ML20049H264
| ML20049H264 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 02/12/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 8202230011 | |
| Download: ML20049H264 (600) | |
Text
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l GESSAk II 22A7007 238 NUCLEAR ISLAND Rev. 0 l
CIIAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT O
8202230011 820212 PDR ADOCK 05000447 K PDR l -- _ _ _ _ _ . _ _ _ _ . _ . _ _ _ , _ _ _ _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rov. O SECTION 1.1 O
CONTENTS Section Title Page
1.1 INTRODUCTION
1.1-1 1.1.1 Type of License Required 1.1-1 3.1.2 Identification of Applicant 1.1-1 1.1.3 Number of Plant Units 1.1-2 1.1.4 Description of Location 1.1-2 1.1.5 Type of Nuclear Steam Supply 1.1-3 1.1.6 Type of Containment -
1.1-3 1.1.7 Core Thermal Power Levels 1.1-3 1.1.8 Scheduled Completion and Operation Dates 1.1-3 ILLUSTRATIONS O Figure Title Page 1.1-1 Mark III Nomenclature 1.1-5 1.1-2 Heat Balance at Rated Power 1.1-6 i
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1.1-i/1.1-ii
- GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l
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( 'l. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT'
)
i
1.1 INTRODUCTION
i i 1.1.1 Type of License Required 4
This General Electric Standard Safety Analysis Report (GESSAR) is submitted in support of the application for a construction permit and facility operating license for the Nuclear Island. portion of a nuclear powered electric generating plant. The Nuclear Island (sometimes referred to as Reactor Island) consists of all buildings which are dedicated exclusively or primarily to housing systems and equipment related to the nuclear system. Under the concept presented herein, there are seven such buildings that comprise the Nuclear Island. These are:
4 (1) Reactor Building (including shield building and
() containment) ;
(2) Fuel Building; (3) Auxiliary Building; (4) Diesel Generator Buildings; (5) Control Building; and
! (6) Radwaste Building. ,
1 The only major system related to the nuclear system that is not housed in one of the seven buildings is the Offgas System which is more appropriately housed in the turbine building since it is
- physically associated with the condenser air ejectors.
/
4 1.1-1 i
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.1.1 Type of License Required (Continued)
For each of the buildings, the General Electric Company (GE) scope of responsibility includes the design of all structures including the foundation mats and everything within these structures.
System boundaries may vary with the nature of the interface, but the general rule for determining interfaces is that boundaries extend to just outside of the building walls. A major factor in the design process is the determination of the exact description of the interface. Parameters such as dimensions and orientation of the pipes, type of connections, and pressures at the interface points are established and identified in Section 1.9. It is expected that the Applicant supply will conform to these estab-lished interfaces.
The GESSAR is written in accordance with Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Revision 3, November 1978.
Where required, information is plant unique; typical information has been included to enhance the reviewer's understanding of the text. The plant-unique information is enclosed in boxes on the appropriate pages.
1.1.2 Identification of Applicant l
Applicant will supply.
l 1.1.3 Number of Plant Units 1
For the purposes ot GESSAR II, only a single standard plant will be considered.
1.1.4 Description of Location l Applicant will supply.
i 1.1-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.1.5 Type of Nuclear Steam Supply
)
This plant will have a boiling water reactor nuclear steam supply system designed and supplied by the General Electric Company and designated as BWR/6.
1.1.6 Type of Containment This plant will have a low-leakage containment vessel completely surrounding a drywell and a pressure suppression pool. The con-tainment vessel is a cylindrical steel structure with an ellip-soidal dome and flat bottom supported by a reinforced concrete a
foundation mat. The containment system is designated as a Mark III containment. Standard Mark III nomenclature is shown in Fig-ure 1.1-1.
1.1.7 Core Thermal Power Levels The information presented in this GESSAR pertains to one reactor unit with a rated power level of 3579 MWt and design power level of 3730 MWt. The station utilizes a single-cycle forced-circulation boiling water reactor (BWR) provided by GE. The heat balance for rated power is shown in Figure 1.1-2. The station is designed to operate at a gross electrical power output of approxi-mately 1269 MWe and a net electrical power output of approximately 1220 MWe.
1.1.8 Scheduled Completion and Operation Dates l
l Applicant will supply.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONSISTENT WITH 1967 ASME STE AM TA8LES O
ASSUMED SYSTEM LOSSES LEGEND THERMAL 1.1 M W
- = F LOW (ib/hr)
F = TEMPERATURE ( F) 1040 P h = ENTHALPY (8tu/lb)
M= % MOISTURE P = PRESSURE (psia) --
MAIN STEAM FLOV' 15.402,000 #
%= ISOLATION VALVES
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ISOLATION 1190.8 h VALVES 98 MAIN FEED FLOW 15,526,000 m 15,372,000 #
31,700.M # 3579 MWt 420 F,397.8 h jL 420 0 F 397.6 h 534 F,428.7 h 1 f
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hin. - 52 7.7 436 F 415 3 h CORE THERMAL POWER 3579.0 MW PUMP HE ATING + 10. 7 RE ACTOR WATER CLE ANUP DEMIN CLE AN UP
- 5.1 I i SYSTEM LOSSES 0 OTHER SYSTEM LOSSES -1.1 )t TURBINE CYCLE USE 3583.5 MWt ROD DRIVE 30,000 #
80 F 154.000 #
FE ED F LOW 48 h 533 F 527.5 h L 1 FROM CONDENSATE STORAGE TANK
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O Figure 1.1-2. Heat Balance at Rated Power l
l 1.1-6 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 SECTION 1.2 CONTENTS Section Title Page 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 Principal Design Criteria 1.2-1 1.2.1.1 General Design Criteria 1.2-1 1.2.1.1.1 Power Generation Design Criteria 1.2-1 8
1.2.1.1.2 Safety Design Criteria 1.2-2 1.2.1.2 System Criteria 1.2-7 1.2.1.2.1 Nuclear System Criteria 1.2-8 1.2.1.2.2 Power Conversion Systems Criteria 1.2-9 1.2.1.2.3 Electrical Power Systems Criteria 1.2-10 1.2.1.2.4 Radwaste System Criteria 1.2-10 1.2.1.2.5 Auxiliary Systems Criteria 1.2-10
! 1.2.1.2.6 Shielding and Access Control Criteria 1.2-11
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h 1.2.1.2.7 Nuclear Safety Systems and Engineered Safety Features Criteria 1.2-11 1.2.1.2.8 Process Control Systems Criteria 1.2-12 1.2.1.2.8.1 Nuclear System Process Control Criteria 1.2-12 1.2.1.2.8.2 Power Conversion Systems Process i Control Criteria 1.2-13 1.2.1.2.8.3 Electrical Power System Process Control Criteria 1.2-14 1.2.1.2.9 Other Plant Design Criteria 1.2-14 l
1.2.2 Plant Description 1.2-15
, 1.2.2.1 Site Characteristics 1.2-15 1.2.2.1.1 Site Location 1.2-15 1.2.2.1.2 Description of Plant Environs 1.2-15 1.2.2.1.2.1 Meteorology 1.2-15 1.2.2.1.2.2 Hydrology 1.2-15 1.2.2.1.2.3 Geology and Seismology 1.2-16 1.2.2.2 General Arrangement of Structures
! Equipment 1.2-16 (O ,j 1.2.2.3 Nuclear Systems 1.2-17
! 1.2-i
GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued)
Section Title Page 1.2.2.3.1 Reactor Core and Control Rods 1.2-17 1.2.2.3.2 Reactor Vessel and Internals 1.2-18 1.2.2.3.3 Reactor Recirculation System 1.2-18 1.2.2.3.4 Residual Heat Removal System 1.2-19 1.2.2.3.5 Reactor Water Cleanup System 1.2-20 1.2.2.3.6 Nuclear Leak Detection System 1.2-20 1.2.2.4 Nuclear Safety Systems and Engineered Safety Features 1.2-20 1.2.2.4.1 Reactor Protection System 1.2-20 1.2.2.4.2 Neutron Monitoring System 1.2-21 1.2.2.4.3 Control Rod Drive System 1.2-21 1.2.2.4.4 Control Rod Drive Housing Supports 1,2-22 1.2.2.4.5 Control Rod Velocity Limiter 1.2-22 1.2.2.4.6 Nuclear System Pressure Relief System 1.2-22 1.2.2.4.7 Reactor Core Isolation Cooling System 1.2-22 1.2.2.4.8 Emergency Core Cooling Systems 1.2-22 1.2.2.4.9 Containment 1.2-25 1.2.2.4.9.1 Functional Design 1.2-25 1.2.2.4.9.2 Heat Removal 1.2-26 1.2.2.4.9.3 Environmental Systems 1.2-26 1.2.2.4.9.4 Containment Spray 1.2-26 1.2.2.4.9.5 Combustible Gas Control 1.2-26 1.2.2.4.10 Containment and Reactor Vessel Isolation Control System 1.2-26 1.2.2.4.11 Main Steamline Isolation Valves 1,2-27 1.2.2.4.12 Main Steamline Flow Restrictors 1.2-28 1.2.2.4.13 Main Steamline Radiation Monitoring System 1.2-28 1.2.2.4.14 Residual Heat Removal System (Containment Cooling) 1.2-28 1.2.2.4.15 Ventilation Exhaust Radiation Monitoring System 1.2-28 1.2.2.4.16 Standby Gas Treatment System 1.2-29 1.2.2.4.17 Auxiliary Building and Fuel Building Isolation Control System 1.2-29 1.2-ii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 CONTENTS (Continued)
Section Title Page
(
-1.2.2.4.18 Standby AC Power Supply 1.2-29 1.2.2.4.19 DC Power Supply 1.2-30 1.2.2.4.20 Standby Liquid Control System 1.2-30
- 1.2.2.4.21 Safe Shutdown from-Outside the Control Room 1.2-30
- 1.2.2.4.22 Main Steam Positiv,e Seal System 1.2-31 1.2.2.5 Power Conversion System 1.2-31 i
1.2.2.5.1 Turbine-Generator 1.2-31 1.2.2.5.2 Main Steam System 1.2-31 i 1.2.2.5.3 Main Condenser 1.2-31 1.2.2.5.4 Main Condenser Evacuation System 1.2-31 1.2.2.5.5 Turbine Gland Sealing System 1.2-31 1.2.2.5.6 Steam Bypass System and Pressure Control System 1.2-32
. 1.2.2.5.7 Circulating Water System 1.2-32 I 1.2.2.5.8 Condensate Storage Facilities 1.2-32
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1.2.2.5.9 Condensate and Feedwater System 1.2-32
,1 1.2.2.6 Electric Power Systems and Instrumentation and Control Systems 1.2-32 1.2.2.6.1 Electric Power Systems 1.2-32 j 1.2.2.6.2 Electrical Power System Process Control and Instrumentation 1.2-32 1.2.2.6.3 Nuclear System Process Control j and Instrumentation 1.2-33 2
1.2.2.6.3.1 Rod Control and Information System 1.2-33 1.2.2.6.3.2 Recirculation Flow Control System 1.2-33 1.2.2.6.3.3 Neutron Monitoring System 1.2-33 1.2.2.6.3.4 Refueling Interlocks 1.2-34 l 1.2.2.6.3.5 Reactor Vessel Instrumentation 1.2-34 1.2.2.6.3.6 Process Computer System 1.2-34 1.2.2.6.4 Power Conversion Systems Process
- Control and Instrumentation 1.2-35 1.2.2.6.4.1 Pressure Regulator and Turbine Generator Control 1.2-35 1.2-iii i
i
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued)
Section Title Page 1.2.2.6.4.2 Feedwater Control System 1.2-35 1.2.2.7 Fuel Handling and Storage Systems 1.2-35 1.2.2.7.1 New and Spent Fuel Storage 1.2-1.2.2.7.2 Fuel Handling System l.2-36 1.2.2.8 Cooling Water and Auxiliary Systems 1.2-36 1.2.2.8.1 Closed Cooling Water System 1.2-36 1.2.2.8.2 Fuel Pool Cooling and Cleanup Systems 1.2-37 1.2.2.8.3 Essential Service Water System 1.2-37 1.2.2.8.4 Ultimate Heat Sink 1.2-37 1.2.2.8.5 Condensate Storage and Transfer System 1.2-38 1.2.2.8.6 Plant Raw Water Treatment and Makeup Water Treatment System 1.2-38 1.2.2.8.7 Potable and Sanitary Waste Water System 1.2-38 1.2.2.8.8 Plant Chilled-Water System 1.2-38 1.2.2.3.9 Process Sampling Systems 1.2-39 1.2.2.8.10 Plant Equipment and Floor Drainage 1.2-39 1.2.2.8.11 Service and Instrument Air Systems 1.2-39 1.2.2.8.12 Normal Auxiliary AC Power 1.2-40 1.2.2.8.13 Diesel Generator Fuel-Oil Storage and Transfer System 1.2-40 1.2.2.8.14 Auxiliary Steam System 1.2-40 1.2.2.8.15 Heating, Ventilating, and Air Conditioning (Environmental) Systems 1.2-40 1.2.2.8.16 Lighting Systems 1.2-42 1.2.2.8.17 Fire Protection System 1.2-43 1.2.2.9 Radioactive Waste Systems 1.2-43 1.2.2.9.1 Gaseous Radwaste System 1.2-43 1.2.2.9.2 Liquid Radwaste 1.2-44 1.2.2.9.3 Solid Radwaste 1.2-44 1.2.2.10 Radintion Monitoring and Control 1.2-45 1.2.2.10.1 Process Radiation Monitoring 1.2-45 1
1.2-iv
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i
i .
1 i
CONTENTS (Continued)
^
\ V Section Title Page
- 1.2.2.10.2 Area Radiation ~ Monitors 1.2-45 1.2.2.10.3 Site Environs Radiation Monitors 1.2-46 l 1.2.2.11 Shielding
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 SECTION 1.2
( ILLUSTRATIONS Figure Title Page 1.2-1 Site Plan 1.2-47 1.2-2 Reactor,. Auxiliary & Fuel Building Arrangement Plan at El (-) 32'-0" 1.2-49 1,2-3 Reactor, Auxiliary & Fuel Building Arrangement Plan at El (-) 6'10" 1.2-50 1.2-4 Reactor, Auxiliary & Fuel Building Arrangement Plan at El 11'-0" 1.2-51 1.2-5 Reactor, Auxiliary & Fuel Building Arrangement Plan at El 28'-6" 1.2-52 3
1.2-6 Reactor, Auxiliary & Fuel Building Plan at El 50'-0" 1.2-53 1.2-7 Reactor, Auxiliary & Fuel Building Plan at El 84'-7" 1.2-54 1.2-8 Reactor, Auxiliary & Fuel Building j
Bldg Arrangement Partial Plans 1.2-55
\ 1.2-9 Reactor, Auxiliary & Fuel Building s-) Bldg Arrangement Section A-A 1.2-56 1.2-10 Reactor, Auxiliary & Fuel Building Bldg Arrangement-Section B-B 1.2-57 1.2-11 Reactor, Auxiliary & Fuel Building Bldg Arrangement Section C-C 1.2-58 1.2-12 Reactor, Auxiliary & Fuel Building Bldg Arrangement Section D-D 1.2-59 1,2-13 Radwaste Building Bldg Arrangement Plan at El (-) 43'10" & (-) 27'-10" 1.2-60 1.2-14 Radwaste Building Bldg Arrangement Plan ,
at El (-) 6'-10" & 9'-2" 1.2-61 l.2-15 Radwaste Building Bldg Arrangement Plan At El 32'-2" & 54'2" 1.2-62 1.2-16 Radwaste Building Bldg Arrangement Sections A-A, B-B & C-C 1.2-63 l 1.2-17 Radwaste Building Bldg Arrangement Sections D-D, E-E & F-F 1.2-64 1,2-18 Diesel Generator Bldgs Div 1, 2 & 3 Plan At El (-)6'-10" 1.2-65 l 1.2-19 Diesel Generator Bldgs Div 1, 2, & 3 Plan at El 10'-10" 1.2-66 t
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ILLUSTRATIONS (Continued)
Figure Title Page 1.2-20 Diesel Generator Bldgs Div 1, 2, & 3 Plan at El 28'-0" 1.2-67 1.2-21 Diesel Generator Buildings Div 1 Sections 1.2-68 1.2-22 Diesel Generator Bldgs Div 2 & 3 1.2-69 1.2-23 Control Building Bldg Arrangement Plan At El (-) 6'-10" 1.2-70 1.2-24 Control Building Bldg Arrangement Plan at El 11'-0" 1.2-71 1.2-25 Control Building Bldg Arrangement Plan at El 28'-6" 1.2-72 1.2-26 Control Building Building Arrangement Sections 1.2-73 0
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GESSAR II '
22A7007 238 NUCLEAR ISLAND Rev. O.
1.2 GENERAL PLANT DESCRIPTION O
r
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1.2.1 Principal Design Criteria The principal design criteria are presented in two ways. First,'
they are classified as either a power generation ' function or a safety function. Second, they are grouped according to sistem.
Although the distinctions between power generation or safety functions are not always clear cut and are sometimes overlapping, ,
the functional classification facilitates safety analyses, while the grouping by system facilitates the understanding of both the system function and design.
1.2.1.1 General Design Criteria 1.2.1.1.1 Power Generation Design Criteria (1) The plant shall be designed to produce steam for direct j
4 O use in a turbine-generator unit.
(2) Heat removal systems are provided with sufficient capacity and operational adequacy to remove heat gen-erated in the reactor core-for the full range of normal operational conditions and abnormal operational l transients.
(3) Backup heat removal systems are provided to remove decay heat generated in the core under circumstances i wherein the normal operational heat removal systems become inoperative. The capacity of such systems shall be adequate to prevent fuel cladding damage.
(4) The fuel cladding in conjunction with other plant sys-tems shall be designed to retain integrity so that any l failures are within acceptable limits throughout the O
1.2-1
GESSAR II 22A7007 238 tiUCLEAR ISLAND Rev. 0 1.2.1.1.1 Power Generation Design Criteria (Continued) range of normal operational conditions and abnornal operational transientr for the design life of the fuel.
(5) Control equipment is provided to allow the reactor to respond automatically to load changes and abnormal opera-tional transients.
(6) Reactor power level is manually controllable.
(7) Control of the reactor is possible from a single location.
(8) Reactor controls, including alarms, are arranged to allow the operator to rapidly assess the condition of the reactor system and locate syster malfunctions.
(9) Interlocks or other automatic equipment are provided as backup to procedural controls to avoid conditions requiring the functioning of nuclear safety systems or engineered safety features.
(10) The station shall be designed for routine continuous operation whereby steam activation products, fission products, corrosion products, and coolant dissociation products are processed within acceptable limits.
1.2.1.1.2 Safety Design Criteria (1) The station design conforms to applicable codes and regulations.
(2) The station is designed, fabricated, erected, and operated in such a way that the release of radioactive 1,2-2
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1.2.1.1.2 Safety Design Criteria '(Continued) >
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pertaining to the release of radioactive materials Dr normal operations and for abnormal transients and I accidents. ,
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(5) Gaseous, liquid, and solid waste disposal facilities 's h' aredesignedsothatthedischargeof' radioactive sx..
d effluents and offsite shipment of radioactive materials ._
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The design provides means by dhich ol. ant operators are (6) alerted when limits on the release ~~~
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(7) Sufficient indications are prevfded s to,
- allow determina-tion that the reactor is operating within the envelope of conditions considered safe by plant analysis.
, s (8) Radiation shielding is provided and access control patterns are established,to allow a properly trained' operating staf f to control radiation oNses within the, ' [
limits of applicable regullitions in Jny mode of normal -, . _ ;
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Gr.dSAR Il 22A7007 238 NUCLEAR ISLAND pey, o 1.2.1.1.2 Safety Design Criteria (Continued)
(9) Those portions of the nuclear system that form part of
, the reactor coolant pressure boundary are designed to retain integrity as a radioactive material containment barrier following abnormal operational transients and
^
accidents.
(19) Nuclear safety systems and engineered safety features shall function to assure that no damage to the reactor coolant pressure boundary results from internal pres-sures caused by abnormal operational transients and accidents.
(11) Where positive, precise action is immediately required in response to abnormal operational transients and accidents, such action is automatic and requires no decision or manipulation of controls by plant operations personnel.
(12) Essential safety actions are provided by equipment of sufficient redundancy and independence so that no single failure of active components, or of passive components in certain cases in the long term, will prevent the required actions. For systems or components to which IEEE-279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, and/or IEEE-308-1974, Criteria for Class 1E Electrical Systems for Nuclear Power Generating Stations, apply, single failures of either active or passive electrical components are con-sidered in recognition of the higher anticipated failure rates of passive electrical components relative to passive mechanical components.
O 1.2-4
= - _ _ . __ .- __ - - . - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.1.1.2 Safety Design Criteria (Continued)
(13) Provisions are made for control of active components of nuclear safety systems and engineered safety features from the control room.
(14) Nuclear safety systems and engineered safety features are designed to permit demonstration of their functional performance requirements.
(15) The design of nuclear safety systems and engineered safety features includes allowances for natural environ-mental disturbances such as earthquakes, floods, and storms at the station site.
(16) Standby electrical power sources have sufficient capacity to power all neulear safety systems and engineered
/ safety features requiring electrical power concurrently.
(17) Standby electrical power sources are provided to allow prompt reactor shutdown and removal of decay heat under circumstances where normal auxiliary power is not available.
(18) A containment is provided that completely encloses the rea e system, drywell, and suppression pool. The con-ta; .it employs the pressure suppression concept.
(19) It is possible to test primary containment integrity and leak tightness at periodic intervals.
(20) A secondary containment is provided that completely encloses the primary containment. This secondary con-tainment contains a system for controlling the release of radioactive materials from the primary containment.
1.2-5
GESSAR II 22A7007 233 NUCLEAR ISLAND Rev. 0 1.2.1.1.2 Safety Design Criteria (Continued) g (21) The primary containment and secondary containment in conjunction with other engineered safety features limit radiological effects of accidents resulting in the release of radioactive material to the containment volumes to less than the prescribed acceptable limits.
(22) Provisions are made for removing energy from the primary containment as necessary to maintain the integrity of the containment system following accidents that release energy to the containment.
(23) Piping that penetrates the primary containment and could serve as a path for the uncontrolled release of radio-active material to the environs is automatically isolated whenever such uncontrolled radioactive material release is imminent. Such isolation is performed in time to limit radiological effects to less than the specified acceptable limits.
(24) Emergency core cooling systems are provided to limit fuel cladding temperature to less than the limits of 10CFR50.46 in the event of a loss-of-coolant accident (LOCA).
(25) The emergency core cooling systems provide for contin-uity of core cooling over the complete range of postu-lated break sizes in the reactor coolant pressure boundary.
(26) Operation of the emergency core cooling systems is initiated automatically when required regardless of the availability of offsite power supplies and the normal generating system of the station.
1.2-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 1.2.1.1.2 Safety Design Criteria (Continued)
O (27) The control room is shielded against radiation so that continued occupancy under accident conditions is possible.
(28) In the event that the control room becomes inaccessible, it is possible to bring the reactor from power range operation to cold shutdown conditions by utilizing the local controls and equipment that are available outside the control room.
(29) Backup reactor shutdown capability is provided independ-ent of normal reactivity control provisions. This backup system has the capability to shut down the reactor from any normal operating condition and subsequently to maintain the shutdown condition.
O (30) Fuel handling and storage facilities are designed to prevent inadvertent criticality and to maintain shielding and cooling of spent fuel.
(31) Systems that have redundant or backup safety functions are physically separated and arranged so that any credible events causing damage to any one region of the reactor island complex has minimum prospect for compro-mising the functional capability of the designated counterpart system.
1.2.1.2 System Criteria The principal design criteria for particular systems are listed in the following subsections.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.1.2.1 Nuclear System Criteria (1) The fuel cladding is a radioactive material barrier designed to retain integrity so that any failures are within acceptable limits throughout the design power range.
(2) The fuel cladding in conjunction with other plant sys-tems is designed to retain integrity so that any fail-urcs are within acceptable limits throughout any abnormal operational transient.
(3) Those portions of the nuclear system that form part of the rcactor coolant pressure boundary are designed to retain integrity as a radioactive material barrier during normal operation and following abnormal opera-tional transients and accidents.
(4) Heat removal systems are provided in sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational transients as well as for abnormal operational transients.
The capacity of such systems ia adequate to prevent fuel cladding damage.
(5) Heat removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inopera-tive. The capacity of such systems is adequate to pre-vent fuel cladding damage. The reactor is capable of being shut down automatically in sufficient time to permit decay heat removal systems to become effective following loss of operation of normal heat removal systems.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
~N 1.2.1.2.1 Nuclear System Criteria (Continued)
(6) The reactor core and reactivity control system are designed so that control rod action is capable of bringing the core suberitical and maintaining it so even with the rod of highest reactivity worth fully withdrawn and unavailable for insertion.
(7) The reactor core is designed so that its nuclear characteristics do not contribute to a divergent power transient.
(8) The nuclear system is designed so there is no ter.dency for divergent oscillation of any operating characteris-tic, considering the interaction of the nuclear system with other appropriate plant systems.
~N 1.2.1.2.2 Power Conversion Systems Criteria Components of the power conversion systems shall be designed to perform the following basic objectives:
(1) produce electrical power from the steam coming from the reactor, condense the steam into water, and return the water to the reactor as heated feedwater with a major portion of its gases and particulate impurities removed; and (2) assure that any fission products or radioactivity associated with the steam and condensate during normal operation are safely contained inside the system or are released under controlled conditions in accordance with waste disposal procedures.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.1.2.3 Electrical Power Systems Criteria Sufficient normal auxiliary and standby sources of electrical power are provided to attain prompt shutdown and continued maintenance of the sta tion in a safe condition under all credible circum-stances. The power sources are adequate to accomplish all required essential safety actions under all postulated accident conditions.
1.2.1.2.4 Radwaste System Criteria (1) The gaseous and liquid radwaste systems are designed to limit the release of radioactive effluents from the station to the environs to the lowest practical values.
Such releases as may be necessary during normal opera-tions are limited to values that meet the requirements of applicable regulations including 10CFR20 and 10CFR50.
(2) The solid radwaste disposal systems are designed so that inplant processing and offsite shipments are in accor-dance with all applicable regulations including 10CFR20, 10CFR71, and 49CFR171 through 179 and DOT Regulations, as appropriate.
(3) The system design provides means by which station opera-tions personnel are alerted whenever specified limits on the release of radioactive material may be approached.
1.2.1.2.5 Auxiliary Systems Criteria (1) Fuel handling and storage facilities are designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel.
(2) Other auxiliary systems, such as service water, cooling water, fire protection, heating and ventilating, 1.2-10
GESSAR II 22A7007 238 NUCLEAR. ISLAND Rev. O 1.2.1.2.5 Auxiliary Systems criteria (Continued)
O communications, and lighting, are designed to function during normal and/or accident conditions.
(3) Auxiliary systems that are not required to effect safe shutdown of the reactor or "aintain it in a safe condi-tion are designed so that a failure of these systems shall not prevent the essential auxiliary systems from performing their design functions.
1.2.1.2.6 Shielding and Access Control Criteria (1) Radiation shielding is provided and access control patterns are established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any normal mode of plant operation.
O' (2) The control room is shielded against radiation so that occupancy is possible under accident conditions.
1.2.1.2.7 Nuclear Safety Systems and Engineered Safety Features Criteria Principal design criteria for nuclear safety systems and engineered safety features are as follows:
(1) These criteria correspond to criteria 10 through 17, 24 through 26, 28, and 29 in Subsection 1.2.1.1.2.
(2) Standby electrical power sources have sufficient capacity to power all Class lE and all engineered safety features requiring electrical power concurrently.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.1.2.7 Nuclear Safety Systems and Engineered Safety Features Criteria (Continued)
(3) Standby electrical power sources are provided as neces-sary for support of all engineered safety feature functions (e.g., decay heat remc val) under all circum-stances where normal auxiliary er is not available.
(4) In the event that the control room is inaccessible, it is possible to bring the reactor from power range opera-tion to a hot shutdown condition by use of controls and equipment that are available outside the control room.
Furthermore, station design includes the ability, in this event, for operators to bring the reactor to a cold shutdown condition from the hot shutdown condition from outside the main control room.
(5) Backup reactor shutdown capability is provided indepen-dent of normal reactivity control provisions. This backup system has the capability to shut down the reactor from any operating condition and subsequently to main-tain the shutdown condition.
1.2.1.2.8 Process Control Systems Criteria The principal design criteria for the process control systems are as follows.
1.2.1.2.8.1 Nuclear System Process Control Criteria (1) Control equipment is provided to allow the reactor to respond automatically to main load changes within design limits.
(2) It is possible to control the reactor power level manually.
1.2-12
t l
, GESSAR II 22A7007 l 238 NUCLEAR ISLAND Rev. 0
) 1.2.1.2.8.1 Nuclear System Process Control Criteria (Continued)
(3) Control of the nuclear system is possible from a central location.
(4) Nuclear systems process controls and alarms are arranged to allow the operator to rapidly assess the condition 1
- of the nuclear system and to locate process system malfunctions.
f 1
1.2.1.2.8.2 Power Conversion Systems Process Control Criteria 4
(1) Control equipment is provided to control the reactor pressure throughout its operating range.
(2) The turbine is able to respond automatically to minor changes in load.
(3) Control egaipment in the feedwater system maintains the water level in the reactor vessel at the optimum level required by steam separators.
(4) Control of the power conversion equipment is possible from a central location.
1.2.1.2.8.3 El?ctrical Power System Process Control Criteria l (1) The Class lE power systems are designed as a split-bus i system with either bus being adequate to safely shut i down the unit.
I j (2) Protective relaying is used to detect and isolate faulted equipment from the system with a minimum of disturbance l ( in the event of equipment failure.
1.2-13 i
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.1.2.8.3 Electrical Power System Process Control Criteria (Continued)
(3) Voltage relays are used on the emergency equipment buses to isolate these buses from the normal electrical system in the event of loss of offsite power and to initiate starting of the Standby Emergency Power System diesel generators.
(4) The standby emergency power diesel generators are started and loaded automatically to meet the existing emergency condition.
(5) Electrically operated breakers are controllable from the control room.
(6) Monitoring of essential generators, transformers, and h circuits is provided in the main control room.
1.2.1.2.9 Other Plant Design Criteria Applicant will supply.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2 Plant Description 1.2.2.1 Site Characteristics l.2.2.1.1 Site Location The plant is 1 coated on a site adjacent to or close to a body of water with sufficient capacity for either once-through or recir-culated cooling or a combination of both methods.
1.2.2.1.2 Description of Plant Environs Applicant will supply.
1.2.2.1.2.1 Meteorology The safety-related structures and equipment are designed to retain required functions for the loads resulting from any tornado with
[ characteristics not exceeding the following:
(1) Translational velocity - 70 mph (max)/5 mph (min)
(2) Rotational velocity - 290 mph (3) Maximum wind velocity - 360 mph (4) Internal differential pressure - 3 psi (at 2 psi /sec).
Tornado missiles are discussed in Section 3.5.
1.2.2.1.2.2 Hydrology The safety design basis of the plant provides that structures of safety significance will be unaffected by the hydrologic parameter envelope defined in Section 2.4.
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GESSAR II 22A7007 238 NUCLEAR TSLAND Rev. 0 1.2.2.1.2.3 Geology and Seismology The structures of safety significance for the plant are designed to withstand a safe shutdown earthquake which results in a free-field peak acceleration of 0.3g and a 1/2 safe shutdown earthquake which results in a free-field peak acceleration of 0.15g.
1.2.2.2 General Arrangement of Structures and Equipment The principal structures located in the plant Nuclear Island are the following:
(1) Reactor Building - includes the Shield Building, con-tainment, drywell, and major portions of the Nuclear Steam Supply System; (2) Auxiliary Building - houses the engineered safety features, the systems equipment and switchgear, Reactor Building switchgear, and portions of the heating and ventilating systems; (3) Fuel Building - houses the fuel storage and shipping area, the standby gas treatment system, the control rod drive pumps, the control rod drive service area, and por-tions of the heating and ventilating systems; (4) Radwaste Building - houses the radioactive waste treat-ment facilities; (5) Control Building - includes the control room, the computer facility, and the cable tunnels; and (6) Diesel Generator Buildings.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.2 General Arrangement of Structures and Equipment
())
q, (Continued)
The arrangement of these structures on the plant site is shown in Figure 1.2-1. Figures 1.2-2 through 1.2-26 show the equipment arrangement in the principal buildings.
1.2.2.3 Nuclear Systems The nuclear system includes a direct-cycle forced-circulation boiling water reactor that produces steam for direct use in the steam turbine. A heat balance showing the major parameters of the nuclear system for the rated power corditions is shown in Figure 1.1-2.
1.2.2.3.1 Reactor Core and Control Rods y Fuel for the reactor core consists of slightly enriched uranium (s,) dioxide pellets sealed in Zircaloy-2 tubes. These tubes (or fuel rods) are assembled into individual fuel assemblies. Gross con-trol of the core is achieved by movable, bottom-entry control rods. The control rods are cruciform in shape and are dispersed throughout the lattice of fuel assemblies. The control rods are positioned by individual control rod drives.
Each fuel assembly has several fuel rods with gadolinia Gd 023 mixed in solid solution with the UO2 The Gd 0 s burnable 23 poison which diminishes the reactivity of the fresh fuel. It is depleted as the fuel reaches the end of its first cycle.
A conservative limit of plastic strain is the design criterion used for fuel rod cladding failure. The peak linear heat genera-tion for steady-state operation is well below the fuel damage limit even late in life. Experience has shown that-the control rods are not susceptible to distortion and have an average life l () expectancy many times the residence time of a fuel loading.
1.2-17 t . _ _ _ _ _ .-.
GESSAR II 22A7007 238 UUCLEAR ISLAND Rev. 0 1.2.2.3.2 Reactor Vessel and Internals The reactor vessel contains the core and supporting structures; the steam separators and dryers; the jet pumps; the control rod guide tubes; the distribution lines for the feedwater, core sprays, and standby liquid control; the in-core instrumentation; and other components. The main connections to the vessel include steamlines, coolant recirculation lines, feedwater lines, control rod drive and in-core nuclear instrument housings, core spray lines, residual heat removal lines, standby liquid control line, core differential pressure line, jet pump pressure sensing lines, and water level i instrumentation.
The reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal opera-ting pressure in the steam space above the separators is 1040 psia.
The vessel is fabricated of low alloy steel and is clad internally with stainless steel (except for the top head, nozzles, and nozzle weld zones which are unclad).
The reactor core is cooled by demineralized water that enters the lower portion of the core and boils as it flows upward around the fuel rods. The steam leaving the core is dried by steam separators and dryers located in the upper portion of the reactor vessel.
The steam is then directed to the turbine through the main steam-lines. Each steamline is provided with two isolation valves in series; one on each side of the containment barrier.
1.2.2.3.3 Reactor Recirculation System The R2 actor Recirculation System consists of two recirculation pump loops external to the reactor vessel. These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps. Each external loop contains one high capacity motor-driven recirculation pump, two motor-operated maintenance 9
1.2-18
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t GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 1.2.2.3.3 Reactor Recirculation System (Continued)
{}
j valves, and one hydraulically-operated flow control valve. The l variable-position hydraulic-flow-control valve operates in con-junction with a low-frequency motor generator set to control i
reactor power level through the effects of coolant flow rate on moderator void content.
The jet pumps are reactor vessel internals. The jet pumps provide
! a continuous internal circulation path for the major portion of
- the core coolant flow. The jet pumps are located in the annular region between the core shroud and the vessel inner wall. Any recirculation line break would still allow core flooding to approximately two-thirds of the active core height - the level of the inlet of the jet pumps.
1.2.2.3.4 Residual Heat Removal System O The Residual Heat Removal (RHR) System is a system of pumps, heat exchangers, and piping that fulfills the following functions:
(1) removes decay and sensible heat during and after plant shutdown; (2) injects water into the reactor vessel following a loss-of-coolant accident to reflood the core independent of other core cooling systems (Subsection 1.2.2.4.8, Emergency Core Cooling Systems);
(3) removes heat from the containment following a loss-of-coolant accident to limit the increase in containment pressure. This is accomplished by cooling and recir-culating the suppression pool water (containment cooling) and by spraying the containment air space (containment spray) with suppression pool water.
)
4 1.2-19
i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l
I 1.2.2.3.5 Reactor Water Cleanup System The Reactor Water Cleanup System (RWCS) recirculates a portion of reactor coolant through a filter-demineralizer to remove particu-late and dissolved impurities from the reactor coolant. It also removes excess coolant from the reactor system under controlled conditions.
1.2.2.3.6 Nuclear Leak Detection System The Nuclear Leak Detection and Monitoring System consists of temperature, pressure, flow, and fission-product sensors with associated instrumentation and alarms. This system detects and annunciates leakage in the following systems:
(1) main steamlines; (2) Reactor Water Cleanup System (RWCS);
(3) residual Heat Removal (RHR) System; (4) Reactor Core Isolation Cooling (RCIC) System; (5) Feedwater System; (6) ECCS Systems; and (7) miscellaneous systems.
Small leaks generally are detected by monitoring the air coolers condensate flow, radiation levels, and drain sump fill-up and pump-out rates. Large leaks are also detected by changes in reactor water level and changes in flow rates in process lines.
1.2.2.4 Nuclear Safety Systems and Engineered Safety Features 1.2.2.4.1 Reactor Protection System The Reactor Protection System (RPS) initiates a rapid, automatic shutdown (scram) of the reactor. It acts in time to prevent fuel cladding damage and any nuclear system process barrier damage 1.2-20
l GESSAR II 22A7007 ,
238 NUCLEAR' ISLAND Rev. 0 ,
i 1.2.2.4.1 Reactor Protection System (Continued) following abnormal operational transients. The reactor protection system overrides all operator actions and process controls and is i based on a fail-safe design philosophy that allows appropriate pro-
) tective action even if a single failure occurs.
1.2.2.4.2 Neutron Monitoring System 4
Those portions of the Neutron Monitoring System (NMS) that are part of the RPS qualify as a nuclear safety system. The inter-mediate range monitors (IRM) and the average power range monitors APRM) which monitor neutron flux via incore detectors provide scram j logic inputs to the Reactor Protection System to initiate a scram in time to prevent excessive fuel clad damage as a result of over-power transients. The APRM system also generates a simulated thermal power signal. Both upscale neutron flux and upscale simulated thermal power are conditions which provide scram logic O signals.
i 1
1.2.2.4.3 Control Rod Drive System i
i When a scram is initiated by the Reactor Protection System, the l Control Rod Drive (CRD) System inserts the negative reactivity l necessary to shut down the reactor. Each control rod is controlled f individually by a hydraulic control unit. When a scram signal is j received, high-pressure water stored in an accumulator in the 4 hydraulic control unit or reactor pressure forces its control rod i into the core.
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__m-~ ._ _ - - , _ _ _ . _ _ _ _ _ __. . . _ _ . . _ _ - _ . . . - . . - _ - _ _ _ . _ _ _ - _ - -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.4.4 Control Ro'? Drive Housing Supports Control rod drive housin g supports are located underneath the reactor vessel near the control rod housings. The supports limit the travel of a control rod in the event that a control rod housing is ruptured. The supports prevent a nuclear excurison as a result of a housing failure and thus protect the fuel barrier.
1.2.2.4.5 Control Rod Velocity Limiter A control rod velocity limiter is attached to each control rod to limit the velocity at which a control rod can fall out of the core should it become detached from its control rod drive. This action limits the rate of reactivity insertion resulting from a rod drop accident. The limiters contain no moving parts.
1.2.2.4.6 Nuclear System Pressure Relief System A pressure relief system consisting of safety / relief valves mounted on the main steamlines is provided to prevent excessive pressure inside the nuclear system for operational transients or accidents.
1.2.2.4.7 Reactor Core Isolation Cooling System The Rector Core Isolation Cooling (RCIC) System provides makeup water to the reactor vessel when the vessel is isolated. The RCIC system uses a s te,4;n-d riven turbine-pump unit and operates automatically in time and with sufficient coolant flow to maintain adequate water level in the reactor vessel for events defined in Section 5.4.
o 1.2.2.4.8 Emergency Core Cooling Systems Four Emergency Core Cooling Systems (ECCS) are provided to main-tain fuel cladding below the temperature limit in 10CFR50.46 in 1.2-22
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 1.2.2.4.8 Emergency Core Cooling Systems (Continued)
O the event of a breach in the reactor coolant pressure boundary a that results in a loss of reactor coolant. The systems are:
(1) High-Pressure Core Spray (HPCS) System - The HPCS System provides and maintains an adequate coolant inventory inside the reactor vessel to limit fuel cladding tempera-tures in the event of breaks in the reactor coolant pressure boundary. The system is initiated by either high pressure in the drywell or low water level in the vessel. It operates independently of all other systems over the entire range of pressure differences from greater-than-normal operating pressure to zero. The HPCS cooling decreases vessel pressure to enable the low-pressure cooling systems to function. The HPCS system pump motor is powered by a diesel generator if auxiliary power is not available and the system may Os also be used as a backup for the RCIC System.
(2) Automatic Depressurization System (ADS) - The ADS rapidly reduces reactor vessel pressure in a loss-of-coolant accident (LOCA) in which the HPCS System fails to maintain the reactor vessel water level. The depres-surization provided by the system enables the low-pressure ECCS to deliver cooling water to the reactor vessel.
The ADS uses some of the relief valves that are part of the nuclear system pressure relief system. The auto-matic relief valves are arranged to open on conditions indicating both that a break in the reactor coolant pressure boundary has occurred and that the HPCS System is not delivering sufficient cooling water to the reactor vessel to maintain the water level above a 1.2-23
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.4.8 Emergency Core Cooling Systems (Continued) preselected value. The ADS will not be activated unless either the low-pressure core spray system (LPCS) or low-pressure coolant injection (LPCI) pumps are operating.
This is to ensure that adequate coolant will be avail-able to maintain reactor water level after the depressurization.
(3) Low Pressure Core Spray System - The LPCS System consists of one independent pump and the valves and piping to deliver cooling water to a spray sparger over the core.
The system is actuated by conditions indicating that a breach exists in the reactor coolant pressure boundary (RCPB) but water is delivered to the core only after reactor vessel pressure is reduced. This system provides the capability to cool the fuel by spraying water into each fuel channel. The LPCS loop functioning in con-junction with the ADS or HPCS can provide sufficient fuel cladding cooling following a LOCA.
(4) Low Pressure Coolant Injection - Low pressure coolant injection is an operating mode of the RHR System, but is discussed here because the LPCI mode acts as an engineered safety feature in conjunction with the other ECCSs. LPCI uses the pump loops of the RHR to inject cooling water into the pressure vessel. LPCI is actuated by conditions indicating a breach in the RCPB, but water is delivered to the core only after reactor vessel pressure is reduced. LPCI operation provides the capability of core reflooding following a LOCA in time to maintain the fuel cladding below the prescribed temperature limit.
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. GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ,
I 1.2.2.4.9 Containment 1.2.2.4.9.1 Functional Design
, The containment design for this plant has been given the name Mark III. This containment design incorporates the drywell/ pressure suppression feature of previous BWR containment designs into a dry-containment-type structure. In fulfilling its design basis as a fission product barrier, the Mark III containment is a low-leakage structure even at the increas'ed pressures that could follow a main steamline rupture or a recirculation line break.
The main features of t,' . containment design include:
(1) a drywell surrounding the reactor pressure vessel (RPV) and a large part of the RCPB; (2) a suppression pool which serves as a heat sink during normal operation and accident conditions; (3) a containment upper pool for shielding, refueling opera-4 tions, and suppression pool makeup; and (4) the containment, a free standing steel structure.
The main features of the secondary containment d'esign include:
l (1) the ECCS pump rooms in Auxiliary Building; I (2) the Fuel Building; and (3) the containment to Shield Building annulus.
The containment function design is described in more detail in Section 6.2.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.4.9.2 IIcat Removal The containment RIIR System is summarized in Subsection 1.2.2.4.14.
1.2.2.4.9.3 Environmental Systems The containment heating and ventilating systems are described in Subsection 1.2.2.8.15.
1.2.2.4.9.4 Containment Spray A containment spray system is provided for containment cooling and for steam condensation in the containment. Steam in the contain-ment is the result of steam bypass of the drywell. The contain-ment spray can be initiated manually or automatically on a high-containment pressure signal to prevent over-pressurization of the containment. The containment spray system consists of two 100%-
capacity redundant subsystems each with its own full-capacity spray header. Each subsystem is supplied from a separate redundant RHR subsystem.
1.2.2.4.9.5 Combustible Gas Control In the event of a LOCA, hydrogen and oxygen will be generated in the reactor. The Gombustible Gas Control System ensures that hydrogen concentrations are kept below the limits specified in NRC Regulatory Guide 1.7. The systems used include a moritoring system, a mixing system, a hydrogen control system. and a purge system.
1.2.2.4.10 Containment and Reactor Vessel Isolation Control System The Containment and Reactor Vessel Isolation Control System automatically initiates closure of isolation valves to close off 1.2-26
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O 1.2.2.4.10 Containment and Reactor Vessel Isolation Control System (Continued) all process lines which are potential leakage paths for radioactive material to the environs. This action is taken upon indication of a breach in the RCPB or main steamlines outside the RCPB.
1.2.2.4.11 Main Steamline Isolation Valves Although all pipelines that both penetrate the containment and offer a potential release path for radioactive material are pro-vided with redundant isolation capabilities, the main steamlines, because of their large size and large mass flow rates, are given special isolation consideration. Automatic isolation valves are provided La each main steamline. Each is powered by both air pres-sure and spring force. These valves fulfill the following objectives:
(1) prevent excessive 3amage to the fuel barrier by limiting the loss of reactos coolant from the reactor vessel resulting from either a major leak from the steam piping outside the containment or a malfunction of the pressure control system resulting in excessive steam flow from the reactor vessel; (2) limit the release of radiocctive materials by isolating the reactor coolant pressure boundary in case of a gross release of radioactive materials from the fuel to the reactor cooling water and steam; and (3) limit the release of radioactive materials by closing the containment barrier in case of a major leak from the nuclear system inside the containment.
O 1.2-27
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.4.12 Main Steamline Flow Restrictors A venturi-type flow restrictor is installed in each steamline.
These devices limit the loss of coolant from the reactor vessel before the main steamline isolation valves are closed in case of a main steamline break outside the containment.
1.2.2.4.13 Main Steamline Radiation Monitoring System The Main Steamline Radiation Monitoring System consists of four gamma radiation monitors located externally to the main steamlines just outside the containment. The monitors are designed to detect a gross release of fission products from the fuel. On detection of high radiation, the trip signals generated by the monitors are used by the Reactor Protection System to initiate a reactor scram and a close the main steamline isolation valves.
1.2.2.4.14 Residual Heat Removal System (Containment Cooling)
The containment cooling subsystem is placed in operation to: (1) limit the temperature of the water in the suppression pool and of the atmospheres in the drywell and suppression chamber following a design basis LOCA; (2) control the pool temperature during nor-mal operation of the safety / relief valves and the RCIC System; and (3) reduce the pool temperature following an isolation transient. In the containment cooling mode of operation, the RHR main system pumps take suction from the suppression pool and pump the water through the RHR heat exchangers where cooling takes place by transferring heat to the service water. The fluid is then discharged back to the suppression pool, to the drywell spray header, to the suppression chamber spray header, or to the RPV.
1.2.2.4.15 Ventilation Exhaust Radiation Monitoring System The Process Ventilation Radiation Monitoring Systems consist of a number of radiation monitors arranged to monitor the activity level 1.2-28
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.4.15 Ventilation Exhaust Radiation Monitoring System
(~)x
(_ (Continued) of the air exhaust from the containment and drywell, Auxiliary Building, fuel-handling areas, and control room.
1.2.2.4.16 Standby Gas Treatment System The Standby Gas Treatment System (SGTS) minimizes exfiltration of contaminated air from the annulus between the containment and Shield Building and from the Auxiliary Building ECCS pump room to the environment following an accident or abnormal condition which could result in abnormally high airborne radiation in these areas.
The Fuel Building ca.. be exhausted to the SGTS.
All necessary equipment and surrounding structures are designed to Seismic Category I specifications.
O) t All components of the SGTS are operable during loss of the offsite power supply.
1.2.2.4.17 Auxiliary Building and Fuel Building Isolation Control System The Auxiliary Building and Fuel Building Isoaltion Control System automatically initiates closure of isolation dampers in all ventilation ducts which are potential leakage paths for radioactive material to the environs. The action is taken upon indication of a potential breach in the nuclear system process barrier or a fuel-handling accident.
1.2.2.4.18 Standby AC Power Supply Standby ac power is supplied by three diesel generators. Each Class lE division is supplied by a separate diesel generator.
[~)
V There are no provisions for transferring Class lE buses between 1.2-29
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.4.18 Standby AC Power Supply (Continued) standby ac pcwer supplies or supplying more than one engineered safety feature (ESP) division from one diesel generator. This one-to-one relationship between diesel generator and ESF division ensures that a failure of one diesel generator can affect only one ESF division. The diesel generators are housed in a Seismic Category I structure to comply with applicable NRC and IEEE design guides and criteria.
1.2.2.4.19 DC Power Supply The plant has four independent Class lE 125-volt de systems.
1.2.2.4.20 Standby Liquid Control System Although not intended to provide prompt reactor shutdown as the control rods are, the Standby Liquid Control System provides a redundant, independent, and alternate way to bring the nuclear fission reaction to subcriticality and to maintain subcriticality as the reactor cools. The system makes possible an orderly and safe shutdown in the event that not enough control rods can be inserted into the reactor core to accomplish shutdown in the normal manner. The system is sized to counteract the positive reactivity effect from rated power to the cold shutdown condition.
1.2.2.4.21 Remote Shutdown System In the event that the control room becomes inaccessible, the reactor can be brought from power range operation to cold shutdown conditions by the use of th local controls and equipment that are available outside the control room.
O 1.2-30
T GESSAR II 22A7007 238 NUCLEAR ISLAND t s
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1.2.2.4.22 Main Steam Positive Leakage Control System N . ,
Following a LOCA, the Main Steam Positive Leakage Control System j (MSPLCS) is operated to prevent the releasejof radioactive steam 3 which could leak through the closed main steam isolatio'n valve 3. .
This is accomplished by establishing a pressurized volume in the main steamlines and maintaining a pressure of' a least~10% over s
that of the reactor at post-LOCA condition. , -i s
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1.2.2.5 Power Conversion System Applicant will supply.
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1.2.2.5.2 Main Steam System
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.5.6 Steam Bypass System and Pressure Control System A turbine bypass system is provided which passes steam directly to the main condenser under the control of the pressure regulator.
Steam is bypassed to the condenser whenever the reactor steaming ratc' exceeds the load permitted to pass to the turbine generator.
'The capacity of the turbine bypass system is 35% of the reactor rated steam flow. The Pressure Regulation System provides main turbine-control valve and bypass valve flow demands to maintain a nearly constant reactor pressure during normal plant operation. It also provides demands to the recirculation system to adjust power level by changing reactor recirculation flow rate.
1.2.2.5.7 Circulating Water System Applicant will supply.
1.2.2.5.8 Condensate Storage Facilities Applican will supply.
1.2.2.5.9 Condensate and Feedwater System Applicant will supply.
1.2.2.6 Electric Power Systems and Instrumentation and Control Systems 1.2.2.6.1 Electric Power Systems Applicant will supply.
1.2.2.6.2 Electrical t'ower System Process Control and Instrumentation s
1.2-32 i
GESSAR II 22A7007
'238 NUCLEAR ISLAND Rev. 0 1.2.2.6.3 Nuclear System Process Control and Instrumentation 1.2.2.6.3.1 Rod Control and Information System
.i The Rod Control and Information System provides the means by which j
control rods are positioned from the control room for power con-trol. The system operates valves in each hydraulic control unit I
to change control rod position. One gang of control rods can be manipulated at a time. The system includes the logic that restricts control rod movement (rod block) under certain conditions as a backup to procedur:1 controls.
1.2.2.6.3.2 Recirculation Flow Control System-During normal power operation, a variable position discharge valve is used to control flow. Adjusting this valve changes the coolant flow rate through the core and thereby changes the core power level.
The system can automatically adjust the reactor power output to O the load demand. For startup and shutdown flow changes at lower power, the pump speed is changed by adjusting the frequency of j the electrical power supply.
I 1.2.2.6.3.3 Neutron Monitoring System The Neutron Monitoring System (NMS) is a system of in-core neutron detectors and out-of-core electronic monitoring equipment.
The system provides indication of neutron flux, which can be cor-related to thermal power level for the entire range of flux con-ditions that can exist in the core. The source range monitors (SRM) and the intermediate range monitors (IRM) provide flux
! level indications during reactor startup and low-power operation.
The local power range monitors (LPRM) and average power range monitors (APRM) allow assessment of local and overall flux condi-tions during power range operation. The Traversing In-core Probe (TIP) System provides a means to calibrate the individual LPRM 1,2-33
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. 0 1.2.2.6.3.3 Neutron Monitoring System (Continued) sensors. The MMS provides inputs to the Reactor Manual Control System to initiate rod blocks if preset flux limits are exceeded and inputs to the RPS to initiate a scram if other limits are exceeded.
1.2.2.6.3.4 Refueling Interlocks A system of interlocks that restricts movement of refueling equip-ment and control rods when tr.e reactor is in the refueling and startup modes is provided to prevent an inadvertent criticality during refueling operations. The interlocks back up procedural and other mechanical controls that have the same objective. The interlocks affect the refueling platform, refueling platform hoists, fuel grapple, and control rods.
1.2.2.6.3.5 Reactor Vessel Instrumentation In addition to instrumentation for the nuclear safety systems and engineered safety features, instrumentation is provided to monitor and transmit information that can be used to assess conditions existing inside the reactor vessel and the physical condition of the vessel itself. This instrumentation monitors reactor vessel pressure, water level, coolant temperature, reactor core differ-ential pressure, coolant flow rates, and reactor vessel head inner-scal ring leakage.
1.2.2.6.3.6 Process Computer System An on-line process computer is provided to monitor and log process variables and to make certain analytical computations.
O 1.2-34
! GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4
4 1.2.2.6.4 Power Conversion Systems Process Control and j Instrumentation 1.2.2.6.4.1 Pressure Regulator and Turbine Generator Control l The pressure regulator maintains control of the turbine control and and turbine bypass valves to allow proper generator and reactor response to system load demand changes while maintaining the nuclear system pressure essentially constant.
j The turbine generator speed-load controls act to maintain the turbine speed (generator frequency) constant and respond to load
. changes by adjusting the reactor recirculation flow control system and pressure regulator setpoint.
] The turbine generator speed-load controls can initiate rapid j closure of the turbine control valves (rapid opening of the tur-
, bine bypass valves) to prevent turbine overspeed on loss of the
) generator electric load.
1.2.2.6.4.2 Feedwater Control System The feedwater control system automatically controls the flow of feedwater into the reactor pressure vessel to maintain the water within the vessel at predetermined levels. A conventional three-
- element control system is used to accomplish this function.
l 1.2.2.7 Fuel llandling and Storage Systems 1.2.2.7.1 New and Spent Fuel Storage l New and spent fuel storage racks are designed to prevent inadvert-I ent criticality and load buckling. Sufficient coolant and shielding
, are maintained to prevent overheating and excessive personnel 1
! exposure, respectively. The design of the fuel pool provides for
!(/
corrosion resistance, adherence to Seismic category I requirements, 1.2-35 l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.7.1 New and Spent Fuel Storage (Continued) and prevention of k from reaching 0.95 under dry or flooded eff conditions. This subject is further discussed in Section 9.1.
1.2.2.7.2 Fuel Handling System The fuel-handling equipment includes a 125-ton cask crane, fuel-handling platform, fuel inspection stand, fuel preparation machine, fuel assembly transfer mechanism, containment refueling platform, 125-ton containment crane, and other related tools for reactor servicing. All equipment conforms to applicable codes and standards.
The only function of the cask crane is to handle the spent fuel cask. The fuel handling platform transfers the fuel assemblies between the transfer pool, storage pools, and cask. Fuel assem-blies are transferred through the transfer tube between the reactor building and the fuel building. The fuel assemblics inside the containment are handled by the refueling platform.
The handling of the reactor head, removable internals, and drywell head during refueling is accomplished using the containment crane.
All tools and servicing equipment necessary to meet the reactor general-servicing requirements are designed for efficiency and safe serviceability.
1.2.2.8 Cooling Water and Auxiliary Systems 1.2.2.8.1 Closed Cooling Water System The Closed Cooling Water System provides cooling water to certain designated equipment located in the containment, the Auxiliary, Fuel, and Radwaste Buildings. Adequate capacity and redundancy is provided in heat exchangers and pumps to ensure performance of the Cooling System under normal modes of plant operation. In the 1.2-36
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1
- 1.2.2.8.1 Closed Cooling Water System (Continued) event of loss of offsite power, but without LOCA, emergency power for the system is available from the onsite emergency diesel gen-erators. The closed loop provides a barrier between nonessential possibly-contaminated systems and the service water discharged to the environment. Heat is removed from the closed loop by the service-water system. Radiation monitors are provided to detect contaminated leakage into the closed systems.
1.2.2.8.2 Fuel Pool Cooling and Cleanup System The Fuel Pool Cooling and Cleanup System maintains acceptable levels of temperature and clarity and minimizes radioactivity levels of the water in the upper containment, fuel storage, and cask pools. The system includes two heat exchangers, each capable of removing one-half of the decay heat generated from an average discharge of spent fuel, and two filter /demineralizers, each unit l
{N ' having the capacity to pass the system flow or greater in order to maintain the desired purity level.
1.2.2.8.3 Essential Service Water System The plant is equipped with a Service Water System that provides cooling water to the RHR heat exchangers, the emergency core cooling pump room air coolers, and the closed cooling water system i heat exchangers. The system is designed with sufficient redundancy to ensure heat removal capability for all modes of operation including normal, shutdown, and hot standby operation, accident conditions, and refueling operations. Redundant power supplies are provided for use in the event of loss of offsite power.
l 1.2.2.8.4 Ultimate Heat Sink
)
i 1 Applicant will supply.
1.2-37
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.8.5 Condensate Storage and Transfer System The Condensate Storage and Transfer System maintains the required capacity and flow of the condensate for the RCIC and HPCS Systems and maintains the required level in the condenser hotwell. The system also stores and transfers upper containment pool water during refueling and cask storage pool water during fuel-shipping cask loading, receives and stores the process effluent from the liquid radwaste system, provides makeup to other plant systems where required, provides storage space for the suppression pool water during plant shutdown, and provides condensate to the Control Rod Drive (CRD) Hydraulic System.
The system consists of a condensate storage tank, two condensate transfer pumps, and the necessary controls and instrumentation.
1.2.2.8.6 Plant Raw Water Treatment and Makeup Water Treatment System Applicant will supply.
O 1.2.2.8.7 Potable and Sanitary Waste Water System Applicant will supply.
1.2.2.8.8 Plant Chilled-Water System The Plant Chilled-Water System consists of the Drywell Chilled-Water System, Control Building Chilled-Water System, and the Nuclear Island and Monessential Chilled-Water System. Chilled water is provided to the drywell coolers, the RWCU pump room coolers, the Control Building air conditioning units cooling coils, the Auxiliary Building electric switchgear room cooling coils, and the Auxiliary, Fuel, Radwaste, and Containment Buildings fan coil units for space cooling and dehumidification to maintain area 0
1.2-38
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
-w 1.2.2.8.8 Plant Chilled-Water System (Continued)
(V \
ambient temperatures. The only safety-related Chilled-Water System is the Control Building Chilled-Water System and it is designed to meet the requirements of Criterion 19 of 10CFR50.
1.2.2.8.9 Process Sampling Systems The process sampling system is furnished to provide process informa-tion that is required to monitor plant and equipment performance and changes to operating parameters. Representative liquid and gas samples are taken automatically and/or manually during normal plant operation for laboratory on-line analyses.
1.2.2.8.10 Plant Equipment and Floor Drainage The Equipment and Floor Drainage Systems are designed to collect g-s liquid waste throughout the plant and discharge the radioactive
(_,/ waste to the Radwaste System for processing. Separate drainage facilities are provided for nonradioactive waste.
The Drainage System is also used to detect abnormal leakage in the ESF equipment rooms and Fuel Building.
1.2.2.8.11 Service and Instrument Air Systems The Service and Instrument Air Systems provide a continuous supply of compressed air of suitable quality and pressure for instrument control and general plant use. The servic, air compressor and the instrument air compressors discharge into their respective air receivers. The air is then distributed throughout the plant.
Instrument air is additionally filtered and dried prior to distribu-tion throughout the plant.
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1.2-39
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.8.12 Normal Auxiliary AC Power Applicant will supply.
1.2.2.8.13 Diesel Generator Fuel-Oil Storage and Transfer System The major components of this system are the fuel-oil storage tanks, pumps, and day tanks. Each diesel generator has its own individual supply components. Each storage tank is designed to supply the diesel needs during the post-LOCA period and each day tank has the capacity for two hours of diesel generator operation. Each fuel-oil pump is controlled automatically by day-tank level and feeds its day tank from the storage tank. Additional fuel-oil pumps supply fuel to each diesel-fuel manifold from the day tank.
1.2.2.8.14 Auxiliary Steam System Applicant will supply.
1.2.2.8.15 Heating, Ventilating, and Air Conditioning (Environmental) Systems The plant Environmental Control Systems control temperature, pressure, humidity, and airborne contamination to ensure the integ-rity of plant equipment, p. ovide acceptable working conditions for plant personnel, and limit offsite releases of airhorne contaminants.
The following Environmental Systems are provided:
(1) Control Room Air Conditioning System consisting of supply, recirculation / exhaust and makeup air cleanup units to ensure the habitability of the control room under normal and abnormal conditions of plant operation; (2) Shield Building Annulus Recirculation and Exhaust System to maintain a negative pressure in the annulus space 1.2-40
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1
s 1.2.2.8.15 Heating, Ventilating, and Air Conditioning
- g_) (Environmental) Systems (Continued) under normal and abnormal operating conditions isolating potential leak sources from the containment; (3) Containment Cooling System to remove heat generated during normal plant operation, shutdown, and refueling periods; (4) Containment Pressure Control-Low-Purge Supply and Exhaust System as well as the High-Flow Purge Supply Exhaust Systems to maintain a negative pressure in Fte contain-ment with respect to atmosphere except during modes of operation when the containment is isolated (areas of the Containment with shielding are held at a negative pressure compared to the unshielded areas to isolate
- potential sources of airborne contamination);
O' (5) Containment Dome Circulating System to prevent thermal stratification in the upper areas of the containment.
l
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(6) Drywell Cooling System to remove heat generated during normal plant operation including reactor scrams, hot standby, shutdown, and refueling periods; 1
(7) Drywell Vent System to transfer air for the containment to the drywell when the containment Pressure Control-Low-Purge Supply and Exhaust System is purging the drywell; (8) the Auxiliary Building Pressure Control Supply and Exhaust System to distribute air so that a negative pressure is maintained in the emergency core cooling equipment roo.ms, thereby isolating the potential airborne contamination in these rooms; 1.2-41
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.8.15 Heating, Ventilating, and Air Conditioning (Environmental) Systems (Continued)
(9) the Auxiliary Building Electrical Equipment Supply and Exhaust System to pressurize the electrical rooms allow-ing exfiltration of air,to the battery rooms for exhaust to the outside atmosphere; (10) the Fuel Building Pressure Control Supply and Exhaust System to maintain the Fuel Building at a negative pressure with respect to the outside atmosphere to prevent the potential release of airborne contamination; (11) the Radwaste Building Supply and Exhaust System to main-tain a negative pressure in the tank rooms and treatment cells to isolate potential airborne contamination; (12) the Diesel Generator Building Air Exhaust System to pro-vide cooling during operation of the diesel generators.
A tempered air supply system controls the thermal environ-ment when the diesel generators are not operating.
(13) cooiers in the steam tunnel and ECCS rooms to remove heat generated during operation of the equipment in these rooms.
1.2.2.8.16 Lighting Systems The design basis for the lighting facilities is the standard of the Illuminating Engineering Society. Special attention is given to areas where proper lighting is imperative during normal and emergency operations. The system design precludes the use of mercury vapor fixtures in the containment and the fuel-handling areas. The normal lighting systems are fed from the unit auxiliary transformers. Emergency power is supplied by engineered safety 1.2-42
__ _ .. - _ _ _ _ _ _ . _ ~ . - . _ . _ _ __
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l
1.2.2.8.16 Lighting Systems (Continued) 4 i
buses backed up by diesel generators. Normal operation and regular simulated offsite power loss tests verify system integrity.
1.2.2.8.17 Fire Protection System The fire protection system is designed to provide an adequate supply of water or chemicals to points throughout the plant where fire protection is required. Diversified fire-alarm and fire-suppression types are selected to suit the particular areas or hazards being protected. Chemical fire fighting systems are also provided as additions to or in lieu of the water fire fighting systems.
1.2.2.9 Radioactive Waste Systems 1.2.2.9.1 Gaseous Radwaste System The purpose of the gaseous radwaste system is to process and con-j trol the release of gaseous radioactive wastes to the site environs I so that the total radiation exposure to persons outside the i controlled area does not exceed the maximum limits of the appli-
- - cable 10CFR regulations even with some defective fuel rods.
The offgases from the main condenser are the major source of gaseous radioactive waste. The treatment of these gases includes volume reduction through a catalytic hydrogen / oxygen recombiner, water vapor removal through a condenser, decay of short-lived i radioisotopes through a holdup line, further condensation and cooling, filtration, adsorption of isotopes on activated charcoal i beds, further filtration through high efficiency filters, and final releases.
1 Continuous radiation monitors are provided which indicate radio-active release from the reactor and from the charcoal adsorbers.
l.2-43 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.2.2.9.1 Gaeous Radwaste System (Continued)
The radiation monitors are used to isolate the Offgas System on high radioactivity in order to prevent releasing gases of unacceptably high activity.
1.2.2.9.2 Liquid Radwaste The purpose of the Liquid Radwaste System is to collect, segregate, store, and process potentially radioactive liquids generated during normal plant operation and anticipated operational occur-rences so that the total radiation exposure to plant personnel and persons outside the controlled area does not exceed the maxi-mum limits of the applicable codes and regulations.
Radioactive liquid wastes are collected as three classes: high conductivity, low conductivity, and detergent or laundry wastes.
The treatment of these radioactive wastes includes filtration, chemical neutralization, demineralization, and distillation depending upon the waste class involved.
Grab samples and instrumentation are utilized to monitor the amount of radioactivity present in various process streams.
Processed waste which would be discharged offsite is collected, sampled, and batch-analyzed prior to discharge.
1.2.2.9.3 Solid Radwaste The purpose of the Solid Radwaste System is to provide solidifica-tion and packaging for radioactive wastes that are produced during shutdown, startup, and normal plant operation and to store these wastes until they are shipped offsite for burial.
The Solid Radwaste System processes waste input from the Reactor Water Cleanup System, the Fuel Pool Cooling and Cleanup System, 1.2-44
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 ]
[ 1.2.2.9.3 Solid Radwaste (Continued) 1 (
the Liquid Radwaste System, resins, and particulate wastes from 1 1
the Condensate Cleanup System. The system also serves to compact and package miscellaneous dry radioactive materials and to package .
I contaminated metallic materials and incompressible solid objects.
Collection, solidification, packaging, and storage of radioactive wastes are done in conformance to the applicable codes and regu-lations for offsite shipment.
1.2.2.10 Radiation Monitoring and Control 1.2.2.10.1 Process Radiation Monitoring Process radiation monitoring systems are provided to monitor and control radioactivity in process and effluent streams and to
\
activate appropriate alarms and controls.
A process radiation monitoring system is provided for indicating and recording radiation levels associated with selected plant process streams and effluent paths leading to the environment.
All effluents from the plant which are potentially radioactive are monitored.
l Process radiation monitoring is also discussed in Chapters 7, 9, l and 11.
l 1.2.2.10.2 Area Radiation Monitors Area radiation monitoring systems alert occupants a'.d the control room personnel of excessive gamma radiation levels a' selected locations within the plant.
1.2-45
__ _ . . . - _ _ .~. _ _ _ _ ___ _-._____ ____ _ _ _ _ . . _ - - _ _ _ _ _ . _ _ . _ . _ _ - . - . . _ _ _ _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O l.2.2.10.3 Site Environs Radiation Monitors Applicant will supply.
1.2.2.11 Shielding Shielding is provided throughout the plant, as required to reduce radiation levels to operating personnel and to the general public within the applicable limits set forth in 10CFR20 and 10CFR100.
It is also designed to protect certain plant components from radiation exposure resulting in unacceptable alterations of material properties or activation.
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1.2-46
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L s!' f6 'l 5 _ ._ ,_ ._i_. __ , p-- - l H SECTi~ N FB i Drawing K-C44, Rev. 4 i - i e i 9 i io i H i er t ! '-* s Figure 1.2-26. Control Building Euilding Arrgt j Sections 1.2-73/1.2-74
GESSAR II ~ 4
- 22A7007 . 238 MUCLEAR ISLAND , Rev. 0 d
SECTION 1.3~ - *
/. '
1 CONTENTS , i
. Section Title
.Page 1.3 COMPARISON TABLES 'l.3-1 l
1.3.1 Comparisons with:Similar Facility Designs 1.3-1 l.3.1.1 Nuclear Steam Supply System Design ' f Characteristics 1.3-1 1 1.3.1.2 Power Conversion System Design i Characteristics 1.3-1 1.3.1.3 Engineered Safety Features Design j Characteristics 1.3-1 1.3.1.4 Containment Desic a Characteristics l'.' 3 - 1 1.3.1.5 Radioactive Waste Management Systems '
; Design Characteristics l.3-2 1.3.1.6 Structural Design Characteristics 1.3-2 1.3.1.7 Instrumentation and Electrical
- Systems Design Characteristics- 1.3-2 g,,/ 1.3.2 Comparison of Final and Preliminary Information TABLES ,
Table Title ,, Page , 1.3-1 Comparison of Nuclear Steam Supply System 1.3-3 Design Characteristics 1.3-2 Comparison of Power Conversion System Design Characteristics 1.3-13 . 1.3-3 Comparison of Engineered Safety Featurep *
, 3 Design Characteristics 1.3-15 1.3-4 Comparison of Containment Desian f^1a r .ctoristics -1.3-19 1.3-5 Radioactive Waste Management Ln.sta$ .; e61gn ,
, Characteristics 1.3-23 1.3-6 Comparison of Structural Design Characteristics l'.3-25 1.3-7 Comparison of Instrumentation and Electrical i Systems Design Characteristics 1.3-26 i () 1.3-8 Comparison of Final and Preliminary Information 1.3-27 1.3-i/1.3-il ,
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.3 COMPARISON TABLES 1.3.1 Comparisons with Similar Facility Designs This subsection highlights the principal design features of the plant and compares it's major features with those of other boiling water reactor facilities. Th'a design of this facility is based on proven technology obtained during the development, design, con-struction, and operation of boiling water reactors of similar types. The data, performance, characteristics, and other informa-tion. presented here represent a current, firm design.
, 1.3.1.1 Nuclear Steam Supply System Design Characteristics Table 1.3-1 summarizes the design and operating characteristics for the Nuclear Steam Supply Systems. Parameters are related to power output for a single plant unless otherwise noted.
) 1.3.1.2 Power Conversion System Design Characteristics g Table 1.3-2 compares the Power Conversion System design characteris-2cs.
1.3.1.3 Engineered Safety Features Design Characteristics Table 1.3-3 compares the engineered safety features design characteristics. 1.3.1.4 Containment Design Characteristics Table 1.3-4 compares the containment design characteristics. O 1.3-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.3.1.5 Radioactivc Waste Management Systems Design Characteristics Table 1.3-5 compares the Radioactive Waste Management Systems design characgeristics. 1.3.1.6 Structural Design Characteristics Table 1.3-6 compares the structural design characteristics. 1.3.1.7 Instrumentation and Electrical Systems Design Characteristics Table 1.3-7 compares the Instrumentation and Electrical Systems design characteristics. 1.3.2 Comparison of Final and Preliminary Information All of the significant changes that have been made in the facility h design since submission of the PSAR are listed in Table 1.3-8. Each item in Table 1.3-8 is cross-referenced to the appropriate portion of the GESSAR which describes the current designs and the bases for them. O 1.3-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are related to rated power output for a single plant unless otherwise noted) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/S BWR/6 Design 238-748 218-560 218-560 251-800 Thermal and Hydraulic Rated power 3,579 2,436 2,436 3,833 (MWt) Design power 3,729 2,550 2,550 4,025 (MWt) (ECCS design basis) Steam flow 15.40 10.03 10.477 16.491 rate, M1b/hr O at 420*F (FW Temp) Core coolant 104.0 78.5 78.5 112.5 flow rate (Mlb/hr) Feedwater flow 15.372 9.998 10.447 16.455 rate (Mlb/hr) System pressure, 1,040 1,020 1,020 1,040 nominal in steam dome (psia) Average power 54.1 51.2 50.51 54.1 density (kW/ liter) Maximum linear 13.4 18.5 13.4 13.4 heat ger. oration rate (kW/ft) Average linear 5.9 7.11 5.40 5.93 l heat generation rate (kW/ft) Maximum heat 361,600 428,300 354,255 361,600 ( s) flux (Btu /hr-ft2) i 1.3-3 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Design 238-748 218-560 218-560 251-800 Thermal and Ilydraulic (Continued) Average heat 159,500 164,410 144,032 160,300 flux (Btu /hr-ft2) Maximum UO 3,435 4,380 3,325 3,435 2 temperature (*P) Average 2,185 2,781 2,130 2,185 volumetric fuel temperature (*F) Average cladding 565 566 566 565 surface tempera-ture (*F) Minimum critical 1.20 N/A 1.24 1.20 power ratio (MCPR) Coolant 527.7 523.7 527.4 527.9 enthalpy at core inlet (Btu /lb) Core maximum exit 79 75 75 76 voids within assemblies Core average 14.7 12.7 13.2 14.6 exit quality (% steam) Feedwater 420 387.4 420 420 temperature ( F) 9 1.3-4
. __ . = _ . . . .- . .- . ._ _ - - . . . - - . .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O
\,
l
) Table 1.3-1 , COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued) i This Plant Hatch 1 Zimmer Grand Gulf 4,
BWR/6 BWR/4 BWR/5 BWR/6 i Design 238-748 218-560 218-560 251-800 _ Thermal and Hydraulic (Continued) Design power peaking factor Maximum relative 1.40 1.40 1.40 1.40 assembly power. Local peaking 1.13 1.24 1.24 1.13 . factor
-Axial peaking 1.40 1.5 1.5 1.40 factor Total Peaking 2.26 2.60 2.43 2.26 4 factor Nuclear (first core)
Water /UO 2.70 2.53 2.55 2.70 volume r$tio (cold) s Reactivity with <0.99 <0.99 <0.99 <0.99 strongest control rod out (keff)
- Dynamic void -7.16 -10.74 -8.57 -7.14 i coefficient (C/%) 40.95 38.0 40.54 41.31 l at core average voids (%)
(EOC-rated output) Fuel temperature -0.412 -0.403 -0.419 -0.396 doppler coeffi-cient (C/ C). (EOC-rated output) - Initial average 1.90 2.34 1.90 1.70
- U-235 enrichment (%)
Initial cycle 9,138 9,413 9,200 7,500
- exposure (mwd /short ton) f
! 1.3-5 i
GESSAR II 22A7007 23'8 NUCLEAR ISLAND Rev. O Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Design 238-748 218-560 218-560 251-800 Core Mechanical Fuel assembly Number of fuel 748 560 560 800 assemblies Fuel rod array 8x8 7x7 8x8 8x8 Overall length 176 176 176 176 inches Weight of UO 2 456 466 466 458 per assembly lb (pellet type) Weight of fuel 697 675 698 697 assembly (1b) Fuel Rods Number per fuel 62 49 63 62 assembly outside diameter 0.483 0.563 0.493 0.483 (in.) Cladding 0.032 0.032 0.034 0.032 thickness (in.) Diametral gap, 0.009 0.012 0.009 0.009 pellet-to-cladding (in.) Length of gas 9.48 16 14 9.48 planum (in.) Cladding Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2 material *
- Free-standing loaded tubes 1.3-6
I GESSAR II 22A7007 { 238 NUCLEAR ISLAND Rev. 0 () Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM
! DESIGN CHARACTERISTICS (Continued)
This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Design 238-748 218-560 218-560 251-800 Core Mechanical (Continued) Fuel Pellets Material UO 2 00 2 UO U0 2 2-Density (% of 95 95 95 95 theoretical) Diameter (in.) 0.410 0.487 0.416 0.410 Length (in.) 0.410 0.5 0.420 0.410 Fuel channel 167.36 166.9 166.9 166.9 () Overall dimen-sion, length (in.) Thickness (in.) 0.120 0.080 0.100 0.120 Cross section 5.45 x 5.45 5.44 x 5.44 5.48 x 5.48 5.45 x 5.45 dimensions (in.) Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Core assembly Fuel weight 341,640 272,850 260,551 365,693 as UO 2 (lb) i Core diameter 185.2 160.2 160.2 191.5 (equivalent) l (in.) Core height 150 144 146 150 l' (active fuel) (in.) l
- O
} t l.3-7 i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.3-1 h COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Design 238-748 218-560 218-560 251-800 Core Mechanical (Continued) Reactor Control System Method of Movable Movable Movable Movable variation of control rods control control control reactor power and variable rods and rods and rods and forced variable variable variable coolant forced forced forced flow coolant coolant coolant flow flow flow Number of 177 137 137 193 movable con-trol rods Shape of Cruciform Cruciform Cruciform Cruciform movabic control rods Pitch of 12.0 12.0 12.0 12.0 movable control rods Control BC BC BC BC 4 4 4 4 material in movable rods granules granules granules granules compacted compacted compacted compacted in SS tubes in SS in SS in SS tubes tubes tubes Type of Bottom Bottom Bottom Bottom control rod entry entry entry entry drives locking locking locking locking piston piston piston piston Type of Burnable Burnable Burnable Burnable temporary poison; poison; poison; poison; reactivity gadolinia- gadolinia- gadolinia- gadolinia-control for urania urania urania urania initial core fuel rods fuel rods fuel rods fuel rods e 1.3-8
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () Table 1.3-1 COMPARISON OF NUCLEAR STE7M SUPPLY SYSTEM DESIGN CHARACTERISTICS l Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Design 238-748 218-560 218-560 251-800 Core Mechanical (Continued) In-core neutron instrumentation Total number 164 124 124 176 of LPRM detec-tors Number of in- 41 31 31 44 core LPRM penetrations Number of LPRM 4 4 4 4 detectors per penetration Number of SRM 4 4 4 6 penetrations Number of IRM C 8 8 8 penetrations Total nuclear 53 43 43 58 instrument penetrations Source range Shutdown through criticality monitor range Intermediate Prior to criticality to low power range monitor range ! Power range Approximately 1% power to 125% power monitors range Local power 164 124 124 176 range monitors O l 1.3-9
GESSAR II 22A7007 238 MUCLEAR ISLAND Rev. O Table 1.3-1 h COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Design 238-748 218-560 218-560 251-800 Core Mechanical (Continued) In-core neutron instrumentation (continued) Average power 4 6 6 8 range monitors Number and type 7 Sb-B'e 5 Sb-Be 5 Sb-Be 7 Sb-Be of in- core neutron sources Reactor Vessel Material Low-alloy Carbon Carbon Low-alloy steel / steel / steel / steel / stainless stainless stainless stainless clad clad clad clad Design pressure 1,250 1,250 .1,250 1,250 (psig) Design tempora- 575 575 575 575 ture (*F) Inside diameter 19-10 18-2 18-2 20-11 (ft-in.) Inside height 70-4 68-9 69-2 72-7 (ft-in.) Minimum base 6.0 6.50 5.375 6.19 metal thickness (cylindrical section) (in.) Minimum cladding 1/8 1/8 1/8 1/8 thickness (in.) . O 1.3-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM Table 1.3-1 DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/S BWR/6 Design 238-748 218-560 218-560 251-800 Reactor Coolant Recirculation Number of recir- 2 2 2 2 culation loops Design pressure inlet leg 1230 1148 1250 1250 (psig) outlet leg 16'20*; 1274 1675*; 1650*; (psig) 1550** 1575** 1550** Design tempera- 575 562 575 575 ture (*F) Pipe diameter 22/24 28 20 24 (}
~ G' (in.)
Pipe material 304/316 304/316 304/316 304/316 (ANSI) Recirculation 42,000 45,200 32,500 44,600 pump flow rate (gpm) Number of jet 20 20 20 24 pumps in reactor Main Steamlines Number of 4 4 4 4 steamlines Design pressure 1,250 1,118 1,250 1,250 (Psig) 4 (} ( ,f
- Pump and discharge piping to and including discharge block valve
** Discharge piping from discharge block valve to vessel i
1.3-11
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Design 238-748 218-560 218-560 251-800 Main Steamlines (Continued) Design tempera- 575 560 575 575 ture (*F) Pipe diameter 26 24 24 28 (in.) Pipe material Carbon Carbon Carbon Carbon steel steel steel steel O O 1.3-12
; GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. 0
() Table 1.3-2 COMPARISON OF POWER CONVERSION SYSTEM DESIGN CHARACTERISTICS This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/S BWR/6 System / Component 238-748 218-560 218-560 251-800 Turbine Generator (Sections 10.2 and 10.4) Rated power 1269 813.0 883 1306 (MWe) (gross) Generator speed 1800 1800 1800 1800 (rpm) 4 Rated steam 15.40 10.47 11.01 15.54
; flow (Mlb/hr)
Inlet pressure 975 950 950 965 (psig)
- Steam Bypass System (Section 10.4.4)
! Capacity 35 25 25 35 l (% design steam
! flow)
Main Condenser l (Section 10.4.1) Heat removal 7996 5717 7053 8506 capacity ! (MBtu/hr) Circulating . Water System l (Section 10.4.5) Applicant 2 3 2 to supply i ' Flow rate Applicant 185,000 150,000 285,500 (gpm/ pump) to supply i l.3-13
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.3-2
)
COMPARISON OF POWER CONVERSION SYSTEM DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 System / Component 238-748 218-560 218-560 251-800 Condensate and Feedwater System Design flow 15.372 10.469 10.971 17.284 rate (Mlb/hr) Number of Applicant 3 3 3 condensate will supply pumps Number of Applicant 3 3 3 condensate will supply booster pumps Number of Applicant 2 2 2 feedwater will supply pumps Number of Applicant None None None feedwater will supply booster pumps Condensate Applicant AC power AC power AC power pump drive will supply Condensate Applicant AC power AC power AC power booster pump will supply drive Feedwater Applicant Turbine Turbine Turbine pump drive will supply Feedwater Applicant booster pump will supply drive 9 1.3-14
GF.SSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/ Table 1.3-3 COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6
, System / Component 238-748 218-560 218-560 251-800 Emergency Core Cooling Systems (sized on design power - Section 6.3) Low Pressure Core Spray Systems Number of loops 1 2 1 1 Flow rate 6000 at 4725 at 4725 at 7000 at (gpm) 122 psid 113 psid 119 psid 122 psid High Pressure Core Spray System Number of loops u 1 l 1 1 Plow rate 1550 at 4250 443 at 1650 at (gpm) 1147 psid 1160 psid 1147 psid 6110 at 4725 at 7000 at 200 psid 200 psid 200 psid Automatic Depressurization System ' i Number of relief 8 7 6 8 valves i Low Pressure Coolant Injection D Number of loops 3 2 3 3 Number of pumps 3 4 3 3 i Flow rate 7100 at 9200 at 5050 at 7450 at I (gpm/ pump) 20 psid 20 psid 20 psid 20 psid i Auxiliary Systems (Sections 5.5 and 9.2) O 1.3-15
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.3-3 h COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 System / Component 238-748 218-560 218-560 251-800 Residual Heat Removal System Reactor shutdown cooling mode Number of loops 2 2 2 2 Number of pumps 2 4 2 2 Plow rate 7100 7700 5050 7450 (gpm/ pump)c Duty (MBtu/hr/ 46.9 30.8 30.8 50.0 heat exchanger)d Number of heat 2 2 2 2 exchangers Primary cor.tain- 7100 7700 5050 7450 ment cooling rode Flow rate (gpm)* Essential Service Water System Flow rate Applicant 8000 5000 25,300 (gpm/ heat will supply total exchanger) Number of pumps Applicant 4 4 2 at 12,000 will supply gpm 1 at 1,300 gpm Reactor Core Isolation Cooling System Flow rate 700 at 400 at 400 at 800 at (gpm) 165-1192 165-1135 165-1135 165-1192 psia psia psia psia reactor reactor reactor reactor pressure pressure pressure pressure 1.3-16
GESSAR II 22A7007 238 NUCLEAR ISLAND ._/. 0 () Table 1.3-3 COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 System / Component 238-748 218-560 218-560 251-800 Fuel Pool Cooling and Cleanup System Capacity 8.0 8.5 6.9 11.8 (MBtu/hr) NOTES
- a. High-pressure coolant injection system utilized
- b. A mode of the RHR system
- c. Capacity during reactor flooding mode wi'h t more than one pump running
- d. Heat exchanger-duty at 20 hours following reactor shutdown
- e. Flow per heat exchanger O
? I O 1.3-17/1.3-18
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () Table 1.3-4 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/S BWR/6 Containment" 238-748 218-560 218-560 251-800 Primary Type Mark IT.I Pressure Over and Mark III free- Suppres- under reinforced standing sion pressure concrete steel with suppres- containment reinforced sion with steel concrete liner shield building Construction Cylindrical Concrete Concrete Reinforced free- with steel pre- concrete standing liner; stressed cylinder steel with steel with with hemi-ellipsoidal structure steel spherical head head; steel () liner lined Drywell Concrete Light Frustum Concrete cylinder bulb / of cone cylinder steel upper vessel portion Pressure- Steel lined Torus / Cylindri- Steel lined suppression concrete steel cal lower concrete chamber cylinder vessel portion cylinder Pressure- 15 56 45 15 suppression chamber internal design pressure (psig) Drywell internal 30 56 45 30 design pressure (psig) Drywell free 275,000 146,240 180,000 270,000 volume (ft ) O 1.3-19
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.3-4 ll COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Containment
- 238-748 218-560 218-560 251-800 Primary (Continued)
Pressure- 1,140,000 110,950 93,000 1,400,000 suppression chamber free volume (ft ) Pressure- 129,600 87,300 120.120 136,000 suppression (upper pool (upper pool water dump = pool dump volume (ft3)(ggg) 34,200 = 72,800 Submergence of 7.5 3.67 10 7.5 vent pipe below pressure pool surface (f t) (IlWL) Design temper- 330 281 340 330 ature of drywell ( F) Design temper- 185 281 275 185 ature of pressure-suppression chamber (*F) Downcomer vent 2.5 - 3.5 6.21 2.17 2.5-3.5 pressure loss factor Break area / 0.012 0.0194 0.008 0.008 total vent area Calculated maxi- 23.0 46.5 40.4 22.0 mum pressure after blowdown to drywell (psig) O 1.3-20
GESSAR II , 22A7007 238 NUCLEAR ISLAND Rev. O () COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Continued) Table 1.3-4 This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Containment
- 238-748 218-560 218-560 251-800 Primary (Continued)
.i
~
Pressure- 8.7 28 35.6 9.0 suppression chamber (psig) Initial pressure- 50 50 35 30 suppression pool temperature rise (*F) Leakage rate 1.0 1.2 0.635 0.35 (% free volume / day) Secondary 4 / Type Controlled Controlled Controlled Controlled k/~) leakage leakage leakage leakage i Construction Lower levels Reinforced Reinforced Reinforced concrete concrete conrete Upper levels Reinforced Steel Steel
- concrete super- super-structure structure l
and and ! siding siding Roof Reinforced Steel Steel concrete decking decking l 4 l l l 1.3-21 f i . ._ - .. _ _ - . , . -
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.3-4 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Containment
- 238-748 218-560 218-560 251-800 Secondary (Continued)
Internal design 0.25 0.25 0.25 0.25 pressure (psig) Design inleakage 100 100 100 100 rate (% free volume / day at 0.25 in. H2 0} NOTE
- Where applicable, containment parameters are based on design rated power.
O O 1.3-22
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () Table 1.3-5 RADIOACTIVE WASTE MANAGEMENT SYSTEMS DESIGN CHARACTERISTICS This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Radwaste 238-748 218-560 218-560 251-800 Gaseous Radwaste (Section 11.3) Design Bases 100,000 100,000 100,000 100,000 annual annual annual annual average average average average Noble gases at 30 min at 30 min at 30 min at 30 min (pCi/sec) Process treatment Recombiner, Recom- Recom- Recombiner, , chilled biner biner Chilled Charcoal Ambient Chilled Charcoal Charcoal Charcoal Number of beds 2 12 5 8 (} Design condenser 30 40 12.5 40 in-leakage (scfm) Release point- Applicant 394 172 29 height above to provide ground (ft) Liquid Radwaste (Section 11.2) Treatment of: * (1) Floor drains E,D & R F,D, & R F,E, & R F,E,D & R (2) Equipment F,D & R F,D, & R F,D, &R F,E,D & R drains (3) Chemical E,D & R F, dis- E,D, con- Neutralized drains charged centrates E, returned E, solid to solid to equip-to rad- radwaste ment drain waste dis- collector tillate R tank O 1.3-23 .i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.3-5 RADIOACTIVE WASTE MANAGEMENT SYSTEMS DESIGN CHARACTERISTICS (Continued) This Plant Hatch 1 Zimmer Grand Gulf BWR/6 BWR/4 BWR/5 BWR/6 Radwaste 238-748 218-560 218-560 251-800 Liquid Radwaste (Continued) (4) Laundry E ar F Diluted Reverse drains discharged and sent osmosis to circu- discharge lating water dis-charge (5) Expected 100 20 10.9 110 annual average release excluding tritium (mci)
- NOTE D = domineralized F= filtered E = evaporator / concentrator R= recycled i.e., returned to condensate storage 9
1.3-24
. __ - .-_ -~ . .
J GESSAR II 22A7007 238-NUCLEAR ISLAND Rev. 0 i l
, Table 1.3-6
) COMPARISON OF STRUCTURAL DESIGN CHARACTERISTICS This Plant Hatch 1 Zimmer ' Grand Gulf BWR/6 _ BWR/4 BWR/5 BWR/6 j Design 238-748 218 60 218-560 251-800 Seismic Design s ! (Section 3. 2)
)
Operating Basis Earthquake (1/2 SSE) i horizontal g 0.15 0.08 0.10 0.075 vertical.g 0.10 0.05 0.07 0.05 Safe Shutdown 1 Earthquake (SSE) horizontal 9 0.30 0.15 0.20 0.15 vertical 9 0.20 0.10 0.14 0.10 Wind Design (Section 3.3)
!!aximum sustained 130 105 90 90 l (mph)
Tornado Design ! Translational 70 max. 60 60 60 l (mph) 5 min. Tangential (mph) 290 300 300 300 l l 1 i !O ! 1.3-25 i - . .
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.3-7 COMPARISON OF INSTRUMENTATION AND ELECTRICAL SYSTEMS DESIGN CHARACTERISTICS Refer to Table 7.1-2 O O 1.3-26
_ _. - - _ _ _ _ _ - . _ .. .- _ . ~~ . _ _ . .. .__ . ._. __ . _.. _ -. - - . ___ _ _-__ he R i Table 1,3-8 l 4 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR l Item Change Reason for Change FSAR Section i Positive Leakage Added systems to supply sealing To resolve NRC staff concerns 6.5.3.2.2 Water Positive. Control System medium (air / water) to pressurize regarding bypass leakage Seal System-space between isolation valves 6.7 Main Steam Positive of select lines thereby prevent- Leakage Control System ing bypass leakage following postulated LOCA Reduced contain- Increased allowable primary Positive Leakage Control 15.6.5.5 Radiological i l ment leak rate containment leakage rate from Systems reduce bypass leak- ' Consequences 0.3%/ day to 1.0%/ day age thus enabling higher w filtered release rates W
$o Suppression pool Incorporated stainless steel Protects against pitting'and 3.8.2.4.3 Corrosion
{ H liner material cladding of the carbon steel corrosion;1 reduces pool- Prevention pQJ j u containment vessel in the wetted maintenance, operating costs to tn i /o 4 areas of the suppression pool over the life of the plant, $( ,
- and crud accumulation gg.
1 en H h' Suppression pool Added system Improvement in reliability 9.5.9 Suppression Pool Cleanup System of plant operations Cleanup System y Containment buck- To incorporate state-of-the-art Improved analytical Appendix 3F Dynamic ling methodology methods for stability analysis methodology Buckling Crieeria for of the containment vessel Containment vessel 4 4 Recirculation Added motor-generator sets to Provides improved operation 7.7.1.3 Recirculation System MG set provide control for reduced System Flow Control . flow during startup and shutdown Recirculation Deleted decoupler from design Analysis demonstrates that 5.4.1.4 Recirculation w pump / motor there are no unacceptable System safety Evalua- yy decoupler consequences for the postu- tion and Appendix 3D, , <4
- lated LOCA event Chapter 3- i o
oa I w w~ m
Table 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR (Continued) Item Change Reason for Change FSAR Section Lcw-low set relief Added logic to assure that no To maintain PSAR design 7.3.1.1.1.2 ADS logic more than one safety / relief basis for containment loads valve cycles subsequent to the and overpressure transients first pressure peak Reduced SRV loads Reduced magnitude of SRV load To reflect recent in-plant Attachment A of definition by approximately 35% test data Appendix 3B, Chapter 3 N Nuclear fuel The number of water rods in each Improved fuel performance 4.2.2 Fuel Design $ fuel bundle has been changed Description ,, from 1 to 2. Five different 2! a g U-235 enrichments are now used Q$ M U1 . in the fuel >> w %' E' The containment cylindrical To make the lower portion of 3.8.2 Steel Containment b Lower portion of m free-standing shell is backed by structural the containment more rigid y[ steel containment concrete below elevation and thus reduce the struc- r backed by concrete (-) 5 ft, 3 in., in the tural response due to SRV $ annular space loadings O Recirculation e Material change e Improve plant reliability 5.4.1 Reactor Recircu-System o Removed bypass lines by reducing cracking lation system CRD System CRD return line to RPV deleted Reduce nozzle cracking 4.6.1.1.2.4 CRD problem Hydraulic System Feedwater The thermal sleeve was changed To eliminate failure, leak- 5.3.3.1.4.5 Reactor sparger to provide improved slip fit age, and provide for possible Vessel Nozzles design of sparger to nozzle in-service inspection Fuel storage Added more fuel storage cast- Increases capacity to handle 9.1.1 New Fuel Storage racks ings for use in spent fuel, more onsite fuel storage 9.1.2 Spent Fuel y and containment pool areas Storage yf y
< sJ
- O O
O sJ O O O
GESSAR II 22I7007 238 NUCLEAR ISLAND Rev 0 SECTION 1.4 CONTENTS Section Title Page 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 Applicant 1.4-1 1.4.2 Architect Engineer - Nuclear Island ' - Design 1.4-1 1.4.3 Nuclear Steam Supply System Supplier 1.4-1 l.4.4 Turbine-Generator Vendor 1.4-2 1.4.5 Consultants 11 4-2 TABLES . Table Title Page O 1.4-1 Commercial Nuclear Reactors Completed, Under Construction or in Design by General Electric
"/
1.4-3 , 3 i 1 t
.. 9 -
I r, i l 1.4-1/l.4-li ,
- l. . . , - _ - . - - . . _-_.-.--.--.-/ >
- GESSAR II 22A7007 238 NUCLEAR ISLAND Rev, 0 p
() 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS Applicant will supply. 1.4.1 Applicant 1.4.2 Architect Engineer - Nuclear Island Design The Nuclear Island design consists of detailed designs for the Reactor Building, Fuel Building, Auxiliary Building, Radwaste Building, Control Building, and the Diesel Generator Buildings. CF Braun & Co (CFB) is subcontracted to GE to provide engineer-ing services related to the design of these Nuclear Island buildings. CFB has been engaged in engineering and construction since 1909. The company offers three major services: design engineering, consulting and research, and worldwide construction. The company serves the process industry in many different fields: () chemical, petroleum, natural gas, and ore processes and offers a full range of services to the electric utility industry. In recent years CFB has designed major facilities for the Atomic Energy Commission. CFB is qualified to provide engineering services to GE for the Nuclear Island facilities. 1.4.3 Nuclear Steam Supply System Supplier GE has been awarded the contracts to design, fabricate, and deliver the direct cycle boiling water nuclear steam supply system, to fabricate the first core of nuclear fuel, and to provide technical direction for installation and startup of this equipment. GE has engaged in the development, design, construction and operation of boiling water reactors since 1955. Table 1.4-1 lists over 80 GE I reactors completed, under construction, or on order. Thus, GE has substantial experience, knowledge, and capability to design, manu-facture, and furnish technical assistance for the installation and () startup of reactors. 1.4-1
GESSAR II 22A7007 l 2 38 NtJCLEAR ISLAND Rev. O ; 1.4.4 Turbine-Generator Vendor i 1.4.5 Consulta:. .s i O' O 1.4-2
! GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4 'i i Table 1.4-1 ! ) COMMERCIAL NUCLEAR REACTORS COMPLETED, UNDER CONSTRUCTION, j OR IN DESIGN BY GENERAL ELECTRIC ' l [! Rating Year of Year of
! Station Utility (MWe) Order Startup Dresden 1 Commonwealth Edison 207 1955 1960 '
llumboldt Bay Pacific G&E 70 1958 1963 ! Kahl Germany 15 1958 1961 ! Garigliano Italy 150 1959 1964 ; Big Rock Point Consumers Power 72 1959 1963 JPDR Japan 11 1960 1963 KRB Germany 237 1962 1967 Tarapur 1 India 190 1962 1969 Tarapur 2 India 190 1962 1969 : , GKN Holland 52 1963 1968 Oyster Creek JCP&L 640 1963 1969 Nine Mile Point 1 Niagara Mohawk 610 1963 1970 i Dresden 2 Commonwealth Edison 794 1965 1970 Pilgrim Boston Edison 670 1965 1972 Millstone 1 NUSCO 652 1965 1971 Tsuruga Japan 340 1965 1970 Nuclenor Spain 440 1965 1971 l Fukushima 1 Japan 439 1966 1971 i BKW KKM Switzerland 306 1966 1972 I Dresden 3 Commonwealth Edison 794 1966 1971 Monticello Northern States 548 1966 1971 ! Quad Cities 1 Commonwealth Edison 789 1966 1972 i Browns Ferry 1 TVA 1067 1966 1974 Browns Ferry 2 TVA 1067 1966 1975 i Quad Cities 2 Commonwealth Edison 789 1966 1972
- Vermont Yankee Vermont Yankee 515 1966 1972
} Peach Bottom 2 Philadelphia Electric 1065 1966 1974
! Peach Bottom 3 Philadelphia Electric 1065 1966 1974 Fitzpatrick PASNY 821 1968 1975 j Bailly NIPSCO 660 1967 1982 Shoreham LILCO 820 1967 1979 Cooper Nebraska PPD 778 1967 1974 Browns Ferry 3 TVA 1067 1967 1977 Limerick 1 Philadelphia Electric 1100 1967 1983 Hatch 1 Georgia 786 1967 1975 Fukushima 2 Japan 762 1967 1974 1 Brunswick 1 Carolina P&L 821 1968 1977 ) Brunswick 2 Carolina P&L 821 1968 1975 Duane Arnold Iowa ELP 545 1968 1975 Fermi 2 Detroit Edison 1093 1968 1980 i Limerick 2 Philadelphia Electric 1100 1967 1985
! Hope Creek 1 PSE&G 1067 1969 1984
, Hope Creek 2 PSE&G 1067 1969 1986 4
O 4
- 1. 4-3 ~
I
GESSAR II 22A7007 238 NUCLEAR IS LAND Rev. O Table 1.4-1 COMMERCIAL NUCLEAR REACTORS COMPLETED, UNDER CONSTRUCTION, OR IN DESIGN BY GENERAL ELECTRIC (Continued) Rating Year of Year of Station Utility (FH1e) Order Startup Zimmer 1 CCDPP 810 1969 1979 Chinshan Taiwan 610 1969 1978 Caorso 1 Italy 822 1969 1977 Hatch 2 Georgia 786 1970 1978 La Salle 1 Commonwealth Edison 1078 1970 1979 La Salle 2 Commonwealth Edison 1078 1970 1980 Susquehanna 1 Pennsylvania P&L 1050 1967 1980 Susquehanna 2 Pennsylvania P&L 1050 1968 1982 Chinshan 2 Taiwan 610 1970 1979 Hanford 2 WPPSS 1100 1971 1980 Nine Mile Point 2 Niagara Mohawk 1100 1971 1982 Grand Gulf 1 Mississippi P&L 1250 1971 1980 Grand Gulf 2 Mississippi P&L 1250 1971 1981 Kaiseraugst Switzerland 915 1971 1982 Fukushima 6 Japan 1135 1971 1979 Tokai 2 Japan 1135 1971 1977 Riverbend 1 Gulf States 940 1972 1983 Riverbend 2 Gulf States 940 1972 1985 Perry 1 Cleveland Electric 1205 1972 1981 h Perry 2 Cleveland Electric 1205 1972 1983 Hartsville 1 TVA 1233 1972 1983 Hartsville 2 TVA 1233 1972 1904 Hartsville 3 TVA 1233 1972 1983 Hartsville 4 TVA 1233 1972 1984 Laguna Verde 1 Mexico 660 1972 1980 Leibstadt Switzerland 940 1972 1980 Kuosheng 1 Taiwan 992 1972 1980 Kuosheng 2 Taiwan 992 1972 1981 Clinton 1 Illinois Power 950 1973 1981 Clinton 2 Illinois Power 950 1973 1988 Allens Creek 1 Houston L&P 1200 1973 1985 Skagit 1 Puget Sound 1290 1973 1984 Skagit 2 Puget Sound 1290 1974 1986 black Fox 1 , Oklahoma 1150 1973 1983 Black Fox 2 Oklahoma 1150 1973 1985 Cofrentes Spain 975 1973 1980 Laguna Verde 2 Mexico 660 1973 1981 Alto Lazio 1 Italy 982 1974 1983 Alto Lazio 2 Italy 982 1974 1984 O 1.4-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O SECTION 1.5 1 CONTENTS Section Title Page 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 Current Development Programs 1.5-1 1.5.1.1 Instrumentation for Vibration 1.5-1 1.5.1.2 Core Spray Distribution 1.5-1 1.5.1.3 Core Spray and Core Flooding Heat Transfer Effectiveness 1.5-1 1.5.1.4 Verification of Pressure Suppression Design- - 1.5-2 1.5.1.5 Boiling Transition Testing 1.5-6
- 1.5.2 PSAR Commitment Items 1.5-7 1.5.3 References 1.5-7
) TABLES Table Title Page 1.5-1 Summary of PSTP Tests 1.5-9 j
1 1.5-2 Commitment Items 1.5-13 i 4 O 1.5-i/1.5-ii
. - __ _ . . . _ _ . _ _ . . - - . _ . ~ _ _ . . ~ . . . _ . . , . . _ . . --.
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i 1.5 REQUIREMENTS FOR PURTilER TECilNICAL INFORMATION 1.5.1 Current Development Programs , 1.5.1.1 Instrumentation for Vibration Vibration testing for reactor internals is performed on all GE BWR
- plants. At the time of issuance of NRC Regulatory Guide 1.20, test f programs for compliance were instituted. The first BWR/6 plant l of each size is considered a prototype design and will be instru-mented and subjected to both cold and hot, two-phase flow testing to demonstrate that flow-induced vibrations similar to those expected during operation will not cause damage. Subsequent i plants which have internals similar to those of the prototypes l
will be tested in compliance to the requirements of Regulatory Guide 1.20 to conform the adequacy of the design with respect to vibration. Further discussion is presented in Section 3.9. O 1.5.1.2 Core Spray Distribution i GE has conducted a program to predict BWR/6 core spray distribu-i tions using a combination of single-nozzle steam and air tests, single- and multiple-nozzle analytical models, and full-scale air tests. This methodology has been confirmed by a full-scale 30* sector steam test. The results have been submitted to the NRC. i An internal NRC memorandum has been issued apporving the use of the GE methodology for application to BWR/6. 1.5.1.3 Core Flooding IIeat Transfer Effectiveness t Due to the incorporation of an 8 x 8 fuel-rod array with unheated water rods, tests have been conducted to demonstrate the effec-l tiveness of ECCS in the new geometry. lO 1.5-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.5.1.3 Core Flooding Heat Transfer Effectiveness (Continued) These tests are regarded as confirmatory only since the geometry change is very slight and the water rods provide an additional heat sink inside the bundle which improves heat transfer effectiveness. The program concerns the testing of core spray and core flooding heat transfer effectiveness. The results of testing with stain-less steel cladding were reported in the Licensing Topical Report NEDO-10001, Modeling the BWR/6 Loss-of-Coolant Accident: Core Spray and Bottom Flooding Heat Transfer Effectiveness, March 1973. The results of testing using Zircaloy cladding were reported in the Licensing Topical Report NEDO-20231, Emergency Core Cooling Tests of an Internally Pressurized, Zircaloy Clad, 8x8 Simulated BWR Fuel Bundle, December 1973. 1.5.1.4 Verification of Pressure Suppression Design f GE has conducted a large scale test program to verify the per-formance characteristics of the Mark III containment. The pur-pose of the Mark III Test Program was to confirm the analytical methods used to predict the drywell and containment pressure response following the postulated LOCA. This Test Program was also used to obtain information on the hydrodynamic loads that are generated in the vicinity of the suppression pool during a LOCA. The General Electric Mark III Containment Pressure Suppression Test Program was initiated in 1971 with a series of small-scale tests. The test apparatus consisted of small-scale simulations of the reactor pressure vessel, drywell, suppression pool, and horizontal vents. Sixty-seven blowdown runs were made. The purpose of these tests was to determine the behavior of the hori-zontal vents and to obtain data for determining the acceleration ll of the water in the test section vents during initial clearing. 1.5-2
- -. . -- _ _ _ - __ . . . _ _ _ - _ ._~ _
GESSAR II 22A7007 238 NUCLEAR I.7 LAND Rev. 0 1.5.1.4 Verification of Pressure Suppression Design (Continued) [} This information was used to establish an analytical model for predicting vent system performance in Mark III and the resulting [ drywell pressure response. In November 1973, testing in the Mark III Pressure Suppression Test Facility (PSTF) began. The PSTP consists of an electrically heated steam generator connected to a simulated drywell which can be heated to prevent steam condensation within its volume during the simulated blowdowns. The drywell is modeled as a cylindrical vessel having a 10-ft' diameter and 26-ft height. A 6-ft diameter vent duct passes from the drywell into the suppression pool and - connects to the simulated vent system. Pool baffles are used to. simulate a scaled or full-scale sector of a Mark III suppression pool. The pool arrangement is such that both vent submergence and pc'l areas can be varied parametrically. O The full-scale PSTF testing performed between November 1973 and February 1974 obtained data for the confirmation of the analytical model. In March 1974 pool swell tests were performed in the PSTF. These full-scale tests involved air blowdown into the drywell and suppression pool to identify bounding pool-swell impact loads and breakthrough elevation (i.e., that elevation at which the water ligament begins to break up and impact loads are signifi-cantly reduced). Impact load data were obtained on selected targets located above the pool. t ! In June 1974, after the PSTF vent and pool system was converted 4 to 1/3-scale, four series of tests were performed to provide i transient data on the interaction of pool swell with flow restric-j tions above the suppression pool surface. Other areas where data
- were obtained included vent clearing, drywell pressurization, and jet forces on pool walls.
l O 4 1.5-3
GESSAR II 22R7007 238 NUCLEAR ISLAND Rev. 0 1.5.1.4 Verification of Pressure Suppression Design (Continued) The next series of 1/3-scale testing began in January 1975 and was directed at obtaining local impact pressures and total loads for typical small structures located over the pressure suppression pool including I-beams, pipes, and grating. Data from this test series expanded the data base from the full-scale air tests. A further series of 1/3-scale tests was added in June 1975 to obtain comparable data on pool-swell velocity and breakthrough elevation to the full-scale air tests. A series of small-scale flow visualization tests was performed in October 1976 in order to qualitatively investigate the steam con-densation phenomena for the Mark III vent configuration. The visual investigation of steam bubble formation and collapse under various bulk pool temperature and vent steam flux conditions pro-vided information for the placing of instrumentation in the vicin-ity of the PSTF drywell vents for subsequent tests. The final three phases of the Mark III Confirmatory Test Program began in November 1976 with a series of 1/3-scale tests under various initial suppression pool temperatures and simulated steam and liquid break sizes to obtain data on the localized conditions associated with the steam condensation portion of the LOCA blow-down. In parallel with this data acquisition, other test data were obtained for use in evaluating the loading conditions on submerged structures located in the suppression pool and for evaluating potential vertical thermal stratification of the sup-pression pool water. The second of the three phases was begun in September 1977. These full-scale tests also provided data on localized steam condensation conditions and thermal stratification. Phase three consists of a 1/9-scale test series in which a nine-vent array is utilized to evaluate multivent effects. In establishing the LOCA-related conditions within the suppression pool, all of the vent stations are conservatively assumed to be in 1.5-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.5.1.4 Verification of Pressure Suppression Design (Continued) phase even though the random nature of the phenomena indicates that some phase separation is expected during the steam condensa-tion process. This final test phase is primarily aimed at confirm-ing that multiple vent loading conditions are not in excess of these identified from single cell tests. The emphasis in some testing was directed at the evaluation of the pool-swell phenomena, while in others the steam-condensation phenomena were evaluated. Each test run consisted of a simulation of the postulated blowdown transient. Various postulated break sizes up to two times the design basis accident for the contain-ment were tested. Data was recorded at selected locations around the test facility suppression pool throughout the blowdown so that the hydrodynamic conditions associated with each phase of the blowdown was available for selecting appropriate design loading conditions. O GE has utilized this data to develop thermal and hydrodynamic loading conditions in the GE Mark III reference plant Pressure Suppression Containment System during the postulated LOCA. In-formation on thermal and hydrodynamic loading conditions during the anticipated safety / relief valve (SRV) discharge and related dynamic events has also been documented. Separate test data have been utilized to establish the SRV air-clearing load prediction model. Information on SRV discharge thermal performance is also provided. The GE reference plant report contains information and guidance to assist the containment designer in evaluating the design conditions for the various structures which form the con-tainment system. Table 1.5-1 identifies all of the LOCA-related tests conducted by GE which form the basis for hydrodynamic loads used. Subsec-tion 1.5-2 identifies the documents referenced in Table 1.5-1 i 1.5-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.5.1.4 Verification of Pressure Suppression Design (Continued) plus other reports containing test data used for non-LCCA-related hydrodynamic load definitions. 1.5.1.5 Boiling Transition Testing Since the formulation of the 1966 dench-Levy Design Limit Lines for use in BWR thermal design, GE has continued to perform exten-sive steady-state and transient boiling transitive test programs. Prior to 1974, over 14,000 data points had been obtained in water and Freon from many test assemblies having various axial heat flux profiles and rod-to-rod power distributions covering all prototypical aspects of reactor operating conditions. Among those, 2100 data points were full-scale simulation of 7x7 and 8x8 BWR fuel assemblics performed in the ATLAS Test Facility. A new boiling transition correlation (GEXL) has been developed and applied to GE BWR thermal design. Detailed information is pro-vided in the approved Licensing Topical Report NEDO-10958A, General Electric BWR Thermal Analysis Basis (GETAB): Data, Corre-lation and Design Application, January 1977. Since the implerientation of GEXL correlation on design in 1974, GE has continued to conduct full-scale 8x8 assembly boiling transition tests, accumulating over 1600 data points after GETAB introduction, to extend the data base and to assure applicability to new 8x8 fuel designs such as the two-water-rod design for BWR/ 2/5 and BWR/6. It has been shown that the 8x8 GEXL correlation with the appropriate R-factors can predict boiling transition critical power data for the two-water-rod assemblies with an accuracy typical of the GEXL correlation predictability for other 8x8 design as described in NEDO-10958-A. O 1.5-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O r3 1.5.2 PSAR Commitment Items ] The NRC staff SER on the 238 Nuclear Island, NUREG-75/110, identified matters required by the NRC and/or committed to by General Electric in the process of acquiring PDA-1. General Electric has pursued closure of these items in order to complete the requirements of PDA-1 and to assure that these items will not impact any BWR/6 operating license review. These items are cate-gorized as: (1) items requiring additional technical information, and (2) development and verification test programs. Each item has been listed and the status is given on Table 1.5-2. Approximately 75% of the items have been submitted to the staff i prior to the FDA submittal and the remainder are addressed in the FDA. A significant portion has already been reviewed and approved. 1.5.3 References
- 1. Mark II Confirmatory Test Program Phase 1 - Large Scale Demonstration Tests, October 1974, (NEDM-13377) (Proprietary Report) .
- 2. Fifth Quarterly ProgrEas Report: Mark III Confirmatory Test Program, July 1974, Supplement 1 (NEDO-20550) (Proprietary Report).
- 3. Mark III Confirmatory Test Program - Full Scale Condensation and Stratification Phenomena - Test Series 5707, August 1978 (NEDE-21853-P) (Proprietary Report) .
- 4. Mark III Confirmatory Test Program 1/3 Scale Three Vent Tests, April 1975 (NEDO-13407) (Proprietary Report) .
- 5. Mark III Confirmatory Test Program 1/3 Scale Pool Swell Impact Tests - Test Series 5805, August 1975 (NEDE-13426-P)
(Proprietary Report). O 1.5-7
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.5.3 References (Continued) g
- 6. Mark III Confirmatory Test Program 1/3 Scale Three Vent Air Tests - Test Series 5806, November 1975 (NEDE-13435-P)
(Proprietary Report).
- 7. Mark III Confirmatory Test Program - 1/d3 Scale Condensation and Stratification Phenomena - Test Series 5807, March 1977 (NEDE-21596-P) (l;roprietary Report).
- 8. Mark III Confirmatory Test Program - 1/9 Area Scale Multivent Pool Swell Tests - Test Series 6002, September 1979 (NEDE-24648P) (Proprietary Report).
- 9. Mark III confirmatory Test Program 1/9 Area Scale Condensa-tion and Stratification Phenomena, Test Series 6003, Novem-ber 1979 (NEDE-24720-P) (Proprietary Report).
- 10. Test Results Employed by GE for BWR Containment and Vertical Vent Loads, October 1975 (NEDE-21078P) (Proprietary Report) .
O O 1.5-8
p) G p)
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^3 i
l Table 1.5-1
SUMMARY
OF PSTF TEST 3 Subsec-Area tion Number Venturi Top Vent q, Initial Number Pool / 1.5.3 Test of Range Submergence Pressure Blowdown of Vent Primary Refer-Series Blowdowns (inch) Range (feet) (psia) Type Vents Scaling Objectives
- ence._
5701 21 2 1/8 - 3 5/8 2.0 - 15.5 1050 Saturated 1 Full 1. Vent 1 steam clearing
- 2. Full scale condensa-tion demo u
LJ
- 3. Drywell C) pressure 2 co o os
,H 5702 17 2 1/8 - 3 5/8 1.93 - 11.97 1050 Saturated 2 Full 1. Vent 1 {y y steam clearing >> c ;c w 5703 3 2 1/2 - 3 5/8 6.77 - 11.05 1050 Saturated 3 Full 1. Vent 1 gH steam clearing g Z 5705 4 1 - 4 1/4 6.0 - 8.0 1065 Air 2 Full 1. Pool 2 C7 swell scoping 5076 7 4 1/4 6.0 - 10.0 1065 Air 2 Full 1. Pool 2 swell
- 2. Impact loading 5707 22 2 1/8 - 3 7.5 1050 Air and 3 Full 1. Chugging 3 steam :o w" 0>
<4
- O O
O -J
Table 1.5-1
SUMMARY
OF PSTF TESTS (Continued) Subsec-Area tion Number Venturi Top Vent PL Initial Number Pool / 1.5.3 Test of Range Submergence Pressure B3cwdown of Vent Primary Refer-Series Blowdowns (inch) Range (feet) (psia) Type Vents Scaling Objectives
- ence 5801 19 2 1/8 - 3 5.0 - 10.0 1050 Saturated 3 1/3 1. 1/3-scale 4 steam demonstra-tion
- 2. Pool swell
- 3. Roof density m and AP =
cO OM g t* Cn
. 5802 3 2 1/8 - 3 6.0 1050 Saturated 3 1/3 1. Pool swell 4 M tn un >>
i steam y NN 5803 2 2 1/8 - 3 5.0 - 7.5 1050 Saturated 3 1/3 '
. 1/3-scale 4 yH liquid demo g Z
- 2. Liquid C blowdown i
5804 5 2 1/8 - 3 5.0 1050 Saturated 3 1/3 1. Roof 4 ! steam density and AP repeat-ability 5805 52 L-3 5.0 - 10.0 1050 Saturated 3 1/3 1. Pool swell 5 steam impact M NM C> '
<w
. O O
O -4 O O O
d O C ) Table 1.5-1 i
SUMMARY
OF PSTF TESTS (Continued)
- Subsec-l Area tion Number Venturi Top Vent ( Initial Number Pool / 1.5.3
- Test of Range Submergence Pressure Blowdown of Vent Primary Refer-Series Blowdowns (inch) Range (feet) (psia) Type Vents Scaling Objectives
- ence
! 5806 12 2 1/2 - 4 1/4 5.0 - 7.5 1065 Air 3 1/3 1. Pool swell 6 j 5807 20 1-3 7.5 1050 Saturated 3 1/3 1. Steam 7 steam and condensa-liquid tion I N 6002 13 2 1/8 - 3 5.0 - 10.0 1050 Saturated 9 1/9 1. Multivent 8 steam effect on co I pool swell z h,
- load CO a o tn H v en
( 6003 12 2 1/2 7.5 1050 Saturated 9 1/9 1. Multivent 9 $$ WW f steam effect on W condensa- HH h N tion loads h z O
*In general tests are not direct prototype simulations, but parametric studies to be used in analytic model evaluations.
I
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1. 5--2 COMMITMENT ITEMS Reference Where Additional Technical Information Item Discussed
- 1. Preop Piping Vil2 ration Test Program . . . . . . . . . . . . . . . Subsection 3.9.2
- 2. Reactor Internals Preep Vibration Test Program Results . . . . . Subsection 3.9.2
- 3. Dynamic Analysis of Reactor Internals and Piping . . . . . . . Section 3.9
- 4. Seismic Oualification of Class 10 Electrical Equipment . . . . . Subsection 3.9.2.2 6 3.10
- 5. Environmental Qualification of Class IE Electrical Equipment . Section 3.11
*6. Electrical isolation Devices Test Program Results . . . . . . . . Chapter 7
- 7. Fuel Experience U[date - NEDO-10505 . . . . . . . . . . . . . . . See No te A
- 8. ruel Surveillance Progran Results . . . . . . . . . . . . . . . See Note B
- 9. Fuel Assembly Cory)nents St ress Report . . . . . . . . . . . . See Note C
- 10. Puel Asserbly Pressure and Temperature Capability . . . . . See Note D
- 11. Fuel Assembly Dynamic Analysis . . . . . . . . . . . . . . See Note C
- 12. Fuel Assembly Analysis Method for Creep-Rupture . . . . . . . See Note E
- 13. Fuel Assem!ly Design Limit for Instability . . . . . . . . See Note D
- 14. ,el channel Deformation Analysis Methods . . . . . . . . . . . See Note F
- 15. Fuel Assembly Stress Limits .............. . . . See Note D 16 ruel Rud 0.000 Inch Deflection Justification . . . . . . . . . See Note D
- 17. Gadolinia Rods Performance l'xperience . . . . . . . . . . . . See Note G
- 18. Process Computer Performance Evaluation Accuracy Update . . . . See Note H
- 19. Lattice Physics Methods Verification . . . . . . . . . . . See Note I
- 20. Dolling Water Reactor Siculator Verification . . . . See Note J
- 21. Void and Dcppler Reactivity Coefficients . . . . . . . . . . See Note K
- 22. Full Pover Scram Reactivity function . . . . . . . . . . . . See Note L
- 23. Fcedwater Flow Rate Uncertainty Justification . . . . . . . . See Note M
- 24. Resolutinn of reedwater Nozzle Design and Verification . . . . See Note N
- 25. Description of WHAM Code and Loads on Internals During LOCA . . . See Note O
- 26. Safety / Relief Valve Surveillance Program Details . . . . . . . . (later)
- 27. U pda t e PCCC LTR NEDO-10466 ...... . . . . . . . . . See Note P
- 28. Analytical Methods of Plant Transient rvaluation . . . . . . . Chapter 15
- 29. ATWS . . . . . ...
. . . . . . . . . . . . . . ce P Q
- 30. Test Progran . . . for Safety / Relief Valve Solenoids . . . . . . . . See Note R
- 31. Fire Protect 2un for PCCC . . . . . . . . . . . . . . . See Note P Appendix 9A
- 32. Prinary Coolant Punp Seals Leakage Characteristics . . . . . . See Note S
- 33. Large Scale Mark III Test . . . . . . . . . . . . . . . . . . Appendix 3B
- 34. Environmental Design of Isolation Valves and Safety Related Equ2pnent . . . . . .. . . . . . . . . . . . See Note T
- 35. Post LOCA Manual Operator Actions . . . . . . . . . . . . . See Note U
*36. Instrument and Control Systens . . . . . . . . . . . . . . . Chapter 7
- 37. HPCS Onsite Electrical Systers . . . . . . . . . . . . . Chapter 8
- 38. Fire Protection for Nuclear Island Conformance . . . . . . . Appendix 9A Development and Verification Test Prograns
- 1. Fuel Surveillance Progran . . .. . . . . . . . . . . . . See Note B
- 2. Safety Relief Valve Surveillance Progran . . . . . . . . . Section 5.2.2
- 3. Core Spray Distribution . . .... . . . . . . . . . . . See Note V
- 4. Fast Scran Design Verification .. . . . . . . . . . . See No te W
- 5. Feedwater Nozzle Design Verification . . . . . . . . . See Note N
- 6. Long Tern Pipe Replacerent Prograr . . . . . . . . . . Chapter 5
- 7. Instrunentation for Vibration and Loose Parts Detection . . . X
- 8. Pressure Suppression Design Veriff . tion . . . . . . . . . Y
- 9. Suppression Pool Dynaries . . . . . . . . . . . . . . . . . . . Appendix 3D
- 10. Evaluate Effects of Relief Valve Blow-Down . . . . . . . . . . . Appendix 3B and Chapter 15 1.5-13
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.5-2 FDA COMMITMENT ITEMS (Continued) NOTES A. (1) Information Report, NEDO-20922-P, Experience with BNR Fuel through September 1974, 7/17/75 (2) Information Report, NEDO-21660-P, Experience with BWR Fuel through Decenber 1976, 12/14/77 B. (1) Letter R. Engel to D. Ross, dated 7/11/77 (2) Letter G. G. Sherwood to D. Ross, dated 4/7/78 C. Licensing Topical Report, NEDO-21175, BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown Earthquake and Loss-of-Coolant Accident Loadings D. Licensing Topical Report, NEDO-20948, BWR/6 Fuel Design, 2/20/76 E. Licensing Topical Report, NEDE-20606-A, Creep Collapse Analysis at BWR Fuel Using Safe-Colaps Model, Approved by NRC 8/76. F. Licensing Topical Report, NEDO-21354, BWR Fuel Channel Mechanical Design and Deflections G. Licensing Topical Report, NEDO-20943, Urania-Gadolinia Nuclear Fuel Physical and Irradiation Characteristics and Material Properties H. Licensing Topical Report, NEDO-20340, Process Computer Performance Evaluation Accuracy, 7/17/74 I. Licensing Topical Report, NEDO-20939-A, Lattice Physics Methods Verification, Approved by NRC 9/22/76. J. Licensing Topical Report, NEDO-20946-A, BWR Simulator Methods Verification, Approved by NRC 9/22/76 K. Licensing Topical Report, NEDO-20964, Generation of Void and Doppler Reactivity Feedback for Application to BWR Design, 2/13/76 L. Appendix A of the Odyn Report M. Letter G. G. Sherwood to E. Case, dated 5'1/78 O 1.5-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 () Table 1.5-2 FDA COMMITMENT ITEMS (Continued) NOTES (Continued) N. Information Report NEDO-21821- A Boiling Water Reactor Feed-water Nozzle /Sparger Final Report. O. Licensing Topical Report, NEDO-24048, Evaluation of Acoustic Pressure Loads on BWR/6 Internal Components, 12/1/77 P. Licensing Topical Report, NEDO-10466-A, Power Generation Control Complex Design Criteria and Safety Evaluation, Approved by NRC 7/31/78 R. (1) Letter G. G. Sherwood to E. Case, dated 2/18/78 (2) Letter S. Varga to G. G. Sherwood, MFN-183-78 (3) Information Report NEDO-23978, ADS Solenoid Valve Reliability Demonstration S. Licensing Topical Report, NEDO-24083, Recirculation Pump 7- s Shaft Seal Leakage Analysis, 12/12/78
' T. Letter G. G. Sherwood To N. Denton, dated, 10/11/78 U. Letter G. G. Sherwood to H. Denton dated 3/22/79 V. (1) NEDO-10846, Boiling Water Reactor Core Spray Distribution (2) NEDO-20566-3 Effect of Steam Environment on BWR Core Spray Distribution W. NEDO-24142, Past Scram Control Rod Drive Qualification Program X. Letter, G. G. Sherwood to E. Case, dated 3/8/78 i
Y. Licensing Topical Report, NEDO-20533, The GE Mark III Pressure l Suppression Containment System Analytical Model, approved by NRC 8/14/75 (Amended 6/30/78) O 1.5-15/1.5-16
i i GESSAR II 22A7007 r i 238 NUCLEAR ISLAND Rev. O. l SECTION 1.6 ; 1
' CONTENTS
. i l
l Section Title Page r r 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1 l TABLE ; Table Title Page 1.6-1 Referenced Reports 1.6-3 l 0 i 4 [ i i I i 1 I i s i 1.6-i/1.6-ii i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.6 MATERIAL INCORPORATED BY REFERENCE l Table 1.6-1 is a list of all GE topical reports and any other report or document which is incorporated in whole or in part by reference in this GESSAR and has been filed with the NRC. ] I l I i I O. 4 l l l l l i l
- O 1
1.6-1/1.6-2
GESSAR .II 22A7007 I 238 NUCLEAR ISLAND Rev. 0 l 1 ( Table 1.6-1 REFERENCED REPORTS Report GESSAR II Number Title Section
- 1. General Electric Company Reports APED-4827 Maximum Two-Phase Blowdown from Pipes 6.2, (April 1965) App. 3D APED-4986 Consequences of Operating Zircaloy-2 Clad 4.2 Fuel Rods Above the Critical Heat Flux (October 1965) (BWR/6 only)
APED-5458 Effectiveness of Core Standby Cooling 5.4 Systems for General Electric Boiling Water Reactors (March 1968) APED-5460 Design and Performance of General 3.9 Electric BWR Jet Pumps (July 1968) APED-5555 Impact Testing on Collect Assembly for 4.6 Control Rod Drive Mechanism 7RDB144A (November 1967) (~~ (m,) APED-5640 Xenon Considerations in Design of Large 4.1, 4.3 Boiling Water Reactors (June 1968) APED-5706 In-Core Neutron Monitoring System for 7.6, 7.7, General Electric Boiling Water Reactors App. ISB (November 1968, Revised April 1969) APED-5750 Design and Performance of General Electric 5.4 Boiling Water Reactor Main Steam Line Isolation Valves (March 1969) APED-5756 Analytical Methods for Evaluating the 15.4, 15.7 Radiological Aspects of General Electric Boiling Water Reactors (March 1976) GEAP-4616 Two-Phase Pressure Drop in Straight Pipes 4.4 and Channels; Water-Steam Mixtures at 600 to 1400 psia (May 1964) GEAP-5620 Failure Behavior in ASTM A106B Pipes 5.2 Containing Axial Through-Wall Flaws (April 1968) GEAP-10546 Theory Report for Creep-Plast Computer 4.1 Program (January 1972) 1.6-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.6-1 REFERENCED REPORTS (Continued) Report GESSAR II Number, Title Section
- 1. General Electric Company Reports (Continued)
GEAP-13112 Thermal Response and Cladding Perform- 4.2 ance of an Internally Pressurized Zircaloy-Clad, Simulated BWR Bundle Cooled by Spray under Loss-of-Coolant Conditions (April 1971) KAPL-2170 hydrodynamic Stability of a Boiling 4.4 Channel (October 1961) KAPL-2208 Hydrodynamic Stability of a Boiling 4.4 Channel, Part 2 (April 1962) KAPL-2290 Hydrodynamic Stability of a Boiling 4.4 Channel, Part 3 (June 1963) KAPL-3070 Hydrodynamic Stability of a Boiling 4.4 Channel, Part 4 (August 1964) KAPL-3072 Reactivity Stability of a Boiling 4.4 Reactor, Part 1 (September 1964) KAPL-3093 Reactivity Stability of a Boiling 4.4 Reactor, Part 2 (March 1965) NEDO-10029 An Analytical Study of Brittle Fracture 5.3 of GE-BWR Vessel Subject to the Design Basis Accident (July 1969) NEDO-10173 Current State of Knowledge, High Per- 4.2, 11.1 formance BWR Zircaloy-Clad UO2 Fuel (May 1973) NEDO-10174 Consequences of a Postulated Fuel Block- 4.2 age Incident in a Boiling Water Reactor, Rev. 1 (October 1977) NEDE-10813-P PDA - Pipe Dynamic Analysis Program for 3.6 Pipe Rupture Movement (Proprietary) (3/73) No non-proprietary equivalent O 1.6-4
3 GESSAR II 22A7007
! 238 NUCLEAR ISLAND Rev. O Table 1.6-1 REFERENCED REPORTS (Continued)
Report GESSAR II j Number Title Section
- l. General Electric Company Reports (Continued)
NEDO-10466-A Power Generation Control Complex Design 1.5, 9.5 Criteria and Safety Evaluation ! (February 1979) NEDO-10505 Experience with BWR Fuel Through 4.2, 11.1 NEDE-10505A September 1971 (May 1972) (3/72)- NEDO-10527 Rod Drop Accident Analysis for Large Boil- 4.3, 15.4 ing Water Reactors (March 1972) Supplement 1 (July 1972), Supplement 2 (January 1973) NEDO-10585 Behavior of Iodine in Reactor Water 15.2, 15.6 During Plant Shutdown and Startup (August 1972) NEDO-10602 Testing of Improved Jet Pumps for the 3.9 BWR/6 Nuclear System (June 1972)
- NEDO-10677 Analysis of Recirculation Pump Overspeed App. 3D in a Typical GE BWR (October 1972)
NEDO-10722A Core Flow Distribution in a Modern Boil- 4.4 ing Water Reactor as Measured in Monticello (August 1976) NEDM-10735 Densification Considerations in BWR Fuel 4.2 4 Design and Performance (December 1972) (Proprietary) NEDM-10735 Supplement 1 (April 1973) (Proprietary) 4.2 NEDO-10739 Methods for Calculating Safe Test Intervals 6.3, and Allowable Repair Times for Engineered App. 15A Safeguard Systems (January 1973) NEDO-10751 Experimental and Operational Confirmation 11.3 NEDO-20ll6 of Offgas System Design Parameters (NP) (10/73) (January 1973) NEDO-10802 Analytical Methods of Plant Transient 4.4, 15.1 Evaluations for General Electric Boiling
, Water Reactor (February 1973)
O 1.6-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.6-1 REFERENCED REPORTS (Continued) Report Gessar II Numbe r Title Section
- 1. General Electric Company Reports (Continued)
NEDM-10848 Mark III Confirmatory Test Program 1.5, Progress Report (8/73) (Proprietary) App. 3B NEDO-10871 Technical Derivation of BWR 1971 Design 11.1 Basis Radioactive Material Source Terms (March 1973) NEDO-10899 Chloride Control in BWR Coolants 5.2 (June 1973) NEDO-10958-A General Electric BWR Thermal Analysis 1.5, 4.2, Basis (GETAB) : Data, Correlation, and 4.3, 4.4, Design Application (January 1977) 15.0, 15.3 NEDO-10976 Mark III Analytical Investigation of 1.5, Small-Scale Tests Progress Report (8/73) App. 3B NEDO-10977 Drywell Integrity Study: Investigation 6.2 cf Potential Cracking for BWR 6/ Mark III Containment (August 1973) NEDE-lll46-P Design Basis for New Gas System 11.3 (July 1971) (Proprietary) NEDM-13377 Mark III Confirmatory Test Program 1.5, No Non- Phase I - Large Scale Demonstration App. 3B proprietary Tests (10/74) (Proprietary) Equivalent NEDM-13407-P Mark III Confirmatory Test Program - 1/3 1.5, NEDO-13407 Scale Three Vent Tests - Test Series App. 3B (6/75) 5801-5804 (May 1975) (Proprietary) NEDE-13426-P Mark III Confirmatory Test Program 1/3 1.5, NEDO-13426 Scale Pool Swell Impact Tests - Test App. 3B (8/75) Series 5805 (8/75) (Proprietary) NEDE-13435-P Mark III Confirmatory Test Program 1.5, NEDO-13435 1/3 Scale Three Vent Air Tests - Test App. 3B (10/75) Series 5806 (10/75) (Proprietary) NEDO-20210B Third Quarterly Progress Report: 1.5, NEDO-20210A Mark III Confirmatory Test Program App. 3B (NP) , (12/13 ) 1.6-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 7\ Table 1.6-1 (_,) REFERENCED REPORTS (Continued) Report Gessar II Number Title Section
- 1. General Electric Company Reports (Continued) .
NEDO-20340 Process Computer Performance Evaluation 1.5, 4.3, Accuracy (June 1974) Amendment 1 16.2 (December 1974) NEDO-20345 Fourth Quarterly Progress Report: 1.5, Mark III Confirmatory Test Program App. 3B (4/74) NEDO-20377 8x8 Fuel Bundle Development Support 4.2 (February 1975) NEDO-20533 The General Electric Mark III Pressure 1.5, 6.2, Suppression Containment System Analytical App. 3B, Model (June 1974), Supplement 1 15.2 (September 1975) NEDO-20550 Fifth Quarterly Progress Report: Mark III 1.5 gS Confirmatory Test Program (7/74) () (Proprietary) NEDE-20566 General Electric Company Model for Loss- 1.5, 3.9, NEDO-20566 of-Coolant Accident Analysis in Accordance 4.3, 6.3 (1/76) with 10CFR50, Appendix K (January 1976) (Proprietary) NEDO-20605 & Creep Collapse Analysis of BWR Fuel 1.5, 4.2 NEDE-20606-P Using Safe Collapse Model (August 1974) (Nonproprietary and Proprietary Versions) NEDM-20609-1 . Liquid Discharge Doses - LIDSR Code 15.7 (August 1976) NEDE-20732-P Seventh Quarterly Progress Report: 1.5, NEDO-20732 Mark III Confirmatory Test Program App. 3B (12/74) (12/74) (Proprietary) NEDE-20853-P Eighth Quarterly Progress Report: 1.5, NEDO-20853 Mark III Confirmatory Test Program App. 3B (4/75) (4/75) (Proprietary) NEDO-20913-A Lattice Physics Method (February 1977) 4.3 NEDE-20913PA (2/77) p t N_Y 1.6-7 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.6-1 REFERENCED REPORTS (Continued) Report GESSAR II Number Title Section
- 1. General Electric Company Reports (Continued)
NEDO-20922 Experience with BWR Fuel Through 1.5, 11.1 NEDE-20922P September 1974 (June 1975) (6/75) NEDO-20939-A Lattice Physics Methods Verification 1.5, 4.3 (January 1977) NEDE-20942-P Safety Relief Valve Discharge Analytical App. 3B NEDO-20942 Models (May 1975) (Proprietary) (5/79) NEDE-20943-P Urania-Gadolinia Nuclear Fuel Physical 1.5, 4.2 NEDO-20943 and Irradiation Characteristics and (1/77) Material Properties (January 1977) (Proprietary) NEDE-20944-1 BWR/4 and BWR/S Fuel Design 4.2 NEDO-20944-1 (January 1977) (Proprietary) (1/77) NEDO-20946A BWR Simulator Methods Verification 1.5, 4.3 (January 1977) NEDE-20948 BWR/6 Fuel Design (February 1976) 15, 4.2 NEDE-20948-P (Proprietary) (6/76) NEDO-20953-A Three Dimensional Boiling Water Reactor 4.3, 4.4, Core Simulator (January 1977) 1.54, App. 15B NEDO-20964 Generation of Void and Doppler Reactivity 1.5, 4.3 Feedback for Application to BWR Plant Transient Analysis (December 1975) NEDO-20994 Peach Bottom Atomic Power Station Units 4.4 2 & 3 Safety Analysis Report for Plant Modifications to Eliminate Sianificant In-Core Vibration (September 1975) NEDE-21056-1 NGG SJAE Offgas Treatment 9ystem 11.3
-P (August 1978) (Proprietar, NEDO-21056-1 -P (8/78) 1.6-8
- _ . ._ - ~ .-
.I 4 ,
J t GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.6-1 Os REFERENCED REPORTS (Continued) Report 4* , GESSAR II Number Title Section
- 1. General Electric Company References (Continued) ,
NEDE-21062-P Comparison of Safety-Relief Valvi App. 3B NEDO-21062 Predictions with Test Data (July 1975) g (7/75) (Proprietary) 4 - NEDE-21078-P Test Results Employed by GE for BWR 'l.5, NEDO-21078 Containment and Vertical Vent Loads ' App. 3B (10/75) (10/75) (Proprietary) , , NEDO-21142 Realistic Accident' Analysis-for General 15.2, 15.4, Electric Poiking Water Reactor - The . 15.6, 15.7 RELAC code and User's Guide (January 1978)^ NEDO-21143 Conservative Radiological Accident 1 .4, 15.6, Evaluasion the CONACOl Code (March 1976) 15.7 NEDE-21156-P Supplemental Information for Planti 4.4 Modification to Eliminate Significant In-Core Vibration (January 1976) e j
,) (Proprietary)
NEDO-21159-2 Airborne Release from EWR's from 11.1, 12.2 Environment Impact Evaluations i (October 1978) , NEDE-21175-P BWR/6 Fuel Assembiy Evlu'ation of Combined 1.b'l'3.i)~,/ NEDO-21175 Safe Shutdown Earthquake (SSE) and Loss-of- 4,R, ' ,f . (11/76) Coolant Accident (LOCA) Loadings , (November 1976) (Proprietary) g . , m _ c-NEDO-21231 Banked Position Withdrawal Sequence 4.3, 15.4, (January 1977) 15.5
,. ,/
4 NEDE-21354-P BWR Fuel Channel Mechanical Design and 1.5, 3 . 9 ,- NEDO-21354 Deflection (September 1976) (Proprietary) 4.2 1~ , (9/76) ,) g, NEDO-21424 238 Nuclear Island Containment Bypass 1.8 Leakage Sealing and Testing Methods (October 1976) NEDE-21471-P Analytical Model for Estimating Drag App. 3B , NEDO-21471 Forces on Rigid Submerged Structures (9/77) Caused by LOCA and Safety / Relief Valve Ramshead Air Discharges (September 1977) / O\ (Proprietary) ' l.6-9 ,
, ._ - - _ - 's-
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.6-1 REFERENCED REPORTS (Continued) Report GESSAR II Number Title Section
- 1. General Electric Company References (Continued)
NEDO-21471-01 Analytical Model for Estimating Drag App. 3B Forces cn Rigid Submerged Structures Caused by LOCA and Safety / Relief Valve Ramshead Air Dischargers (October 1979) NEDE-21472-P Analytical Model for Liquid Jet Properties App. 3B NEDO 21472 for Predicting Forces on Rigid Submerged (9/77) Structures (9/77) (Proprietary) NEDO-21506 Stability and Dynamic Performance of the 4.1, 4.4 General Electric Boiling Water Reactor (January 1977) NEDE-21526-P SCAM - Subcompartment Analysis Method 6.2 No Non- (January 1977) (Proprietary) proprietary Equivalent! NEDE-21564-P Stability Criteria for Primary Metal Con- App. 3F NEDO 21564- tainment Vessels under Static and Dynamic I (8/77) Loads (January 1977) (Proprietary) NEDE-21596-P Mark III Confirmatory Test Program 1/3 1.5, NEDO 21596 Scale Condensation and Stratification App. 3B (3/77) Phenomena - Test Series 5807 (3/77) (Proprietary) NEDE-21606-P Mark III One-Third Area Scale Submerged App. 3B NEDO-21606 - Structure Tests (October 1977) (10/77) (Proprietary) NEDO-21617-A Analog Transmitter / Trip Unit System for 7.1, 1:ngineered Safeguard Sensor Trip Inputc App. 3B (December 1978) NEDO-21660 Experience with BWR Fuel through Decem- 1.5, 4.2, NEDE-21660-P ber 1976 (July 1977) 11.1 (7/77) NEDO-21778-A Transient Pressure Rises Affecting Frac- 5.3 ture Toughness Requirements for Boiling Water Reactors (January 1979) NEDE-21821-A Boiling Water Reactor Feedwater Nozzle / 5.3, 1.5 l NEDO-21821-A Sparger, Final Report (February 1980) (2/80) (Proprietary) 1.6-10
- -- . . - = - - .- .-
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i j. Table 1.6-1 , REFERENCED REPORTS (Continued) ! Report GESSAR II j Number Title Section
- 1. General Electric Company Reports (Continued) i NEDE-21851-P Chugging Parametric Test Report - Small App. 3B NEDO-21851 Scale (6/78) (Proprietary)
(6/78) NEDE-21853-P Mark III Confirmatory Test Program Full 1.5 NEDO-21853 Scale Condensation and Stratification (8/78) Phenomena - Test Series 5707, (8/78) (Proprietary) NEDE-21878-P Mark I Containment Program Analytical App. 3B i NEDO-21878 Model for Computing Air Bubble.and Bound-(1/79) ary Pressures Resulting from an S/RV Dis-charge Through a T-Quencher (Proprietary) (January 1979) NEDE-23014 HEX 01 Users Manual (July 1976) 15.2 No.Non- (Proprietary) proprietary O Equivalent NEDE-23542 Fuel Assembly Evaluation of Shipping and 4.2 No Non- Handling Loadings (March 1977) proprietary Equivalent NEDE-23610-P Analytical Model for Estimating Drag App. 3B ' Forces on Rigid Submerged Structures Caused by Condensation oscillation and Chugging (May 1977) (Proprietary) NEDO-23649 Application of Pipe Break Criteria for 3.6 l Major Piping Systems Inside Containment t for the BWR/6 218, 238, and 251 Mark III l Product Line Plants (August 1977) NEDE-23786-1 Fuel Rod Prepressurization, Amendment 1 4.2
-P (May 1978) (Proprietary and Nonproprietary)
NEDO-23786-1 i NEDO-23909 Control Room Accident Exposure Evaluation 15.6 CRDS Program NEDO-23978 ADS Solenoid Valve Reliability Demon- 1.5 stration (October 1978)
)
4 1.6-11
- - - - - -.-------c r .,, - - , - - . . , , , -, -~-,-,,.n.-,m, ~ - - ---,--e , . , e . -- ----.,,m,-- -,a,v--
+
GESSAR II 22A7007 2 38 NUCLEAR ISLAND Rev. O Table 1.6-1 REFERENCED REPORTS (Continued) Report GESSAR II Number Title Section
- 1. General Electric Company Reports (Continued)
NEDE-240ll-P General Electric Boiling Water Reactor 4.2, 15.4
-A-1 Generic Reload Fuel Application (July NEDO-240ll-A 1979) (Procrietary) -1 (7/79)
NEDO-24048 Evaluation of Acoustic Pressure Loads on 1.5 BWR/6 Internal Components (September 1978) NEDE 24057-P Assessment of Reactor Internals Vibration 3.9
-A in BWR/4 and BWR/S Plants (April 1981)
NEDO 24057-A (Proprietary) (4/81) NEDO-24083 Recirculation Pump Shaft Seal Leakage 1.5 Analysis (December 1978) NEDO-24142 Fast Scram Control Rod Drive Qualifica- 1.5 tion Program (October 1978) g NEDO-24154 Qualification of One-Dimensional Core 4.3, 15.1 Transient Model for BWR (October 1978) NEDO-24548 Technical Description Annulus Pressuriza- 6.2 tion Load Adequacy Evaluation (January 1979) NEDE-24645-P Analysis of Full-Scale Test Facility for App. 3B Condensation Oscillation Loading (July 1979) (Proprietary) NEDE-24648-P Mark III Confirmatory Test Program - 1.5, NEDO-24648 1/9 Area Scale Multivent Pool Swell App. 3B Tests - Test Series 6002, (9/79) (Proprietary) NEDE-24720-P Mark III Confirmatory Test Program, 1.5 1/9 Area Scale Condensation and Strati-fication Phenomena, Test Series 6003 (1/80) (Proprietary) NEDE-24757-P Caorso Safety / Relief Valve Discharge Tests, App. 3B NEDO-24757 Phase I Test Data (May 1979 ( (Proprietary) (7/80) O 1.6-12
i- [ 22A7007
~
GESSAR II _ 238 NUCLEAR ISLAND Rev. O
- Table 1.6-1 I REFERENCED REPORTS. (Continued)
Report GESSAR II ) Number Title Section L. General Electric Company Reports (Continued) NEDE-25100-P Caorso Safety /Relier Valve Discharge App. 3B NEDO-25100 Tests, Phase I Test Data (May 1979) (8/79) (Proprietary) j NEDO-25153 Analytical Model for Estimating Drag ll.3B
- Forces on Rigid Submerged Structures 1
Caused by Steam Condensation and Chug-ging (July 1979)
- 2. Other Referenced Reports
{ AE-RTL-788 Void Measurements in the Region of Sub- 4.4
- cooled and Low Quality Boiling (April 1966)
ANL-5522 The Effect of Pressure on Boiling 4.4
- Density in Multiple Rectangular Channel j (February 1956)
ANL-5621 Boiling Density in Vertical Rectangular 4.4 i Multichannel Sections with Natural Circu-lation (November 1956) ANL-6385 Power-to-Void Transfer Functions 4.4 (July 1961)
- BRH/ DER 70-1 Radiological Surveillance Studies at a 11.1 Boiling Water Nuclear Power Reactor (March 1970)
BMI-ll63 Vapor Formation and Behavior in Boiling 4.4 Ileat Transfer (February 1957) BC-TOP-9 Design of Structures for Missile Impact 3D (October 1972 and July 1973) CF 59-6-47 Removal of Fission Product Gases from 11.3 (ORNL) Reactor Off-Gas Streams by Adsorption IDO-ITR-105 The Response of Waterlogged UO2 Fuel Rods 4.2 - to Power Bursts (April 1969) l ! IN-ITR-lll The Effects of Cladding Material and Heat 4.2 l Treatment on the Response of Waterlogged UO2 Puel Rods to Power Bursts (January 1970) 1.6-13 : _ . . - _ _ _ . _ - _ . _ . _ _ _ _ _ , _ _ _ _ _ . . _ _ . . . _ . _ _ _ _ . _ . . _ _ _ _ . _ _ . . . . _ . . _ _ =
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.6-1 REFERENCED REPORTS (Continued) Report GESSAR II Number Title Section
- 2. Other Referenced Reports (Continued)
STL-372-38 Kinetic Studies of Heterogeneous Water 4.4 Reactors (April 1966) TID-4500 Relap 3 - A Computer Program for Reactor 3.6 Blowdown Analysis IN-1321 (June 1970) TID-7672 ANS Topical Meeting, Nuclear Performance 4.3 of Power Reactors (September 1976) WACP-6065 Melting Point of Irradiated Uranium 4.2 Dioxide (February 1965) WAPD-BT-19 A Method of Predicting Steady-State 4.4 Boiling Vapor Fractions in Reactor Coolant Channels (June 1960) WAPD-TM-283 Effects of High Burnup on Zircaloy- 4.2 Clad, Bulk UO2 Plate Fuel Element Samples (September 1962) WAPD-TM-629 Irradiation Behavior of Zircaloy-Clad 4.2 Fuel Rods Containing Dished End UO2 Pellets (July 1967) O 1.6-14
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 SECTION 1.7 (} CONTENTS Section Title Page 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1.7-1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7-1 1.7.2 Piping and Instrumentation Diagrams' l.7-1 1.7.3 Abbreviations and Symbols 1.7-1 TABLES Table Title Page 1.7-1 P&ID and Electrical System Logic Diagrams 1.7-3 1.7-2 Abbreviations 1.7-13 ILLUSTRATIONS Figure Title Page 1.7-1 Electrical Drawing Symbols 1.7-23 1.7-2 Nuclear Island Logic Diagram Symbols 1.7-25 1.7-3 Nuclear Island Elementary Diagram Symbols and Standard Logic Modules 1.7-26 1.7-4 NI Piping and Instrumentation Flow Diagram Symbols 1.7-28 1.7-5 Nuclear Island Structural Drawing Symbols and Abbreviations 1.7-29 , 1.7-6 Nuclear Islar.d Piping Layout Drawing Symbols and Abbreviations 1.7-30 1.7-7 Architectural Drawing Symbols and Abbreviations 1.7-31 1 l 1.7-i/1.7-ii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O v 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1.7.1 Electrical, Instrumentation, and Control Drawings Elementary diagrams are listed and provided in Appendix 7A. System Logic diagrams and IEDs are listed in Table 1.7-1. Fig-ures 1.7-la and 1.7-lb define the symbols used on electrical drawings; Figures 1.7-2, 1.7-3a, and 1.7-3b define the symbols used on the instrumentation and control drawings. 1.7.2 Piping and Instrumentati n Diagrams Table 1.7-1 contains a list of system P&I diagrams provided in this GESSAR; Figure 1.7-4 defines symbols used on, piping and instru-mentation diagrams. 1.7.3 Abbreviations and Symbols O Table 1.7-2 is a list of the abbreviations used in this GE Standard Safety Analysis Report (GESSAR). Figure 1.7-5 defines symbols and abbreviations used on structural drawings. Figure 1.7-6 defines symbols and abbreviations used on piping lay-out drawings. Figure 1.7-7 defines symbols and abbreviations used on architec-tural drawings. O
- 1. 7-1/1. 7-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.7-1 V P&ID AND ELECTRICAL SYSTEM LOGIC DIAGRAMS GESSAR NI Figure Drawing Dwg. No. No. Rev. Title Type 4.6-Sa K-104A 5 Control Rod Drive Hydraulic System P&ID 4.6-5b K-104B 6 Control Rod Drive Hydraulic System P&ID 4.6-Sc K-104C 5 Control Rod Drive Hydraulic System P&ID 5.1-3a K-102A 7 Nuclear Boiler System P&ID 5.1-3b K-102B 7 Nuclear Boiler System P&ID 5.1-3c K-102C 7 Nuclear Boiler Systeni P&ID 5.1-3d K-102D 7 Nuclear Boiler System P&ID P 5.2-11 768E324C 3 Nuclear Boiler System D 5.2-15a K-lllA 2 Leak Detection System P&ID 5.2-15b K-lllB 2 Leak Detection System P&ID n 5.2-15c K-lllC 2 Leak Detection System P&ID ( l s.J 5.4-2a K-103A 5 Reactor Recirculation System P&ID 5.4-2b K-103B 5 Reactor Recirculation System P&ID 5.4-2c K-103C 5 Reactor Recirculation System P&ID 5.4-8a K-110A 6 Reactor Core Isolation Cooling P&ID System 5.4-8b K-110B 6 Reactor Core Isolation Cooling P&ID System 5.4-12a K-107A 8 Residual Heat Removal System P&ID 5.4-12b K-107B 8 Residual Heat Removal System P&ID 5.4-12c K-107C 8 Residual Heat Removal System P&ID 5.4-15a K-ll2A 6 Reactor Water Cleanup System P&ID 5.4-15b K-112B 6 Reactor Water Cleanup System P&ID p 5.4-17a K-ll3A 6 RWCU Filter Demineralizer System P&ID
'w - 5.4-17b K-ll3B 6 RWCU Filter Demineralizer System P&ID 1.7-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.7-1 P&ID AND ELECTRICAL SYSTEM LOGIC DIAGRAMS (Continued) GESSAR NI Figure Drawing Dwg. No. No. Rev. Title Type 6.2-60 K-152 3 Suppression Pool Makeup System P&ID 6.3-8 4 High Pressure Core Spray System P&ID 6.3-9 1 Low Pressure Core Spray System P&ID 6.5-12 K-158 2 Air Positive Seal-ISO Valve P&ID 6.5-13 K-157 2 Water Positive Seal-ISO Valve P&ID 6.5-1 K-160 4 Standby Gas Treatment System P&ID 6.7-la K-156A 2 MSIV Posiuive Leakage Control P&ID System 6.7-lb K-156B 2 MSIV Positive Leakage Control P&ID System 6.8-1 K-128 5 Pneumatic Supply System P&ID 7.1-1 Block Diagram Master Trip Circuit 7.1-2 Block Diagram Calibration Unit with Remote Display 7.2-1 BWR/6 NSPS Control Power Scheme ! 7.2-2 RPS Scram Functions l 7.2-3 Typical Arrangement of Analog Channels and Logic 7.2-4 Typical Arrangement of Digital Channels and Logic 7.2-5 Typical Configuration of MSIV l Closure Reactor Trip Coincident Logic O L 1.7-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i Table 1.7-1 O. P&ID AND ELECTRICAL SYSTEM LOGIC DIAGRAMS (Continued) i GESSAR NI Figure Drawing Dwg. No. No. Rev. Title Type 7.2-6 Arrangement of Channels and Logic FCD j 7.2-7 828E317 6 Reactor Protection System IED 7.2-8 Neutron Monitoring and RPS Logic I 7.2-9 TSV Closure Reactor Trip Coincident Logic 7.3-1 828E314 8 HPCS FCD j 7.3-2 828E151 6 Nuclear Boiler System FCD 7.3-3a I-780A- 4 Hydrogen Mixing, Drywell Vac Relief LOG and b I-780B 0 and Containment Vac Relief Sys DIAG 7.3-4 828E153 6 LPCS FCD 7.3-5 828E304 5 RHR FCD 7.3-6 865E305 2 MS-PLCS FCD 7.3-7a I-650A- 3 Suppression Pool Make-up System Log [\- and b I-650B 3 DIAG 7.3-8a I-630A- 1 Standby Gas Treatment System LOG thru c I-630C 1 DIAG 7.3-9 I-790 4 Shield Annulus Return / Exhaust LOG System and Plant System and DIAG Plant Vent 7.3-10a I-720A- 3 Auxiliary Bldg ECCS Area Pressure LOG thru d I-720D 0 Control System DIAG , 7.3-lla I-760A- 4 Fuel Building HVAC Systems LOG thru c I-760C 2 DIAG 7.3-12a I-690A- 1 Air Positive Seal Isolation Valve LOG thru c I-690C 1 Leakage Control Sys DIAG t 7.3-13 I-680 1 Water Positive Seal Isol. Valve LOG Leakage Control Sys DIAG 7.3-14a I-770A- 4 DG Rooms and Switchgear Room LOG l thru c I-770C 4 Heating and Ventilating DIAG 7.3-15a I-530A- 6 Essential Service Water System LOG and b I-530B 1 DIAG 7.3-16a I-722A 2 AB Elec Areas, Corridors, Stm Tunn LOG
-s thru c I-722C 1 and Elevator Tower HVAC Systems DIAG
( ,/ 7.3-17 I-600 4 Pneumatic Supply System LOG DIAG l 1.7-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.7-1 P&ID AND ELECTRICAL SYSTEM LOGIC DIAGRAMS (Continued) GESSAR NI Figure Drawing Dwg. No. No. Rev. Title Type 7.3-18a I-710A 1 Control Building HVAC System LOG thru e I-710E 1 DIAG 7.3-19a I-610A 1 Control Building Chilled Water LOG thru C I-610C 1 System DIAG 7.4-1 828E306 4 RCIC FCD 7.4-2 762E434 8 Standby Liquid Control System FCD 7.4-3a I-170A 3 Remote Shutdown System IED 7.4-3b I-170B- 3 Remote Shutdown System IED 7.4-3c I-170C 3 Remote Shutdown System IED 7.4-3d I-170D 3 Remote Shutdown System IED 7.4-3e I-170E 3 Remote Shutdown System IED 7.4-3f I-170G 3 Remote Shutdown System IED 7.4-3g I-170G 2 Remote Shutdown System IED 7.4-3h I-170H 0 Remote Shutdown System IED 7.6.1 828E300 5 Neutron Monitoring System IED 7.6-2 828E316 7 Neutron Monitoring System FCD 7.6-5 Block Diagram - IRM Channel 7.6-6 828E300C 5 Neutron Monitoring System IED DATA 7.6-9 APRM Block Diagram 7.6-10a I-200A 3 Proces Radiation Monitoring IED System NSSS 7.6-10b I-200B 3 Process Radiation Monitoring IED System NSSS 7.6-10c I-200C 3 Process Radiation Monitoring IED System NSSS 7.6-lla I-202A 3 Process Radiation Monitoring IED System - Nuclear Island 7.6-llb I-202B 3 Process Radiation Monitoring IED System - Nuclear Island 1.7-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.7-1 (~} v P&ID AND ELECTRICAL SYSTEM LOGIC DIAGRAMS (Continued) GESSAR NI Figure Drawing Dwg. No. No. Rev. Title Type 7.6-12 851E256 3 Leak Detection System IED 7.6-13 Area Temperature Monitoring Block Diagram 7.6-14a I-740A 1 Containment Cooling, Pressure LOG Control & Purge System DIAS 7.6-14b I-740B 1 Containment Cooling, Pressure LOG Control & Purge System DIAG 7.6-14c I-740C 1 Containment Cooling, Pressure LOG Control & Purge System DIAG 7.6-14d I-740D 0 Containment Cooling, Pressure LOG Control & Purge System DIAG
/~ 7.6-15 I-220 2 Containment Atmosphere Monitoring IED
(_], System 7.6-16 I-670 Suppression Pool Temperature IED Monitoring System ! 7.7.2 762E429 8 Control Rod Drive Hydraulic FCD System 7.7-3b 865E916 Rod Control and Information BLOCK o System Operation DIAG 7.7-5 866E304 1 Reactor Recirculation System FCD 7.7-5b 865E352 1 Reactor Recirculation System IED 7.7-6 828E171CE 2 Feedwater Control System IED Block Diagram - SRC Channel 8.3-20 828E546 4 NSPS Power Supply LOGIC DIAG
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- O .
1.7-7 1 I
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.7-1 P&ID AND ELECTRICAL SYSTEM LOGIC DIAGRAMS (Continued) GESSAR NI Figure Drawing Dwg. No. No. Rev. Title Type 9.1-24a K-ll5A 5 FPCCU System P&ID 9.1-24b K-115B 5 FPCCU System P&ID 9.1-24c K-ll5C 5 FPCCU System P&ID 9.1-25a K-116A 5 FPCCU Filter /Demineralizer System P&ID 9.1-25b K-ll6B 5 FPCCU Filter /Demineralizer System P&ID 9.2-la K-121A 8 ESW System, Division 1 P&ID 9.2-lb K-121B 8 ESW System, Division 2 P&ID 9.2-2 K-122 7 HPCS Service Water System P&ID Division 3 9.2-3a K-120A 6 Closed Cooling Water System P&ID 0 9.2-3b K-120B 6 Closed Cooling Water System P&ID 9.2-4a K-124A 7 Demineralized Water & Condensate Flow Distribution Diag 9.2-4b K-124B 5 Demineralized Water & Condensate Flow Distribution Diag 9.2-4c K-124C 4 Demineralized Water & Condensate Flow Distribution Diag l l 9.2-5 K-126 5 Potable Water System P&ID 9.2-6 K-051 0 Sanitary Water System P&ID 9.2-7a K-123A 6 Drywell Chilled Water System P&ID 9.2-7b K-123B 6 Drywell Chilled Water System P&ID l 9.2-8a K-129A 5 CB Chilled Water System P&ID 9.2-8b K-129B 5 CB Chilled Water System P&ID 9.2-9a K-125A 5 RI Chilled Water System P&ID 9.2-9b K-125B 5 RI Chilled Water System P&ID 1.7-8
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Table 1.7-1
)
P&ID AND ELECTRICAL SYSTEM LOGIC DIAGRAMS (Continued) GESSAR NI Figure Drawing Dwg. No. No. Rev. Title Type 9.2-9c K-125C 5 RI Chilled Water System P&ID 9.2-10a K-127A 6 Heated Water Distribution System P&ID
; 9.2-10b K-127B 1 Heated Water Distribution System P&ID L
l 9.3-1 K-131 6 Service Air Distribution System P&ID
- 9.3-2a K-130A 3 Instrument Air Distribution System P&ID 9.3-2b K-130B 3 Instrument Air Distribution System P&ID 9.3-2c K-130C 3 Instrument Air Distribution System P&ID 9.3-2d- K-130D 3 Instrument Air Distribution System P&ID 9.3-2e K-130E O Instrument Air Distribution System P&ID 9.3-5 K-105 5 Standby Liquid Control System P&ID 9.4-la K-162A 2 Control Building HVAC P&ID 9.4-lb K-162B 2 Control Building HVAC P&ID 9.4-lc K-162C 0 Control Building HVAC P&ID 9.4-2 K-169 5 FB HVAC Systems P&ID 9.4-3 K-163 5 Auxiliary ECCS Area Pressure Control P&ID i
9.4-4a K-164A 6 Auxiliary Building Electrical P&ID l Area, Steam Tunnel ! 9.4-4b K-164B 0 Auxiliary Building Electrical P&ID Area, Steam Tunnel 9.4-5 K-166 5 Drywell Cooling System P&ID 9.4-6 K-165A 4 Cont. Cool, Pressure Control & P&ID Purge System 9.4-7 K-165B 2 Cont. Cool, Pressure Cnntrol & P&ID Purge System 9.4-8 K-168 4 H Mixing & Vacuum Relief 2 P&ID 1.7-9
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.7-1 P&ID AND ELECTRICAL SYSTEM LOGIC DIAGRAMS (Continued) GESSAR NI Figure Drawing Dwg. No. No. Rev. Title Type 9.4-9 K-167 5 Shield Annulus Return & Exhaust P&ID System 9.4-10 K-171 4 Radwaste Building IIVAC P&ID 9.4-ila K-170A 0 Diesel Generator Building Ileat & P&ID Vent 9.4-11b K-170B 0 Diesel Generator Building IIcat & P&ID Vent 9.5-1 K-151A 5 Wet Standpipe Fire Protection P&ID System 9.5-2 K-151B 3 Wc t Standpipe Fire Protection P&ID System 9.5-3 K-153 3 Carbon Dioxide Fire Protection P&ID System 9.5-10 K-132 4 Diesel Fuel Transfer System P&ID 9.5-11 K-133 1 Division 3 Diesel Generator Fuel P&ID Transfer System 9.5-12 K-136 1 Diesel Generator Jacket Water P&ID System 9.5-13 K-137 1 Division 3 Diesel Generator Jacket P&ID Water System 9.5-14 K-138 1 Diesel Generator Starting Air P&ID System 9.5-15 K-139 3 Diesel Generator Starting Air P&ID System Division 3 9.5-16 K-134 1 Diesel Generator Lube Oil System P&ID 9.5-17 K-135 1 Diesel Generator Lube Oil System P&ID Division 3 9.5-18 K-172 1 Suppression Pool Cleanup System P&ID 11.2-2 Liquid Waste Management System P&ID ll.2-3a Kil8A 4 CRW Drain System P&ID ll.2-3b Kll8B 4 CRW Drain System P&ID 1.7-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Pev. 0 () Table 1.7-1 P&ID ELECTRICAL SYSTEM LOGIC DIAGRAMS (Continued) GESSAR NI Figure Drawing Dwg. No. No. Rev. Title Type ll.2-3c Kll8C 4 CRW Drain System P&ID ll.2-4a K119A 4 DRW Drain System P&ID ll.2-4b K119B 4 DRW Drain System P&ID 11.2-4c K119C 4 DRW Drain System P&ID 11.2-4d K119D 3 DRW Drain System P&ID ll.2-4e K119E 4 DRW Drain System P&ID 11.2-5 K-155 4 Detergent Drain System P&ID 11.3-2 7 Offgas System P&ID O i i l O 1.7-11/1.7-12
GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 l () Table 1.7-2 ABBREVIATIONS AB Auxiliary Building ac Alternating Current ACI American Concrete Institute ACRS Advisory Committee for Reactor Safeguards ADS Automatic Depressurization System AISC American Institute of Steel Construction ALARA As Low As Reasonably Achievable ANS American Nuclear Society ANSI American National Standards Institute APED Atomic Power Equipment Department (GE) APRM Average Power Range Monitor APSS Air Positive Seal System ARM Area Radiation Monitor ASCE American Society of Civil Engineer ASME American Society of Mechanical Engineers ASTM American Society for Testing Materials AWS American Welding Society BOC Beginning of Cycle BOL Beginning of Life BOP Balance of Plant Btu /hr British thermal units per hour BWR Boiling Water Reactor cal /gm Calories per gram CB Control Building CCW Closed Cooling Water CFR Code of Federal Regulations 4 cfs Cubic feet per second CHF Critical Heat Flux Ci/yr Curies per year i CLOC Closed Loops Outside Containment CMFA Common Mode Failure Analysis CP Construction Permit CPR Critical Power Ratio 1.7-13 J
. . . . - , .. -- - ,,. .- .. - ~. --
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.7-2 ABBREVIATIONS (Continued) CRDS Control Rod Drive System CRPI Control Rod Position Indicator CRVICS Containment and Reactor Vessel Isolation Control System CRW Clean Radioactive Waste CVCS Chemical and Volume Control System DBA Design Basis Accident dc Direct current AT Differential Tenporature DG Diesel Generator DOP Dioctyl Pthalate DRW Dirty Radioactive Waste EBS Emergency Boration System ECA Engineering Change Authorization ECCS Emergency Core Cooling System ECN Engineering Change Notice EFCV Excess Flow Check Valve EHC Blectro-Hydraulic Control EOC End of Cycle EOL End of Life ESF Engineered Safety Feature ESWS Essential Service Water System FA Full Arc (mode of TCV operation) FB Puel Building FCD Functional Control Diagram FDA Final Design Approval FDDR Field Deviation Disposition Request FDI Field Disposition Instruction FHA Fuel Handling Accident PLECHT Full Length Emergency Cooling Heat Transfer FMEA Failure Modes and Effects Analysis FO Fuel Oil O 1.7-14
I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.7-2 ABBREVIATIONS (Continued) FPCCU Fuel Pool Cooling and Cleanup FPS Fire Protection System PSAR Final Safety Analysis Report GDC AEC General Design Criteria GE General Electric HCU Hydraulic Control Unit HEPA High Efficiency Particulate Air / Absolute - Referring to Filters HPCS High Pressure Core Spray HX Heat Exchanger H&V Ileating and Ventilating HVAC Heating, Ventilating and Air Conditioning HWL liigh Water Level IAC Interim Acceptance Criteria (NRC) IDS Instrument Data Sheet IED Instrumentation and Equipment Diagram , IEEE Institute of Electrical and Electronics Engineers
) ILRT Integrated Leakage Rate Test-IRM Intermediate Range Monitor ISI Inservice Inspection IVLCS Isolation Valve Leakage Control System K Effective multiplication factor eff KV Kilovolts KW Kilowatt LCO Limiting Condition for Operation LCS Leakage Control System LDS Leak Detection System LHGR Linear Heat Generation Rate LO Lube Oil LOCA Loss of Coolant Accident LOPP Loss of Preferred Power LOOP Loss of Offsite Power LPCI Low Pressure Coolant Injection I
LPCS Low Pressure Core Spray LPRM Local Power Range Monitor LPZ Low Population Zone 1.7-15
GESSAR II 22A7007 238 NUCLEAR ISLAND R;v. O Table 1.7-2 ABBREVIATIONS (Continued) LSSS Limiting Safety System Setting LWL Low Water Level MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCC Motor Control Center MCHFR Minimum Critical Heat Flux Ratio MCPR Minimum Critical Power Ratio MG Motor-Generator Set MLD Mean Low Water Datum MLHGR Maximum Linear Heat Generation Pate MM Modified Mercalli (earthquake intensity) MOV Motor-operated Valve MPC Maximum Permissible Concentration Mrem Millirem MSDL Main Steam Drain Line MSL Main Steam Line MSL Mean Sea Level MSPLCS Main Steam Positive Leakage Control System MSIV Main Steam Isolation Valve MW Meg watts electrical e MW t Megowatts thermal NB Nuclear Boiler NBR Nuclear Boiler Rated (power) NBS National Bureau of Standards NBS Nuclear Boiler System NDT Nil-Ductility Transition NED Nuclear Energy Division (GE) NFPA National Fire Protection Assn. NI Nuclear Island NMS Neutron Monitoring System NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NSSSS Nuclear Steam Supply Shutoff System NSOA Nuclear Safety Operational Analysis 1.7-16
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 () Table 1.7-2 ABBREVIATIONS (Continued) OBE Operating Basis Earthquake OT Operational Transient PA Public Address (System) PCIOMR Preconditioning Cladding Interim Operating Management Recommendation PCS Process Computer System PCT Peak Cladding Temperature PCV Pressure Control Valve PDA Preliminary Design Approval PGCC Power Generation Cont ? Complex PMF Probable Maximum Flood P&ID Piping and Instrumentation Drawing POC Product of Combustion PRM Power Range Monitor PSAR Preliminary Safety Analysis Report () PSF PSID Pounds per square foot Pounds per square inch differential PSTF Pressure Suppression Test Facility PVS Plant Vent Stack PWR Pressurized Water Reactor RB Reactor Building RBM Rod Block monitor RCIC Reactor Core Isolation Cooling RC&IS Rod Control and Information System RCPB Reactor Coolant Pressure Boundary Rem Roentgen equivalent man RFCS Recirculation Flow Control System RH Relative Humidity i RHR Residual Heat Removal RI Reactor Island (Nuclear Island) RPCS Rod Pattern Control System l RPS Reactor Protection System i () RPIS RPV Rod Position Information System Reactor Pressure Vessel 1.7-17 t
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O Table 1.7-2 ABBREVIATIONS (Continued) RRRC Regulatory Requirements Review Committee RSO Reactor System Outline RWCS Reactor Water Cleanup System RWCU Reactor Water Cleanup System RWM Rod Worth Minimizer SACF Single Active Component Failure SAR Safety Analysis Report SBE Small Break Event SCFM Standard Cubic Ft/ Minute SCG Startup Coordinating Group SCR Silicon Controlled Rectifier SEP Single Equipment Failure SGTS Standby Gas Treatment System SLCS Standby Liquid Control System SOF Single Operator Failure SPCU Suppression Pool Cleanup SPMU Suppression Pool Makeup System SRM Source Range Monitor SRP Standard Review Plan SRSS Square Root of the Sum of the Squares SRV Safety / Relief Valve SS Safe Shutdown SSE Safe Shutdown Earthquake SW Service Water TCV Turbine Control Valve TG Turbine-Generator TIP Traversing Incore Probe TMil Trolley Mounted floist UO 2 Uranium Dioxide WG Water Gage WPSS Water Positive Seal System O 1.7-18
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