Text

Amends 92 & 85 to Licenses DPR-42 & DPR-60,respectively, Deleting cycle-specific Parameter Limits from Tech Specs 1.0,3.1,3.10 & 6.7 & Associated Bases & Relocating Limits in Core Operating Limits Rept
	ML20012C389
Person / Time
Site:	Prairie Island
Issue date:	03/13/1990
From:	Thoma J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20012C390	List: ML20012C389; ML20012C392; ML20012C394;
References
	NUDOCS 9003210176
	Download: ML20012C389 (52)
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.j NUCLEAR REGULATORY COMMISSION 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. 92 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 November 17, 1989, 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 1; 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 Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. DPR-42 is hereby amended to read as follows:

t 9003210176 900313 PDR ADOCK 05000282 p

PDC l

. Technical Specifications' The Technical Specifications contained in Appendix A, as revised through Amendment No.92, are hereby incorporated in the license The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance, FOR THE NUCLEAR REGULATORY COMMISSION

.(

W J n 0. Thoma, Acting Director Preject Directorate III-1 Division of Reactor Projects - III, IV, V & Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 13, 1990 l

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ATTACHMENT TO LICENSE AMENDMENT NO. 92 FACILITY OPERATING LICENSE NO. DPR-42 DOCKET NO. 50-212 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 marginal lines indicating-the area of change REMOVE INSERT TS-ix TS-ix TS-xlii TS-xiii TS.1-2 TS.1-2 TS 2.3-3 TS 2.3-3 TS.3.1-12 TS.3.1-12 Table TS 3.5-2 Table TS 3.5-2 TS.3.10-1 TS.3.10-1 TS.3.10-2 TS.3.10-2 TS.3.10-3 TS.3.10-3 TS.3.10 TS.3.10-4 TS.3.10-5 TS.3.10-5 TS.3.10-7 TS.3.10-7 TS.3.10-8 TS.3.10-B TS.6.7-4 TS.6.7-4 TS.6.7'S

- TS.6.7-5 TS.6.7-6 B.2.1-2 B.2.1-2 B.3.10-1 B.3.10-1 B.3.10-3 B.3.10-3 B.3.10-4 B.3.10-4 B.3.10-6 B.3.10-4 B.3.10-10 B.3.10-10

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TS ix 4

M OF CONTENTS (Continued)

TS $ECTION M

PAGE 6.7 Reporting Requirements TS.6.7 1 A. Routine Reports TS.6.7 1

1. Annual Report TS.6.7-1

a. Occupational Exposure Report-TS.6.7-1

b. Report of Safety and Relief Valve Failures and Challenges TS.6.7-1

c. Primary Coolant Iodine Spike Report TS.6.7-1

2. Startup Report TS.6.7 2

3. Monthly Operating Report TS,6.7-2

4. Semiannual Radioactive Effluent Release Report TS.6.7-3

5. Annual Summari,es of Meteorological Data TS.6.7-4

6. Core Operating Limits Report TS.6.7 4 B. Reportablo Events TS.6.7-5 L

C. Environmental Reports TS.6.7-5

1. Annual Radiation Environmental Monitoring Reports TS.6.7-5

2. Environmental Special Repotta TS.6.7 6

3. Other Environmental Reports TS.6.7-6 (non-radiological, non-aquatic)

D. Special Reports TS 6.7 6 I

Prairie Island Unit 1 - Anendment flo. 50, 59,63,73,80,91,92 Prairie Island Unit 2 - Amendment No. 64,53,59,66,,73,84,85

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TSoxiii APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FICURES TS FICURE M

2.1 1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop Operation 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 EQUIVAlINT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I 131 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents 3.10 1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4 1 Shield Building Design In Leakage Rate 6.1 1 NSP Corporate Organizational Relationship to On Site Operating Organizations 6.1 2 Prairie Island Nuclear Generating Plant Functional. Organization for On-Site Operating Group Prairie Island Unit 1 - Amendaent No. 91,92 Prairie Island Unit 2 - Amendaent Ho. 84,85 i

1 I

I

TS.I.2 CONTAINMENT INTEGRITY CONTAINMENT INTEGRITY shall exist when:

1.

Penetrations required to be isolated during accident conditions are either:

a.

Capable of being closed by an OPERABLE containment automatic isolation valve system, or b.

Closed by manual valves, blind flan 6es, or deactivated automatic valves secured in their closed positions, except as provided in Specifications 3.6.0 and 3.6.D.

2.

Blind flanges required by Table TS.4.4-1 are installed.

3.

The equipment hatch is closed and sealed.

I 4.

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

5.

The containment leakage rates are within their required limita.

COLD SHUTDOWN l-l A reactor is in the COLD SHUTDOWN condition when the reactor is subcriti-cal by at least in A kj0c and the reactor coolant average temperature is less than 200*F.

CORE ALTERATION CORE ALTERATION is the movement or manipulation of any component within l

_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.

l CORE OPERATING LIMITS REPORT l

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

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

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

l l

Prairie Island Unit 1 - Amendment No. 9,91,92 Prairie Island Unit 2 - Amendment No. 4,84,85 w--

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1 2.3.A.2.g.

Open reactor coolant pump motor breaker.

1.

Reactor coolant pump bus undervoltage -

275% of normal voltage.

2.

Reactor coolant pump bus underfrequency -

258.2 Hz h.

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 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-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 l

Prairie Island Unit 1 - Amendment No. 28,87,91,92 Prairie Island Unit 2 - Amendment No. 22,80,84,85

TS.3.1 12

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l 3.1.F.

ISOTHERMAL TEMPERATURE COEPTICIENT (ITC) 1.

When the reactor is critical, the isothermal temperature coefficient shall be less than 5 pcm/*F with all rods withdrawn, except during low power PHYSICS TESTS and as specified in 3.1.F.2 and 3.

l 2.

When the reactor is above 70 percent RATED THERMAL POWER with all rods withdrawn, the isothermal temperature coefficient shall be negative, e.xcept as specified in 3.1.F.3.

3.

If the limits of 3.1.F.1 or 2 cannot be met. POWEP OPERATION may continue provided the following actions are taken:

a. Establish and maintain control rod withdrawal limits sufficient to restore the ITC to less than the limits specified in specification 3.1.F.1 and 2 above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT SRUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addittor, to the insertion limits specified in the CORE OPERATING LIMITS REPORT.

b. Maintain the control rods within the withdrawal limits established above until a subsequent calculation verifies that the ITC has been restored to within its limit for the all rods withdrawn condition.

c. Submit a special report to the Commission within 30 days, describing the value of the measured ITC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the ITC to within its limit for the all rods withdrawn condition.

Prairie Island Unit 1 - Anendment No. 52,73,80,91,92 Prairie Island Unit 2 - Anendnent No. 46,66,73, B4,85

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

INSTRUMENT OPERATING CONDITIONS FOR REACTOR TRIP 1

2 3

4 MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF OPERABLE DECREE OF BYPASS CONDITIONS OF COLUMN FUNCTIONAL UNIT CllANNELS REDUNDANCT CONDITIONS (1) 1 OR 2 CANNOT BE MET

13. Undervoltage 4KV RCP Bus 1/ bus 1/ bus Maintain hot shutdown

14. Underfrequency 4KV Bus 1/ bus 1/ bus Maintain het shutdown

15. Control Rod Misalignment Monitor

a. Rod position deviation 1

log data required by

b. Quadrant power tilt 1

TS 3.10 I and TS 3.10 J

16. RCP Breaker Open 2

1 Maintain hot shutdown

17. Safety Injection Actuation Signal 2

1 Maintain hot shutdown

18. Automatic Trip Logic including Reactor Trip Breakers **

2 1

Notes 3, 4 Note 1: Automatic permissives not listed Note 2: When bypass condition exists, maintain normal, operation Note 3: With the number of operable channels one less than the minimum operable channels requirement, be in at least hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one 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 testing per Specification 4.1, provided the other channel is operable.

Note 4: When in the hot shutdown cmndition with the number of operable channels one less than the minimum operable channels requirement, restore the inoperable channel to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor Qp trip breakers within the next hour.

55 F.P. - Full Power n -4

- One additional channel may be taken out of service for low power physics testing o.

wy

- Includes both undervoltage and shunt trip circuits and if either circuit becomes inoperable the respective y

ni channel shall be considered inoperable.

Prairie Island Unit 1 - Amendment No. /5,82,$

Prairie Island Unit 2 - Amendment No. 68,N 0

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TS.3.10 1 j

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS 1

Annlicabiliev Applies to the limits on core fission power distribution and to the limits on control rod operations.

Obiective i

To assure 1) core suberiticality after reactor trip, 2) acceptable core power distributions during power operation, and 3) limited pottntial j

reactivity insertions caused by hypothetical control rod ejection.

i Snacification

)

A.

Shutdown Marnin i

The shutdown margin with allowance for a stuck control rod assembly j

shall exceed the applicable value shown in Figure TS.3.10 1 under all steady state operating condh ions, except for PHYSICS TESTS, from zero to full power, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at HOT SHUTDOWN conditions if all l

control rod assemblies were tripped, assuming that the highest worth l

control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron concentration, i

B.

Power Distribution Limits i

1.

At all times, except d ring I w power PHYSICS TESTING, measured hot channel factors, and H, as defined below and in the bases, shall meet the ollowing limits:

% x 1.03 x 1.05 $ (FqhP)xK(Z)

F$Hx1.04<F4H x [1+ PFDH(1 P))

where the following definitions apply:

F is the F limit at RATED THERMAL POWER specified in the CORE o

o OPERATING LIMITS REPORT.

i RTP F3g s the FAH limit at RATED THERMAL POWER specified in the CORE i

OPERATING LIMITS REPORT.

PFDHisthePowerFactorMultiplierforPfp~specifiedintheCORE OPERATING LIMITS REPORT.

n K(Z) is a normalized function that' 11mits F (z) axially as o

specified in the CORE OPERATING LIMITS REPORT.

- Z is the core height location.

1 P is the fraction o RATED THERMAL POWER at which the core is operating.

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

Prairie Island Unit 1 - Amendment No. 35,44,66,77,81,84,91,92 Prairie Island Unit 2 - Amendment No. 29,33,60,70,74,77,84,05

TS.3.10-2 FNorF[H 3.10.B.1.

wkththesmallestmarginorgreatesteheessohrespectively, is defined as the measured F or F 3 limit.

E

'- 1.03 is th engineering hot channel factor, F, app, lied to the toaccountformanufacturingtolehance.

measured Pq N

1.05 is applied to the measured Fg to account for measurement uncertainty.

N 1.04 is applied to the measured F to account for measurement uncertainty.

N 2.

Hot channel factors, Fo andF$H,shallbemeasuredandthetarget flux difference determined, at equilibrium conditions according to the following conditions, whichever occurs first:

(a) At least once po'r 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.

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

the core:

o Pq (equil) x V(Z) x 1.03 x 1.05 $ (Fq 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 f1 trip set point by 16 for each percent that the measured or by the factor specified in the CORE OgERATING IMITS REPORT for each percent that the measured F5' H exceeds the 3.10.B.1 limit. Then follow 3.10.B.3(c).

(b)Ifthemeasuredd(equil)exceedsthe3.10.B.2limitsbutnot the 3.10.B.1 limif, 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 satis-fied, or 2.

Reducereactorpowerandthehighneutronfluxthip s epoint by 14 for each percent that the measured (equil) x 1.03 x 1.05 x V(Z) exceeds the limit.

Prairie Island Unit 1 - Anendment t!o. 35,44,66,77,81,84,91,92 Prairie Island Unit 2 - Amendment flo. 29,38,60,70,74,77,84,85

TS.3.10 3 3.10.B.3 (c) If subsequent in core mapping cannot, within a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, demonstrate that the hot channel factors are met, the reactor shall be brought to a HOT SHUTDOWN condition with return to power authorized up to 50% of RATED THERMAL POWER for the purpose of PHYSICS TESTING.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above 50% of RATED ERMAL POWER. THERMAL POWER may then be increased provided P or E is demonstrated through in core AH mapping to be within its limits.

(d) If two successiv measurements indicate an increase in the peak rod power aH with exposure, either of the following actions shall be taken:

Fl05forcomparisontothelimitspecifiedin3.10B.2,or 1.

(equil) shall be multiplied by 1.02 x V(1) x 1.03 x 1.

2.

(equil) shall be measured at least once per seven e factive full power days until two successive maps N

indicate that the peak pin power PAH, is not increasing.

4 Except during PHYSICS TESTS, and except as provided by specifica.

tions 5 through 8 below, the indicated axial flux difference for at least three operable excore channels shall be maintained within the target band about the target flax difference. The target band is specified in the CORE OPERATING LIMITS REPORT.

5.

'Above 90 cereent of RATED THERMAL POWER?

If the indicated axial flux difference of two OPERABLE excore channels deviates from the target band, within 15 minutes either l

eliminate such deviation, or reduce THERMAL POWER to less than 90 percent of RATED THERMAL POWER.

6.

Between 50 and 90 nercent of RATED THERMAL POWER!

a.

The indicated axial flux difference may deviate from the l

target band for a maximum of one* hour (cumulative) in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period provided that the difference between the indicated axial flux difference about the target flux difference does not exceed the envelope specified in the CORE OPERATING LIMITS REPORT.

b.

If 6.a is violated for two OPERABLE excore channels then the THERMAL POWER shall be reduced to less than 50% of RATED THERMAL POWER and the high neutron flux setpoint reduced to

' less than 55% of RATED THERMAL POWER.

May be extended to 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months during incore/excore calibration.

Prairie Island Unit 1 - Amendment No. 35,44,91,92 Prairie Island Unit 2 - Amendment No. 29,38,84,85 4

Ts.3.10 4 3.10.B.6. c.

A power increase to a level greater than 90 percent of rated power is contingent upon the indicated axial flux difference of at least three OPERABLE excore channels being within the target band.

7.

lass than 50 careent of RATED THERMAL POWER

l a.

The i.ndicated axial flux difference may deviate from.the l

target band.

b.

A power increase to a level greater than 50 percent of RATED

]

THERMAL POWER is contingent upon the indicated axial flux difference of at least three OPERABLE excore channels not being outside the target band for more than one hour (cumula-tive) out of the preceding 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

)

8.

In applying 6a and 7b above, penalty deviations outside the l

target band shall be accumulated on a time basis of:

1 a.

One minute penalty deviation for each one minute of power operation outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and b.

One half minute penalty deviation for each one minute of power operation outside of the target band at THERMAL POWER levels between 154 and 50% of RATED THERMAL POWER.

9.

If alarms associated with monitoring the indicated axial flux difference deviations from the target band are not operable, the l

indicated axial flux difference value for each OPERABLE excore channel shall be logged at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and half hourly thereafter until the alarms are returned to an OPERABLE status.

For the purpose of applying this specifica-tion, logged values of indicated axial flux difference must be assumed to apply during the previous interval between loggings.

C. OUADRANT POWER TILT RATIO I

1.

Except for PHYSICS TESTS, if the QUADRANT POWER TILT RATIO exceeds 1.02 but is less than 1.07, the rod position indication shall be monitored and logg,ed once each shift to verify rod position within each bank assignment and, within two hours, one of the following steps shall be taken:

a.

Correct the QUADRANT POWER TILT RATIO to less than 1.02.

1 b.

Restrict core power level so as not to exceed RATED THERMAL POWER less 2% for every 0.01 that the QUADRANT POWER TILT RATIO exceeds 1.0.

l Prairie Island Unit 1 - Amendment No. 29,44,91,92 Prairie Island Unit 2 - Anendment No. 23,38,84,85

TS.3.1005 3.10.C.2.

If the QUADRANT POWER TILT RATIO exceeda 1.02 but is less than 1.07 for a sustained period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or if such a tilt recurs intermittently, the reactor shall be brought to the HOT SHUTDOWN condition.

Subsequent operation below 50% of rating, for testing, shall be piraitted.

3.

Except for PHYSICS TESTS if the QUADRANT POWER TILT RATIO exceeds t

1.07, the reactor shall be brought to the HOT SHUTDOWN condition.

Subsequent operation below 50% of rating, for testing, shall be permitted.

4.

If the core is operating above 854 power with one excore nuclear channel inoperable, then the core quadrant power balance shall be determined daily and after a 10% power change usins either 2 movable detectors or 4 core thermocouples per quadrant, per Specification 3.11.

D. Rod Insertion Limits 1.

The shutdown rods shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT when the reactor is critical or approaching criticality.

2.

When the reactor is critical or approaching criticality, the i

control banks shall be limited in physical insertion as specified I

in the CORE OPERATING LIMITS REPORT.

l 3.

Insertion limits do not apply during PHYSICS TESTS or during periodic exercise of individual rods.

The shutdown margin shown l

j in Figure TS.3.10 1 must be maintained except for low power l

PHYSICS TESTING.

For this test the reactor may be critical with l

all but one high worth full length control rod inserted for a period not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per year provided a rod drop test is run on the high worth full length rod prior to this particular low power PHYSICS TEST.

Prairie Island Unit 1 - Anendnent No. 32,44,91,92 Prairie Island Unit 2 - Amendnent No. 26,38,84,65 l

1 l

TS.3.10,7 3.10.G. Inocerable Rod Limitatione 1.

An inoperable rod is a rod which (a) does not trip (b) is i

. declared inoperable under specification 3.10.E. or 3.10.H. or (c) cannot be moved by its drive mechanism and cannot be corrected within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

~

2.

The reactor shall be brought to the HOT SHUTDOWN condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months should more than one inoperable rod be discovered during POWER OPERATION.

3.

If the inoperable rod is located below the 200 step level and is capable of being tripped, or if the rod is located below the 30 step level whether or not it is capable of being tripped, then the insertion limits specified in the CORE OPERATING LIMITS REPORT l

pply.

. 4.

If the inoperable rod cannot be located, or if the inoperable rod is located above the 30 step level and cannot be tripped, then the insertion limits specified in the CORE OPERATING LIMITS REPORT l

apply.

5.

If POWER OPERATION is continued with one inoperable rod, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days unless the rod is earlier made OPERABLE.

The analysis shall include due allowance for nonuniform fuel depletion i

in the neighborhood of the inoperable rod.

If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, THERMAL POWER shall be reduced to a level consistent with the safety analysis.

H.

Rod Dron Tina l

At operating temperature and full flow, the drop time of each RCCA shall be no greater than 1.8 seconds from loss of stationary gripper i

coil voltage to dashpot entry.

If the time is greater than 1.8 seconds, the rod shall be declared inoperable.

I l

l Prairie Island Unit 1 - Amendment No. 44,91,92 Prairie Island Unit 2 - Amendment No. 38,84,85 l

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TS.3 10 8

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i 3.10.I. Monitor Inonerabiliev'Reouirements 1.

If the rod bank insertion limit monitor is inoperable, or if the j

rod position deviation monitor is inoperable, individual rod i

positions shall be logged once per shift, after a load change greater than 10 percent of RATED THERMAL POWER, and after 30 inches or more of rod motion.

2.

If both the rod position deviation monitor and one or both of the quadrant power tilt monitors are inoperable for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or more, the nuclear overpower trip shall be reset to 93% of RATED THERMAL POWER in addition to the increased surveillance requirements.

3.

If one or both of the quadrant power tilt monitors is inoperable, individual upper and lower excore detector calibrated outputs and l

the calculated power tilt shall be logged every two hours after a r

load change greater than 10% of RATED THERMAL POWER I

J. DNB Parameters 1

I The following DNB.related parameters limits shall be maintained during POWER OPERATION:

L a.

Reactor Coolant System Tavg $564*F b.

Pressurizer Pressure 22220 psia

l c.

Reactor Coolant Flow 2the value specified in the CORE OPERATING LIMITS REPORT With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER I

to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Compliance with a and b. is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Compliance with c. is demonstrated by verifying that the parameter is I

within its limit after each refueling cycle.

1

Limit not applicable during either a THERMAL POWER ramp increase in excess of (54) RATED THERMAL POWER per minute or.a THERMAL POWER step increase in excess of (10%) RATED THERMAL POWER Prairie Island Unit 1 - Amendment No. 16,19,44,77,91,92 Prairie Island Unit 2 - Amendnent No. 10,13,38,70,84,85

TS.6.7 4 6.7.A.5. Annual S

-ries of Mateorolonical Bata An annual summary of meteorological data shall be submitted for the previous calendar year in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability at the request of the Commission.

6.7.A.6. Core Doeratine Limita Reoort

a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining.part of a reload cycle for the following:

1. Heat Flux Hot Channel Factor Limit (F

), Nuclear Enthalpy q

RTP (Specifications 3.10.B.1, 3.10.H ), PFDH, K(2) and V(2)

Rise Hot Channel Factor, Limit (FA B.2 and 3.10 B.3)

2. Axial Flux Difference Limits and Target Band (Specifications 3.10 B.4 through 3.10.B.9)

3. Shutdown and Control Bank Insertion Limits (Specification 3,10.D)

4. Reactor Coolant System Flow Limit (Specification 3.10.J)

b. The analytical me'thods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NSPNAD 8101 A, " Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version)

NSPNAD 8102 A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units" (latest approved version)

WCAP 9272 P A, Westinghouse Reload Safety Evaluation Methodology", July,1985 WCAP-10054 P A, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code". August, 1985 WCAP-10924-P A, " Westinghouse Large Break thCA Best Estimate Methodology" December, 1988 XN NF-77-57 (A), XN NF-77 57, Supplement 1 (A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II", May, 1981 c..The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

l

(

Prairie Island Unit 1 - Amendment flo. 54,59,73,91,92 Prairie Island Unit 2 - Amendment flo. 48,53,66,84,85 m.m.m a m.-

..h

.z

TS.6.7 5 f

d. The CORE OPERATING LIMITS REPORT, including any mid cycle revisions or supplements thereto, shall be supplied upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

B. REPORTABLE EVENTS The following' actions shall be taken for REPORTABLE EVENTS:

a.

The Commission shall be notified by a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the Operations Committee and the results of this review shall be submitted to the Safety Audit Committee and the Vice President Nuclear Generation.

C. Environmental Renorts The reports listed below shall be submitted to the Administrator of the appropriate Regional NRC Office or his designate:

1.

Annual Radiation Environmental Monitorine Renort (a) Annual Radiation Environmental Monitoring Reports covering the operation of the pro 5 ram during the previous calendar year shall be submitted prior to May 1 of each year.

(b) The Annual Radiation Environmental Monitoring Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

The reports shall also include the results of land use censuses required by Specification 4.10.B.1.

If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

(c) The Annual Radiation Envirenmental Monitoring Reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period.

In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted as soon as possible in a supplementary report.

Prairie Island Unit 1 -- Amendment No. 54,59,73,91,92 Prairie Island Unit 2 - Amendment No. 48,53,66,84,85

Ts.6.7 6 l.

(d) The reports shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensees participation in the Interlaboratory Comparison Program, required by Specification 4.10.C.1.

2.

Environmental Seacial Raeorts (a) When radioactivity levels in samples exceed limits specified in Table 4.10 3, an Environmental Special Report shall be submitted within 30 days from the end of the affected calendar quarter.

For certain cases involving long analysis time, determination of quarterly averages may extend beyond the 30 day period.

In these cases the potential for exceeding the quarterly limits will be reported within the 30 day period to be followed by the Environmental S'pecial Report as soon as practicable.

3.

Other Environmental Recorts (non radiolorical. non.acuatic)

Written reports for the following items shall be submitted to the appropriate NRC Regional Administrator:

a.

Environmental events that indicate or could result in a significant environa' ental impact casually related to plant operation. The following are examples:

excessive bird impaction; onsite plant or animal disease outbreaks; unusual nortality of any species protected by the Endangered Species Act of 1973; or increase in nuisance organisms or conditions. This report shall be submitted within 30 days of the event and shall (a) describei analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (b) describe the probable cause of the event. (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses, b.

Proposed changes, test or experiments which may result in a significant increase in any adverse environmental impact which was not previously reviewed or evaluated in the Final Environmental Statement or supplements thtd'eto. This report shall include an evaluation of the environmental impact of the proposed activity and shall be submitted 30 days prior to implementing the proposed change, test or experiment.

D.

Seecial Reoorts Unless otherwise indicated, special reports required by the Technical Specifications shall be submitted to the appropriate NRC Regional Administrator within the time period specified for each report.

Prairie Island Unit 1 -' Amendment No. 92 Prairie Island Unit 2 - Anendnent No. Ub

.B.2.1 2 i

1 2.1 SAFETY LIMIT. REACTOR CORE

&&111 continued power levels of 914 and 744 respectively.

For the 2235 psig and 2385 psig curves, the' coolant average temperature at the core exit is equal to 650'T below power levels of 644 and 734 respectively.

The third and fourth criteria are evaluated using standard DNB metho.

dology.

For all four curves the DNBR is limiting at higher power levels.

. The area of safe operation is below these curves.

The plant conditions requ. ired to violate the limits in the lower power range are precluded by the self actuated safety valves.on the steam generators. The highest nominal setting of the steam generator safety valves is 1129 psig (saturation temperature 560'F). At zero power the difference between primary coolant and secondary coolant is zero and at full power it is 50'F.

The reactor conditions at which steam generator safety valves open is shown as a dashed line on Figure TS.2.1 1.

,m 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 a,re conservative for the following nuclear hot channel factors:

N AT RTP F H-F

[1+PFDH(1P));andTh=F H

where:

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

RTP F

is the F limit at RATED THERMAL POWER specified in the CORE OhkRATINGLIMkSREPORT.

PFDH is the Power Factor Multiplier for PfH 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.

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 becauae 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.

Prairie Island Unit 1 - Amendnent No. 91,92 Prairie Island Unit 2 - Anendment No. 84,85

~,..

5.3.10 1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases i

Throughout the 3.10 Technical Specifications, the terms " rod (s)" and

'RCCA(s) are synoriymous.

A.

Shutdown Margi,n Trip shutdown reactivity is provided consistent with plant safety analyses assumptions._ One percent shutdown margin is adequate except for the steam break analysis, which requires more shutdown reactivity due to the more negative moderator temperature coefficient at end of life (when boron concentration is low).

Figure TS.3.10 1 is drawn j

accordingly, i

B.

Power Distribution Control The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operations) and II (Incidents of Moderate l

frequency) events by: (a) maintaining the minimum DNBR in the core of J

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 assumed design criteria.

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

One calculation at the 954 probability level was performed as well as one calculation with all the required features of 10 CFR Part 50, Appendix K.

The 954 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 LIMITS REPORT for the F limit specified in the CORE o

OPERATING LIMITS REPORT. Maintaining I) peaking factors below the Fq 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 954 probability level assures compliance with the ECCS analysis.

Duringopeqation,theplantstaffcomparesthemeasuredhotchannel factors, F'n and 'aH, (described later) to the limits determined in the transient and IDCA 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 t

corrected for engineering, calculational, and measurement uncertainties.

I F is the measured Nuclear Hot Channel Factor, defined as the maximum q

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, i

l The K(Z) function specified in the CORE OPERATING LIMITS REPORT is a normalized function that limits Fq axially. The K(Z) value is based on large and small break 1DCA analyses.

Prairie Island Unit 1 - Amendment No. 91,92 Prairie Island Unit 2 - Amendment No. 84,85 1

l

i B.3.10 3 j

3.10 CONTROL ROD AND POWER DISTRIBUTIO.4 LIMITS Bases continued WhenameasurementofFfH is taken, measurement error must be allowed for and 4 percent is the appropriate allowance for a full core map l

taken with the movable incore detector flux mapping system.

Measurements of the hot channel factors are required as part of startup l

PHYSICS TESTS, at least once each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors, The incore map taken following initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns.

The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would otherwise affect these bases.

We For normal operation, it is not necessary to measure these quantities.

Instead it has been determined that, provided certain conditions are observed, the hot channel factor limits will be met; these conditions are as follows:

1.

Control rods in a eingle bank move together with no individual rod insertion differing by more than 1$

inches from the bank demand position. An accidental misalignment limit of 13 steps precludes a rod misalign-ment greater than 15 inches with consideration of maximum instrumentation error.

2.

Control rod banks are sequenced with overlapping banks as described in Technical Specification 3.10, 3.

The control bank insertion limits specified in the CORE OPERATING LIMITS REPORT are not violated.

4 Axial power distribution control procedures, which are given in i

terms of flux difference control and control bank insertion limits are observed.

Flux difference refers to the difference in signals between the top and bottom halves of two section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

Prairie Island Unit 1 - Amendment No. 91,92 Prairie Island Unit 2 - Anendment No. 84,85

B.3.10 4 3.10 CORTROL ROD AND POWER DISTRIBUTION LIMITS

&&Ltt continued 8.

Power Distribution Control (continued)

The permitted relaxatiota in F$H and Fl allows for radial power shape changes with rod inserrion to the insertion limits.

It has been determined that provided the above conditions 1 through 4 are obse ed, these hot channel factor limits are met.

In specification 3.10, is arbitrarily limited for P less than er equal to 0.5 (except for lo power PHYSICS TESTS).

the procedures for axial power distribution control referred to above aro designed to minimize the effects of xenon redistribution on the axial power distribution during load. follow maneuvers. Basically control of flux difference is required to limit the difference between the current value of Flux Difference (AI) and a reference value which corresponds to the full power equilibrium value of Axial Offset (Axial Offset = AI/ fractional power). The reference value of flux difference varies with power level and burnup but expressed as axial offset it varies only with burnup.

The technical specifications on power distribution control assure that the F limit is not exceeded and xenon distributions are not developed q

which at a later time, would cause greater local power peaking even though the flux difference is then within the limits specified by the procedure.

The target (or reference) value of flux difference is determined as follows: At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with the full length rod control rod bank more than 190 steps withdrawn (i.e., normal full power operating position appropriate for the time in life, usually withdrawn farther as burnup proceeds).

This value, divided by the fraction of full power at which the core was operating is the full power value of the target flux difference. Values for all other core power levels are obtained by multiplying the full power value by the fractional power.

Since the indicated equilibrium was noted, no allowances for excore detector error are necessary and indicated deviation from the indicated reference value but within the target band is permitted.

The allowed deviation from the target flux difference as a function of THERMAL POWER is specified in the CORE OPERATING LIMITS REPORT.

Prairie Island Unit 1 - Amendnent No. 91,92 Prairie Island Unit 2 - Amendment No. 84.85 m ogse+

e e-e e.**

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5.3.10 6 3.10 CONTROL ROD AND POVT.R DIS 1h1BUTION LIMITS haggg continued B.

Power Distribution Control (continued)

In some instances of rapid plant power reduction, automatic rod motion will cause the flux difference to deviate from the target band when the reduced power leval is reached. This does not necessarily affect the xenon distribution sufficiently to change the anvelope of peaking factors which can be reached on a subse.

quent return to full power within the target band, however to simplify the specification, a limitation of one hour in any period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is placed on operation outside the band. This ensures that the resulting xenon distributions are not significantly different from those resulting from operation within the target band.

The consequences of being outside the target band but within the limits specified in the CORE OPERATING LIMITS 4EPORT for power levels between 50% and 904 has been evaluated and determined to result in acceptable peaking factors. Therefore, while the deviatien exists the power level is limited to 90 percent or lower depending on the indicated axial flux difference.

In all cases the target band is the Limiting Condition for Operation. Only when the target band is violated do the limits specified in the CORE OPERATING LIMITS REPORT apply.

If, for any reason, the indicated axial flux difference is not control-led within the target band for as long a period as one hour, then xenon l

distributions may be significantly changed and operation at or below 50 percent is required to protect against potentially more severe consequences of some accidents.

As discussed above, the essence of the procedure is to waintain the xenon distribution in the core as close to the equilibrium full power condition as possible. This is accomplished by using the boron system to position the full length control rods to produce the required indicated flux difference.

For Condition II events the core is protected from overpower and a minimum DNBR of 1.30 for Exxon fuel and 1.17 for Westinghouse fuel by an automatic protection system. Compliance with operating procedures is assumed as a precondition for Condition II transients, however, operat'or error and equipment malfunctions are separately assumed to lead to the cause of the transients considered.

C.

QUADRANT POWER TILT RATIO QUADRANT POWER TILT RATIO limits are based on the following considera-tions. Frequent power tilts are not anticipated durin6 normal operation since this phenomenon is caused by some asymmetric perturbation, e.g.

rod misalignment, x.y xenon transient, or inlet temperature mismatch.

A dropped or misaligned rod will easily be detected by the Rod Position Indication System or core instrumentation per Specification 3.10.F, and Prairie Island Unit 1 - Amendment flo. 91,92 Prairie Island Unit 2 - Amendment flo. 84,85

[

ip 1

8.3.10 10 1

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS AAAAA continued H.

Rod Drop Time i

The required drop time to dashpot entry is consistent with the safety analysis.

)

i I.

Monitor Inoperability Requirements 1

If either the rod bank insertion limit monitor or rod position devia.

eion monitor are inoperable, additional surveillance is required to 1

ensure adequate shutdown margin is maintained.

1 If the rod position deviation monitor and quadrant power tilt monitor (s) are inoperable, the overpower reactor trip setpoint is reduced (and also j

power) to ensure ghet adequate core protection is provided in the event that unsatisfactory conditions arise that could affect radial power distribution.

Increased surveillance is required, if the quadrant power tilt monitors l

are inoperable and a load change occurs, in order to confirm satisfac.

tory pcwer distribution behavior.

The automatic alarm functions related to QUADRANT POWER TILT aust be considered incapable of alerting thh operator to unsatisfactory power distribution conditions.

)

J.

DNB Parameters TheRCSflowrate,T,yEs,sumptions.

and Pressurizer Pressure requirements are based on transient analyses The flow rate shall be verified by s

calorimetric flow data and/or elbow taps.

Elbow taps are used in the reactor coolant system as an instrument device that indicates the status of the reactor coolant flow. The basic function of this device is to j

provide information as to whether or not a reduction in flow rate has occurred.

If a reduction in flow rate is indicated below the value l

specified in the CORE OPERATING LIMITS REPORT, shutdown is required to investigate adequacy of core cooling during operation.

l Prairie Island Unit 1 - Anendnent flo. 91,92 Prairie Island Unit 2 - Anendnent flo. 84,85

j0aseg

1 k

UNITED $TATES p

NUCLE AR REGULATORY COMMISSION

)

a g

,f wAswiwoTow. o. c. 20sss s *...* e NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 85 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 November 17, 1989, 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 Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:

e

2-i I

Technical Specifications The Technical Specifications contained in Appendix A, as revised 1

through Amendment No.85,, are hereby incorporated in the license.

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

i 3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Mhn O. Thoma, Acting Director Project Directorate III-1 Division of Reactor Projects - III, IV, V & Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 13, 1990 W

l ATTACWENT TO LICENSE AMENDMENT N0. 85

[ACILITYOPERATINGLICENSENO.DPR-60 p0CKETNO.50-306 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendrent nuiaber and contain marginal lines indicating the area of change

, REMOVE INSERT TS-ix TS-ix TS-xiii TS-xiii TS.1-2 TS.1-2 TS 2.3-3 TS 2.3-3 TS.3.1-12 TS.3.1-12 Table TS 3.5-2 Table TS 3.5-2 TS.3.10-1 TS.3.10-1 TS.3.10-2 TS.3.10-2 TS.3.10-3 TS.3.10-3 TS.3.10-4 TS.3.10-4 6

TS.3.10-5 TS.3.10-5 TS.3s10-7 TS.3.10-7 TS.3.10-8 TS.3.10-8 TS.6.7-4 TS.6.7-4 TS.6.7-5 TS.6.7-5 TS.6.7-6 B.2.1-2 B.2.1-2 B.3.10-1 B.3.10-1 B.3.10-3 B.3.10-3 B.3.10-4 B.3.10-4 B.3.10-6 B.3.10-4 B.3.10-10 B.3.10-10 l

l a

~Q E

4 4.

T$eix l

l TABLI 0F CONTENTS (Continued)

I TS SECTION IIILE PACE 6.7 Reporting Requirements TS.6.7 1 A. Routine Reports TS.6.7 1

1. Annual Report TS.6.7 1

a. Occupational Exposure Report TS.6.7 1

b. Report of Safety and Relief Valve L

Failures and Challenges TS.6.7 1

c. Primary Coolant Iodine Spike Report TS.6.7 1

2. Startup Report TS.6.7 2

3. Monthly Operating Report TS.6.7 2

4. Semiannual Radioactive Effluent Release Report TS.6.7 3

5. Annual Summaries of Meteorological Data TS 6.7 4

6. Core Operating Limits Report TS.6.7 4 B. Reportable Events TS.6.7 5 C. Environmental Reports TS.6.7 5 1

1. Annual Radiation Environmental Monitoring Reports TS 6.7 5

2. Environmental Special Reports TS.6.7 6

3. Other Environmental Reports TS.6.7 6 (non radiological, non aquatic)

D. Special Reports TS.6.7 6 I

Prairie Island Unit 1 - Anendment No. 50, 59,63,73,80,91,92 Prairie Island Unit 2 - Amendnent No. 44,53,59,66,,73,84,85

~

'~

TS xiii APPENDIX A TECHNICAL SPECIFICATIONS LIST OF TICUREE TS FICURE M

2.1 1 Safety Limits Reactor Core, Thermal and Hydraulic Two Loop Operation 3.1 1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1 2 Unit 1 and Unit 2 Reector Coolant System Cooldown Limitations 3.1 3 DOSE EQUIVALENT I 131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uC1/ gram DOSE EQUIVALENT I 131 3.9 1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9 2 Prairie Island Nuclear Generating Plant Site Boundary for caseous Effluents 3.10 1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4 1 Shield Building Design In Leakage Rate 6.1 1 NSP Corporate Organizational Relationship to On Site Operating Organizations 6.1 2 Prairie Island Nuclear Canerating Plant Punctional Organization for on Sits Operating Group l

[

l Prairie Island Unit 1 - Anendment No. 91.92 Prairie Island Unit 2 - Amendoent No. 84,85 l

~.

l-

[

TS.1 2 l

CONTAINMENT INTECRITY CONTAINMENT INTEGRITY shall exist when:

1.

Penetrations required to be isolated during eccident conditions are either:

a.

Capable of baing closed by an OPERABLE containment automatic 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.

Blind flanges required by Table 75.4.4 1 are installed.

3.

The equipment hatch is closed and sealed.

4.

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

5.

The containment leakage rates are within their required limits.

COLD SFUTDOWN A reactor is in the COLD SHUTDOWN condition when the reactor is suberiti.

cal by at least 14 angst and the reactor coolant average tempereture is less than 200*F.

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 reactivicy. 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 provided core operating limits for the current operating reload cycle.

These cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.7.A.6 Plant operation within these operating limits is addressed in individual specifications.

Prairie Island Unit 1 - Anendnent No. 9,91,92 Prairie Island Unit 2 - Amendment No. 4,84,85 1

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

Open reactor coolant pump actor breaker.

1.

Reactor coolant pump bus undervoltage -

275% of normal voltage.

2.

Reactor coolant pump bus underfrequency.

258.2 Hz h.

Power range neutron flux rate.

1.

Positive rate

$15% of RATED THERP.AI. PokT.R with a time constant 22 seconds 2.

Negative rate 57% of RATED THERMAL P0k'ER with a time constant 22 seconds 3.

Other reactor trips High pressurizer water level 5906 of narrow a.

range instrument span, b.

Low low steam generator water level 25% of narrow range instrurtent 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 l

Prairie Island Unit 1 - Amendment No. 28,87,91,92 Prairie Island Unit 2 - Amendment No. 22,80.84,85 i

s.

TS.3.1 12 3.1.F.

ISOTHERMAI_ TFRPERATURE COE m CfrRT (ITC) 1.

Wen the reactor is critical, the isothermal temperature coefficient shall be less than 5 pcm/'T with all rods withdrawn, except during low power PHYSICS TESTS and as specified in 3.1.F.2 and 3.

2.

When the reactor is above 70 percent RATED THERMA 1. POWER with all rods withdrawn, the isothermal temperature coefficient shall be negative, e,xcept as specified in 3.1.F.3.

3.

If the limits of 3.1.F.1 or 2 cannot be met. POWER OPERATION may continue provided the following actions are taken:

a. Establish and maintain control rod withdrawal limits sufficient to restore the ITC to less than the limits specified in Specification 3.1.F.1 and 2 above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition to the insertion limits specified in the CORE OPERATING 1.IMITS REPORT.

b. Maintain the control rods within the withdrawal limits established above until a subsequent calculation verifies that the ITC has been restored to within its limit for the all rods withdrawn condition.

c. Submit a special report to the Commission within 30 days, describing the value of the measured ITC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the ITC to within its limit for the all rods withdrawn condition.

s i

l l

l Prairie Island Unit 1 - Anendment flo. 52,73,80,91,92 Prairie Island Unit 2 - Amendnent flo. 46,66,73,84,85 4

e,

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TABLE TS. 3. 5 -2 ( Page 2 o f 2)

INSTRUMENT OPERATING CONDITIONS FOR REACTOR TRIP 1

2 3

4 MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF OPERABLE DECREE OF,

BYPASS CONDITIONS OF COIllMN FUNCTIONAL UNIT C11ANNELS REDUNDANCY CONDITIONS (1) 1 OR 2 CANNOT BE NET

13. Undervoltage 4KV RCP Bus 1/ bus 1/ bus Maintain hot shutdown

14. Underfrequen.:y 4KV Bus 1/ bus 1/ bus Maintain hot shutdown

15. Control Rod Misalignment Monitor

a. Rod position deviation 1

Ing data required by

b. Quadrant power tilt 1

TS 3.10 1 and TS 3,10 J

16. RCP Breaker Open 2

1 Maintain hot shutdown

17. Safety Injection Actuation Signal 2

1 Maintain hot shutdown

18. Automatic Trip Ingic including Reactor Trip Breakers **

2 1

Notes 3, 4 Note 1: Automatic permissives not listed Note 2: When bypass condition exists, maintain normal operation Note 3:

With the number of operable channels one less than the minimum operable channels requirement, be in at least hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one 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 testing per Specification 4.1 provided the other channel is operable.

Note 4:

When in the hot shutdown condition with tl.e number of operable channels one less than the minimum operable channels requirement, restore the inoperable channel to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

Qis? :

F.P. - Full Power EEi co *

- One additional channel may be taken out cf service for low power physics testing o." !

- Includes both undervoltage and shunt trip circuits and if either circuit becomes inoperable the respective g

channel shall be considered inoperable.

to -

Prairie Island Unit 1 - Amendment No. 75,8i,0 Prairie Island Unit 2 - Amendment No. 68,8(

1-O 1r

['

TS.3.10 1

'0 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Aeolicability Applies to the limits on core fission power distribution and to the limits j

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.

Snecification A.

Shutdown Marrin l

^-

The shutdown margin with allowance for a stuck aontrol rod assembly f-shall exceed the applicable value shown in Figure TS.3.10 1 under all a

steady state operating conditions, except for PHYSICS TESTS, from zero to full power including effects of axial power distribution.

The shutdown margin as used here is defined as the amount by which the reactor core would be suberitical at HOT SHUTDOWN conditions if all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron concentration.

B.

Power Distribution Limits 1.

At all times, except d ring low power PHYSICS TESTING, measured hot channel factors, and aH, as defined below and in the bases, shall meet the ollowing limits:

Fh' x '1.03 x 1.05 5 (F P) x K(Z) q F$gx1.045F x [1+ PFDH(1 P))

H where the following definitions apply:

-F is the Fo limit at RATED THERMAL POWER specified in the CORE o

OPERATING LIMITS REPORT.

-F is the F3 limit at RATED THERMAL POWER specified in the CORE y

OPERATING LIM TS REPORT.

- PFDH is the Power Factor Multiplier for dH specified in the CORE OPERATING LIMITS REPORT.

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

- Z is the core height location.

- P is the fraction of RATED THERMAL POWER at which the core is operating. In the limit determination when P 50.50, set P = 0.50.

Prairie Island Unit 1 - Amendment No. 35,44,66,77,81,84,91,92 Prairie Island Unit 2 - faendment No. 29,30,60,70,74,77,84,05

1 TS.3.10 2 EorF[H is defined as the measured Fq or F3 3.10.B.1.

wkththesmallestmarginorgreatestexcessofrespectively, limit.

E 1.03 is th engineering hot channel factor, F,,ppygge e, gy, toaccountformanufacturingtoleharco.

measured q

1.05isappliedtothemeasured%toaccountfor.otsurement uncertainty.

-1.04isappliedtothemeasuredFfH to account for measurement uncertainty.

N 2.

Not channel factors, F andF[H,shallbemeasuredandthetarget o

j flux difference determined, at equilibrium conditions according to the following conditions, whichever occurs first:

(a) At least once pe'r 31 effective full power days in conjunction with the p rget 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$(equil)shallmeetthefollowinglimitforthemiddleaxial80%

of the core:

%(6quil)xV(Z)x1.03x1.05<(F P) x K(Z) q where V(2) is specified in the CORE OPERATING LIMIT 3 REPORT and other terms are defined in 3.10.B.1 above.

.g 3.

(a) If either measured hot channel factor exceeds its limit specified in 3.10.B.1, reduce reactor power and the high neutron f1 trip set point by it for each percent that the measured or by the factor specified in the CORE ORERATING IMITS REPORT for each percent that the measured j'

FyHexceedsthe3.10B.1 limit.

Then follow 3.10.B.3(c).

1 (b)IfthemeasuredF$(equil)exceedsthe3.10.B.2limitsbutnot the 3.10.B.1 limit, take one of the following actions:

i.

l 1.

Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months place the reactor in an equilibrium L

configuration for which Specification 3.10.B.2 is satis-fied, or 2.

Reduce reactor power and the high neutron flux trip sgtpoint by 14 for each percent that the measured F" (equil) x 1.03 x 1.05 x V(2) exceeds the limit.

q Prairie Island Unit 1 - Amendment flo. 35,44,66,77,81,84,91,92 Prairie Island Unit 2 - Amendment flo. 29,38,60,70,74,77,84,85 1

=

TS.3.10 3 3.10.B.3. (c) If subsequent in core mapping cannot, within a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, demonstrate that the hot channel factors are met, the reactor shall be brought to a HOT SHUTDOWN condition with return to power authorized up to 50% of RATED THERMAL POWER for the purpose of PHYSICS TESTING. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above 50% of RATED POWER.

THERMAL POWER may then be increased provided or H is demonstrated through in core mapping to be within its 1 mits.

(d) If two successiv measurements indicate an increase in the peak rod power AH with exposure, either of the following actions shall be taken:

%(equil)shallbemultipliedby1.02xV(Z)x1.03x 1.

1.05 for comparison to the limit specified in 3.10.B.2, or 2.

d(equil)shallbemeasuredatleastonceperseven effective full power days until two successive maps indicate that the peak pin power' E, is not increasing.

AH 4

Except during PHYSICS TESTS, and except as provided by specifica-tions 5 through 8 below, the indicated axial flux difference for at least three operable excore channels shall be maintained within the target band about the target finx difference.

The target band is specified in the CORE OPERATING LIMITS REPORT.

5.

Above 90 cereent of RATED THERMAL POWER!

If the indicated axial flux difference of two OPERABLE excore channels deviates from the target band, within 15 minutes either l

eliminate such deviation, or reduce THERMAL POWER to less than 90 percent of RATED THERMAL POWER.

6.

Between 50 and 90 cereent of RATED THERMAL PdWER a.

The indicated axial flux difference may deviate from the l

target band for a maximum of one* hour (cumulative) in any 24-hour period provided that the difference between the indicated axial flux difference about the target flux difference does not exceed the envelope specified in the CORE OPERATING LIMITS REPORT.

b.

If 6.a is violated for two OPERABLE excore channels then the THERMAL POWER shall be reduced to less than 50% of RATED THERMAL POWER and the high neutron flux setpoint reduced to

' less than 55% of RATED THERMAL POWER.

May be extended to 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months during incore/excore calibration.

Prairie Island Unit 1 - Amendrmnt No. 35,44,91,92 Prairie Island Unit 2 - Amendment No. 29,38,84,85 l

j i

xq TS.3.10 4 3.10.3.6. c.

A power increase to a level greater than 90 percent of rated power is contingent upon the indicated axial flux difference of at least three OPERABLE excore channels being within the

~

target band.

7.

Iass than 50 nereent of RATED THERMAL POVERt a.

The i.ndicated axial flux difference may deviate from the-

- l target band, b.

power increase to a level greater than 50 percent of RATED THERMAL POWER is contingent upon the indicated axial flux difference of at least three OPERABLE excore channels not being outside the target band for more than one hour (cumula-tive) out of the preceding 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

8.

In applying 6a and 7b above, penalty deviations outside the l

gP target band shall be accumulated on a time basis of:

One minute penalty deviation for each one minute of power a.

operation outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and b.

One h61f minute penalty deviation for each one minute of power operation outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

9.

If alarus associated with monitoring the indicated axial flux difference deviations from the target band are not operable, the l

indicated axial flux difference value for each OPERABLE excore channel shall be logged at least once per hour for Phe first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and half hourly thereafter until the alarms are returned to an OPERABLE status.

For the purpose of' applying this specifica-tion, logged values of indicated axial flux difference must be assumed to apply during the previous interval between loggings.

C. OUADRANT POWER TILT RATIO 1.

Except for PHYSICS TESTS, if the QUADRANT POWER TILT RATIO exceeds 1.02 but is less than 1.07, the rod position indication shall be monitored and logg,ed once each shift to verify rod position within each bank assignment and, within two hours, one of the following steps shall be taken:

a.

Correct the QUADRANT POWER TILT RATIO to less than 1.02.

b.

Restrict core power level so as not to exceed RATED THERMAL POWER less 24 for every 0.01 that the QUADRANT POWER TILT RATIO exceeds 1.0.

Prairie Island Unit 1 - Amendment No. 29,44,91,92 Prairie Island Unit 2 - Amendment No. 23,38,84,85 p &

a at e q se gessee meanimeeguerge n e

esguea w egs p eeme = +ce m

e s.,-w.

-d W

~

TS.'3.10o5 3.10.C.2.

'If the QUADRANT POWER TILT RATIO exceeds 1.02 but is less than 1.07 for a sustained period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or if such a tilt recurs intermittently, the reactor shall be brought to the HOT SHUTDOWN condition.

Subsequent operation below 50% of rating, for testing, shall be permitted.

3.

Except for PHYSICS TESTS if the QUADRANT POWER TILT RATIO exceeds 1.07, the reactor shall be brought to the HOT SNUTDOWN condition.

Subsequent operation below 50% of rating, for testing, shall be permitted.

4 If the core is operating above 854 power with one excore nuclear channel inoperable, then the core quadrant power balance shall be determined daily and after a lot power change using either 2 movable detectors or 4 core thermocouples per quadrant, per Specification 3.11.

D. Rod Insertion Limits 1.

The shutdown rods shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT when the reactor is critical or approaching criticality.

2.

When the reactor is critical or approaching criticality, the control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT.

3.

Insertion limits do not apply during PHYSICS TESTS or during periodic exercise of individual rods. The shutdown margin shown in Figure TS.3.101 must be maintained except for low power PHYSICS TESTING.

For this test the reactor may be critical with all but one high worth full-len6th control rod inserted for a period not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per year provided a rod drop test is run on the high worth full-length rod prior to this particular low power PHYSICS TEST.

Prairie Island Unit 1 - Anendment No. 32,44,91,92 Prairie Island Unit 2 - Amendment No. 26,38,84,85 l

l

e TS.3.1097 3.10.G. Inocerable Rod Limitations 1.

An inoperable rod is a rod which (a) does not trip, (b) is

. declared inoperable under specification 3.10.E. or 3.10.H. or (c) cannot be moved by its drive mechanism and cannot be corrected within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

2.

The reactor shall be brought to the HOT SHUTDOWN condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months should more than one inoperable rod be discovered during POWER OPERATION.

3.

If the inoperable rod is located below the 200 step level and is capable of being tripped, or if the rod is located below the 30 step level whether or not it is capable of being tripped, then the insertion limits specified in the CORE OPERATING LIMITS REPORT.

l apply.

4 If the inoperable rod cannot be located, or if-the inoperable rod is located above the 30 step level and cannot be tripped, then the insertion limits specified in the CORE OPERATING LIMITS REPORT l

apply.

5.

If POWER OPERATION is continued with one inoperable rod, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days unless the rod is earlier made OPERABLE.

The analysis shall include due allowance for nonuniform fuel depletion in the neighborhood of the iroperabl,e rod.

If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis THERMAL POWER shall be reduced to a level consistent with the safety analysis.

H.

Rod Dron Time At operating temperature and full flow, the drop time of each RCCA shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to dashpot entry.

If the time is greater than 1.8 seconds, the rod shall be declared inoperable.

Prairie Island Unit 1 - Nnendment flo. 44,91,92 Prairie Island Unit 2 - Amendment !!o. 38,84,85 4

l.

l j

TS.3 10 8 3.10 I.' Monitor-Inonerability'Recuirements 1.

If the rod bank insertion limit monitor is inoperable, or if the rod position deviation monitor is inoperable, individual rod positions shall be logged once per shift, after a load change grar.ter than 10 percent of RATED THERMAL POWER, and after 30 inches or more of rod action.

2.

If both the rod position deviation monitor and one or both of the quadrant power tilt monitors are inoperable for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or more, the nuclear overpower trip shall be reset to 934 of RATED THERMAL POWER in addition to the increased surveillance requirements.

3.

If one or both of the quadrant power tilt monitors is inoperable, individual upper and lower excore detector calibrated outputs and the calculated power tilt shall be logged every two hours after a load change greater than 10% of RATED THERMAL POWER J. DNB Parameters The following DNB related parameters limits shall be maintained during POWER OPERATION:

a.

Reactor Coolant System Tavg <564'F b.

Pressurizer Pressure 22220 psia

c.

Reactor Coolant Flow 2the value specified in the CORE OPERATING LIMITS REPORT With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Compliance with a. and b is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Compliance with c. is demonstrated by verifying that the parameter is within its limit after each refueling cycle.

Limit not applicable during either a THERMAL POWER ramp increase in excess of (54) RATED THERMAL POWER per minute or,a THERMAL POWER step increase in excess of (10%) ltATED THERMAL POWER Prairie Island Unit 1 - Amendment No. 16,19,44,77,91,92 Prairie Island Unit 2 - Amendment No. 10,13,38,70,84,85 l

l

.=

TS.6.7-4 6.7.A.5, Annum 1 5 - -ries of Meteorolonical Data An annual susanary of meteorological data shall be submitted for the previous calendar year in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability at the request of the Commission.

6.7.A.6. Core Deeratine Limits Renort

a. Core operating limits shall be established and documented in the CORE OPERATING 1.IMITS REPORT before each reload cycle or any i

remaining part of a reload cycle for the following:

1. Heat Flux Hot Channel Factor Limit (F

), Nuclear Enthalpy Rise Hot Channel Factor, Limit (F

), PFDH, K(Z) and V(Z)

(Specifications 3.10.B.1,3.10.5.2and3.10.B.3)

2. Axial Flux Difference Limits and Target Band (Specifications 3.10 B.4 through 3.10.B.9)

3. Shutdown and Control Bank Insertion Limits (Specification 3.10.D)

4. Reactor Coolant System Flow Limit (Specification 3.10.J) e

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NSPhD-8101-A, " Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version)

NSPNAD-8102 A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units" (latest approved version)

WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology", July, 1985 WCAP 10054-P A, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", August, 1985 WCAP-10924 P A, "Westin6 ouse Large-Break LOCA Best Estimate h

Methodology", December, 1988 XN-NF-77-57 (A), XN NF 77-57, Supplement 1 (A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II", May, 1981 c..The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

Prairie Island Unit 1 - Amendment flo. 54,59,73,91,92 Prairie Island Unit 2 - Amendment flo. 48,53,66,84,85

TS.6.705 d'. The CORE OPERATING LIMITS REPORT, including'any mid cycle revisions or supplements thereto, shall be supplied upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the

{

Regional Administrator and Resident Inspector.

1 B. REPORTABLE EVENTS The following' actions shall be taken for REPORTABLE EVENTS:

The Commission shall be notified by a report submitted pursuant a.

to the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the Operations Committee and the results of this review shall be submitted to the Safety Audit Committee and the Vice President Nuclear Generation.

C. Environmental Reeorta i

The reports listed below shall be submitted to the Administrator of the appropriate Regional NRC Office or his designate:

1.

Annual Radiation Environmental Monitorine Reeert i

l (a) Annual Radiation Environmental Monitoring Reports covering f

[

the operation of the program during the previous calendar year shall be submitted prior to May.1 of each year.

(b) The Annual Radiation Environmental Monitoring Reports shall include summaries, interpretations, and an analysis of l-trends of the results of the radiological environmental l

surveillance activities for the report period, including a comparison with preoperational studies, operational-controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts-of the plant' operation on the environment.

The reports shall also include the results of land use censuses l

required by Specification 4.10.B.1.

If harmful effects or

(

evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

(c) The Annual Radiation Environmental Monitoring Reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period.

In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted as soon as possible in a supplementary report.

Prairie Island Unit 1 - Amendment No. 54,59,73,91,92 Prairie Island Unit 2 - Amendment No. 48,53,66,84,85

TS.6.706

[,

(d) 'The reports shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table

.giving distances and directions from one reactor; and the results of licensees participation in the Interlaboratory comparison Program, required by Specification 4.10.C.1.

2.

Environmental Snecial Renorts (a) When radioactivity levels in samples exceed limits specified in Table 4.10 3, an Environmental Special Report shall be submitted within 30 days from the-end of the affected calendar quarter.

For certain cases involving long analysis time, determination of quarterly averages may extend beyond the 30 day period. -In these cases the potential for exceeding the quarterly limits will be reported within the 30 day period to be followed by the Environmental S'pecial Report as soon as practicable.

3, Other Environmental Retorts (non-radiolorical. non-acuatie)

Written reports for the following items shall be submitted to the appropriate NRC Regional Administrator:

a.

Environmental events that indicate or could result in a significant environmental impact casually related to plant operation. The following are examples:

excessive bird impaction; onsite plant or animal disease ou,tbreaks; unusual mortality of any species protected by the Endangered Species Act of 1973; or increase in nuisance organisms or conditions. This report shall be submitted within 30 days of the event and shall (a) describe l analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.

b.

Proposed changes, test or experiments which may result in a significant increase in any adverse environmental impact which was not previously reviewed or evaluated in the Final Environmental Statement or supplements thereto. This report shall include an evaluation of the environmental impact of the proposed activity and shall be submitted 30 days prior to implementing the proposed change, test or experiment.

D.

Seeelal Recorts Unless otherwise indicated, special reports required by the Technical Specifications shall be submitted to the appropriate NRC Regional Administrator within the time period specified for each report.

Prairie Island Unit 1 - Amendment No. 92 Prairie Island Unit 2 - Amendnent No. Ob

__=

________._________-______-__._______m---m_

- - - = - - -

.B.2.1-2 2.1 SAFETY LIMIT. REACTOR CORE A1111 continued power levels of 91% and 74% respectively.

For the 2235 psig and 2385 psig curves, the coolant average temperature at the core exit is equal to 650'F below power levels of 64% and 73% respectively.

The third and fourth criteria are evaluated using standard DNB metho-dology. For all four curves the DNBR is limiting at higher power levels.

. The area of safe operation is below these curves.

The plant conditions required to violate the limits in the lower power range are precluded by th's self-actuated safety valves.on the steam generators.

The highest nominal setting of the steam generator safety valves is 1129 psig (saturation temperature 560*F). At zero power the difference between primary coolant and secondary coolant is zero and at full power it is 50*F.

The reactor conditions at which steam generator safety valves open is shown as a dashed line on Figure TS.2.1-1.

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:

N R P [1 + PFDH(1 P)] ; and Fh - FRTP F

,p where:

RTP

-F is the F limit at RATED THERMAL POWER specified in the CORE o

g OPERATING LIMITS REPORT.

RTP F

is the F limit at RATED THERMAL POWER specified in the CORE OhkRATINGLIMkTSREPORT.

H

-PFDHisthePowerFactorMultiplierforFfHspecifiedintheCORE 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.

This combination of hot channel factors is hi her than that calculated 6

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

1s 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.

Prairie Island Unit 1 - Amendment flo. 91,92 Prairie Island Unit 2 - Anendnent flo. 84,85

B.3.10 1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS AA111 Throughout the 3.10 Technical Specifications, the terms ' rod (s)" and "RCCA(s)" are synonymous.

A.

Shutdown Margi,n Trip shutdown reactivity is provided consistent with plant safety analyses assumptions.

One percent shutdown margin is adequate except for the steam break analysis, which requires more shutdown reactivity due to the more negative moderator temperature coefficient at and of life (when boron concentration is low). Figure TS.3.10 1 is drawn accordingly.

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 mtchanical properties to within assumed design criteria.

The ECCS analysis was performed in accordance with SECY 83-472. One calculation at the 954 probability level was performed as well as one calculation with all the required features of 10'CFR Pagt 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 LIMITS REPORT for the To limit specified in the CORE OPERATING LIMITS REPORT. Maintaining I) peaking factors below the Fq 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 954 probability level assures compliance with the ECCS analysis.

During opeqation, the plant staff compares the measured hot channel factors, F'q and aH, (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' is the measured Nuclear Hot Channel Factor, defined as the maximum q

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(2) function specified in the CORE OPERATING LIMITS REPORT is a normalized function that limits Fq axially. The K(Z) value is based on large and small break LOCA analyses.

Prairie Island Unit 1 - Amendment No. 91,92 Prairie Island Unit 2 - Amendment No. 84,85 l

B.3.10 3 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Aaasa continued WhenameasurementofF[H is taken measurement error must be allowed for and 4 percent is the appropriate e.llowance for a full core map taken with the movable incore detector flux mapping system.

Measurements of the hot channel factors are required as part of startup PHYSICS TESTS, at least once each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors, The incore map taken following initial loading provides confirmation of the basic' nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would otherwise affect these bases.

For normal operation, it is not necessary to measure these quantities.

Instead it has been determined that, provided certain conditions are observed, the hot channel factor limits will be met; these conditions are as follows:

1.

Control rods in a single bank move together with no individual rod insertion differing by more than 15 inches from the bank demand position. An accidental misalignment limit of 13 steps precludes a rod misalign-ment greater than 15 inches with consideration of maximum instrumentation error.

2.

Control rod banks are sequenced with overlapping banks as described in Technical Specification 3.10.

3.

The control bank insertion limits specified in the CORE OPERATING LIMITS REPORT are not violated.

4 Axial power distribution control procedures, which are given in i

terms of flux difference control and control bank insertion limits are observed.

Flux difference refers to the difference in signals between the top and bottom halves of two section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as-the difference in normalized power between the top and bottom halves of the core.

Prairie Island Unit 1 - Amendment flo. 91,92 Prairie Island Unit 2 - Anendment flo. 84,85 e

t's a s e eae a ee e

e,

, g e as sie ene.%,,me,,,

o g,e., ewe

.ee me.

,-c 5

B.3.10-4 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS R&111 continued B.

Power Distribution Control (continued)

The permitted relaxation in F$H and F( allows for radial power shape changes with rod insertion to the insertion limits.

It has been determined that provided the above conditions 1 through 4 are obse ed.

these hot channel factor limits are not.

In specification 3.10, is arbitrarily limited for P less than or equal to 0.5 (except for lo power PHYSICS TESTS).

The procedures for axial power distribution' control referred to above are designed to minimize the effects of xenon redistribution'on the axial power distribution during load. follow maneuvers. Basically control of flux difference is required to limit the difference between the current value of Flux Difference (AI) and a reference value which corresponds to the full power equilibrium value of Axial Offset (Axial Offset - AI/ fractional power).

The reference value of flux difference varies with power level and burnup but expressed as axial offset it varies only with burnup, j

The technical specifications on power distribution control assure that the F limit is not exceeded and xenon distributions are not developed q

which at a later time, would cause greater local power peaking even though the flux difference is then within the limits specified by the procedure.

The target (or reference) value of flux difference is determined as follows: At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with the full l'

length rod control rod bank more than 190 steps withdrawn (i.e., normal full power operating position appropriate for the time in life, usually l

withdrawn farther as burnup proceeds). This value, divided by the l

fraction of full power at which the core was operating is the full i

power value of the target flux difference. Values for all other core power levels are obtained by multiplying the full power value by the fractional power.

Since the indicated equilibrium was noted, no allowances for excore detector error are necessary and indicated deviation from the indicated reference value but within the target band

'is permitted.

The allowed deviation from the target flux difference as a function of THERMAL POWER is specified in the CORE OPERATING LIMITS l

REPORT.

Prairie Island Unit 1 - Amendnent flo. 91,92 Prairie Island Unit 2 - Amendment No. 84,85 l

B.3.10 6 3.10 CONTROL ROD AND POVER DIST'IBUTION LIMITS R

IAL11 continued B.

Power Distribution control (continued)

In some instances of rapid plant power reduction, automatic rod motion will cause the flux difference to deviate from the target band when the reduced power level is reached.

This does not necessarily affect the xenon distribution sufficiently to change the envelope of peaking factors which can be reached on a subse-quent return to full power within the target band, however to simplify the specification, a limitation of one hour in any period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is placed on operation outside the band.

This ensures -

that the resulting xenon distributions are not significantly different from those resulting from operation within the target band.

The consequences of being outside the target band but within the limits specified in the CORE OPERATING LIMITS REPORT for power levels between 50% and 904 has been evaluated and determined to result in acceptable peaking factors.

Therefore, while the deviation exists the power level is limited to 90 percent or lower depending on the indicated axial flux difference.

In all cases the target band is the Limiting Condition for Operation.

Only when the target band is violated do the limits specified in the CORE OPERATING. LIMITS REPORT apply.

If, for any reason, the indicated axial flux difference is not control-led within the target band for as long a period as one hour, then xenon l,

distributions may be significantly changed and operation at or below 50 percent is required to protect against potentially more severe consequences of some accidents.

As discussed above, the essence of the procedure is to maintain the xenon distribution in the core as close to the equilibrium full power condition as'possible.

This is accomplished by using the boron system to position the full length control rods to produce the required indicated flux difference.

For Condition II events the core is protected from overpower and a minimum DNBR of 1.30 for Exxon fuel and 1.17 for Westinghouse fuel by an automatic protection system.

Compliance with operating procedures is assumed as a precondition for Condition II transients, however, operator error and equipment malfunctions are separately assumed to lead to the cause of the~ transients considered.

C.

QUADRANT POWER TILT RATIO QUADRANT POWER TILT RATIO limits are based on the following considera-tions.

Frequent power tilts are not anticipated during normal operation since this phenomenon is caused by some asymmetric perturbation, e.g.

rod misalignment, x-y xenon transient, or inlet temperature mismatch. A dropped or misaligned rod will easily be detected by the Rod Position Indication System or core instrumentation per Specification 3.10.F, and Prairie Island Unit 1 - Amendment No. 91,92 Prairie Island Unit 2 - Amendment No. 84,85

B.3.10 10 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS 11111 continued H.

Rod Drop Time The required drop time to dashpot entry is consistent with the safety analysis.

I.

Monitor Inoperability Requirements If either the rod bank insertion limit monitor or rod position devia-tion monitor are in6perable, additional surveillance is required to ensure adequate shutdown margin is maintained.

If the rod position deviation monitor and quadrant power tilt monitor (s) are inoperable, the overpower reactor trip setpoint is reduced (and also I

power) to ensure that adequate core protection is provided in the event that unsatisfactory conditions arise that could affect radial power distribution.

L Increased surveillance is required, if the quadrant power tilt monitors are inoperable and a load change occurs, in order to confirm satisfac-tory power distribution behavior.

The automatic alarm functions related to QUADRANT POWER TILT must be considered incapable of alerting thb l

operator to unsatisfactory power distribution conditions.

1 L

J.

DNB Parameters The RCS flow rate T and Pressurizer Pressure requirements are based ontransientanalyse,yEs,sumptions. The flow rate shall be verified by s

calorimetric flow data and/or elbow taps.

Elbow caps are used in the reactor coolant system as an instrument device that indicates the status of the reactor coolant flow.

The basic function of this device is to provide information as to whether or not a reduction in flow rate has occurred.

If a reduction in flow rate is indicated below the value j

specified in the CORE OPERATING LIMITS REPORT, shutdown is required to l

investigate adequacy of core cooling during operation.

l Prairie Island Unit 1 - Anendment No. 91,92 Prairie Island Unit 2 - Anendnent No. 84,85

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ML20012C389

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