ML23291A410: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot insert)
 
(StriderTol Bot change)
 
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:SSES-FSAR 1.3 COMPARISON TABLES 1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS This subsection highlights the historical principal design features of the plant and compares its major features with other boiling water reactor facilities. The design of this facility was based on proven technology obtained during the development, design, construction, and operation of boiling water reactors of similar types.
{{#Wiki_filter:SSES-FSAR
 
1.3 COMPARISON TABLES
 
1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS
 
This subsection highlights the historical principal design features of the plant and compares its major features with other boiling water reactor facilities. The design of this facility was based on proven technology obtained during the development, design, construction, and operation of boiling water reactors of similar types.
 
The data, performance, characteristics, and other information presented here represent the then current Susquehanna Steam Electric Station design as it compared to similar designs available at that time. To maintain this original design comparison, these tables will not be revised to reflect current plant design, other than the previous addition of the fifth ("E") emergency diesel generator to Tables 1.3-6 and 1.3-7.
The data, performance, characteristics, and other information presented here represent the then current Susquehanna Steam Electric Station design as it compared to similar designs available at that time. To maintain this original design comparison, these tables will not be revised to reflect current plant design, other than the previous addition of the fifth ("E") emergency diesel generator to Tables 1.3-6 and 1.3-7.
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 rated power output for a single plant unless otherwise noted.
1.3.1.2 Power Conversion System Design Characteristics Table 1.3-2 compares the power conversion system design characteristics.
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.
1.3.1.5 Radioactive Waste Management Systems Design Characteristics Table 1.3-5 compares the radioactive waste management design characteristics.
Rev. 49, 04/96                          1.3-1


SSES-FSAR 1.3.1.6 Structural Design Characteristics Table 1.3-6 compares the structural design characteristics.
1.3.1.1 Nuclear Steam Supply System 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 design between submission of the last PSAR revision and Revision 0 of the FSAR are listed in Table 1.3-8. Each item in Table 1.3-8 is cross-referenced to the appropriate portion of the FSAR which describes the changes and the bases for them.
Table 1.3-1 summarizes the design and operating characteristics for the nuclear steam supply systems. Parameters are related to rated power output for a single plant unless otherwise noted.
Rev. 49, 04/96                       1.3-2
 
1.3.1.2 Power Conversion System Design Characteristics
 
Table 1.3-2 compares the power conversion system design characteristics.
 
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.
 
1.3.1.5 Radioactive Waste Management Systems Design Characteristics
 
Table 1.3-5 compares the radioactive waste management design characteristics.
 
Rev. 49, 04/96 1.3-1 SSES-FSAR


SSES-FSAR Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
1.3.1.6 Structural Design Characteristics
SSES                HATCH 1                       ZIMMER                GESSAR BWR 4                  BWR 4                        BWR 5                  BWR 6 251-764                218-560                      28-560                238-732 THERMAL AND HYDRAULIC DESIGN Rated power, MWt                                                3293                  2436                        2436                    3579 Design power, MWt (ECCS design basis)                            3439                  2550                        2550                    3758 Steam flow rate, lb. hr.                                    13.48 x 106            10.03 x 106                  10.477 x 106          15.396 x 106 Core coolant flow rate, lb/hr.                              100.0 x 106            78.5 x 106                  78.5 x 106            105.0 x 106 Feedwater flow rate, lb/hr.                                13.574 x 106          10.445 x 106                  10.477 x 106              15.358 System pressure, nominal in steam dome, psia                    1020                  1020                        1020                    1040 Average power density, KW/liter                                  48.7                  51.2                        50.51                  56.0 Maximum thermal output, KW/ft.                                  13.4                    13.4                        13.4                  13.4 Average thermal output, KW/ft.                                  5.34                    7.11                        5.45                  6.04 Maximum heat flux, Btu/hr-ft2                                  361,000                428,300                      354,000                354,300 Average heat flux, Btu/hr-ft2                                  144,100                164,700                      143,900                159,600 Maximum UO2 temperature, F                                      3330                  4380                        3325                    3337 Average volumetric fuel temperature, F                          1100                  1100                        1100                    1100 Average cladding surface temperature, F                          558                    558                          558                    558 Minimum critical power ratio (MCPR)                              1.23                    1.9*                        1.21                  1.24
* For Hatch minimum critical heat flux (MCHFR) ws used.
Coolant enthalpy at core inlet, Btu/lb                          521.8                  526.2                        527.4                  527.9 Core maximum exit voids within assemblies                        76                      79                          75                    76 Core average exit quality, % steam                              13.2                    12.9                        13.6                  14.9 Feedwater temperature, F                                        383                  387.4                        420                    420 THERMAL AND HYDRAULIC DESIGN Design Power Peaking Factor:
Maximum relative assembly power                              1.40                  1.40                        1.40                    1.40 Local peaking factor                                        1.15                    1.24                        1.24                  1.13 Axial peaking factor                                        1.40                  1.50                        1.40                    1.40 Total peaking factor                                        2.51                    2.60                        2.43                  2.22 NUCLEAR DESIGN (First Core)
Water/U02 volume ratio (cold)                                    2.80                    2.53                        2.41                  2.70 Reactivity with strongest control rod Keff                      F0.99                  F0.99                        F0.99                  F0.99 Moderate void coefficient:
Hot, no voids k/k - % void                              -1.0 x 10-3            -1.0 x 10-3                  -1.0 x 10-3            -0.3 x 10-3 At rated output, k/k - % void                            -1.7 x 10-3            -1.6 x 10-3                  1.6 x 10-3            -1.0 x 10-5 Fuel temperature doppler coefficient:
At 68 F, k/k - F fuel                                  -1.2 x 10-5            1.3 x 10-5                  -1.3 x 10-5            -1.6 x 10-5 Rev. 55, xx/xx                                                                                                                              Page 1 of 4


SSES-FSAR Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
Table 1.3-6 compares the structural design characteristics.
SSES                HATCH 1                      ZIMMER              GESSAR BWR 4                  BWR 4                        BWR 5                BWR 6 251-764                218-560                      28-560              238-732 Hot, no voids, k/k F fuel                                -1.2 x 10-5          -1.2 x 10-5                  -1.2 x 10-5          -1.3 x 10-5 At rated output, k/k,F fuel                              -1.2 x 10-5          -1.3 x 10-5                  -1.3 x 10-5          -1.2 x 10-5 Initial average U-235 enrichment wt. %                              1.88                  2.23                        1.90                  1.90 Fuel average discharge exposure, MWd/short ton                    16,200                19,000                      15,053                13,000 CORE MECHANICAL DESIGN (First Core)
Fuel Assembly:
Number of fuel assemblies                                      764                    560                          560                  732 Fuel rod array                                                8x8                    7x7                          8x8                  8x8 Overall dimensions, in.                                        176                    176                          176                  176 Weight of U02 per assembly lb. (pellet type)            458 (chamfered)      490.4 (undished)                  465.15                  472 483.4 (dished)                                      (chamfered)
Weight of fuel assembly, lb.                                    600            681 (undished)                      698 675 (dished)
Fuel Rods:
Number per fuel assembly                                        62                    49                          63                    63 Outside diameter, in                                          0.483                  0.563                        0.493                0.493 Cladding thickness, in                                        0.032                  0.032                        0.034                0.034 Gap, pellet to cladding, in                                  0.0045                0.006                        0.0045                0.009 Length of gas plenum, in                                        10                    16                          14                    12 Cladding material*                                          Zircaloy-2            Zircaloy-2                  Zircaloy-2            Zircaloy-2
    *Free-standing loaded tubes Fuel Pellets:
Material                                                        U02                    U02                          U02                  U02 Density, % of theoretical                                        95                    95                          95                    94 Diameter, in                                                  0.410                  0.487                        0.416                0.416 Length, in                                                    0.410                    0.5                        0.420                0.420 Fuel Channel:
Overall dimension, length, in                                166.9                  166.9                        166.9 Thickness, in                                                0.080                  0.080                        0.100                0.120 Cross section dimensions, in                              5.48 x 5.48            5.44 x 5.44                  5.48 x 5.48          5.52 x 5.52 Material                                                    Zircaloy-4            Zircaloy-4                  Zircaloy-4            Zircaloy-4 Core Assembly:
Fuel Weight as UO2 lb                                      349,000                272,850                      250,538              345,500 Core diameter, (equivalent), in                                187.1                  160.2                        160.2 Core height (active fuel) in                                    150                    144                          146                  148 Rev. 55, xx/xx                                                                                                                              Page 2 of 4


SSES-FSAR Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
1.3.1.7 Instrumentation and Electrical Systems Design Characteristics
SSES                        HATCH 1                             ZIMMER                                GESSAR BWR 4                        BWR 4                              BWR 5                                  BWR 6 251-764                      218-560                              28-560                                238-732 Reactor Control System:
Method of variation of reactor power                    Movable control rods and      Movable control rods and      Movable control rods and variable      Movable control rods and variable variable forced coolant flow  variable forced coolant flow            forced coolant flow                    forced coolant flow Number of movable control rods                                      185                          137                                137                                    177 Shape of movable control rods                                    Cruciform                    Cruciform                            Cruciform                              Cruciform Pitch of movable control rods                                      12.0                        12.0                                12.0                                  12.0 Control material in movable rods                              B4C granules                B4C granules                        B4C granules                            B4C granules compacted in SS              compacted in SS                    compacted in SS                        compacted in SS tubes                        tubes                              tubes                                  tubes Type of control rod drives                              Bottom entry locking piston  Bottom entry locking piston        Bottom entry locking piston            Bottom entry locking piston Type of temporary reactivity control for initial core  Burnable poison; gadolinia- Burnable poison; gadolinia-urania Burnable poison; gadolinia-urania fuel Burnable poison; gadolinia-urania fuel urania fuel rods                fuel rods                              rods                                  rods Incore Neutron Instrumentation:
Number of incore neutron detectors (fixed)                          172                          124                                124                                    164 Number of incore detector assemblies                                43                          31                                  31                                    41 Number of detectors per assembly                                      4                            4                                    4                                      4 Number of flux mapping neutron detectors                              5                            4                                    4 Range (and number) of detectors:
Source range monitor                                  Source to 0.001% power (4)    Source to 0.001% power (4)          Source to 0.001% power (4)              Source to 0.001% power Intermediate range monitor                              0.001% to 10% power (8)      0.001% to 10% power (8)            0.001% to 10% power (8)                0.001% to 10% power Local power range monitor                                5% to 125% power (172)        5% to 125% power (124)              5% to 125% power (124)                    5% to 125% power Average power range monitor                              5% to 125% power (6)*        2.5% to 125% power (6)*            2.5% to 125% power (6)*                  2.5% to 125% power
    *Channels of monitors from LPRM detectors.
Number and types of incore neutron sources                      7 Sb-Be                      5 Sb-Be                            5 Sb-Be REACTOR VESSEL DESIGN Material                                                  Carbon steel/Stainless Clad  Carbon steel/Stainless Clad        Carbon steel/Stainless Clad            Carbon steel/Stainless Clad Design pressure, psig                                                  1250                          1265                                1250                                  1250 Design temperature, F                                                  575                          575                                575                                    575 Inside diameter, ft-in.                                                20-11                        18-2                                18-2                                  19-10 Inside height, ft-in.                                                  72-11                        69-4                                69-4                                  70-10 Minimum base metal thickness(cylindrical section) in                    6.19                        5.53                                5.375                                  5.70 Minimum cladding thicknesses, in                                        1/8                          1/8                                  1/8                                    1/8 REACTOR COOLANT RECIRCULATION DESIGN Number of recirculation loops                                            2                            2                                    2                                      2 Design pressure:
Rev. 55, xx/xx                                                                                                                                                                  Page 3 of 4


SSES-FSAR Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
Table 1.3-7 compares the instrumentation and electrical systems design characteristics.
SSES                HATCH 1                      ZIMMER                GESSAR BWR 4                  BWR 4                          BWR 5                BWR 6 251-764                218-560                        28-560              238-732 Inlet leg, psig                                              1250                  1148                          1250                1250 Outlet leg, psig                                              1500                  1274                    1675*;1575**          1675*;1575**
* Pump and discharge piping to and including discharge block valves.
    **    Discharge piping from discharge block valve to vessel Design temperature, F                                            575                    562                          575                  575 Pipe diameter, in                                                  28                    28                            20                  22/24 Pipe material, ANSI                                            304/316                304/316                      304/316                304 Recirculation pump flow rate, gpm                                45,200                42,200                        33,880              35,400 Number of jet pumps in reactor                                    20                    20                            20                  20 MAIN STEAMLINES Number of streamlines                                              4                      4                            4                    4 Design pressure, psig                                            1250                  1146                          1250                1250 Design temperature, F                                            575                    563                          575                  575 Pipe diameter, in                                                  26                    24                            24                  26 Pipe material                                                Carbon Steel          Carbon Steel                Carbon Steel          Carbon Steel Rev. 55, xx/xx                                                                                                                              Page 4 of 4


SSES-FSAR Table 1.3-2 COMPARISON OF POWER CONVERSION SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION
SSES                HATCH                    ZIMMER              GESSAR BWR 4                BWR 4                    BWR 5                BWR 6 251-764              218-560                  218-560              238-732 TURBINE GENERATOR (See Sections 10.2 and 10.4)
Rated power, MWt                                      3293                  2550                    2550
* Rated power, MWe (gross)                              1085                  813                      883
* Generator Speed, RPM                                  1800                  1800                    1800
* Rated steam flow, lb/hr                              13.4 x 106          10.48 x 106              11.0 x 106
* Inlet pressure, psig                                    965                  950                      950
* STEAM BYPASS SYSTEM (See Section 10.4.4)
Capacity, % design steam flow                            25                  25                      25
* MAIN CONDENSER (See Section 10.4.1)
Heat removal capacity, Btu/hr                        7890 x 106          5720 x 106              7053 x 106
* CIRCULATING WATER SYSTEM (See section 10.4.5)
Number of pumps                                          4                    2                        3
* Flow rate, gpm/pump                                  112,000              185,000                  150,000
* CONDENSATE AND FEEDWATER SYSTEM Design flow rate, lb/hr                            13.44 x 106          10.096 x 106            10.971 x 106
* Number of condensate pumps                                4                    3                        3
* Number of condensate booster pumps                    None                    3                        3
* Number of feedwater pumps                                3                    2                        2
* Number of feedwater booster pumps                      None                  None                    None
* Condensate pump drive                                AC Power            AC Power                AC Power
* Booster pump drive                                      NA                AC Power                AC Power
* Feedwater pump drive                                  Turbine              Turbine                  Turbine
* Feedwater booster pump drive                            NA                    NA                      NA                    *
*See applicants SAR Rev. 49, 04/96                                                                                                                Page 1 of 1


Table 1.3-3 COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS Security-Related Information Table Withheld Under 10 CFR 2.390
All of the significant changes that have been made in the facility design between submission of the last PSAR revision and Revision 0 of the FSAR are listed in Table 1.3-8. Each item in Table 1.3-8 is cross-referenced to the appropriate portion of the FSAR which describes the changes and the bases for them.


Table 1.3-4 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS Security-Related Information Table Withheld Under 10 CFR 2.390
Rev. 49, 04/96 1.3-2


Table 1.3-5 RADIOACTIVE WASTE MANAGEMENT SYSTEMS DESIGN CHARACTERISTICS Security-Related Information Table Withheld Under 10 CFR 2.390
Table 1.3-3 COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS


SSES-FSAR Table 1.3-6 COMPARISON OF STRUCTURAL DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
Security-Related Information Table Withheld Under 10 CFR 2.390 Table 1.3-4 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS
SSES                      HATCH 1                        ZIMMER                GESSAR BWR-4                      BWR 4                          BWR 5                  BWR 6 251-764                    218-560                        218-560                238-732 SEISMIC DESIGN* (See Section 3.7)
Operating Basis Earthquake
      - horizontal g                            0.05                        0.08                          0.10                    0.15
      - vertical g                              0.033                        0.05                          0.07                    ----
Safe shutdown earthquake
      - horizontal g                            0.10                        0.15                          0.20                    0.30
      - vertical g                              0.067                        0.10                          0.14                    ----
WIND DESIGN (See Section 3.3)
Maximum sustained - mph                            80                        105                            90                    130 TORNADOS (See Section 3.3)
Translational - mph                                60                          60                            60                      70 Tangential - mph                                  300                        300                          300                    290
*Some of the tabluated values differ for the design of the Diesel Generator E Facility.
Rev. 49, 04/96                                                                                                                        Page 1 of 1


Table 1.3-7 COMPARISON OF ELECTRICAL POWER SYSTEM DESIGN CHARACTERISTICS Security-Related Information Table Withheld Under 10 CFR 2.390
Security-Related Information Table Withheld Under 10 CFR 2.390 Table 1.3-5 RADIOACTIVE WASTE MANAGEMENT SYSTEMS DESIGN CHARACTERISTICS


SSES-FSAR TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*
Security-Related Information Table Withheld Under 10 CFR 2.390
FSAR PORTION IN WHICH CHANGE IS ITEM                                CHANGE                        REASON FOR CHANGE                          DISCUSSED Recirculation flow measurement      The recirculation flow measurement    To improve flow measurement          7.3.1, 7.6.1 design was changed from a flow        accuracy.
element to an elbow-tap type.
Recirculation system                The pressure interlock for RHR        NRC Requirement for diversity.        7.3.1, 7.6.1 shutdown mode was changed.
Nuclear fuel                        The number of fuel pins in each fuel  Improved fuel performance by          4.2 bundle has been changed from 7 x 7    increasing safety margins.
to 8 x 8.
Nuclear boiler                      An additional test mode was added      Verifies that the spring force on the 5.4 for closing MSIVs one at a time to    valves will cause them to close 90% of full open in the fast mode      under loss-of-air conditions.
(close in slow mode already existed).
Main steam line isolation          A main condenser low vacuum            NRC requirement                      7.3.1 initiation of the main steam line isolation was added.
Main steam line isolation          Reactor isolation was deleted for      To provide improved plant            5.4 high water level initiation actuation. availability.
Main steam line drain system        A main steam line drain system was    Prevent accumulation of condensate 5.4 improved.                              in an idle line outboard of MSLIV.
Feedwater sparger                  The thermal sleeve was changed to      To eliminate vibration, failure, and  5.3 provide improved design of sparger    leakage.
to nozzle.
Standby liquid control (SLC) system Interlocks on the SLC system were      To prevent inadvertent boron          9.3.5 and 7.4.1 revised.                              injection during system testing.
Rev. 49, 04/96                                                                                                                        Page 1 of 3


SSES-FSAR TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*
Table 1.3-7 COMPARISON OF ELECTRICAL POWER SYSTEM DESIGN CHARACTERISTICS
FSAR PORTION IN WHICH CHANGE IS ITEM                            CHANGE                      REASON FOR CHANGE                      DISCUSSED RCIC & HPCI steam supply          A warmup bypass line and valve      Permits pressurizing and pre-    5.4 and 6.3 was added.                          warming of the steam supply line downstream to the turbine during reactor vessel heatup.
RCIC & HPCI vacuum breaker        A vacuum breaker system was        To prevent backup of water in the 5.4 and 6.3 system                            added to the turbine exhaust line  pipe and consequential high into the suppression pool.          dynamic pipe loads and reactions.
RCIC & HPCI system                Each component has been made        Improved testability              5.4 and 6.3 capable of functional testing.
Automatic depressurization system The interlocks on the automatic    To meet IEEE-279 requirements. 7.3.1 (ADS)                            depressurization system were revised.
RPV code                          The RPV was partially updated to    Update to applicable code as much 5.2 ASME 1971 code and Summer 1971 as practical.
addenda.
Level instrumentation            The RPV level instrumentation was Improve ECCS separation per IEEE    7.3.1 revised to eliminate Yarway columns 279 and improve reliability.
and replace them with a conventional condensing chamber type; also, separation and redundancy features were added.
Leak detection system            The leak detection system was      To meet IEEE-279 requirements. 7.6.1 revised to upgrade the capability.
Reactor vibration monitoring      A confirmatory vibration monitoring NRC requirement                  14.2 test was added.
Rev. 49, 04/96                                                                                                              Page 2 of 3


SSES-FSAR TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*
Security-Related Information Table Withheld Under 10 CFR 2.390}}
FSAR PORTION IN WHICH CHANGE IS ITEM                                CHANGE                        REASON FOR CHANGE                    DISCUSSED Primary Containment Concrete        Delineation of compressive          Update to reflect current engineering 3.8B strengths for pozzolan vs. non-      design requirements.
pozzlan Type II Portland cements.
RPV Insulation                      Correct the RPV Insulation          Revised support beams on as-build 5.3.3.1.4 Description                          RPV Insulation Panels Safety Related Conduits & Trays    Correct separation statements for    Question 7.4 of Amend. #5 of PSAR 3.12 conduits and trays.                  (Revised per requirement of Reg.
Guide 1.75 - 1974).
Tornado Loading                    Revised Tornado Loading              To reflect latest NRC                3.3 combinations.                        recommendations in the Standard Review Plan.
*NOTE:        Design changes listed are only those which have occurred between the last SSES PSAR Amendment and Revision 0 of the FSAR.
The NRC has been notified of all other design changes prior to the last PSAR amendment by previous amendments to the PSAR.
Rev. 49, 04/96                                                                                                                    Page 3 of 3}}

Latest revision as of 09:49, 13 November 2024

1 to Updated Final Safety Analysis Report, Chapter 1, Section 1.3, Comparison Tables
ML23291A410
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 10/12/2023
From:
Susquehanna
To:
Office of Nuclear Reactor Regulation
Shared Package
ML23291A105 List: ... further results
References
PLA-8081
Download: ML23291A410 (1)
v • d • e


Text

SSES-FSAR

1.3 COMPARISON TABLES

1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS

This subsection highlights the historical principal design features of the plant and compares its major features with other boiling water reactor facilities. The design of this facility was based on proven technology obtained during the development, design, construction, and operation of boiling water reactors of similar types.

The data, performance, characteristics, and other information presented here represent the then current Susquehanna Steam Electric Station design as it compared to similar designs available at that time. To maintain this original design comparison, these tables will not be revised to reflect current plant design, other than the previous addition of the fifth ("E") emergency diesel generator to Tables 1.3-6 and 1.3-7.

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 rated power output for a single plant unless otherwise noted.

1.3.1.2 Power Conversion System Design Characteristics

Table 1.3-2 compares the power conversion system design characteristics.

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.

1.3.1.5 Radioactive Waste Management Systems Design Characteristics

Table 1.3-5 compares the radioactive waste management design characteristics.

Rev. 49, 04/96 1.3-1 SSES-FSAR

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 design between submission of the last PSAR revision and Revision 0 of the FSAR are listed in Table 1.3-8. Each item in Table 1.3-8 is cross-referenced to the appropriate portion of the FSAR which describes the changes and the bases for them.

Rev. 49, 04/96 1.3-2

Table 1.3-3 COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS

Security-Related Information Table Withheld Under 10 CFR 2.390 Table 1.3-4 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS

Security-Related Information Table Withheld Under 10 CFR 2.390 Table 1.3-5 RADIOACTIVE WASTE MANAGEMENT SYSTEMS DESIGN CHARACTERISTICS

Security-Related Information Table Withheld Under 10 CFR 2.390

Table 1.3-7 COMPARISON OF ELECTRICAL POWER SYSTEM DESIGN CHARACTERISTICS

Security-Related Information Table Withheld Under 10 CFR 2.390

Retrieved from "https://kanterella.com/w/index.php?title=ML23291A410&oldid=3693818"