ML20202H084

Amends 77 & 70 to Licenses DPR-42 & DPR-60,respectively, Revising Tech Specs to Incorporate Change in Fuel Suppliers

ML20202H084
Person / Time
Site:	Prairie Island
Issue date:	04/03/1986
From:	Lear G Office of Nuclear Reactor Regulation
To:
Shared Package
ML20202H091	List: ML20202H084 ML20202H102
References
NUDOCS 8604150090
Download: ML20202H084 (44)

Text

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.e NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.77 License No. DPR-42 1.

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

A.

The application for amendment by Northern States Power Company (thelicensee)datedJanuary 13, 1986, supplemented March 25, 1986, 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; 8.

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 (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in 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, i

8604150090 860403 PDR ADOCK 05000282 P

PDR J

4 4 2.

Accordingly, the license is amended by changes to the Technical

^

Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.

DPR-42 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 77, 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.

OR THE NUCLEAR REGU TORY COMMISSION

/

George E. Lear, Director I

PWR Project Directorate #1 Division of PWR Licensing-A

Attachment:

Changes to the Technical Specifications Date of Issuance: April 3,1986 i

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m.

,a ATTACHMENT TO LICENSE AMENDMENT NO. 77 FACILITY OPERATING LICENSE NO. DPR-42 DOCKET NO. 50-282 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT TS-x TS-x TS.2.1-1 TS.2.1-1 TS.2.1-2 TS.2.1-2 TS.2.1-3 TS.2.3-6 TS.2.3-6 Figure TS.2.1-1 Figure TS.2.1-1 TS.3.1-17 TS.3.1-17 TS.3.1-18 TS.3.1-18 TS.3.3-1 TS.3.3-1 TS.3.10-1 TS.3.10-1 TS.3.10-2 TS.3.10-2 TS.3.10-8 TS.3.10-8 TS.3.10-9 TS.3.10-9 TS.3.10-10 TS.3.10-10 TS.3.10-11 TS.3.10-11 TS.3.10-13 TS.3.10-13 TS.3.10-17 TS.3.10-17 Figure TS.3.10-5 Figure TS.3.10-5 Figure TS.3.10-7 Figure TS.3.10-8 Figure TS.3.10-7*

• Existing Figure TS.3.10-7 will be deleted by this change. Existing Figure TS.3.10-8 will be renumbered to TS.3.10-7.

i

4

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TS-x REV APPENDIX A TECHNICAL S1'ECIFICATIONS LIST OF FIGURES TS FIGURE 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 I and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 Effect of Fluence and Copper Content on Shift of RT NDT Reactor Vessel Steels Exposed to 550*F Temperature 3.1-4 Fast Neutron Fluence (E >l MeV) as a Function of Full Power Service Life 3.1-5 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >l.0 uCi/ gram DOSE EQUIVALENT I-131 3.9-1 Prairie Island Nucisar Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island N' clear Generating Plant Site Boundary for u

Gaseous Effluents 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3.10-2 Control Bank Insertion Limits 3.10-3 Insertion Limits 100 Step Overlap with One Bottomed Rod 3.10-4 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10-7 V(Z) as a Function of Core Height l

4.4-1 Shield Building Design In-Leakage Rate 6.1-1 NSP Corporation Organization 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 - Anendment No. 52, 59, 66, #9, 73, 77 Prairie Island Unit 2 - Amendment No. 66,53,69,$$,66,70 j

1 l

TS.2.1-1 REV 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 2.I SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of thermal power, reactor coolant system pressure and coolant temperature during operation.

Objective To maintain the integrity of the fuel cladding.

Specification 1.

The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure TS.2.1-1.

The safety limit is exceeded if the point defined by the combination of I

reactor coolant system average temperature and power level is at any time above the appropriate pressure line.

l Basis

\\

To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions. This is accomplished by operating the hot regions of the core within the nucleate boiling regime of heat transfer wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation te=perature. The upper boundary of the nuc1'este boiling regime is termed departure from nucleate boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation. Therefore, the observable parameters; thermal power, reactor coolant temperature and pressure have been related to DNB through the W-3 and WRB-1 DNB correlations.

The W-3 DNB correlation is used for Exxon Nuclear fuel. The WRB-1 DNB correlation is used for Westinghouse fuel. The W-3 and WRB-1 DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB heat flux ratio, I

defined as the ratio of the heat flux that would cause DNB at a particular I

core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNii ratio, DNBR, during steady state operation, nor=al operational transients, and anticipated transients is limited to 1.30 for the Exxon Nuclear fuel and to 1.17 for the Westinghouse fuel.

l These limits correspond to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

Prairie Island Unit 1 - Amendment No. 77 Prairie Island Unit 2 - Amendment No. 70

TS.2.1-2 REV The solid curves of Figure TS 2.1-1 represent the loci of points of thermal power, coolant pressure, and coolant average temperature

~

for which either the coolant enthalpy at the dore exit is limiting or the DNB ratio is limiting.

For the 1685 psig and 1985 psig curves, the coolant average enthalpy at the core exit is equal to saturated water enthalpy below 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. For all four curves, the DNBR is limiting at higher power l

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 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.

Except for special tests, power oneration with only one loop or with natural circulation is not alloweu.

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:

F

=

+. ( - )] ; and F - 2.30 aH 9

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 cal,culated full power for,the range from all control rods fully withdrawn to at 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 by Figure TS.3.10-1 assure that the DNB ratio 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 - Amendment No. 77 Prairie Island Unit 2 - Amentment No. 70

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0 20 40 60 80 100 120

% Rated Core Power SAFETY LIMITS. REACTOR CORE, TH:=RMAL AND HYDRAUIIC WO-LOOP OPERATION l

FIGURE TS.C.1-1 Prairie Island Unit 1 - Anendment No. 77

.i TS.2.3-6 REV The other reactor trips specified in A.3.

above provide additional protection.

The trip initiated by steam /feedwater flow mismatch in coincidence with low steam gsnerator water level is designed for protection from a sudden loss of the rsactor's heat sink. The safety injection signal trips the reactor to decrease the severity of the accident condition. The reactor is tripped when the turbine generator trips above a power level equivalent to the load rejection capacity of the steam dump valves.

This reduces the severity of the loss-of-load transient.

The positive power range rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power 1svel.

Specifically, this trip compliments the power range nuclear flux l

high and low trip to assure that the criteria are met for rod ejection from-1 partial power.

The negative power range rate trip provides protection satisfying all IEEE criteria to assure that minimum DNBR is maintained above 1.30 for Exxon Nuclear j

]

fuel and above 1.17 for Westinghouse fuel for all multiple control rod drop accidents. Analysis indicates (Section 14.1.3) that in the case of a single rod drop, a return to full power will be initiated by the automatic reactor control l

l system in response to a continued full power turbine load demand and it will not result in a DNBR of less than 1.30 for Exxon Nuclear fuel or 1.17 for Westinghouse fuel. Thus, automatic protection for a single rod drop is not required.

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

j l

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Rsferences:

(1)

FSAR 14.1.1 i

(2)

FSAR Page 14-3 (3)

FSAR. 14.2.6 (4)

FSAR 14.3.1 (5)

FSAR 14.1.2 (6)

FSAR

?.2, 7.3 j

(7)

FSAR 3.2.1 i

(8) FSAR 14.1.9 i

(9)

FSAR 14.1.11 1

Prairie Island Unit 1 - Amendment No. 33, 77 Prairie Island Unit 2 - Amendment No. 27, 70

FIGURE TS.3.10-7 REV i

1.20 1.18 1.16 (11.2.1.15) 1.14 j

1.12 j

/l V(Z) 1.10 (9.25,1.11) 1.08 l

1.06 1.04 1.02 1.00 0

2 4

6 8

10 12 Core Height (Ft)

V(Z) as a Function of Core Height Prairie Island Unit No.1 - Amendment No. 35, 77

~

Prairie Island Unit No. 2 - Amendment No. 29, 70

n TS.3.1-17 REV F.

MINIMUM CONDITIONS FOR CRITICALITY Specification 1.

Isothermal Temperature Coefficient (ITC) l a.

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.1.b and c.

l 1

b.

When the reactor is above 70 percent rated thermal power with all rods withdrawn, the isothermal temperature coefficient shall be negative, except as specified in 3.1.F.1.c.

c.

If the limits of 3.1.F.1.a or b cannot be met, Power Operation may continue provided the following actions are taken:

i (1) Establish and maintain control rod withdrawal limits sufficient to restore the ITC to less than the limits specified in Specification 3.1.F.1.a and b above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be 1

in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These withdrawal i

limits shall be in addition to the insertion limits of Figure TS.3.10-2.

(2) 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.

(3) 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.

1 2.

The reactor shall not be brought to a critical condition until the 4

reactor coolant temperature is higher than that defined by the criti-cality limit line shown in Figure TS.3.1-1.

I Basis l

At the beginning of a fuel' cycle the moderator temperature coefficient has its most positive or least ne,gative value.

As the boron concentration is reduced throughout the fuel cycle, the moderator temperature coefficient becomes more negative.

The isothermal temperature coefficient is defined l

as the reactivity change associated with a unit change in the moderator and fuel temperatures.

Essentially, the isothermal temperature coefficient is the sum of the moderator and fuel temperature coefficients. This co-efficient is measured directly during low power physics tests in order to verify analytical predictions. Theunitsogtheisothermaltemperature coefficient are pcm/*F, where I pcm = 1x10 AK/K.

Prairie Island Unit No.1 - Amendment No. 35, 77 i

Prairie Island Unit No. 2 - Amendment No. 29, 70

- ~__

s TS.3.1-18 REV For extended optimum fuel burnup it is necessary to either load the reactor with burnable poisons or increase the boron concentration in the reactor coolant system. If the latter approach is emphasized, it is possible that a positive isothermal temperature coeffieient could exist at beginning of cycle (BOC). Safety analyses verify the accep-tability of the isothermal temperature coefficient for limits specified in 3.1.F.1.

Other conditions, e.g., higher power or partial rod insertion would cause the isothermal coefficient to have a more negative value. These analyses demonstrate that applicable criteria in the NRC Standard Review Plan (NUREG 75/087) are met.

t Physics measurements and analyses are conducted during the reload startup test program to (1) verify that the plant will operate within safety analyses assumptions and (2) establish operational procedures to ensure safecy analyses assumptions are met.

The 3.1.F.1 requirements are waived during low power physics tests to permit measurement of reactor temperature coefficient and other physics design parameters of Special operating precautions will be taken during these (3) interest.

' physics tests.

In addition, the strong negative Doppler coefficient and the small integrated ak/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

The requirement that the reactor is not to be made critical except as specified in Figure TS.3tl-1 provides increased assurance tLat the l

proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility temperature range.

Heatup to this temperature will be accomplished by operating the reactor coolant pumps and by the pressurizer heaters.

The pressurizer heater and associ-ated power cables have been sized for continuous operation at full heater power. The shutdown margin in Specification 3.10 precludes the possibility of accidental criticality as a result of an moderatortemperatureoradecreaseofcoolantpressure.{gyreaseof

References:

(1) FSAR Figure 3.2-10 (2)

FSAR Table 3.2-1

. Prairie Island Unit No. 1 - Amendment No. 35, 77 Prairie Island Unit No. 2 - knendment,No. 27, 70

l TS.3.3-1 REV 3.3 ENGINEERED SAFETY FEATURES Applicability Applies to the operating status of the engineered safety features.

1 Objective

)

To define those limiting conditions that are necessary for operation of engineered safety features:

(1) to remove decay heat from the core in an emergency or normal shutdown situations, and (2) to remove heat from containment in normal operating and emergency situations.

Specifications A.

Safety Injection and Residual Heat Removal Systems

1. A reactor shall not be made or maintained critical nor shall it be heated or maintained above 200*F unless the following condi-tions are satisfied except as permitted in Specification 3.3.A.2.

a.

The refueling water tank contains not less than 200,000 gallons of water with a boron concentration of at least 1950 ppm.

b.

Each reactor coolant system accumulator shall be operable when reactor coolant system pressure is greater than 1000 psig.

Operability requires:

(1) The isolation valve is open (2) Volume is 1270 2 20 cubic feet of borated water l

(3) A minimum boron concentration of 1900 ppm (4) A nitrogen cover pressure of at least 700 psig c.

Two safety injection pumps are operable except that pump control switches in the control room shall meet the require-ments of Section 3.1.G whenever the reactor coolant system temperature is less than MPT.

d.

Two residual heat rem: val pumps are operable.

e.

Two residual heat exchangers are operable.

f.

Automatic valves. interlocks and piping associated with the 1,

above components and required to function during accident conditions, are operable.

4 g.

Manual valves in the above systems that could (if one is improperly positioned) reduce injection flow below that assumed for accident analyses, shall be blocked and tagged in the proper position for injection.

RHR system valves, however, may be positioned as necessary to regulate plant heatup or cooldown rates when the reactor is subcritical. All changes in valve position shall be under direct administrative control.

Prairie Island Unit No.1 - Amendment No. 33, SI, 77 Prairie Island Unit No. 2 - Amendment No. 32, 55, 70

4 s

TS.3.10-1 REV i

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability l

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

i Objective 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 ejecticn.

Specification A.

Shutdown Reactivity The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown in Figure TS.3.10-1 under all 2

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 j

reactor core would be subcritical at hot shutdown conditions if all j

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

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

B.

Power Distribution Limits 1.

At all times, except dgring g power physics testing, measured

)

hot channel factors, T' and T H, as defined below and in the bases,shallmeetthehollowinglimits:

F x 1.03 x 1.05 <(2.30/P) x K(Z)

I*

+

(~

F H*

where the following definitions apply:

- K(Z) is the axial dependence function shown in Figure TS.3.10-5.

- Z is the core height location.

1

- P is the fraction of rated power at which the core is operating. In the limit determination when P s.50, set P = 0.50.

4 i

(2.21/P) shall be used for Westinghouse Standard assemblies for Unit 2 Cycle 10.

• • 1.55 x (1 + 0.2 (1-P)] shall be used for Westinghouse Standard assemblies i

for Unit 2 Cycle 10.

Prairie Island Unit No.1 - Amendment No. 75, $$, $$, 77 Prairie Island Unit No. 2 - Amendment No. 29, 38, 60, 70

..-.--,__-.s

s TS.3.10-2 REV or [ is defined as the measured F or F respectively, I

- [kth the smallest margin or greatest ekcess N limit.

w l

engineering hot channel factor, F, applied to the I

- 1.03 is t toaccountformanufacturingtolehance.

measured q

to account for measurement I

- 1.05 is applied to the measured [q uncertainty.

- 1.04 is applied to the measured N t acc unt f r measurement l

H uncertainty.

Hotchannelfactors,[$ned,atand [ H, shall be measured and the targe 2.

flux difference determ equilibrium conditions according to the following conditions, whichever occurs first:

(a) At least once per 31 effective full-power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions after exceeding the reactor power at which target flux differance was last.

determined, by 10% or more of rated power.

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

o9thecore:

[ (equil) x V(Z) x 1.03 x 1.05 1(2.30/P) x K(Z)

)

where V(Z) is defined Figure 3.10-7 and other terms are l

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 flp trip setpoint by 1% for each percent that the 2

measured r or by 3.33% for each percent that the measured exceed the 3.10.B.1 limit. Then follow 3.10.B.3(c).

F'[H i

i (b) If the measured [o (equil) exceeds the 3.10.B.2 limits but not I

the 3.10.B.1 limit, take one of the following actions:

1 1.

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

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

l t

Prairie Island Unit No.1 - Amendment No. 35, ##, 66, 77 Prairie Island Unit No. 2 - Amendment No. 29,38,60,70

FIGURE TS.3.10-5 REV i

1.2 I

(6.0,1.0)

(12.0,0.92) s 1.0 (10.9,0.941) 1

)

,i Q$ \\

.8

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$gg

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0.6 Qg

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(12.0,0.452) 0.4 0.2

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6 8

10 12 Core Height (ft)

HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE Prairie Island Unit No.1 - Amendment No. 33, $$, 77 Prairie Island Unit No. 2 - Amendment No. 29, $0, 70

f\\

FIGURE TS.3.10-7 REV l

1.20 1.18

~

1.16 (11.25.,1.15) 1.14

[

1.12 j

/t (9.25,1.11)

V(Z) 1.10 1.08 1.06 1.04 1.02

' 1.00 O

2 4

6 8

10 12 Core Height (Ft)

V(Z) as a Function of Core Height Prairie Island Unit No.1 - Amendment No. 35, 77 Prairie Island Unit No. 2 - Amendment No. 29, 70

s TS.3.10-8 REV 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 power.

J.

DNB Parameters The following DNB related parameters limits shall be maintained during power operation:

a.

Reactor Coolant System Tavg 1564*F b.

Pressurizer Pressure 12220 psia

• c.

Reactor Coolant Flow 1,178,000 gpm With'any of the above parameters exceeding its limit, restore.the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal power to less than 5% of rated thermal power using normal shutdown procedures.

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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

Bases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and "RCCA(s)" are synonomous.

Shutdown Reactivity Trip shutdown reactivity is provided consistent with plant safety analyses assumptions.

One percent shutdown 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 accordingly.

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 2.1.30 for Exxon Nuclear fuel and 2J.17 for Westinghouse fuel during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding

• Limit not applicable during either a THERMAL POWER ramp increase in excess of (5%) RATED THERMAL POWER per minute or a THERMAL POWER step increase in excess of (10%) RATEL ~HERMAL POWER.

Prairie Island Unit No.1 - Amendment No. J6, J9,'((, 77 Prairie Island Unit No. 2 - Amendment No. JS, J3, 38, 70

.~

r TS.3.10-9 REV mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is

)

not exceeded.

During opegation,Nthe plant staff compares the measured hot channel i

transientkndLOCAa,na(describedlater)tothelimitsdeterminedinthe factors, F and F lyses. The terms on the right side of the j

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 10calheatfluxonthesurfaceofafuelroddividedbytheaverage heat flux in the core. Heat fluxes are derived from measured neutron fluxes and fuel enrichment.

The K(Z) function shown in Figure TS.3.10-5 is a normalized function that limits F axially. The K(Z) specified for the lowest six (6) feet l

n of the core.iH abritrarily flat since the lower part of the core is generally not limiting. Above that region, the K(Z) value is based on large and small break LOCA analyses.

V(Z) is an axially depgndent function applied to the equilibrium i

N measured F to bound F"'s that could be measured at non-equilibrium conditions 9 This funckion is based on power distribution control analyses that evaluated the effect of burnable poisons, rod position, axial effects, and xenon worth.

F Engineering Heat Flux Hot Channel Factor, is defined as the allow-abc,eonheatfluxrequiredformanufacturingtoler~ances. The engi-neering factor allows for local veriations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.

The 1.05 multiplier accounts for uncertainties associated with measure-ment of the power distribution with the moveable incore detectors and the use of those measurements to establish the assembly local power distribution.

F (equil) is the measured limiting F obtained at equilibrium conditions dOringtargetfluxdetermination.

O F, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio ohtheintegraloflinearpoweralongtherodwiththehighestintegrated power to the average rod power.

Prairie Island Unit No.1 - Amendment No. 35, ##, 66, 68, 77 Prairie Island Unit No. 2 - Amendment No. 29, 3$, 60, $2, 70

/.

/

8 TS.3.10-10 REV When a measurement of F is taken, measurement error must be allowed for l

AH and 4 percent is the appropriate allowance 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 pattarns. The periodic monthly incore mapping provides additional. assurance that the s

nuclear design bases remain inviolate and identify operational anomalies which would otherwise affect these bases.

For tsormal 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 1

inches from the bank demand position. An accidental i

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 are not violated.

4.

Axial power distribution control procedures, which are given in terms of flux difference control and control bank inser-tion 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 dif-ference in normalized power between the top and bottom halves of the core.

Prairie Island Unit No.1 - Amendment No. 35, ##, 77 Prairie Island Unit No. 2 - Amendment No. 29, SS, 70 9

+

2.m-a

- J

-4 a-

-ha l

-s 4

TS.3.10-11 REV and F allows for radial power The permitted relaxation in F shapechangeswithrodinsertkHon to the insertion limits.

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

In.specifica-tion 3.10, F is arbitrarily limited for P <0.5 (except for low powerphysicktests).

I The procedures for axial power distribution control referred to j

above are designed to minimize the effects of xenon redistribution 1

I on the axial power distribution during load-follow maneuvers.

Basically control of flux difference is required to limit the l

difference between the current value of Flux Difference (AI) and a reference value which corresponds to the full po'wer equilibrium value of Axial Offset (Axial Offset = AI/ fractional power). The reference value of flux difference varies with power level and j

i burnup but expressed as axial offset it varies only with burnup.

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

l developedOhichatalatertime,wouldcausegreaterlocalpower peaking even though the flux differen'ce is then within the limits specified by the procedure.

The target (or reference) value of flux difference is determined as 1

follows: At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with the full I

length rod control rod bank more than 190 steps withdrawn (i.e.,

normal full power operating position appropriate for the time in l

life, usually withdrawn f arther as burnup proceeds). This value, divided by the fraction of full power at which the core was oper-r ating 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 of 25 percent AI are permitted from the indicated reference value. Figure TS.3.10-6 shows the allowed deviation from the target flux difference as the function of thermal power.

i i

Prairie Island Unit No.1 - Amendment No. 35,##,66,77 Prairie Island Unit No. 2 - Amendment No. 29, 38, 60, 70 r

l i

.n

-+

---

,,,,..-y,,

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. ~.

l TS.3.10-13 REV i

l The consequences of I

i resulting from operation within the target band.

j being outside the 15% target band but within the Figure TS.3.10-6 limit for power levels between 50% and 90% has been evaluated and determined i

to result in acceptable peaking factors. Therefore, while the deviation l

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

In all cases the 15 percent target band is the Limiting Condition for Operation.

Only when the target band is violated do the limits under Figure TS.3.10-6 apply.

l If, for any reason, the indicated axial flux difference is not controlled within the t5 percent band for as long a period as one hour, then xenon t

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

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 j

as.possible. This is accomplished by using the boron system to position the full length control rods to produce the required indicated flux l;

difference.

l 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 l

protection system.

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

Quadrant Power Tilt Limits j

Quadrant power tilt limits are based on the following considerations. Fre-i quent power tilts are not anticipated during normal operation since this j

phenomenen 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. core limits protected per Specification 3.10.E.

A quadrant tilt by some other means (x-y xenon tran-sient, etc.) would not appear instantaneously, but would build up over several hcurs and the quadrant tilt limits are set to protect against this situation. They also serve as a backup protection against the dropped or misaligned rod.

i Operational experience shows that normal power tilts are lens than 1.01.

l Thus, sufficient time is available to recognize the presence of a tilt and correct the cause before a severe tilt could build up.

During start-up and powar escalation, however, a large tilt could be initiated.

i Therefore, the Technical Specification has been written so as to prevent escalacion above 50 percent power if a large tilt is present.

l;.

Prairie Island Unit No.1 - Amendment No. J6, ##, 77 Prairie Island Unit No. 2 - Amendment No. J9, 38, 70

r j

TS.3.10-17 REV 4

If the rod position deviation monitor and quadrant power tilt monitor (s) are inoperable, the overpower reactor trip setpoint is reduced (and also power) to ensure that adequate core prot 6ction is provided in the event that unsatisfactory conditions arise that could affect radial power distribution.

j Increased surveillance is required, if the quadrant power tilt monitors are inoperable and a load change occurs, in order to confirm satisfac-4 tory power distribution behavior. The automatic alarm functions related to quadrant power tilt must be censidered incapable of alerting the operator to unsatisfactory power distribution conditions.

DNB Parameters j

l The RCS flow rate, T and Pressurizer Pressure requirements are based

)

on transient analyse!'Es,sumptions. The flow rate shall be verified by calorimetric flow data and/or elbow taps.

Elbow taps are used in the l

reactor coolant system as an instrument device that indicates the status i

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 specifica-tion value indicated, shutdown is required to investigate adequacy of core cooling during operation.

l 1

i l

i 4

i i

i, Prairie Island Unit No. 1 - Amendment No. ##, 77 Prairie Island Unit No. 2 - Amendment No. 38, 70

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i NORTHERN STATES POWER COMPANY I

i DOCKET NO. 50-306 j

PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. DPR-60 j

1.

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

l l!

A.

The application for amendment by Northern States. Power Company (the licensee) dated January 13, 1986, supplemented March 25, 1986, complies with the standards and requirements of the Atomic i

Energy Act of 1954, as amended (the Act), and the Comission's i

rules and regulations set forth in 10 CFR Chapter I; B.

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

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public. and (ii) that such activities will be conducted in compliance with the

.Comission's regulations; D.

The issuance of this amendment will not be inimical to the 1

comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR

)

Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

j l

l

~

l

r 2-2.

Accordingly, the license is amended by changes to the. Technical

~

4 Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.

DPR-60 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 70, 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

/

George E. Lear, Director PWR Project Directorate #1 Division of PWR Licensing-A

Attachment:

Changes to the Technical Specifications Date of Issuance: April 3,1986 j

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4 ATTACHMENT TO LICENSE AMENDMENT NO. 70 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 Revise Appendix A Technical Specifications by removing the pages identified below and ihserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT TS-x TS-x TS.2.1-1 TS.2.1-1 TS.2.1-2 TS.2.1-2 TS.2.1-3 l

TS.2.3-6 TS.2.3-6 Figure TS.2.1-1 Figure TS.2.1-1 TS.3.1-17 TS.3.1-17 TS.3.1-18 TS.3.1-18 TS.3.3-1 TS.3.3-1 TS.3.10-1 TS.3.10-1 i

TS.3.10-2 TS.3.10-2 TS.3.10-8 TS.3.10-8 TS.3.10-9 TS.3.10-9 TS.3.10-10 TS.3.10-10 TS.3.10-11 TS.3.10-11 TS.3.10-13 TS.3.10-13 TS.3.10-17 TS.3.10-17 Figure TS.3.10-5 Figure TS.3.10-5 Figure TS.3.10-7 Figure TS.3.10-8 Figure TS.3.10-7*

j i

l

• Existing Figure TS.3.10-7 will be deleted by this change. Existing Figure TS.3.10-8 will be renumbered to TS.3.10-7.

l I

j s

TS-x REV _

APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE l

2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop I

Operation 3.1-1 Unit I and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit I and Unit 2 Reactor Coolant System Cooldown Limitations j

3.1-3 Effect of Fluence and Copper Content on Shift of RT NDT j

Reactor Vessel Steels Exposed to 550*F Temperature i

3.1-4 Fast Neutron Fluence (E >l MeV) as a Function of Full Power Service Life 3.1-5 DOSE EQUIVALENT 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 Prairit Island Nuclear Generating Plant Site Boundary for j

Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3.10-2 control Bank Insertion Limits 3.10-3 Insertion Limits 100 Step Overlap with One Bottomed Rod 3.10-4 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10-7 V(Z) as a Function of Core Height l

4.4-1 Shield Building Design In-Leakage Rate 6.1-1 NSP Corporation Organization Relationship to On-Site Operating Organizations 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization j

for On-Site Operating Group 4

4 i

Prairie Island Unit 1 - Anendment No. 62, 69, 66, #S, 73, 77 Prairie Island Unit 2 - Anendment No. (6, 63, 60, 64, 66, 70 t

I i

s I

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

-l 2.1 SAFETY LIMIT, REACTOR CORE i

Applicability 1

Applies to the limiting combinations of thermal power, reactor coolant system pressure and coolant temperature during operation.

Objective

]

To maintain the integrity of the fuel cladding.

1 Specification i

1.

The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure TS.2.1-1.

The I

safety limit is exceeded if the point defined by the combination of reactor coolant system average temperature and power level is at any time above the appropriate pressure line.

Basis To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions.

This is accomplished by operating the hot regions of the core within the nucleate boiling regime of heat j

transfer wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation temperature.

The upper boundary of the nuc1'eate boiling regime is termed departure from nucleate boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of I

clad failure. DNB is not, however, an obscrvable parameter during reactor operation. Therefore, the observable parameters; thermal power, I

reactor coolant temperature and pressure have been related to DNB i

through the W-3 and WRB-1 DNB correlations.

The W-3 DNB correlation is used for Exxon Nuclear fuel. The WRB-1 DNB correlation is used for Westinghouse fuel. The W-3 and WRE-1 DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNB ratio DNBR, during steady state operation, normal operational transients, and anticipated transients is limited to 1.30 for the Exxon Nuclear fuel and to 1.17 for the Westinghouse fuel.

l These limits correspond to a 95% probability at a 95" confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for i

all operating conditions.

1 l

Prairie Island Unit 1 - Amendment No. 77 Prairie Island Unit 2 - Amendment No. 70

s TS.2.1-2 REV The solid curves of Figure TS 2.1-1 represent the loci of points of thermal power, coolant pressure, and coolant average temperature for which either the coolant enthalpy at the core exit is limiting or the DNB ratio is limiting. For the 1685 psig and 1985 psig curves, the coolant average enthalpy at the core exit is equal to saturated water enthalpy below 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. For all four curves, the DNBR is limiting at higher power l

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 the self-actuated safety valves on the steam generators. The highest nominal setting of the steam generator safety 1

valves is 1129 psig (saturation temperature 560*F). At zero power the l

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 4

l tests will be determined as a part of the test procedure.

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

F

= 1.60 [1 + 0.3(1-P)] ; and F = 2.30 H

q Use of these factors results in more conservative safety limits than 4

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 i

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 j

specified by Figure TS.3.10-1 assure that the DNB ratio 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 i

I a DNB ratio of less than 1.30 for Exxon Nuclear fuel and less than 1.17 i

for Westinghouse fuel.

i Prairie Island Unit 1 - Amendment No. 77 Prairie Island Unit 2 - Amentment No. 70 i

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1 SAFETY LIMITS, REACTOR CORE, THERMAL AND HYDRACIC TWO-LOOP OPERATION i

FIG"JRE T5.0.1-1 Prairie Island Unit 1 - Anendment No. 77

_ _ - - --Pr.ai..rie Island Unit 2 _ Amendment No. 70, _ _ _ _ _ _. _. _ _ _ _ _ _. _ _ - _. _ _ _ _

3.

s 1

TS.2.3-6 REV The other reactor trips specified in A.3.

above provide additional protection.

Th2 trip initiated by steam /feedwater flow mismatch in coincidence with low steam gsnerator water level is designed for protection from a sudden loss o'T the

'rocctor's heat sink. The safety injection signal trips the reactor to decrease t

tha severity of the accident condition. The reactor is tripped when the turbinc gsnerator trips above a power level equivalent to the load rejection capacity of tha steam dump valves. This reduces the severity of the loss-of-load transient.

4

}

Th2 positive power range rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power livel.

Specifically, this trip compliments the power range nuclear flux l

high and low trip to assure that the criteria are met for rod ejection from partial power.

t l

Th2 negative power range rate trip provides protection satisfying all IEEE criteria to assure that minimum DNBR is maintained above 1.30 for Exxon Nuclear I

fual and above 1.17 for Westinghouse f2el for all multiple control rod drop i

tecidents. Analysis indicates.(Section 14.1.3) that in the case of a single rod drop, a return to full power will be initiated by the automatic reactor control l

l cystem in response to a continued full power turbine load demand and it will not I

result in a DNBR of less than 1.30 for Exxon Nuclear fuel or 1.17 for Westinghouse fuel. Thus, automatic protection for a single rod drop is not required. Admini-strative limits in Specification 3.10 require a power reduction if design power j

distribution limits are exceeded by a single misaligned or dropped rod.

i f

i I

Roferences:

)

(1) FSAR 14.1.1 l

(2) FSAR Page 14-3 i

(3) FSAR 14.2.6 j

(4) FSAR 14.3.1 (5)

FSAR 14.1.2 (6)

FSAR 7.2, 7.3 i

(7)

FSAR 3.2.1 (8) FSAR 14.1.9

}

(9)

FSAR 14.1.11 i

i Prairie Island Unit 1 - Amendment No. 33, 77 Prairie Island Unit 2 - Amendment No. 27, 70 q

i i

.')

o..

1.18.

I

_....._l_.._-

i,

.r 1.16

=

i i

._:I

.li ~

.. _. L ( -11.25.1.15)

-.._t

-! n:

,. 1, t

a

.,c.

t.

1*12

=

i :i; -l (9.25.1.11)

/::!;

.I--

l.: :r*:.

1,.

. ['.

..l :. - - -

..r:.

n...: =

V(Z) 1.10

.,s :...

s. -.--....

-.t :.

. =. :..=. :

,.[.-i; u= =.:

1.08

.p

..:.:l.:-

l

.:::g:: -

'.r.

.t.

1.06

-+.

.,l.

gg

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oe 1.04

.7{..;;.

.1;

t:..

... ;l.;

sa

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aa em 73

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_s s Core Height (Ft)

H MM ea e

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V(Z) as a Function of Core Height w

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TS.3.1-17 J

REV

(

F.

MINIMUM CONDITIONS FOR CRITICALITY Specification

(

1.

Isothermal Temperature Coefficient (ITC)

~

a.

When the reactor is critical, the isothermal temperatttre coefficient l

shall be less than 5 pcm/*F with all rods withdrawn, except during low power physics tests and as specified in 3.1.F.1.b and c.

b.

When the reactor is above 70 percent rated thermal power with all rods withdrawn, the isothermal temperature coefficient shall be negative, except as specified in 3.1.F.1.c.

c.

If the limits of 3.1.F.1.a or b cannot be met, Power Operation may continue provided the following actions are taken:

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

limits shall be in addition to the insertion limits of Figure TS.3.10-2.

(2) Maitcain the control rods within the withdrawal limits er:ablished above until a subsequent calculation verifies that the ITC has been restored to within its limit for the all rods withdrawn condition.

(3) Submit a special report to the Commission within 30 days, t

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.

2.

The reactor shall not be brought to a critical condition until the reactor coolant temperature is higher than that defined by the criti-cality limit line shown in Figure TS.3.1-1.

l Basis At the beginning of a fuel cycle the moderator temperature coefficient has its most positive or least negative value. As the boron concentration is i

reduced throughout the fuel cycle, the moderator temperature coefficient becomes more negative. The isothermal temperature coefficient is defined l

as the reactivity ahange associated with a unit change in the moderator and fuel temperatures. Essentially, t.he isothermal temperature coefficient is the er. of the moderator and fuel temperature coefficients.

This co-i i

efficient is measured directly during low power physics tests in order to g

verify analytical predictions. The units f the isothermal temperature g

coefficient are pcm/*F, where 1 pcm = 1x10 AK/K.

Eg E

[

Prairie Island Unit No.1 - Amendment No. 33, 77 Prairie Island Unit No. 2 - Amendment No. 29, 70

g TS.3.1-18 REV For extended optimum fuel burnup it is necescary to either load the reactor with burnable poisons or increase the boron concentration in the reactor coolant system.

If the latter approach is emphasized, it is possible that a positive isothermal temperature coeffieient.could exist at beginning of cycle (BOC).

Safety analyses verify the accep-tability of the isothermal temperature coefficient for limits specified in 3.1.F.1.

Other conditions, e.g., higher power or partial rod insertion would cause the isothermal coefficient to have a more negative value.

These analyses demonstrate that applicable criteria in the NRC Standard Review Plan (NUREG 75/087) are met.

Physics measurements and analyses are conducted during the reload startup test program to (1) verify that the plant will operate within safety analyses assumptions and (2) establish operational procedures to ens'ure safety analyses assumptions are met.

The 3.1.F.1 requirements are waived during low power physics tests to permit measurement of reactor temperature coefficient and other physics design parameters of interest.

Special operating precautions will be taken during these (3) physics tests.

In addition, the strong negative Doppler coefficient and the small integrated dk/k would limit the magnitude of a power excurston resulting from a reduction of moderator density.

The requirement that the reactor is not to be made critical except as specLfied in Figure TS.3.1-1 provides increased assurance that the proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility temperature range. Heatup to this temperature will be accomplished by operating the reactor coolant pumps and by.the pressurizer heaters.

The pressurizer heater and associ-ated power cables have been sized for continuous operation at full heater power. The shutdown margin in Specification 3.10 precludes the possibility of accidental criticality as a result of an moderatortemperatureoradecreaseofcoolantpressure.{gyreaseof l

References:

(1) FSAR Figure 3.2-10 (2)

FSAR Table 3.2-1 Prairie Island Unit No.1 - Amendment No. 35, 77 Prairie Island Unit No. 2 - Anendment No. 29, 70

TS.3.3-1 REV s

3.3 ENGINEERED SAFETY FEATURES 1

Applicability Applies to the operating status of the engineered safety features.

Obj ective To define those limiting conditions that are necessary for operation of engineered safety features:

(1) to remove decay heat from the' core in an emergency or normal shutdown situations, and (2) to remove heat from containment in normal operating and emergency situations.

Specifications A.

Safety Injection and Residual Heat Removal Systems

1. A reactor shall not be made or maintained critical nor shall it be heated or maintained above 200*F unless the following condi-tions are satisfied except as permitted in Specification 3.3.A.2.

a.

The refueling water tank contains not less than 200,000 gallons of water with a boron concentration of at least 1950 ppm.

b.

Each reactor coolant system accumulator shall be operable when reactor coolant system pressure is greater than 1000 psig.

Operability requires:

(1) The isolation valve is open (2) Volume is 1270 1 20 cubic feet of borated water

[

(3) A minimum boron concentration of 1900 ppm (4) A nitrogen cover pressure of at least 700 psig Two safety injection pumps are operable except that pump c.

control switches in the control room shall meet the require-ments of Section 3.1.G whenever the reactor coolant system temperature is less than MPT.

d.

Two residual heat removal pumps are operable.

Two residual he_at exchangers are operable.

e.

f.

Automatic valves, interlocks and piping associated with the above components and required to function during accident conditions, are operable.

g.

Manual valves in the above systems that could (if one is improperly positioned) reduce injection flow below that assumed for accident analyses, shall be blocked and tagged in the proper position for injection. RHR system valves, however, may be positioned as necessary to regulate plant heatup or cooldovn rates when the reactor is suberitical. All changes in valve position shall be under direct administrative centrol.

Prairie Island Unit No.1 - Amendment No. 38, 67, 77 Prairie Island Unit No. 2 - Amendment No. 32, 55, 70

TS.3.10-1 REV 3.10 CONTROL R0D AND POWER DISTRIBUTION LIMITS Applicability Applies to the limits on core fission power distribution and to the limits on control rod operations.

~

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

Specification A.

Shutdown Reactivity The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown in Figure TS.3.10-1 under all 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 dgring igw power physics testing, measured hotchannelfactors,TyandFAH, as defined below and in the bases, shall meet the following limits:

F x 1.03 x 1.05 3(2.30/P) x K(Z)

N F

x.04 s1.60 x [1+ 0.3(1-P)]

H where~the following definitions apply:

- K(Z) is the axial dependence function shown in Figure TS.3.10-5.

- Z is the core height location.

- P is the fraction of rated power at which the core is operating.

In the limit determination when P 5 50, set P = 0.50.

(2.21/P) shall be used for Westinghouse Standard assemblies for Unit 2 Cycle 10.

• • 1.55 x [1 + 0.2 (1-P)] shall be used for Westinghouse Standard assemblies for Unit 2 Cycle 10.

Prairie Island Unit No.1 - Amendment No. $$, ($, $$, 77 Prairie Island Unit No. 2 - Amendment No. 29, 38, 69, 70

s TS.3.10-2 REV _

is defined as the s:easured F or F respectively.

'l or [3

- [9th the smallest margin er greatest ekcess o9 limit.

3 w

E

- 1.03 is thg engineering hot channel factor, F, applibd to'the I

measuredTftoaccountformanufacturingtole9ance.

- 1.05 is applied to the measured [q to account for measurement I

uncertainty.

- 1.04 is applied to the measured N t account for measurement j

H uncertainty.

Hotchannelfactors,[9ned,atandFfH,shallbemeasuredandthetarget 2.

flux difference determ equilibrium conditions according to the following conditions, whichever occurs first:

(a) At least once per 31 effective full-power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions af ter exceeding the reactor power at which target flux differance was last.

determined, by 10% or more of rated power.

[kthecore:

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

o

[ (equil) x V(Z) x 1.03 x 1.051(f. 30/P) x K(Z) where V(Z) is defined Figura 3.10-7 and other terms are l

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 f1p trip setpoint by 1% for each percent that the measured r or by 3.33% for each percent that the measured T']H exceed 9the3.10.B.1 limit. Then follow 3.10.B.3(c).

(b) If the measured [9,(equil) exceeds the 3.10.B.2 limits but not the 3.10.B.1 limi take one of the following actions:

1.

'n'ithin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> place the reactor in an equilibrium configuration for which Specification 3.10.B.2 is satis-fied, or i

2.

Reduce reactor power and the high neutron flux trip sgtpoint by 1% for each percent that the measured 9 (equil) x 1.03 x 1.05 x V(Z) exceeds the limit.

l 7

Prairie Island Unit No.1 - Amendment No. 36,(4,66,77 Prairie Island Unit No. 2 - Amendment No. 29, 38, 60, 70

g FIGURE TS.3.10-5 REV l

i 1.2 I

(6.0,1.0)

(12.0,0.92:

1.0 o

(10.9,0.941) i

\\

\\$ \\

.8 I,E I

$4 K(Z)

=%\\

e ck?zs og \\

0.6 fQ~\\

\\

l l

\\

(12.0.0.452) i j

0.4 O.2 0.0 0

2 4

6 8

10 12 Core Height (ft)

HOT CHAMTEL FACTOR NORMALIZED OPERATING EhTELOPE Prairie Island Unit No. 1 - Amendment No. 33, 66, 77 Prairie Island Unit No. 2 - Amendment No. 29, 60, 70

.l FIGURE TS.3.10-7 REV

~ l

)

1.20 1.18 1.16 (11.25.1.15)r 1.14

/

1.12

/

/'

(

}

V(Z) 1.10 1.08 1.06 1.04 1.02 1.00 0

2 4

6 8

10 12 Core Height (Ft)

V(Z) as a Function of Core Height Prairie Island Unit No. 1 - Amendment No. 29, 44, 56. EB, 67, 77 Prairie Island Unit No. 2 - Amendment No. 73, $$, EP, EZ, 57, 70

s TS.3.10-8 REV 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 houra after a load change greater than 10% of rated power.

J.

DNB Parameters The following DNB related parameters limits shall be maintained during power operation:

a.

Reactor Coolant System Tavg 1564*F b.

Pressurizer Pressure

>2220 psia

• c.

Reactor Coolant Flow

>178,000 gpm

. With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal power to less than 5% of rated thermal power using normal shutdown procedures.

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

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

Bases Throughout the 3.10 Technical Specifications, the terms " rod (s)" and "RCCA(s)" are synonomous.

Shutdown Reactivity Trip shutdown reactivity is provided consistent with plant safety analyses assumptions.

One percent shutdowa 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 accordingly.

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 >1.30 for Exxon Nuclear 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

• Limit not applicable during either a THERMAL Pok'ER ramp increase in excess of (5%) RATED THER%\\L P0k'ER per minute or a THERMAL P0k'ER step increase in excess of (100) RATE: THERMAL P0k'ER.

Prairie Island Unit No.1 -- Amendment No. J6, 79, (J, 77 Prairie Island Unit No. 2 - Amendment No. JB, J3, 38, 70

s' l

TS.3.10-9 i

REV J

mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the,LOCA i

analyses are met and the ECCS acceptance criteria limit of 2200'F is i

not exceeded.

i During op je ation,Nthe plant staff compares the measured hot channel transientOndLOCkH,(describedlater)tothelimitsdeterminedinthe f actors, F-~ and F I

analyses. The terms on the right side of the j

equations in Section 3.10.B.1 represent the analytical limits. Those 4

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

[0calheatfluxonthesurfaceofafuelroddividedbytheaverage is the measured Nuclear Hot Channel Factor, defined as the maximum 1

heat flux in the core. Heat fluxes are derived from measured neutron fluxes ard fuel enrichment, The K(Z) function shown in Figure TS.3.10-5 is a normalized function that limits F axially. The K(Z) specified for the lowest six (6) feet l

n of the core iM abritrarily flat since the lower part of the core is generally not limiting. Above that region, the K(Z) value is based on large and small break LOCA analyses.

1 V(Z) is a axially depgndent function applied to the equilibrium measured to bound F 's that could be measured at non-equilibrium j

conditions 9 This funckion is based on power distribution control analyses that evaluated the effect of burnable poisons, rod position, axial effects, and xenon worth.

i 7, Engineering Heat Flux Hot Channel Factor, is defined as the allow-i a0ceonheatfluxrequiredformanufacturingtolerances. The engi-nearing factor allows for local veristions in enrichment, pellet-

{

density and diameter, surface area of the fuel rod and eccentricity of i

the gap between pellet and clad. Combined statistically the net effect j

is a factor of 1.03 to be applied to fuel rod surface heat flux.

i l

The 1.05 multiplier accounts for uncertainties associated with measure-ment of the power distribution with the moveable incore detectors and the use of those measurements to establish the assembly local power l

distribution.

[hringtargetfluxdetermination.(equil) is the measured limiting [0 obtained at equilibrium conditions j

d

)

[ftheintegraloflinearpoweralongtherodwiththehighestintegrated

, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio j

o i

power to the average rod power.

i Prairie Island Unit No. 1 - Amendment No. 36, (4, 66, 68, 77 Prairie Island Unit No. 2 - Amendment No. 29, 38, 69, 62, 70 i

..4

s TS.3.10-10 REV When a measurement of F is taken, measurement error must be allowed for l

H and 4 percent is the appropriate allowance 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 ir.;0re 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.

j l

2.

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

3.

The control bank insertion limits are not violated.

4.

Axial power distribution control procedures, which are given in terms of flux difference control and control bank inser-tion 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 dif-ference in normalized power between the top and bottom halves of the core.

Prairie Island Unit No.1 - Amendment No. 35, Ad, 77 Prairie Island Unit No. 2 - Amendment No. 29, 3S, 70

s TS.3.10-11 REV i

N N

and F allows for radial power The permitted relaxation in F,H shapechangeswithrodinsertiontotheinsertionlimits.

It has been determined that provided the above conditions I through 4 are observed, thgse hot channel factor limits are met.

In specifica-tion 3.10, F is arbitrarily limited for P 10.5 (except for low I

powerphysichtests).

The procedures for axial power distribution control referred to above are designed to minimize the effects of xenon redistribution j

on the axial power distribution during load-follow maneuvers.

Basically control of flux difference is required to limit the J

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 l

developedOhichatalatertime,wouldcausegreaterlocalpower peaking even though the flux difference is then within the limits specified by the precedure.

l l

The target (or reference) value of flux differance is determined as follews: 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 oper-ating 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 indicafkd equilibrium was noted, no allowances for excore detector error are i

necessary and indicated deviation of t5 percent AI are permitted

{

from the indicated reference value. Figure TS.S.10-6 shows the allowed deviation from the target flux difference as the function of thermal power.

]

1 Prairie Island Unit No.1 - Amendment No. 35,##,66,77 Prairie Island Unit No. 2 - Amendment No. 29, 38, 60, 70

s

)

TS.3.10-13 i

REV l

resulting from operation within the target band. The consequences of being outside the 15% target band but within the Figure TS.3.10-6 limit for power levels between 50% and 90% has been evaluated and determined-l to result in acceptable peaking factors. Therefore, while the deviation l

exists the power level is limited to 90 percent or lower depending on I

the indicated axial flux difference.

In all cases the 15 percent target band is the Limiting Condition for Operation.

Only when the target band j

is violated do the limits under Figure TS.3.10-6 apply.

I If, for any reason, the indicated axial flux difference is not controlled l

within the 15 percent band for as long a period as one hour, then xenon i

distributions may be significantly changed and operation at 50 percent is f

required to protect against potentially more severe consequences of some 1

accidents.

\\

j 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 t

as possible.

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

difference.

I For Condition II events the core is protected from overpower and a minimum j

DNBR of 1.30 for Exxon fuel and 1.17 for Westinghouse fuel by an automatic l

protection system. Compliance with operating procedures is assumed as a precondition for Condition II transients, however, operator error and equip-ment malfunctions are separately assumed to lead to the cause of the transients il considered.

l 1

)

Quadrant Power Tilt Limits Quadrant power tilt limits are based on the following considerations.

Fre-quent power tilts are not anticipated during normal operation since this phenomenen is caused by some asymmetric perturbation, e.g. rod misalignment.

i x-y xenon transient, or inlet temperature mismatch. A dropped or misaligned i

rod will easily be detected by the Rod Position Indication System or core.

t instrumentation per Specification 3.10.F. and core limits protected per Specification 3.10.E.

A quadrant tilt by some other means (x-y xenon tran--

1 sient, etc.) would not appear instantaneously, but would build up over i

several hcurs and the quadrant tilt limits are set to protect against this situation. They also serve as a backup protection against the dropped or misaligned rod.

ll Operational experience shows that normal power tilts are lens than 1.01.

I Thus, sufficient time is availabic to recognize the presence of a tilt 4

and correct the cause before a severe tilt could build up.

During atart-l up and power escalation, however, a large tilt could be initiated.

1 Therefore, the Technical Specification has been written so as to prevent I

escalation above 50 percent power if a large tilt is present.

I 1

Prairie Island Unit No.1 - Amendment No.16, #4, 77 Prairie Island Unit No. 2 - Amendment No. J0, 38, 70

s s

TS.3.10-17 REV If the rod position deviation monitor and quadrant power tilt monitor (s) are inoperable, the overpower reactor trip setpoint is reduced (and also power) to ensure that 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 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 the operator to unsatisfactory power distribution conditions.

f DNB Parameters The RCS flow rate, T and Pressurizer Pressure requirements are based ontransientanalyse!"Es,sumptions. The flow rate shall be verified by calorimetric flow data and/or elbcv 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 provide information as to whether or not a reduction in flow rate has occurred. If a reduction in flow rate is indicated below the specifica-tion value indicated, shutdown is required to investigate adequacy of core cooling during operation.

i t

4 i

Prairie Island Unit No. 1 - Amendment No. ##, 77 Prairie Island Unit No. 2 - Amendment No. 38, 70 1

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