild result in potential offsite exposures. The Potable Water System is not directly connected with any other system or equipment that could be a source of accidental, radio-logical contamination. There are no potable water lines located in buildings other than the Control Building. In areas where showers are required, potable water is supplied on a dead-end basis. No potable water is supplied to buildings or areas from any portion of the system that has been routed through areas where radioactive operations are conducted. Also, check-valves are installed, where required, to prevent radioactive back-flow from entering the remainder of the system. 9.2.4.4 Tests and Inspections t The Potable Water and Sanitary Water Systems are proved operable by their use during normal plant operation. The portions of each system normally closed to flow can be tested to ensure operability and integrity of the system. 9.2-20
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O. ( ) 9.2.4.5 Instrumentation Application QJ Instrumentation within the control building is limited to a pres-sure control valve and to a thermostat on the water heater. All other BOP instrumentation is the responsibility of the Applicant to describe. 9.2.5 Ultimate Heat Sink ( Appl icant to supply) 9.2.6 Condensate Storage Facilities and Distribution System Design of the Condensate Storage facility is the respontibility of the Applicant. A condensate distribution system exists within the Nuclear Island and is described below. (Remainder of subsection to be completed by the Applicant.) v Within the Nuclear Island, the Condensate Distribution System sup-plies condensate from the BOP to the RCIC and HPCS Systems, the CPD System and the containment fire protection network. The sup-pression pool is the alternate backup system to the RCIC and HPCS Systems, and high pressure fire protection water is the alternate for containment fire protection. A redundant supply within the BOP is provided for the CRD System. Appropriate surge capacity for the HPCS and RCIC pump suctions is provided. Other uses of condensate within the Nuclear Island are for general service, equipr.ent and line flushing, pump sealing, etc. Conden-sato is also used for refueling activities and for refilling the Water Positive Seal System tank and several Radwaste system tanks. s_- 9.2-21
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.6.1 Design Bases 9.2.6.1.1 Safety Design Bases (1) The condensate distribution system is not safety related except for two parts - (a) the pipe penetration into containment which is an engineered safety feature, and (b) the storage provisions described in the following paragraph. (2) A 7000-gal. surge volume shall be provided in the piping to the RCIC and HPCS systems. This volume allows for automatic switchover to the suppression pool upon loss of water supply from the condensate storage tank. The 1000-gal. surge volume header located in the Auxiliary Building and the piping from the header to the RCIC and HPCS pumps shall be designed to Seismic Category I and ASME Code Section III, Class 2, Quality Group B and Quality Assurance B requirements. Isolation valves and penetration piping for the line into the drywell and con-tainment structures shall also meet the above requirements. (3) Pipes shall be touted to the HPCS and RCIC pump rooms so that a break will not flood or damage any safety-related equipment outside the pump rooms. A break inside one of these pump rooms will flood only that pump room. 9.2.6.1.2 Power Generation Design Bases (1) A minimum water volume of 150,000 gallons is required in the condensate storage tank. This volume of water is capable of removing, over an 8-hr period, the decay heat generated by a reactor scram from rated powe and also acts as the preferred water source for the RCIC and HPCS systems. 9.2-22
GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 0 9.2.6.2 System Description [\ /~') The description of the supply portion of the condensate system is the responsibility of the Applicant. The distribution system within the Nuclear Island is described schematically in Fig-urec 9.2-4a, b, and c (K-124A, B, C). Minimum water volume inventory of 150,000 gal. is required in the condensate storage tank for the RCIC and HPCS Systems. This water is used to remove the decay heat generated over an 8-hr period from the time the reactor is scrammed from rated power. The only outlets below the 150,000-gal. minimum water volume of the condensate stor-age tank are for the RCIC and HPCS pumps. The takeoffs above the 150,000 gal minimum water volume are for non-safety functions. Manual valves for HPCS, RCIC and CRD pump suction condensate lines in the BOP shall be locked open to prevent them from being inadver-tently closed. Manual valves in the HPCS, RCIC and FPCCU pump dis-() charge lines returning to the condensate tank in the BOP shall be locked open to prevent over-pressurizing the lines. See Subsec-tions 6.3, 5.4.6 and 4.6, respectively, for further discussion. A 7000-gal. surge volume, made up of a section of 36-in.-diam pipe, is provided in the Auxiliary Building as a header for suctica of the HPCS and RCIC pumps. A level switch and alarm on the surge volume notifies the control room operator of a low surge volume condition and automatically switches the suction of these pumps over to the suppression pool volume. See Subsections 5.4.6 and 6.3.2.2.1 for further details. As stated previously, the other uses of condensate within the Nuclear Island are for the CRD pump supply, containment fire pro-tection and general uses (e.g., tank and live flushing, utility stations and tank makeup). The use of condensate for containment fire protection is discussed further in Appendix 9A. s_- 9.2-23
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.6.2 System Description (Continued) The various uses and flow requirements for the Condensate Distribu-tion System within the Nuclear Island are outlined in Table 9.2-6. General use points are shown schematically on the P& ids. All flow rates given are based on intermittent operations. Water supply pressure and flow requirements from the BOP are as follows: Flow Pressure Use (gpm) psig (min) HPCS 7800 7 RCIC 700 7 CRD 200 12 Other 1900 120 Radwaste 550 120 Water quality requirements are a minimum temperature of 40 F and a maximum of 100 F, a maximum conductivity of 3.0 micromho/cm at 25 C, maximum chloride concentration of 0.05 ppm and a pH range of 5.3 to 7.5 at 25 C. It is the responsibility of the Applicant to conform to above parameters and to monitor, and make necessary cor- t rections for, the above properties. Piping in the Condensate Distribution System is primarily designed to ANSI B31.1, Quality Group D requirements, except as specified. Piping and associated isolation valves for containment penetrations are designed to Seismi'c Category I, ASME Code, Section III, Class 2, Quality Group B, Quality Assurance B requirements. The same cri-teria apply to piping associated with the HPCS and RCIC systems, including the surge volume header. Piping and/or valves that are part of the secondary containment boundary are designed to Seismic Category I, ASME Code, Section III, Class 3, Quality Group B and Quality Assurance B requirements. 9.2-24
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ( 9.2.6.3 Safety Evaluation The system has no safety-related function. Operation of the Condensate Distribution System within the Nuclear Island is not required in order to assure: (1) integrity of the reactor coolant pressure boundary; (2) capability to shut down the reactor safety; or (3) capability to prevent or mitigate the consequences of events which could result is potential offsite exposures. That portion of the system which connects to the HPCS and RCIC Systems is designed to appropriate seismic and ASME Code require-ments. Redundant level instrumentation on the surge volume ()
%^J automatically alarms and transfers the suction of the HPCS and RCIC pumps to the suppression pool. Air-operated, fail-closed valves also isolate the return lines from those systems in the event of low surge volume and consequent switchover to the suppression pool supply. As such, a malfunction of the Condensate Storage Facili-ties and Distribution System cannot preclude operation of the HPCS and RCIC Systems.
Redundant, divisional, Class lE-powered containment isolation valves close on a LOCA containment isolution signal. Also, the design of the condensate lines which penetrate secondary contain-ment is adequate to maintain the required secondary containment negative pressure. 9.2.6.4 Tests and Inspections The Condensate Distribution System is proven operable by its con-
/}
kJ tinuous use during normal plant operation. Portions of the system 9.2-25 t
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.6.4 Tests and Inspections (Continued) normally closed to flow can be tested to ensure operability and system integrity. Test connections are provided only at the con-tainment isolation valve penetrations in order to verify their leak tigh tness . The air-operated isolation valves are capable of being tested to assure their operating function by manual actuation of a switch located in the control room and by observation of associated posi-tion indication lights. 9.2.6.5 Instrumentation Application The pipe penetration through containment is provided with two iso-lation valves, one in each side of the containment. The isolation valves are automatically actuated by a LOCA isolation signal. A remote manual switch and open/ closed position lights are provided in the control room for verification of proper valve operation A normally closed, manual valve is provided for the pipe penetra-tion thrcugh the drywell. The level switches are provided for the 7000-gal. surge volume of the RCIC and HPCS Systems (see Subsections 5.4.6 and 6.3.2.2.1, respectively). The low level alarm in the Control Room indicates low surge volume conditions and provides for automatic switchover to the alternate water source. Air-operated, fail-closed valves isolate the condensate suctions and return lines in this event. The remainder of the Condensate Distribution System is balanced by manual valves during normal operation. Local instruments provide temperature, pressure and flow indications where required. O 9.2-26
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/~N 9.2.7 Plant Chilled Water Systems
( )
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The Plant Chilled Water Systems are composed of three separate systems: (1) Drywell Chilled Water System (2) Control Building Chied Water System (3) Reactor Island Chilled Water System Each system is discussed separately in the following subsections. 9.2.7.1 Design Bases 9.2.7.1.1 Power Generation Design Bases 9.2.7.1.1.1 Drywell Chilled Water System
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The Drywell Chilled Water System (nonsafety-related) shall provide chilled water to the coils of the drywell coolers and the RWCU pump room coolers to maintain design thermal environments during normal and upset conditions. The supply temperature is 60 F, the return temperature is 70 F. 9.2.7.1.1.2 Control Building Chilled Water System The Control Building Chilled Water System (safety-related) shall provide chilled water to the cooling coils of the Control Building air conditioning units, and to the Auxiliary Building Electrical Switchgear Room coolers. The supply temperature is 45 F, the return temperature is 55 F. 9.2.7.1.1.3 Reactor Island Chilled Water System The Reactor Island Chilled Water System (nonsafety-related) shall s provide chilled water to the Auxiliary Building, Fuel Building, l ~, i 9.2-27 i I
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.7.1.1.3 Reactor Island Chilled Water System (Continued) Radwaste Building, and containment fan coil uni ts for space cooling and dehumidification. The supply temperature is 45 F, the return temperature is 55*F. 9.2.7.1.2 Safety Design Bases On ly the Control Building Chilled Water System performs a safety design function. (1) The Control Building Chilled Water System shall deliver 45 F chilled water to the Control Building air condi-tioning unit coo'ing coils and one of the auxiliary electrical switchgear room unit cooling coils in both Division 1 and Division 2 during shutdown of the reactor, ope rating modes and abnormal reactor conditions including LOCA. (2) Suf ficient redundancy and electrical and mechanical separation shall be provided to ensare proper operations under all conditions. (3) The system shall be designed and constructed in accord-ance with Seismic Category I, ASME Code, Section III, Class 3 requirements. (4) The system shall be powered from Class lE buses. 9.2.7.2 System Description 9.2.7.2.1 Dryuell Chilled Water System The Drywell Chilled Water system components are listed in Table 9.2-7 and shown in Figure 9.2-7a&b (K-12 3 A&B) . System 9.2-28
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O gT 9.2.7.2.1 Drywell Chilled Water System (Continued) components consist of two 100% chillers, each with pumps, serving a common chilled water distribution system connected to the chilled water cooling coils in six drywell coolers and two RWCU pump room coolers. Condenser cooling is from the nonessential portion of the ESW System. Each chiller and pump set has a three-way mixing valve for automatically controlling the temperature of the chilled water delivered to the dryuell and the RWCU pump room cooling coils. During startup following any thermal transient, the control room operator can reset the three-way valve to limit the tempera-ture of the water returned to the chiller evaporator. Each chiller evaporator is designed, f abricated and certified in accordance with the ASME Code Sec: ion VIII, Division 1. An air separator and expansion tank with nitrogen gas blanket limits air adsorption. No chemical feed tank is required. Makeup water is from the Demineralized Water System. O Isolation valves and piping for primary containment penetrations are designed to seismic Category I, ASME Code, Section III, Class 2, Quality Group B, Quality Assurance B requirements. The supply line penetration has a Division 1 isolation valve outside containment, and Class 2 piping into the drywell. The return line penetration has divisional isolation valves inside and outside containment. Piping for penetrations of secondary containment is to Seismic Category I, ASME Code, Section III, Class 3, Quality Group B and Quality Assurance B requirements. Equipment, controlL and instruments are connected to the diesel-generator bus during loss of preferred power. No diesel-generator power is available to this system during a LOCA. The capacity of only three drywell coolers and one RWCU cooler operating provide () V suf ficient cooling for safe reactor shutdown during loss of 9.2-29
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.7.2.1 Drywell Chilled Water System (Continued) preferred power, with nonemergency power available from one diesel generator. See Subsection 9.4.5 for further detaila. 9.2.7.2.2 Control Building Chilled Water System The Control Building Chilled Water system consists of completely redundant subsystems, Division 1 and Division 2. Each subsystem consists of a 100% chiller unit, expansion tank and air separator, pump, instrumentation and distribution piping and valves to corresponding division Control Building Chiller Room cooling coils, Control Building air conditioning unit cooling coil and Auxiliary Building Electrical Switchgear Room cooling coile. No chemical feed tanks are required. Equipment is listed in Table 9.2-8 and on Figure 9.2-8a&b (K-129A&B). Each cooling coil has a three-way valve controlled by a room thermostat. The subsystems are designated Divison 1 and Division 2 and receive their power trom Division 1 and Division 2 power, respactively. One compressor is the operating unit, while the other is on standby. Condenser cooling is from the corresponding division of ESW. Piping and valves for the Control Building Chilled Water System, as well as the cooling water lines from the ESW System, designed entirely to ASME Code, Section III, Class 3, Quality Group C, Quality Assurance B requirements. The extent of this classifica-tion is up to and including drainage block valves. There are no primary or secondary containment penetrations within the system. Low flow of chilled water in the redundant division of the Control Building HVAC, low flow in the Auxiliary Building electrical area self-contained air conditioning unit or high temperature in the Auxiliary Building electrical area causes the chiller unit to start 1 9.2-30
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.7.2.2 Control Building Chilled Water Systems (Continued) [)) automatically. Makeup water is supplied from the Demineralized Water System, at the expansion tank. 9.2.7.2.3 Reactor Island Chilled Water System The Reactor Island Chilled Water System components are listed in Table 9.2-9 and Figure 9.2-9a,b,c (K-125A,B,C). The system consists of two 50% capacity chillers, coils and piping. Hermetically sealed, centrifugal compressors, piped in parallel and with freon refrigerant, are supplied. Both chillers operate simultaneously in carrying the total design load. Chilled water is circulated to the area fan coil units by three 50% capacity (one on standby), hori-l zental split case centrifugal pumps connected in parallel to the supply header for the cooling coils. An air separator and an expansion tank with a nitrogen gas blanket limits air entrainment. (~') L.J No chemical feed tank is required. Primary containment penetration design criteria are identical with that of the Drywell Chilled Water System. 9.2.7.3 Safety Evaluation 9.2.7.3.1 Drywell Chilled Water System Operation of the Drywell Chilled Water System is not required to assure the following conditions: (1) integrity of the reactor coolant pressure boundary; (2) capability to shut down the reactor and maintain it in a safe shutdown condition; and O 9.2-31
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.7.3.1 Drywell Chilled Water System (Continued) (3) ability to prevent or mitigate the consequences of events which could result in potential offsite radio-logical exposures. The Drywell Chilled Water System is not safety-related. However, it does incorporate features that assume reliable operation over the full range of normal plant operations. Also, freon relief valves are vented to the outside atmosphere. Portions of the chilled water system which penetrate the contain-ment and drywell are provided with isolation valves and penetra-tions which are Seismic Category I, Safety Class 2. The valves may be manually operated from the control room, except when a LOCA signal assumes control. Portions of the system which is part of secondary containment boundary are Seismic Category I, ASME III, Class 3. 9.2.7.3.2 Control Building Chilled Water System The Control Building Chilled Water System is enclosed in a Seismic Category I structure, protected from flooding and tornado missiles. All essential components of the system are designed to be operable during a loss of normal power by connection to the ESP buses. Redundant components are provided to ensure that any single component failure does not preclude system operation. The sys tem is designed to meet the requirements of Criterion 19 of 10CFR50. Each chiller is isolated in a separate room, and there are no cross-ties between the divisions. O 9.2-32
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Reactor Island Chilled Water System ()' 'v 9.2.7.3.3 Operation of the Reactor Island Chilled Water System is not a requirement in assuring the following: (1) integrity of the reactor coolant pressure boundary; (2) capability to shut down the reactor and maintain it in a safe shutdown condition; and (3) ability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures cited in 10CFR100, floweve r , the system does incorporate features that assure its reliability over the full range of operation. These features include the installation of multiple components for equipment such (}' '" as chiller, pumps and fan coil units. Additional features include failsafe position on the system controls and equipment safety con-trols. Certain standby equipment is provided for use during main-tenance. A normally closed valve cross-connection with the standby drywell chiller allows operation during chiller maintenance. Auto-matic functioning of the valve restores the drywell chiller to drywell service during loss of normal power. Portions of the chilled water system which penetrate the contain-ment are provided with isolation valves and penetrations. These valves are actuated during LOCA and may be operated manually from the main control room provided that a LOCA signal is not in present. O v 9.2-33
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.7.4 Tests and Inspections 9.2.7.4.1 Drywell Initial testing of the system includes performance testing of the chillers, pumps and coils for conformance with design heat loads, water flows, and heat transfer capabilities. An integrity test is performed on the system upon completion. Provision is made for periodic inspection of major components to ensure the capability and integrity of the system. Local display devices are provided to indicate all vital parameters required in testing and inspections. The chillers are tested in accordance with ASHRAE Standard 30 (Methods of Testing for Rating Luid Chilling Packages). The pumps are tested in accordance with standards of the Hydraulic Institute. ASME Section VIII and TEMA C standards apply to the heat exchangers. The cooling coils are tested in accordance with ASHRAE Standard 33 (Methods of Testing for Rating Forced Circulation Air-Cooling and Heating Coils). 9.2.7.4.2 Control Building Initial testing of the system includes performance testing of the chillers, pumps and coils for conformance with design capacity water flows and heat transfer capabilities. An integrity test is performed on the system upon completion. Provision is made for periodic inspection of major components to ensure the capability and integrity of the system. Local display devices are provided to indicate all vital parameters required in testing and inspections. Standby features are periodically tested by initiating the transfer sequence during normal operation. O 9.2-34
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (T 9.2.7.4.2 Control Building (Continued) C/ The chillers are tested in accordance with ASHRAE Standard 30. The pumps are tested in accordance with standards of the Hydraulic Institute. ASME Section VIII and TEMA C standards apply to the heat exchangers. The cooling coils are tested in accordance with ASHRAE Standard 33. 9.2.7.4.3 Reactor Island Initial testing of the system includes performance' testing of the chillers, pumps and coils for conformance with design tonnages, water flows, and heat-transfer capabilities. An integrity test is performed on the system upon completion. Provision is made for periodic inspection of major components to ensure the capability and integrity of the system. Local display (\ devices are provided to indicate all vital parameters required in '\ '/ testing and inspections. The chillers are tested in accordance with ASHRAE Standard 30. The pumps are tested in accordance with standards of the Hydraulic Institute. ASME Section VIII and TEMA C standards apply to the heat exchangers. The cooling coils are tested in accord-ance with ASHRAE Standard 33. 9.2.7.5 Instrumentation Application 9.2.7.5.1 Drywell Chilled Water System 9.2.7.5.1.1 Instrumentation Application A regulated supply of domineralized makeup water adds water to the expansion tank by water level controls, and the chiller units are controlled individually by remote manual switches. Chilled water (]N Q 9.2-35
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.7.5.1.1 Instrumentation Application (Continued) temperature is controlled by a three-way valve positioned by thermostatic controls in the drywell. A temperature controller and flow switch continuously monitors the discharge of the evaporator. If the temperature of the chilled water drops below a specified level, the controller automatically adjusts the temperature control inlet guide vanes of the chiller compressor. Flow switches prohibit the chiller from operating unless there is water flow through both evaporator and condenser. See subsection 3.11.1 for temperature requirements. Chilled water flow into and out of the containment is controlled by isolation valves which close automatically upon LOCA signal. Each chiller is provided with a booster pump on the condenser water circuit which also has a three-way valve to maintain uniform condenser water temperature. Condenser water is provided from the essential service water system. The three-way valve on the chilled water circuit controls the temperature of the chilled water to the drywell cooling coils from the zone averaging thermocouple controller. The thermocouples are located in each zone of the drywell. The control room operator car. adjust the three-way valve position during startup and whenever high chilled water return temperatures are indicated and alarmed. Remote controlled solenoid actuated pneumatic valves permit isola-tion of any drywell cooling coil in the event of the coil develop-ing a detectable leak. 9.2.7.5.2 Control Building Chilled Water System A regulated supply of makeup water for the chilled water circuit is provided. O 9.2-36
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.7.5.2 Control Building Chilled Water System (Continued) (v) The chilled water pumps are locally controlled. The standby chiller is equipped with an interlock which automatically starts the standby chiller and pump upon failure of the operating unit. The chiller units can be controlled individually from the main control room by a remote manual switch. Chilled water temperature is controlled by inlet guido vanes on each chiller refrigerant circuit. Condenser water flow is controlled by a three-way valve to provide constant inlet condensate water temperature. A temperature controller and flow switch continuously monitors the discharge of each evaporator. If the temperature of the chilled water drops below a specified level, the controller automatically adjusts the position of the compressor inlet guide vancs. Flow switches prohibit the chiller from operating unless (J\ there is water flow through both evaporator and condenser. Each cailler is provided with a booster pump on the condenser water circuit utilizing water from the essential service water system. See Subsection 3.11.1 for temperature requirements. 9.2.7.5.3 Reactor Island Chilled Water System A regulated supply of makeup water for the chilled water circuit is provided. The chilled water pumps are controlled locally. The standby pump is equipped with an interlock to automatically start the standby pump upon f ailure of either or both of the two operating pumps. The chiller units can be controlled individually by a local switch or by a two-position selector, which enables the cperator to select () the order in which the chillers are to operate. The lead eniller 9.2-37
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.7.5.3 Reactor Island Chilled Water Systcm (Continued) operates until its maximum load is reached, and then the next chiller in line automatically starts and shares the load. Chilled water temperature is controlled by the number of chiller units in operation. However, the individual temperature controller in each chiller unit overrides the chilled water temperature control on low chilled water supply temperature. Condenser outlet tempera-ture is monitored and regulates condenser cooling water flow. A temperature controller and flow switch continuously monitor the discharge of each evaporator. If the temperature of the chilled water drops below a specified level, the controller automatically adjusts the position of the inlet guide vanes in the refrigerant circuit of the chiller. Flow switches prohibit the chiller from operating unless there is water flow through both evaporator and condenser. See Subsection 3.11.1 for temperature requirements. 9.2.8 Heated Water Systems 9.2.8.1 Design Bases A Heated Water Distribution System is provided within the Nuclear Island to deliver heated water to preheat coils, reheat coils, unit heaters and air handling units. The Heated Water Distribu-tion System services no safety-related components. Description of the supply portion of the Heated Water System is the responsi-bility of the applicant. 9.2.3.2 System Description The equipment served by the Heated Water Distribution System is shown in Figure 9.2-10 (K-127A&B). Specific requirements for flows and heat loads are given in Table 9.2-10. Flows and heat loads given are the maximum anticipated, which do not occur 9.2-38
GESSAR II 22A7007 ' 238 NUCLEAR ISLAND Rev. O
~ ~t ;
s 9.2.8.2 System Description (Continued) -
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, V ,
during normal operation. Th,e'naximom }oad occurs during
'~. : ,
containment high purge operaCions. ., W System interface requirements are for $ 195'P supply, 155'P return. Supply pressure shall,H&JLeiss th'an the., system <hsign , pressure of 150 psig, and pressu);e drop is 40 psig. Flow require-ments are 367 gpm at the AuxiJiary Building interface and 276 gpm, at the Radwaste Building. NdterqualityshallbethesameasfCkk s the Domineralized Water System. It is the respons'ibility of the,;,- Applicant to monitor, and make corrections for, thd'above <
~
parameters. - *7 *
~
c . s.. Piping and valves within the sysCem .5 are generally designed;Co - ANSI B31.1, Power Piping Code' requirements. Piping and "acuum breakers at secondary containnent penetrations are designed'(o
~~
Seismic Category I, ASME code,.6e'gtioq III, Class 3, Quality Group C and Quality Assurance b , requirements..- s 9.2.8.3 Safety Evaluation . .
+
s Operation of the !!cated Water Distributi^on System is not requir,ed ,
' ~'
to assure any of the following conoitions:" * ; s's , ' s s . _ . , - - (1) integrity of the reactorcoolantprehsuleboun'darkij
.x (2) capability to shut down the reactnr and. maintain it in a A
safe shutdown condition; or y, (3) ability to prevent or mitigate thJ consequences ~ of events
~
which could result in pc'antial offsite exposures. 4 Consequently, the distribution system itself 'Is n6t safdty-related.
^
The only safety-related portions are the secondary containment pene-trations, which are designed to appropriate requirements. 1 9.2-39
GhdSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.8.4 Tests and Inspections The Heated Water Distribution System is proved operable by its use during normal plant operation. Portions of the systems normally closed to flow can be tested to ensure operability and integrity of the system. Adequate, accessible flow, temperature and pressure indicators are provided to allow sufficient inspec-tion of normal system operation. Initial system tests include flow balancing and coil capacity verification. 9.2.8.5 Instrumentation Application Local temperature indicators and pressure taps are provided at each coil. Temperature indicators are provided in the supply and return lines to the building interfaces. Plow test points are provided. These components are readily accessible during normal plant operation. Waterflow to all reheat coils is controlled by a three-way modu-lating control valve, located in the return line. Balancing valves are provided in the bypass line around the coils, as well as in the return line downstream of the control valve. The return line balancing valves are provided with adjustable memory stops to allow the valve to be fully closed and later reopened to the exact balancing position. Coils which are subjected to 100% outside air and the majority of coils that heat a combination of return air and outside air are controlled by two-way control valves. The control valve opens fully whenever the outside air temperature is belcw the thermostat set pcint. Air temperature control is achieved through f ace and bypass dampers which vary the amount of air passing through the heating coil. A balancing valve is provided in the return line from each coil. 9.2-40
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.8.5 Instrumentation Application (Continued) A temperature sen9or downstream of each preheat coil sounds an . alarm and closes the outside air damper when the preheat coil [ 1eaving temperature reaches a preset minimum value. i An outside air temperature sensor shuts off the fan and closes the inlet air damper on low temperature to protect the coil from freezing. ,
- Unit heatern are not provided with a control valvo. Water con-
- tinuously flows through the coils and space conditions are mair.-
tained by a thermostat which cycles the unit heater f an. l The pressure control valve in the distribution loop bypass line 3 s maintains a constant pressure differential between the supply and l return lines. l 9.2.9 Nuclear Island / BOP Interfaces This section describes in detail the safety-related interface requirement for water systems in the BOP required by the Nuclear { l Island. l f 9.2.9.1 Essential Service Water System I ae Nuclear Island Essential Service Water (Distribution) System f shall be supplied by the BOP. This system shall provide both the ! essential and nonessential service water requirements of the I Nuclear Island. A Division 1, Division 2 and HPCS (Division 3) { l Service Water System shall be provided. Flow, pressure and tem- ! perature requirements for the various reactor modes are given in Subsection 9.2.1. lllI f I r 9.2-41 l i l' ! l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.9.1.1 Design Criteria (1) The ESW is cJassified as a Safety Class 3, Seismic Category I system. Essential piping, pumps, valves and vessels shall be designed and fabricated to ASME Sec-tion III Class 3 requirernents. (2) "he ESW System shull be tornado, wind and flood protected. (Site-specific parameters may be used.) (3) Protection shall be provided for both internal and external missiles. (4) The ESW System shall function following LOCA or loss of offsite power. (5) Provirions shall be made for periodic inspection and testing. (6) The ESW System shall be designed to limit leakage to the environment of radioactive contamination that may enter the ESW f rom the Rilk System. 9.2.9.1.2 Interfacas The ESW System interface is shown on Figures 9.2-1 and 9.2-2 (K-121 and K-122). Tables 9.2-1 and 9.2-2 define the system flow and heat dissipation requirements. Instrumentation and control interface requirements are discussed in Subsection 7.8 and electrical power interface is in Subsec-tion 8.3.2.3. O l 9.2-42
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 9.2.9.2 Ultimate Heat Sink ( The BOP shall provide the ultimate heat sink to provide cooling water to dissipate reactor decay heat and essential cooling system head loads after a normal reactor ahutdown or a shutdown following an accident, including LOCA. Ccoling water is delivered to the Nuclear Island and returned via the ESW System. Note that the ESW System also provides coolant for dissipation of normal cooling system heat loads (nonessential) during normal operation. Nuclear Island heat loads for sizing the ultimate heat sink are given in Subsection 9.2.1. The ultimate heat sink shall be capable of providing sufficient cooling for more than 100 days to permit safe shutdown of the plant and to maintain it in a safe shutdown condition. The ultimate heat sink is classified as Safety Class 3 and Seismic () Category I. 9.2.9.3 Nuclear Island Condensate Distribution System - BOP Condensate Storage Facilities Interface The Applicant shall provide the condensate storage and transfer facilities. Distribution in the Nuclear Island is provided by the Condensate Distribution System (Subsection 9.2.6). 9.2.9.3.1 Design Criteria The condensate storage facility shall be designed by the Applicant to store condensate for the RCIC and HPCS systems, maintain the i level of condensate in the condenser hotwell and provide condensate to other plant systems given in Subsection 9.2.6. The condensate storage tank shall be designed in accordance with codes and standards described in Subsection 3.2. 9.2-43 I
r-____ _ - _ _ - - _ _ _ - _ _ _ - - . _ . GESSAR 11 22A7007 l 238 NUCLEAR ISLAND Rev. 0 9.2.9.3.1 Design Criteria (Continued) ( ! The minimum volume in the condensate storage tank will be ( l 150,000 gallons of water for the RCIC and IIPCS systems. > c The condensate storage and transfer systems are not safety-related. 1 There shall be no takeoff lines below the minimum water level of the condensate storage tank, except for the supply to the RCIC and IIPCS pumps. Manual valves at the condensate storage tank fer IIPCS, RCIC and CRD pump suction }ines shall be locked open. All of the pneumatic-operated valves shall be designed to be fail safe and shall not require a continuous air supply under emergency or abnormal conditions. Supply requirements are given in Subsection 9.2.6. 9.2.9.3.2 Specific Interfaces Specific interfaces between the Nuclear Island and BOP are shown on Figure 9.2-4a and are listed below: P-10 Core cooling water (RCIC and IIPCS preferred supply) from condensate storage tank. P-ll Core cooling pumps (RCIC and IIPCS) bypass to condensate storage tank. ,. P-84 CRD water from condensate storage tank. P-85 CRD and fuel pool water to main condenser. P-86 Supply header from condensate transfer system. 9.2-44
__=_____ ___ _ _ l d GESSAR II 22A7007 ! 238 NUCLEAR ISLAND Rev. O Table 9.2-1 i ESSENTIAL SERVICE WATER f 1 DIVICION 1, TRAIN A, NORMAL OPERATION (l} f I l Flow Heat j Equipment figure No. In (each) (each) Name Fn. Location Use (gpm) (iO6Btu /h) l i Auxiliary building self- X7 3- ACUO 3 b r6 0 100 0(4) l contained air-conditioning unit l Air positive seal computer X6 3-BB015 1-J8 0 8 0
- room cooler l l Air positive seal computer W/CC001A 1-Js a 20 0 i seal cooler l ,
i Control B chiller cond W/P45-ZZ001A 1 - 11 6 1 540 2.16(4) ! I Diesel-generator heat W/R4 3-S001A 1-C2 0 2000 0(3) [ l exchanger l 112 rnix b1r air, oil cooler T41-CC008A 2 - 11 5 0 40 0 FPCCU pump room cooler X63-BB010A 1-J3 1 19 0.05 Fuel pool cooling heat G41-B001A 1 - 11 3 1 1750 8.0(2) f exchanger ! LPCS pump room cooler X 7 3-f,B00 4 1-G8 0 190 0 i Radwaste monitor panel D17-J005 1-Cll 1 0 0 , RCIC pump room cooler X73-BB003 1-111 0 0 30 0 Remote shutdown air- X73-ACU05 1 - 11 8 0 9 0 l conditioning h l t F.;1R heat exchanger E12-B001A&C 1-B9 0 0 0 i I RHR purcp room cooler A X73-BB006 1-C10 0 150 0 [ dilR pump seal cooler A W/E12-C002A 1-E10 0 20 0 ! SGTS room cooler A X63-BB002A 1 - 11 3 0 17 0 Shield annulus fan cooler X63-BB011A 1-J5 0 5 0 Fire protection NA Various (6) 0 0 [ Puel pool makeup 2" ESW127-ADC 1 - 11 1 (7) 0 0 Water positive seal makeup 1" ESW40-ADC 1-G4 (8) 0 0 , 4 Total Essential 4898 10.21 i I i Drywell chiller condenser W/P44-ZZ001A 1-C6 1 1224 4.80(4) ' f Radwaste evaporator G17-D555A 1-B6 1 1400 21.35(5,9) ! condenser j RI closed cooling water P42-BB001A 1-D3 1 3160 23.51(4) l heat exchanger j R: chiller condenser W/P39-ZZ001A 1-06 1 1224 4.80 I l j Steam tunnel cooler X73-BB0llA 1-C6 1 110 0.34 l ) Total Nonessential 7118 54.80 ' 1 I ! Total per Division 1 12016 65.01 i .i G , t [ i 1 9.2-45 i I ' 2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 9.2-1 ESSENTIAL SERVICE WATER (Continued) DIVISION 2, TRAIN B, NOldtAL OPI: RATIO!J Plow 11ea t Equipment rigure No. in (each) (each) Name No. Location Use (qpm) (106 Btu /hr) Auxiliary building self- X73-ACUO4 2-G13 1 100 0.4(4) l l contained air-conditioning , r unit Air positive seal corputer X 7 3- BB017 2-G13 0 8 0 I rocm cooler Air positive seal W/CC0 0 lli 2-;il 3 0 20 0 , corr p r e s s o r seal cooler { I i control B chiller W/P45-ZZ00lb 2-011 0 540 O condenser Di e se l-gene r a to t heat W/R43-S001B 2 -Is 3 0 2000 0( I exchanger 11 , nix blower air, oil T41-CC008B 2-G3 0 40 0 choler ITCCU pump room coo 1er X63-BB010B 2 - 11 3 0 19 0 ruel pool coolinq heat G41-B001B ? - 11 4 1 1750 8' exchanger Rad tr.onitor panel D17-J006 2-C11 1 0 0 PilP heat exchanqer 012-b001B&D 2-B9 0 0 0 X73-BB007 2-C9 0 150 0 kilP ?> ump rer .rn cooler B RIIR pump roon cooler C X73-BB008 2-B10 0 120 0 PliR pump seal cooler B W/E12-C002B 2-E10 0 20 0 FilR pump seal coole r C W/E12-C002C 2-G10 0 20 0 X63-BB002B 2-G3 0 17 0 SGTE room cooler B X63-BBollB ?-J3 0 5 0 Shield annulus fan cooler Iire protection Various Various (6) 0 0 Fuel pool nakeup 2" ESW130-ADC 2-G2 (7) _ 0 0 Total Essential 1809 8.4 1224 O I' Drywell chiller condenner W/P44-ZZ001B 2-D6 0 Radwaste evaporator G 17- D 5 5 515 2-B6 1 1400 21.35( condenser I I RI closed cooler water P42-BB001B 2-D4 0 3160 O heat exchanger RI chiller condenser W/P 3 9 -Z:'0 0 lb 2-C6 1 1224 4.80
)
l 85 0.34 is W b Idq se 1 f- con t a ined V41-ACUO2 2-Al3 1 air conditioner f 110 0.34 steam tunnel cooler X7 3-BB0llB 2-C6 0 Tot al !Jcnessent ia l 7203 26.83 Total per Division 2 12012 35.23 0 9.2-46
(-. - _ _ _ - - - . - . _ _ . _ . - _ . _ _ _ I GESSAR II 22A7007 { 238 NUCLEAR ISLAND Rev. 0 9 ESSENTIAL SERVICE WATER (Continued) Table 9.2-1 DIVISION 1, TRAIN A, INITIAL S!!UTDOWN (20 hr)III Flow licat Fquipment Figure No. In (each) (each) l Name
~.
No. Location Use (gpm) (106Btu /h) Auxiliary building self- X73-ACUO3 1-G6 0 100 0 cortained air-conditioning } unit > Air positive seal computer X63-BB015 1-58 1 8 0 I room cooler Air positive seal computer W/CC001A 1-J9 1 20 0.12 i seal cooler Control B chiller W/P45-ZZ001A 1 - 11 6 1 540 2.16 condenser Diesel-generator heat W/R43-3001A 1-C2 0 2000 O II - exchanger H2 mix blower air, oil T41-CC008A 2 - 11 5 0 40 0 coole r , FPCCU pump room cooler X63-BB010A 1-J3 1 19 0.05 l Fuel pool cooling heat G41-B001A 1-H3 1 1750 8.0(2) l exchanger i LPCS pump room cooler X73-BB004 1-G8 0 190 0 j D17-J005 i Rad monitor panel 1-C11 1 0 0 , I I RCIC pump room cooler X73-BB003 1-H10 0 30 0.07 Remote shutdown air X73-ACUO5 1-H8 1 9 0.05 I conditioner I l RilR heat exchanger E12-B001A&C 1-B9 1 7300 61.0 ; B11R pump room cooler A X73-BB006 1-C10 1 150 0.55 RiiR pump seal cooler A W/E12-C002A 1-E10 1 20 0.1 SGTS room cooler A X63-BB002A 1 - 11 3 0 17 0 Shield annulus fan cooler X63-BB011A 1-J5 0 5 0 Fire protection NA Various (6) 0 0 i Fuel pool makeup 2" ESW127-ADC 1 - 11 1 (7) 0 0 Water positive seal makeup 1" ESW40-ADC 1-G4 (8) 0 0 j Total Essential 12198 72.11 ! Drywell chiller condenser W/P44-ZZ001A 1-C6 1 1224 4.80 , Radwaste evaporator C17-D555A 1-B6 1 1400 21.35 ' l condenser i RI closed cooling water P42-BB001A 1-D3 1 3160 23.51 heat exchanger RI chiller condenser W/P39-ZZ001A 1-D6 1 1224 4.80 l Steam tunnel cooler X73-BB0llA 1-C6 1 110 0.34 i Total Nonessential 7118 54.80 Total per Division 1 19316 126.91 I i
- 9.2-47 i
C ________________________________!
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 9.2-1 ESSENTIAL SERVICE WATER (Continued) DIVISION 2, TRAIN B, INITIAL SilUTDOWN (20 hr)II' l'l ow lic a t Equipment l'iqure No. In (each) (each) Name No. Location Use (qpm) (106 Btu /h) Auxiliary building self- X73/ACUO4 2-G13 1 100 0.4( contained air-conditioninq l unit 8 0 Air positive neal computer X73-BB017 2-Cl3 1 room cooler Air positive seal W /CC00 iB 2-1I13 1 20 0.12 corrp r e s s o r seal cooler I W/P45-ZZn01B 2-E13 0 540 O control B chiller condenser I ' Diesel-generator heat W/R43-5001B 2-B3 0 2000 O exchanger T41-CC008B 2-G5 0 40 0 II, mix blower air, oil choler X63-BD010B 2 -l! 3 0 19 0 PPCCU pump room cooler ' Fuel pool cooling heat G41-B001B 2 - 11 4 1 1750 8 exchanger D17-J006 2-C11 1 0 0 Radiation monitor panel idlR Pea t exchanger C12-It001B&D 2-B9 1 7300 61.0 RilR pump room cooler B X73-BB007 2-C9 1 150 0.56 X73-BB008 2-l!10 0 120 0 RilR pump room cooler W/E12-C002B 2-010 20 0.1 R!!R pump seal cooler B 1 2-G10 0 20 0 Pl!R pump neal cooler C W/E12-C002C X63-BB002B 2-G3 0 17 0 SGTS room cooler il 2-J3 0 5 0 Shield arnulus fan cooler X63-BBollB Various Various (6) 0 0 Pire protection 2-G2 (7) 0 0 ruel pool makeup 2" ESW130-ADC Total 1:snentia1 12109 70.18 IN W/P44-ZZ001B 2-D6 0 1224 O D rywe l l chiller condenser ' Radwaste evaporator G17-D555B 2-b6 1 1400 21.35 condenser I4I RI closed cooling water P42-BU001B 2-D4 0 3160 O heat exchanger W/P 39 -Z Z001B 2-C6 1 1224 4.80 l RI chiller condenser Radwaste builfing self- V41-ACUO2 2-A13 1 85 0.34 centaini"! air-conditioner X73-BB0llB 2-C6 1 110 0.34 Steam tunnel cooler Total Nonensential 7203 26.83 Total per Division 2 19312 97.01 l O l 9.2-48 j l l _ _ _ _ _ _ _ _
l \ i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 9.2-1
- ESSENTIAL SERVICE WATER (Continued) '
i Div1SION 1, TRAIN A, EXTENDED SilVTDOWN (+ 23 hr) ( ' l Flow Heat j Equipment Pigure No. In (each) (each) j g Name No. Location Une (gpm) (106Btu /h) I Auxiliary l>uilding self- X73-ACUO3 1-G6 0 100 O l contained air-conditioning unit ' Air positive seal computer X63-BB015 1-J8 1 8 0 room cooler Air positive seal W/CC001A 1-J9 1 20 0.12 9 compressor seal cooler l Control B chiller W/P45-ZZ001A 1 - 11 6 1 540 2.16 IN i ! condenser I Diesel-generator heat W/R43-5001A 1-C2 0 2000 O II j exchanger l II2 mix blower air, oil T41-CC008A 2-il5 0 40 0 ! cooler i I FPCCU pu.:1p room cooler X63-BB010A 1-J3 1 19 0.05 , Puel pool cooling heat G41-B001A 1 - 11 3 1 1750 8.0 (2) i exchanger i LPCS pump room coole r X73-BB004 1-G8 0 190 0 I Radiation monitor panel D17-J005 1-Cll 1 0 0 I
- cooler l RCIC purr.p t oom cooler X73-BB003 1-H10 0 30 0 i
Remote shutdown air- X73-ACUO5 1 - 11 8 1 9 0.05 conditioner R!lR heat exchanger E12-B001A&C 1-B9 1 7300 46.9 RHR pump room cooler A X73-BB006 1-C10 1 150 0.56 RHR pump seal cooler A W/C12-C002A 1-E10 1 20 0.1 I SGTS room cooler A X63-BB002A 1-H3 0 17 0 ! < i Shield mnulus fan cooler X63-BB011A 1-J5 0 5 0 ' l Fire protection NA Various (6) 0 0 , ) Puol pool makeup 2" ESW127-ADC 1-111 (7) 0 0 [ I wat< positive see.1 makeup 1" ESW40-ADC 1-G4 (8) 0 0 f
; Total Essential 12198 57.94 -
1 ! l Drywell chiller condenser W/P44-ZZ001A 1-C6 1 1224 1.0 I4I ' Radwaste evaporator G17-D555A 1-B6 1 1400 21.35 I condenser RI closed cooling water P42-BB001A 1-D3 1 3160 1.0 I4I neat exchanger RI chiller condenser W/P39-ZZ001A 1-D6 1 1224 4.80 5 Steam tunnel cooler X73-BB011A 1-C6 0 110 0 ! i Total Nonessential 7118 28.15 j G Total per Division 1 19316 86.09 t i i 9.2-49 l 1, Lac-- rr-. - - - - - - , - . - - . - , - - , . - - - ~ - . - , , - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --- - -=-
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i ! Table 9.2-1 ESSENTIAL SERVICE WATER- (Continued) DIVISION 2, 'I PA I N h, 1:XTI:NDI:D SilUTDOW'i (+ 20 hr)' I Flow heat 1:rj u i pren t Ftqure No. In (each) (each)
';ane No. Location Une (qpm).- (106 Btu /h) l Auxiliary building se l f - X7 3- ACUQ 4 2-Cl3 1 100 0.4(4) contained air-condit. toning unit
/s i r oonitive seat corputer ;73-Bh017 2-G13 0 8 0 room cooler I Air positive seal W/CC00lb 2-111 1 0 20 0 r cor pres sor neal cooler Control h chiller W/P45-ZZOOlh 2-1:13 0 540 0 condenner ,
Diesel-generator heat W/H4 3-500lb 2-83 0 2000 0 exchanoer 2-G5 40 0 112 mix blower air, oil T41-c00RB 0 ccoler I I CCU pur p t o ,n coo le r Y 6 3 -It h010 h 2 - 11 3 0 19 0 I Iuel pool c oo l i m, heat c. 4 ! -iM i h 2 - 11 4 1 1750 H exchc;qer Unliatson ronttor pane! D17 .P:06 2-C11 1 0 0 PiiP heat exchanger E12-Is00186D 2-83 1 7300 46.9 PHP p ur p r oon coo l e r h X73-hh007 2-C9 1 150 0.56 oaler C X 7 3-h:100 8 2-h10 0 120 0 l PHP purp roun t l PhD pump seal cooler b W / I. l .'-C0 0 2 h 2-!: 10 1 20 0.1 l P!IP pur.p seal cooler C .'/I: 12 -C0 0 2 C 2-G10 C 20 0 f SGTS roun cooler h X6 3-Bh002B 2-G3 0 17 0 X63-Bh0llh 2-J3 0 5 0 Shie ld ar.nulus fan croler V.s r t r >u s Vartous (6) 0 0 ftle protection 2 OSW 130- ADC 2 - ( '. 2 (7) 0 0 l't e l pool n.ikeup Total 1: aent i al 12109 55.96 Dryw"ll chiller condenaer W/P44-ZZ00lb 2-D6 0 1224 0( ' Padwante eeapo ra t i,r G17-D5'J,B 2-b6 1 1400 21.15 condenser PI c l ose-d ct ol i rol w. ster P 4 2 - 1:.h 0 0 l b 2-D4 G 1160 0 h e .i t e x ckmge r PI chiller condenser ' /P 39-ZZ00lb
< 2-C6 1 1224 4.80 Radwaste buildinq clt- /41-ACCO2 2-A13 1 H5 0.34 icntainmi a t r-et nd i t t one r S t e .irr t a r.ne l cc.oler X7 3-Isha llh 2-C6 0 110 0 Total
- one: sential 7'J3 26.49 Tot 1 per Divisten 2 19312 82.45 ,
O: 9.2-50
I l
.1 GESSAR II 22A7007 i l 238 NUCLEAR ISLAND Rev. O I
t 4 b I i
\
i i l Table 9.2-1 l 1 l j ESSENTIAL SERVICE WATER (Continued) l t i i ! DIVISION 1, TRA!!i A , POST-LOCA kid LOPP I l ! J Flow Heat ! Equipment Figure No. In (each) (each) !
] Name No. Location Use (gpm) (106Btu /h) {
Auxiliary building self- X73-ACUO3 1-G6 0 100 0 { contained air-conditioning i unit l Air positive seal computer X63-BB015 1-J9 1 8 0.16 I j room cooler l Air positive seal computer W/CC001A 1-J9 1 20 0.12 seal cooler Control B chiller I*' W/P45-ZZn01A 1-H6 1 540 2.16
- condenser 8
Diesel-generator heat W/R43-S001A 1-C2 1 2000 23.23 [ exchanger j H2 mix blower air, oil T41-CC008A 2-HS 2 40 0.2'#I cooler PPCCU pump room cooler X63-BB010A 1-J3 1 19 0.05 l ! Fuel pool cooling heat G41-B001A 1-H3 1 1750 8.0 I' ! exchanger ! i l LPCS pump room coole r X73-BB004 1-G8 1 190 0.48 i Radiation monitor panel D17-J005 1-C11 1 0 0 ! RCIC pump room cooler X73-BD003 1-H10 1 30 0.07 1 Remote shutdown air- X73-ACU05 1-H8 1, 9 0.05 coniitioner i t RHR heat exchanger E12-8001A&C 1-89 1 7300 130.0 RHR pump room cooler A X73-DB006 1-C10 1 150 0.56 RHR pump seal cooler A W/E12-C002A 1-E10 1 20 0.1 l SGTS room coole r A X63-B0002A 1-H3 1 17 0.03 ; Shield annulus fan cooler X63-BB011A 1-J5 1 5 0.01 Fire protection NA various (6) 0 0 ; Fuel pool makeup 2" ESW127-ADC 1-H1 (7) 0 0 Water positive seal makeup 1" ESW40-ADC 1-G4 (8) 0 0 Total Essertial 12198 165.22 Drywell chiller condenser W/P44-ZZ001A 1-C6 0 0 0 Radwaste evaporator G17-D555A 1-b6 0 0 0 condenser RI closed cooling water P42-BB001A l-D3 0 0 0 heat exchanger R1 chillet condenser W/P39-ZZ001A 1-D6 0 0 0 l Steam tunnel cooler X73-BB011A 1-C6 0 0 0 Total Nonessential 0 0 9.2-51
l l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 9.2-1 ESSENTIAL SERVICE WATER (Continued) DIVISION 2, 7 PAIN II , POST-LOCA AND LOPP ' Flow IIc a t Equiprent Figure No. In (each) (each) Name No. Location Use (gpm) (106 Btu /h) Auxiliary building self- X 7 3-/sCUO 4 2-G13 1 100 0.4(4} contained air-conditioning unit Al- positive seal computer X73-BB017 2-G13 0 8 0 room cooler l Air positive seal W/CC00lb 2-111 3 0 20 0 compresso r seal coole r control B chiller W/P45-ZZ0018 2-E13 6 540 O I4I condenser Diesel-gener ster heat W/R4 3-S00lb 2-B3 1 2000 23.23 exchan<1er I 11 , mix blower air, cil T11-CC008B 2-G3 0 40 O cooler T CCU pump room cooler x63-DB010B 2-83 0 19 0 luel pool cooling heat Gil-800lb 2 - 11 4 1 1750 8(2) exchanger l Padiation nonitor p.snel D17-J006 2-Cll 1 0 0 PilR heat exchanger 1:12 - 80 01 B l. D 2-B9 1 7300 130.0 Pit R pu.mp room cooler B X73-BB007 2-C9 1 150 0.56 PHR pump room cooler.C X7 3-bB00 8 2-1110 1 120 0.36 Pl'H pur-p sea l cooler B W/E12-C002B 2-E10 1 20 0.1 PHR purp neal cooler C W/E12-C002C 2-G10 1 20 0.1 SGTS roon cooler B X63-BB002B 2-G3 0 17 0 (4 Chield annulus fan cooler X63-BB0llB 2-J3 0 5 0 Fire protection Various Various (6) 0 0 ruel pool makeup 2" I:SW 13 0- ADC 2-G2 (7) 0 0 Total Essential 12109 162.75 Drywell chiller ccndenser W/P44-ZZO0lB 2-D6 0 0 0 Radwaste evaporator G17-D555B 2-B6 0 0 0 condenser RI closed cooling water P42-BE001B 2-D4 0 0 0 heat exchanger RI chiller condenser W/P39-ZZ001B 2-C6 0 0 0 Radwaste building self- V41-ACUO2 2-A13 0 0 0 contained air-conditicner Steam tunnel cooler X73-BB0llB 2-C6 0 0 0 i Total Nonessential 1 1 Total per Division 2 12109 162.75 1 O 9.2-52
.. _ _ _ .m _ _ _ . _ - _ . _ _ _ _ . _ _ . _ . _ ______ _ ___ _ .___ _ _ __ __ -
I !
- l l GESSAR II 22A7007 j 238 NUCLEAR ISLAND Rev. 0 4
1 l ! l Table 9.2-1 l ESSENTIAL SERVICE WATER (Continued) ' 1 i 6 DIVISION 1, TRAIN A, NONACCIDENT LOPP Flow Heat j Equipment Figure No. In (each) (each) i Name No. Location Use (gpm) (106Btu /h) ! j Auxiliary building self- X73-ACUO3 1-G6 0 100 O IN ! . contained air-conditioning l unit l Air positive seal computer X63-b8015 1-J8 1 8 0 ) room cooler l Air positive seal W/CC001A 1-J9 1 20 0.12 [ l compresser seal cooler Centrol B chiller W/P45-ZZn01A 1-H6 1 540 2.16( ' l condenser ; I Diesel-ger.erator heat w/R4 3-S001 A 1-C2 1 2000 23 23 ! exchanger i l
! !!2 mix blower air, oil T41-CC008A 2-HS 0 40 0
! cooler ;
!"PCCU purrp room cooler X63-BB010A 1-J3 1 19 0.05 !
- Fuel pool cooling heat G41-8001A 1-113 1 1750 8.0 I}
l exchanger ! LPCS pump room cooler X73-BB004 1-G8 0 190 0 Radiation monitor panel D17-J005 1-Cll 1 0 0 i RCIC pump room cooler X73-BB003 1-H10 1 30 0.07 ; Pemote shutdown X73-ACUO5 1-H8 1 9 0.05 l l air-conditioner t I Rl!R heat exchanger E12-B001A&C 1-B9 1 ~300
< 61 RHR pump room cooler A X73-BB006 1-C10 1 153 0.56 RilR pump seal cooler A W/E12-C002A 1-E10 1 20 0.1 SGTS room cooler A X63-BB002A 1 - 11 3 0 17 0 Shield annulus fan cooler X6 3-14B0llA 1-JS 0 5 0 Pire protection NA Various (6) 0 0 [
L ruel pool makeup 2" ESW127-ADC 1-111 (7) 0 0 l I Water positive seal makeup 1" ESW40-ADC 1-G4 (8) 0 0 Total Essential 12198 95.34 l Drywell chiller condenser W/P44-ZZ001A 1-C6 1 1224 4.8( Radwaste evaporator G17-D555A 1-B6 0 1400 0(93 condenser I RI closed cooling water P42-BB001A 1-D3 1 3160 23.51 I4I l heat exchanger l RI chiller condenser W/P39-3Z001A 1-D6 0 1224 0 l t 1-C6 110 Steam tunnel cooler X73-BB0llA 0 0 i Total Nonessential 7118 28.31 l Total per Division 1 19316 123.65 9.2-53 w'm - ---ssave_ e--res----- m- . - , - - -- - - - - - _ _ _ _ _ . - - - ------ ,__
GESSAR II 22A7007 ' 238 NUCLEAR ISLAND Rev. O Table 9.2-1 ESSENTIAL SERVICE WATER (Continued) DIVISION 2, TRAIN B, NONACCIDENT AND LOPP Plow lle a t Equipment Iigure No. In (each) (each) Name No. Location Use (qpm) (106 Btu /h) 100 I4I Auxiliary building self- X7 3- ACUO 4 2-G13 1 0.4 contained air-conditioning unit 2-G13 0 0 Air positive seal computer X7 3-BB017 8 room cooler Air positive sea 1 W/CC00H 2 ill3 0 20 0 compresser seal cooler I I Control B chiller W/P45-ZZ001B 2-013 0 540 O condenser Diese l-generator heat W/R43-S00lb 2-B3 1 2000 23.23 exchanger 11 , nix blowcr air, oil 'I41-CC008B Z-G5 0 40 0 cI_;o l e r X6 3-DB010 B 2-HJ 0 19 0 FI CCU pump room cooler luel pool cooling heat G41-B00lb 2 - 11 4 1 1750 8(2) exchanger D17-J006 2-Cll 1 0 0 Radiation ronitor panel E12-D001B&D 2-B9 1 7300 61.0 PBR heat exchanger X71-Bb007 2-C9 1 150 0.56 R!lR pump roon cooler B PJIR pump roon cooler C X7 3-BB00 8 2 -111 0 0 120 0 RilR puTp seal cooler B W /E 12 -C0 0 2 8 2-E10 1 20 0.01 FitP. pump seal cooler C W/Cl2-C002C 2-G10 0 20 0 X6 3-BBO'12 B 2-G3 0 17 0 SGTS roor cooler B X63-BB0llB 2-J3 0 5 0 Shield annulun fan cooler Fire irotection Various Various (6) 0 0 Puel pool nakeup 2" E SW 13 0 - AN? 2-G2 (7) 0 0 Total Essential 12109 93.29 Drywell chiller cendenser W/P44-ZZ001B 2-D6 0 1224 0' Padwanto evaporator G17-D555B 2-B6 0 1400 0 condenser PI closed cooling water P42-BB001B 2-D4 0 3160 0 heat exchanger RI chiller condenser W P 37-Z7"o l B 2-n- 0 1224 0 V41-ACIN2 2-Al? O H5 0 H a'tw as t o buildinq self-contained ai r-con li t ioner X 7 3-BB0118 2-C6 0 110 0 Steam tunnel cooler , _ - 7203 0 Total Nenescential 19312 93.29 Total per Division J 9 9.2-54 I i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i () Table 9.2-1 ESSENTIAL SERVICE WATER (Continued) j NOTES: ) (1) In all tables where service water flow is indicated, it should I be maintained continuously even though no heat load is ) indicated, with the exception of the Division 1 and 2 diesel 4 generators and radwaste evaporators under conditions where note (3) is applied. This will assure a balance of waterflow to the various equipment. (2) The heat duty shown is the maximum used for design. It will 3 be reduced as the decay heat of the stored fuel diminishes. (3) The service water flow is required only during operating of the Division 1 or Division 2 diesel generators. Testing of the diesel generators will be required periodically during ! normal plant operation. When operating the diesel-l generators, their heat load is as shown in Table 9.2-1 for LOPP. (4) The heat load may be on Division 1 or Division 2 but never on both simultaneously, since there are two 100% units with one () (5) on each division. The heat load shown is the maximum used for design. The waterflow and heat load may be reduced to zero by the opera-tor in either or both divisions whenever the evaporators are not in service. , (6) Division 1 and 2 will provide 150 gpm of ESW for fire pro-l tection purposes. The ESW for fire protection will be used intermittently for testing, and will not be counted in the ' total. (7) An intermittent flow of 40 gpm (equivalent to 2.3 gpm continuous) is required to supply makeup water to the fuel pool during the post-LOCA period, and will not be counted in ! the total. (8) An intermittent flow of 10 gpm (equivalent to 0.05 gpm con-tinuous) is required to supply makeup water to the water positive seal tank, and will not be counted in the total. l (9) The radwaste evaporator condensor is installed in Unit 1 of l a dual-unit plant only. !O 1 I j 9.2-55/9.2-56
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 /~'N Table 9.2-2 b DIVISION 3 IIPCS SERVICE WATER SYSTEM Item No. Name Equipment Number 1 IIPCS Pump Room Cooler X73-ECUO8 2 IIPCS Diesel Generator E22-S001 Plow i; 3at Item No. In (each) (each) Condition No. ___ Use (gpm) (106Btu /h) Normal operation 1 0 0 0 2 0 0 0 Initial shutdown (20 hrs) 1 0 0 0 2 0 0 0 Extended shutdown (+20 hrs) 1 0 0 0 2 0 0 0 Post-LOCA and LOPP 1 1 302 0.76 2 1 1020 8.55 Total 1322 9.31 Nonaccident LOPP 1 1 302 0.76 2 1 1020 8.55 Total 1322 9.31 0 9.2-57
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 9.2-3 ESSENTI AL SERVICE WATER SYSTEM ACTIVE FAILURE ANALYSIS Single Active Failure Analysis Failure of diesel generator to The other service water pumps start or failure of all power are powered and controlled from to a single Class lE power other buses which are energized system bus from other independent diesel generators and DC buses and, therefore, provide sufficient cooling for the essential equip-ment. The independent service water systems are mechanically and electrically separated to prevent damage to one system from another system. Failureof pump auto start Same analysis as above, signal Failure of ECCS Pump Room air Essential plant cooling require-cooler ments are met by the redundant ECC systems which have their own independently cooled pump rooms. Failure of a single service Essential plant cooling require-water pump during normal plar.t ments are met by the remaining operation operable, redundant ESW pump. ESSENTIAL SERVICE WATER SYSTEM PASSIVE FAILURE ANALYSIS Single Passive Failure Analysis Failure of any ESW System Essential plant cooling require-supply or return piping monts are met by the remaining intact ESW System, which includes its own independent supply and raturn service water headers. The redundant systems are mechanically and electrical-ly separated to prevent damage to one system from another system. Failure of ESW to RHR heat Essential plant cooling require-exchanger ments are met by the remaining intact redundant RHR System, which includes its own 100% capacity heat exchanger. Failure of service water piping Essential plant cooling require-to or from the air cooler for ments are met by the redundant an ECCS pump room ECC systems which have their own independently cooled pump rooms. 9.2-58
O O O Table 9.2-4 CLOSED COOLING WATER SYSTEM COMPONENT Heat Design Design Flow Exchanger Tank Pressure Temperature Capacity Duty Volume Component (psig) (OF) (gpa) (Btu /hr) (gal.) CCW Heat Exchanger (2) 150 Shell 212 Shell 1953 Shell Each -
& Tube & Tube 3160 Tube Each 23,600,000 CCW Surge Tar.k 150 212 - -
330 RWCU System Nonregenerative Heat Exchanger (2) 150/1410 370/575 875 17,500,000 - M Reactor Recirc Pump N Seals and Motor Bearings 150 212 190 1,575,000 - ya w . OM e Reactor Recirc Pump pm ) Motor Windings 150 212 550 2,750,000 - xx w RWCU Cleanup Pump gg i Seals and Bearings 150 212 50 220,000 - ms CRD Pump Seals and Bearings 150 212 100 40,000 - h Drywell Equipment Drain Sump 150 212 100 500,000 - Containment Equipment Drain Sump 150 212 40 300,000 - Concentrated Waste Tank Coil 150 212 25 80,000 - Reactor Water Sample Cooler 150/420 212 25 80,000 - Concentrated Waste Tank Vent Cooler 150 212 5 100,000 - Detergent Evaporator Sample Cooler 150 275 20 150,000 -
,N O Da
. O O
O4
O O O Table 9.2-5 DEMINERALIZED WATER DISTRIBUTION SYSTEM USER REQUIREMENTS Minimum Pressure Service Item At User Floor Building Equipment or Service Number (psig) gpm* Elevation Containment RWCU sample station G33-ZO20 25 5 48 ft 7 in. Refuel equipment service - 67 50(3) 84 ft 7 in. box Type III Auxiliary RI chilled water P39-AA001 25 40(1) 28 ft 6 in. Expansion tank U Drywell chilled water P44-AA001A 25 40(1) 28 ft 6 in. m Expansion tank z ' Drywell chilled water P44-AA001B 25 40(1) 28 ft 6 in. 8@
. Expansion tank gm y Division 1 Battery Room -
25 6 11 ft 0 in. >> m Division 2 Battery Room - 25 6 11 ft 0 in. W5 Division 4 Battery Room - 25 6 11 ft 0 in. HH Nondivision Battery Room RHR System flush Radiation monitor purge 25 70 6 150(3) 11 ft 0 in. (-) 11 ft 9 in. f" 2 D17-J006 40 2 (-) 32 ft 0 in. O Radiation monitor purge D17-J005 40 2 (-) 32 ft 0 in. Radiation monitor purge D17-J009 40 2 (-) 32 f t 0 in. CRD maintenance area test - 200 5(4) (-) 6 ft 10 in. water CRD maintenance area flush - 25 25 (-) 6 ft 10 in. water Fuel CCW expansion tank P42-AA001 25 40(1) 101 ft 0 in. (3) (4) Radiation monitor purge D17Z-JJ001 40 2 (-) 32 ft 0 in. Radiation monitor purge D17-J008 40 2 11 ft 0 in. mU Refuel equipment box I - 67 50(3) 11 ft 0 in. c>y Refuel equipment box IV - 67 50(3) 11 ft 0 in. - o Cask Decontamination System - 25 20 (-) 32 ft 0 in. o
Table 9.2-5 DEMINERALIZED WATER DISTRIBUTION SYSTEM USER REQUIREMENTS (Continued) Minimum Pressure Service Item At User Floor Building Equipment or Service Number (psig) gpm* Elevation Division I Engine jacket water R43-S001 40 40(1) (-) 6 ft 10 in. Diesel Generator Division 2&3 Engine jacket water R43-S002 40 40(1) (-) 6 ft 10 in. Diesel- Engine jacket water E22-S001 40 40(1) " (-) 6 ft 10 in. Generator Division 3 Battery Room - 40 6 (-) 6 ft 10 in. m
=
, Radwaste Flush water to radiation Gl?-J001 50 1 (-)27 ft 10 in. @@ monitor sample &m Line flush RD 128 - 50 100 (-)27 ft 10 in. $$ W% $ Radiation monitor sample G17-J002 50 1 (-)27 ft 10 in. purge HH Sample station G17-Z200 25 5 (-) 6 ft 10 in. Excess water tank at BOP - 40 20 (-) 9 ft 0 in. k Mix / fill station washdown - 40 10 (-) 6 ft 10 in. 6 spray Safety shower - 50 30 (-)27 ft 10 in. General service - 30 3 32 ft 2 in. Radiation monitor purge D17-J007 40 2 (-)27 ft 10 in.
- Flow rates are maximum based on intermittent operation.
w MN O>
<4
. O O
04 9 9 O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Table 9.2-5 [V) DEMINERALIZED WATER DISTRIBUTION SYSTEM USER REQUIREMENTS (Continued) NOTES (1) Flow rate shown is for initial fill only. Normal flow is minimal. (2) Provide 1-in. connection. (3) Water supply from BOP demineralizer water transfer pump by bypassing the BOP demineralizer water head tank. (CCW expansion tank initial fill, etc.) (4) Water supply from Nuclear Island domineralizer water booster pump discharge. (CCW expansion tank normal makeup, etc.) (5) Flows given are per unit for single-unit plant. For duwl units, subtract radwaste services for second unit. O l 9.2-63/9.2-64
o O O Table 9.2-6 CONDENSATE DISTRIBUTION SYSTEM FLOW REQUIREMENTS Reference Pressure Building Equipment or Service Item No. P&ID (psig) gpm(3) Reactor Precoat tank fill G36-A001 K-ll3B 80 40 Resin feed tank fill G36-A002 K-ll3B 80 25 Backwash receiving tank water G36-A003 K-ll3A 80 100 spray Precoat pump seal flush G36-C002 K-113B 80 1 Resin metering pump flush G36-C003 K-ll3B 80 3 Drain pump flush G36-CC004 K-ll3A 80 25 RWCU filter demineralizer, G36-D001 K-ll3A 80 560 w backwash and fill m RWCU filter demineralizer, G36-D002 K-ll3B 80 560 z backwash and fill cc
? RB transfer pool makeup -
K-ll5C 25 150 OE y m Drywell lower level equipment B33-C001A&B K-124A 35 25 each sy decontamination W% Fire Protection - K-124A 65 500 ss Line flushes - - 45 200 each "" SLC storage tank C41-A001 K-105 45 45 h SLC test tank and pump suction C41-A002 K-105 45 45 O Reactor well wall Wetter System - K-124B 25 90 Refueling equisment - - 66 50 connections ( ) General service (for each - - 25 (1) outlet) Suppression pool makeup - K-152 25 40 System A containment spray - K-107B 45 150 flush System B containment spray - K-107C 45 150 flush CRD scram header flush - K-104B 80 25 N CRD scram header flush - K-104B 80 25 $$
<w
*8 ou
Table 9.2-6 CONDENSATE DISTRIBUTION SYSTEM FLOW REQUIRE.iENTS (Continued) Reference Pressure Building Equipment or Service Item No. P&ID (psig) gpm(3) Reactor CRD removal area - K-124A 40 10 (Cont'd) DRW sump - K-124A 40 10 CRW sump - K-124A 40 10 Auxiliary HPCS pump E22-C001 K-109A 45 7800 RCIC pump E51-C001 K-110A 45 700 RHR System A supply flush E12-C002A K-107B 45 200 RHR System A, B&C return flush E12-C002A, K-107B 45 200 w B&C
- RHR System B supply flush E12-C002B K-107B 45 200 =
e RHR System C supply flush E12-C002C K-107B 45 200 $$ RCIC head spray line flush - K-107B 45 200 gg Line flushes - K-108A,109B 45 200 each >> g Positive Seal System supply - K-157 45 10 Fuel CKU pump supply CRD pump return Cll-C001 Cll-C001 K-104A K-104A 25 200 20 fz Fuel storage pool makeup - K-115A 45 150 C Cask pool fill - K-ll5A 45 150 Cask pool water pumpout G41-C001A K-ll5B - 1100(2) Cask pool water pumpout G41-C001B K-115B - 1100(2) Precoat tank fill G46-A001 K-ll6B 80 150 Resin tank fill G46-A002 K-ll6B 80 25 Backwash receiving tank water G46-A003 K-116A 80 135 spray Resin metering pump flusn G46-C003 K-ll6B 80 5 Drain pump flush G46-CC004 K-ll6A 80 50 Backwash tank drain pump seal G46-CC004 K-ll6A 45 5 water xw FPCC filter /demineralizer G46-D001 K-ll6A 80 1650 Q> y backwash and fill g 04 O O O
O O O Table 9.2-6 CONDENSATE DISTRIBUTION SYSTEM FLOW REQUIREMENTS (Continued) Reference Pressure Building Equipment or Service Item No. P&ID (psig) gpm(3) Fuel FPCC filter /demineralizer G46-D002 K-ll6B 80 1650 (Cont'd) backwash and fill Strainer backwash - K-ll6A&B 80 40 Cask decontamination - K-154 40 40 Refueling equipment connections Type I, III, - 66 50 IV Type II - 66 20 General service (for each - - 25 (1) w outlet) $ z Radwaste Radwaste precoat tank fill G17-A275 K-ll7D 50 80 co ? Body feed tank fill G17-A305 K-117E 50 10 Condensate tank fill G17-A577 K-ll7H 50 20 yy $ Concentrated waste tank spray G17-A700 K-ll7J 50 16 xx Caustic tank makeup G17-A730 K-ll7J 50 10 ss Chemical add tank makeup G17-A755A K-ll7J 50 10 mH Chemical add tank makeup G17-A755B K-ll7J 50 10 h Chemical add tank makeup G17-A755C K-117J 50 10 6 High conductivity oil separator G17-D015 K-ll7A 50 35 line flush Low conductivity oil separator G17-D035 K-ll7A 50 35 line flush Cleanup phase separator fill G17-D061A K-117B 50 200 Cleanup phase separator fill G17-D061B K-ll7B 50 200 Waste filter decon spray and G17-D260A K-ll7D 100 50 belt wash Waste filter decon spray and G17-D260B K-ll7E 100 50 belt wash Waste demineralizer resin G17-D326A K-ll7F 50 200 w sluicing my Waste demineralizer resin G17-D326B K-ll7F 50 200 <g sluicing o
Table 9.2-6 CONDENSATE DISTRIBUTION SYSTEM FLOW REQUIREMENTS (Continued) Reference Pressure Building Equipment or Service Item No. P&ID (psig) gpm(3) Radwaste Waste evaporator spray G17-D510A K-ll 711 50 20 (Cont'd) Waste evaporator spray G17-D510B K-117I 50 20 Screw conveyor flush G17-D738 K-ll7J 100 50 Distillate domineralizer resin G17-D790 K-ll7K 50 200 sluicing Detergent evaporator spray G17-D900 K-ll7L 50 10 Sample line flushes - K-ll7D,E 50 5 each connection w Pipe line flushes - K-ll7B,C, 50 200 each $ J,K connection , Instrument line flushes - K-ll/B,H, 5 5 each Eo ' J,L connection @$ N o Radiation monitor purge Makeup tank G17-A764 I-200 K-ll7L 50 50 2 yy 1 xx Iligh conductivity tanks G17 -A 4 41 A,B , C K-ll7G 50 180 gs Mix / fill station G17-D739 K-ll7J 100 25 mH Pump seal water G17-C061 K-117B 50 2 $ Pump seal water G17-C086 K-117B 50 2 $ Pump seal water G17-C213A&B K-ll7C 50 4 Pump seal water G17-C186 K-ll7C 50 4 Pump seal water G17-C275A&B K-ll7D 50 4 Pump seal water G17-C441A&B K-ll7G 50 4 Pump seal water G17-C461A&B K-ll7G 50 4 Pumo seal water G17-C522A K-ll7H 50 4 Pump seal water G17-522B K-ll7I 50 4 Pump seal water G17-C700 K-ll7J 50 2 NOTES: w (1) One-in. connections, yy (2) Output of pumps G41-C001A&B corbine for a total flow of 2200 gpm. <w (3) Flows are for single unit. Por dual units, subtract radwaste flows from second unit. *$ ow O O
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O t} Table 9.2-7 l DRYWELL CilILLED WATER SYSTEM COMPONENT DESCRIPTION Drywell Chillers Type Centrifugal hermetic Quantity 2 (100% capacity units) Capacity (tons, refrigeration)
- Chilled water flow (gpm) 790 each Supply temperature (OF) 60 Condenser water flow (gpm) 750 each Supply temperature (OF) 100 Cooling media Freon Control Inlet guide vane Motor horsepower (hp)
- Condenser Shell and tube Evaporator Shell and tube
() V Drywell Chilled Water Pumps Quantity 2 (100% capacity units) Type Centrifugal, horizontal Capacity (gpm) 790 Motor horsepower (hp)
- Drywell Condenser Booster Pumps Quantity 2 (100% capacity units)
Type Centrifugal, horizontal Capacity (gpm) 750 Motor horsepower (hp) *
- Vendor Data By ipplicant
'w g
a 9.2-69
GESSAR II 22A7007 238 11UCLEAR ISLAND Rev. O Table 9.2-8 CONTROL BUILDING CIIILLED WATER SYSTEM COMPOMENT DESCRIPTION Control Buildina Chill _ erg Type Centrifugal hermetic Quantity 2 (100% capacity units) Capacity (tons, refrigeration) Chilled water flow (gpm) 380 Supply temperature (UF) 45 Condenser water flow (gpm) 370 Supply temperature (OF) 100 Cooling media Freon Control Inlet guide vane Motor horsepower (hp) Condenser Shell and tube Evaporator Shell and tube Control Building Chilled Water Pumps Quantity 2 (1001 capacity units) Type Centrifugal, horizontal Capacity (gpm) 380 Motor horsepower (hp) Control Building Condenser Water Pumps Quantity 2 (100% capacity units) Type Centrifugal, horizontal Capacity (gpm) 370 Motor horsepower (hp)
- Vendor Data By Applicant O
9.2-70
. _ _ . . = - _ - _ . -- --. _ -_ - .. - - _ . _ . _ - - - .. - - . . -
i I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () Table 9.2-9 REACTOR ISLAND NONESSENTIAL CHILLED WATER SYSTEM COMPONENT DESCRIPTION Nonessential Chillers Type Centrifugal hermetic Quantity 2 Capacity (tons, refrigeration)
- Chilled water flow (gpm) 880 each j Supply temperature (OF) 45 Condenser water flow (gpm) 825 each j Supply temperature (OF) 100
- Cooling media Freon Control Inlet guide vane Motor horsepower (hp)
- I Condenser Shell and tube
, Evaporator Shell and tube i, , Nonessential Chilled Water Pumps Quantity 3 (50% capacity units) ] Type Centrifugal, horizontal Capacity (gpm) 030 each Motor horsepower (hp)
- j Nonessential Condenser Water Pumps 4
Quantity 2 (1 per chiller) Type Centrifugal, horizontal capacity (gpm) 825 each Motor horsepower (hp) *
- Vendor Data By Owner Applicant O
l j 9.2-71/9.2-72
(G) U w.) Table 9.2-10 HEATED WATER SYSTEM HEAT LOAD AND FLOW REQUIREMENTS Heat Load Component Item No. gpm (Btu /hr) Containment pressure control supply air heating T41-BB002A 16.2 324,000(4) coil Containment pressure control supply air heating T41-BB002B(1) 16.2 324,000(4) coil Containment high purge heating coil T41-BB004 108 2,160,000(4) Radwaste outdoor air preheat coil V41-BB001 160 3,201,000 w Radwaste zone heating coil V41-BB003A 15 293,000
- Radwaste zone heating coil V41-BB003B 10 187,000 5o om P
M Radwaste zone heating coil V41-BB003C 17 333,000 yy Radwaste zone heating coil V41-BB003D 11 f 207,000 mx Radwaste Control Room heating coil V41-ACU02 29.4 586,000 (( Radwaste Mechanical Equipment Room unit heater Radwaste Mechanical Equipment Room unit heater V41-BB006A V41-BB006B 3.6 3.6 70,000 70,000 f 0 Radwaste Mechanical Equipment Room unit heater V41-BB006C 3.6 70,000 Radwaste Mechanical Equipment Room unit heater V41-BB006D 3.6 7^,000 Fuel Building air conditioning unit heating coil X63-BB003 40 800,000 Fuel Building air conditioning unit heating coil X63-BB005 40 800,000 Auxiliary Building pressure control supply air X73-BB001A 17.5 350,000 heating coil Auxiliary Building pressure control supply air X7 3-BB001B (1) 17.5 350,000 heating coil ' x" o> Auxiliary Building Mechanical Equipment Room unit X73-BB005A 4.7 93,000 <
,g heater o o .a
Table 9.2-10
. HEATED WATER SYSTEM HEAT LOAD AND FLOU REQUIREMENTS (Continued)
Heat Load Component Item No. gpm (Btu /hr) Auxiliary Building Mechanical Equipment Room unit X73-BB005B 4.7 93,000 heater Auxiliary Building Mechanical Equipment Room unit X73-BB005C 4.7 93,000 heater Auxiliary Building Mechanical Equipment Room unit X73-BB005D 4.7 93,000 heater Auxiliary Building area preheat coil X'3-BB013A 11.6 200,230 Auxiliary Building area preheat coil X73-BB013B 23.2 465,960 gg y Fuel Building area reheat coil X73-BB014 8.1 162,310 @$ N mm Fuel Building area reheat coil X73-BB015 3 60,480 >> q mz A Fuel Building Mechanical Equipment Room unit X63-BB0013 5 100,000 -~ heater m~ Fuel Building Mechanical Equipment Room unit X63-BB0014 5 100,000 b z heater C Fuel Building zone heating coil X63-BB007 18 359,000 Fuel Building zone heating coil X63-BB008 13 260,000 Fuel Building zone heating coil X63-BB009 5.7 113,400 Radwaste truck loading area unit heater V41-BB009A 5 95,500 Radwaste truck loading area unit heater V41-BB009B 5 95,500 Radwaste truck loading area unit heater V41-BB009C 2.3 45,000 Radwaste Mechanical Equipment Room unit heater V41-BB006E 3.6 70,000 w zw Radwaste Mechanical Equipment Room unit heater V41-BB006F 3.6 70,000 @>y
- O 643.l(3) 11,785,520(4) o O O O
; GESSAR II 22A7007
- i. 238 NUCLEAR ISLAND Rev. 0 i
i w i Table 9.2-10 l IIEATED WATER SYSTEM llEAT LOAD AND l FLOW REQUIREMENTS (Continued) ( l ' ?!OTES { (1) Standby component. l (2) Flow passes through all the coils. ; , (3) Maximum heat load based on containment high purge mode. (4) fleat load only as applied to either containment high purge i mode or normal containment pressure control mode. (5) Above totals for single-unit plants. For dual-unit plants, subtract Radwaste components for second unit. 1 l, . l 1 .e nv l 1
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