ML20117J080

Amends 123 & 116 to Licenses DPR-42 & DPR-60,respectively, Revising TS by Changing Pressurizer & Main Steam Safety Valve Lift Setting Tolerance from +1 Percent to +3 Percent (as Found Only),Revising Safety Limit Curves

ML20117J080
Person / Time
Site:	Prairie Island
Issue date:	05/21/1996
From:	Wetzel B NRC (Affiliation Not Assigned)
To:
Shared Package
ML20117J082	List: ML20117J080 ML20117J085
References
NUDOCS 9605300147
Download: ML20117J080 (50)

Text

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4.

• 4 UNITED STATES g

,, g NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 30886-0001 o%

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NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 123 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 May 4, 1995, as supplemented November 27, 1995, and March 1, 1996, 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; i

D.

The issuance of this amendment will nat be inimical to the common defense and security or to the health and safety of the public; and 1

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:

9605300147 960521 PDR ADOCK 05000282 p

PDR d

O O j 1

l Technical Specifications The Technical Specifications contained in Appendix A, as revised j

through Amendment No.123, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the i

Technical Specifications.

3.

This license amendment is effective as of the date of issuann, with full implespentation within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Jw 0% p Beth A. Wetzel, Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications

)

Date of Issuance:

May 21, 1996

a ATTACHMENT TO LICENSE AMENDMENT NO.123 FACILITY OPERATING LICENSE NO. DPR-42 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-1 TS-1 TS-viii TS-viii TS-x TS-x TS-xiii TS-xiii TS.2.1-1 TS.2.1-1 TS.2.2-1 Figure TS.2.1-1 Figure TS.2.1-1 TS.2.3-2 TS.2.3-2 TS.2.3-3 TS.2.3-3 TS.3.4-1 TS.3.4-1 Table TS.4.1-2A Table TS.4.1-2A (Page 1 of 2)

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

TS.6.4-1 B.2.1-1 B.2.1-1 B.2.1-2 B.2.1-2 B.2.1-3 B.2.1-4 B.2.1-5 Figure B.2.1-1 B.2.2-1 B.2.2-1 B.3.1-2 B.3.1-2 B.3.1-3 B.3.1-3 B.3.4-1 B.3.4-1 B.3.4-2

TS-i l

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS i

i l

TS SECTION TITLE PACE

)

1 1.0 DEFINITIONS TS.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING TS.2.1-1 l

2.1 Safety Limits TS.2.1-1 i

2.2 Safety Limit Violations TS.2.1-1 2.3 Limiting S4f;ty System Settings, Protectivo Instrumer.thilon TS.2.3-1 A. Protective Instrumentation Settings for Reactor Trip TS.2.3-1 B. Protective Instrumentation Settings for Reactor Trip Interlocks TS.2.3-4 C. Control Rod Withdrawal Stops TS.2.3 4 l

l l

Prairie Island Unit 1 Amendment No. 123 i

l Prairie Island Unit 2 Amendment No. 116 l

[

l i

TS-viii TABII OF CONTENTS (Continued)

TS SECTION IIILE PACE 6.0 ADMINISTRATIVE CONTROLS TS.6.1-1 6.1 Organization TS 6.1-1 6.2 Review and Audic TS.6.2-1 A. Safety Audit Committee (SAC)

TS.6.2-1

1. Membership TS 6.2-1

2. Qualifications TS 6.2-1

3. Meeting Frequency TS.6.2-2

4. Quorum TS.6.2-2

5. Responsibilities TS.6.2-2

6. Audit TS.6.2-3

7. Authority TS.6.2-4

8. Records TS.6'.2-4

9. Procedures TS.6.2-4 B. Operations Committee (DC)

TS.6.2-5

1. Membership TS.6.2-5

2. Meeting Frequency TS.6.2-5

3. Quorum TS.6.2-5

4. Responsibilities TS.6.2-5

5. Authority TS.6.2 6

6. Records TS.6.2-6

7. Procedures TS.6.2-6 C. Maintenance Procedures T5.6.2-7 6.3 Special Inspections and Audits TS.6.3 1 6.4 Deleted 6.5 Flant Operating Procedures TS.6.5-1 A. Plant Operations TS.6.5-1

5. Radiological TS.6.5-1 C. Maintenance and Test TS.6.5-4 D. Deleted E. Offsite Dose Calculation Manual (0DcN)

TS.6.5-4 F. Security TS.6.5-5 C. Temporary Changes to Procedures TS.6.5-5 H. Radioactive' Effluent Controls Program TS.6.5-6 I. Explosive Gas and Storage Tank Monitoring rrogram TS.6.5-7 6.6 Flant Operating Records TS.6.6-1 A. Records Retained for Five Years TS.6.6-1 B. Racords Retained for the Life of the Plant TS.6.6-1 Prairie Island Unit 1 Amendment No. 91,96,fff,123 Prairie Island Unit 2 Amendment No. 84,89,116,116

TS-x TABLE OF CONTENTS (continued)

TS BASES SECTION TITLE EagE 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits B.2.1-1 A.

Reactor Core Safety Limits B.2.1-1 B.

Reactor Coolant System Pressure Safety Limits B.2.1-5 2.2 Safety Limit Violations B.2.2-1 2.3 Limiting Safety System Settings, Protective B.2.3-1 Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 E. Maximum Reactor Coolant Oxygen, Chloride B.3.1-8 and Fluoride Concentration F. Isothermal Temperature Coefficient (ITC)

B.3.1-9 3.2 Chemical and Volume Control System E.3.2-1 3.3 Engineered Safety Features B.3.3-1 3.4 Steam and Power Conversion Systems B.3.4 1 3.5 Instrumentation System B.3.5-1 3.6 Containment System B.3.6-1 3.7 Auxiliary Electrical System B.3.7-1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10-9 F. Inoperable Rod Position Indicator Channels B.3.10-9 C. Control Rod Operability Limitations B.3.10-9 l

H. Rod Drop Time B.3.10-10 I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.11-1 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13-1 3.14 Deleted 3.15 Event Monitoring Instrumentation B.3.15-1 Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment 'No. 116

-. -. ~ - - -.

TSoxiii APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Reactor Core Safety Limits 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activ'ty Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131

'3.8-1 spent Fuel Fool Unrestricted Region Minimum Burnup Requirements 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In Leakage Rate 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell layout 5.6-2 Spent Fuel Fool Checkerboard Region Minimum Burnup Requirements B.2.1-1 Origin of Safety Limit Curves at 2235 psigwith delta-T Trips and Locus of Reactor Conditions at which SG Safety Valves Open Prairie Island Unit 1 Amendment No. 105.108.122,123 Prairie Island Unit 2 Amendment No. 98,101.115,116

TS.2.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 2.1 SAFETY LIMITS A.

Reactor Core Safety Limits In MODES 1 and 2, combination of thermal power (measured in aT), pressurizer pressure, and the highest reactor coolant system loop average temperature shall not exceed the limits shown in Figure TS.2.1-1.

B.

Reactor Coolant System Pressure Safety Limit In MODES 1, 2, 3, 4, and 5, the reactor coolant system pressure shall not exceed 2735 psig.

2.2 SAFETY LIMIT VIOLATIONS A.

If SAFETY LIMIT 2.1.A. is violated, restore compliance and be in MODE 3 I

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B.

If SAFETY LIMIT 2.1.B. is violated:

1. In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2. In MODE 3, 4, or 5. restore compliance within 5 minutes.

C.

If a SAFETY LIMIT is violated, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notify the NRC Operations Center in accordance with 10CFR50.72.

D.

If a SAFETY LIMIT is violated, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notify the Vice President Nuclear Generation, and the Chairman of the Safety Audit Committee or their designated alternates.

E.

If a SAFETY LIMIT is violated, within 30 days a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73. The LER shall be submitted to the NRC, the Vice President Nuclear Generation and the Safety Audit Committee.

F.

If a SAFETY LIMIT is violated, operation of the unit shall not be resumed until authorized by the NRC.

Prairie Island Unit 1 Amendment No. 77,91,105,123 Prairie Island Unit 2 Amendment No. 70,84,98,116

Figura TS.2.1-1 660 650

=-

640 630

+

.+-

,u.

Q 620 O

h e

+"

F-610 2

slii 600 u

g.

Ee 590 F-a) cn 2

580 e

2385 psig I

570 560

-l1985 100% Flow (68.2 x 10'lb/hr)

.jg'885l 550

- 1785l 540 i

i i

i i

i i

O 10 20 30 40 50 60 70 80 delta-T (T -T )

F 3 c Reactor Core Safety Limits Figure TS.2.1-1 Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

l l

TS.2.3-2 2.3.A.2.d Cont.

and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chamber, with gains to be selected based on measured instrument response during plant startup tests, such that where qt and qb are the percent I

power in the top and bottom halves of the core, respectively, and qs + q3 is total core power in percent of rated power:

1.

for qt - q3 within -12% and +94, f (AI) - 0, and l

2.

for each percent that the magnitude of qt - q3 exceeds

+94 the AT trip set point shall be automatically reduced by an equivalent of 2.5 percent of RATED THERMAL POWER.

3.

for each percent that the magnitude of qs - q3 exceeds

-124, the T trip set point shall be automatically reduced by an equivalent of 1.5 percent of RATED THERMAL POWER.

e.

Overpower A T K t sT 33

'T 8 'To IE4~

~K6(T-r') - r (A2 1 p

1 + tas where AT, Indicated AT at RATED THERMAL POWER 1

T Average temperature, *F T'

567.3'F K.

s 1.10 K

0.0275 for increasing T; O for decreasing T 3

Ks 0.002 for T > T',

O for T < T' 10 sec t3 f(AI) - as defined in d. above f.

Low reactor coolant flow per loop - t90% of normal indicated loop flow as measured at loop elbow tap.

l i

l Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

~

a 1

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.

Power range neutron flux rate.

1.

Positive rate - $15% of RATED THERMAL POWER with a time constant 22 seconds 2.

Negative rate - $7% of RATED THERMAL POWER with a time constant of 22 seconds

3. Other reactor trips a.

High pressurizer water level - 590% of narrow range instrument span.

b.

Low-low steam generator water level - 25% of narrow range i

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 i

1 1

I i

l Prairie Island Unit 1 Amendment No. 77, 77, 777,123 Prairie Island Unit 2 Amendment No. F#, 77, 79#,116

_r.._

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

Obiective To specify minimum conditions of steam-relieving capacity and auxiliary feed-water

~

supply necessary to assure the capability of removing decay heat from the reactor, and to limit the concentration of activity that might be released by steam relief to the atmosphere.

Specification A.

Steam Generator Safety and Power Onorated Relief Valves 1.

A reactor shall not be made or maintained critical nor shall reactor l

coolant system average temperature exceed 350'F unless the following conditions are satisfied (except as specified in 3.4.A.2 below):

Ten steam generator safety valves shall be OPERABLE with lift settings a.

of 1077, 1093, 1110, 1120 and ll31.psig 34 except during testing.

l

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

2.

During STARTUP OPERATION or POWER OPERATION, the following condition of inoperability may exist provided STARTUP OPERATION is discontinued until OPERABILITY is restored.

If OPERABILITY is not restored within the time specified, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce reactor coolant system average temperature below 350*F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

a. One steam generator power-operated relief valve may be inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

B.

Auxiliary Feedwater System 1.

A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 350*F unless the following conditions are satisfied (except as specified in 3.4.B.2 below):

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

b.

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

c. Valves and piping associated with the above components are OPERABLE except that during STARTUP OPERATION necessary changes may be made in motor-operated valve position. All such changes shall be under direct administrative control, l

Prairie Island Unit 1 Amendment No. 123 i

Prairie Island Unit 2 Amendment No. 116

l

@3 Table TS.4.1-2A (Page 1 of 2) 77 XX ss MINIMUM FREOUENCIES FOR EOUIPMENT TESTS FSAR Sect.

EE Eaulement Test Freauency Reference l

a. o.

i

@F

1. Control Rod Assemblies Rod Drop Times of full length All rods during each refueling shutdown or 7

rods following each removal of the reactor vessel head; affected rods following maintenance u-on or modification to the control rod drive system which could affect performance of those specified rods

2. Control Rod Assemblies Partial movement of all rods Every Quarter 7

I

3. tressurizer Safety Verify OPERABLE in accordance Per ASME Code,Section XI Inservice Testing Velves with the Inservice Testing Program l

Program (i 34). Following l

testing, lift settings shall be within ilt

4. Main Steam Safety Verify each required lift Per ASME Code,Section XI Inservice Testing Valves setpoint in accordance with Program the Inservice Testing Program (1 34). Following testing, lift settings shall be within ilt

^Y

• e

5. Reactor Cavity Water Imvel Prior to moving fuel assemblies or control S$

rods and at least once every day while the

((

cavity is flooded.

"d

=.

e,.

6. Pressurizer PORV Functional Quarterly, unless the block valve has been

~

• h u'

Block Valves s

closed per Specification 3.1.A.2.c.(1).(b).2

[

or 3.1.A.2.c.(1).(b).3.

O O

{'

7. Pressurizer PORVs Functional Every 18 months su

9 mm l

9E Table TS.4.1-2A (Page 2 of 2) rr33 em MINIMim yiinnUENCIES FOR EDUIPMENT_TasTS yg t

aa~"

FSAR Sect.

i Ecutoment Test Frecuency Reference l

8. Deleted w-

9. Primary System Imakage Evaluate Daily 4

10. Deleted

11. Turbine stop valves, Functional Turbine stop valves, governor valves and 10 governor valves, and intercept valves are to be tested at a intercept valves.

frequency consistent with the methodology (Part of turbine presented in WCAP-11525 "Probabilistic overspeed protection)

Evaluation of Reduction in Turbine Valve test Frequency", and in accordance with the established NRC acceptance criteria for the i

probability of a turbine missle ejection incident of 1.0x10'8 per year. In no case shall the turbine valve test interval exceed one year.

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4 B.2.1-1 2.1 SAFETY LIMITS A. Reactor Core Safety TAmits Aaaan i

To maintain the integrity of the fuel cladding and prevent fission product I

release, it is necessary to prevent overheating of the cladding under all 1

operating conditions. This is accomplished by operating the hot regions of the core within the nucleate boiling regime of heat transfer wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation temperature. Tho upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation. Therefore, the observable parameters; thermal power, reactor coolant temperaturs and pressure have been related to DNB through the W-3 and WRB-1 DNB correlations. The W-3 DNB correlation is used for Exxon fuel.

The WRB-1 DNB correlation is used for Westinghouse fuel.

The W-3 and WRB-1 DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNB ratio, DNBR, during steady state operation, normal operational transients, and anticipated transients is limited to 1.30 for the Exxon fuel using the W-1 3 correlation and to 1.17 for the Westinghouse fuel using the WRB-1 correlation. There is a third DNBR limit specifically for the steam line break accident but it does not apply to the safety limit curve calculations.

These limits correspond to a 954 probability at a 954 confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The safety limit curves of Figure TS.2.1-1 define the regions of acceptable operation with respect to average temperatures, power, and pressurizer pressure. These boundaries of acceptable operations are limited by the thermal overpower limit (fuel melting), thermal overtemperature limit (cladding damage based on DNB considerations), and the locus of points where the steam generation safety valves open. These limits are used to set the overpower and overtemperature AT trip setpoints.

The safety limit curves of Figure TS.2.1-1 comprise the most limiting of the following four criteria:

1) Vessel Exit Temperature < 650*F This is the design temperature limit. This limit defines the portion of the safety limit curves from 0 AT to the first knee for the 2235 and 2385 psig curves. At these pressures, the temperature limit of 650* is more Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

4 B.2.1-2 A.

Reactor Core Safety Limits RAAAA continued limiting than the Tk.s limit. The locus of points is calculated from a heat balance with the minimum RCS flow specified in TS 3.10.J.

2) Vessel Exit Temperature < T;.s This limit ensures that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated water which ensures that the AT measured by instrumentation used by the RPS as a measure of core thermal power is proportional to core power. This limit defines the portion of the safety limit curves from 0 AT to the first knee for the 1985, 1885 and 1785 psig curves. The locus of points is calculated from a heat balance with the minimum RCS flow specified in TS 3.10.J.

3) Minimum DNBR > 1.3 or 1.17 whichever is applicable As mentioned before,1.3 is the DNBR limit for Exxon fuel using the W-3 critical heat flux correlation and 1.17 is the DNBR limit for Westinghousa fuel using the WRB-1 critical heat flux correlation. The locus of points past the first knee at all pressures represents the thermal-hydraulic conditions abo.ve which the hot channel has a DNBR less than the limit. The conditions are evaluated using approved DNB methodology. The assumptions used in the calculation include a bypass flow of 64, an Fin greater than 1.75, and a rod bow penalty of 2.64.

The very shallow knee at full power AT occurs because the Fin (hot channel power) is allowed to increase for core power less than RATED THERMAL POWER as described in TS 3.10.8.1.

4) Hot Channel Exit Quality < 15% or 304.whichever is applicable This limit is typically not the most restrictive because it is generally approached at lowir powers where the T,as < T,.s or 650*F is more limiting. However, it is considered when the DNB calculations described above are performed using approved DNB methodology. This limit is determined by the range of the channel exit quality for the critical heat flux correlations. The maximum channel exit quality limit is 154 for the W-3 correlation and 304 for the WRB-1 correlation.

Operation above the safety limit curves of Figure TS 2.1-1 is not acceptable.

At each pressure the safety limit curve is the most restrictive combination of the four limits discussed above. The area of acceptable operation below the safety limit curves is bounded by the OTAT trip, the OPAT trip, and the locus of points where the steam generator (main steam) safety valves open.

The AT trips are set conservatively with respect to the safety limit curves to protect the core from exceeding the safety limits. The locus of points at which the steam generator safety valves open defines the thermodynamic limit of temperature conditions in the RCS based on the maximum pressure in the steam generators. For this calculation, it is assumed that the pressure in the steam generator is 1195 psig which is 110% of design pressure.

It is Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

- -. - ~. - -. -..__ -. - - -.- -

~. - - - -. - - -

j a

8.2.1-3 A.

Reactor Core Safety TAmits lagga continued i

required that the steam generator safety valves protect the pressure from exceeding 1104 of design pressure so using 1195 psig in the calculations is conservative.

Thus, the reactor is protected from violating the safety-limits by the physical limit of the AT trips and the opening of the steam generator safety valves.

As an example, all the limits for the 2235 psig curve are plotted in Figure B.2.1-1 along with the AT trips and the locus of points where the steam generator safety valves open. This plot demonstrates that the AT trips and the steam generator safety valves do protect the reactor from exceeding the safety limits. Note, however, that the OTAT trip locus on that plot is for 3

steady state conditions and that the locus will drop in response to the rate at which the AT is increasing.

In addition, f(AI) increasing will also lower l

the OTAT trip 2ocus.

l The safety limit curves are plotted with AT on the x-axis for the following l

two reasons:

1.) the full power AT is different at different temperatures and pressures because water properties are nonlinear. This makes it difficult to plot the curves at'each pressure using the same scale for the percent power axis.

2.) the AT trip setpoints which the reactor protection system actually calculates is based on the AT, not the percent power.

Except for special tests, POWER OPERATION with only one loop or with natural circulation is not allowed.

Safety limits for such special tests will be determined as a part of the test procedure.

The curves are conservative for the following nuclear hot channel factors:

78 - 7"TF (1 + PFDH(1-P)) ; and 78 - Pert as as o

o where:

- Farr is the Fo limit at RATED THERMAL POWER specified in the CORE OPhTINGLIMITSREPORT.

- Farr is the Fas limit at RATED THERMAL POWER specified in the CORE as OPERATING LIMITS REPORT.

- PFDH is the Power Factor Multiplier for 7'g specified in the CORE OPERATING LIMITS REPORT Use of these factors results in more conservative safety limits than would result from power distribution limits in Specification TS.3.10.

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

r" i

l l

l l

B.2.1 4 A.

Reactor Core Safety Limits l

11111 continued l

This combination of hot channel factors is higher than that calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. The control rod insertion limits are covered by Specification 3.10.

Adverse power distribution factors could occur at lower power levels because additional control rods are in the core. However, the control rod insertion limits specified in the CORE OPERATING LIMITS REPORT assure that the DNB ratio l

is always greater at part power than at full power.

The Reactor Control and Protective System is designed to prevent any anticipated combination of transient conditions that would result in a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 for Westinghouse fuel.

l l

I l

l Prairie Island Unit 1 Amendment No. 123 j

Prairie Island Unit 2 Amendment No. 116 i

1 l

l l

t.

8.2.1-5 l

l l

B.

Reactor Coolant System Pressure Safety Limit l

I l

l M

i l

The reactor coolant system (Reference 1) serves as a barrier preventing radionuclides contained in the reactor coolant from reaching the atmos-phere.

In the event of a fuel cladding failure the reactor coolant system is the primary barrier against the release of firrion products.

By establishing a system pressure limit, the continued integrity of the reactor coolant system is assured. The maximum transient pressure allowable-in the reactor coolant system pressure vessel under the ASME Code,Section III is 110% of design pressure.

The maximum transient pressure allowable in the reactor coolant system piping, valves and fittings under USAS Section 531.1 is 120% of design j

pressure. Thus, the safety limit of 2735 psig (110% of design pressure) has been established (Reference 2).

The nominal settings of the powir-operated relief valves, the reactor l

high pressure trip and the safety valves have been established to assure that the pressure never reaches the reactor coolant system pressure safety limit.

In addition, the reactor coolant system safety valves (Reference 3) are sized to prevent system pressure from exceeding the design pressure by more than 10 percent (2735 psig) in accordance with Section III of the ASME Boiler and Pressure Vessel Code, assuming complete loss of load without a direct reactor trip or any other control, except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valves settings.

As an assurance of system integrity, the reactor coolant system was hydrotested at 3107 psig prior to initial operation (Reference 4).

l References 1.

USAR, Section 4.1 2.

USAR, Section 4.1.3.1 l

3.

USAR, Section 4.4.3.2 4.

USAR, Section 4.1 J

i i'

Prairie Island Unit 1 Amendment No. 91,123 Prairie Island Unit 2 Amendment No. 84,116 i

Figura B.2.1-1 660 650

..j.

s

. s..

640 -.-..\\..

... Exit Temp 7

Limit 650 F

..i.

.s....>.

630

+

. a...

p Q 620 O

h

..i.

t 6

+j F-610 e

Locus of Reactor E5 Conditions at which j

600 the SG Safety

+

g Valves Open Ee 590 s

~

s F-e m

E 580 e

Q

~

DNB Limit 570 Region of Trip OTDT 560 1

.o.

~ Operation t

550 OPDT

. Trip 540 i

i i

i i

i I

O 10 20 30 40 50 60 70 80 delta-T (T -T,)

F n

(

Origin of Safety Limit Curves at 2235 psig with delta-T Trips and Locus of Reactor Conditions at which the SG Safety Valves Open Figure B.2.1-1 Prairie-Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

i B.2.2-1 2.2 SAFETY LIMIT VIOLATIONS l

11111 If the reactor core SAFETY LIMIT 2.1.A is violated, the requirement to go to MODE 3 places the unit in a MODE in which this FAFETY LIMIT is not applicable.

The allowed completion time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing the unit to a MODE of operation where this SAFETY LIMIT is not applicable, and reduces the probability of fuel damage.

If the Reactor Coolant System pressure SAFETY LIMIT 2.1.5 is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Exceeding the Reactor Coolant System pressure SAFETY LIMIT may cause immediate Reactor Coolant System failure and create a potential for radioactive releases in excess of 10CFR100, " Reactor Site Criteria", limits.

The allowable completion time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

If the Reactor Coolant System pressure SAFETY LIMIT 2.1.5 is exceeded in MODE 3, 4, or 5, Reactor Coolant System pressure must be restored to within the SAFETY LIMIT value within 5 minutes. Exceeding the Reactor Coolant System pressure SAFETY LIMIT in MODE 3, 4, or 5 is more severe than exceeding this SAFETY LIMIT in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SAFETY LIMIT within 5 minutes. The action does not require reducing MODES, sic:

this would require reducing 4

temperature, which would compound the roblem by adding thermal gradient stresses to the existing pressure stre 5.

If either SAFETY LIMIT in 2.1. A or 2.1.5 is violated, the NRC Operations Center must be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in accordance with 10CFR50.72.

l If either SAFETY LIMIT in 2.1.A or 2.1.5 is violated, the Vice President Nuclear Generation shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provides time for the plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to senior management.

If either SAFETY LIMIT in 2.1.A or 2.1.8 is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC and the Vice President Nuclear Generation. This requirement is in accordance with 10CFR50.73.

If either SAFETY LIMIT in 2.1.A or 2.1.5 is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation.

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

B.3.1-2 3.1 REACTOR COOIANT SYSTEM Annan continued A.

Operational Components (continued)

Reactor coolant pump start is restricted to RCS conditions where there is pressurizar level indication or low differential temperature across the SG tubee to reduce the probability of positive pressure surges causing overpressurization.

The pressurizer is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated transients.

Each of the pressurizer safety valves is designed to relieve 325,000 lbs per hour of saturated steam at the valve set point.

These valves are considered OPERABLE at i 34 of their setpoint of 2485 l

psig. Following testing the valve lift settings are restored within a nominal i 14 of their setpoint. Below 350*F and 450 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure. If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequate defense against over-pressurization of the reactor coolant system for reactor coolant.

temperatures less than 350*F.

The combined capacity of both safety valves is greater than the maximum surge rate resulting from complete loss of load (Reference 1).

The requirement that two groups of pressurizar heaters be OPERABLE provides assurance that at least one group will be available during a loss of offsite power to maintain natural circulation.

Backup heater group "A" is normally supplied by one safeguards bus. Backup heater group "B" can be manually transferred within minutes to the redundant safeguards bus. Tests have confirmed the ability of either group to maintain natural circulation conditions.

The pressurizer power operated relief valves (PORVs) operate to. relieve reactor coolant system pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The PORVs are pneumatic valves operated by instru-ment air. They fail closed on loss of air or loss of power to their DC solenoid valves.

The PORV block valves are motor operated valves supplied by the 480 volt safeguards buses.

The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:

Manual control of PORVs to control reactor coolant pressure. This is a.

a function that is used for the steam generator tube rupture accident and for plant shutdown.

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

.c.

B.3.1-3 3.1 REACTOR.C00iANT SYSTEM 11111 continued A.

Operational Components (continued) b.

Maintaining the integrity of the reactor coolant pressure boundary.

This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage, c.

Manual control of the block valve to:

(1) unblock an icolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item a above), and (2) isolate a PORV with excessive seat leakage (Item b. above).

d.

Manual control of a block valve to isolate a stuck-open PORV.

The OPERABILITY of two PORVs or an RCS vent opening of at least 3 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the RCS temperature is less than 310*F*.

The PORV control switches are three position switches, Open Auto-Close. A PORV is placed in manual control by placing its control switch in the closed position.

The minimum pressurization temperature (310*F*) is determined from Figure TS.3.1-1 and is the temperature equivalent to the RCS safety relief valve setpoint pressure.

The RCS safety valves and normal setpoints on the pressurizer PORV's do not provide overpressure protection for certain low temperature operational transients.

Inadvertent pressurization of the RCS at temperatures below 310*F* could result in the limits of Figures TS.3.1-1 and TS.3.1-2 being exceeded. Thus the low temperature overpressure protection system, which is designed to prevent pressurizing the RCS above the pressure limits specified in Figures TS.3.1-1 and TS.3.1-2, is enabled at 310*F*.

Above 310*F* the RCS safety valves would limit the pressure increase and would prevent the limits of Figures TS.3.1-1 and TS.3.1-2 from being exceeded.

The setpoint for the low temperature overpressure protection system is derived by analysis which models the performance of the low temperature overpressure protection system assuming various mass input and heat input transients. The low temperature overpressure protection system setpoint is updated whenever the RCS heatup and cooldown curves (Figures TS.3.1-1 and TS.3.1-2) are revised.

The 3 square inch RCS vent opening is based on the 2.956 square inch cross sectional flow area of a pressurizer PORV. Because the RCS vent opening specification is based on the flow capacity of a PORV, a PORV maintained in the open position may be utilized to meet the RCS vent requirements.

• Valid until 20 EFPY Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

=

B.3.4-1 3.4 jlEAM AND POWER CONVERSION SYSTEMS Bases A reactor shutdown from power requires removal of decay heat. Decay heat removal requirements are normally satisfied by the steam bypass to the condenser and by continued feedwater flow to the steam generators. Normal feedwater flow to the steam generators is provided by operation of the turbine-cycle feedwater system.

The ten steam generator safety valves have a total combined rated capability of 7,745,000 lbs/hr.

The total full power steam flow is 7,094,000 lbs/hr; therefore, the ten steam generator safety valves will be able to relieve the total steam flow if necessary (Reference 1).These valves are considered OPERABLE at 34 of their specified setpoint. Following testing the valve lift settings are restored within a nominal 14 of their setpoint.

In the unlikely event of complete loss of offsite electrical power to either or both reactors, continued removal of decay heat would be assured by availability of either the steam-driven auxiliary feedwater pump or the motor-driven auxiliary feedwater pump associated with each reactor, and by steam discharge to the atmosphere through the steam generator safety valves.

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

During STARTUP OPERATIONS, the Auxiliary Feedwater motor-operated injection valves maybe less than full open as necessary to faciliate plant startup.

The minimum amount of water specified for the condensate storage tanks is sufficient to remove the decay heat generated by one reactor in the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of shutdown.

Essentially unlimited replenishment of the condensate storage supply is available from the intake structures through the cooling water system.

The two steam generator power-operated relief valves located upstream of the main steam isolation valves are required to remove decay heat and cool the reactor down following a steam generator tube rupture event (Reference 3) and following a high energy line rupture outside containment (Reference 2).

The steam generator power operated relief valves are provided with manual upstream block valves to permit testing at power and to provide a means of isolation.

In order to assure timely response to a steam generator tube rupture event, a

steam generator power operated relief valve is considered operable when it is capable of being remotely operated and when its associated block valve is open.

Isolation dampers are required in ventilation ducts that penetrate those rooms containing equipment needed for a high energy line rupture outside containment.

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

i l

l

=

i 1

l 1

l l

l B.3.4-2 The limitations on secondary system specific activity ensure that the resultant l

off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses.

References l

1. USAR,,Section 11.9.4

2. USAR, Appendix I

3. USAR, Section 14.5.4 l

l i

l l

I l

i I

l Prairie Island Unit 1 Amendment No. 91,97,123 Prairie Island Unit 2 Amendment No. 84,90,116

1 curo lt UNITED STATES y

j j

NUCLEAR REGULATORY COMMISSION I E WASHINGTON, D.C. 20086 0001

%...../

NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 116 License No. DPR-60 1.

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

A.

The application for amendment by Northern States Power Company (the licensee) dated May 4. 1995, as supplemented November 27, 1995, and March 1, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the 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 Comission'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-60 is hereby amended to read as follows:

i

. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment N9.116, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license arandment is effective as of the date of issuance, with full implementation within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Ja 64g Bei.h A. Wetzel, Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 21, 1996 l

l i

l i

1 ATTACHMENT TO LICENSE AMENDMENT N0.116 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 j

Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by i

amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT TS-i TS-1 TS-viii TS-viii TS-x TS-x TS-xiii TS-xiii TS.2.1-1 TS.2.1-1 TS.2.2-1 Figure TS.2.1-1 Figure TS.2.1-1 TS.2.3-2 TS.2.3-2 TS.2.3-3 TS.2.3-3

~

TS.3.4-1 TS.3.4-1 Table TS.4.1-2A Table TS.4.1-2A (Page 1 of 2)

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

TS.6.4-1 B.2.1-1 B.2.1-1 B.2.1-2 B.2.1-2 B.2.1-3 B.2.1-4 B.2.1-5 Figure B.2.1-1 B.2.2-1 B.2.2-1 B.3.1-2 B.3.1-2 B.3.1-3 B.3.1-3 B.3.4-1 B.3.4-1 B.3.4-2

1 TS-i TECHNICAL SPECIFICATIONS TABLE OF CONTENTS TS SECTION M

PACE 1.0 DEFINITIONS TS.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING TS.2.1-1 2.1 Safety Limits TS.2.1-1 2.2 Safety Limit Violations TS.2.1-1 2.3 Limitinr; Safety System Settings, Protective Instrwantation TS.2.3-1 A. Protective Instrumentation Settings for Reactor Trty TS.2.3-1 S. Protective Instrumentation Settings for Reactor Trip Interlocks TS.2.3-4 C. Control Rod Withdrawal Stops TS.2.3-4 Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116 4

TSoviii TABLE W C5nin.nls (Contimiad)

TS SECTIGI III}d FACE 6.0 ADMINISTRATIVE CONTROLS TS.6.1 1 6.1 Organizaties TS.6.1-1 6.2 Review and Audit TS.6.2-1 A. Safety Audit Committee (SAC)

TS.6.2-1

1. Membership TS.6.2 1

2. Qualifications TS.6.2-1

3. Neeting Frequency 75.6.2-2

4. Quorum TS.6.2-2

5. Responsibilities TS.6.2-2

6. Audit TS.6.2-3

7. Authority TS.6.'2-4

8. Records TS.6.2-4

9. Procedures TS.6.2-4

3. Operatises Committee (OC)

TS.6.2-5

1. Membership TS.6.2-5

2. Meeting Frequency TS.6.2-5

3. Quorian T5.6.2-5

4. Responsibilities TS.6.2 5

5. Authority TS.6.2-6

6. Records TS.6.2 6

7. Procedures TS.6.2-6 C. Maintenance Procedures TS.6.2-7 6.3 Special Inspections and Audits TS.6.3-1 6.4 Deleted 6.5 Plant operating Procedures TS.6.5-1 A. Plant Operations TS.6.5 1

3. Radiological TS.6.5-1 C. Maintenanee and Test TS.6.5-4 D. Deleted E. Offsite Dese Calculation Manual (0DCN)

TS.6.5-4 F. Security TS 6.5-5 C. Temporary Changes to Procedures TS.6.5-5 M. Radioactive Effluent Controls Program TS.6.5-6

2. Explosive cas and Storage Tank Monitoring Program TS.6.5-7 6.6 Plant Operatisg Records TS.6.6-1 A. Records Retained for Five Tears TS.6.6 1

3. Records Retained for the Life of the Plant T5.6.6-1 Prairie Island Unit 1 Amendment No. 91.96,Iff.123 Prairie Island Unit 2 Amendment No. 84,89,11$,116

3 TS-x i

j TABLE OF CONTENTS (continued)

TS BASES SECTION TITLE FAfrI

)

2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits B.2.1-1 A.

Reactor Core Safety Limits B.2.1-1 B.

Reactor Coolant System Pressure Safety Limits B.2.1-5 2.2 Safety Limit Violations B.2.2-1 2.3 Limiting Safety System Settings, Protective B.2.3-1 4

Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION q

3.0 Applicability B.3.0-1 l

3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 E. Maximum Reactor Coolant Oxygen, Chloride B.3.1-8 and Fluoride Concentration F. Isothermal Temperature Coefficient (ITC)

B.3.1-9 3.2 Chemical and Volume. control System B.3.2-1 3.3 Engineered Safety Features B.3.3-1 3.4 Steam and Power Conversion Systems B.3.4 1 3.5 Instrumentation System B.3.5-1 3.6 Containment System B.3.6-1 3.7 Auxiliary Electrical System B.3.7-1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10-9

)

F. Inoperable Rod Position Indicator Channels B.3.10-9 1

C. Control Rod Operability Limitations B.3.10-9 i

H. Rod Drop Time B.3.10-10 I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 i

3.11 Core Surveillance Instrumentation B.3.11-1 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13-1 3.14 Deleted 3.15 Event Monitoring Instrumentation B.3.15-1 l

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment.:No. 116 l

TSoziii APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURE $

TS FICURE TITLE 2.1-1 Reactor Core Safety Limits 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE IQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL "WER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131

~3.5-1 Spent Fuel Fool Unrestricted Region Minimum Burnup Requirements 3.10-1 Required shutdown Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In 14akage Rate 5.6-1 Spent Fuel Fool Burned / Fresh Checkerboard Cell Layout 5.6-2 spent Fuel Fool Checkerboard Region Miniaua Surnup Requirements B.2.1-1 Origin of Safety Limit Curves at 2235 psigwith delta-T Trips and Iocus of Reactor Conditions at which SG Safety Valves Open Prairie Island Unit 1 Amendment No. 105,108,122,123 Prairie Island Unit 2 Amendment No. 98,10f,125,116

V TS.2.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 2.1 SAFETY LIMITS A.

Reactor Core Safety tf=fts In MODES 1 and 2, combination of thermal power (measured in 4T), pressurizer pressure, and the highest reactor coolant system l

loop average temperature shall not exceed the limits shown in I

Figure TS.2.1-1.

B.

Reactor Coolant System Pressure Safety Limit In MODES 1, 2, 3, 4, and 5, the reactor coolant system pressure shall not exceed 2735 psig.

I 2.2 SAFETY LIMIT VIDIATIONS A.

If SAFETY LIMIT 2.1. A. is violated, restore c:tpliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B.

If SAFETY LIMIT 2.1.B. is viplated:

1. In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2. In MODE 3, 4, or 5, restore compliance within 5 minutes.

C.

If a SAFETY LIMIT is violated, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notify the NRC Operetions Center in accordance with 10CFR50.72.

D.

If a SAFETY LIMIT is violated, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notify the Vice President Nuclear Generation, and the Chairman of the Safety Audit Committee or their designated alternates.

E.

If a SAFETY LIMIT is violated, within 30 days a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73. The LER shall be submitted to the NRC, the Vice President Nuclear Generation and the Safety Audit Committee.

F.

If a SAFETY LIMIT is violated, operation of the unit shall not be resumed until authorized by the NRC.

Prairie Island Unit 1 Amendment No. 77,91,105,123 Prairie Island Unit 2 Amendment No. 70,84,95,116

Figura TS.2.1-1 660 650

-p s

s 640 630 i.

s-p.

Q 620 g.

F

+

H' 610

~

e slii 600 u

j.e. -.

h 590 F-g E

580 2385 psig e

/

I

[

2235 570 560

- 1985 8

100% Flow (68.2 x 10 lb/hr) 11885l 550 117851 i

i i

i i

i i

i i

540 0

10 20 30 40 50 60 70 80 delta-T (T J,) T n

Reactor Core Safety Limits Figure TS.2.1-1 Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

~

4 I

TS.2.3-2 a

1 2.3.A.2.d Cont.

and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chamber, with gains to be selected based on measured instrument response during plant startup tests, such that where qs and q are the percent power in the top and bottom halves of the core, respectively, and qs + gn is total core power in percent of rated power:

1.

for qs - q within -124 and +96, f (AI) = 0, and l

2.

for each Forcent that the magnitude of qs - gn exceeds

+94 the A't trip set point shall be auto 6atically reduced by an equivalent of 2.5 percent of RATED THERMAL POWER.

l 3.

for each percent that the magnitude of qs - gn exceeds

-124, the T trip set point shall be automatically reduced by an equivalent of 1.5 percent of RATED THERMAL POWER.

4 i

e.

Overpower A T i

K t sT 33 ar s are tr4 -

- r6(f*T') ~ f W 3 3 p

1 + t s-3 where AT.

Indicated AT at RATED THERMAL POWER T

Average temperature, 'F T'

567.3*F K.

s 1.10 K

0.0275 for increasing T; O for decreasing T 3

Ke 0.002 for T > T', O for T < T' 10 see t

f(AI) - as defined in d. above f.

Low reactor coolant flow per loop - 290% of normal indicated loop flow as measured at loop elbow tap.

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

\\

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 1.

Power range neutron flux rate.

1.

Positive rate - 515% of RATED THERMAL POWER with a time constant 22 seconds 2.

Negative rate - 57% of RATED THERNAL POWER with a time constant of 22 seconds

3. Other reactor trips a.

High pressurizer water level - 590% of narrow range instre-3nt span.

b.

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

Turbine Generator trip 1.

Turbine stop valye indicators - closed 2.

Low auto stop oil pressure - 245 psig i

d.

Safety injection - See Specification 3.5 Prairie Island Unit 1 Amendment No. 77, 77, 777,123 Prairie Island Unit 2 Amendment No. 77, 77, 777,116

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

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

Snecification A.

Steam Generator Safety and Power Onorated Relief V* vaa l

1.

A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 350*F unless the following conditions are satisfied (except as specified in 3.4.A.2 below):

a. Ten steam generator safety valves shall be OPERABLE with lift settings of 1077, 1093, 1110, 1120 and 1131 psig i 3e except during testing.

l

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

2.

During STARTUP OPERATION or POWER OPERATION, the following condition of inoperability may exist provided STARTUP OPERATION is discontinued until OPERABILITY is restored.

If OPERABILITY is not restored within the time specified, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce reactor coolant system average temperatius below 350'F within the j

following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

a. One steam generator power-operated relief valve may be inoperable

{

for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

B.

Auxiliary Feedwater Systen j

1.

A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 350*F unless the following conditions are satisfied (except as specified in 3.4.5.2 below):

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

l b.

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

c. Valves and piping associated with the above components are OPERABLE except that during STARTUP OPERATION necessary changes may be made in motor-operated valve position. All such changes shall be under direct administrative control.

]

Prairie Island Unit 1 Amendment No. 123 i

Prairie Island Unit 2 Amendment No. 116

C 4

l

,,22 Table TS.4.1-2A (Page 1 of 2)

XX E#

sw MINIMUM FREOUENCIES FOR EQUIPMENT TESTS EE t

FSAR Sect.

Eautoment Inst Freauency Reference l

o. a.

1. Control Rod Assemblies Rod Drop Times of full length All rods during each refueling shutdown or 7

??

rods following each removal of the reactor vessel head; affected rods following maintenance w,_.

on or modification to the control rod drive l

system which could affect performance of those specified rods

2. Control Rod Assemblies Partial movement of all rods Every Quarter 7

3. Pressurizer Safety Verify OPERABLE in accordance Per ASME Code,Section XI Inservice Testing Valves with the Inservice Testing Program Program (1 34). Following i

testing, lift settings shall be within ilt

4. Main Steam Safety Verify each required lift Per ASME Code,Section II Inservice Testing Valves setpoint in accordance with Program I

the Inservice Testing Program (i 34). Following testing, lift settings shall be within ilt

5. Reactor Cavity Water Imvel Prior to moving fuel assemblies or control E~

rods and at least once every day while the cavity is flooded.

"d S, b

6. Pressurizer PORY Functional Quarterly, unless the block valve has been "g

$S Block Valves ue closed per Specification 3.1.A.2.c.(1).(b).2 or 3.1.A.2.c.(1).(b).3.

oo

((

7. Pressurizer PORVs Functional Every 18 months 5' tY

'xs 'xs l

22 Table TS.4.1-2A (Page 2 of 2) g 44 XX sw MINIMUM REOUENCIES FOR EQUIPMENT TESTS ED

@B na s'ece.

~

Eautoment Ies_t Frecuency Be.ference l vn

1. 4

8. Deleted ww

9. Primary System Imakage Evaluate Daily 4

10. Deleted

11. Turbine stop valves, Manettonal Turbine stop valves, governor valves and 10 governor valves, and intercept valves are to be tested at a intercept valves, frequency consistent with the methodology (Part of turbine presented in UCAP-11525 "Probabilistic overspeed protection)

Evaluaicion of Reduction in Turbine Valve test Frequency", and in accordance with the established NRC acceptance criteria for the probability of a turbine missle ejection incident of 1.0x10-5 per year. In no case shall the turbine valve test interval exceed one year.

QY A. 5 oP

a. a.

me na y

? ?.

W&

Ch W

a B.2.1-1 2.1 SAFETY LIMITC A. Reactor Core Safety Limits I1111 To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions. This is accomplished by operating the hot regions of the core within the nucleate boiling regime of heat transfer wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation. Therefore, the observable parameters; thermal power, reactor coolant temperature and pressure have been related to DNB through the W-3 and WRB 1 DNB correlations. The W-3 DNB correlation is used for Exxon fuel. The VRB-1 DNB correlation is used for Westinghouse fuel.

The W-3 and WRB-1 DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNB ratio, DNBR, during steady state operation, normal operational transients, and anticipated transients is limited to 1.30 for the Exxon fuel using the W-3 correlation and to 1.17 for the Westinghouse fuel using the WRB-1 correlation. There is a third DNBR limit specifically for the steam line break accident but it does not apply to the safety limit curve calculations.

These limits correspond to a 954 probability at a 954 confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The safety limit curves of Figure TS.2.1-1 define the regions of acceptable operation with respect to average temperatures, power, and pressurizar pressure. These boundaries of acceptable operations are limited by the thermal overpower limit (fuel melting), thermal overtssperature limit (cladding damage based on DNB considerations), and the locus of points where the steam generation safety valves open. These limits are used to set the overpower and overtemperature AT trip setpoints.

The safety limit curves of Figure TS.2.1-1 comprise the most limiting of the following four criteria:

1) Vessel Exit Temperature < 650*F This is the design temperature limit. This limit defines the portion of the safety limit curves from 0 AT to the first knee for the 2235 and 2385 psig curves. At these pressures, the temperature limit of 650' is more Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

-s i

C B.2.1-2 A.

Reactor Core Safety Limirs Bases continued limiting than the T..s limit. The locus of points is calculated from a heat balance with the minimum RCS flow specified in TS 3.10.J.

2) Vessel Exit Temperature < T..s This limit ensures that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated water which ensures that the AT measured by instrumentation used by the RPS as a measure of core thermal power is proportional to core power. This limit defines the portion of the safety limit curves from 0 AT to the first knee for the 1985, 1885 and 1785 peig curves. The locus of points is calculated from a heat balance with the minimum RCS flow specified in TS 3.10.J.

j

3) Minimum DNBR > 1.3 or 1.17 whichever is applicable As mentioned before, 1.3 is the DNBR limit for Exxon fuel using the W-3 critical heat flux correlation and 1.17 is the DNBR limit for Westinghouse fuel using the WRB-1 critical heat flux correlation.

The locus of points past the first knee at all pzessures represents the thermal hydraulic conditions above which the hot channel has a DNBR less than the limit. The conditions are evaluated using approved DNB methodology. The assumptions used in the calculation include a bypass flow of 64, an Fis greater than 1.75, and a rod bow penalty of 2.64.

The very shallow knee at full power AT occurs because the Fin (hot channel pcver) is allowed to increase for core power less than RATED THERMAL POWER as described in TS 3.10.B.1.

4) Hot Channel Exit Quality < 15% or 304 whichever is applicable This limit is typically not the most restrictive because it is generally approached at lower powers where the T us < T..s or 650*F is more limiting. However, it is considered when the DNB calculations described above are performed using approved DNB methodology. This limit is determined by the range of the channel exit quality for the critical heat flux correlations. The maximum channel exit quality limit is 154 for the V-3 correlation and 304 for the VRB-1 correlation.

Operation above the safety limit curves of Figure TS 2.1-1 is not acceptable.

At each pressure the safety limit curve is the most restrictive combination i

of the four limits discussed above. The area of acceptable operation below l

the safety limit curves is bounded by the OTAT trip, the OPAT trip, and the locus of points where the steam generator (main steam) safety valves open.

The AT trips are set conservatively with respect to the safety limit curves to protect the core from exceeding the safety limits. The locus of points at which the steam generator safety valves open defines the thermodynamic limit of temperature conditions in the RCS based on the maximum pressure in the steam generators. For this calculation, it is assumed that the pressure in the steam generator is 1195 psig which is 110% of design pressure.

It is Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

7

=

5.2.1-3 A.

Reactor Core Safety Limits

}g3Ag continued required that the steam generator safety valves protect the pressure from exceeding 110% of design pressure so using 1195 psig in the calculations is conservative.

Thus, the reactor is protected from violating the safsty limits by the physical limit of the AT trips and the opening of the steam generator safety valves.

l As an example, all the limits for the 2235 pais curve are plotted in Figure B.2.1-1 along with the AT trips and the locus of points where the steam generator safety valves open. This plot demonstrates that the AT trips and the steam generator safety valves do protect the reactor from exceeding the safety limits. Note, however, that the OTAT trip locus on that plot is for steady state conditions and that the locus will drop in response to the rate at which the AT is increasing.

In addition, f(AI) increasing will also lower the OTAT trip locus.

The safety limit curves are plotted with AT on the x-axis for the following two reasons:

1.) the full power AT is different at different temperatures and pressures because water properties are nonlinear. This makes it difficult to plot the curves at each pressure using the same scale for the percent power axis.

2.) the AT trip setpoints which the reactor protection i

system actually calculates is based on the AT, not the percent power.

I 1

Except for special tests, POWER OPERATION with only one loop or with rstural circulation is not allowed. Safety limits for such special tests will be determined as a part of the test procedure.

The curves are conservative for the following nuclear hot channel factors:

P'

- @" [1 + PFDH(1-P)] ; and 78 - 7""

as as e

o where:

- P'" is the To limit at RATED THERMAL POWER specified in the CORE OPbTINCLIMITSREPORT.

- 7"" is the Fas limit at RATED THERMAL POWER specified in the CORE as OPERATING LIMITS REPORT.

- PFDH is the Power Factor Multiplier for 78 specified in the CORE g

OPERATING LIMITS REPORT Use of these factors results in more conservative safety limits than would result from power distribution limits in Specification TS.3.10.

I i

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116 i

l B.2.1 4 A.

Reactor Core Safety Limits Bases continued This combination of hot channel factors is higher than that calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. The control rod insertion lie'.ts are covered by Specification 3.10.

Adverse power distribution f ctors could occur at lower power levels because additional control rods are in the core. However, the control rod insert 19u Itaits specified in the CORE OPERATING LIMITS REPORT assure that the DNB ratio is always greater at part power than at full power.

j i

The Reactor Control and Protective System is designed to prevent any anticipated combination of transient conditions that would result in a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 for Westinghouse fuel.

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

B.2.1-5 l

l 3.

33uttpor coolant System Pressure Safety Limit Bases The reactor coolant system (Reference 1) serves as a barrier preventing

)

radionuclides contained in the reactor coolant from reaching the atmos-phere.

In the event of a fuel cladding failure the reactor coolant l

system is the primary barrier against the release of fission products.

By establishing a system pressure limit, the continued integrity of the reactor coolant system is assured. The maximum transient pressure allowable in the reactor coolant system pressure vessel under the ASME Code,Section III is 1106 of design pressure.

The maximum transient pressure allowable in the reactor coolant system piping, valves and fittings under USAS Section 531.1 is 120% of design pressure. Thus, the safety limit of 2735 psig (110% of design pressure) has been established (Reference 2).

{

The moninal settings of the power-operated relief valves, the reactor hi h pressure trip and the safety valves have been established to assure 8

that the pressure never reaches the reactor coolant system pressure safety limit.

In addition, the reactor coolant system safety valves (Reference 3) are sized to prevent system pressure from exceeding the design pressure by more than 10 percent (2735 psig) in accordance with Section III of the ASME Boiler and Pressure Vessel Code, assuming complete loss of load without a direct reactor trip or any other control, except that the safety valves on the secondary plant are assumed to open when the eteam pressure reaches the secondary plant safety valves settings.

As an assurance of system integrity, the reactor coolant system was hydrotested at 3107 psig prior to initial operation (Reference 4).

References 1.

USAR, Section 4.1 2.

USAR, Section 4.1.3.1 3.

USAR, Section 4.4.3.2 4.

USAR, Section 4.1 i

i Prairie Island Unit 1 Amendment No. 91,123 Prairie Island Unit 2 Amendment No. 84,116

y Figura B.2.1-1 660 650

..r.

..g.

640 Exit Temp Limit 650 *F 630

5... -i -

p.

Q 620 U

..j.

. {..

i F'.610 e)

Locus of Reactor s

Conditions at which j

600 the SG Safety g

Valves Open b

590 H

e O) a 580 - -.

ID

+. DNB Limit 570 -

Region of OTDT 560 -. -

.t.

' Operation Trip 550

~ OPDT

. Trip 540 i

i i

i i

i I

O 10 20 30 40 50 60 70 80 delta-T (T -T,)

F 3

Origin of Safety Limit Curves at 2235 psig with detta-T Trips and Locus of Reactor Conditions at which the SG Safety Valves Open Figure B.2.1-1 Prair1e-Island Uriit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

2 B.2.2-1 2.2 SAFETY LIMIT VIO1ATIONS BA181 If the reactor core SAFETY LIMIT 2.1.A is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SAFETY LIMIT is not applicable.

The allowed completion time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing the unit to a MODE of operation where this SAFETY LIMIT is not applicable, and reduces the probability of fuel damage.

If the Reactor Coolant System pressure SAFETY LIMIT 2.1.5 is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Exceeding the Reactor Coolant System pressure SAFETY LIMIT may cause immediate Reactor Coolant System failure and create a potential for radioactive releases in excess of 10CFR100, " Reactor Site Criteria", limits.

The allowable completion time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

If the Reactor Coolant System pressure SAFETY LIMIT 2.1.5 is exceeded in MODE 3, 4, or 5, Reactor Coolant System pressure must be restored to within the SAFETY LIMIT value within 5 minutes. Exceeding the Reactor Coolant System pressure SAFETY LIMIT in MODE 3, 4, or 5 is more severe than exceeding this SAFETY LIMIT in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SAFETY LIMIT within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

If either SAFETY LIMIT in 2.1.A or 2.1.5 is violated, the NRC Operations Center must be motified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in accordance with 10CFR50.72.

If either SAFETY LIMIT in 2.1.A or 2.1.5 is violated, the Vice President Nuclear Generation shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provides time for the plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to senior management.

I If either SAFETY LIMIT in 2.1.A or 2.1.5 is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC and the Vice President Nuclear Generation. This requirement is in accordance with 10CFR50.73.

If either SAFETY LIMIT in 2.1.A or 2.1.5 is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation.

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

n-

~~

u B.3.1 2 3.1 REACTOR C001 ANT SYSTFM 3333g continued A.

Operational Components (continued) i Reactor coolant pump start is restricted to RCS conditions where there is pressurizer level indication or low differential temperature across i

the SC tubes to reduce the probability of positive pressure surges causing i

overpressurization.

The pressurizer is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated transients.

Each of the pressurizer safety valves is designed to relieve 325,000 lbs per hour of saturated steem at the valve set point.

These valves are considered OPERABLE at i 34 of their setpoint of 2485

)

psig. Following testing the valve lift settings are restored within a nominal i 16 of their setpoint. Below 350*F and 450 psig in the reactor coolant system, the residual heat removal system can remove decay heat and j

thereby control system temperature and pressure. If no residual heat were removed by any of the means available, the unount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequate defense against over pressurization of the reactor coolant system for reactor coolant l

temperatures less than 350*F.

The combined capacity of both safety valves is greater than the maximum surge rate resulting from complete loss of load (Reference 1).

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

The pressurizer power operated relief valves (PORVs) operate to relieve reactor coolant system pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The PORVs are pneumatic valves operated by instru-ment air. They fail closed on loss of air or loss of power to their DC solenoid valves.

The PORV block valves are motor operated valves supplied by the 480 volt safeguards buses.

The OPERABILITY of the FORVs and block valves is determined on the basis of their being capable of performing the following functions:

a.

Manual control of PORVs to control reactor coolant pressure. This is a function that is used for the steam generator tube rupture accident and for plant shutdown.

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116 i

1 8.3.1-3' 3.1 REACTOR C001 ANT SYSTEM R1111 continued A.

Operational Components (continued) b.

Maintaining the integrity of the reactor coolant pressure boundary.

This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage, c.

Manual control of the block valve to:

(1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system i

pressure (Item a above), and (2) isolate a PORV with excessive seat leakage (Ices b. above).

d.

Manual control of a block valve to isolate a stuck-open PORV.

The OPERABILITY of two PORVs or an RCS vent opening of at least 3 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix C to 10 CFR Part 50 when the RCS temperature is less than 310'F*.

The PORV control switches are three position switches, Open-Auto-Close. A PORV is placed in manual control by placing its control switch in the Closed position.

The minimum pressurization temperature (310*F*) is determined from Figure TS.3.1-1 and is the temperature equivalent to the RCS safety relief valve setpoint pressure. The RCS safety valves and normal setpoints on the pressurizer PORV's do not provide overpressure protection for certain low temperature operational transients. Inadvertent pressurization of the RCS at temperatures below 310*F* could result in the limits of Figures TS.3.1-1 and TS.3.1-2 being exceeded. Thus the low temperature overpressure protection system, which is designed to prevent pressurizing the RCS above the pressure limits specified in Figures TS.3.1-1 and TS.3.1 2, is enabled at 310*?*.

Above 310*F* the RCS safety valves would limit the pressure increase and would prevent the limits of Figures TS.3.1-1 and 7s.3.1-2 from being exceeded.

The setpoint for the low temperature overpressure protection system is derived by analysis which models the performance of the low t m perature overpressure protection system assuming various mass input and heat input transients. The low temperature overpressure protection syst== eetpoint is updated whenever the RCS heatup and cooldown curves (Figures TS.3.1-1 and TS.3.1-2) are revised.

The 3 square inch RCS vent opening is based on the 2.956 square inch cross sectional flow area of a pressurizer PORV.

Because the RCS vent opening specification is based on the flow capacity of a PORV, a PORV maintained in the open position may be utilized to meet the RCS vent requirements.

• Valid until 20 RFFY Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

n --

i B.3.4-1 3.4 STEAM AND POWER CONVEILSION SYSTEMS Asana A reactor shutdown from power requires removal of decay heat. Decay heat removal requirements are normally satisfied by the steam bypass to the condenser and by continued feedwater flow to the steam generators. Normal feedwater flow to the steam generators is provided by operation of the turbine-cycle feedwater system.

The ten steam generator safety valves have a total combined rated capability of 7,745,000 lbs/hr. The total full power steam flow is 7,094,000 lbs/hr; therefore, the ten steam generator safety valves will be able to relieve the total steam flow if necessary (Reference 1).These valves are considered OPERABl.E at i 34 of their specified setpoint. Following testing the valve lift settings are restored within a nominal i 14 of their setpoint.

In the unlikely event of complete loss of offsite electrical power to either or both reactors, continued removal of decay heat would be assured by availability of either the steam driven auxiliary feedwater pump or the motor-driven auxiliary feedwater pump associated with each reactor, and by steam discharge to the atmosphere through the steam generator safety valves.

• One auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from one reactor. The motor-driven auxiliary feedwater pump for each reactor can be made available to the other reactor. During STARTUP OPERATIONS, the Auxiliary Feedwater motor-operated injection valves maybe less than full open as necessary to faciliate plant startup.

The minimum amount of water specified for the condensate storage tanks is sufficient to remove the decay heat generated by one reactor in the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of shutdown.

Essentially unlimited replenishment of the condensate storage supply is available from the intake structures through the cooling water system.

The two steam generator power-operated relief valves located upstream of the main steam isolation valves are required to remove decay heat and cool the reactor down following a steam generator tube rupture event (Reference 3) and following a high energy line rupture outside containment (Reference 2).

The steam generator power operated relief valves are provided with====a1 upstream block valves to permit testing at power and to provide a means of isolation.

In order to assure timely response to a steam generator tube rupture event, a

steam generator power operated relief valve is considered operable when it is capable of being remotely operated and when its associated block valve is open.

Isolation dampers are required in ventilation ducts that penetrata those rooms containing equipment needed for a high energy line rupture outside containment.

Prairie Island Unit 1 Amendment No. 123 Prairie Island Unit 2 Amendment No. 116

B.3.4-2 l

1 The limitations on secondary system specific activity ensure that the resultant i

off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 gym primary to secondary tube leak in the steam generator of the affected steam line. These valuas are consistent with the assumptions used in the accident analyses.

References l

1. USAR, Section 11.9.4

2. USAR, Appendix I

3. USAR Section 14.5.4 l

l Prairie Island Unit 1 Amendment No. 91,97,123 Prairie Island Unit 2 Amendment No. 84,90,116 1