ML20008F373

Amends 46 & 40 to Licenses DPR-42 & DPR-60,respectively, Incorporating New Requirements for Instruments & Equipment & Shift Manning,Resulting from NRC Assessment of TMI-2 Accident

ML20008F373
Person / Time
Site:	Prairie Island
Issue date:	03/02/1981
From:	Clark R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20008F374	List: ML20008F373 ML20008F375
References
TAC-12428, TAC-12429, NUDOCS 8103130146
Download: ML20008F373 (32)

Text

.

llNill D SI AT L S NtJCLE AR REGUt ATORY COMMISSION wawnraants. t> c w,

NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.46 License No. DPR-42 l.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Northern States Power Cogany (the licensee) dated December 19, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in cogliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

!810813 ONb

. 2.

Accordingly, the license is anended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-42 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.46 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION i

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R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date.of Issuance:

March 2, 1981 I

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t NORTilERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 40 License No. DPR-60 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Northern States Power Cor:pany (the licensee) dated December 19, 1980, conplies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in conpliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.-

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

9 e

. 2.

Accordingly, the license is amended by changes to the Technical Specificatiens as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment f40. 40, are hereby incorporated in the license.

The licensee shall operate the f acility in accordance with the Technical Specifications.

3. -This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/.J (.

ku, R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications

- Date of Issuance: March 2, 1981 t

1 ATTACHMENT TO LICENSE AMENDMENTS i

AMENDMENT NO. 46 TO FACILITY OPERATING LICENSE NO. DPR-42 AMENDMENT NO. 40 TO FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NOS. 50-282 AND 50-306 i'

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised pages are identified by Amendment Number and contain vertical lines indicating the area of change.

Page L

TS-i TS-ii TS-i i i TS.3.1-2 TS.3.1-3 TS.3.1-3A (new)

TS.3.4-1 TS.3.4-2 TS.3.4 TS.3.5-3 TS.3.5-4 TS.3.5-5 (new)-

TABLE TS.3.5-1 TABLE TS.3.5-3.(Page 1 of 2)

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

TABLE TS.3.5-4

. TS.3.15-1 (new)-

TABLE TS.3.15.1-(new)

TABLE TS.4.1-1 (Page 5 of 5)-

TABLE TS.4.1-2A TS.4.6-1A TS.4.6-3 (new)

TS.4.8-1 TS.4.8-2(new)

TS.6.1-2 TABLE TS.6.1-l' TS.6.5-2

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TS-i APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS TS SECTION TITLE PAGE 1.0 De f init ions 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, Prot ect ive TS.2.3-1 Ins t rument ation 3.0 Limiting Conditions for Operation TS.3.1-1 3.1 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 Cont ainment 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 Requirements TS.4.1-1 4.1 Ope rational. 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 Leaksge 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 T5.4.16-1 DPR-4: - Amendment No. /(, 43, 46 DPR Amendment No. 10, 31, 40

TS-li APPENDIX A TECllNICAI SPECIFICATIONS TAllLE OF CONTENTS (Continued) 1S SEC1

,,s TITLE PAGE 5.0 Design Features TS.S.1-1 5.1.

Site TS.S.1-1 5.2 containment System TS.5.2-1 5.3 Reactor TS.S.3-1 5.4 Engineered Safety Features TS.S.4-1 5.5 Radioactive Waste System TS.S.5-1 TS.S.6-1 5.6 Fuel llandling 6.0 Administrative Controls TS.6.1-1 TS.6.1-1 6.1 Organization 6.2 Review.and Audit TS.6.2-1

~6.3' Special Inspections and Audits TS.6.3-1 TS.6.4-l' 6.4 Safety-Limit Violation TS.6.5-1 6.5 Plant Operating Procedures TS.6.6-1 6.6 Plant Operating Records TS.6.7-1 6.7 Reporting Requirc*nents 6.8 Environmental Qualification TS.6.8-1 l

i I

.DPR Amendment No. 32, 46

'DPR-60 '-: Amendment No.' 26, 40-E

TS-iii 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 Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Ins t rument Operating Conditions for E=ergency Cooling System 3.5~4 Ins trument 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 Caseous 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 Requirements 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 Ef fluents Frc- 'rairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated a

. elease of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant l

(Per Unit) 6.1-1 Minimum Shif t Crew Composition 6.7-1 Special Esports l

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.DPR Atwndment No. 43, 46 l

DPR Aundment No. /Is 40 l

r TS.3.1-2 i

3.

P re:,su r i n e r a.

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

1.

Steam bu,bble 2.

Pressurizer heater gro.ps "A" and "11" 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 operabic 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 ilot Shutdown within the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

With the pressurizer inoperable for any other reason than c.

(b) above, the reactor shall be placed in at least liot Shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

f d.

At Icast one pressurizer safety valve shall be operabic whenever the head is on the reactor vessel, except during hydrostatic tests.

Both pressurizer safety valves shall be operabic whenever average reactor coolant system temperature is above 350 F or the reactor is critical.

Pressurizer safety valve lift setting shall be 2485 paig 1%.

Except as specified in (f) and (g) below, two power operated e.-

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 within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, With one or more block valves inoperable, within one hour either g.

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 within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

f DPR Amendment ~No.

46 DPR Amendment No. 40

TS.3.1-3 Basis I

khen 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 suf ficient 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 approxiuately one-half hour.

" Steam Generator Tube Surveillanc t", 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 powe7 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 dug co 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 release 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 ient 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 3

350 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 tne 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 pressurizacion of the reactor coolant system for reactor coolant temperatures less than 350 F.

The combined. capacity f rom complete loss of load.yster than the maiximum surge rate resulting of both safety valves ir gr D1'R Amendment No.46 DPR Amendment No. 40

TS.3.1-3A Basis (continued)

Th e requirement that two groups of pressurizer heaters be operable provides assurance that at least one group will be available during a loss of of f site 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.

Re fe rence s 1FSAR, Section 14.1.9 2Testimony by J Knight in the Prairie Island Public Hearing on January 28, 1975.

DPR Amendment No. 46 DPR Amendment No. 40

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TS.3.4-1 3.4 STEAM AND POWER CONVERSION SYSTEM Aeolicability 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 carability of removing decay heat from the reac:or, 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 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.

d.

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

DPR-42 Amendment No. 17, 46 DPR Amendment No. 11, 40

TS.3.4-2 e.

For Unit 1 operation motor ope-ated 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 position 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 Rad iochemis 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 ope rable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or one unit 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 />.

Basis A reactor shutdown from power requires removal of decay heat.

Decay heat 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.

e DPR-47 Amendment No. J /, 46 DPR-60 Amendment No. J J,40 -

TS.3.4-3 Th. ten main steam safety valves have a total combined rated capability of 7,745,000 lbs /hr.

Die total full power steam flow is 7,094,000 lbs /hr; th ere fo re, the ten main steam {9jety valves will be able to relieve the t ot al steam flow il necessary.

In the unlikely event of complete loss of of f site electrical power to either or both reactors, continued removal of decay heat would be assured by avail-ability of either the steam-drivs a 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 can be made available to the other reactor.

The 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.

The 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 rooms containing equip-ment needed for the accident.

The secondary coolant activity is based on a postulated release 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.

The 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:

X/Q Dose '(rem) =

C-V B(t)

DCF 10 Where: C = secondary coolant activity, 0.30 uCi/cc 3

3 V = water volume in one steam generator = 3510 f t = 99 M B(t) = breathing rate, 3.47 x 10 ' M /sec

~

3 X/Q = 9.8 x 10 ' sec/m

~

6 131 DCF = 1.50 x 10 rem /Ci 1 inhaled The resulting dose is 1.5 rem.

s Re fe renc e s (1)

FSAR, Section 10.4 (2)

FSAR, Appendix I (3)

FSAR, Section 14 ITI:-/./ Amendaent No. 46 lu b 00 Amendment No.

40-

TABLE TS.3.5-1 EE iM ENGINEERED SAFETY FEATURES INITI ATION INSTRUMENT LIMITING SET POINTS E S-FUNCTIONAL UNIT CHANNEL LIMITING SET POINTS

• O f4

['['l liigh Containment Pressure (Hi)

Safety inje'et ion *

<4 psig E i

((2

'lligh Cont ainment Pressure (Hi-Hi)

a. Containment Spray

<23 psig b.b

b. Steam Line Isolation

<l7 psig

.$5 E of Both Lines

.II %; 3 Pressurizer Low Pressure Safety Injection *

>l815 psig

$1104 Low Steam Line Pressure Safety Injection *

>$00 psig Lead Time Constant 712 seconds a= er Lag Time Constant 72 seconds i

O Os 5

High Steam Flow in a Steam Line Steam Line Isolation d/p correspgnding to Coincident with Safety Injection of Affected Line

<0.745 x 10 lb/hr and Low T

, 1005 psig at avg

>540 F N

6 liigh-high Steam Flow in a Steam Line Isolation ij/pcorrespgnding pg Steam Line Coincident with of Affected Line to 4.5 x to Ib/hr M

l Safety Injection at 735 psig D!

L l

7 liigh Pressure Dif ference Between Containment Vacuum

<0.5 psi i

Shield Building and Containment Breakers v

~

7 8

liigh Temperature in Ventilation Ducts Ventilation System

<!20 F Isolat ion Dampers t

9 liigh Radiation in Containment Exhaust Containment Ventilation icount rate corresponding to Air..

Isolation 500 mrem / year whole body and 3000 mrem / year skin due to noble gases at the site l

boundarv i

• Init iates also containment isolation, feedwater line isolat ion and s tart ing of all cont ainment fana.

d/p m. ans di f ferent ial pressure

TS.3.5-3 Steam Line Isolation In the event of a steam line break, the steam line stop valve of the affected line is automatically 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 injection or high steam flow (Hi-Hi) in coincidence with safeEy injection. Adequate protection is afforded for breaks ir. side or outs ide the co..tainment even when it is assumed that the steam line check valves do not function properly.

Containment Ventilation Isolation Valve s in the containment purge and inservice purge systems automatically close on receipt of a Safety 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 redundant isolation dampers in certain ventilation ducts are closed. {gyent, 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 the 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.

DPR-42.- Amendment No. 46 DPR-60'- Amendment No. 40 e

TS.3.5-4 Limiting Instrument Setpoints 1.

The high containment pressure limit is set at about 10% of the maximum interne. pressurg2) Initiation of Safety Injection protects against loss of coolant or steam line break accidents as 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 gggayandSteamLineIsolationprogg$tsagainst large loss Containment of coolant or steam line break accidents as discussed in the safety analysis.

3.

The pressurizer low pressure limit is set substantially below system it is suf ficiently high to pro as shown in the safety analysis.gggt operating pressure limits.

Howeve r, against a loss of coolant accident 4

The' steam line low pressure signal is lead / lag compensated and its is set well above the pressure expected in the event of as shown in the safety analysis.(g) setpoint large steam line break accident 5.

The high steam line flow limit is set at approximately 20% of nominal full-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 ordet to protect against large steam break accidents. The coincident low T setting limit for steam line isolation initiation is set below"Its hot shutdown value.

The safety of a

- analysis shows tha[3ghese settings provide protection in the event 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 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 Isolation circuity are set in' accordance with the Radioactive Effluent Technical Specifications. The setpoints are established to' prevent exceeding the' limits of 10 CFR Part 20 at the site boundary.

DPR-42 Amendment No. 36, 46

- DPR-60 Amendment No. 39, 40 z

f TS.3.5-5 Instrument Operating Conditions 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 not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design. This specification outlines limiting conditions for operation necessary to preserve the ef fectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with suf ficient redundancy to provide the capability for channel calibration and test at power.

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 circuit becomes a one-out-of-two circuit.

The source and intermediate 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 trip condition exists in a concurrent channel.

Re fe rence s (1) FSAR - Section 7.5 (2)

FSAR - Section 14.3 (3) FSAR - Seeiton 14.2.5 (4) FSAR - Appendix I DPR Amendment No. 46 DPit Amendment No. 40

55 7Y

o C ta TABLE TS.3.5-3 sF

@ s' INSTRUMENT OPERATING CONDITIONS FOR EMERGENCY COOLING SYSTEM

!E

=3 1

2 3

4 yy MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTIDN IF OPERABLE DBGREE OF BYPASS CONDITIONS OF COLUMN g }"

FUNCTIONAL UNIT CHANNELS REDUNDANCY CONDITIONS 1 or 2 CANNOT BE MET 1.

SAFETY INJECTION

e. t.

a.

Manual 2

1 Ilot shutdown **

o cn b.

High Containment Pressure 2

I Hot shutdown **

c.

Steam Generator Low Steam 2

I primary pressure Hot shutdown **

Pressure / Loop less than 2000 psig d.

Pressurizer Low Pressure 2

1 primary pressure llo t shutdown **

less than 2000 psig 2.

CONTAINMENT SPRAY a.

Manual 2

Ilot shutdown **

pg m

b.

Hi-Hi Containment Pres-ilor shutdown *

• g sore (Containment Spray)

Channel a 2

1 Channel b 2

1 Y

Channel c 2

1 Logic 2

1

[f 2

O r%

TABLE TS.3.5-3 (continued)

'E 5 77 INSTRUMENT OPERATING CONDITIONS FOR EMERGENCY COOLING SYSTEMS z-e,.

1 2

3 4

pp

q q MINIMUM MINIMUM PERMISSIBLE OPERAlCR ACTION IF

[ ~

OPERATING DEGREE OF BYPASS CONDITIONS OF COLT:MN

'@(

FUNCTIONAL UNIT CHANNELS REDUNDANCY CONDITIONS 1 OR 2 CANNOT hE MET

E E 22_3.

AUXILIARY FEEDWATER

? ?

a.

Steam Generator Low-Low 2

1 Hot shutdown

<3 C'

Water Level b.

Undervoltage on 4.16 KV 2/ bus 1/ bus llot shutdown Bu se s 11 and 12 (21 and 22 Unit 2)

(Start Turbine Driven Pump only) c.

Trip of Main Feedwater Pumps 2/ pump 1/ pump Ilo t shutdown d.

Safety Injection (See Ites No. 1)

Hot shutdown e.

Manual 2

1 Ilot shotdown N

is in Ul 9

• - Must actuate two switches simultaneously.

• • - If minimum conditions 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

g to place the unit in cold shutdown conditions, y

a FJ ev.

L'

TABLE TS.3.5-4 EE INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNr? IONS e~

l 2

3 4

MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF II OPE RABLE DEGREE OF BYPASS CONDITIONS OF COLUMN

_h h FUNCTIONAL UNIT CHANNELS REDUNDANCY CONDITIONS 1 OR 2 CANNOT BE MET is k {

l.

CONTAINMENT ISOLATION ff a.

Safety Injection (See Item No. I of Table TS.3.5-3) 110t shutdown **

b.

Manual 2

1 Ilo t shutdown 2.

CONTAINMENT VENTILATION ISOLATION a.

_ Safety Injection (See Item No. I of Table TS.3.5-3)

Maint ain l' urge 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 cet 3.

STEAM LINE ISOLATION a.

Hi-Ili Steam Flow with 2

1 llot shutdown **

Safety Injection b.

Hi Steam Flow and 2 of 4. Low 2

1 liot shutdown **

Tavg with Safety Injection i

U

{

c.

Hi Containment Pressure 1/ loop i

llot shutdown **

d.

Manual 1/ loop Hot shutdown **

E!

L 4.

EMERGENCY COOLDOWN EQUIPMENT ROOM L

ISOLATION l.

a.

High temperature in ventilation 2

1 llot shutdown **

system ducts

• • - If minimum conditions 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 place the unit in cold shutdown conditions.

TS.3.15-1 3.15 EVENT MONITORING IhTRUl!ENTATION 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.

Ob je c t ive s To ensure that suf ficient information is available to operators to de te rmine the effects 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 Nwmber 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 leas t 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 Task Force Status Report and Short Term Recommendations."

111'R Amendment No. 46 lin'H Amendment No. 40

TABLE TS.3.15-1 EVENT MONITORING INSTRUMENTATION

==

g2 Required Total No.

Minimum Channels s

Instrument of Channels Operable c re yg 1.

Pressurizer Water Level 2

1 e c

==

g@

2..Auxiliacy Feedwater Flow to Steam Generators 2/ steam gen 1/ steam gen 55 (One Channel Flow and One Channel Wide Range EN Level for Each Steam Generator) te z oc 3.

Reactor Coolant System Subcooling Margin ***

2 I

nu 4.

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

5.

Pressurizer Power Operated Relief Block Valve 2/ valve 1/ valve

, Position (One Common Channel Temperature, One Channel Limit Switch per Valve, and One Channel Acoustic Sensor per Valve *)

6.

Pressuriser Safety Valve Position 2/ valve 1/ valve (One Channel Temperature per Valve and Common Acoustic Sensor **)

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

g

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

P

• • • - Fully qualified input instrumentation is being installed in accordance with the HRC's TM1 Act ion Plan. Until installation is completed, this function will be y

satisfied using the plant process computer.

TABLE TS.4.1-1 (Page 5 of 5)

==

==77

[f]

Channel Functional

Response

Description Check Calibrate

__ Test Test Remarks Ik

j!.35. Event Honitoring M

R NA NA Includes all those in FSAR

_g(.

Instrumentation Table 7.7-2 and Table TS.3.15-1 E-not included elsewhere in this Table

,36.

Steam Exclusion W

R-M NA See FSAR Appendix 1,

!'2' Actuation System-Section 1.14.6 wu

?.5 37.

Pressurizer PORV NA R

M NA Instrument Ch annel s for PORV Control Control Including Overpressure r

ca cn Mitigation System i

S Each Shift D

Daily W

hkekly M

Monthly g

m Q

Quarterly N

R Each refueling shutdown d.

P Prior to each startup if not done previous week

{

f Prior to each startup following shutdown in excess of 2 days if not done in the previous 30 days j

m NA Not Applicable u

See Specification 4.1.D O

=

1 TABLE TS.4.1-2A MINIMUM FREQUENCIES FOR EQUIPMENT TESTS FSAR Section Test Frequency Reference 1.

Control Rod Assemblies Rod drop times All rods during each 7

of full length re fueling shutdown rods or following each removal of the reac-tor vessel head; affected rods following maintenance on or modification to the control rod drive system which could af fect performance of those specific rods la. Reactor Trip Breakers Open trip Monthly 2.

Control Rod Assemblies Partial move-Every 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 PORY Functional Quarterly Block Valsas l

6.

Pressurizer PORV's Functional Every 18 months 7.

(Deleted) 8.

(Deleted)

9.. Primary System Leakage Evaluate Daily 4

10. (Del. *d)

11. Turbine stop valves, Funct iona l Monthly 10 governor valves, and intercept valves.

(Part of turbine overspeed protection)

12. (Deleted)

NOTES:

I

• See Specification 4.1.D.

DPH Amendment No. //, 25, 26, 46 DPR Amendment No, f f,19, 70, 40

TS.4.6-1A B.

Stat ion 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: the density and height of electrolyte in every cell, the amount of water added to each cell, and the temperature of each fifth 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 manuf acturer'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 shall 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 :.harger is running.

C.

Pressurizer Heater Emergency Power Supply 4

The emergency pressurizer heater supply shall be demonstrated operable at least once every 18 months by transferring Backup Heater Group "B" f rom its normal bus to its safeguacds bus and energizing the hea*.ers.

DPR Amendment No. 25, 46 DPR Amendment No. If, 40

?

IS.A.6-3 l

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 voit bus.

In an emergency, a manual transfer switch can be used to supply the heater group from a safeguards supply bus.

i l

l l

1 l

l l

DPR Amendment No. 46 l

DPR Amendment No. 40 s

TS.6.8-1 4.8 STEAM 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 conditions shall be tested at intervals of one month.

5.

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

6.

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

7.

Af ter each cold shutdown and prior to exceeding 10 powe r, a test is required to verify the normal flow path from the primary AFW source to the steen generators.

This test may consist of maintaining steam generator level during startup with the auxiliary feed pumps.

8.

At least once every 18 months during shutdown verify that each pump starts as designed automatically and each automatic valve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.

B.

' Power Oporated Relief Valves Each power operated main steam relief valve shall be isolated and tested monthly.

DPH-41 Auenacent No. M DPR-hD Amendment No. @

I l

TS.4.8-2 I

r C.

Steam Exclusion System Isolation dampers in each duct that penetrates rooms containing eq uipeent required for a high energy line rup:ure 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 feedwatar 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 isolation dampers required to function for the postulated rupture of a high energy line will also be tested.

At 18-month intervals, pump starting and valve positioning is verified using test signals to simulate each of the automatic actuation parameters.

4 4

t.

f i

i Re ference FSAR, Sections 6.6,14, and Appendip 1.

Di'it Amendment No.' 46.

Dl'R Amendment No. 40-

. = _

4 TS.6.1-2 D.

Each member of the plant staff 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.

I E.

A training program for the fire brigade shall be maintained under the direction of a designated member of Northern States Power management.

l l

i 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.

i i

i i

i DPR Amendment No. 26, 46 DPR Amendment No. 29. 40

TABLE TS.6.1-1 E 3' MINIMUM SHIFT CREW COMPOSITION (Note 1 and 3)

Yf 80 CATECORY BOTH UNITS IN COLD SHUTDOWN ONE UNIT IN COLD SHUTDOWN BOTH UNITS ABOVE COLD k$

OR REFUELING SHUTDOWN OR REFUELING SHUTDOWN AND SHUTDOWN ONE UNIT ABOVE COLD SHUTDOWN ii 22

~

No. Licensed Senior 2 (Note 2) 2 (Note 2) 2

~~

ff Operators (LS0)

El S' Total No. Licensed 4

4 5

Operators (LSO & LO) u z.

C Cn Total No. Licensed &

6 7

8 Unlicensed Operators Shif t Technical Advisor 0

1 1

NOTES:

1.

Shif t 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 composit ion 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.

.IMch LSO and LO shall be licensed on each unit.

TS.6.5-2 Pa ragraph 20.203 "Caut ion s igns, labels, signals and controls".

In 1.

a.

lieu of the " Control device" or alarm signal required by paragraph

20. 203(c )( 2 ), each high radiat ion area in which the intensity of radiation is 1000 mrem /hr or less chall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation b'ork 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 be 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.

Ihis program shall include the following:

Provisions establishing preventive maintenance and periodic visual a.

inspection requirements, and b.

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

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 determine the airborne iodine concentration in essential plant areas under accident conditions. This program shall include the following:

a.

Training of personnel, l

b.

Procedures for monitoring, and d.

Provisions for mair *anance of sampling and analysis equipment.

(

I 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, 1979, "1/1/80 Lessons Learned Implementation Additional Information".

DPR Amendment No. 25, 46 DPR Amendment No. If, 40 i

_