ML19340D334

Proposed Changes to Tech Specs Sections 3,4 & 6 Re TMI Lessons Learned Category a Procedural & Equipment Requirements

ML19340D334
Person / Time
Site:	Prairie Island
Issue date:	12/19/1980
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML19340D332	List: ML19340D331 ML19340D333 ML19340D334
References
RTR-NUREG-0578, RTR-NUREG-578 TAC-12428, TAC-12429, NUDOCS 8012300310
Download: ML19340D334 (28)

Text

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EXHIBIT B License Amendment Request dated December 19, 1980 Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Exhibit B consists of revised pages for the Prairie Island Nuclear Generating Plant Technical Specifications, Appendix A, as listed below showing the proposed changes:

TS-i TS-iii TS-3.1-2 IS.3.1-3 TS.3.1-3A

• TS.3.4-1

• TS.3.4-2

• TS.3.4-3

• • TABLE TS.3.5-1

• • TABLE TS.3.5-3 (Page 1 of 2)

TABLE TS.3.5-3 (Page 2 of 2)

TABLE TS.3.5-4

• • TS.3.5-2 TS.3.5-3 TS.3.5-4 TS.3.5-5 TS.3.15-1 TABLE TS.3.15-1 TABLE TS.4.1-1 (Page 5 of 5)

TABLE TS.4.1-2A TS.4.6-1A TS.4.6-3

• TS.4.8-1

• TS.4.8-2 TS.6.1-2 TABLE TS.6.1-1 TS.6.5-2

• - This page previously submitted with License Amendment Request dated December 31, 1979

, ** - This page previously submitted with License Amendment Request dated May 6, 1980 l

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TS-1 RE7 APPENDIX A T1CHNICAL CPECIFICATIONS TABLE OF CONTENTS TS SECTION TITLE PAGE 1.0 Definitions TS.1-1 2.0 Safety Limits and Limiting Safety System TS.2.1-1 Settings 2.1 Safety Limit, Reactor Core TS.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure TS.2.2-1 2.3 Limiting Safety System Settings, Protective TS.2.3-1 Ins t rument ation 3.0 Limiting Conditions for Operation TS.3.1-1 2.1 Reactor Coolant System TS.3.1-1 3.2 Chemical and Volume Control System TS.3.2-1 3.3 Engineered Safety Features TS.3.3-1 3.4 Steam and Power Conversion System TS.3.4-1 3.5 Ins trumentation System TS.3.5-1 3.6 Containment System TS.3.6-1 3.7 Auxiliary Electrical Systems TS.3.7-1 3.8 Refueling and Fuel Handling TS.3.8-1 3.9 Radioactive Effluents TS.3.9-1 3.10 Control Rod and Power Distribution Limits TS.3.10-1 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Shock Suppressors (snubbers) TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13-1 3.14 Fire Detection and Protection Systems TS.3.14-1 3.15 Event Monitoring Instrumentation TS.3.15-1 l 4.0 Surveillance Requirement s TS.4.1-1 4.1 Operational Safety Review TS.4.1-1 4.2 Primary Sys tem Surveillance TS.4.2-1 4.3 Reactor Coolant System Integrity Testing TS.4.3-1 4.4 Containment System Tests TS.4.4-1 4.5 Engineered Safety Features TS.4.5-1 4.6 Periodic Testing of Emergency Power System TS.4.6-1 4.7 Main Steam Stop Valves TS.4.7-1 4.8 Steam and Power Conversion System TS.4.8-1 l 4.9 Reactivity Anomalies TS.4.9-1 4.10 Radiation Environmental Monitoring Program TS.4.10-1 4.11 Radioactive Source Leakage Test TS.4.11-1 4.12 Steam Generator Tube Surveillance TS.4.12-1 4.13 Shock Suppressors (snubbers) TS.4.13-1 4.14 Control Room Air Treatment System Tests TS.4.14-1 4.15 Spent Fuel Pool Special Ventilation System TS.4.15-1 4.16 Fire Detection and Protection Systems TS.4.16-1

TS-lii REV APPENDIX A TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE 3.1-1 Unit 1 Reactor Vessel Toughness Data 3.1-2 Unit 2 Reactor Vessel Toughness Data 3.5-1 Engineered Safety Features Initiation Instrument Limit ing Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Instrument Operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating Conditions for Isolation Functions 3.5-5 Instrument Operating Conditions for Ventilation Systems 3.9-1 Radioactive Liquid Waste Sampling and Analysis 3.9-2 Radioactive Gaseous Waste Sampling and Analysis 3.12-1 Safety Related Shock Suppressors (Snubbers) 3.14-1 Safety Related Fire Detection Instruments 3.15-1 Event Monitoring Instrumentation 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requiremants 4.4-1 Unit 1 and Unit 2 Penetration Designation for Leakage Tests 4.10-1 Prairie Island Nuclear Generating Plant-Radiation Environmental Monitoring Program Sample Collection and Analysis Environmental Monitoring Program 4.12-1 Steam Generator Tube Inspection 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating Plant (Per Unit) 5.3-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shif t Crew Composition 6.7-1 Special Reports l

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5 TS.3.1-2 REV

3. Pressurizer

a. Whenever average reactor coolant system temperature is above 350 F or the reactor is critical, the pressurizer shall be operable with:

1. Steam bubble

2. Pressurizer heater groups "A" and "B" and their associated safeguards power supplies operable

3. At least one operable spray

b. With the pressurizer inoperable due to an inoperable heater group restore the equipment to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or place the reactor in at least Hot Shutdown.

c. With the pressurizer inoperable for any other reason than (b) above, the reactor shall be placed in at least Hot Shutdown.

d. At least one pressurizer safety valve shall be operable whenever the head is on the reactor vessel, except during hydrostatic tests. Both pressurizer safety valves shall be operable whenever average reactor coolant system temperature is above 350 F or the reactor is critical. Pressurizer safety valve lift setting shall be 2485 psig ! 1%.

e. Except as specified in (f) and (g) below, two power operated relief valves (PORV's) and their associated block valves shall be operable whenever average reactor coolant system temperature is above 350 F or the reactor is critical.

f. With one or more PORV's inoperable, within one hour either restore the PORV(s) to operable status or close the associated block valve (s). If this cannot be done, place the reactor in the Cold Shutdown condition.

g. With one or more blocks valves inoperable, within one hour either restore the block valve (s) to operable status or close the valve. If this cannot be done, place the reactor in the Cold Shutdown condition.

TS.3.1-3 REV Basis When the boron concentration of the reactor coolant system is to be reduced, the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place.

The residual heat removal pump will circulate the equivalent of the primary system volume in approximately one-half hour.

" Steam Generator Tube Surveillance", Technical Specification 4.12, identifies steam generator tube imperfections having a depth >50% of the 0.050-inch tube wall thickness as being unacceptable for power operation. The results of steam generator burst and tube collapse tests submitted to the staff have demonstrated that tubes having a wall thickness greater than 0.025-inch have adequate margins of safety against failure due to loads imposed by normal plant operation and design basis accidents.'

Part A of the specification requires that both reactor coolant pumps be operating when the reactor is critical to provide core cooling in the event that a loss of flow occurs. In the event of the worst credible coolant flow loss (loss of both pumps from 100% power) the minimum calculated DNBR remains well above 1.30. Therefore, cladding damage and rel?ase of fission products to the reactor coolant will not occur. Critical operation, except for low power physics tests, with less than two pumps is not planned. Above 10% power, an automatic reactor trip will occur if flow from either pump is lost. Below 10% power, a shutdown under administrative control will be made if flow from either pump is lost.

The pressurizer is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated t rans ien t s . Each of the pressurizer safety valves is designed to relieve 325 000 lbs per hour of saturated steam at the valve set point. Below 3503 F and 450 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure. If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves ' capacity. One valve therefore provides adequate defense against over pressurization of the reactor coolant system for reactor coolant temperatures less than 350 F. The combined capacity ,

of both safety val.as is grgater than the maiximum surge rate resulting from complete loss of load.

t TS.3.1-3A REV Basis (continued)

The requirement that two groups of pressurizer heaters be operable provides assurance that at least one group will be available during a loss of of fsite power to maintain natural circulation. Backup heater group "A" is normally supplied by one safeguards bus. Backup heater group "B" can be manually transferred within minutes to the redundant safeguards bus. Tests have confirmed the ability of either group to maintain natural circulation conditions.

The pressurizer power operated relief valves (PORV's) operate to relieve reactor coolant system pressure below the setting of the pressurizer Code safety valves. These relief valves have remotely operated block valves to provide a positive shutof f capability should a relief valve become inoperable.

The PORV's are pneumatic valves operated by instrument air. They fail closed on loss of air or loss of power to their DC solenoid valves. The PORV block valves are motor operated valves supplied by the 480 volt safeguards buses.

References

! 1 FSAR, Section 14.1.9 2

j Testimony by J Knight in the Prairie Island Public Hearing on January 28, 1975. .

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TS.3.4-1 REV 3.4 STEAM AND POWER CONVERSION SYSTEM Applicability Applies to the operating status of the steam and power conversion system.

Objective To specify minimum conditions of steam-relieving capacity and auxiliary feed-water supply necessary to assure the capability of removing decay heat from the reactor, and to limit the concentration of activity that might be released by steam relief to the atmosphere.

Specification A. A reactor shall not be heated above 350 F unless the following conditions are satisfied:

1. Safety and Relief Valves

a. Rated relief capacity of ten steam system safety valves is available for that reactor, except during testing.

b. Both steam generator power-operated relief valves for that l reactor are operable.

2. Auxiliary Feed System

a. For single unit operation, the turbine-driven pump associated with that reactor plus one motor-driven pump are operable,

b. For two-unit operation, all four auxiliary feedwater pumps are operable.

c. Valves and piping associated with the above components are operable except that during Startup Operation necessary changes may be made in motor-operated valve pos it ion. All such changes shall be under direct ad-ministrative control. -

I d. A minimum of 100,000 gallons of water is available in l the condensate storage tanks and a backup supply of river water is available through the cooling water system.

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TS.3.4-2 REV

e. For Unit 1 operation motor operated valves MV32242 and MV32243 shall have valve position monitor lights operable and shall be locked in the open position by having the motor control center supply breakers manually locked open. For Unit 2, correspond-ing valve conditions shall exist.

f. Essential features including system piping, valves, and inter-locks directly associated with the above components are operable,

g. Manual valves in the above systems that could (if one is im-properly positioned) reduce flow below that assumed for acci-dent analysis shall be locked in the proper pcsition for emergency use. During power operation, changes in valve position will be under direct administrative control.

3. Steam Exclusion System Both isolation dampers in each ventilation duct that penetrates rooms containing equipment required for a high energy line rupture outside of containment shall be operable or at least one damper in each duct shall be closed.

4. Radiochemis try The iodine-131 activity of the water on the secondary side of either steam generator for that reactor does not exceed 0.30 uCi/ce.

B. If, during startup operation or power operation, any of the conditions of Specification 3.4.A., except as noted below for 2.a or 2.b cannot be, met, startup operations shall be discontinued and if operability cannot be res tored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the af fected reactor shall be placed in the cold shutdown condition using normal operating procedures.

With regard to Specifications 2a or 2b, if a turbine driven AFW pump is not operable, that AFW pump shall be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the affected reactor shall be cooled to less than 350 F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If a motor driven AFW pump is not operable, that AFW pump shall be restored to operable status within 7 days, or one unit shall be cooled to le'ss than 350 F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Basis A reactor shutdown from power requires removal of decay heat. Decay beat removal requirements are normally satisfied by the steam bypass to the-con-denser and by continued feedwater flow to the steam generators. Normal feedwater flow to the steam generators is provided by operation of the turbine-cycle feedwater system.

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tS.3.4-3

. REV The ten main steam safety valves have a total combined rated capability of 7,745,000 lbs /hr. ne total full power steam flow is 7,094,000 lbs/hr; l therefore, the ten main steam I total steam flow if necessary.gety valves will be able to relieve the In the unlikely event of complete loss of of fsite electrical power to either or both reactors, continued removal of decay heat would be assured by avail-ability of either the steam-driven auxiliary feedwater pump or the motor-driven auxiliary feedwater pump associated with each reactor, and by steam discharge to the atmosphere through the main steam safety valves.

One auxiliary feedwater pump can supply suf ficient feedwater for removal of decay heat from one reactor. The motor-driven auxiliary feedwater pump for each reactor dan be made available to the other reactor.

L e minimum amount of water specified for the condensate storage tanks is sufficient to remove the decay heat generated by one reactor in the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of shutdown. Essentially unlimited replenishment of the condensate storage supply is available from the intake structures through the cooling water system.

We two power-operated relief valves located upstream of the main steam isolation valves are required to remove decay heat and cool the reactor down following a high energy line rupture outside containment (2). Isolation dampers l are required in ventilation ducts that penetrate those tooms containing equip-ment needed for the accident.

The secondary coolant activity is based on a postulated re' ease of the contents of one steam generator to the atmosphere.(3) This could happen, for example, l as a result of a steam break accident combined with failure of a steam line isolation valve. The limiting dose for this case results from iodine-131 because of its low MPC, and because its long half-life relative to the other iodine isotopes results in its greater concentration in leakage fluid. La accident is assumed to occur at zero load when the steam generators contain maximum water. With allowance for plate-out retention of iodine in water droplets, one-tenth of the contained iodine is assumed to reach the site boundary. The maximum inhalation dose at the site boundary is then as follows:

Dose (rem) = C- V -

B(t) -

X/Q -

DCF 10 bhere: C = secondary coolant activity, 0.30 uCi/cc V = water volume in one steam generator = 3510 f t3 , 99 33 B(t) = breathing rate, 3.47 x 10 M /sec

-4 3 l X/Q = 9.8 x 10 sec/m 6

DCF = 1.50 x 10 rem /Ci 1 131 inhaled l

l The resulting dose is 1.5 rem.

l Re fe rences

! (1) FSAR, Section 10.4

(2) FSAR, Appendix I l (3) FSAR, Section 14 t

TABLE TS.3.5-1 ENGINEERED SAFETY FEATURES INITIATION INSTRUMENT LIMITING SET POINTS FUNCTIONAL UNIT CHANNEL LIMITING SET POINTS

• 1 High Containment Preasure (lii) Safety Injection

• f,4 psig 2 liigh Containment Pressure ( H i-lii ) a. Containment Spray j,23 psig

b. Steam Line Isolation f,17 psig of Both Linen 3 Pressurizer Low Pressure Safety Injection * >l815 psig 4 Low Steam Line Pressure Safety injection * >500 psig Lead Time Constant 712 seconds Lag Time Constant [,2 seconds 5 liigh Steam Flow in a Steam Line Steam Line Isolation d/p correspgnding to Coincident with Safety Injection of Affected Line f,0.745 x 10 lb/hr and Low T at 1005 psig avg

~>540 F E!

6 liigh-h igh Steam Flow in a Steam Line Isolaticn f,d/pcorrespgnding El Steam Line Coincident with of Affected Line to 4.5 x 10 lb/hr M Safety Injection at 735 psig pl ta 7 liigh Pressure Dif ference Between Containment Vacuum f,0.5 psi L, Shield Building and Containment Breakers $

8 liigh Temperature in Vent ilat ion Ducts Ventilation System f,120 F Isolation Dampers

• Initiates also containment isolation, feedwater line isolation and starting of all containment fans.

d/p means dif ferential pressure

TABLE TS.3.5-3 INSTRUMENT OPERATING CONDITIONS FOR EMERGENCY COOLING SYSTEM 1 2 3 4 MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF OPERABLE DEGREE OF BYPASS CONDITIONS OF COLUMN FUNCTIONAL UNIT CllANNELS REDUNDANCY CONDITIONS 1 or 2 CANNOT BE MET

1. SAFETY INJECTION

a. Manual 2 1 Hot shutdown **

b. High Containment Pressure 2 1 Ilot shutdown **

c. Steam Generator Low Steam 2 1 primary pressure Hot shutdown **

Pressure / Loop less than 2000 psig

d. Pressurizer Low Pressure 2 1 primary pressure llot shutdown **

less than 2000 psig

2. CONTAINMENT SPRAY

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a. Manual 2 llot shutdown ** N en E

b. Ili-Ili Containment Pres- llot shutdown ** g sure (Containment Spray) .

Giannel a 2 1 F Channel b 2 1 Y Qiannel c 2 1 Logic 2 1

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TABl.E TS.3.5-3 (continued)

INSTRUMENT OPERATING CONDITIONS FOR EMERGENCY C001.ING SYSTEMS 1 2 3 4 MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF OPERATING DEGREE OF BYPASS CONDITIONS OF COLUMN FUNCTIONSL UNIT CilANNELS REDUNDANCY CONDITIONS 1 OR 2 CANNOT BE MET

3. AUXILI ARY FEEDWATER

a. Steam Generator Low-Low 2 1 llot shutdown Water I.evel

b. Undervoltage on 4.16 KV 2/ bus 1/ bus llot shutdown Buses 11 and 12 (21 and 22 Unit 2)

(Start Turbine Driven Pump only)

c. Trip of Main Feedwater Pumps 2/ pump 1/ pump Ilot shutdown

d. Safety Injection (See Item No. 1) llot shutdown

e. Manual 2 1 ilot shutdown NY "O

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• - Must actuate two, switches simultaneously.

• • - If minimum conditions are not met with in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> , steps shall be taken on the affected unit g to place the unit in cold shutdown conditions, g e

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s TABLE TS.3.5-4 -

INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS 1 2 3 4 MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF OPERABLE DEGREE OF BYPASS CONDITIONS OF COLUMN FUNCTIONAL UNIT CHANNELS REDUNDANCY CONDITIONS 1 OR 2 CANNOT BE MET

1. CONTAINMENT ISOLATION

a. Safety Injection (See Item No. I of Table TS.3.5-3) llot shutdown **

b. Manual 2 1 Ilot shutdown

2. CONTAINMENT VENTILATION ISOLATION

a. Safety Injection (See Item No, I of Table TS.3.5-3) Maintain Purge and Inservice Purge Valves

b. High Radiation in Exhaust Air 2 1 closed if conditions of (a), (b), or (c)

c. Manual 2 1 cannot be met

3. STEAM LINE ISOLATION

a. Ili-fli Steam Flow with 2 1 liot shutdown **

Safety Injection

b. Ili Steam Flow and 2 of 4 Low 2 1 llot shutdown **

Tavg with Safety Injection

c. Ili Containment Pressure 1/ loop 1 llot shutdown **

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d. Manual 1/ loop -

tiot shutdown **

4. EMERGENCY C00LDOWN EQUIPMENT ROOM h ISOLATION s-

a. liigh temperature in ventilation 2 1 liot shutdown **

system ducts

• • - If minimum cor.ditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps shall be taken on the affected unit to p' lace the unit in cold shutdown conditions.

k TS.3.5-2 REV Safety Injection The Safety Injection System is actuated automatically to provide emergency cooling and reduction of reactivity in the event of a loss-of-coolant accident or a steam line break accident.

Safety injection in response to a loss-of-coolant accident (LOCA) is provided by a high containment pressure signal backed up by the low pressurizer pressure signal. These conditions would accompany the depressurization and coolant loss during a LOCA.

Safety injection in response to a steam line break is provided directly by a low steam line pressure signal, backed up by the low pressurizer pressure signal and, in case of a break within the containment, by the high containment pressure signal.

The safety injection of highly borated water will of fset the temperature-induced reactivity addition that could otherwise result from cooldown following a steam line break.

Containment Spray Containment sprays are also actuated by a high containment pressure signal (Hi-Hi) to reduce containment pressure in the event of a loss of coolant or steam line break accident inside the containment.

The containment sprays are actuated at a higher containment pressure (approximately 50% of design containment pressure) than is safety injection (10% of design). Since spurious actuation of containment spray is to be avoided, it is initiated on coincidence of high containment pressure sensed by three sets of one-out-of-two containment pressure signals provided for its actuation.

Containment Isolation A containment isolation signal is initiated by any signal causing automatic initiation of safety injection or may be initiated manually. The containment isolation system provides the means of isolating the various pipes passing through the containment walls as required to prevent the release of radioactivity to the environment in the event of a loss-of-coolant accident.

TS.3.5-3 REV Steam Line Isolation In the event of a steam line break, the steam line stop valve of the af fected line is autematically isolated to prevent continuous, uncontrolled steam release from more than one steam generator. The steam lines are isolated on high containment pressure (Hi-Hi) or high steam line flow in coincidence with low T and safety injectiot. or high steam flow (Hi-Hi) in coincidence with =afefy injection. Adequate protection is af forded for breaks inside or outside the containment even when it is assumed that th e steam line check valves do n)t function properly.

Containment Ventilatioc Isolation Valves in the containment p4rge and inservice purge systems automatically close on receipt of a Safecy Injection signal or a high radiation signal. Gaseous and particulate monitors in the exhaust stream or a gaseous monitor in the exhaust stack provide the high radiation signal. .

Ventilation System Isolation In the event of a high energy line rupture outside of conta{g9ent, redundant isolation dampers in certain ventilation ducts are closed.

Auxiliary Feedwater System Actuation The following signals automatically start the pumps and open the steam admission control valve to the turbine driven pump of the affected unit:

1. Low-low water level in either steam generator

2. Trip of both main feedwater pumps

3. Safety Injection signal

4. Undervoltage on both 4.16 KV normal buses (turbine driven pump only)

Manual control from both the control room and t$e Hot Shutdown Panel are also available. The design provides assurance that water can be supplied to the steam generators for decay heat removal when the normal feedwater system is not available.

TS.3.5-4 REV i

Limiting Instrument Setpoints

1. The high containment pressure limit is set at about 10% of the maximum internal against loss pressurg2) or steam line break accidents asInitiation of Safety Injection of coolant discussed in the safety analysis.

2. The Hi-Hi containment pressure limit is set at about 50% of the maximum internal pressure for initiation of containment spray and at about 30% for initiation of steam line isolation. Initiation of Containment (gyay and Steam Line Isolation pro [ggts against large loss of coolant or steam line break accidents as discussed in the safety analysis.

3. The pressurizer low pressure limit is set substantially below system operating pressure limits. Howeve r , it is sufficiently high to pro against a loss of coolant accident as shown in the safety analysis.[ ggt

4. The steam line low pressure signal is lead / lag compensated and its setpoint is set well above the pressure expected in the event of large steam line break accident as shown in the safety analysis.(g)

5. The high steam line flow limit is set at approximately 20% of nominal tull-load flow at the no-load pressure and the high-high steam line flow limit is set at approximately 120% of nominal full-load flow at the full load pressure in order to protect against large steam break accidents. The coincident low T setting limit for steam line isolation initiation is set below"Ets hot shutdown value. The safety analysis shows tha[33hese settings provide protection in the event of a large steam break

6. Steam generator low-low water level and 4.16 KV Bus 11 and 12 (21 and 22 in Unit 2) low bus voltage provide initiation signals for the l

Auxiliary Feedwater System. Selection of these setpoints is discussed in Section 2.3 of the Technical Specifications.

7. High radiation signals providing input to the Containment Ventilation Isg}3 pion circuity are set in accordance with the Radioactive Ef fluent Technical Specifications. The setpoints are established to prevent exceeding the limits of 10 CFR Part 20 at the site boundary.

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TS.3.5-5 REV I

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Instrume tt Operating Conditions i

During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is tot compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design. This specific.ation outlines limiting conditions for operation necessary to preserve the ef fectiveness of the Reaccor Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with sufficient redunda.tcy J to provide the capability for channel calibration and test at power.

I Exceptions are backup channels such as reactor coolant pump breakers. The removal of one trip channel 'on process control equipment is accomplished by placing that channel bistable in a tripped mode; e.g., a two-out-of-three 4 circuit becomes a one-out-of-two circuit. The source and internediate .

range nuclear instrumentation System channels are not intentionally placed ,

in a tripped mode since these are one-out-of-two trips, and the trips are therefore bypassed during testing. Testing does not trip the. system unless a tri,, condition exists in a concurrent channel.

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Re fe renc es t

(1) FSAR - Section 7.5 (2) FSAR - Section 14.3 i

(3) FSAR - Seciton 14.2.5 l (4) FSAR - Appendix I i

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_ ._ _ . ._ ,- , _ _._ ,. - . . _ - . .__ . - ~ . . - _ - - - - - _ _ _ _ _ , . . . - - . - .

TS.3.15-1 REV 3.15 EVENT MONITORING INSTRUMENTATION Applicability Applies to plant instrumentation which does not perform a protective function, but which provides information to monitor and assess important parameters during the following an accident.

Objectives To ensure that suf ficient information is available to operators to determine the ef fects of and determine the course of an accident to the extent required to carry out required manual actions.

Specification A. The event monitoring instrumentation channels specified in Table TS.3.15-1 shall be Operable.

B. With the number of Operable event monitoring instrumentation channels less than the Required Total Numbe r of Channels shown on Table TS.3.15-1, either restore the inoperable channels to Operable status within seven days, or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. With the number of Operable event monitoring instrumentation channels less than the Minimum Channels Operable requirements of Table TS.3.15-1, either restore the minimum number of channels to Operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Basis The operability of the event monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of NUREG-0578, "TMI-2 Lessons Learned l Task Force Status Report and Short Te rm Recomme nda t ion s . "

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TABLE TS.3.15-1

• EVENT MGNITORING INSTRUMENTATION Required Total No. Minimum Channels Instrument of Channel s Operable

l. Pressurizer Water Level 2 1

2. Auxiliary Feedwater Flow to Steam Generators 2/ steam gen 'l/ steam gen (One Channel Flow and One Channel Wide Range i Level for Each Steam Generator)

3. Reactor Coolant System Subcooling Margin *** 2 1

4. Pressurizer Power Operated Relief Valve Position 2/ valve 1/ valve (One Channel Temperature, One Channel Limit t Switch, and One Channel Acoustic Sensor per Valve *)

5. Pressurizer Power Operated Relief Block Valve 2/ valve 1/ valve Position (One Channel Temperature, One Channel Limit Switch, and One Channel Acoustic Sensor per Valve *)

6. Pressurizer Safety Valve Position 2/ valve 1/ valve (One Channel Temperature per Valve i and Common Acoustic Sensor **)

4

• - A common acoustic sensor provides backup position indication for each pressurizer power operated relief valve and its associated block valve. wg N en

• • - The acoustic sensor channel is common to both valves. When operable, the acoustic N sensor may be considered as an operable channel for each valve.

• • • - Fully qualified input instrumentation is being installed in accordance with the F NRC's TMI Action Plan. Until installation is completed, this function will be y satisfied using the plant process computer, r*

' TABLE TS.4.1-1 -

(Page 5 of 5)

Qiannel Functional Re s ponse Description dieck Calibrate Test Test , Remarks

35. Event Monitoring M R NA NA Includes all those in FSAR Instrumentation Table 7.7-2 and Table TS.3.15-1 not included elsewhere in this Table

, 36. Steam Exclusion W R M NA See FSAR Appendix 1, Actuation System Section 1.14.6

37. Pressurizer PORV NA R H NA Instruient Channels for POkV Control Control including Overpressure l Mitigation System 4

S -

Each Shift ,

D -

Daily W -

Weekly M -

Monthly gg

< tn Q

Quarterly N R -

Each refueling shutdown .d P -

Prior to each startup if not done previous week {

T -

Prior to each startup following shutdown in excess of 2 days if not done in the previous 30 days Q 5*

NA -

Not Applicable vi See Specification 4.1.D R O

1

IABLE TS.4.1-2A REV MINIMUM FREQUENCIES FOR EQUIPMENT TESTS FSAR Section Test Frequency Re fe rence

1. Control Rod Assemblies Rod drop times All rods during each 7 of full length refueling shutdown rods or following each removal of the reac-tor vessel head; .

affected rods following maintenance on or modification *.

the control rod drive system which could affect performance of those specific rods la. Reactor Trip Breakers Open trip Monthly -

2. Control Rod Assemblies Partial move- Eve ry 2 weeks 7 ment of all rods

3. Pressurizer Safety Set point Each refueling 4 Valves shutdown

4. Main Steam Safety Set point Each refueling 10 Valves shutdown

5. Pressurizer PORV Functional Quarterly -

Block Valves

6. Pressurizer PORV's Functional Every 18 months -

7. (Deleted)

8. (Deleted)

9. Primary System Leakage Evaluate Daily 4

10. (Deleted)

11. Turbine stop valves, Functional Monthly 10 governor valves, and intercept valves. (Part of turbine overspeed protection)

12. (Deleted)

NOTES:

• See Specification 4.1.D.

TS.4.6-1A REV B. Station Batteries

1. Each battery shall be tested each month. Tests shall include measuring voltage of each cell to the nearest hundredth volt, and measuring the temperature and density of a pilot cell in each battery.

2. The following additional measurements shall be made every three months: ';he density and height of electrolyte in every cell, the amounu of water added to each cell, and the temperature of each fif th cell.

3. All measurements shall be recorded and compared with previous data to detect signs of deterioration or need of equalization charge according to the manufacturer's recommendation.

4 The batteries shall be subjected to a performance test discharge during the first refueling and once every five years thereaf ter.

Battery voltage , hall be monitored as a function of time to establish that the battery performs as expected during heavy discharge and that all electrical connections are tight.

5. Integrity of Station Battery fuses shall be checked once each day when the battery charger is running.

C. Pressurizer Heater Emergency Power Supply The emergency pressurtzer heater supply shall be demonstrated operable at least once every 18 months by transferring Backup Heater Group "B" from its normal bus to its safeguards bus and energizing the heaters.

_ . . ._ = .. .- _ - . _ -. . _. .-

TS.4.6-3 REV The surveillance specified for the pressurizer heater power source provides assurance that Backup Heater Group "B" can be transferred to its emergency bus. Normally, this group of heaters is supplied from a normal plant 480 volt-bus. In an emergency, a manual transfer switch can be used to supply the heater group from a safeguards supply bus.

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TS.4.8-1 REV 4.8 SIZAM AND POWER CONVERSION SYSTEMS Applicability Applies to periodic testing requirements of the auxiliary feedwater, steam generator power operated relief valves, and steam exclusion systems.

Objective To verify the operability of the steam and power conversion systems required for emergency shutdown cooling of the plant.

Specification A. Auxiliary Feedwater System

1. Each motor-driven auxiliary feedwater pump shall be started at intervals of one month and full flow to the steam generators shall be demonstrated once every refueling shutdown.

2. The steam turbine-driven auxiliary feedwater pump shall be started at intervals of one month and full flow to the steam generators shall be. demonstrated once each year when steam from the steam generators is available.

3. The auxiliary feedwater pumps discharge valves shall be tested by operator action at intervals of one month.

4. Motor-operated valves required to function during accident l conditions shall be tested at intervals of one month.

5. These tests shall be considered satisfactory if control board l indication and subsequent visual observation of the equipment demon-strate that all components have operated properly.

5. During power operation, for the manual valves outside containment, that could reduce AFW flow, if improperly positioned, to less t than assumed in the accident analysis, monthly inspection are required to verify the valves are locked in the proper position required for emerency use.

l l 7. Af ter each cold shutdown and prior to exceeding 10% power, a test is required to verify the normal flow path from the primary AFW source to the steam generators. This test may consist of maintaining steam generator level during st r' g eith the auxiliary feed pumps.

i B. Power Operated Relief Valves i Each power operated sein steam relief valve shall be isola, ed and tested monthly.

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. . . _ , ,-. _ - - - - , , . . . , . ~ , ,

e TS.4.8-2 REV P

C. Steam Exclusion System Isolation dampers in each duct that penetrates rooms containing equipment required for a high energy line rupture outside of con-tainment shall be tested for operability once each month.

In addition, damper mating surf aces will be examined visually at each reactor refueling shutdown to assure that no physical change has occurred that could affect leakage.

Basis Monthly testing of the auxiliary feedwater pumps, monthly valve inspections, and startup flow verification provide assurance that the AFW system will meet emergency demand requirements. The discharge valves of the pumps are normally open, as are the suction valves from the condensate storage tanks. Proper opening of the steam admission valve on each turbine-driven pump will be demonstrated each time a turbine-driven pump is tested. Ventilation system l isolation dampers required to function for the postulated rupture of a high energy line will also be tested.

Re fe rence FSAR, Sections 6.6, 14, and Appendix I.

I TS.6.1-2 REV D. Each member of the plant staf f shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Superintendent Radiation Protection who shall meet or exceed the qualifications of Regulatory Guide 1.8, September, 1975, and (2) the Shif t Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents. The training program shall be under the direction of a designated member of Northern States Power management.

E. A training program for the fire brigade shall be maintained under the direction of a designated member of Northern States Power management . l This program shall meet the requirements of Section 27 of the NFPA Code - 1976 with the exception of training scheduling. Fire brigade training shall be scheduled as set forth in the plant training program.

- , ~ , - , - -

,. ,m.- , , -e - - m o ,-m, - ,-

TABLE TS.6.1-1 .

MINIMUM SilIFT CREW COMPOSITION (Note I and 3)

CATEGORY BOTil UNITS IN COLD Sl!UTDOWN ONE UNIT IN COLD SituTDOWN BOTil UNITS ABOVE COLD OR REFUELING SHUTDOWN OR REFUELING Sl!UTDOWN AND SilUTDOWN ONE UNIT ABOVE COLD SilUTDOWN No. Licensed Senior 2 (Note 2) 2 (Note 2) 2 5 Operators (LS0)

Total No. Licensed 4 4 5 Operators (LSO & LO)

Total No. Licensed & 6 7 8 Unlicensed Operators Shift Technical Advisor 0 1 1

(

NOTES:

1. Shift crew composition may be one less than the minimum requirements for a period of time not to exceed two hours in order to accomodate an unexpected absence of one duty shif t crew member provided immediate action is taken to restore the shif t crew composition to within the minimum requirements specified.

2. Does not include the licensed Senior Reactor Operator, or Senior Reactor Operator Limited to Fuel Handling, supervising refueling operatons.

3. Each LSO and to shall be licensed on each unit.

O TS.6.5-2 REV

1. a. Paragraph 20.203 " Caution signs, labels, signals and controls". In lieu of the " Control device" or alarm signal required by paragraph 20.203(c)(2), each high radiation area in which the intensity of radiation is 1000 mrem /hr or less s'..all be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (or continuous escort by a qualified person for the purpose of making a radiation survey) and any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr, except that locked doors shall be provided to prevent unauthorized entry into these areas and the keys to these locked doors shall he maintained under the administrative control of the Plant Manager.

2. A program shall be implemented to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

This program shall include the following:

a. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

b. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

A program acceptable to the Commission was described in letters from L 0 Mayer, NSP, to Director of Nuclear Reactor Regulation, dated December 31,1979 " Lessons Learned Implementation" and March 13, 1980, "1/1/80 Lessons Learned Implementation Additional Information".

3. A program shall be implemented which will ensure the capability to accurately i determine the airborne iodine concentration in essential plant areas under

! accident conditions. This program shall include the following:

l

a. Training of personnel, i

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b. Procedures for monitoring, and

d. Provisions for maintenance of sampling and analysis equipment.

A program acceptable to the Commission was described in letters from l L 0 Mayer, NSP, to Director of Nuclear Reactor Regulation, dated i

December 31, 1979 " Lessons Learned Implementation" and March 13, 1979, l

"1/1/80 Lessons Learned Implementation Additional Information".

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