ML022470190

Virgil C Summer Unit 1, Technical Specification Pages for Amendment 160 Spent Fuel Pool Expansion

ML022470190
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Issue date:	08/30/2002
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves.......................................................................................

3/4 7-1 Emergency Feedwater System.............................................................

3/4 7-4 Condensate Storage Tank....................................................................

3/4 7-6 A ctivity..................................................................................................

7-7 Main Steam Line Isolation Valves.........................................................

3/4 7-9 Feedwater Isolation Valves...................................................................

3/4 7-9a 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION.....

3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................................

3/47-11 3/4.7.4 SERVICE WATER SYSTEM................................................................

3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK........................................................................

3/4 7-13 3/4.7.6 CONTROL ROOM NORMAL AND EMERGENCY AIR HANDLING SYSTEM.

3/47-14 3/4.7.7 SNUBBERS..........................................................................................

3/47-16 3/4.7.8 SEALED SOURCE CONTAMINATION................................................

3/4 7-23 3/4.7.9 AREA TEMPERATURE MONITORING................................................

3/4 7-37 3/4.7.10 WATER LEVEL - SPENT FUEL POOL................................................

3/4 7-39 3/4.7.11 SPENT FUEL POOL VENTILATION SYSTEM.....................................

3/4 7-40 3/4.7.12 SPENT FUEL ASSEMBLY STORAGE.................................................

3/4 7-42 3/4.7.13 SPENT FUEL POOL BORON CONCENTRATION...............................

3/4 7-44 Amendment No. 23,79, 160 SUMMER - UNIT 1 Vill

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION................................................................

3/4 9-1 3/4.9.2 INSTRUMENTATION...........................................................................

3/4 9-2 3/4.9.3 DECAY TIM E.......................................................................................

3/4 9-3 3/4.9.4 REACTOR BUILDING PENETRATIONS..............................................

3/4 9-4 3/4.9.5 COMMUNICATIONS............................................................................

3/4 9-5 3/4.9.6 MANIPULATOR CRANE......................................................................

3/4 9-6 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High W ater Level.................................................................................

3/4 9-7 Low W ater Level...................................................................................

3/4 9-8 3/4.9.8 REACTOR BUILDING PURGE AND EXHAUST ISOLATION SYSTEM....

3/4 9-9 3/4.9.9 WATER LEVEL - REFUELING CAVITY AND FUEL TRANSFER CANAL.

3/4 9-10 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.........................................................................

3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS 3/4 10-2 3/4.10.3 PHYSICS TESTS.................................................................................

3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS............................................................

3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN...............................

3/4 10-5 Amendment No. 4-9 160 SUMMER - UNIT 1 X

INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYC LE.................................................................................

B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION.....

B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................................

B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM................................................................

B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK........................................................................

B 3/4 7-3 3/4.7.6 CONTROL ROOM NORMAL AND EMERGENCY AIR HANDLING SYSTEM.............................................

B 3/4 7-4 3/4.7.7 SNUBBERS..........................................................................................

B 3/4 7-4 3/4.7.8 SEALED SOURCE CONTAMINATION................................................

B 3/4 7-6 3/4.7.9 AREA TEMPERATURE MONITORING................................................

B 3/4 7-6 3/4.7.10 WATER LEVEL - SPENT FUEL POOL................................................

B 3/4 7-6 3/4.7.11 SPENT FUEL POOL VENTILATION SYSTEM.....................................

B 3/4 7-6 3/4.7.12 SPENT FUEL ASSEMBLY STORAGE.................................................

B 3/4 7-7 3/4.7.13 SPENT FUEL POOL BORON CONCENTRATION...............................

B 3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS...................................................

B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTION DEVICES.........................

B 3/4 8-3 Amendment No. 38,63, 79, 160 XIV SUMMER - UNIT 1

INDEX BASES SECTION 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION................................................................

3/4.9.2 INSTRUM ENTATIO N..........................................................................

3/4.9.3 D ECAY TIM E.......................................................................................

3/4.9.4 REACTOR BUILDING PENETRATIONS.............................................

3/4.9.5 CO M M UNICATIO NS...........................................................................

3/4.9.6 MANIPULATOR CRANE......................................................................

3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION...........

3/4.9.8 REACTOR BUILDING PURGE AND EXHAUST ISOLATION SYSTEM.....

3/4.9.9 WATER LEVEL - REACTOR VESSEL........................................................

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOW N MARGIN........................................................................

3/4.102 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS 3/4.10.3 PHYSICS TESTS................................................................................

3/4.10.4 REACTOR COOLANT LOOPS............................................................

3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN...............................

Amendment No. 49 "160 PAGE B 3/4 9-1 B 3/4 9-1 B 3/4 9-1 B 3/4 9-1 B 3/4 9-1 B 3/4 9-2 B 3/4 9-2 B 3/4 9-2 B 3/4 9-2 B 3/4 10-1 B 3/4 10-1 B 3/4 10-1 B 3/4 10-1 B 3/4 10-1 XV SUMMER - UNIT 1

PLANT SYSTEMS 3/4.7.10 WATER LEVEL-SPENT FUEL POOL LIMITING CONDITION FOR OPERATION 3.7.10 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY:

Whenever irradiated fuel assemblies are in the spent fuel pool.

ACTION:

With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.10 The water level in the spent fuel pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the spent fuel pool.

Amendment No. 160 SUMMER - UNIT 1 3/4 7-39

PLANT SYSTEMS 3/4.7.11 SPENT FUEL POOL VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.11 Two independent spent fuel pool ventilation sub-systems shall be OPERABLE with at least one sub-system in operation.

APPLICABILITY: Whenever irradiated fuel is being moved in the spent fuel pool and during crane operation with loads over the pool.

ACTION:

a.

With one spent fuel pool ventilation sub-system inoperable, fuel movement within the spent fuel pool or crane operation with loads over the spent fuel pool may proceed provided the OPERABLE spent fuel pool ventilation sub-system is capable of being powered from an OPERABLE emergency power source and is in operation and discharging through at least one train of HEPA filters and charcoal adsorbers.

b.

With no spent fuel pool ventilation sub-system OPERABLE, suspend all operations involving movement of fuel within the spent fuel pool or crane operation with loads over the spent fuel pool.

c.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11 The above required spent fuel pool ventilation sub-systems shall be demonstrated OPERABLE:

a.

At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that each sub-system operates for at least 15 minutes.

b.

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical -release in any ventilation zone communicating with the system by:

1.

Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 30,000 ACFM +/- 10%.

SUMMER - UNIT 1 Amendment No. 42, 159, "160 3/4 7-40

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM D3803-1989, at a relative humidity of 95%

and 300C with a methyl iodide penetration of <2.5%.

3.

Verifying a system flow rate of 30,000 ACFM +/- 10% during system operation when tested in accordance with ANSI N510-1975.

c.

Prior to the movement of fuel or crane operation with loads over the pool by verifying that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of ASTM D3803-1989, at a relative humidity of 95% and 300C with a methyl iodide penetration of <2.5%.

Subsequent to each initial analysis (which must be completed prior to fuel movement or crane operation with loads over the pool), during the period of time in which there is to be fuel or crane movement with loads over the pool, verify charcoal adsorber operation every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> by obtaining and analyzing a sample as described above. These subsequent analyses are to be completed within thirty one (31) days of sample removal.

d.

At least once per 18 months by:

1.

Verifying that the pressure drop across the combined HEPA and roughing filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 30,000 ACFM +/- 10%.

2.

Verifying that on a loss of offsite power test signal, the system automatically starts.

3.

Verifying that the system maintains the spent fuel pool area at a negative pressure greater than or equal to 1/8 inches Water Gauge relative to the outside atmosphere during system operation.

e.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 30,000 ACFM +/- 10%.

f.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 30,000 ACFM +/- 10%.

SUMMER - UNIT 1 3/4 7-41 Amendment No. 42, !31, 440,159,160

PLANT SYSTEMS 3/4.7.12 SPENT FUEL ASSEMBLY STORAGE LIMITING CONDITION FOR OPERATION 3.7.12 The combination of initial enrichment and cumulative burnup for spent fuel assemblies stored in Region 2 shall be within the acceptable domain of Figure 3.7-1.

APPLICABILITY:

Whenever irradiated fuel assemblies are in the spent fuel pool.

ACTION:

a.

With the requirements of the above specification not satisfied, suspend all other movement of fuel assemblies and crane operations with loads in the fuel storage areas and move the non-complying fuel assemblies to Region 1. Until these requirements of the above specification are satisfied, boron concentration of the spent fuel pool shall be verified to be greater than or equal to 2000 ppm at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.12 The burnup of each fuel assembly stored in Region 2 shall be ascertained by careful analysis of its burnup history prior to storage in Region 2. A complete record of such analysis shall be kept for the time period that the ioel assembly remains in Region 2 of the spent fuel pool.

SUMMER - UNIT 1 3/4 7-42 Amendment No. 27,116, 160

z 0

W 0

ca 2

W 0o Cfo 45 40 35 30 25 20 15 10 5

0QL1 2.0 2.5 3.0 3.5 4.0 4.5 5.0 MAXIMUM NOMINAL U-235 ENRICHMENT (W/0)

Notes: 1. Fuel assemblies with enrichments less than 2.0 W/0 must meet the burn-up requirements of 2.0 W/0 U-235 assemblies.

2. Use of the following polynomial fit is acceptable, where E = Enrichment (W/O):

Assembly Discharge Burnup = 0.1246 E3 - 1.91 E2 + 20.9205 E - 30.2482 FIGURE 3.7-1 REQUIRED FUEL ASSEMBLY BURN-UP AS A FUNCTION OF INITIAL ENRICHMENT TO PERMIT STORAGE IN REGION 2 Amendment No. 160 SUMMER - UNIT 1 3/4 7-43

PLANT SYSTEMS 3/4.7.13 SPENT FUEL POOL BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.7.13 The boron concentration in the spent fuel pool, the fuel transfer canal, and the cask loading pit shall be maintained at a boron concentration greater than or equal to 500 ppm.

APPLICABILITY:

Whenever new or irradiated fuel is being moved (non-refueling movement) in the spent fuel pool, fuel transfer canal, or cask loading pit.

ACTION:

With the requirements of the above not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the spent fuel pool, the fuel transfer canal, and the cask loading pit until the boron concentration in the area where fuel is being moved shall be verified greater than or equal to 500 ppm.

SURVEILLANCE REQUIREMENTS 4.7.13 The boron concentration of the spent fuel pool, fuel transfer canal, or cask loading pit shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when moving new or irradiated fuel in the spent fuel pool, transfer canal, or cask loading pit.

SUMMER - UNIT 1 3/4 7-44 Amendment No.

"160

REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical a period of time within the acceptable domain of Figure 3.9-1, but not less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

APPLICABILITY:

During movement of irradiated fuel in the reactor pressure vessel.

ACTION:

With the reactor subcritical for less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, immediately suspend all movement of irradiated fuel in the reactor pressure vessel. With the reactor subcritical for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> but not within the acceptable domain of Figure 3.9-1, immediately suspend movement of irradiated fuel in the reactor pressure vessel.

SURVEILLANCE REQUIREMENTS 4.9.3.1 The reactor shall be determined to have been subcritical for a period of time within the acceptable domain of Figure 3.9-1 by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

4.9.3.2 Prior to moving irradiated fuel from the reactor pressure vessel, and at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during movement of irradiated fuel, verify the CCW temperature at the inlet to the Spent Fuel Pool Cooling System heat exchanger is within the acceptable domain of Figure 3.9-1.

Amendment No.160 SUMMER - UNIT 1 3/4 9-3

150 140 130 120 110 100 90 80 70' 60 85 95 100 CCW TEMPERATURE (OF)

Note: The use of linear interpolation between CCW temperatures reported above is acceptable to determine the minimum incore hold time.

FIGURE 3.9-1 REQUIRED IN-CORE HOLD TIME AS A FUNCTION OF COMPONENT COOLING WATER (CCW) TEMPERATURE Amendment No.'160 SUMMER - UNIT 1 03 rr D

0

-r 0 I-w 0

0 Y K*

105 314 9-3a

PLANT SYSTEMS BASES 3/4.7.8 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.7.9 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures.

Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of 20F.

3/4.7.10 WATER LEVEL - SPENT FUEL POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

3/4.7.11 SPENT FUEL POOL VENTILATION SYSTEM The limitations on the spent fuel pool ventilation system ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analysis.

B 3/4 7-6 SUMMER - UNIT 1 Amendment No. 7-9, 160

PLANT SYSTEMS BASES 3/4.7.12 SPENT FUEL ASSEMBLY STORAGE The restrictions placed on spent fuel assemblies in Region 2 of the spent fuel pool ensure Ke~f remains less than 0.95. The minimum burnup bounds the use of Burnable Poison Rod Assemblies (BPRA), Wetted Annular Burnable Absorbers (WABA), Integral Fuel Burnable Absorbers (IFBA), and Erbia.

An axial burnup shape penalty is also included in the burnup requirement.

3/4.7.13 SPENT FUEL POOL BORON CONCENTRATION A minimum boron concentration is required in the spent fuel pool, fuel transfer canal, or cask loading pit whenever new 4.95 W/0 fuel is being moved to ensure Keff remains less than 0.95 during this normal condition of fuel movement.

The minimum boron concentration in the spent fuel pool, fuel transfer canal, or cask loading pit also is sufficient to maintain Keff less than 0.95 for postulated accident condition consisting of a dropped or a mispositioned fuel assembly. This requirement is a direct result of the analysis performed for pool criticality during evolutions performed while the Spent Fuel Pool is isolated from the reactor cavity and is due to geometry, materials and poisons being different in the spent fuel pool than those in the reactor. During periods of direct communication between the pool and the reactor, Specification 3.9.1 shall be followed when the refueling cavity is filled and the transfer canal blind flange is removed.

Sampling to determine boron concentration is required only for those specific areas where fuel is being moved, e.g. in the spent fuel pool, in the fuel transfer canal, or in the cask loading area.

Amendment No. 2 160 SUMMER - UNIT 1 B 3/4 7-7

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that: 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses. The value of 0.95 or less for Keff includes a 1 percent delta k/k conservative allowance for uncertainties. Similarly, the boron concentration value of 2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron. Valves in the reactor makeup system are required to be closed to minimize the possibility of a boron dilution accident.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. The minimum decay time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is consistent with the assumptions used in the accident analysis.

The tabulated hold times associated with Component Cooling Water (CCW) temperature ensure that the spent fuel heat load is reduced sufficiently to allow the spent fuel pool cooling system to maintain the bulk pool temperature below 170 0F. These hold times ensure that adequate cooling is provided to the Spent Fuel Pool under the highest possible heat load conditions. The hold times are based on the performance of the cooling system, which is dependent upon CCW temperature and recognizes that the spent fuel pool cooling system is capable of increased flow rates up to 2400 gpm during single loop operation. This higher flow rate may be required when only a single cooling loop is operable during a refueling outage.

The CCW temperature limits defined in Figure 3.9-1 are adjusted for uncertainty in the implementing procedure.

3/4.9.4 REACTOR BUILDING PENETRATIONS The requirements on reactor building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of reactor building pressurization potential while in the REFUELING MODE.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

Amendment No. 160 SUMMER - UNIT 1 B 3/4 9-1

REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that:

1) manipulator cranes will be used for movement of control rods and fuel assemblies,

2) each crane has sufficient load capacity to lift a control rod and fuel assembly, and

3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be in operation ensures that 1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140OF as required during the REFUELING MODE, and 2) sufficient coolant circulation is maintained thru the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.8 REACTOR BUILDING PURGE SUPPLY AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the reactor building vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the reactor building. The OPERABILITY of this system is required to restrict the release of radioactive material from the reactor building atmosphere to the environment.

3/4.9.9 WATER LEVEL - REACTOR VESSEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

Amendment No. 160 SUMMER - UNIT 1 B 3/4 9-2

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 157 fuel assemblies. Each fuel assembly shall consist of 264 Zircaloy-4 or ZIRLO(TM) clad fuel rods with an initial composition of uranium dioxide with a maximum nominal enrichment of 4.95 weight percent U-235 as fuel material. Limited substitutions of Zircaloy-4, ZIRLO(TM) and/or stainless steel filler rods for fuel rods, if justified by a cycle specific reload analysis using an NRC-approved methodology, may be used. Fuel assembly configurations shall be limited to those designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or cycle-specific reload analyses to comply with all fuel safety design bases. Reload fuel shall contain sufficient integral fuel burnable absorbers such that the requirements of Specifications 5.6.1.1a.2 and 5.6.1.2.b are met. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core locations.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 48 full length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b.

For a pressure of 2485 psig, and

c.

For a temperature of 6501F, except for the pressurizer which is 6801F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 9914 +/- 100 cubic feet at an indicated Tavg of 587.41F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

SUMMER - UNIT 1 5-6 Amendment No. 27-1-554-62, 74, 105, 116, 119, 160

DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks consist of 1712 individual storage cells. The cells are grouped into two regions, which are determined based on storage cell spacing as defined below. The spent fuel storage racks are designed, and shall be maintained, with a Ke, less than or equal to 0.95 when flooded with unborated water, which includes conservative allowances for uncertainties and biases. This is ensured by maintaining the following for each region:

a.

REGION 1 - designated for storage of fresh fuel assemblies and fuel assemblies with a cumulative burnup less than the required cumulative burnup for storage in Region 2.

1.

A nominal 10.867 inch center-to-center distance between fuel assemblies placed in the storage rack.

2.

A maximum nominal initial enrichment of 4.95 weight percent U-235.

b.

REGION 2 - designated for storage of discharged fuel assemblies.

1.

A nominal 9.07 inch center-to-center distance between fuel assemblies placed in the storage rack.

2.

A cumulative burnup with the acceptable domain defined by Figure 3.7-1.

5.6.1.2 The new fuel storage racks consist of 60 individual cells, each of which accommodates a single assembly. The new fuel pit storage racks are designed and shall be maintained with a Kf, less than or equal to 0.95 when flooded with unborated water and less than or equal to 0.98 for low density optimum moderation conditions, including conservative allowances for uncertainties and biases. This is ensured by maintaining:

a.

A nominal 21 inch center-to-center distance between new fuel assemblies placed in the storage rack.

b.

A nominal enrichment of 5.0 weight percent U-235.

DRAINAGE 5.6.2 The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 460'3".

CAPACITY 5.6.3 The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 1712 fuel assemblies, with 200 assemblies in Region 1 and 1512 assemblies in Region 2.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

SUMMER - UNIT 1 5-7 Amendment No. 27 74-,116, 444, 160

DESIGN FEATURES This page intentionally left blank.

Amendment No. 160 SUMMER UNIT 1 5-8

DESIGN FEATURES This page intentionally left blank.

Amendment No. 1 60 SUMMER UNIT 1 5-9