Text

Amends 111 & 104 to Licenses DPR-42 & DPR-60,respectively, Revising TSs & Associated Bases for Surveillance Test Intervals & Allowed Outage Times for ESFs
	ML20072C076
Person / Time
Site:	Prairie Island
Issue date:	08/10/1994
From:	Marsh L; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20072C078	List: ML20072C076; ML20072C090;
References
	NUDOCS 9408170067
	Download: ML20072C076 (63)
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UNITEC STATES e

NUCLEAR REGULATORY COMMISSION

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t WASHINGTON. D.C. 20555 0001 NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT N0 d AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.111 License No. DPR-42 1.

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

A.

The application for amendment by Northern States Power Company (the licensee) dated September 21, 1992, as revised December 29, 1992, November 24, 1993, May 17, 1994, and June 21, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 Commission; 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 compliance with the Commission's regulations;.

D.

The issuance of this amendment will not be inimical to the common 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended 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:

4,,.

7 9400170067 940810 T'

PDR ADOCK 05000282!

P PDR

, Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.111, 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 issuance.

FOR T E CLEAR REGULATORY COMMISSION

+0F L. B. Marsh, Director Project Directorate 111-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 10, 1994 i

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ATTACHMENT TO LICENSE AMENDMENT N0.111 FACILITY OPERATING LICENSE N0. DPR-42 i

DOCKET NO. 50-282 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

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

REMOVE INSERT TS-xii TS-xii TS.1-1 TS.1-1 TS.1-2 TS.1-2 TS.1-3 TS.1-3 TS.1-4 TS.1-4 TS.1-5 TS.1-5 TS.1-7 TS.1-7 TS.1-8 TS.1-8 TABLE TS.1-1 TS.2.3-3 TS.2.3-3 TS.2.3-4 TS.2.3-4 TS.3.5-1 TS.3.5-1 TABLE TS.3.5-2 (Pages 1 & 2)

TABLE TS.3.5-2A (Pages 1 - 6)

TABLE TS.3.5-2B (Pages 1 - 9)

TABLE TS.3.5-3 (Pages 1 & 2)

TABLE TS.3.5-4 (Pages 1 & 2)

TABLE TS.3.5-5 TABLE TS.3.5-6 TS.3.10-1 TS.3.10-1 TS.3.10-2 TS.3.10-2 TS.4.1-1 TS.4.1-1 TABLE TS.4.1-1 (Pages 1 - 5)

TABLE TS.4.1-1A (Pages 1 - 5)

TABLE TS.4.1-1B (Pages 1 - 7)

TABLE TS.4.1-lC (Pages 1 - 4)

TABLE TS.4.1-2B (Pages 1 & 2)

B.2.3-2 B.2.3-2 B.2.3-3 B.2.3-3 B.3.5-1 B.3.5-1 B.3.5-2 B.3.5-2 B.3.5-3 B.3.5-3 B.3.5-4 B.3.5-4 B.3.5-5 B.3.5-5 B.3.10-1 B.3.10-1 i

B.3.10-2 B.3.10-2 B.4.1-1 B.4.1-1 B.4.1-2

TS-nii TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE IIILE l-1 Operational Modes 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2A Reactor Trip System Instrumentation 3.5-2B Engineered Safety Feature Actuation System Instrumentation 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Gaseous Effluent Monitoring instrumentation 3.14-1 Safety Related Fire Detection Instruments 3.15-1 Event Monitoring Instrumentation - Process & Containment 3.15-2 Event Monitoring Instrumentation - Radiation 4.1-1A Reactor Trip System Instrumentation Surveillance Requirements 4.1-1B Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements 4.1-lC Miscellaneous Instrumentation Surveillance Requirements 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements 4.10-1 Radiation Environmental Monitoring Program (PE:!P)

Sample Collection and Analysis 4.10-2 RFMP - Maximum Values for the Lower Limits of Detection 4.10-3 RFMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection 4.13-1 Snubber Visual Inspection Interval 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 4.17-3 Radioactive Liquid Waste Sampling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program j

5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Crew Composition Prairie Island Unit 1 Amendment No. $$,J07, 111 Prairie Island Unit 2 Amendment No. 9J J00, 104

i TS.1-1 1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are 1

applicable throughout these Technical Specifications.

ACTION ACTION shall be that part of a Specification which prescribes remedial i

measures required under designated conditions.

AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY shall exist when:

1.

Single doors in the Auxiliary Building Special Ventilation Zone are locked closed, and 2.

At least one door in each Auxiliary Building Special Ventilation Zone air lock type passage is closed, and 3.

The valves and actuation circuits that isolate the Auxiliary Building Normal Ventilation System following an accident are ODERABLE.

4 The Auxiliary Building Special Ventilation System is iRABLE.

CHANNEL CHECK i

CHANNEL CHECK is a qualitative determination of acceptable OPERABILITY by observation of channel behavior during operation. This determination shall include comparison of the channel with other independent channels measuring the same variable.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST consists of injecting a simulated signal into the channel as close to the primary sensor as practicable to verify that it is OPERABLE, including alarm and/or trip initiating action.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL RESPONSE TEST A CHANNEL RESPONSE TEST consists of injecting a simulated signal into j

the channel as near the sensor as practicable to measure the time for j

electronics and relay actions, including the output scram relay.

Prairie Island Unit 1 Amendment No. 73, 9J,111 Prairie Island Unit 2 Amendment No. 66, Bf,104

TS.1-2 CONTAINMENT INTEGRITY CONTAINMENT INTEGRITY shall exist when:

1.

Penetrations required to be isolated during accident conditions are either:

Capable of being closed by an OPERABLE containment automatic a.

isolation valve system, or b.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Specifications 3.6.C and 3.6.D.

2.

The equipment hatch is closed and sealed.-

3.

Each air lock is in compliance with the requirements of Specification 3.6.M.

4.

The containment leakage rates are within their required limits.

CORE ALTERATION CORE ALTERATION is the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel, which may affect core reactivity.

Suspension of CORE' ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle-1 in accordance with Specification 6.7.A.6.

Plant operation within these operating limits is addressed in individual specifications.

Prairie Island Unit 1 Amendment NO.

t07' 111 Prairie Island Unit 2 Amendment No.

100', 104 d

TS.1-3 DOSE EOUIVALENT I-131 DOSE EQUIVALENT I-131 is that concentration of I-131 (uci/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and 1-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites".

E-AVERAGE DISINTEGRATION ENERGY E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

FIRE SUPPRESSION WATER SYSTEM The FIRE SUPPRESSION WATER SYSTEM consists of: Water sources; pumps; and distribution piping with associated sectionalizing isolation valves.

Such valves include yard hydrant valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe, or spray system riser.

CASEOUS RADVASTE TREATMENT SYSTEM The CASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

Prairie Island Unit 1 Amendment No. 9J,'111 Prairie Island Unit 2 Amendment No. EA, 104

TS.1-4 LIMITING SAFETY SYSTEM SETTINGS LIMITING SAFETY SYSTEM SETTINGS are settings, as specified in Section 2.3, for automatic protective devices related to those variables having significant safety functions.

MEMBERS OF THE PUBLIC MSMBERS OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational occupational, or other purposes not associated with the plant.

0FFSITE DOSE CALCULATION MANUAL (ODCM1 The ODCM is the manual containing the methodology and parameters to be.used in the calculation of offsite doses due to radioactive liquid and gaseous effluents, in the calculation of liquid and gaseous effluent monitoring instrumentation alarm and/or trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program.

Prairie Island Unit 1 Amendment No. AS, SJ,111 Prairie Island Unit 2 Amendment flo. A3, pg,104 t,,,

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TS.1-5 OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:

(1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s),

subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this paragraph.

The OPERABILITY of a system or component shall be considered to be estab-lished when:

(1) it satisfies the Limiting Conditions for Operation in Specification 3.0, (2) it has been tested periodically in accordance with Specification 4.0 and has met its performance requirements, and (3) its condition is consistent with the two paragraphs above.

OPERATIONAL MODE - MODE An OPERATIONAL HODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table TS.I.1.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental characteristics of the core and related instrumentation.

PHYSICS TESTS are conducted such that the core power is sufficiently reduced to allow for the perturbation due to the test and therefore avoid exceeding power distribution limits in Specification 3.10.B.

Low power PHYSICS TESTS are run at reactor powers less than 2% of rated power.

Prairie Island Unit 1 Amendment No. 49, SJ,111 Prairie Island Unit 2 Amendment No. 43, $9,104

TS.1-7 PATED THERMAL POWER RATED THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant of 1650 megawatts thermal (MWt).

REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

SHIELD BUILDING INTEGRITY SHIELD BUILDING INTEGRITY shall exist when:

1.

Each door in each access opening is closed except when the access opening is being used for normal transit entry and exit, then at least one door shall be closed, and 2.

The shield building equipment opening is closed.

3.

The Shield Building Ventilation System is OPERABLE.

SRUTDOWN MARGIN SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which:

1) the reactor is suberitical or

2) the reactor would be suberitical from its present condition assuming all rod cluster control assemblies are fully inserted except for the rod cluster control assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

10LIDIFICATION SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel responce when the channel sensor is exposed to a source of increased radioactivity.

Prairie Island Unit 1 Amendment No. 59, $J, 111 Prairie Island Unit 2 Amendment No. EA, Eg, 104

i TS.1-8 j

i STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the specified Surveillance Frequency so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

For example, the surveillance frequency for the automatic trip and interlock logic specifies that the functional testing of that system is monthly and that each train shall be tested at least every two months on a STAGGERID TEST BASIS.

Per the definition above, for the automatic trip and interlock logic, the Surveillance Frequency interval is monthly and the number of trains (channels) is 2 (n-2).

Therefore, STAGGERED TEST BASIS roquires one train be tested each month such that after two Surveillance Frequency intervals (two months) both trains will have been tested.

STARTUP OPERATION The process of heating up a reactor above 200*F, making it critical, and bringing it up to POWER OPERATION, THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNRESTRICTED AREAS An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional and/or recreational purposes.

VENTILATION EXMAUST TREATMENT SYSTEM A VENTIIATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment.

Such a system is not considered to have any effect on noble gas effluents.

Engineered safety feature atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

Prairie Island Unit 1 Amendment No. 59, 92, 111 Prairie Island Unit 2 Amendment No. E3, 59, 104

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TABLE TS,1 1 TABLE TS.1-1 OPERATIONAL MODES REACTOR tRATED AVERAGE VESSEL HEAD REACTIVITY THERMAL COOLANT CIDSURE BOLTS HQ_DE TITLE CONDITION POWER TEMPERATURE FULLY TENSIONED Q

1 POWER OPERATION Critical

> 24 NA YES 2

HOT STANDBY **

Critical 5 24 NA YES 3

HOT SHUTDOWN **

Suberitical NA 2 350*F YES 4

INTERMEDIATE Suberitical NA

< 350*F YES SHUTDOWN **

2 200*F 5

COLD SHUTDOWN Suberitical NA

< 200*F YES 6

REFUELING NA*

NA NA NO

Boron concentration of the reactor coolant system and the refueling cavity sufficient to ensure that the more restrictive of the following conditions is met:

a. K rr 5 0.95, or e

b. Boron concentration 2 2000 ppm,

- Prairie Island specific MODE title, not consistent with Standard Technical Specification MODE titles. MODE numbers are consistent with Standard Technical Specification MODE numbers.

Prairie Island Unit 1 Amendment No. 111 Prairie Island Unit 2 Amendment No. 104

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TS.2.3-3 2.3.A.2.g.

Reactor coolant pump bus undervoltage - 275% of normal voltage.

h.

Open reactor coolant pump motor breaker.

Reactor coolant pump bus underfrequency - 258.2 Hz i

1.

Power range neutron flux rate.

1.

Positive rate - s15% of RATED THERMAL POWER with a time constant h2 seconds 2.

Negative rate 57% of RATED THERMAL POWER with a time constant k2 seconds 3.

Other reactor trips High pressurizer water level - 590% of narrow a.

range. instrument span.

b.

Low-low steam generator water level - 25% of narrow range instrument span.

c.

Turbine Generator trip 1.

Turbine stop valve indicators - closed 2.

Low auto stop oil pressure - 245 psig d.

Safety injection - See Specification 3.5 t

i i

Prairie Island Unit 1 Amendment No. S7, SJ, 92,111 Prairie Island Unit 2 Amendment No. $p, $A, $5,104

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TS.2.3 4 r

2.3.B. Protective instrumentation settings for reactor trip interlocks shall be as follows:

1. P-6 Interlock:

4 Source range high flux trip shall be unblocked whenever inter-mediate range neutron flux is s10-10 amperes.

2. P-7 Interlock:

"At power" reactor trips that are blocked at low power (low

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pressurizer pressure, high pressurizer level, and loss of flow for one or two loops) shall be unblocked whenever:

a.

Power range neutron flux is 212% of RATED THERHAL POWER or, b.

Turbine load is 210% of full load turbine impulse pressure.

3. P-8 Interlock:

Low power block of-single loop loss of flow is permitted whenever

-l power range neutron flux is s10% of RATED THERMAL POWER.

4. P 9 Interlock:

Reactor trip on turbine trip shall be unblocked whenever power range neutron flux is 250% of RATED THERMAL POWER.

5. P-10 Interlock:

Power range high flux low setpoint trip and intermediate range high flux trip shall be unblocked whenever power range ntatron flux is 59% of RATED THERMAL POWER.

C. Control Rod Withdrawal Stops

1. Block automatic rod withdrawal:

a.

P-2 Interlock:

.i Turbine load 515% of full load turbine attpulse pressure.

' Prairie Island Unit 1 Amendment No. 33, SJ, 111 Prairie Island Unit 2 Amendment No. 77, E9, 104

TS.3.5-1 3.5 INSTRUMENTATION SYSTEM Applicabil ity Applies to protection system instrumentation.

Obiectives To provide for automatic initiation of the engineered safety features in the event the principal process variable limits are exceeded, and to delineate the conditions of the reactor trip and engineered safety feature instrumentation necessary to ensure reactor safety.

Specification A.

Limiting set points for instrumentation which initiates operation of the engineered safety features shall be as stated in Table TS.3.5-1.

B.

For on-line testing or in the event of failure of a sub-system instrumentation channel, plant operation shall be permitted to continue at RATED THERMAL POWER in accordance with Tables TS.3.5-2A and TS.3.5-2B.

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Prairie Island Unit 1 Amendment No. A9, SJ, 111 Prairie Island Unit 2 Amendment No f), Ef, 104

7, o,7 TABLE TS.3.5-2A (Page 1 of 6) o TY REACTOR TRIP SYSTEM INSTRUMENTATION 88 g

MINIMUM gg TOTAL NO.

CHANNELS CHANNELS APPLICABLE gg FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

&W N

1.

Manual Reactor Trip 2

1 2

1, 2 1

N'*

2 1

2 3(*3, 4t*3, 5(*)

8 2.

Power Range, Neutron Flux a.

High Setpoint 4

2 3

1, 2 2

b.

Low Setpoint 4

2 3

ltb),2 2

+

3.

Power Range, Neutron Flux,.

4 2

3 1, 2 2

High Positive Rate 4.

Power Range, Neutron Flux, 4

2 3

1, 2 2

High Negative Rate 5.

Intermediate Range, Neutron Flux 1

2 1cb>,p 3

i f f 6.

Source Range, Neutron Flux i

a.

Startup 2

1 2

2(*3 4

gg b.

Shutdown 2

1 2

3(*3, 4(*),

5(*3 5

2z oo 7.

Overtemperature AT 4

2 3

1, 2 6

3%ye

8.

Overpower AT 4

2 3-1, 2 6

o g g EE

~ *

(a) When.the Reactor Trip Breakers are closed and the Control Rod Drive System is capable of rod eg

N withdrawal.

o.

.- w S

(b) Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

O (c) Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1 w

N, o,N o

QQ TABLE TS.3.5-2A (Page 2 of 6)

REACTOR TRIP SYSTEM INSTRUMENTATION

~~

-a Of O, ao MINIMUM EE TOTAL NO.

CllANNELS CilANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION m-9.

Low Pressurizer Pressure 4

2 3

1 6

10. liigh Pressurizer Pressure 3

2 2

1, 2 6

11. Pressurizer High Water Level 3

2 2

1 6

12. Reactor Coolant Flow Low 3/ loop 2/ loop 2/ loop 1

6

13. Turbine Trip a.

Low AST 011 Pressure 3

2 2

1 6

b.

Turbine Stop Valve Closure 2

2 1

1 6

EE

14. Lo-Lo Steam Generator 3/SG 2/SG in 2/SG in 1, 2 6

Water Level any SG each SG

15. Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one 1

11 11 and 12 (Unit 2: 21 and 22) both bus 55 buses IN w-n

$N as mg o.

P% f aso' e u.

4" w e

TABLE TS.3.5-2A (Page 3 of 6) mb

3. 2.

REACTOR TRIP SYSTEM INSTRUMENTATION EU "m ~

m 55 MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE gg g;

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

~~

16. Loss of Reactor Coolant Pump a.

RCP Breaker Open 1/ pump 1

1/ pump 1

1 b.

Underfrequency 4kV bus 2/ bus 1/ bus on 2 on one 1

11 both bus buses 2

1 2

1, 2 7

17. Safety Injection Input from ESF

18. Automatic Trip and Interlock Logic 2

1 2

1, 2 7

2 1

2 3( *), 4(*), S'*)

8 EE om

19. Reactor Trip Breakers 2

1 2

1, 2 9

2 1

2 3(*), 4'*),

5(*)

8 hh

20. Reactor Trip Bypass Breakers 2

1 1

(d) 10 55

e n n om (a) When the Reactor Trip Breakers are closed and the Control Rod Drive System is capable of rod gyg um gg R

withdrawal.

d (d) When the Reactor Trip Bypass Breakers are racked in and closed for bypassing a Reactor Trip Breaker gg and the Control Rod System is capable of rod withdrawal.

my

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,2 m TABLE 3.5-2A (Page 4 of 6) 55

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Action Statements eo

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ACTION 1: With the number of OPERABLE channels ACTION 3: With the number of channels OPERABLE one g[g[

one less than the Total Number of less than the Total Number of Channels and Channels, restore the inoperable channel with the THERMAL POWER level:

c SLgt to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be co ro in at least HOT SHUTDOWN within the next a.

Below the P-6 (Intermediate Range N*

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Neutron Flux Interlock) Setpoint, restore the inoperable channel to OPERABLE status prior to increasing ACTION 2: With the number of OPERABLE channels THERHAL POWER above the P-6 Setpoint.

less than the Total Number of Channels HOT STANDBY and/or POWER OPERATION may b.

Above the P-6 (Intermediate Range proceed provided the following Neutron Flux Interlock) Setpoint but conditions are satisfied:

below the P-10 (Power Range Neutron Flux Interlock) Setpoint, restore the a.

The inoperable channel is placed in inoperable channel to OPERABLE status the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; prior to increasing THERMAL POWER above the P-10 Setpoint.

b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed ACTION 4: With the number of OPERABLE channels one l[f[

for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance less than the Total Number of Channels gg testing of other channels per suspend all operations involving positive gg Specification 4.1; and reactivity changes.

55 c.

If THERMAL POWER is above 85% of pp RATED THERMAL POWER, then determine ACTION 5: With the number of OPERABLE channels the core quadrant power balance in one less than the Total Number of

,,s, to w=

accordance with the requirements of Channels, suspend all operations m -s s

,[,

Specification 3.10.C.4.

involving positive reactivity changes, Ejf f ts;w and restore the inoperable channel to em d.

One additional channel may be taken OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or eg ta de out of service for low power PHYSICS within the next hour open the reactor o-TESTS.

trip breakers.

OY 2"

W

77 TABLE 3.5-2A (Page 5 of 6)

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Action Statements eo sw

[7{s ACTION 6: With the number of OPERABLE channels ACTION 9:

a.

With one of the diverse trip features one less than the Total Number of (Undervoltage or Shunt Trip g(g[

Channels, HOT STANDBY and/or POWER Attachment) inoperable, restore it to mL 3, OPERATION may proceed provided the OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or following conditions are satisfied:

declare the breaker inoperable and r e+

apply the requirements of b below.

na wa I

a.

The inoperable channel is placed in The breaker shall not be bypassed the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , while one of the diverse trip features

. inoperable, except for the time and

s required for perfor7ing maintenance b.

The Minimum Channels OPERABLE and testing to restore the diverse requirement is met; however, the trip feature to OPERABLE status.

inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance b.

With one of the Reactor Trip Breakers testing of other channels per otherwise inoperable, be in at least Specifiestion 4.1.

HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one Reactor Trip Breaker may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for ACTION 7: With the number of OPERABLE channels one surveillance testing per Specification less than the Total Number of Channels, 4.1, provided the other Reactor Trip f(

restore the inoperable channel to Breaker is OPERABLE.

g[g OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in gg at least HOT SHUTDOWN within the next 6 gg hours; however, one channel may be ACTION 10: With the Reactor Trip Bypass Breaker bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for inoperable, restore the Reactor Trip pp surveillence testing per Specification Bypass Breaker to OPERABLE status 4.1 provided the other channel is prior to using the Reactor Trip y, s, re ta OPERABLE.

Bypass Breakar to bypass a Reactor m-s

(,*,

Trip Breaker.

If the Reactor Trip Ejf f ts wa ACTION 8: With the number of OPERABLE channels one Bypass Breaker is racked in and am less than the Total Number of Channels closed for bypassing a Reactor Trip vig wa re restore the inoperable channel to Breaker and it becomes inoperable, be o-

~

OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open in at least HOT SHUTDOWN within 6

$3d the reactor trip breakers within the hours. Restore the Bypass Breaker to ES[

next hour.

OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Bypass Breaker within the following hour.

~

77 T!LBLE 3.5-2A (Page 6 of 6) 11 Action Statements to m EE gg ACTION 11: With the number of OPERABLE channels ACTION 19: NOT USED gg less than the Total Number of Channels, POWER OPERATION may proceed provided 31 the following conditions are satisfied:

.9 co N-a.

The inoperable channel (s) is placed i

in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and b.

The Minimum Channels OPERABLE requirement is met; however,-the inoperable channel (s) may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.1.

i ACTION 12: NOT USED i

ACTION 13: NOT USED aa aa m CD gg ACTION 14: NOT USED EE ACTION 15: NOT USED

_e NEh N

ACTION'16: NOT USED

$5

- ~

.a o

ACTION 17: NOT USED "h>

ACTION 18: NOT USED

r 1

I, I.

i onQQ TABLE TS.3.5-28 (Page 1 of 9)

.m o t

O.jNEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

~~

a MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE EE FUNCTIONAL UNIT OF CHANNELS

-TO TRIP OPERABLE MODES ACTION 3;

1.

SAFETY INJECTION ms a.

Manual Initiation 2

1 2

1,2,3,4 23 b.

High Containment Pressure 3

2 2

1,2,3,4 24 f

c.

Steam Line Low Pressure 3/ Loop 2 in any 2/ Loop 1, 2, 3(*)

24 Loop i

I d.

Pressurizer Low Pressure 3

2 2

1, 2, 3(*)

24 i

e.

Automatic Actuation Logic 2

1 2

1,2,3,4 20 and Actuation Relays i

2.

CONTAINMENT SPRAY NN a.

Manual Initiation 2

2 2

1,2,3,4-23 b.

Hi-Hi Containment Pressure 3 channels 1 sensor 1 sensor 1,2,3,4 21 with 2 per per EE sensors per-channel channel channel-in all'3 in all 3 UM channels channels

,g m *e g '

c.

Automatic Actuation Logic and 2'

1 2

1, 2,'3, 4

20

$ e MI,

-Actuation Relays

~s w

o.,."

e w s

,a e v' :

oa

' ~h

, (a) Trip ' function may be blocked in this MODE below a Reactor Coolant System Pressure. of 2000 psig.

?

??

11 TABLE TS.3.5-2B (Page 2 of 9) to to ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION y

==

MINIMUM gg TOTAL NO.

CHANNELS CHANNELS APPLICABLE 3, s FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION ee ro *

3.

CONTAINMENT ISOLATION a.

Safety Injection See Functional Unit I above for all Safety injection initiating functions and requirements.

b.

Manual 2

1 2

1,2,3,4 23 c.

Automatic Actuation Logic and 2

1 2

1,2,3,4 20 Actuation Relays 4.

CONTAINMENT VENTIIATION IS01ATION a.

Safety Injection See Functional Unit I above for all Safety injection initiating functions and requiremmts.

b.

Manual 2

1 2

(b) 22 mm c.

Manual Containment Spray See Functional Unit 2a above for Manual Containment Spray requirements.

3g S El to to gS d.

Manual Containment Isolation See Functional Unit 3b above for Manual Containment Isolation requirements.

ff High Radiation in Exhaust Air 2

1 2

(b) 22 e.

ww W

f.

Automatic Actuation Logic 2

1 2

(b) 22 and Actuation Relays jf

~ '

o e-(D (T3 N

(b) Whenever CONTAINMENT INTEGRITY is required and either of the containment purge systems are in S [w

~

gr operation.

3y 3

??

TABLE TS.3.5-2B (Page 3-of 9)

$. S.

3. 3.

ENGINEERED SAFETY FFATURE ACTUATION SYSTEM INSTRUMENTATION mm

^

W MINIMUM pg TOTAL NO.

CHANNELS CHANNELS APPLIC/ ALE EE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION CC t

?

3. 3.

5.

STEAM LINE iso 1ATION ee a.

Manual 1/ Loop 1/ Loop 1/ Loop 1, 2, 3(*)

27 b.

Hi-Hi Containment Pressure 3

2 2

1, 2, 3(*3 24 c.

Hi-Hi Steam Flow with Safety Injection 1.

Hi-Hi Steam Flow 2/ Loop 1 in any 1/ Loop 1, 2, 3(*)

29 l

Loop i

2.

Safety Injection See Functional Unit I above for all Safety Injection initiating functions and requirements.

l d.

Hi Steam Flow and 2 of 4 Lo-Lo T,, with Safety Inj ection:

1.

Hi Steam Flow 2/ Loop 1 in any 1/ Loop 1, 2, 3(d8 29 RE Loop aa em SE 2.

Lo-Lo T.,,

4 2

3 1, 2, 3(d) 24

z :z

??

3.

Safety Injection See Functional Unit I above for all Safety injection initiating functions and requirements.

win wne EIN Sb tw w (c) When either main steam isolation valve is open.

uH i

?*.*

(d) When reactor coolant system average temperature is greater than 520*F and either main steam isolation g-valve is open.

ma o-4" o,

...- m m

YI El TABLE TS.3.5-2B (Page 4 of 9) 11 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION EE CC SE MINIMUM TOTAL No.

CHANNELS CHANNELS APPLICABLE cc 11 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION co co 5.

STEAM LINE ISOLATION (continued) e.

Automatic Actuation Logic and 2

1 2

1, 2, 3(*)

25 Actuat sn Relays 6.

FEEDUATER IS01ATION a.

Hi-Hi Steam Generator Level 3/SG 2/SG in 2/SG in 1, 2 24 any SG each SG b.

Safety Injection See Functional Unit I above for all Safety injection initiating functions and requirements.

c.

Reactor Trip with 2 of 4 gg Low T.,,

(Main Valves only):

mmh 1.

Reactor Trip 2

1 2

1, 2 28

<+ co 2.

Low T,y 4

2 3

1, 2 24 55 d.

Automatic Actuation Logic 2

1 2

1, 2 28 gg and Actuation Relays g

EED RE

%G

>a o.

w (c) When either main steam isolation valve is open.

m,w m?

SC W

oo 22 TABLE TS.3.5-2B (Page 5 of 9)

~4 "4 7$

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM

"]$

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION EE AA 7.

AUXILIARY FEEDWATER ro.a a.

Manual 2

1 2

1,2,3 26 b.

Steam Generator Lo-Lo 3/SG 2/SG in 2/SG in 1,2,3 24 Water Level any SG each SG Undervoltage on 4.16 kV Buses 2/ bus 1/ bus on 2 on one 1, 2 29 c.

11 and 12 (Unit 2: 21 and 22) both bus (Start Turbine Driven Pump buses only) d.

Trip of Both Main Feedwater Pumps EE

1. Turbine Driven 2

2 2

1, 2 26

$S

2. Motor Driven 2

2 2

1, 2 26 e.

Safety Injection See Functional Unit I above for all Safety injection initiating functions and requirements.

55 f.

Automatic Actuation Logic 2

1 2

1,2,3 30 g

and Actuation Relays gyg 77

$N y

,, L, N-

~

II 1 3.

TABLE TS.3.5-2B (Page 6 of 9)

2. 2.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTREMENTATION 77 77 MINIMtM EE TOTAL NO.

CHANNELS CHANNELS APPLICABLE cc FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

3. "

8.

LOSS OF POWER m-

a. Degraded Voltage 4/ Bus 2/ Bus 3/ Bus 1,2,3,4 31, 32, 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases)

b. Undervoltage 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, 4 31, 32, 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases)

ET a*

aa sn

.O O

23

%M og RE AL

.6 t, 2 "h

n-w

TABLE 3.5-2B (Page 7 of 9)

,2,3 55 j[][

Action Statements mo cw w eke 5 ACTION 20: With the number of OPERABLE channels ACTION 23: With the number of OPERABLE channels one less than the Total Number of one less than the Total Number of Channels, restore the inoperable Channels, restore the inoperable EE channel to OPERABLE status within 6 channel to OPERABLE status within 48

?$$

hours or be in at least HOT SHUTDOWN hours or be in at least HOT SHUTDOWN N*

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

however, one channel may be bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance testing per Specification 4.1, provided ACTION 24: With the number of OPERABLE channels the other channel is OPERABLE.

one less than the Total Number of Channels, operation in the applicable MODE may proceed provided the following ACTION 21: With the number of OPERABLE channels conditions are satisfied:

less than the Total Number of Channels, The inoperable channel is placed in operation may proceed provided the a.

inoperable channel (s) is placed in the the tripped condition within 6 tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and hours, and, the Minimum Channels OPERABLE requirement is met.

One inoperable b.

The Minimum Channels OPERABLE channel may be bypassed at a time for requirement is met; however, the up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing inoperable channel may be bypassed gpg per Specification 4.1.

for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per oo

[pfp Specification 4.1.

mm ACTION 22: With the number of OPERABLE channels S5 less than the Total Number of Channels, g;;s gg operation may continue provided the containment purge supply and exhaust

<:g h em E;d valves are maintained closed.

,, 9 u~

o Hs 63 0

TABLE 3.5-2B (Page 8 of 9)

Il Action Statements 1 3.

mo

7;7 ACTION 25: With the number of OPERABLE channels ACTION 28: With the number of OPERABLE channels

7;7 one less than the Total Number of one less than the Total Number of g[g[

Channels, restore the inoperable Channels, restore the inoperable channel to OPERABLE status within 6 channel to OPERABLE status within 6 c: c:

ELS.

hours or be in at least HOT SHUTDOWN hours or be in at least HOT SHUTDOWN r* r*

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Operation in within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . However, one HOT SHUTDOWN may proceed provided the channel may be bypassed for up to 8 N "'

main steam isolation valves are closed, hours for surveillance testing per if not, be in at least INTERMEDIATE Specification 4.1, provided the other SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , channel is OPERABLE.

However, one channel may be bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance testing per Specification 4.1, provided ACTION 29: With the number of OPERABLE channels the other channel is OPERABLE.

less than the Total Number of Channels, operation in the applicable MODE may proceed provided the following ACTION 26: With the number of OPERABLE channels conditions are satisfied:

one less than the Total Number of Channels, declare the associated a.

The inoperable channel (s) is placed auxiliary feedwater pump inoperable and in the tripped condition within 6 take the action required by hours, and, Specification 3.4.2.

b.

The Minimum Channels OPERABLE jf!I requirement is met; however, one ACTION 27: With the number of OPERABLE channels inoperable channel may be bypassed one less than the Total Number of at a time for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for Channels, restore the inoperable surveillance testing of other channel to OPERABLE status within 48 channels per Specification 4.1 m-H O2$

85 hours9.837963e-4 days 0.0236 hours 1.405423e-4 weeks 3.23425e-5 months or be in at least HOT SHUTDOWN 3$

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and close the

$;:0 associated valve, md i o

Sm -n mf 3Y 0

l

TABLE 3.5-2B (Page 9 of 9) oo jfjf Action Statements 11 mo

7;7 ACTION 30: With the number of OPERABLE channels ACTION 33: If the requirements of ACTIONS 30 or 31 gy g7 one less than the Total Number of cannot be met within the time g[j[

Channels, declare the associated specified, or with the number of auxiliary feedwater pump inoperable and OPERABLE channels three less than the c: c:

3, 3, take the action required by Total Number of Channels, declare the Specification 3.4.2.

However, one associated diesel generator (s) channel may be bypassed for up to 8 inoperable and take the ACTION required ha ra hours for surveillance testing per by Specification 3.7.B.

Specification 4.1, provided the other channel is OPERABLE.

ACTION 31: With the number of OPERABLE channels one less than the Total Number of Channels, operation in the applicable MODE may proceed provided the inoperable channel is placed in the bypassed condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 32: With the number of OPERABLE channels two less than the Total Number of Channels, operation in the applicable MODE may proceed provided the following conditions are satisfied:

Ed One inoperable channel is placed in a.

the bypassed condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and, wna

@7$

@f F b.

The other inoperable channel is

%M placed in the tripped condition y}

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and, eg Ei.a All of the channels associated with

Y, c.

the redundant 4kV Safeguards Bus Sl l are OPERABLE.

\\

TS.3.10-1 l

i 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to the limits on core fission power distribution and to the limits on control rod operations.

Obiective To assure 1) core suberiticality after reactor trip, 2) acceptable core power distributions during POWER OPERATION, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.

Specification A.

Shutdown Martin

1. Reactor Coolant System Averare Temperature > 200*F The SHUTDOWN MARGIN shall be greater than or equal to the applicable value shown in Figure TE.3.10 1 when in HOT SHUTDOWN and INTERMEDIATE SHUTDOWN.

2. Reactor Coolant System Averare Temperature s 200'F The SHUTDOWN MARGIN shall be greater than or equal to 1%Ak/k when in COLD SHUTDOWN.

3. With the SHUTDOWN MARGIN less than the applicable limit specified in 3.10.A.1 or 3.10.A.2 above, within 15 minutes initiate boration to restore SHUTDOWN MARGIN to within the applicable limit.

B.

Power Distribution Limits 1.

At all times, except during low power PHYSICS TESTING, measured hot channel factors, F*n and F%3, as defined below and in the bases, shall meet the following limits:

RTP F*n x 1.03 x 1.05 s (Fn / P) x K(Z)

RTP F%f x 1.04 s Fa x (1+ PFDH(1-P))

where the following definitions apply:

RTP

- Fn is the Fo limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.

RTP

- Fa is the Fa limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.

- PFDH ie the Power Factor Multiplier for F$a specified in the CORE OPERATING LIMITS REPORT.

- K(Z) is a normalized function that limits Fg(z) axially as specified in the CORE OPERATING LIMITS REPORT.

Prairie Island Unit 1 Amendment No. B4, 91, 92,111 Prairie Island Unit 2 Amendment No. 77, $f, B5,104

TS.3.10 2 3.10.B.1. - Z is the core height location.

P is the fraction of RATED THERMAL POWER at which the core is N

operating.

In the F g limit determination when P 50.50, set P - 0.50.

N F*g or F g is defined as the measured Fg or Fan respectively, with the smallest margin or greatest excess of limit.

E 1.03 is the engineering hot channel factor, F n, applied to the 8

measured F g to account for manufacturing tolerance.

1.05 is applied to the measured F*n to account for measurement uncertainty.

- 1.04 is applied to the measured F%a to account for measurement uncertainty.

N

2. Hot channel factors, Fg and F shall be measured and the target flux difference determined, at equilibrium conditions according to the following conditions, whichever occurs first:

I (a) At least once per 31 effective full-power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions after exceeding the reactor power at which target flux difference was last determined, by 10% or more of RATED THERMAL POWER.

F*n (equil) shall meet the following limit for the middle axial 80%

of the core:

RTP 8Fg (equil) a V(Z) x 1.03 x 1.05 s (Fg / P) x K(Z) where V(Z) is specified in the CORE OPERATING LIMITS REPORT and other terms are defined in 3.10.B.1 above.

3. (a) If either measured hot channel factor exceeds its limit specified in 3.10.B.1, reduce reactor power and the high neutron flux trip set-point by 1% for each percent that the measured F*n or by the factor specified in the CORE OPERATING LIMITS REPORT for each percent that the measured NFg exceeds the 3.10.B.1 limit. Then follow 3.10.B.3(c).

(b) If the measured F*n (equil) exceeds the 3.10.B.2 limits but not the 3.10.B.1 limit, take one of the following actions:

1.

Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months place the reactor in an equilibrium configuration for which Specification 3.10.B.2 is satisfied, or 2.

Reduce reactor power and the high neutron flux trip setpoint by 1% for each percent that the measured F*n (equil) x 1.03 x 1.05 x V(2) exceeds the limit.

Prairie Island Unit 1 Amendment No. BA, SJ, 92,111 Prairie Island Unit 2 Amendment No. //, $$, EE,104

TS.4.1-1 4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Obiective To specify the minimum frequency and type of surveillance to be applied to plant equipment and conditions.

Specification A.

Calibration, testing, and checking of instrumentation channels and testing of logic channels shall be performed as specified in Tables TS.4.1-1A, 4.1 1B and 4.1-1C.

B.

Equipment tests shall be conducted as specified in Table TS.4.1-2A.

C.

Sampling tests shall be conducted as specified in Table TS.4.1-2B.

D.

Whenever the plant condition is such that a system or component is not required to be OPERABLE the surveillance testing associated with that system or component may be discontinued.

Discontinued surveillance tests shall be resumed less than one test interval before establishing plant conditions requiring OPERABILITY of the associated system or component, unless such testing is not practicable (i.e., nuclear power range calibration cannot be done prior to reaching POWER OPERATION) in which case the testing will be resumed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of attaining the plant condition which permits testing to be accomplished.

Prairie Island Unit 1 Amendment No. J/, $J, JpJ, 111 Prairie Island Unit 2 Amendment No. JJ, EA, $A, 104

(

~

??

TABLE TS.4;1-1A (Page 1 of 5) l 11 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS ea

. y;;;*

. RTNCTIONAL

RESPONSE

MODES FOR WHICH g;;;'

FUNCTIONAL UNIT CHECK CALIBRATE TEST

-TEST SURVEILIANCE IS REQUIRED ss aa 0) 5")

0),4 1.

Manual Reactor Trip N.A.

N.A.

RU3)

N.A.

1, 2, 3 c_ c_

ao 7

2.

Power Range, Neutron Flux a) High Setpoint S

DO 7)

Qos R

1, 2 g(6, 7) q(7. e)

R(7) i b) Low Setpoint S

R(7)

S/U"7)

R

10) 2 i

3.

Power Range, Neutron Flux, N.A.

R(7)

Q R

1, 2 High Positive Rate 4

I 4.

Power Range,-Neutron Flux, N.A.

R(7)

Q R

1, 2 g

High Negative Rate m.

mu Q. O.

5.

Intermediate Range, S

R(7)

S/U(')

R 1(3) 2 l

55 Neutron Flux

=z 6.

Source Range, Neutren Flux wm a.

Startup S

R(7)

S/U(')

R 2(2)

mg b.

Shutdown S

R(7)

Quo)

R 30) u, 50)

Q7@

5N mw HH 7.

Overtemperature AT S

R-Q R

1, 2 o."

h..

m,z~.

. v ".

m 1

8.

Overpower AT S

R Q

R 1, 2 5;'-

1 I

??

TABLE 4.1-1A (Page 2 of 5)

==

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS ru to EE FUNCTIONAL

RESPONSE

MODES FOR WHICH UI FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED Ra 9.

Low Pressurizer Pressure S

R Q

N.A.

1 gg

+

((

10. High Pressurizer Pressure S

R Q

N.A.

1, 2

11. Pressurizer High Water Level S

R Q

N.A.

1

12. Reactor Coolant Flow Low S

R Q

N.A.

1

13. Turbine Trip S/U* 11)

N.A.

1 t

a.

Low AST Oil Pressure N.A.

R 113 N.A.

1 b.

Turbine Stop Valve N.A.

R S/U('-

Closure 14 Lo-Lo Steam Generator S

R Q

N.A.

1, 2 Water Level

15. Undervoltage 4KV RCP Bus N.A.

R Q

N.A.

I aa 22 P. P.

.E.E O

bb G7B PP TE

=

un k&

o O~

+

I 7

TABLE TS.4.1-1A (Page 3 of 5) a m.

11 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS mo j

EE FUNCTIONAL

RESPONSE

MODES FOR WHICH TU RINCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED oo aa 16.-Loss of Reactor Coolant Pump cc

" 3.

a.

RCP Breaker Open N.A.

R S/U(')

N.A.

1 I

ro --

b.

Underfrequency 4KV Bus N.A.

R Q

N.A.

1 l

t

17. Safety Injection Input N.A.

N.A.

R N.A.

1, 2

18. Automatic Trip and Interlock N.A.

N.A.

M(')

R 1, 2, 3(1), 4(1) 5(1)

Logic

[

t M s.12 R

1, 2, 3(13, 4(13, 5(1)

I t

19. Reactor Trip Breakers N.A.

N.A.

t l

20. Reactor Trip Bypass Breakers N.A.

N.A.

Mil'8 R(153 See Note (16) an me dM N

MQ".

bb

<mb 1

um 4~

%G w

.m.

we b$

o*

a m c~

v 5'

1 t

w

ooy}[

TABLE 4.1-1A (Page 4 of 5) s, 57 27 TABLE NOTATIONS

-US FREQUENCY NOTATION oa ss NOTATION FREOUENCY EE S

Shift no -

D Daily M

Monthly Q

Quarterly S/U Prior to each reactor startup R

Each Refueling Shutdown N.A.

Not applicable.

TABLE NOTATION (1)

When the Reactor Trip Breakers are (6)

Single point comparison of incore to excore closed and the Control Rod Drive System is for axial off-set above 15% of RATED THERHAL capable of rod withdrawal.

POWER. Recalibrate if the absolute difference is greater than 2%.

(2)

Below P-6 (Intermediate Range Neutron Flux f[f Interlock) Setpoint.

(7)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

ss ff (3)

Below P-10 (Low Setpoint Power Range Neutron ss Flux Interlock) Setpoint.

(8)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

fifi (4)

Prior to each startup following shutdown in excess of two days if not done in previous 30 (9)

Each train shall be tested at least every gg days.

two months on a STAGGERED TEST BASIS.

E5 (5)

Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER.

eg o.

Adjust excore channel gains consistent with calorimetric power if absolute difference is

[-[

greater than 2%.

s 4,>

+L-4 I?

ELIL TABLE 4.1-1A (Page 5 of 5) 11 TABLE NOTATIONS Continued)

XX ET EI TABLE NOTATION (Continued) wo C CL c: c:

(10) Quarterly surveillance in MODES 3, 4 and 5 (17)

Prior to each startup if not done previous EL EL shall also include verification that week.

permissives P-6 and P-10 are in their required state for existing plant conditions (18)

Including quadrant power tilt monitor.

by observation of the permissive annunciator window.

(19) Not Used (11)

Setpoint verification is not applicable.

(12) The Functional Test shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor

' Trip Breakers.

(13) The Functional Test shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual g2p Reactor Trip Punction. The test shall also

'me verify the OPERABILITY of the Bypass Breaker-gg trip circuit (s).

mm 5L5!

(14) Manually trin the undervoltage trip attachment remotely (i.e.,

from the gy gg protection system racks).

8 '.

(15) Automatic undervoltage trip.

g;;g

<=w.

(16) Whenever the Reactor Trip Bypass Breakers are 5E

. racked in and closed for bypassing a Reactor vi -i Trip Breaker and the Control Rod' Drive System o."'

is capable of rod withdrawal.

Ut M8 l

v s I

J

.A I

TABLE TS.4.1-1B (Page 1 of 7) mo 22 TT ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS

-CX FUNCTIONAL P'.PONSE MODES FOR VHICH FUNCTIONAL UNIT CHECK

" LIBRATE TEST TEST SURVEILLANCE IS REQUIRED cu c.

oo 1.

SAFETY INJECTION 55 2%

a.

Manual Initiation N.A.

N.A.

R(20)

N.A.

1,2,3,4 eo >--

b.

High Containment Prese. ire S

R Q

N.A.

1,2,3,4 c.

Steam Line Low Pressure S

R Q

N.A.

1, 2, 3(21) d.

Pressurizer Low Pressure 3

R Q

N.A.

I 1, 2, 3t21) e.

Automatic Actuation Logic N.A N.A.

Mt22)

N.A.

1,2,3,4 and Actuation Relays 2.

CONTAINMENT STRAY a.

Manual Initiation N.A.

N.A.

R N.A.

1, 2, 3, 4 b

Hi-Hi Containment S

R Q

N.A.

1,2,3,4 F $"

Pressure c.

Automatic Actuation Logic N.A.

N.A.

Mt22)

N.A.

1,2,3,4 and Actuation Relays er O

..ga d a. r m

Hd V2 O-*

-~

w

1

??

TABLE TS.4.1-IB (Page 2 of 7) 11.

11'

. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS em 77 FUNCTIONAL

RESPONSE

MODES FOR WHICH 77 F11NCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED

.EE cc 3.

CONTAINMENT ISOLATION

3. 3.

a. Safety Injection See Functional Unit I above for all Safety injection Surveillance Requirements ro w b.

Manual N.A.

N.A.

R N.A.

1, 2, 3, 4 c.

Automatic Actuation Logic N.A.

N.A.

M!22)

N.A.

1, 2, 3, 4 and Actuation Relays 4.

CONTAINMENT VENTIIATION IS01ATION a.

Safety Injection See Functional Unit I above for all Safety injection Surveillance Requirements b.

Manual N.A.

N.A.

R N.A.

See Note (26) c.

Manual Containment Spray See Functional Unit 2a above for all Manual Containment Spray Surveillance Requirements d.

Manual Containment' See Functional Unit 3b above for all Manual Containment isolation Surveillance Requirements gg mm Isolation EE I

e.

High Radiation in Dt25)

R M

N.A.

See Note (26)

~I 55

. Exhaust Air mz f?

f.

Automatic Actuation Logic N.A.

N.A.

Mt22)

N.A.

See Note (26) and Actuation Relays ww

?.*

wea EO G7%

%M

%4

?.*

~s 82

. o."

m e-

~

m

l 1

i yy TABLE TS.4.1-1B (Page 3 of 7) n o, t

11 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS to to gg FUNCTIONAL

RESPONSE

MODES FOR WHICH gg FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED t

oo

a. a S.

STEAM LINE ISOLATION oo NN a.

Manual N.A.

N.A.

R N.A.

1, 2, 3(23) tw b.

Hi-Hi Containment S

R Q

N.A.

1, 2, 3(28) t Pressure i

c.

Hi-HL Steam Flow with Safety Injection 4

1.

Hi-Hi Steam Flow S-R' Q

N.A.

1, 2, 3t23)

I 2.

Safety Injection See Functional Unit I above for al! Safety Injection Surveillance Requirements d.

Hi Steam Flow and 2 of 4 Lo-Lo T.,, with Safety Inj ection kk 1.

Hi Steam Flow S

R Q

N.A.

1, 2,'3(23) oo a

aagg 2.

Lo-Lo T.,,

S

'R Q

N.A.

1, 2, 3(24) 5A 3.

Safety Injection See Functional Unit I above for all Safety injection surveillance Requirements 22 l

oo e.

Automatic Actuation Logic N.A.

N.A.

Mt22)

N.A.

1, 2, 3(283 I

fa "

and Actuation Relays

_ gg e l

<=$

WM

%g Mw W"

43H g

h w-v e H

l l

I

?, o?

TABLE TS.4.1-1B (Page 4 of 7) o n ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ee.

EE FUNCTIONAL

RESPONSE

MODES FOR VHICH

~ #7 FUNCTIONAL' UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REQUIRED la cc 6.

FEEDWATER IS01ATION l

3.3.

a.

Hi-Hi Steam Cenerator S

R Q

N.A.

1, 2 Level k

b.

Safety Injection See Functional U,it I above for all Safety injection Surveillance Requirements c.

Reactor Trip with 2 of 4 Low T.,, (Main Valves only) 1.

Reactor Trip N.A.

N.A.

R N.A.

1, 2 2.

Low T.,,

S R

Q N.A.

1, 2 d.

Automatic Actuation Logic N.A.

N.A.

Mt22)

N.A.

1, 2.

and Actuation Relays Ie El CL Po Po

==

M r+

W^d

.O.

meg

<a r 3,.:

m

&H o-NY G

??

TABLE TS.4.1-1B (Page 5 of 7; 11 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS mo pp FITNCTIONAL

RESPONSE

MODES FOR WHICH gg FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED al 7.

AUXILIARY FEEDWATER

c. c-

3. 3.

a.

Manual N.A.

N.A.

R N.A.

1, 2, 3

~~

b.

Steam Generator Low-Low S

R Q

N.A.

1, 2, 3 Water Level c.

Undervoltage on 4.16 kV N.A.

R R

N.A.

1, 2 Buses 11 and 12 (Unit 2:

21 and 22) (Start Turbine Driven Pump only) d.

Trip of Both Main Feedwater Pumps 1.

Turbine Driven N.A.

N.A.

R N.A.

1, 2 2.

Motor Driven N.A.

N.A.

R N.A.

1, 2 mo RR e.

Safety Injection See Functional Unit I above for all Safety Injection Surveillance Requimnents S$

55 f.

Automatic Actuation Logic N.A.

N.A.

M(22)

N.A.

1, 2, 3 22 and Actuation Relays m.

M^d G7s

=*

PP ar am

~

.D.

LA *)

m V

b" w

p

??

TABLE TS.4.1-1B (Page 6 of 7) 11 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE REQUIREMENTS mm 77 FUNCTIONAL

RESPONSE

MODES FOR WHICH 77 FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED EE ec 8.

LOSS OF POWER

" 3.

a.

Degraded Voltage N.A.

R M

N.A.

1,2,3,4 4kV Safeguards Bus b.

Undervoltage N.A.

R M

N.A.

1, 2, 3, 4 4kV Safeguards Bus mn Q. CL RR "E

EE DM M9" bb

<=M um b~

%M

.s o=

Oa-me

t

'I mm 3 2 '.

TABLE 4.1-1B (Page 7 of 7)

TABLE NOTATIONS FREQUENCY NOTATION o c-NOTATION FREOUENCY E5 S

Shift D

Daily M

Monthly Q

Quarterly R

Each Refueling Shutdown N.A.

Not Applicable i

^

TABLE NOTATION (20) One manual switch shall be tested at each (26) Whenever CONTAINMENT INTEGRITY is required refueling on a STAGGERED TEST BASIS.

and either of the containment purge systems are in operation.

(21) Trip function may be blocked in this MODE below a reactor coolant system pressure of (27) Not Used 2000 psig.

(28) Not Used (22) Each' train shall be tested at least every two months on a STAGGERED TEST BASIS.

(29) Not Used no co RR (23) When either main steam isolation valve is open.

aa (24) When reactor coolant system average 22??

temperature is greater than 520*F and either gyg main steam isolation valve is open.

gg

. u --~

g (25)

See Table 4.17-2.

m m,e 0-4

,e y

-a-w e

+

v

-1

-,2 e 22-T.T TABLE TS.4.1-1C (Page l ' of 4) t

..me MISCELIANEOUS INSTRUMENTATION SURVEILIANCE REQUIREMENTS "s *s E$

r FUNCTIONAL

RESPONSE

MODES FOR VHICH l

FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REOUIRED L

on rw 1.

Control Rod Insertion Monitor M

R S/U(38)

N.A.

1, 2 2.

Analog Rod Position S

R S/U(388 N.A.

1, 2, 3(313, 4(313, 5(31) 3.

Rod Position Deviation M

N.A.

S/U(383 N.A.

1, 2 Monitor 4.

Rod Position Bank S(32)

N.A.

1, 2, 3(31) 4(31) 5(323 Counters 5.

Charging Flow S

R N.A.

N.A.

1, 2, 3, 4 t

6.

Residual Heat Removal S

R N.A.

N.A.

4(373, 5(37), 6(373 Pump Flow FF 7.

soric Acid Tank tevel o

R(33)

M(33>

N.A.

1, 2, 3, 4 i

em SE gg 8.

Refueling Water Storage W

R M

N.A.

1,2,3,4 j

ss Tank Level or

-. 2

,E o 9.

Volume Control Tank Level S

R N.A.

N.A.

1, 2, 3, 4 un ww

10. Annulus Pressure N.A.

R.

R N.A.

See Note (39)

'g'g (Vacuum Breaker)

{j{

k k

11. Auto Load Sequencers N.A.

N.A.

M N.A.

1, 2, 3, 4.

",M ts'es 0e

?*

12. Boric Acid Make-up Flow-N.A.

R N.A.

N.A.

1, 2, 3, 4 e 's y

. F> U Channel a -~

n I

l

..c

-- -)

9

-h I

r o,a I

11 TABLE TS.4.1-1C (Page -2 of 4) l ll MISCELIANEOUS INSTRUMENTATION SURVEILIANCE REOUIREMENTS 11 s:

o o-FUNCTIONAL

RESPONSE

MODES FOR WHICH I

FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILLANCE IS REOUIRED

. ee

13. Containment Sump A, B and C N.A.

R R

N.A.

1,2,3,4 Level 1

14. Accumulator Level and S

R R

N.A.

1, 2, 3, 4 f

Pressure

[

15. Turbine First Stage S

R Q

N.A.

1 Pressure

16. Emergency Plan Radiation M

R M

N.A.

1,2,3,4, 5, 6 Instruments (353

17. Seismic Monitors R

R N.A.

N.A.

1, 2, 3, 4, 5,.6

18. Coolant Flow - RTD-S-

R M

N.A.

1, 2, 3(3')

gg Bypass Flowmeter mm oagg

19. CRDM Cooling Shroud S

N.A.

R N.A.

1, 2, 3(313, 4(32), 5'30 Exhaust Air Temperature or gg

20. Reactor Gap Exhaust Air S

N.A.

R N.A.

1,2,3,4 i

Temperature i

SS wnn

21. Post-Accident Monitoring M

R N.A.

N.A.

1, 2 gyg Instruments

%M R*

(Table TS.3.15-1)(353 mg l

MN o.

5,5

22. Post-Accident Monitoring D

R M

N.A.

1, 2 m,d Radiation Instruments e e, m-v RC (Table TS.3.15-2) g m

.m 4

d

_p_'--

??

TABLE TS.4.1-1C (Page 3 of 4) 11

~ *

MISCELI.ANEOUS INSTRUMENTATION SURVEILLANCE PEOUIREMENTS E U.

Q# CS EE FUNCTIONAL

RESPONSE

MODES FOR WHICH cc FUNCTIONAL UNIT CHECK CALIBRATE TEST TEST SURVEILIANCE IS REQUIRED oo

23. Post-Accident Monitoring M

R N.A.

N.A.

1, 2 Reactor Vessel Level Instrumentation (Table TS.3.15-3)

24. Steam Exclusion Actuation W

Y M

N.A.

1, 2, 3

25. Overpressure Mitigation N.A.

R R

N.A.

4:38) 5

26. Auxiliary Feedwater N.A.

R R

.N.A.

1, 2, 3 Pump Suction Pressure

27. Auxiliary Feedwater N.A.

R R

N.A.

1,2,3 Pump Discharge Pressure

'gg

28. NaOH Caustic Stand Pipe W

R M

N.A.

1, 2, 3, 4 mo Level EE

29. Hydrogen Monitors S

Q M

N.A.

1, 2 S5 gg

30. Containment Temperature

'M R

N.A.

1, 2, 3, 4 Monitors f

31. Turbine Overspeed N.A.

R M

N.A.

I yy Protection Trip Channel gg AS ve

'Eh

.h V '

ho n

a, 7, 7 o

((

TABLE 4.1-1C (Page 4 of 4) ee TABLE NOTATIONS g

-a FREQUENCY MCTATION aa yE NOTATION FREQUENCY

+

5 Shift m-D Daily V

Weekly M

Monthly Q

Quarterly S/U Prior to each startup Y

Yearly R

Each refueling shutdown N.A.

Not applicable TABLE NOTATION (30)

Prior to each startup following shutdown in (36)

Except for containment hydrogen monitors excess of two days if not done in previous 30 which are separately specified in this table.

days.

(37) When RHR is in operation.

(31) When the reactor trip system breakers are closed and the control rod drive system is (38) When the reactor coolant system average g @p capable of rod withdrawal.

temperature is less than 310*F.

o EE (32)

Following rod motion in excess of six inches (39) Whenever CONTAINMENT INTEGRITY is required.

AR when the computer is out of service.

2z (33) Transfer logic to Refueling Water Storage jh g.

gg Tank.

OH (34) When either main steam isolation valve is

, ca open.

m,e

$7 (35)

Includes those instruments named in the g

emergency procedure.

Table TS.4.1-2B (Page 1 of 2) l TABLE TS.4.1-2B MINIKUM FREOUENCIES FOR SAMPLING TESTS TEST FREOUENCY l

1.

RCS Gross 5/ week Activity Determination i

2.

RCS Isotopic Analysis for DOSE 1/14 days (when at power) l EQUIVALENT I-131 Concentration 3.

RCS Radiochemistry i determination 1/6 months (1) (when at power) 4.

RCS Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever Including I-131, I-133, and 1-135 the specific activity ex-ceeds 1.0 uCi/ gram DOSE EQUIVALENT I-131 or 100/1 uCi/ gram (at or above cold shutdown), and j

b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) 5.

RCS Radiochemistry (2)

Monthly 6.

RCS Tritium Activity Weekly 7.

RCS Chemistry (Cl*,F*, 02) 5/ Week 8.

RCS Boron Concentration *(3) 2/Woek (4) 9.

RUST Boron Concentration Weekly

10. Boric Acid Tanks Boron Concentration 2/ Week

11. Caustic Standpipe NaOH Concentration Monthly

12. Accumulator Boron Concentration Monthly nne)

13. Spent Fuel Pit Boron C)nr.entration Monthly / Weekly l

Required at all times.

, +.

Prairie Island Unit 1 Amendment No. 99, J0E,111 l

Prairie Island Unit 2 Amendment No. 92, J0J,104

Table TS.4.1-2B (Page 2 of 2)

TABLE TS.4.1 2B MINIMUM FREQUENCIES FOR SAMPLING TESTS TEST FREOUENCY 14.

Secondary Coolant Gross Weekly Beta-Camma activity 15, Secondary Coolant Isotopic 1/6 months (5)

Analysis for DOSE EQUIVALENT I-131 concentration 16.

Secondary Coolant Chemistry pH 5/ week (6) pH Control Additive 5/ week (6)

Sodium 5/ week (6)

Notes:

1.

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

2.

To determine activity of corrosion products'having a half-life greater than 30 minutes.

3.

During REFUELING, the boron concentration shall be verified by chemical analysis daily.

4 The maximum interval between analyses shall not exceed 5 days.

5.

If activity of the samples is greater than 10% of the limit in Specification 3.4.D, the frequency shall be once per month.

6.

The maximum interval between analyses shall not exceed 3 days.

7.

The minimum spent fuel pool boron concentration from Specification 3.8.B.l.b shall be verified by chemical analysis weekly while a spent fuel cask containing fuel is located in the spent fuel pool.

8.

The spent fuel pool boron concentration shall be verified weekly, by chemical analysis, to be within the limits of Specification 3.8.E.2.a when fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in the spent fuel pool and a spent fuel pool verification has not been performed since the last movement of any fuel assembly in the spent fuel pool.

Prairie Island Unit 1 Amendment No. 99, 108,111 Prairie Island Unit 2 Amendment No. 92, J01.104

B.2.3-2 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION Bases continued The overpower and overtemperature protection setpoints include the effects of fuel densification on core safety limits.

1 A loss of coolant flow incident can result from a mechanical or electrical failure in one or more reactor coolant pumps, or from a fault in the power supply to these pumps.

If the reactor is at power at the time of the incident, the immediate effect of loss of coolant flow is a rapid increase in coolant temperature.

This increase could result in departure from nucleate boiling (DNB) with subsequent fuel damage if the reactor is not tripped promptly.

The following trip circuits provide the necessary protection against a loss of coolant flow incident:

a.

Low reactor coolant flow b.

Low voltage on pump power supply bus c.

Pump circuit breaker opening (low frequency on pump power supply bus opens pump circuit breaker)

The low flow reactor trip protects the core against DNB in the event of either a decreasing actual measured flow in the loops or a sudden loss of one or both l

reactor coolant pumps. The set point specified is consistent with the value used in the accident analysis (Reference 7).

The low loop flow signal is caused by a condition of less than 90% flow as measured by the loop flow instrumentation.

The reactor coolant pump bus undervoltage trip is a direct reactor trip (not a reactor coolant pump circuit breaker trip) which protects the core against DNB in the event of a loss of power to the reactor coolant pumps. The set point specified is consistent with the value used in the accident analysis (Reference 7).

The reactor coolant pump breaker reactor trip is caused by the reactor coolant pump breaker opening as actuated by either high current, low supply voltage or low electrical frequency, or by a manual control switch.

The significant feature of the reactor coolant pump breaker reactor trip is the frequency set l

point, 258.2 cps, which assures a trip signal before the pump inertia is reduced to an unacceptable value.

The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief. The specified set point allows adequate operating instrument error (Reference 2) and transient level overshoot beyond their trip setting so that the trip function prevents the water level from reaching the safety valves.

The low. low steam generator water level reactor trip protects against loss of feedwater flow accidents.

The specified set point assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedvater system (Reference 8).

Prairie Island Unit 1 Amendment No. 91,111 Prairie Island Unit 2 Amendment No. B4,104

B.2.3-3 2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION Bases continued The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations. The prescribed set point above which these trips are unblocked assures their availability in the power range where needed. The reactor trips related to loss of one or both reactor coolant pumps are unblocked at approximately 10% of RATED THERMAL p0WER.

The other reactor trips specified in 2.3.A.3. above provide additional protection. The safety injection signal trips the reactor to decrease the severity of the accident condition.

The reactor is tripped when the turbine generator trips above a power level equivalent to the load rejection capacity of the steam dump valves. This reduces the severity of the loss-of-load transient.

The positive power range rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level.

Specifically, this trip compliments the power range nuclear flux high and low trip to assure that the criteria are met for rod ejection frem partial power.

The negative power range rate trip provides protection against DNB for control rod drop accidents. Most rod drop events will cause a sufficiently rapid decrease in power to trip the reactor on the negative power range rate trip signal. Any rod drop events which do not insert enough reactivity to cause a trip are analyzed to ensure that the core does not experience DNB.

Administrative limits in Specification 3.10 require a power reduction if design power distribution limits are exceeded by a single misaligned or dropped rod.

1 References 1.

USAR, Section 14.4.1 2.

USAR, Section 14.3 3.

USAR, Section 14.6.1 4

USAR, Section 14.4.1 5.

USAR, Section 7.4.1.1, 7.2 6.

USAR, Section 3.3.2 7.

USAR, Section 14.4.8 8.

USAR, Section 14.1.10 Prairie Island Unit 1 Amendment No. 9J, 111 Prairie Island Unit 2 Amendment No. E#,104

l l

B.3.5-1 3.5 INSTRUMENTATION SYSTEM Bases l

Instrumentation has been provided to sense accident conditions and to l

initiate reactor trip and operation of the Engineered Safety Features l

(Reference 1).

The OPERABILITY of the Reactor Trip System and the Engineered Safety System instrumentation and interlocks ensures that: (1) l the associated ACTION and/or reactor trip will be initiated when the i

parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features l

instrumentation and, (3) sufficient system functions capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the safety analysis.

Specified surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System",

and supplements to that report. Out of service times were determined based on naintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

The evaluation of surveillance frequencies and out of service times for the reactor protection and engineered safety feature instrumentation described in WCAP 10271 included the allowance for testing in bypass.

The evaluation assumed that the average amount of time the channels within a given trip function would be in bypass for testing was 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 offset the temperature-induced reactivity addition that could otherwise result from cooldown following a steam line break.

Prairie Island Unit 1 Amendment No. SJ,111 Prairie Island Unit 2 Amendment No. Ef,104

B.3.5-2 3.5 INSTRUMENTATION SYSTEM Bases continued 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 auto-matic 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.

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, uncon-trolled steam release from more than one steam generator.

The steam lines are isolated on high containment pressure (Hi-HL) or high steam line flow in coincidence with low T,y and safety injection or high steam flow (Hi-Hi) in coincidence with safety injection. Adequate protection is afforded for breaks inside or outside the containment even when it is assumed that the steam line check valves do not function properly.

Containment Ventilation Isolation Valves in the containment purge and inservice purge systems automati-cally 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 containment, redundant isolation dampers in certain ventilation ducts are closed (Reference 4).

Safeguards Bus Voltage Relays are provided on buses 15, 16, 25, and 26 to detect undervoltage and degraded voltage (the voltage level at which safety related equipment may not operate properly).

Relays are not provided on 4 kV safeguards bus 27 to detect undervoltage and degraded voltage since voltage is monitored on the 4 kV source safeguards bus (i.e., bus 25 or bus 26 ) to which it is Prairie Island Unit 1 Amendment No. 9J. J03,111 Prairie Island Unit 2 Amendment No. E,4, 95,104

B.3.5-3 3.5 INSTRUMENTATION SYSTEM Bases continued Safeguards Bus Voltage (continued) connected. Upon receipt of an undervoltage signal the automatic voltage restoring scheme is actuated after a short time delay which prevents actuation during normal transients (such as motor starting) and which allows protective relaying operation during faults.

When degraded voltage is sensed, two time delays are actuated. The first time delay is long enough to allow for normal transients. The first time delay annunciates that a sustained degraded voltage condition exists and enables logic which will ensure that voltage and timing are adequate for safety injection loads by automatically performing the following upon receipt of a safety injection signal:

1.

Auto start the diesel generator; 2.

Separate the bus from the grid; 3.

Load the bus onto the diesel generator; and 4

Start the load sequencer (including safety injection loads).

The second longer time delay is used to allow the degraded voltage condition to be corrected by external actions within a time period that will not cause damage to operating equipment.

If voltage is not restored within that time period, the logic automatically performs the following:

1.

Auto start the diesel generator; 2.

Separate the bus from the grid; 3.

Load the bus onto the diesel generator; and 4.

Start the load sequencer.

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.

i 1.

Low low water level in either steam generator 2.

Trip of both main feedwater pumps j

3.

Safety Injection signal j

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 provi. des assurance that water can be supplied to the steam generators for decay heat removal when the normal feedwater system is not available.

Underfrequency 4kV Bus The underfrequency 4kV bus trip does not provide a direct reactor trip signal to the reactor protection system. A reactor coolant pump bus i

underfrequency signal from both buses provides a trip signal to both reactor coolant pump breakers. Trip of the reactor coolant pump breakers results in a reactor trip.

The underfrequency trip protects against postulated flow coastdown events.

Prairie Island Unit 1 Amendment No. ST, LOT,111 Prairie Island Unit 2 Amendment No. 24, 96, 104

B.3.5-4 3.5 INSTRUMENTATION SYSTEM Bases continued Limiting Instrument Setpoints (continued) 2.

The Hi-Hi containment pressure limit is set at about 50% of the maximum internal pressure for initiation of containment spray and at i

about 30% for initiation of steam line isolation.

Initiation of Containment Spray and Steam Line Isolation protects against large loss of coolant (Reference 2) or steam line break accidents (Reference 3) as discussed in the safety analysis.

3.

The pressurizer low pressure limit is set substantially below system operating pressure limits.

However, it is sufficiently high to protect against a loss of coolant accident as shown in the safety analysis (Reference 2).

4.

The steam line low pressure signal is lead / lag compensated and its set-point is set well above the pressure expected in the event of a large steam line break accident as shown in the safety analysis (Reference 3).

5.

The high steam line flow limit is set at approximately 20% of nominal full-load flew 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 its hot shutdown value.

The safety analysis shows that these settings provide protection in the event of a large steam break (Reference 3).

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 the Bases of Section 2.3 of the Technical Specification.

l 7.

High radiation signals providing input to the Containment Ventilation Isolation circuitry 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.

8.

The degraded voltage protection setpoint is 294.8% and 596.2% of nominal 4160 V bus voltage. Testing and analysis have shown that all safeguards loads will operate properly at or above the minimum degraded voltage setpoint. The maximum degraded voltage setpoint is chosen to prevent unnecessary actuation of the voltage restoring scheme at the minimum expected grid voltage.

The first degraded voltage time delay of 8 0.5 seconds has been shown by testing and analysis to be long enough to allow for normal transients (i.e., motor starting and fault clearing).

It is also longer than the time required to start the safety injection pump at minimum voltage.

The second degraded voltage time delay is provided to allow the degraded voltage condition to be corrected within a time frame which will not cause damage to permanently connected Class lE loads.

M.

?

Prairie Island Unit 1 Amendment No. Sd. tM 111 Prairie Island Unit 2 Amendment No. 34, ps,104 i

B.3.5-5 3.5 INSTRUEENTATION SYSTEM gases continued Limiting Instrument Setpoints (continued)

The undervoltage setpoint is 75 2.5% of nominal bus voltage.

The minimum setpoint ensures equipment operates above the limiting value of 75% (of 4020 V) for one minute operation. The 75% maximum setpoint is chosen to prevent unnecessary actuation of the voltage restoring scheme during voltage dips which occur during motor starting.

The undervoltage time delay of 4 1.5 seconds has been shown by testing and analysis to be long enough to allow for normal transients and short enough to operate prior to the degraded voltage logic, providing a rapid transfer to an alternate source.

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 effectiveness 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 sufficient redundancy to provide the capability for CHANNEL CALIBRATION and test at j

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.

/

Eeferences 1.

USAR, Section 7.4.2 2.

USAR, Section 14.6.1 3.

USAR, Section 14.5.5 4.

FSAR, Appendix I Prairie Island Unit 1 Amendment No, t03,111 Prairie Island Unit 2 Amendment No. 96, 104

B.3.10-1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and "RCCA(s)" are synonymous.

A.

Shutdown Margin A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made suberitical from all opexating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARCIN requirements vary throughout core life as a function of fuel depletion, reactor coolant system boron concentration and reactor coolant average temperature.

The most restrictive condition occurs at end of life and is associated with a postulated steam line break accident and resulting uncontrolled reactor coolant system cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN (shown in Figure TS.3.10-1 as a function of equilibrium hot full power boron concentration) is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirements are based upon this limiting conditien and are consistent with plant safety analysis assumptions. With reactor coolant system average temperature less than 200*F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 14 Ak/k SHUTDOWN MARGIN provides adequate protection.

i In POWER OPERATION and HOT STANDBY, with k,gg a 1, SHUTDOWN MARGIN is ensured by complying with the rod insertion limitations in Specification 3.10.D.

In HOT SHUTDOWN, INTERMEDIATE SHUTDOWN and COLD SHUTDOWN, the SHUTDOWN MARGIN requirements in Specification 3.10.A are applicable to j

provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above.

For REFUELING, the shutdown reactivity requirements are specified in Table TS.1-1.

1 When in POWER OPERATION and HOT STANDBY, SHUTDOWN MARGIN is determined assuming the fuel and moderator temperatures are at the nominal zero power design temperature of $47'F.

With any rod cluster control assembly not capable of being fully inserted, the reactivity worth of the rod cluster control assembly must be accounted j

for in the determination of SHUTDOWN MARGIN.

B.

Power Distribution Control The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operations) and II (Incidents of Moderate frequency) events by: (a) maintaining the minimum DNBR in the core of greater than or equal to 1.30 for Exxon fuel and 1.17 for Westinghouse fuel during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumeu design criteria.

The ECCS analysis was performed in accordance with SECY 83 472.

One calculation at the 95% probability level was performed as well as one calculation with Prairie Island Unit 1 Amendment No. $1,$2,111 Prairie Island Unit 2 Amendment No. $4,$5,104

B.3.10 2 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bsses continued B.

Power Distribution Control (continued) all the required features of 10 CFR Part 50, Appendix K.

The 95%

probability level calculation used the peak linear heat generation rate specified in the CORE OPERATING LIMITS REPORT.

The Appendix K calculation used the peak linear heat generation rate specified in the CORE OPERATING l

LIMITS REPORT for the Fo limit specified in the CORE OPERATING LIMITS REPORT.

Maintaining 1) peaking factors below the Fn limit specified in the CORE OPERATING LIMITS REPORT during all Condition I events and 2) the peak linear heat generation rate below the value specified in the CORE OPERATING LIMITS REPORT at the 95% probability level assures compliance

]

with the ECCS analysis.

During operation, the plant staff compares the measured hot channel factors, F"n and F"a, (described later) to the limits determined in the transient and LOCA analyses.

The terms on the right side of the equations in Section 3.10.B.1 represent the analytical limits.

Those terms on the left side represent the measured hot channel factors corrected for engineering, calculational, and measurement uncertainties.

F"o is the measured Nuclear Hot Channel Factor, defined as the maximum local heat flux on the surface of a fuel rod divided by the average heat flux in the core. Heat fluxes are derived from measured neutron fluxes and fuel enrichment.

The K(Z) function specified in the CORE OPERATING LIMITS REPORT is a normalized function that limits Fn axially.

The K(Z) value is based on large and small break LOCA analyses.

V(Z) is an axially dependent function applied to the equilibrium measured F*o to bound F*n's that could be measured at non-equilibrium conditions.

This function is based on power distribution control analyses that evaluated the effect of burnable poisons, rod position, axial effects, and xenon worth.

EF n, Encineerine Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances.

The engineering factor allows for local variations in enrichment,-pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.

The 1.05 multiplier accounts for uncertainties associated with measurement of the power distribution with the movable incore detectors and the use of those measurements to establish the assembly local power distribution.

F"o (equil) is the measured limiting F*n obtained at equilibrium conditions during target flux determination.

F"a, Nuclear Enthalov Rise Hot Channel Factor, is defined as the ratio of the integrel of linear power along the rod with the highest integrated power to the average rod power.

Prairie Island Unit 1 Amendment No. 9J, 92,111 Prairie Island Unit 2 Amendment No. BA, $5,104

B.4.1-1 4.1 OPERATIONAL SAFETY REVIEW pases CHANNEL CHECK Failures such as blown instrument fuses, defective indicators, faulted amplifiers which result in " upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system.

Furthermore, such failures are, in many cases, revealed by alarm or annunciator action, and a check supplements this type of built-in surveillance.

Based on experience in operation of both conventional and nuclear plant systems, when the plant is in operation, the minimum checking frequencies set forth are deemed adequate for reactor and steam system instrumentation.

CHANNEL CALIBRATION Calibration is performed to ensure the presentation and acquisition of accurate information.

The nuclear flux (linear level) channels daily calibration against a thermal power calculation will account for errors induced by changing rod patterns and core physics parameters.

Other channels are subject only to the " drift" errors induced within the instrumentation itself and, consequently, can tolerate longer intervals between calibration.

Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at intervals of each refueling shutdown.

j Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

CHANNEL FUNCTIONAL TESTS The specified surveillance intervals for the Reactor Protection and Engineered Safety Features instrumentation have been determined in accordance with WCAP 10271, " Evaluation of Surveillance Frequencies and out of Service Times for the Reactor Protection Instrumentation System",

and supplements to that report.

Surveillance intervals were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

CHANNEL RESPONSE TESTS Measurement of response times for protection channels are performed to assure response times within those assumed for accident analysis (USAR, Section 14).

Prairie Island Unit 1 Amendment No. 97,111 Prairie Island Unit 2 Amendment No. Ep, 104

=..

'po mo u

'^

S UNITED STATES NUCLEAR REGULATORY COMMISSION

%q...../

WASHINGTON, D.C. 20555 0001 NORTHERN STATES POWER COMPANY DOCKET N0. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UN!T NO. 2 MENDMENT TO FACILITY OPERATING LICENSE Amendment No.104 License No. DPR-60 1.

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

A.

The application for amendment by Northern States Power Company (the licensee) dated September 21, 1992, as revised December 29, 1992, November 24, 1993, May 17, 1994, and June 21, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set l

forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the application, tne provisions of the Act, and the rules and regulations of the Commission; 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 compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common 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 Commission's regulations and all applicable requirements F we been satisfied.

2.

Accordingly, the license is amended 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-60 is hereby amended to read as follows:

(. I

i Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.104, 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 issuance.

FOR T CLEAR REGULATORY COMMISSION k

tf L. B. Marsh, Director Project Directorate 111-1 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 10, 1994 1

1

ATTACHMENT TO LICENSE AMENDMENT N0.104 FACILITY OPERATING LICENSE N0. DPR-60 DOCKET N0. 50-306 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

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

~

REMOVE INSERT TS-xii TS-xii TS.1-1 TS.1-1 TS.1-2 TS.1-2 TS.1-3 TS.1-3 TS.1-4 TS.1-4 TS.1-5 TS.1-5 TS.1-7 TS.1-7 TS.1-8 TS.1-8 TABLE TS.1-1 i

TS.2.3-3 TS.2.3-3 TS.2.3-4 TS.2.3-4

'l TS.3.5-1 TS.3.5-1 TABLE TS.3.5-2 (Pages 1 & 2)

TABLE TS.3.5-2A (Pages 1 - 6)

TABLE TS.3.5-2B (Pages 1 - 9)

TABLE TS.3.5-3 (Pages 1 & 2)

TABLE TS.3.5-4 (Pages 1 & 2)

TABLE TS.3.5-5 TABLE TS.3.5-6 TS.3.10-1 TS.3.10-1 TS.3.10-2 TS.3.10-2 TS.4.1-1 TS.4.1-1 TABLE TS.4.1-1 (Pages 1 - 5)

TABLE TS.4.1-1A (Pages 1 - 5)

TABLE TS.4.1-1B (Pages 1 - 7)

TABLE TS.4.1-lC (Pages 1 - 4)

TABLE TS.4.1-2B (Pages 1 & 2)

TABLE TS.4.1-28 (Pages 1 & 2)-

B.2.3-2 B.2.3-2 B.2.3-3 B.2.3-3 B.3.5-1 B.3.5-1 B.3.5-2 B.3.5-2 B.3.5-3 B.3.5-3 B.3.5-4 B.3.5-4 B.3.5-5 B.3.5-5 B.3.10-1 B.3.10-1 B.3.10-2 B.3.10-2 B.4.1-1 B.4.1-1 B.4.1-2

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