ML19312A220

Chapter 16 of S&W SWESSAR-P1, Tech Specs.

ML19312A220
Person / Time
Site:	05000495
Issue date:	11/29/1978
From:	NEW YORK STATE ELECTRIC & GAS CORP., STONE & WEBSTER, INC.
To:
References
NUDOCS 7909060016
Download: ML19312A220 (62)

Text

SWESSAR-P1 CHAPTER 16 TECHNICAL SPECIFICATIONS LIST OF EFFECTIVE PAGES Amendment Amendment Page, Table (T) , No.

or Figure (F) No.___

16-a 39 1

16-i 13 16-ii 1

16-iia 4

16 -iii Orig 16.1-1 Orig 16.2-1 Orig 16.3-1/2 9 16.3-3/4 23 16.3-5 through 6B 13 16.3-6C/D 8 16.3-7/8 11 16.3-9/10 Orig 16.3-11 through 16 T16.3-1 (2 sheets) 35 8

16.4-1/2 16 16.4-3 34

" 16.4-4 16.4-5 through 6A 21 16.4-6B 33 16.4-7 through 10 8 8

16.4 -10A/10B 8 16.4-11 16.4-12 through 15 Orig 4

T16.4.2-1 6 T16.4.2-2f,3 8

F16.4.2-1 16.5-1 through 4 Orig 16.6-1 Orig

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16-a Amendment -

7/14/5

SWESSAR-P1 CIIAPTER 16 TECllNICAL SPECIFICATIONS TABLE OF CONTENTS Section Page 16.1 DEFINITIONS 16.1-1 16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEli SETTINGS 16.2-1 16.3 LIMITING CONDITIONS FOR OPERATION 16.3-1 16.3.1 Peactor Coolant System 16.3-1 16.3.2 Chemical and Volume Control System 16.3-1 16.3.3 Engineered Safety Features 16.3-1 16.3.3.1 Safety Injection and Residual Heat Removal ,

Systems 16.3-1 16.3.3.2 Containment IIeat Removal and Iodine Removal Systems 16.3-1 16.3.3.3 Reactor Plant Component Cooling Water System 16.3-3 16.3.3.4 Reactor Plant Service Water System 16.3-4 16.3.3.5 Ultimate IIeat Sink 16.3-6 16.3.3.6 Extended Maintenance 16.3-6 16.3.4 Steam and Power Conversion System 16.3-6 16.3.5 Containment System 16.3-8 16.3.6 Electrical System 16.3-9 16.3.7 Initial Fuel Loading and Refueling 16.3-10 16.3.8 Effluent Release Limits 16.3-13 16.3.8 1 Release of Radioactive Liquid Waste 16.2-13 16.3.8.2 Release of Radioactive Gaseous Waste 16.3-15

'6.3.9 Control Rod and Power Distributior Limits 16.3-16 16-i Amendment 1 7/30/74 670 l$0

SWESSAR-P1

'IABLE OF CONTENTS (CONT)

P Sectis Page 16.3.10 Core Surveillance Instrtrnentation 16.3-16 16.4 SURVt.1LIJ4C1. kEQUIRM1ENTS 16.4-1 16.4.1 Operat.Lonal Safety Review 16.4-1 16.4.2 hules for Inservice Inspection and Test-ing af Nuclear Reactor Coolant Systems and Other Applicable Systems 16.4-1 16.4.3 heactor Coolant System Integrity Testing 16.4-3 16.4.4 Con + a inment Structure Leakage Rate Tests 16.4-4 Reference Ior Section 16.4.4 16.4-6 16.4.5 1.ngineered Satety Features 16.4-6 13 16.4.6 kJaergency Power System Tests 16.4-8 16.4.6.1 Diesel Generators 16.4-8 16.4.6.2 Fuel Oil Storage Tanks for Diesel Generators 16.4-9 16.4.6.3 Station Batteries 16.4-9 16.4.7 Main Steam Isolation Valves 16.4-16A 16.4.8 Auxiliary Feedwater Pumps 16.4-10B 16.4.9 Reactivity Anomalies 16.4-12 16.4.10 Ef fluent Sampling and Radiation Monitoring Systems 16.4-12 16.5 DESIGN Ft.ATURES 16.5-1 16.5.1 Site 16.5-1 16.5.2 Containment 16.5-1 16.5.2.1 Containment 16.5-1 16.5.2.2 Cbntainment Penetrations 16.5-1 16.5.2.3 Containment Systems 16.Q-2 7 (j/ij i c' ! s 16-ii Amendment 13 6/30/75

SWESSAR-P1 TIELE OF CONTENTS (CONT)

Section Page 16.5.3 Reactor 16.5-3

16. 5. f4 Fuel Storage 16.5-3 16.6 ADMINISTRATIVE CONTROLS 16.6-1 I

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i uu 16-iia Amendment 1 7/30/74

SWESSAR-P1 LIST OF TABLES Tables 16.3-1 Technical Specification Limits 16.4.2-1 Identification of Code Class 2 Systems Requiring 4

Inservice Inspection 16-iii Amendment 4 11/1/74

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SWESSAF.-p1 16.1 DEFINITIONS Refer to Section 16.1 of the NSSS Vendor's SAR.

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SWESSAR-P1 16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS Refer to Section 16.2 of the NSSS Vendor's SAR.

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SWESSAR-P1 16.3 LIMITING CONDITIONS FOR OPERATION 16.3.1 Reactor Coolant System This material is discussed in the NSSS Vendor's SAR.

16.3.2 Chemical and Volume Control System This material is discussed in the NSSS Vendor's SAR.

16.3.3 Engineered Safety Features 16.3.3.1 Safety Iniection and Residual Heat Removal Syste:as This material is discussed in the NSSS Vendor's SAR.

16.3.3.2 Containment Heat Removal And Iodine Removal Systems Applicability Applies to the operational status of the containment heat removal and iodine removal systems.

Obiective To define those conditions of the containment heat removal and iodine removal systems necessary to ensure safe unit operation.

Specifica tion The following specifications apply except during low temperature physics tests.

A. The reactor shall not be made critical unless the following conditions are met:

1. All the containment spray subsystems shall be operable.

2. All the containment atmosphere recirculation (CAR) subsystems shall be operable.

3. The refueling water storage tank (RWST) shall contain at least the quantity of borated ' water as soecifi ed in Table 16.3-1. The water shall have a boron concentration not less than that specified in Table 16.3-1 to ensure that the reactor is in the refueling shutdown condition when all control rod assemblies are inserted and the reactor has been cooled down.

4. The refueling water chemical addition tank (CAT) shall contain at least the quantity of sodium hydroxide solution specified in Table 16.3-1 with a concentration of not less than that specified in Table 16.3-1.

16.3-1 3mr 670 i3

SWESSAR-P1 oi ._ !. O!9

5. All valves and piping associated with the above components shall be operable.

B. During power operation Specification 16.3.3.2-A may be modified to allow one subsystem to be out of service provided that immediate action is taken to restore the affected subsystem to operable status. If the subsystem is not restored to operable status within the outage time period specified in Table 16.3-1, the reactor shall be placed in the hot shutdown condition. If the subsysten is not restored to operable status within the supplementary time period as specified in Table 16.3-1, the reactor shall be placed in the cold shutdown condition using normal operating procedures.

1. The other containment spray subsystems shall be tested in acccordance with Specification 16.4.5 to demonstrate operability coincident with initiating rapair of the inoperable subsystem.

2. The other CAR subsystems operate during normal unit operation; therefore, testing of these subsystems is not required.

3. If the RWST volume cannot be maintainad at the level specified in Table 16.3-1 by makeup, the reactor shall be placed in hot shutdown until water level is reestablished.

Bases The containment heat removal systems consist of the following subsystems: separate containment spray subsystems, separate Crik subsystems, and low head safety injection (:LHSI) subsystems as described in Chapter 16 of the NSSS Vendor's SAR.

Each containment spray subsystem draws from the RWST. The water in the RWST is 120 F or less.

In each containment spray subsystem, water flow is fror. the RWST through a containment spray pump and through a containment soray header into the containment atmosphere. After the RWST inventory is exhausted, each containment spray pump takes suction from a separate ESF sump. Each CAR subsystem fan takes suction from the containment atmosphere and discharges the cooled air back into the ccntainment atmosphere. Cooling is effected by the reactor plant component cooling water system.

During accident conditions, heat is also removed from the containment by the residual heat removal (RHR) heat exchangers dis cussed in Chapter 16 of the NSSS Vendor's SAR. With minimum engineered safety features (ESF) as described in Table 6.1 , the containment heat removal systems are capable of cooling and depressurizing the containment to a low pressure within one day i following a design basis accident (DBA). With the above 16.3-2

SWESSAR-P1 equipment operating, the containment heat removal systems are capable of maintaining this low pressure in the containment indefinitely following a DBA.

In addition to supplying water to the containment spray subsystem, the RWST is also the source of water for both high and low head safety injection systems during the injection phase immediately following an accident. This water is borated to a concentration that ass ures the minimum reactor shutdown margin required when all control rod assemblies are inserted and when the reactor is cooled down for refueling.

Reference Sections Title No.

Containment Functional Design 6.2.1 Containment Heat Removal System 6.2.2 Containment Spray System-Iodine Removal 6.2.3.2 9

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16.3-3/4 Amen dment 9 4/30/75

SWESSAR-P1 16.3.3.3 Reactor Plant Component Cooling Water System Applicability Applies to the operational status of the reactor plant component cooling water system.

Obiective To define limiting conditions for the component cooling water system necessary to ensure safe operation of the reactor during startup, power operation, and cooldown.

Specifications The following specifications apply except during low temperature physics tests.

A. The reactor shall not be made critical unless operating conditions for the component cooling water system are as follows:

1. The B component cooling water pump in each engineered safety features area is operable.

2. All component cooling water heat exchangers are operable.

3. Each component cooling water train is operable for immediate supply of cooling water to the designated 23 equipment on Table 9.2.2-6.

B. During power opera tion , Specification 16.3.3.3A may be modified to allow one of the required components to be inoperable provided immediate attention is directed to making repairs. If the system is not restored to operable status within the outage time period specified in Table 16.3-1, the reactor is placed in the hot shutdown condition . If the repairs are not completed within the supplementary time period specified in Table 16.3-1, the reactor is placed in the cold shutdown condition.

i. The B component cooling water pump in one engineered safety features train may be out of service, provided this pump is restored to operable status within the specified time period and the remaining train is demonstrated operable.

2. One component cooling water heat exchanger may be out of service provided that it is restored to operable status within the time period specified or the reactor plant service water temperature is sufficiently low as identified in Table 16.3-1 to 16.3-5 7 , ,g Amendment 23 0 / ;,U i/U 3/31/76

SWESSAR-P1 enable the remaining heat exchanger to handle the 7A.lg maximum design heat load.

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3. Any valve required for the functioning of the system during and following accident conditions may be inoperable provided that it is restored to operable status within the time period specified and all valves in the system that provide the duplicate function are demonstrated to be operable.

C. During operation of the reactor coolant pumps, loss of reactor plant component cooling water to one or more reactor coolant pump requires:

23 1. The affected reactor coolant pump (s) must be shut down within the time period specified in Table 16.3-1.

2. The affected reactor coolant pump (s) shall not be restarted until component cooling water is restored and pump thermal conditions are normal.

Bases The reactor plant component cooling water system provides an intermediate loop for removing heat 'from reactor plant auxiliary systems and transferring it to the reactor plant service water system. The component cooling water system is subdivided into two distinct trains. Power for each train is provided by an independent emergency bus.

During normal operation, the A pump in each train is operated.

The B pump in each train is started manually on a low pressure alarm. The B pump in each train must be operated to supply the residual heat removal exchanger.

Sufficient redundancy is provided so that the loss of a complete train, as the ass.xmed single active f ailure, does not prevent the safety function of the system from being achieved.

For the Combustion Engineering NSSS, analysis has shown that the reactor coolant pump will remain operable for at least 45 minutes 23 following loss of component cooling water. Requiring operator action before 30 minutes ensures that the plant may be safely shut down.

16.3.3.4 Reactor Plant Service Water System Applicability Applies to the operational statut of the reactor plant service water system.

Obiective 16.3-6 Amendment 23 3/31/76

SWESSAR-P1 To define limiting conditions for the service water system n< cessary to ensure safe operation of the reactor during startup, pa er operation, and cooldown.

Specification The following specifications apply except during low temperature physics tests.

A. The reactor shall not be made critical unless operating conditions for the service water system are as follows:

1. One pump in each train is operable.

2. Each train is operable for Unmediate supply of cooling water to the following:

a. Reactor plant component cooling water heat exchanger

b. Emergency diesel generator cooler

c. Control room air conditioning chiller

d. Safety related unit coolers B. During power operation, Specification 16.3.3.4 A may be modified to allow one of the required components to be inoperable provided immediate attention is directed to making repairs. If the system is not restored to operable status within the outage time period specified in Table 16.3-1, the reactor is placed in the hot shutdown condition. If the repairs are not completed within the supplementary time period specified in Table 16.3-1, the reactor is placed in the cold shutdown condition.

1. Both service water pumps in the same train may be out of service, provided one pump is restored to operable status within the specified time period and the remaining trains are demonstrated operable.

2. Any valve required for the functioning of the system during and following accident conditions may be inoperable provided that it is restored to operable status within the specified time period and all valves in the system that provide the duplicate function are demonstrated to be operable.

Bases The service water system provides cooling f or reactor plant auxiliary systems and consists of separate trains. Each train consists of two pumps and its associated cooling train. Power for each train is provided by an independent emergency bus.

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16 . 3 -6 A Amencment 23 3/31/70

[b[ bz h swESSAR-P1 During normal operation, one pump in each train may be necessary for operation. The second pump in each train is started manually upon a low pressure or flow alarm.

Sufficient redundancy is provided so that the loss of a complete train, as the assumed single active failure, does not prevent the safety function of the system from being achieved.

s 16.3-6B Amendment 23 3/31/76

SWESSAR-P1 16.3.3.5 Ultimate heat 51nk This will be described in the Utility-Applicant's SAR.

16.3.3.6 Extended Maintenance The extended maintenance time period for the engineered sarety teatures is oescribed in preceding subsections of Section 16.3.3.

16.3.4 Steam and Power Conversion System Applicability Applies to the operating status of the main steam system (Section 10.3) and auxillary feedwater system (Section 10.4.10).

Ubiective To define the concitions required in the main steam and auxiliary feedwater systems during loss or power to prevent release of reactor coolant through the pressurizer safety valves, and to ensure the capability to remove sensible and decay heat trom the core by heat exchange in the steam generators.

Specif ica tion A The reactor shall not be made critical except for low power physics testing unless operating conditions for the steam and power conversion system are:

1. All auxiliary reedwater pumps, motor-driven and turbine-driven, and the auxiliary feedwater pump shall be operable.

2. The minimum inventory of water as specitied in Table 16.3-1 shall be available in the auxiliary feedvater storage tank (AFST) to supply emergency water to the auxiliary feedwater pump suctions.

3. All main steam safety valves shll be operable.

4. System piping and valves directly associated with the aoove components shall be operable.

B. The reactor coolant system temperature shall not be raised I" above the temperature / pressure of the residual heat removal system's capabiilty unless the steam generators are able to perform their heat transfer function to the main. steam system.

C. During power operation of the reactor, the requirements or Specification 16.3.4A are modified to allow:

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SWESSAR-P1

1. Any one auxiliary feedwater pump and its associated the tlow path may be inoperable, provided operability of the other auxiliary feedwater pumps and their associated flow paths is demonstrated once per day.

basis The steam supply to the turbine driven auxiliary teecwater pump is connectea to the steam generators.

Each motor-driven auxiliary feedwater pump receives power from an independent emergency bus in the emergency power system (Section 8.3) . 'Ihe auxiliary feedwater flow rate is established to ensure that each auxiliary feedwater pump has suf f icient capacity for removal of core sensible and decay heat.

U The capability to supply feedwater to the steam generators and is normally provided by the operation of the condensate feedwater sy'3tems (Section 10.4.7) . In the event of complete loss of electric power, sensible and decay heat removal continues of the auxiliary to be ensured by the availability of one feedwater pumps and the auxiliary feedwater storage tank.

The minimum capacity of the AFST is sufficient for sensible and the decay heat removal and cooldown to the conditions at which residual heat removal system can commence operation following a The reactor trip and loss of all offsite electric power.

condensate stroage tank may be available as an alternate supply of feedwater to the auxiliary feedwater pump suctions.

The main steam atmospheric dump valves have a combinea capacity as given in Table 10.3-1. If the atmospheric dump valves fall to open, the main steam safety valves have a combined capacity as The combined capacity of the main steam given in Table 10.3-1.

safety valves and the main steam atmospheric dump valves required b[h rn i a >;.

16.3-6D Amendment 13 6/30/75

SNESSAR-P1 by this specification alwa ys exceeds the total steam flow cor responding to the maximum steady-state power that can be obtained during operation.

The availability of the auxiliary feedwater pumps, the auxiliary feedwater storage tank, and the main steam safety valves adequately ensure that suf ficient sensible and decay heat removal capability is available when required.

Reference Sections Title No.

Reactor Coolant System Chapter 5 (NSSS)

Residual Heat Removal System Chapter 5 (NSSS)

Onsite Power Systems 8.3 l8 Main Steam System 10.3 Condensate and Feedwater Systems 10.4.7 Auxiliary Feedwater System 10.4.10 Auxiliary Feedwater Pumps '5.4.8 16.3.5 Containment System Applicability Applies to the integrity and operating pressure of the containment structure.

Obiective To define the limitina operating status of the containment structure for unit operation.

Specification A. Containment Integrity

1. The containment integrity shall not be violated unless the reactor is in the cold shutdown condition.

2. The containment integrity shall not be violated when the reactor vessel head is removed unless the shutdown margin is greater than that specified in Table 16.1-1.

3. Positive reactivity changes shall not be made by rod drive motion when the containment integrity is not intact, unless the shutdown margin is greater than that specified in Table 16.3-1.

4. Positive reactivity changes shall not be made by boron dilution when the containment integrity is not inta ct ,

unless the shutdown margin is creater than that specified in Table 16.3-1.

16.3-7 Amendment 8 b0 l0f

SWESSAR-P1 B. Containment Pressure If the containment interna l pressure exceeds 15.7 psia, the condition shall be corrected or the reactor shall be shut down.

Bas is With the reactor coolant system (RCS) in the cold shutdown condition , there can be no steam generation and thus no pressure buildup in the containment if a loss of coolant accident (LOCA) occurs, b[h !Or iGJ t

16.3-8 Amendment 8 3/28/75

SWESSAR-P1 the maximum containment pressure allowable during normal At operation , containment integrity is assured should a LOCA occu r .

Reference Section Title No.

Containment Systems 6.2 16.3.6 Electrical System Applicability Applies to availability of electrical power for operation of station auxiliaries and safe shutdown of the plant.

Obiective To define those conditions of electrical power availability necessary to provide for safe reactor operation.

Specif ica tion A. The rea ctor shall not be made critical unless the following conditions are satisfied:

1. Sufficient normal power is available to meet plant requirements outlined in these technical specitications.

2. All emergency a-c systems are energized.

3. All emergency d-c systems are energized including one charger per battery.

4. All 120 V vital systems are energized.

5. All emergency diesel generators are operable at maximum rated load with a 7 day fuel supply in each underground storage tank.

6. All preferred of f site supplies are energized.

B. The reactor may remain critical with the following limitations:

1. Sufficient normal power is available to meet the requirements of the nuclear steam supply system technical specifications.

2. One of the two redundant bottery chargers associated with each channel may be unavailable.

3. Any one inverter or the regulating tran sformer associated with each channel may be unavailable.

16.3- 9 Amendment 11 b[fj l]6 5/30/75

SWESSAR-P1

4. One of the two redundant emergency diesel generator fuel oil transfer purps associated with each diesel generator may be unavailable. ,

5. One or more sources of onsite and/or offsite power may be unavailable. However, this would place the plant beyond limiting conditions of operation and time 11 constraints on the continued operation of the plant, prior to shutdown, would be required. These are operatino constraints and shall be described in the Technical Specification of the Utility-Applicant.

6. Loss of any emergency a-c, d-c, or 120 V a-c vital system bus shall be treated as condition 5 above.

Bases See Sections 8.3.1, 8.3.2, and 8.3.3.

16.3.7 Initial Fuel Loading and Refueling Applicability Applies to operating limitations during initial fuel loading and during refueling operations.

Obiective To ensure that no accident can occur, during initial fuel loading or during refueling operations, that would affect public health and safety.

Specifica tion A. During fuel loading or refueling operations the following conditions will be satisfied:

1. Whenever core geometry is being changed, the equipment hatch and at least one door of each personnel access lock in the containment structure shall be properly closed. In addition, automatic containment isolation valves shall be cycled to demonstrate operability and manual containment isolation valves shall be closed and tagged, except for the isolation valves in those lines which penetrate the containment structure and which must be used during initial core loading or refueling or which would be required to mitigate the consequences of a fuel handling accident, which shall be tagged open.

2. The containment purge air system and the containment purge air exhaust monitors, which initiate isolation of the containment purge air system, shall be demonstrated 16.3-10 Amendment 11 f 5/30/75

SWESSAR-P1 to be operable immediately prior to refueling operations (Sections 9.4.5 and 11.4).

3. Source range neutron de*c7 tors shall be in service as specified by the NSSS Vendor.

4. Rea ctor cavity manipulator crane area radiation levels ,

airborne activity levels within the containment structure (Section 12.2) , and airborne activity levels in the fuel building exhaust air and containment purge air system exhaust ducts shall be continuously monitored during refueling. A high activity alarm from either of the redundant containment purge air exhaust monitors will automatically stop the containment purge air fans and automatically close the containment purge air system containment isolation valves. Hourly readings shall be taken on the purge air exhaust monitors and a plot of the readings shall be maintained to allow early identification of any change.

5. Fuel pool surface area radiation levels and fuel building ventilation exhaust air subsystem activity levels shall be continuously monitored when fuel is stored in the fuel pool. During refueling operations the fuel building ventilation exhaust air subsystem (Section 9.4.6) shall be continuously passed through one of the filter banks of the supplementary leak collection and release system (Section 6.2.3.1) prior to discharge through the ventilation vent.

6. At least one residual heat removal pump and one residual heat exchanger flow path shall be operable.

7. When the reactor vessel head is unbolted, a minimum boron concentration as specified by the NSSS Vendor, shall be maintained in any filled portion of the reactor coolant system. This boron concentra tion shall be checked by sampling with a frequency as specified by the NSSS Vendor.

8. Direct communication between the control room and the cab of the reactor cavity manipulator crane shall be available whenever changes in core geometry are taking place.

9. There shall be no movement of fuel in the reactor core until the reactor has been suberitical for a period of time as specified by the NSSS Vendor.

10. No load exceeding 110 percent of the weight of a fuel assembly will be moved over new fuel, and only one fuel assembly shall be handled at any one time over the reactor or the new f uel storage racks.

16.3-11 670 133

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11. The water level in the refueling cavity shall be at least (later) ft above the reactor vessel flange.

B. If any of the limiting conditions for refueling specified in Specification 16.3.7 are not met, refueling of the reactor shall cease, work shall be initiated to correct the conditions so that the specified limits are met, and no operations which' increase the reactivity of the core shall take place.

C. After initial fuel loading and after each core refueling operation and prior to reactor operation at greater than 75 percent of rated power, the moveable in-core neutron flux detector system shall be used to verif y proper power distribution.

Basis Detailed instructions, the above specifications, and the design of the fuel handling equipment, which incorporates built-in interlocks and safety features, ensure that an accident which could result in a hazard to public health and safety cannot occur during fuel loading and refueling operations. When no change is being made in core geometry, one neutron detector is sufficient to monitor the core. This permits maintenance of the instrumentation not in operation. Continuous monitoring of rad iation levels and neutron flux provides immediate indication of an unsafe condition. Containment structure high airborne radioactivity levels automatically stop and isolate the containment purge air system and cause actuation of audible and visual alarms in the control room and locally. The fuel building ventilation exhaust is filtered through one of the filter banks of the supplementary leak collection and release system when refueling is in progress.

At least one residual heat removal pump and residual heat exchanger flow path is required for cooling and mixing of the reactor coolant contained in the reactor vessel so as to maintain a uniform boron concentration and to remove residual heat. At least one reacter plant component cooling pump and heat exchanger flow path is required for cooling the residual heat removal system. One reactor plant service water pump in the associated train is required to provide cooling for the component cooling heat exchanger.

Specification 16.3.7 A.8 allows the operator in the control room to inform the reactor cavity manipulator crane operator of any impending unsafe condition as detected by the main control board indicators.

In addition to the above specifications, interlocks are used during fuel loading and refueling to ensure safe handling of the fuel assemblies. An excess weight interlock is provided on the '

reactor cavity manipulator crane hoist drive to prevent movement 16.3-12

SWESSAR-P1 of more than one fuel assembly at a time. Also, the fuel transfer mechanism is able to accommodate only one fuel assembly at a time.

On completion of core loading and installation of the reactor vessel head and prior to initial criticality, specific mechanical and electrical tests will be performed on the rod cluster control assenbly drive mechanisms, the rod position indication system, reactor trip circuits, in-core thermocouples, and neutron flux instrumentation system to ensure continuity, safety, and the ability of equipment to perform design functions.

The requirements detailed in this Specification ensure that re-fueling conditions will conform to the limiting operating condi-tions assumed in the fuel handling accident analysis (Section 15.1.23).

Additional bases are described in the USSS Vendor's SAR.

Reference Sections Title Number Containment Structure 3.8.1 Reactor 4 (NSSS) dupplementary Leak Collection and Release System 6.2.3.1 Containment Isolation Systems 6.2.4 Fuel Handling System 9.1.4 Reactor Plant Component Cooling Water System 9.2.2 Containment Purge Air System 9.4.5.2 Fuel Building Ventilation 9.4.6 Process and Effluent Radiation Monitoring Systems 11.4 Area Radiation Monitoring 12.1.4 Airborne Radiation Monitoring 12.2.4 Inadvertent Loading of a Fuel Assembly into an Improper Position 15.1.15 Fuel Handling Accident 15.1.23 16.3.8 Effluent Release Limits 16.3.8.1 Release of Radioactive Liquid Waste 16.3-13 670 170

SWESSAR-P1 Applicability Applies to the controlled release of radioactive liquid wastes discharged from the plant during normal operation.

Obiective To establish conditions for the release of liquid waste contain-ing radioactive materials and to ensure that such releases are within the concentration or dose limits specified in 10CFR20 and 10CFR50.

Specifica tion A. If actually experienced rates of release of radioactive material except tritium in liquid effluents from the plant, averaged over any calendar quarter, are such that the estimated ann tal quantities or concentrations of radioa ctive ma terials in liquid effluents are likely to exceed twice the design objective quantities and concentrations, the licensee shall:

1. Make an investigation to identify the causes for such release rates.

2. Define and initiate a program of action to reduce such release rates to the design level.

3. Report these actions to the USAEC within 30 days.

B. If a ctually experienced rates of release of radioactive material except tritium in liquid effluents, averaged over any calendar quarter, are such that estimated annual quantities or concentrations of radioactive material in liquid effluents are likely to exceed 8 times the design objective quantities and concentrations, the licensee shall:

1. Make an investigation to identify the causes for such release rates.

2. Define and initiate a program of action to reduce such release rates to the design level.

3. Submit a report to the USAEC with regard to the quanti-ties and concentrations of the principal radionuclides released to unrestricted areas as soon as practicable.

C. Tritium concentration at the point of discharge shall be con-trolled such that doses to an individual shall be less than those specified in 10CFR50.

D. The rate of release of radioactive materials in liquid waste k from the plant shall be controlled such that the instantaneous concentration of radioactivity in liquid waste I '

b/d t ' I 16.3-14

SWESSAR-P1 does not exceed the values listed in 10CFR20, Appendix B, Table II, Column 2.

Basis The releases of radioactive materials in liquid waste are based on the necessity to ensure the safety of the public and as such are based on the limits of 10CFn50. An analysis indicating that these limits are not exceeded is contained in Section 11.2.

16.3.8.2 Release of Radioactive Gaseous Waste Applicability Applies to the controlled release of radioactive gaseous waste discharged to the atmosphere trom the plant during lormal operation.

Obiective To establish conditions for the release of gaseous waste con-taining radioactive materials and to ensure that such releases are within the concentration limits specified in 10CFR20 and 10CFR50.

Specification A. If actually experienced rates of release of radioactive mat-erial in gaseous effluents from the plant, averaged over any calendar quarter, are such that the estimated annual quantities or concentrations of radioactive materials in gaseous effluents are likely to exceed twice the design objective quantities and concentrations, the licensee shall:

1. Make an investigation to identify the causes for such release rates.

2. Define and initiate a program of action to reduce such release rates to the design level.

3.

Report these actions to the USAEC within 30 days.

B. If actually experienced rates of release of radioactive material in gaseous effluents, averaged over any calendar quarter, are such that estimated annual quantities or concentrations of radioactive material in gaseous effluent are likely to exceed 8 times the design objective quantities and concentrations, the licensee shall:

1. Make an investigation to identify the causes for such release rates.

2. Define and initiate a program of action to reduce such release rates to the design level.

i^i u,v 16.3-15

SWESSAR-P1

3. Submit a report to the USAEC with regard to the quanti-ties and concentratior.s of the principal radionuclides released to unrestricted areas as soon as practicable.

C. The rate of release of radioactive materials in gaseous waste from the plant shall be controlled such that the instantan-eous concentration of radioactivity at the boundary of the restricted areas does not exceed the values listed in 10CFR20, Appendix B, Table II, Column 1.

Basis The releases of radioactive materials in gaseous wastes are based on the necessity to ensure the safety of the public and a. such are based on the limits of 10CFR20 and 10CFR50. An analysis indicating this is contained in Section 11.3.

Reference Sections Title No.

Control of Releases of Radioactive Materials to Environment 3.1.51 Radioactive Liquid Waste System 11.2 Radioactive Gaseous Waste System 11.3 16.3.9 Control Rod and Power Distribution Limits This material is discussed in the NSSS Vendor's SAR.

16.3.10 Core Surveillance Instrumentation This material is discussed in the NSSS Vendor's SAR.

6 16.3-16 4 7 fi UID !n7

\/J

SWESSAR-P1 TABLE 16 3-1 TECHNICAL SPECIFICATION LIMITS NSSS Vendor Item B&W C-E W-41 W-3S Containment Heat Removal and Iodine Removal System Minimum RWST 402,400 384,000 416,500 366,000 l35 capacity, gal Minimum boron con- See NSSS Vendor SAR centration, ppm Maximum boron See NSSS Vendor SAR concentration. ppm 9,200 10,900 5,000 Minimum CAT 4,600 f35 usable cap-acity, gal 13.4 25.0 35 Minimum NaOH 25.0 26.0 concentration, %wt Outage time period (1) 72 hr 24 hr 72 hr 48 hr 35 Supplementary 54 hr 48 hr 4 8 hr 24 hr time period Reactor Plant Component Cooling Water System Outage time period (1)

Pumps 72 hr 24 hr 72 hr 48 hr Heat exchangers 72 hr 24 hr 7 days 48 hr Valves 72 hr 24 hr 24 hr 48 hr 35 Supplementary 6 hr 60 hr 48 hr 24 hr time period Service water 74 hr 7s NA-loss of one 74 temperature below component cool-which no technical ing water heat specification limit exchanger re-is required for sults in a reactor plant complete loss component cooling of one safety water heat train.

@' exchanger, F 1 of 2 ,n, Amend nent 35 6 /,U i,4 10/6/77

SWESSAR-P1 TABLE 16.3-1 (CONT)

NSSS Vendor Item B&W C-E W-41 W-3S Maximum allow- 4 min. 30 min.(2310 min. 10 min. 35 able time before shutdown of a re-actor coolant pump following the ter-mination of com-ponent cooling water (See Section 9.2.2.6.) '5 3 Reactor Plant Service Water System Outage time period (3)

Pumps 72 hr 24 hr 72 hr 48 hr i Valves 35 72 hr 24 hr 24 hr 48 hr i Water temperature See Utility - Applicant's SAR.

Supplementary 6 hr 60 hr 48 hr 24 hr time period l35 Steam and Power Conversion System Minimum AFST 200,000 300,000 250,000 225,000 35 capacity, gal

. I Containment System Shutdown margin required,Sk/k See NSSS Vendor SAR Note 1: These time periods are based on the technical specifications limits of the emergency core cooling system (containment cooling system for BSW) of each NSSS Vendor and not the requirements of Stone & Webster.

2: Prior to shutdown of the last reactor coolant pump, a sufficient amount of boron shall be introduced into the reactor coolant system to facilitate cooldown to shutdown cooling system operating conditions.

2 of 2 Amendment 35 10/6/77 670 i95

""=uus

=a E

16.

f ~l f'; 1n' Vi U 1 / O,

SWESSAR-P1 16.4 SURVEILLANCE REQUIREMENTS 16.4.1 Operational Safety Peview This material is discussed in the Utility-Applicant's SAR.

16.4.2 Rules fo~r Inservioq1nypection and Testino of ? uclear o

Feact;or Coolar.t SystTs and Other Applicable Systorm Thic section applies to all systems which recuire inservice inspection and testing in accordance with ASME Section XI and all  :

applicable addenda thereto, hereinafter called ASME XI. The governing edition of ASME XI wi2? include the addenda in effect on the date of docketing for a construction permit. Inservico inspections are performed to ensure continuing integrity of the plant systems. Prior to initial plant operation, nondest ructive ,

inspection and preoperational + esting in accordance with the e

requirements of ASME XI will be performed. to establish the reference or baseline condition of the plant for comparison with subsequent inservice inspections and tests. lc Subsequent inservice inspections, system leakage test s ,

hydrostatic pressure tests, and inser vice tests will bo ,a g

performed. Records of all inspectiens and testa will be kept in accordance with the requiremen'.s of ASME XI, Inservice Inspection 8 The tentative program of inservice inspection anticipated considers the items, components, me thods , and frequency of inspection as given in ASME XI, Subsection IWB, Article IWS-2000 for Code Class 1 systems and Table 16.4.2-1 and ASME XI, Subsection IWC, Article IWC-2000 for Code Class 2 systems. Code Class 3 systems will be inservice examined in accordance with the requirements of ASME XI, Subsection IWD, Article IWD-2000. The g final program of inse7 vice inspection will be presented in the Utility-Applicant's final technical specifications.

Inservice inspection requirements beyond that specified in ASME XI are applied to the fcllowing portions of high energy fluid system piping that penetrate the containment structure:

A. Piping between containment isolation valves; B. Fhere no isolation valve exists inside containment, piping between the first rigid pipe connection to the containment penetration, or the first pipe whip restraint inside containment, and the outside isolation valve.

This additional inservice inspection will consist of a 100 percent volumetric inspection of 100 percent of the 16.4-1 Amendment 8 3/28/75 (7n' l' tn, l//

SWESSAR-P1 circumferential and longitudinal welds in these regions durino euch 10 year inspection interval. .

At the present time, the most advanced technique for volumetric inspection of reactor pressure vessels is ultrasonics. lio th ultrasonic techniques and radiography can be used for most of the other components, but ultrasonic techniques will probably be used ior the following reasons:

1. The pipe or vessel need not bo drained.

2. The inspection can be automated, reducing radiation exposure to personnel.

3. Less time is usually required than for radiographic in s p ection.

4. Radiation levels make radiography difficult and sometimes impossible.

S. Adequate reproducible records can be obtained with either manual or automated techniques.

Althouqh ultrasonic techniques can be used for most of the volumetric inspection, radiography may be used on piping and s &uctures and in other areas where ultrasonic techniques cannot be used.

Th< method of inspection planned for euch area - volumetric, surface, or visual - and the detailed procedures will be prepared at the time of preoperational inspection to permit incorporation of the latest available techniques.

Inservice Testing Code Class 1, 2, and 3 pumps powered from an emergency power source require inservice testing in accordance with the procedures and f requencies given in ASME XI, Subsection IWP, Articles IWP-3000 and IWP4 000. Where normanently in stalled instrumentation is able to function over the ranges stated in Table IWP-3100-2 within the nominal maximum errors st.ated in 0

Table IWP4110--1, that instrumentation may be used for the insorvice testing. Where permanently installed instrumentation is not used for inservice testing, portable station instruments with the code required accuracies will be used. Incal pressure taps and test wells are provided where required for the portable instruments.

Code Class 1, 2, and 3 valves require inservice testing in accordance with the procedures and frequencies given in ASME XI, Subsection IWV, Article IWV-3000. In addition, containment isolation valves, which are Code Class 2, require testing in .

(, 'l '

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16.4-2 Amendment 8 3/28/75

SWESSAR-P1 accordance with the requirements for Type C tests as stated in Section 16.4.4.

Whenever possible, the Type C test will be used to satisfy the requirements of the inservice test, with corrections for ditterential pressure being made according to the provisions of Article IWV-3420.

The final program of inservice testing will be presented in tl.e Utility-Applicant 's SAR.

16.4.3 Reactor Coolant System Integrity Testing This material is discussed in the NSSS Vendor's SAR.

16.4.4 Containment Structure Leakage Rato Te d I.pplicability Applies to containment structure leakage rate testing with a partial secondary containment.

Obiective

1. To ensure that the leakage rate of the primary con-tainment structure and associated ' systems is held within allowable limits.

2. To ensure that perionic leakaae rate testing of contain-ment penetrations and isolation valves in performed during the service life of the containment structure.

3. To apply Appendix J of 10CFR50 to light water nuclear power plants with partial secondary containments.

Appendix J does not address partial secondary containmmts .

16

4. To assure that the direct leakage rate to the environ-ment f rom the primary containment structure is below the limit stated herein.

Specification A testing program shall be implemented for Type A, L, and C con-tainment structure leakage tests, as defined in paragraphs II.F, II.G, and II .li of Appendix J to 10CFR50, " Primary Reactor Containment Leakage Testing for Water-cooled Power Reactors,"

published in the Federal Register, Volume 38,1;o. 30, on February 14 , 1973, and as corrected on March 6, 1973 (herea tter referred to as Appendix J) . Preoperational and periodic leakage rate tests shall be conducted in accordance with Appendix J with exceptions as noted below. In addition, Type C tests shall 16.4-3 Amendment 16 8/29/75 67U 79

SWESSAR-P1 conform to the inservice testing requirements stated in Section 16.4.2.

The test program may also include (Utility-Applicant. option)

Type C' tests as detined below. Alternately, if Type C' tests are not perf ormed, the type C' leakage rates shall then be assumed zero. Type C' tests are not required when an enclosure building is used. A discussion of Type A, B, and C tests for Options A and B are included in Sections A16.4.4 and B16.4.4 of Appendixes A and B respectively.

A. Type a Tests Pretest requirements shall comply with paragraph 111.A.1 of Appendix J except that Type B, C, and C' tests shall be perrormed prior to the Type A test. Repairs or adjustments shall be made betore testing to correct abnornualities found in the pretest inspection.

Periodic Type A leakage rate tests shall be scheduled in accord-ance with paragraph III.D.1 of Appendix J.

All Type a tests shall be conducted in accordance with ANS1 N45.4-1972 " Leakage Rate Testing of Containment Structures for Nuclear Reactors" with the following exceptions-

1. Scheduling of leakage rate tests to account for the eftects of weather conditions is not necessary for a concrete containment structure.

2. Leakage rate shall be calculated f rom a linear least sguares tit to the calculated mass of containment air as a function of timo (refer to Section 6.2.6). An instrument error analysis shall also be performed.

3. Leakage rate shall be based on the reference volume or 34 absolute method. The makeup air method shall be used for supplemental verification (ref er to Section 6.2.6) .

4. Reference vessel system if used shall be at a pressure below 0.95 Pa at the start of a Type A test and shall reuuin below P, throughout the test.

Preoperational and periodte Type A tests shall be performed in demrdance with the peak pressure program defined in Appendix J, paragraphs III.A.4 (a) . (2) and III .A.5 (a) . (2) . Test pressure shall be as specitied in Table 6.1-1, the calculated peak containment internal pressure (Pa ). The design basis accident leakage rate (L,) shall be less than 0.2 weight percent per day 6 7 f! ?00g 16.4-4 Amendment 34 7/22/77

SWESSAR-P1 of the containment structure air content at pressure P . The acceptance criteria for Type A tests shall satisfy the requirements of paragraph III. A.4 (b) and III . A.5 (b) of Appendix J. The measured leakage rate (Lam) shall be less than 0.75 L 3 (L am <0.15 percent per day).

The partial secondary containment is assumed to collect 50 percent of the leakage from the primary containment. To establish an upper limit on the uncollected, unfiltered release, the measured leakage rate (Lam) shall be less than 0.1 percent per day.

The Utility-Applicant may propose a test to quantify bypass leakage (Type C' test). If approved for the Utility-Applicant's 21 SAR , the measured primary leakage rate shall be less than 0.1 percent per day, plus the total Type C' measured leakage rate (L c'm) , as described under Type C' tests. Thus, the maximum allowable Type A test measured leakage rate L am 18 1 0.1 + L',*

the minimum of  ; 1 vt %/ day 0.15 The additional leakage testing requirements of paragraph III.A.6 of Appendix J shall be satisfied in the event of Type A test tailure.

b. Type B Tests Preoperational and periodic Type B tests shall be performed at a test pressure equal to P a. Test methods are described in Section 6.2.6. Type B tests on air locks shall be conducted at intervals specified in " Reactor Containment Leakage Testing Requirements," Draft 1, by ANS Committee N274, Work Group 56.8, dated April 22, . 75, paragraph 5.3.3 (1) as follows:

" Personnel air locks shall be tested prior to initial criti-cali y and at six month intervals thereaf ter at an internal pressure of P a. Air locks opened during periods when containment integrity is not required need be tested only at the end of these periods. For air locks opened when containment integrity is required, the air locks shall be tested within three days after such opening. For air locks opened more frequently than once every three days, the air lock shall be tested at least once every three days.

a. For air lock doors having testable seals, testing the seals fulfills the three day test requirements. The test pressure shall be in accordance with door manuf a cturer 's recommenda tions . Seal tests shall not be substituted for the six month air lock test.

16.4-5 Amendment 21 2/20/76 67^ ^^l

SWESSAR-P1 b ., For containments utilizing continuous leakage monitoring systems, only the six month testing requirements need apply to air locks.

Other Type B tests shall be conducted during each reactor shutdown for refueling at intervals not greater than two years.

The a cceptance criteria for Type B tests shall satisfy the requirements of paragraph III.B.3 of Appendix J. The total of Type B and C measured leakage shall be less than 0.6 L g, which equals 0.12 percent per day.

C. Type C Tests Preoperational and periodic Type C tests shall be performed at a test preasure equal to P 3 Test methods are described in Section 6.2.6. Type C tests shall be conducted during each reactor shutdown for refueling but in no case at intervals greater than two years. The acceptance criteria for Type C tests shall satisfy the requirements of paragraph III.C.3 of Appendix J. The total of Type B and C measured leakage shall be less than 0.6 L,, which equals 0.12 percent per day.

D. Type C' Tests Type C' tests shall include only isolation valves which can be tested in a manner such that the leakage measured is the minimum leakage from the primary containment to the secondary containment through a particular leakage path. Test methods are described in Section 6.2.6.

Leakage through containment isolation valves in lines where bypass leakage may occur shall not be included in the total Type C' leakage (L cs, ) .

Type C' tests shall only be performed if the Utility-Applicant proposes a test program to quantify the bypass 21 leakage, and this program is approved for the Utility-Applicant's SAR.

E. Special Testing Requirements Type A, B, and C tests, as applicable, shall be conducted following containment structure modifications in accordance with paragraph IV (A) of Appendix J.

f ,r n07 O / u, 'J' 16.4-6 Amendment 21 2/20/76

SWESSAR-P1 F. Inspection and Reporting on Tests A general pretest inspection of the containment structure shall be performed in accordance with paragraph V (A) of Appendix J. Technical reports for preoperational and periodic tests shall be submitted in accordance with the requirements of paragraph V (B) of Appendix J.

Bases

1. Appendix J to 10CFR50

2. The maximum allowable containment leak rate (L a ) of

02. percent per day is chosen to ensure that the radio-logical consequences of the design basis accident are below the limits suggested in 10CFR100.

3. The maximum allowable measured leakage rate (L ,) is based on the following:

L = Primary containment design leak rate d

= 0.2 wt %/ day L

= Maximum allowable primary containment leak rate = 0.2 wt %/ day L.*

= Maximum allowable measured primary containment leak rate

= 0.75 L a = 0.15 wt %/ day However, the radiological consequences of an accident are con-servatively based on 0.1 p' . cent per day uncollected leakage and 0.1 percent per day leakage that is collected and treated before release. Therefore, a more restrictive Type A test limit than La = 0.15 percent / day is defined which is 0.1 wt %/ day. This 21 va$ue of Lam may be increased by the amount of any leakage that can be definitely determined to be collected and processed before release if the Utility-Applicant has proposed an acceptable test program in his SAR.

4. Type B and C tests ensure that leakuge through contain-ment isolation valves and penetrations is less than 0.6 L a. This provides a high probability that the Type A measured rate (Lam) will be below 0.75 L .

5. Type C' tests ensure that any Type A measured leakuge above 0.1 percent per day will be collected and treated before release.

6. The peak containment internal pressure related to the design basis accident (Pa ) is calculated by means of the LOCTIC computer code as described in Section 6.2.1.

b ~! b, nn 16.4-6A 4 d ) Amendment 21 2/20/76

SWdSSAR-P1 heterence Sections Title No.

Con t.alnmen t. Icakage Monitoring System o.2.6 Contaalment Funct lonal Design o.2.1 16.4.5 dnylneered Sa r eta Features npplies to the testing or the engineered safety Icatures (dSF) systems.

Objective

'Ib verify that the r.SF systems respond promptly and perrorm their design tunction, it required.

Specitication A. Con tainment Spray System

1. Each containment spray pump shall be rlow tested at design flow rate at least once a month 1. accordance with the method stated in Section 6.2.2.4 and the requirements stated in Section 16.4.2.

2. All valves in the containment spray system shall be periodically tested as specified in Section 16.4.2. In audition, all containment isolation valves in the containment spray system shall be given Type C tests as specified in Sec. ion 16.4.4. The Type C tests on the containment isolation valves are describea in Section 6.2.6.

33

3. Tne nozzles in tne containment spray system headers shall be tested by air flow every 5 years tor blockage.

(3 l 0 u n r\ A

4. U T O

16.4-6B Amendment 33 6/30/77

SWESSAR-P1 The air flow test of the containment spray system nozzles shall be considered satisfactory if flow can be detected through each nozzle.

B. Containment Atmosphere Recirculation System Each containment atmosphere recirculation subsystem shall be tested to verify the proper operation of all features which are required during a DBA. This test is performed prior to each startup from cold shutdown.

C. Emergency Core Cooling System (ECCS)

Refer to Chapter 16 of the NSSS Vendor's SAR for ECCS tests.

D. Containment Isolation System

1. The containment as a whole, certain penetrations, and the containment isolation valves shall be test ed for leakage using test methods, acceptance criteria, and time intervals specified in Section 16.4.4.

2. All valves in the containment isolation system shall be tested periodically as specified in Section 16.4.2.

The test of an automatic trip valve in the containment isolation system shall be considered satisfactory if that valve's status light on the main control board is actuated, thus indicating that the valve has tripped.

E. Combustible Gas Control System

1. Each hydrogen recombiner shall be tested at least once a year as specified in Section 6.2.5.

2. The hydrogen analyzers shall be tested at least once per month as specified in Section 6.2.5.

3. Safety Class 2 valves in the combustible gas control system shall be periodically tested as s pecified in Section 16.4.2 'for all valves) and Section 16.4.4 (for containment iso.ation valves).

F. Containment Pressure Monitoring System

1. Each pressure transmitter shall be calibrated at least once a year by pressurizing with air and comparing the pressure indication to that of an accurate mechanical gage.

The pressure transmitter shall be considered satisfactory upon successful calibration.

16.0-7 f, ,qg Amendment 8 0/gV CVJ 3/28/75

SWESSAR-P1 Bases The ESF systems are required to operate following a LOCA or a steam line break accident inside the containment. Complete systen tests cannot be performed when the reactor is operating.

Therefore, the method.of ensuring operability of the systen is to combine system tests that can be performed at plant shutdown with more frequent component tests that can be performed durina reactor operation.

The flow testing of each containment spray pump is performed by opening the normally closed valve in the containment spray pump test line to the RWST. The containment spray pump is operated, and a quantity of water sufficient to obtain adequate test results is returned to the RWST through a strainer. This ensur es that there is no particulate material in the RWST large enough to clog the spray no les inside the containment. Clogging is determined by a decrease in flow rate relative to the base flow rate determined during preoperational testing.

The hydrogen analyzer is tested independently of the hydrogen recombiner system using sample gases.

Reference Sections Title No.

Containment Punctional Design 6.2.1 Containment Heat Bemoval Systems 6.2.2 Containment Isolation Systems 6.2.4 Combustible Gas Control in Containment 6.2.5 Containment Pressure Monitoring System 7.3.3.9 1

Eules for Inservice Inspection and Testing 16.4.2 8 Containment Structure Leakage Rate Tests 16.4.4 16.4.6 Emergency Power System Tests Apolicability Applies to periodic testing and surveillance requirements of the emergency power system.

Obiective To verify that the emergency power system will respond promptly and properly when required. Specification Specification The rollowing tests and surveillance shall be performed as stated:

16.4.6.1 Diesel Generators k

1. Test and Frequencies /7r, ,n ,

b/ U CUO 16.4-8 Amendment 8 3/28/75

SWESSAR-P1

a. The diesel generator will be started, synchronized with other plant power sources, and loaded up to the nameplate rating by manual operation. This test will be conducted monthly on each diesel generator. Normal plant operation will not be af fected by this test .

b. Combination of accident signals and/or loss of preferred power signal will be simulated. This test will be conducted monthly during normal plant operation. The operability of the automatic start of the diesel generator along with all the components within the bus load shedding and load reapplication sequence will be verified. The verification of automatic trip and automatic close of the driven equipment electrical breakers may be a part of this test or conducted as a separate overlapping test.

c. Combination of accident signals and/or loss of pref erred power signal will be simulated. The test will be conducted at approximately one year intervals normally during rea ctor shutdown. The operability of the automatic start of the diesel generator along with load shedding tollowed by the emergency generator automatically restoring power to the bus and acceptina the reapplication of load in the proper sequence will be verified. The diesel generator voltage and frequency will be observed during this test to verify compliance with Regulatory Guide 1.9 (Section 3A.1-1. 9) .

d. Each diesel genera tor shall be given a thorough inspection at least annually following the manufacturer's service.

recommendations for this class of standby

2. Acceptance Criteria The above tests will be considered satisf actory if all applicable e;uipment operates as designed.

16.4.6.2 Fuel Oil Storage Tanks for Diesel Generators A mininum fuel oil supply shall be maintained in each underground storage tank to assure full power operation of each diesel generator for seven days. Each fuel oil storage tank will be smmpled monthly to detect the presence of water.

16.4.6.3 Station Batteries

1. Test and Frequencies

a. The specific gravity, electrolyte temperature, and cell voltage of the pilot cell in each 60-cell battery, and 16.4-9 Amendment 8

, 3/29/75 670 c0,/

SWESSAR-P1 the d-c bus voltage of each battery shall be measured and recorded weekly.

b. Each month the voltage of each battery cell in each 60-cell battery shall be measured to the nearest 0.01 V and recorded.

c. Each month the specific gravity of each battery cell, the electrolyte temperature reading of every fifth cell, the height of electrolyte of each cell, and the amount of water added to any cell shall be measured and recorded.

d. Twice a year, during normal operation, each battery charger shall be turned off for approximately 30 min, and the battery voltage and current shall be recorded at the beginning and end of the test.

e. Once a year, during the normal ref ueling shutdown, each battery shall be subjected to a simulated load test without the battery charger. The battery voltage and current as a function of time shall be monitored.

2. Acceptance Criteria

a. Each test shall be considered satisfactory if the new data when compared to the old data indicate no signs of abuse or deterioration.

b. The load tests in 1d and le shall be con sidered satisfactory if the batteries perform within acc eptable limits as establisned by the manufacturer's discharoe characteristic curves.

Basis The tests specified are designed to demonstrate that the emergency diesel generators will provide power for operation of ESF equipment. They also assure that the emergency diesel generator system controls and the control systems for the ESF e quipment will function automatically in the event of a loss of all normal station service power.

The tests specified will be performed often enough to identify and correct any mechanical or electrical deficiency before it can result in a system failure. The fuel supply and starting circuits and controls are continuously monitored and any taults are clarm indica ted . An abnormal condition in these systems would be signaled without having to place the emergency diesel generators themselves on test.

S tation batteries may deteriorate with time, but precipitous f ailure is extremely unlikely. The surveillance specified is that which has been demonstrated by experience to provide an '

16.4-10 Amendmer.t 8 3/28/75 1 /v ano Gyu

SWESSAR-P1 indication of a cell becoming unserviceable long before it fails.

In addition, alarms are provided to indicate low battery voltage and low current from the chargers which would make it extremely unlikely that deterioration would go unnoticed.

Section 8.3.2 provides further amplification of the basis.

Reference Sections Title No.

A-c Power Systems 6.3.1 8 D-c Power Systems 8.3.2 16.4.7 Main Steam Isolation Valves Applicability Applies to periodic testing of the main steam isolation valves.

Obiective To verify the ability of the main steam isolation valves to close on signal.

Specifica tion A. Tests and Frecuencies

1. Each main steam isolation valve shall be tested ror full closure under cold conditions approximately once during each scheduled refueling shutdown.

2. Each main steam isolation valve shall be inservice tested for partial closure at least once a month. This test is performed locally.

3. Each main steam isolation valve shall be leakage tested as specified in Sections 16.4.2 and 16.4.4. 8 B. Acceptance Criteria

1. A full closure test of a main steam isolation valve shall be considered satisfactory if the valve closes fully within the time period specified by the NSES Vendor (see Fig. 10.1-1 for specified time) .

2. A partial closure inservice test of a ma in steam isolation valve shall be considered satisfactory if the valve can be stroked at least 6 degrees f rom its f ull open position.

16.4-10A Amendment 8 670 ?no 3 28 7s LU/

SWESST.R-P 1 8

3. Leakage test acceptability shall be as specified in Sections 16.4. 2 and 16.4.4 Basi _s The main steam isolation valves serve to limit the rea ctor coolant system cooldown rate and resultant reactivity in sertion following a main steam pipe break accident. Their ability to close fully is verified at each scheduled refueling shutdown.

The specified closure time is consistent with the requirement to prevent accidental main steam system depressurization, an described in Section 15.1.14. The inservice testing of a pa rtial valve stroke takes place to verify the freedom or the valve to function as req uired . During inservice testing, the valve is prevented f ran entering the main steam system flow stream enouah to impede steam flow. Verification of correct opera tion is obtained f rom indicating lights and an annunciator in the control room as well as f raa local observation.

Ref erence Section Title No.

Main Steam System 10.3 16.4.8 Auxiliary Feedwater Pumps Applicability Applies to periodic testing requirements of the turbine-driven and motor-driven auxiliary feedwater pumps.

Obiective To verify the operability of the auxiliary feedwater pumps and their ability to respond properly when required.

Specification A. Tests and Frecuency

1. The motor-driven auxilia ry feedwater pumps shall be g tested on a monthly ba sis to demonstrate their operability, os specified in Section 16.4.2.

2. The turbine-dr iven auxiliary feedwater pumo shall be g tested on a monthly basis to demonstrate its operability, as specified in Section 16.4.2.

i 3. The auxiliary feedwater pump discharge valves shall be B ,

exercised on a monthly basis, as part of the pump test.

k

b. Acceptance Criteria / ~7 n n .

O/'. ciU 16 .4 - 10 B Amendment 8 3/28/75

SWESSAR-P1 These te sts shall be considered satisfactory it the requirements specified in Section 16.4.2 are met. 8 Basis The auxiliary feedwater pumps are tested monthly to demonstrutr their operability by recirculation to the auxiliary f eedwa t o r storage tank.

Proper functioning of the steam turbine drive admission valves, f eedwater pump discharge valves , and the ability of the auxiliary feedwater pumps to start demonstrate the integrity of the system.

Verification or correct operation can be made from 16.4- 11 Amendment 8 bl[] a.,

cii 3/28/75

SNESSAR-P1 instrumentation within the control room, on the auxilia ry shutdown panel, and from direct visual observation of the pumps.

Reference Section Title No.

Auxiliary Feedwater System 10.4.10 16.4.9 Rea cti vity Anomalies This section is discussed in the NSSS Vendor's SAR.

16.4.10 Effluent Sampling and Radiation Monitorina Systems Applicability Applies to the surveillance requirements for controlling radioactive liquid and gaseous waste releases and potential paths to the environment.

Design Obiectives To verify that the annual average discharge of radioactive materials to the environment is maintained as low as practicable and to comply with 10CFR20 and 10CFR50, AEC General Design Criterion 64, and Regulatory Guide 1.21.

Specifications A. Liquid Effluents

1. Treatment and Monitoring

a. Liquid waste discharged from the waste test tank is monitored during release. The liquid effluent monitor readings are compared with the allowable reading of each discharge batch. The monitors are tested daily and calibrated during each refueling.

The calibration procedure consists of exposing the detectcr to a referenced calibration source in a controlled, reproducible geometry. The sources and geometry are referenced to the original monitor calibration which provides the applicable calibration curves.

b. The liquid waste monitor is set to alarm and automatically close the waste discharge valve to assure the requirements of Section 16.3.8 are met.

In the event of a malfunction in the monitor, an alarm sounds in the control room and the operator may manually close the valve.

7n ,

16.4-12 b/U c

SWESSAR-P1

c. Steam generator blowdown discharge is continuously monitored except that, during periods when the monitor is not operating, daily grab samples are taken.

2. Sampling and Analysis In addition to the above monitoring requirements, the following specific sampling and radionuclide analyses are performed:

a. Radioactive liquid wastes discharged from the radioa ctive liquid waste system test tanks are sampled and analyzed as follows:

(1) Each batch is sampled and analyzed for gross beta and gamma activity. (2) One batch per month is analyzed for dissolved radioactive gases. (3) Representative samples from each batch released over the period of a week are composited for analysis of I-131, La-140, and Ba-140. For each batch released over the period of a month representative samples are composited for laboratory activity analysis. This composite sample is analyzed for gross alpha, beta, and gamma activity, and for identification of individual radionuclides. Individual radionuclide identification is such that at least 90 percent of the gross beta and gamma radioactivity and the tritium activity in a sample is accounted for by the analysis.

b. Liquid effluents discharged from the steam and power conversion systems are sampled and analyzed as follows:

(1) Blowdown and condensate are sampled weekly and analyzed for gross beta and gamma activity and tritium. (2) Blowdown and condensate samples are taken monthly and analyzed for dissolved noble gases. (3) Representative samples from blowdown and condensate are taken weekly and composited for laboratory activity analysis. These composite samples are analyzed monthly for gross alpha, beta, and gamma activity and for identification of individual radionuclides. Individual radionuclide identification is such 16.4-13 f/

SWESSAR-P1 that at least 90 percent of the gross beta and gamma radioactivity and the tritium in a sample is accounted for by the analysis.

c. Records are maintained and reports submitted in eccordance with Section 16.6.

B. Gaseous Effluents

1. Treatment and Monitorina

a. During release of radioactive gaseous waste from the discharge line to +he ventilation vent, the following conditions are met:

(1) The process gas (hydrogenated) , process vent (aerated) , and steam jet air ejector monitors, and the iodine absorber (for laboratory analysis) in the discharge line to the ventilation vent are normally operable. If any of these monitors are temporarily inoperative, grab samples from the respective system are taken by the operator. The normal response of the above monitors is verified by comparison with the prior cample analyses. The monitors are tested daily when radioactive gases are being released and are calibrated during refueling. The calibration procedure consists of exposing the detectors to a referenced calibration source in a controlled reproducible geometry. The sources and geometry are referenced to the original monitor calibration which provides the applicable calibration curves. (2) The process gas and process vent gaseous waste is filtered through high efficiency particulate filters.

b. Process gas waste is processed through charcoal decay beds with a minimum selective holdup of 60 days for xenon and 4 days for krypton.

c. During power operation, the steam jet air ejector discharge is monitored for gross activity.

d. Ventilation flows from the charging pump cubicles in the annulus building and from the annulus, fuel, and solid waste and decontamination buildings are monitored individually.

e. Gases discharged through the ventilation vent are monitored for gross noble gas and particulate I activities by a ventilation vent airborne radiation 16.4-14 e-g [ r;u ?

c. '1 $

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SRESSAR-P1 monitor. If this monitor is inoperable, appropriate grab samples will be taken and analyzed daily.

2. Sampling and Analysis In addition to the above monitoring requirements, the following rpecific sampling and radionuclide analyses are performed:

a. The radioactive gaseous waste discharge line to the vent is sampled continuously for iodines and particulates. The iodine sample is analyzed weekly for I-131 and monthly for I-133 and I-135 and the particulate samples are collected weekly for gross beta, gamma activity and composited for monthly analysis. This monthly analysis consists of gross alpha, gross beta, and gamma measurements, and identification of individual radionucliden.

Individual radionuclides are such that at least 90 percent of the gross beta and gamma activity in a sample is accounted for by the analysis. In addition, a sample from the discharge line is analyzed quarterly for tritium vapor.

b. Prior to purging, the containment atmosphere is sampled for noble gases, I-131, tritium vapor, and particulates. The radioactive gases are analyzed for gross beta and garra activity and identification of individual radionuclides. The particulate sample is analyzed for gross alpha, gross beta, and gamma activity, and for identification of individual radionuclides.

Reference Sections Title No. Air Conditioning, Heating, Cooling, and 9.4 Ventilation Systems Condenser Evacuation System 10.4.2 Radioactive Liquid Waste System 11.2 Radioactive Gaseous Waste System 11.3 Process and Effluent Radiation 11.4 Monitoring Systems Effluent Release Limits 16.3.8

                                                                 ,,c 67C      c" 16.4-15

SWESSAR-P1 TABLE 16.4.2-1 IDENTIFICATION OF CODE CLASS 2 SYSTES REQUIRITU I!4 SERVICE INSPECTION

1. Pressure containing components of the reactor coolant pressure boundary not covered in Code Class 1

2. Residual heat removal system

3. Letdown and makeup portion of the chemical and volume control system

• 4. Containment spray system
• 5. Low pressure safety injection system
• 6. High pressure safety injection system

7. Main steam system from and including the steam generator 4 secondary up to and including the outermost main steam isolation valve

8. Feedwater and auxiliary feedwater systems f rom the outermost containment isolation valves up to the steam generator.

• 9. Combustible gas control system

10. Reactor plant component cooling and service water systems inside the containment structure and servicing safety related components .

For the Code Class 2 systems listed above, inservice inspection of components and parts will be performed in accordance with the methods and frequencies given in ASME XI, Subsection IWC, Article IWC-2000.

• Code Class 2 systems marked with an asterisk (*)

require only visual inspection provided that the chemis try of the contained fluid is periodically verified by sampling to ensure that it is non-corrosive to the piping system. ,

                                                    ,i/

1 of 1 ()7 O LLU Amendment 4 11/1/74

Sb1SSAR-P1 Tables 16.4.2-2 and 16 .4.2-3 have been deleted. 6 a1 7

                          /0-[ rU LII        A:r.endment 6 1/17/75

SWESSAR-P1 Fig. 16.4.2-1 has been deleted. E ni0 (- Amenchnent 8

                      / i 3/28/75

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SWESSAR-P1 16.5 DESIGN FEATURES 16.5.1 Site This material is discussed in the Utility-Applicant's SAR. 16.5.2 Containment Applicability Applies to those design features of the containment and containment systems relating to operational and public safety. Obiective To define the significant design features of the containment and containnent systems. Specifications 16.5.2.1 Containment

1. A containment is provided, which completely encloses the reactor and reactor coolant system, to ensure that an acceptable upper limit for leakage of radioactive raaterials to the environment is not exceeded even if gross failure of the reactor coolant system occurs. The containment provides biological shielding for loth normal operation and postulated accident conditions.

The containment is designed for internal pressures up to 63 psia.

2. The containment is designed for the reactor core operating at the ultimate rated thermal power.

3. In combination with the internal design pressure and other appropriate loads, the containment is designed to withstand loads resulting from a safe shutdown earthquake (SSE) .

16.5.2.2 Containment Penetrations

1. All penetrations through the containment for pipe, electrical conductors, ducts, and access hatches are of the double barrier type.

a. The equipment hatch is of a bolted flange, double gasketed cons truction , with means provided for introducing Preon gas between the gaskets for test purposes.

b. The personnel hatches are double-closure penetration with interlocks so that when one door o nn 16.5-1 b70 L'd

SWESSAR-P1 is open, the other door cannot be opened during normal operation. ,

2. Containment isolation valves in the automatically actuated containment isolation system are designed to close as outlined below. The containment isolation system is designed such that no single component failure prevents accomplishing containment isolation if required. Refer to Section 7.3 for setpoint values or the various signals which actuate containment isolation valves.

a. A containment isolation phase A signal (CIA) closes the containment isolation valves in lines which are not required for an orderly and safe hot shutdown of the unit. This protects all equipment required for resumption of operation once the initiating cause for the CIA signal has been corrected.

b. A containment isolation phase B (CIB) signal completes conta inment. isolation (except for containment isolation valves required to be open for operation of engineered safety features) .

c. Containment isclation is accomplished manually from the mein control board in the control room if the automatic signals fail to actuate the above valves.

d. When instrument air supply is lost, all air operated containment isolation valves fail in the closed position.

Manual containment isolation valves normally kept closed are under administrative control. Table 6.2.4-1 lists all containment piping penetrations and the signals which actuate the various containment isolation valves. Table 6.2.4-2 lists all instrument lines penetrating the containment. 16.5.2.3 Containment Systems

1. Following a loss-of-coolant accident, the minimum ESF given in Table 6-1 for the containment spray system shall distribute 3,500 gpm borated spray water containing sodium hydroxide for removing heat and iodine from the containment atmosphere.

2. The containment ventilation system consists of the containment atmosphere recirculation system, containment atmosphere filtration system, and a containment purge air syste;a. During normal operation these systems are ' ' ,

designed to limit ambient air temperature inside the 16.5-2 - rr. 1 (j / y CL I

SWESSAR-P1 con ta inmen t, provide a suit able environment for personnel and equipment, and limit the spread of any radioacti contamination. The containment atmosphere recircula t - :>n system is required to operate during a LOCA in conjunction with the containment spray system to cool the containment atmosphere. It is therefore an engineered safety feature.

3. The combustible gas control system consists of the hydrogen recombiner subsystem and dilution air subsystem. The hydrogen recombiner system consists of two redundant hydrogen recombiners and is used after a LOCA to decrease the concentration of hydrogen in tho containment atmosphere. The dilution air subsystem serves as a backup to the hydrogen recombiner system.

4. The containment leakage monitoring system is used to determine the leakage rate out of the containment structure during periodic testing.

5. The containment pressure monitoring system initiates the engineered safety features actua tion system.

Reference Sections Title No. Seismic Design 3.7 Concrete Containment 3.8.1 Containment Functional Design 6.2.1 Containment IIeat Removal Systems 6.2.2 Containment Isolation System 6.2.4 Combustible Gas Control System 6.2.5 Containment Leakage Monitoring System 6.2.6 Engineered Safety Features Systems 7.3 Containment Pressure Monitoring System 7.3.3.9 Containment Atmosphere Recirculation System 9.4.5.1 Containment Purge Air System 9.4.5.2 Containment Atmosphere Filtration System 9.4.5.3 Shielding 12.1 16.5.3 Reactor This material is discussed in the NSSS Vendor's SAR. 16.5.4 Fuel Storage Anplicability Applies to the design of the new and spent fuel storage areas. 16.5-3 h li n G '77 LL-

SWESSAR-P1

'sbiective To   define those d.esign aspects      of fuel   storage designed to prevent criticality in fuel storage areas, to    prevent mechanical damage to the spent        fuel elements, and to prevent inadvertent draining of water from the spent fuel storage    area.

Specification A. The Seism:;c Category I portion of the fuel buildinc, including the spent fuel storage racks, shall be designed to withstand the safe shutdown earthquake.

b. The new and spent fuel storage racks shall be designed to store the fuel vertically in an array with sufficient center-to-center distance between assemblies tc, ensure K 50.95, even if covered with unborated water. The design shall be such that fuel assemblies can only be inserted in their proper locations.

C. The fuel building crane rails shall be located such that the spent f uel shipping cask cannot be transported over the fuel pool. D. The fuel pool shall be designed such that failure or malfunction of permanently connected systens canno t cause the uncovering of the stored fuel. E. All penetrations into the fuel pool shall be at least 11 ft above the top of the spent fuel. Reference Sections Title No. New Fuel Storage 9.1.1 Spent Puel Storage 9.1.2 Fuel Pool Cooling and Purification System 9.1.3 Fuel Handling System 9.1.4

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SWESSAR-P1 16.6 ADMINISTRATIVE CONTROLS This material is discussed in the Utility-Applicant's SAR. $ 'h W 7 j s (j j i; L'I 16.6-1}}