ML20207J555

Amends 84 & 77 to Licenses DPR-42 & DPR-60,respectively, Changing Tech Specs by Increasing Hot Channel Enthalpy Factor & Total Peaking Factor Fq to 1.70 & 2.50

ML20207J555
Person / Time
Site:	Prairie Island
Issue date:	09/16/1988
From:	Diianni D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20207J557	List: ML20207J555 ML20207J572 ML20207J579
References
NUDOCS 8809280040
Download: ML20207J555 (9)

Text

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION o

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WASHINGTON, D. C. 20666

%,..... p' NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 84 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 July 5, 1988, as supplemented August 5, 1988, 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-42 is hereby amended to read as follows:

8809260040 880816,

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Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 84, 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 m

Dominic C. Dilanni, Acting Director Project Directorate 111-1 Division of Reactor Projects - III, IV, V

& Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 16, 1988

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UNITED STATES 8

NUCLEAR REGULATORY COMMISSION n

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WASHING TON, D. C. 20665

• %,.....,f NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 77 License No OPR-60 1.

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

A.

The application for amendment by Northern States Power Company (the licensee) dated July 5, 1988, 's supplemented August 5, 1988, 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. OPR-60 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.

FOR THE NUCLEAR REGULATORY COMMISSION Q

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Dominic C. Dilanni, Acting Director Project Directorate !!I-1 Division of Reactor Projects - III, IV, V

& Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 16, 1988

ATTACHMENT TO LICENSE AMENDMENT NOS.gg 77 FACILITY OPERATING LICENSES NOS. DPR-42 AND OPR-60 00CKETS NOS. 50-282 AND 50-306 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

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

REMOVE INSERT TS-x TS-x TS.3.10-1 TS.3.10-1 TS.3.10-2 TS.3.10-2 TS.3.10-9 TS.3.10-9

TS x APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES r

I TS FIGURE TITLE 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 a-'. "-it 2 Reactor 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 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 L

Gaseous Effluents i

3.10 1 Required Shutdown Reactivity Vs Reactor Boron Concentration l

3.10 2 control Bank Insertion Limits 3.10 3 Insertion Limits 100 Step overlap with One Bottomed Rod 3.40 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 Corporate Organizational Relationship to On Site Operating Organizations 6.1 2 Prairie Island Nuclear Generating Plant Functional Organization for On site Cperating Group i

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Prairie Island Unit 1 Anendment No. JJ.JJ,79,7J,77,79.JJ 37 i

PrairieIslandUnit2AmendmentNo.JJ,J9,J#,JJ,79,7J,7#//

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

Obiectives 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.

Seecification 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 nuberitical at hot shutdown condicions 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 Pfg, as defined below and in the bases, shall meet the ollowing limits:

Ffx1.03x1.05$(2.50/P)K(Z)

Ffgx1.04$1.70x[1+0.3(1P))

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 Ps.50, set P - 0.50.

Prairie Island Unit 1 Amendment No. JJ ##,##,77,77. 84.

Prairie Island Unit 2 Amendment No. 77 JJ J9,79,7# 1/

TS.3.10-2

-FNor7$H is defined as the measured F or FA wkththesmallestmarginorgreatestekeessofrespectively, limit.

l engineering hot channel factor Th, applied to the l

- 1.03 is th measured to account for manufacturing tolerance.

l N

1.05 is applied to the measured F to account for measurement g

uncertainty.

l 1.04isappliedtothemeasuredF$gto.

for measurement uncertainty.

2.

Hot channel factors, En andF$H,shallbe ad the target f

flux difference determined, at eqciUW

',,.us according l

to the following conditions, chew-it:

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(a) At least once per 31 effs.t.

f 2.ays in conjunction with the target flux differe-t.

.n, or (b) Upon reaching equilibrium condi.

. N r exceeding the l

reactor power at which target fh orference was last determined, by 10% or pure of rated power.

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F$(equil)shallmeetthefollowinglimitforthemiddleaxial804

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of the core:

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l'h(equil)xV(2)x1.03x1.055(2.50/P)xK(Z) l where V(Z) is defined Figure 3.10 7 and other terms are j

defined in 3.10.8.1 above.

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3.

(a) If either measured hot channel factor exceeds its limit j

specified in 3.10.B.1, reduce reactor power and the high i

neutron f1 x trip setpoint by 14 for each percent that the

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mgasured ) or by 3.334 for each percent that the measured i

F"LH exceed the 3.10.B.1 limit. Then follow 3.10 B.3(c).

(b) If the measured No (equil) exceeds the 3,10.B.2 limits but not f

the 3.10.B.1 11mie, take one of the following actions:

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.

j fled, or l

t 2.

Reduce reactor power and the high neutren flux trip j

s tpoint by 14 for each percent that the measured

[

(equil) x 1.03 x 1.05 x V(2) exceeds the lim t.

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i Prairie Island Unit 1 Amendment No. 77,##,((,77,U, 84 Pratrie Island Unit 2 Amendment No. 77.JJ.J9,79,7#,77

TS.3.10-9 mechanical properties to within assumed design cri iria.

The ECCS analysis was performed in accordance with SECY 83 472. Ons calculation at the 95%

j 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 1

calculation used a peak linear heat generation rate of 14.2 kw/ft. The Appendix K calcolation used a peak linear heat generation rate of 15.8 kw/ft for the F limit of 2.5.

Maintaining 1) peaking factors below the F o

q limit of 2.5 during all Condition I events and 2) the peak linear heat generation rate below 14.2 kv/ft at the 954 probability level assures compliance with the ECCS analysis.

During ope ation, the plant staff compares the measured hot channel j

factors, q and E, (described later) to the limits determined in the aH i

transient and LOCA analyses. The terms on the right side of the J

equations in Section 3.10.B.1 represent the e.nalytical limits. Those l

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

q l

ThisthemeasuredNuclearHotChannelFactor,definedasthemaximum local heat flux on the surface of a fuel rod divided by the average l

heat flux in the core. Heat fluxes are derived from measured neutron j

fluxes and fuel enrichment.

j The K(Z) function shown in Figure TS.3.10 5 is a normalized function j

that limits Fq axially. The K(Z) specified for the lowest six (6) feet

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of the core is arbitrarily flat since the lower part of the core is l

l generally not limiting. Above that region, the K(Z) value is based on j

small break LOCA analyses.

l 1

i V(2) is an a~ially dep ndent function applied to the equilibrium measured F#y bound

's that could be measured at nonequilibrium i

conditions.

This fune ion is based on power distribution control j

analysis that evaluated the effect of burnable poisons, rod position.

l axial effects, and xenon worth, f

Fh,EnrineerineHeat Flux Hot Channel Factor, is defined as the allow-ance on heat flux required for manufacturing tolerances.

The engi.

i neering factor allows for local variations in enrichment, ps11et l

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 multipli... accounts for uncertainties associated with measure-ment of the pruer distribution with the movable incore detectors and l

the use of tho!.= measurements to establish the assembly local power distribution.

N FO (equil) is the measured limiting E obtained at equilibtium conditions ddring target flux determination.

E/g,NuclearEnthalovRiseHotChannelTactor, is defined as the ratio of the integral of linear power along the rod with the highest incegrated power to the average rod power.

Prairie Island Unit 1 Amendment No. 77.##.1#.97.77, 84 Prairie Island Unit 2 Amendment No. 77,77,79,77,79, 77