ML19274E462
| ML19274E462 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/02/1979 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19274E463 | List: |
| References | |
| NUDOCS 7903260211 | |
| Download: ML19274E462 (7) | |
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UNITED ST ATES o.S NUCLEAR REGULATORY COMMISSION 3's W ASHINGTON, D. C. 20555
%, ' c ;P NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICEijSE Amendment No. 33 License No. DPR-42 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated November 3,1978, 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 reascnable 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.
7903260211
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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 License No. DPR-42 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 33, 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
,/'
iC W 'd $
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Q A. Schwencer, Chief Operating Reactors Branch =1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: March 2, 1979
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UNITED STATES a
j NUCLEAR REGULATORY COMMISSION
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.e WASHINGTON, D. C. 2r555 NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 27 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 Novemoer 3,1978, complies.
with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and tne Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with tne 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 conducteo 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 t'een satisfied.
- 4 2.
Accordingiy, the license is amended by changes tc tne Technical Specifications as indicated in the attachment to nis license amey.cnent, and paragraph 2.C(2) of Facility Liter.se No. DPR-60 is hereby amended to read as follows:
(2) Technical Specificati,og The Technical Specifications contained in Ap;endices A and B, as revised through Amendment No. 27, are hereby incorporated in the license. The licensee shall operate the facility in accordance witr. the Technical Specifications.
3.
This license amendment is effective as of the cats of its issuance.
FOR THE NUCLEAR REG._ATORY COMMISSION
$!!!W'YY A. Schwencer, Chief Operating Reactors iranch el Division of Operatir) Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 2, 1979 i
ATTACHMENT TO LICENSE AMENDMENT NOS. 33 AND 27 FACILITY OPERATING LICENSE N05. DPR-42 AND DPR-60 DOCKET N05. 50-282 AND 50-306 Remove and insert the below indicated pages of the Appendix A Technical Specifications for the above indicated licenses.
Each area of change is reflected by a line in the right margin of the page.
Remove Insert 2.3-4 2.3-4 2.3-6 2.3-6
5 i
TS.2.3-4 2.
Low power block of single loop loss of flow is permitted whenever power range neutron flux is 610% of rated power.
3.
Power range high flux low setpoint trip and intbrmediate range high flux trip shall be unblocked whenever power range neutron flux is 4 9% of rated power.
4.
Source range high flux trip shall b-I O""
" **~
mediate range neutron flux is 6 10 amperes.
5.
Reactor trip on turbine trip shall be unblocked whenever power range neutron flux is350% of rated power.
C.
Control Rod Withdrawal Stops 1.
Block automatic rod withdrawal:
l a.
Turbine load 615% of full load turbine impulse pressure.
Basis i
The power range high flux reactor trips (low set point) provides redundant Thistripwasusedinthesafetyanalysis.ggjursionbeginningfromlowpowcr.
protection in the power range for a power The intermediate and source range high flux reactor trips provide additional protection agains t. uncontrolled s tartup excursions.
As power level increases, during s tartup, these trips are manually blocked to prevent unnecessary plant 3
trips.
The power range high flux (high set point) reactor trip protects the reactor core against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.
The prescribed set point, with allowance {gr errors, is consistent with the trip point assumed in the accident analysis.g The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted.
The high' pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves (2485 psig) such I
that the reactor is tripped before the safety valves actuate. The low pressur-izer pressure reactor trip 4frips the reactor in the unlikely event of a los s-o f-c oolant ac cide nt The overt empe rature a T reactor trip provides core protection against DNB for all co=binations of pressure, power, coolant temperature, and axial power distrib ution, provided only that (1) the transient is slow with respect to piping transit delays free the core to the teep-aconcs,
' and (2. pressure is within crature detecters 2 bout 1
the range between t r. e high and low pressure reactcr trips.
With i
nor=al axial power distribution, the re t tor trip limit, with allowance l
Amendment No. 33, Unit '
I Amendment No. 27, Unit :
TS.2.3-6 The other reactor trips specified in A.3. above provide additional protectiom.
The trip initiated by steam /feedwater flow mismatch in coincidence with low steam generator water level is designed for protection from a sudden loss of the reactor'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 level.
Specifically, this trip complements the power range nuclear flux high and low trip to assure that the criteria are met for rod ejection from partial power.
The negative power range rate trip provides protection satisfying all IEEE criteria to assure that minimum DNER is maintained above 1.30 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 f ull power will be indicated by the automatic reactor control system in response to a continued full power turbine load demand and it will not result in a DNBR of less than 1.30.
Thus, autornatic protection for a single rod drop is not required. Ad minis trative limits in Specification 3.10 require a power reduction if design power distribution limits are exceeded by a single misaligned or dropped rod.
References:
(1)
FSAR 14.1.1 (2)
FSAR Page 14-3 (3)
FSAR 14.2.6 (4)
FSAR 14.3.1 (5)
FSAR 14.1.2 (6)
FSAR 7.2, 7.3 l
(7)
FSAR 3.2.1 (8)
FSAR 14.1.9 (9)
FSAR 14.1.11 y
l Amendment No. 33, Uni:
Amendment No. 27, Uni j
I