ML20217B520

Amend 193 to License DPR-40,revising TS Sections 2.10.4,3.1 & Table 3-3 to Increase Min Required RCS Flow Rate & Change Surveillance Requirements for RCS Flow Rate

ML20217B520
Person / Time
Site:	Fort Calhoun
Issue date:	10/06/1999
From:	Stephen Dembek NRC (Affiliation Not Assigned)
To:
Shared Package
ML20217B517	List: ML20217B511 ML20217B520 ML20217B540
References
NUDOCS 9910120282
Download: ML20217B520 (7)

Text

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p 'i, UNITED STATES g g NUCLEAR REGULATORY COMMISSION o,

• WASHINGTON, D.C. 20566 4001 1

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QMAHA PUBLIC POWER DISTRICT l

DOCKET NO. 50-285 FORT CALHOUN STATION. UNIT NO.1 i AMENDMENT TO FACILITY OPERATING LICENSE I

Amendment No.193 License No. DPR-40 I

1. The Nuclear Regtiatory Commission (the Commission) has found that:  ;

l A. The application for amendment by the Omaha Public Power District (the  ;

licensee) dated March 31,1999, 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; 4

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 license 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.

9910120282 99100625 DR hDOCK 05

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2. Accordingly, Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-40 is hereby amended to read as follows: i B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.193 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/  %

Stephen Dembek, Chief, Section 2 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: October 6, 1999 i

I-l ATTACHMENT TO LICENSE AMENDMENT NO. 193 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the fohcwing pages of Appendix A Technical Specifications with the attached revised pages. The revised pages ara identified by amendment number and contain vertical lines indicating the areas of change.

REMOVE INSERT 2-57c 2-57c 2-57e 2-57e l

3-2 3-2 3-16 3-16

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F 2.0 LIMITING CONDITIONS FOR GPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)

(5) DNBR Marnin Durine Power Operation Above 15% of Rated Power (a) The following limits on DNB-related pcrameters shall be maintained: l (i) Cold Leg Temperature as specified in the COLR l (Core Inlet Temperature) l (ii) Pressurizer Pressure 22075 psia

• l (iii) Reactor Coolant Flow rate 2206,000 gpm indicated l l (iv) Axial Shape Index as specified in the COLR l (b) With any of the above parameters exceeding the 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 power to less than 15% of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, i BMin The limitation on linear heat rate ensures that in the event of a LOCA, the peak teraperature of the fuel cladding will not exceed 22NfF.

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System, or the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution l

and are capable of verifying that the linear heat rate does not exceed its limit. The Excore Detector Monitoring System performs this function by continuously monitoring the axial shape index with the operable quadrant symmetric excore neutron flux detectors and verifying that the axial shape index is maintained within the allowable limits of the Limiting Condition for Operation for Excore Monitoring of LHR Figure provided in the COLR as adjusted by Specification 2.10.4(1)(c) for the allowed linear heat rate of the Allowable Peak Linear Heat Rate vs. Burnup Figure provided in the T

COLR, RC Pump configuration, and F,[ of the Fa , F,/ and Core Power Limitations Figure provided in the COLR. In conjunction with the use of the excore monitoring system and in establishing the axial shape index limits, the following assumptions are made. (1) the CEA insertion limits of Specification 2.10.1(6) and long term insertion limits of Specification 2.10.1(7) are satisfied, (2) the flux peaking augmentation factors are as shown in Figure 2-8, and (3) the total planar radial peaking factor does not exceed the limits of Specification 2.10.4(3).

Limit not applicable during either a thermal power ramp in excess of 5 % of rated thermal power per minute or a thermal power step of greater than 10% of rated thermal power.

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2-57c Amendment No. 32,43,57,70,193

r 2.0 LIMITING CONDITIONS FOR OPERATIQN l 2.10 Reactor Core (Contmued) 2.10.4 Power Distribution Limits (Continued) l In order for these objectives to be met, the reactor must be operated consistent with the operating limits specified for margin to DNB.

The parameter limits given in (5) and the F,7, F,,7and Core Power Limitations Figure provided in the COLR along with the parameter limits on quadrant tilt and control element assembly position (Powei Dependent Insertion Limit Figure provided in the {

COLR) provide a high degree of acurance that the DNB overpower margin will be j maintained during steady state operao m.

The actions specified assure that the reactor is brought to a safe condition.

I The Reactor Coolant System flow rate of 206,000 gallons per minute is the indicated value. It does not include instrumentation uncertainties. l The calorimetric methodology shall be used to measure the Reactor Coolant System flow rate, l

2-57e Amendment No. 32,57,141,157, 169,193 (next page is 2-59) l l

3.0 SURVEILLANCE REOUIREMENTS 3.1 Imitrumentation and Control (Continued)

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

The minimum calibration frequencies of once-per-day (heat balance adjestment only) for the power range safety channels, and once each refueling shutdown for the process system channels, are considered adequate.

The minimum testing frequency for those instrument channels connected to the Reactor Protective System and Engineered Safety Features is based on ABB/CE probabilistic risk analyses and the accumulation of specific operating history. The quarterly frequency for 1 1

the channel functional tests for these systems is based on the analyses presented in the NRC approved topical report CEN-327-A, "RPS/ESFAS Extended Test Interval Evaluation,"

as supplemented, r.nd OPPD's Engineering Analysis EA-FC-93-064, "RPS/ESF Functional Test Drift Analysis."

The low temperature setpoint power operated relief valve (PORV) CHANNEL FUNCTIONAL TEST verifies operability of the actuation circuitry using the installed test switches. PORV actuation could depressurize the reactor coolant system and is not .

required.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently l installed sensing element. The CHANNEL CALIBRATION may be perfor med by means of any series of sequential, overlapping, or total chantiel steps so that the entire channel is calibrated. ,

Calculation of the Reactor Coolant System (RCS) total flow rate by performance of a ,

l precision calorimetric heat balance once every 18 months verifies that the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate (Table 3-3, Item 15, Reactor Coolant Flow).

The Frequency of 18 months reflects the importance of verifying flow after a refueling outage when the core has been altered, Steam Generator tubes plugged, or other activities, which may have caused an alteration of flow resistance, This requirement is modified by a footnote that requires the surveillance to be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 795 % reactor thermal power (RTP) following power escalation from a refueling outage. The footnote is necessary to allow measurement of the flow rate at normal operatin; condMons at power in MODE 1.

3-2 Amer.dment No. 1 0 ,182,193 July 15,1999

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