ML20235J876
| ML20235J876 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/08/1987 |
| From: | Wigginton D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20235J877 | List: |
| References | |
| TAC-65078, TAC-65079, NUDOCS 8707160041 | |
| Download: ML20235J876 (10) | |
Text
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UNITED STATES
',g NUCLEAR REGULATORY COMMISSION y
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j WASHINGTON, D. C. 20555 NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NO. 1 j
AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 81 l
License No. DPR-42 l
c.
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated April 13, 1987, complies with the standards and i
requirements of the Atomic Energy Att of 1954, as amended (the Act),
i and the Commission's rules and regulations set forth in 10 CFR i
Chapter I; I
B.
The facility will operate in conformity with the application, i
the provisions of the Act, and the rules and regulations of I
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 l
of the Commission's regulations and all applicable requirements have been satisfied.
8707160041 070708 PDR ADOCK 0500 ' 2 P
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. 81 are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
I i
3.
This license amend..,ent is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
-s David L.
gYhton, Acting Director
]
Project Directorate III-3 1
Division of Reactor Projects
)
Attachment:
Changes to the Technical Specifications l
Date of Issuance: July 8, 1987 1
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,...s fg UNITED STATES
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NUCLEAR REGULATORY COMMISSION y
'.j W ASHINGT ON, D. C. 205S5 5
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated April 13, 1987, 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, l
the provisions of the Act, and the rules and regulations of i
the Commission; 1
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 l
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 l
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.
L
l
.g.
I 2.
Accordingly, the license is amended by changes to the Technical
)
Specifications as indicated in the attachment to this license
{
amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 74, are hereby
~~
incorporated in the license.
The licensee shall operate the facility in accordance with the Technical f
Specifications.
This license amendment is effective as of the date of its issuance.
3.
FOR HE NUCLEAR REGULATORY COMMISSION l
David L.
nton, Acting Director Project Directorate III-3 j
Division of Reactor Projects 1
Attachment:
f Changes to the Technical l
,(
Specifications i
pc Date of Issuance: July 8, 1987 k
p h I!
1 l
1 l
I ATTAC'iMENT TO LICENSE AMENDMENT N05. 81 AND 74 TO FACILITY OPERATING LICENSE N0. 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-x TS-x TS.2.1-2 TS.2.1-2 TS.3.10-1 TS.3.10-1 TS.3.10-2 TS.3.10-2 Figure TS.3.10-8 c>.
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1 l
l TS-x 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 I and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit I and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL P0'n'ER with the Primary Coolant Specific Activity >1.0 uC1/ gram DOSE EQUIVALENT I-131 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary fer Liquid e
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 j
3.10-5 Hot Channel Factor Normalized Operating Envelope j
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 Acceptable Values of F (F and F l
Shield Building Design 1n S)eakage kg(F )
3.10-8 9 3 S
4.4-1 ate 6.1-1 NSP Corporate Organizational Relationship to On-Site Operating Organizations 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-site Operating Group I
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I Prairie Island Unit 1 - Amendment No. E9,EE,70,73,77,P,A 81 Prairie Island Unit 2 - Amendment No. E3,ED,EA,$E,70,73,74 1
TS.2.1-2 The solid curves of Figure TS 2.1-1 represent the loci of poi 61s 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 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 diff'erence 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 l
natural circulation is not allowed.
Safety limits for such special j
tests will be determined as a part of the test procedure.
i
)
The curves are conservative for the following nuclear hot channel factors:
1 2.50 I
F
= 1. 70 [1 + 0.3(1-P)) ; and F g
q Use of these factors results in more conservative safety limits than vould result from power distribution limits in Specification TS.3.10.
This combination of het channel factors is higher than that calculated at full power for the range from all control rods fully withdrawn to caximum 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. liovever, 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. Bh, 81 Prairie Island Unit 2 - Amentment No h/), 74
T S. 3 '.10- 1 3.10 CONTROL ROD AND P0k'ER DISTRIBUTION LIMITS Applicability Applies to the limits on core fission power distribution and to the limits on control rod operations.
1 Objectives 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.
I B.
Power Distribution Limits 1.
At all times, except dgring log power physics testing, measured hot channel factors, F and F33, as defined below and in the bases, shall meet thehollowinglimits:
F x 1.03 x 1.05 $[F (F q oH
$ g (F ) x [1 + 0.3(1-F)]
]
F x 1.04 F
H I
where the following definitions apply:
]
- K(Z) is the axial dependence function shown in Figure TS.3.10-5.
1 J
- Z is the core height location.
1
- P is the fraction of rated power at which the core is j
g operating.
In the F limit determination when P 3 50, set q
P - 0.50.
1 and F function [Fg (F limits are shown F u 31 9.4H oH Prairie Island Unit 1 - Amendment No. 3E,AA,66,77,81 Prairie Island Unit 2 - Amendment No. 5,3$,60,70, 74
\\ TS. 3.10- 2 l or F"H -F"kththesmallestmarginorgreatest is defined as the measured F respectively, excessoflimit. 0 6 w - 1.03 is thq engineering hot channel factor, F, applied to the ) measured F to account for manufacturing tolerance. - 1.05 is applied to the measured F"q to account for measurement uncertainty. - 1.04 is applied to the measured F t account for measurement H uncertainty. N N 2. Hot channel factors, F and FAH, shall be measured and the target n flux difference determined, at equilibrium conditions according to the following conditions, whichever occurs first: c, (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 j teactor power at which target flux difference was last j determined, by 10" or more of rated power. { F (equil) shall meet the following limit for the middle axial 80% o9thecore: F (equil) x V(Z) x 1.03 x 1.05 5[F (F33)/P] x K(Z) q where V(Z) is defined Figure 3.10-7 and other teres are defined in 3.10.B.1 above. 3. (a) If either measured hot channel factor exceeds its limit specified in 3.10.B.1, reduce reactor power and the high neutron flgx trip setpoint by 1% for each percent that the eqasured F or by 3.33% for each percent that the measured ( FiH"*** (b) If the measured F ( ) exceeds the 3.10.B.2 limits but not the3.10.B.1limik,equil take one of the following actions: 1. Within 48 hours place the reactor in an equilibrium configuration for which Specification 3.10.B.2 is satis-fied, or 2. Reduce reactor power and the high neutron flux trip s tpoint by 1% for each percent that the measured F (equil) x 1.03 x 1.05 x V(Z) exceeds the limit. Prairie Island Unit No.1 - Amendment No. EE,//,ES,//, 81 Prairie Island Unit No. 2 - Amendment No. 29,3E,E0,70,74
FIGURE TS.3.10-8
- '/////// R S //B // / d
- 2., n
,I I ,' f/ l(2.4.1.6). / Operation NOT / Allowed / 2.35- ~ ~~~ -- - ~ - "'j' / i l '/ I / o b O r LL. 2.30- ~ ~ ~ ~ - - ~ ~ ~ Operation Allow e d /[ i l(2.3 2.1.6 8)~ /'
~~-~-~--l-(-
2.25- ~r - lwM ~ ~" 2.20, 1.55 1.60 1.65 1.70 F 3g(F ) a FIGURE 3.10-8 Acceptable Values of F (Fg AH) and FAH(F ) 4 q Prairie Island Unit 1 - Amendment No. 81 Prairie Island Unit 2 - Amendment No. 74}}