ML20078G688
| ML20078G688 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/03/1983 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078G691 | List: |
| References | |
| NUDOCS 8310130076 | |
| Download: ML20078G688 (11) | |
Text
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o UNITED STATES NUCLEAR REGULATORY COMMISSION o
h WASHINGTON, D. C. 20555 k +... + p$
NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 66 License No. DPR-42 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Northern States Power Company (thelicensee)dat'edJune 24, 1983 as revised July 29, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the 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.
e 8310130076 831003 PDR ADDCK 05000282 P
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.. 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. 66, are hereby in,
corporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR RE ULATORY COMMISSION C
James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 3, 1983 t
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I UNITED STATES h
NUCLEAR REGULATORY COMMISSION 3
j WASHINGTON. D. C. 20555 e
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 60 License No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by "srthern States Power Company (the licensee) dated June 24, 1983 as revised July 29, 1983, 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 (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 i
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.
--.y
)
.. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendnent, 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. 60, are hereby in-corporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR GULATORY COMMISSION James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 3, 1983
_,-,_m...
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NOS. 66 AND 60 TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of changes.
Remove Insert TS-iv TS-iv 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.3.10-11 TS.3.10-11 Figure TS.3.10-5 Figure TS.3.10-5 4
TS-iv APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES 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 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' 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 >1.0 pCi/ 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 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 Normalized Exposure Depencent Function BU(E ) for Exxon Nuclear Company Fuel 3.10-8 V(Z) as a Function of Core Height 4.4-1 Shield Building Design In-Leakage Rate 4.10-1 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 4.10-2 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 6.1-1 NSP Corporate Organizational Relationship to On-site Operating Organization 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-site Operating Group Unit 1 - Amendment No. 52, $/g, 66 Unit 2 - Amendment No. 46, fry, 60 e
TS,3.10-1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Appilcability 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 in-sertions 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 powr.r, including effects of dxial 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 lgw power physics testing, measured hot channel factors, F and Fg, as defined below and in the O
bases, shall meet the rollowing limits:
F" x 1.03 x 1.05 1 (2 32/P)*x K(Z) x BU (E )
q F
1 H*
where the following definitions apply:
(a) K(Z) is the axial dependence function shown in Figure TS.3.10-5.
(b) Z is the core height location.
(c) E is the maximum pellet exposure in fuel rod j for which 3
N tne F is being measured.
q (d) BU(E ) is the normalized exposure dependence function for 4
Exxos. Nuclear Company fuel shown in Figure TS.3.10-7.
For Westinghouse fuel, BU(E ) = 1.0 (e) P is the fraction of full power at which the core is operating.
N In the F limit determination when P-1 50, set P = 0.50.
q
- (2.21/P) shall be used for Westinghouse assemblies.
l Prairie Island Unit 1 - Amendment No. 35, 44, 66 Prairie Island Unit 2 - Amendment No. 29, 38, 60 i
TS.3.10-2 N
N N
N (f)
F rF is defined, as the measured F or Fg, respectively, o
AH n
wIth the smallest. margin'or greatest eMeess or limit (g) 1.03 is the engi cring hot channel factor, F, applied to the measured toaccountformanufacturibgtolerance.
N (h) 1.05 is applied to the measured F to account for cieasurement q
uncertainty.
N (i) 1.04 is applied to the measured F to account for measure-AH ment uncertainty N
N 2.
Hot chanr.el factors, F and F shall b'e measured and the target fluxdifferencedeterm9ned,ak"e,quilibriumconditionsaccording to the following conditions, whichever occurs first:
(a) At least once per 31 ef fective 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 power.
F (equil) shall meet the following limit for the middle axial 80%
o9thecore:
N q (equil) x V(Z) x 1.03 x 1.05 < (2.32/P) x K(Z) x BU(E )
F where V(Z) is defined Figure 3.10-8 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 neutgon flug trip set-q or F 'd
- d' point by 1% for each percent that the measured F the 3.10.B.1 limit. Then follow 3.10.B.3(c).
the 3.10.B.1 limik,(equil) exceeds the 3.10.B.2 limits but not (b)
If the measured F 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 /> placc the reactor in an equilibrium c-afigura-tion for which Specification 3.10.B.2 is satisfied, or 2.
Reducereactorpowerandthehighneutrongluxtripsetpoint x1.05xV(Z)exceedsthe(2.32/P)xK(Z)k(equil)x1.03 l
by 1% for each percent that the measured F BU(E ) limir..
Praitle Idand Unit 1 - Amendment No. $$, $$, 66 Prairie Island Unit 2 - Amendment No. 29,38, 60
1 TS.3.10-9 i
mechanical properties to wi. thin assumed design criteria.
In addition, limiting the peak lin**r pawar density during Cendition i events pro-vides assurance that the initial conditions assumed fog the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.
1 During opagation,Nthe plant staff compares the measured hot channel i
factors, F and F (described later) to the limit determined inthetraksientNdLOCAanalyses. He limiting F (Z) includes measurement, O
engineering, and calculational uncertainties. He 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 (Z), Height Dependent Heat Flux Hot Channel Factor, is defined as the n
m3ximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.
The maximum value of F (Z) is 2.32/P l
0 for the Prairie Island reactors. This value is restricted f0rther by the, K(Z) and BU(E ) functions d.escribed below.
He product of these three factors is j
F (Z).
9 The K(Z) function shown in Figure TS.3.10-5 is a normalized function that 4
j limits F (Z) axially for three reasons. He K(Z) specified for the 9
lowest stx (6) feet of the core is based on large break LOCA analyses, a
j Above this region the K(Z) value is based on DNBR requirements since the minimum DNBR would be expected in this region of the core, based on power, pressure, and tersperature.
The K(Z) value in the uppermost region of the core is based on the small break LOCA analyses.
F (Z) 0 i
in the uppermost region is limited to reduce the PCT expected during a small break LOCA since this region of the core is expected to uncover temporarily for some small break LOCA's.
The BU(E.) function shown in Figure TS.3.10-7 is a normalized function that lia}ts F (Z) based on exposure dependent analyses for the ENC fuel.
0 hese analyses consider pin internal pressure uncertainties, fuel swelling, rupture pressures, and flow blo,ckage.
F is the measured Nuclear Hot Channel Factor, defined as the maximum
'lbealheatfluxinthecoredividedbytheaverageheatfluxinthecore.
Heat fluxes are derived from measured neutron fluxes and fuel enrichment.
i V Z) is an axjally dependent function applied to the equilibrium measured to bound F 's that could be measured at non-equilibrium conditions.
q is function is based on power distribution control analyses that eval-usted the effect of burnable poisona, rod position, axial ef fects, and xenon worth, i
E F, Engineering Heat Flux Hot Channel Factor, is defined as the l
aklowanceonheat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.
Prairie Island Unit 1 - Amendment No.15 /99 66 Prairie Island Unit 2 - Amendment No. 29, /fg,, 60
TS.3.10-11
~ inches from the banW demand p' sition. An accidental o
l misalignment limit of 13 steps precludes a rod misalign-i 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 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.
N N
The permitted relaxation in Fg and F allows for radial power shapechangeswithrodinsertiontotkeinsertionlimits.
It has been determined that provided the above conditio'ns 1 through 4 are obsegved, these hot channel f actor limits are met.
In specification 3.10, F arbitrarily limited for P < 0.5 (except for low power physics testk)is The procedures for axial power distribution control referred to above are designed to minimize the ef fects of xenon redistribution on the axial power distribution during load-follow maneuvers.
Basically control of flux difference is required to limit the difference between i
the current value of Flux Difference ( tkI) 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.
l The technical specifications on power distribution control assure that TS.3.0(Z) upper bound envelope of 2.32/P times Figures TS.3.10-5 and l
the F IO-7 is not exceeded and xenon distributions are not developed which at a later time, would cause greater local power peaking even l
though the flux difference is then within the limits specified by the procadure.
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 i.
length rod control rod bank more than 190 steps withdrawn (i.e.,
normal full power operating position appropriate for the time in life, l
usually withdrawn farther as burnup proceeds).
This value, divided by l
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 of +5 percent A I 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.
Prairie Island Unit 1 - Amendment No. 35, 99, 66 Prairie Island Unit 2 - Amendment No. 29,28,60
FIGURE TS.3.10-5 1
- 1. 2 nn,,
1 (6.0,1,0)
,,, r e (10.8,.942)*~
1.O
+
-~.-
O.8
.-~-
1: -
K(Z)
_ '. =
Li 0.6 D
g.-.
- +.
%g (12.0,.430h 0.4 wu m.-
0.2
=. _w_.._...m
- a
-~,:
g t-:
m
- =de-
=.
. m_..
- 0. 0 '
~ ~~~ ~~~.....%
~
~~~
~"~"
- "~ ~ -
0 2
4 6
8 10 12 Core Height (ft)
HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE FOR FQ = 2.32 Prairie Island Unit 1 - Amendment No. 35, 66 Prairie Island Unit 2 - Amendment No. 29, 60
- _ _ _ _ _ _ _ _ _ _ _ _