ML20198S373

Amend 189 to License DPR-40,revising TS 2.1.6 to Restrict Number of Inoperable Main Steam Safety Valves When Reactor Critical & Revising Associated Basis

ML20198S373
Person / Time
Site:	Fort Calhoun
Issue date:	12/31/1998
From:	Wharton L NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198S366	List: ML20198S362 ML20198S373 ML20198S377
References
NUDOCS 9901110294
Download: ML20198S373 (6)

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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 001 sg..../

OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.189 License No. DPR-40

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

A. The application for amendment by the Omaha Public Power District (the licensee) dated March 26,1997, as supplemented by letters dated March 18,1998, and November 17,1998, 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 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.

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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:

B; Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.189, 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.

FOR THE NUCLEAR REGULATORY COMMISSION W

L. Raynard Wharton, Project Manager Project Directorate IV-2 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: December 31, 1998 I

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ATTACHMENT TO LICENSE AMENDMENT NO.189  ;

FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 ,

1 Revise Appendix "A" Technical Specifications as indicated below. The revised pages are identified by amendmen! number and contain vertical lines indicating the area of change.

REMOVE INSERT 2-15 2-15 2-15a 2-15a 2-16 2-16 i

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L 2.0 LIMITING CONDITIONS FOR OPERATION l 2.1. Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves j

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Apolicabihty l Applies to the status of the pressurizer and main steam safety valves.

Obiective l

To specify minimum requirements pertaining to the pressurizer and main steam safety valves.  ;

Specifications i To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met: '

l l (1) The reactor shall not be made critical unless the two pressurizer safety valves are operable with their lift settings adjusted to ensure valve opening at 2485 psig 11 % and 2530 psig 11 %.m l l (2) Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one

! operable safety valve shall be installed on the pressurizer. However, when in at least the cold l l shutdown condition, safety valve nozzles may be open to containment atmosphere during i l performance of safety valve tests or maintenance to satisfy this specification.

[ (3) At least four of the five Main Steam Safety Valves (MSSVs) associated with each steam generator shall be OPERABLE in MODES I and 2. Lift settings shall be at 985 psig +3/-2%,

1 1000 psig +3/-2%,1010 psig +3/-2%,1025 psig +3/-2%, and 1035 psig +3/-2%.m l a. With less than four of the five MSSVs associated with each steam generator l OPERABLE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN

! within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(4) Two power-operated relief valves (PORVs) shall be operable during heatups and cooldowns when the RCS temperature is less than 515*F, and in Modes 4 and 5 whenever the head is on the reactor vessel and the RCS is not vented through a 0.94 square inch or larger vent, to prevent violation of the pressure-temperature limits designated by Figures 2-1A and 2-1B.

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a. With one PORV inoperable during heatups and cooldowns when the RCS temperature is less than $15'F, restore the inoperable PORV to operable within 7 days or be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least

a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

. b. With both PORVs inoperable during heatups and cooldowns when the RCS temperature

! is less than 515'F, be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

! c. With one PORV inoperable in Modes 4 or 5, within one hour ensure the pressurizer L steam space is greater than 53% volume (50.6% or less actual level) and restore the l inoperable PORV to operable within 7 days. If adequate steam space cannot be

. established within one hour, then restore the inoperable PORV to operable within 24

- - hours. If the PORV cannot be restored in the required time, depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

2-15 Amendment No. 39,47,54,146,161, 189 i

- wr ummMUFUUNumUN5 FUR UPERATION 2.1 Reactor Coolam System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued)

d. With both PORVs inoperable in Modes 4 or 5 depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

(5) Two power-operated relief valves (PORVs) and their associated block valves shall be operable in Modes 1,2, and 3.

a. With one or both PORV(s) inoperable because of excessive seat leakage, within I hour either restore the PORV(s) to operable status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />,

b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore PORV to operable status or close its associated block valve and remove power from the block valve; restore the PORV to operable status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN with the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

c. With both PORVs inoperable due to causes other than excessive seat leakage, within I hour either restore at least one PORV to operable status or close both block valves, remove power from the block valves, and be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the fo!10 wing 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

d. With one or both block valve (s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the block valve (s) to operable status or place the associated PORV(s) in the closed position. Restore at least one block valve to operable status within the next hour if both block valves are inoperable; restore the remaining inoperable block valve to operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Otherwise, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

B31l i The highest reactor coolant system pressure reached in any of the accidents analyzed resulted from a complete loss of turbine generator load without simultaneous reactor trip while operating at 1500 MWt.* This pressure was less than the 2750 psia safety limit and the ASME Section III upset pressure limit of 10% greater than the design pressure.m The reactor is assumed to trip on a "High Pressurizer Pressure" trip signal.

The pressurizer safety valves are required to be calibrated to within i 1% of the specified setpoint value using ASME Section XI test methods. ASME Section XI requires that valves in steam service use steam as the test medium for establishing the setpoint. With the presence of a water-filled loop seal, establishing the valve setpoint with steam may result in in-situ valve actuation at pressures outside the i 1% tolerance specified. Under transient conditions, it is expected that the valve (s) will actuate at no less than 4 % below, nor greater than 6% above, the specified setpoint, which is within the tolerance assumed in the safety analysis.m These analyses are based on a minimum of any four of the five main steam safety valves on each main steam header being OPERABLE.

The power-operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

l 2-15a Amendment No. " '" ' " "'

189

2U LIMIIINb UUNUI'IIUN5 FUM UPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued)

Action statements (5)b. and c. include the removal of power from a closed block valve to preclude any inadvertent opening of the block valve at a time the PORV may not be closed due to maintenance.

However, the applicability requirements of the LCO to operate with the block valve (s) closed with power maintained to the block valve (s) are only intended to permit operation of the plant for a limited period of time not to exceed the next refueling shutdown (Mode 5), so that maintenance can be performed on the PORV(s) to eliminate the seat leakage condition.

To determine the maximum steam flow, the only other pressure relieving system assumed operational is the main steam safety valves. Conservative vpies for all systems parameters, delay times and core moderator coefficients are assumed. Overpressure protection is provided to portions of the reactor coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads.

If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve lift pressure would be less than half of the capacity of one safety valve. This specification, therefore, provides adequate defense against overpressurization when tne reactor is subcritical.

Performance of certain calibratiot ad maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removed, either operability of the other safety valve or maintenance of at least one nozzle open to atmospher: will assure that sufficient relief capacity is available. Use of plastic or other similar material to prevent the entry of foreign material into the open nozzle will not be construed to violate the "open to atmosphere" provision, since the presence of this material would not significantly restrict the discharge of reactor coolant.

Tia total relief capacity of the ten main steam safety valves is 6.606 x 106 lb/hr. If, following testing, l the as found setpoints are outside +/-1% of nominal nameplate values, the valves are set to within the

+/-1% tolerance. The main steam safety valves were analyzed for a total loss of main feedwater flow while operating at 1500 MWtm to ensure that the peak secondary pressure was less than 1100 psia, the ASME Section III upset pressure limit of 10% greater than the design pressure. At the power of 1500 MWt, sufficient relief valve capacity is available to prevent c,verpressurization of the steam system on loss-of-load conditions.*

The power-operated relief valve low setpoint will be adjusted to provide sufficient margin, when used in conjunction with Technical Specification Sections 2.1.1 and 2.3, to prevent the design basis pressure transients from causing an overpressurization incident. Limitation of this requirement to scheduled cooldown ensures that, should emergency conditions dictate rapid cooldown of the reactor coolant system, inoperability of the low temperature overpressure protection system would not prove to be an inhibiting factor. The effective full flow area of an open PORV is 0.94 inz, Removal of the reactor vessel head provides sufficient expansion ve' une to limit any of the design basis pressure transients. Thus, no additional relief capacity is requirr References (1) Article 9 of the 1968 ASME Boiler and Pressure Vessel Code, Section 111 (2) USAR, Section 14.9 (3) USAR, Section 14.10 (4) USAR, Sections 4.3.4, 4.3.9.5 2-16 Amendment No. 39 " '54,145,151, 1 W9 l