ML20138R538

Proposed Tech Specs 3/4.4.1.2 Re Number of Reactor Coolant Loops to Be Operating at Low Power & in Mode 3.Certificate of Svc Encl

ML20138R538
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	12/27/1985
From:	PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20138R530	List: ML20138R529 ML20138R535 ML20138R538
References
TAC-56567, NUDOCS 8512310352
Download: ML20138R538 (6)

Text

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. , LCA 1130 Rev. 1 o Attachment 1 Page 1 of 5 s

INDEX BASES SECTION Page 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION. . . . . . . . . . . . . . B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION . . . . . . B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION. . . . . . . . . . . . . . B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS . . . . . . . . . . . . . . . . B 3/4 4-1 3/4.4.2 and 3/4.4 3 SAFETY AND RELIEF VALVES . . . . . . . . . B 3/4 4-la l 3/4.4.4 PRESSURIZER . . . . . . . . . . . . . . . . . . . . . B 3/4 4-2 j 3/4.4.5 STEAM GENERATORS. . .;. . . . . . .......... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE. . . . . . . . . . . . B 3/4 4-2 3/4.4.7 CHEMISTRY . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.8 SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.9 PRESSURE / TEMPERATURE LIMITS . . . . . . . . . . . . . B 3/4 4-4 3/4.4.10 STRUCTURAL INTEGRITY. ................ B 3/4 4-10 3/4.4.11 REACTOR VESSEL HEAD VENTS . . . . . . . . . . . . . . B 3/4 4-10 7

a 8512310352 851227 PDR ADOCK 05000344 P PDR TROJAN-UNIT 1 X Amendment No. 93,

a

. , LCA 113, Rev. 1 Attachment 1 Page 2 of 5 REACTOR COOLANT SYSTEM ACTION (Continued) b) Place the following reautor trip system and ESFAS instrumentation channels, associated with the loop not in operation, in their tripped conditions:

1) Overpower AT channel.

2) Overtemperature AT channel.

3) - Low-Low channel used in the coincidence Tavg circu it with steam Flow - High for Safety Injection.

4) Steam Line Pressure - Low channel used in the coincidence circuit with Steam Flow - High for Safety Injection.

5) Steam Flow-High channel used for Safety Injection.

6) Differential Pressure Between Steam Lines - High Channel used for Safety Injection (trip all l bistables which indicate low active loop steam pressure with respect to the idle loop steam pressure).

c) Change the P-8 interlock setpoint from the value specified in Table 3.3-1 to 175% of RATED THERMAL POWER.

2. THERMAL POWER is restricted to $70% of RATED THERMAL POWER.

Below P-7:

a. With Keff 11.0, operation below P-7 may proceed provided'at least four reactor coolant loops and associated pumps are in l operation.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

TROJAN-UNIT 1 3/4 4-2 Amendment No. 54 i

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• LCA 113, Rev. 1

, Attachment 1 i Page 3 of 5 '

REACTOR COOLANT SYSTEM l HOT STANDBY

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LIMITING CONDITION FOR OPERATION 3.4.1.2 a. With any control rod drive mechanism (CRDM) energized, all four reactor coolant loops shall be in operation.

b. With no CRDMs energized, at least two reactor coolant loops shall be OPERABLE with one reactor coolant loop in operation.*

APPLICABILITY: MODE 3 ACTION:

a.

With less than de-energize four reactor coolant loops in operation, imediately {

all CRDMs.

b. With less than two reactor coolant loops OPERABLE when the CRDMs are de-energized, restore the required loops to OPERABLE status within 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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. With no reactor coolant 1 cop in operation, suspend all operations [

involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4 . 4 .1. 2.1 All CRDMs shall be verified de-energued at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if less than four reactor coolant loops are in operation.

4.4.1.2.2 At least two reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker [

alignments and indicated power availability.

4.4.1.2.3 At least one cooling loop shall be verified to be in operation I and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is

maintained at 10*F below saturation temperature.

l TROJAN-UNIT 1 3/4 4-2b Amendment No. H

LCA 113, Rev. 1

. Attachment 1 Page 4 of 5 3/4.4 REACTOR COOLANT SYSTEM BASES i 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.73 during all normal operations and anticipated transients. With one reactor coolant loop not in operation, THERMAL POWER is restricted to <58 percent of RATED THERMAL POWER until the Overtemperature AT trip is reset. Either action ensures that the DNBR will be maintained above 1.73. A loss of flow in two loops will cause a reactor trip if operating above F-7 (10 percent of RATED THERf1AL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (39 percent of RATED THERMAL POWER).

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removitig decay heat; however, single failure con-siderations require that two loops be OPERABLE. Four loops in operation while control rod drive mechanisms are energized ensures that the DNB design basis can be met for a bank withdrawal from subcritical or low power accident.

In MODES 4 and 5, a single reactor coolant loop or RHR loop pro-vides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one RHR pump provides adaquate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 290*F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water vol-ume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures, or (3) by restricting starting of an RCP unless another RCP is running.

TROJAN-UNIT 1 B 3/4 4-1 Amendment No. 5#, 55, 78

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LCA 113, Rsv. 1 o Attachment 1 Page 5 of 5 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES and 3/4.4.3 SAFETY AND RELIEF VALVES l The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 pSig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at 110% of the valve's setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, pror! des overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (ie, no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the p$ovisions of Section XI of the ASME Boiler and Pressure Code.

TROJAN-UNIT 1 B 3/4 4-la Amendment No. 5#, 7#

r UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

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PORTLAND CENERAL ELECTRIC COMPANY, ) Docket 50-344 THE CITY OF EUGENE, OREGON, AND ) Operating License WPF-1 PACIFIC POWER & LICHT COMPANY )

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(TROJAN NLCLEAR PLANT) )

CERTIFICATE OF SERVICE I hereby certify that copies of License Change Application 113 Revision 1 to the Operating License for Trojan Nuclear Plant, dated December 27, 1985, have been served on the following by hand delivery or by deposit in the United States mail, first class, this 27th day of December 1985:

Mr. Lynn Frank, Director State of Oregon Department of Energy Labor & Industries Blds, Rm 102 Salem OR 97310 Mr. Robert L. King Chairman of County Commissioners Columbia County Courthouse St. Helens OR 97051

/v .w A. Zinmerman, Manager lear Regulation Branch ear Safety & Regulation Subscribed and sworn to before me this 27th day of December 1985.

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NOTA 2 PUGU$ EGON t.ty commis: ion txpes. N'8f88

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