ML20207J924
| ML20207J924 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 12/16/1986 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20207J920 | List: |
| References | |
| NUDOCS 8701090030 | |
| Download: ML20207J924 (8) | |
Text
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^g UNITED STATES
[
g NUCLEAR REGULATORY COMMISSION
- j WASHINGTON, D. C. 20555
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PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY DOCKET NO. 50-344 TROJAN NUCLEAR PLANT AMENDMENT TO FACILITY CPERATING LICENSE Amendment No.122 License No. NPF-1 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Portland General Electric Company, et al., (the licensee) dated November 24, 1984, as superseded Ueceniber 27, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the 4.pplication, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
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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 Operating License No. NPF-1 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.122, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of the date of its issuance.
F0 THE NUCL GULATORY COMMISSICN I
I Eg har ven PWR Project Direct r!te No. 3 i
Division of PWR Li sing-A, NRR
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 16, 1986 1
i
l pa ury gjg UNITED STATES 3,
NUCLEAR REGULATORY COMMISSION o
g
- y WASHINGTON, D. C. 20555 1
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ATTACHMENT TO LICENSE AMENDMENT N0.122 TO FACILITY OPERATING LICENSE NO. NPF-1 DOCKET NO. 50-344 Revise Appendix A as follows:
Remove Pages Insert Pages TS X
TS X
3/4 4-2 2/4 4-2 3/4 4-2b 3/4 4-2b B 3/4 4-1 B 3/4 4-1 B 3/4 4-la B 3/4 4-la I
1 J
~
JNDEX BASES I
SECTION
_ Pace 3/4.3 INSTRUNENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION..............
B 3/4 3-1 3/4. 3. 2 ENGINEERED SAFETY FEATURE INSTRUMENTATIO 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.........
3 B3/44-lal 3/4.4.4 PRESSURIZER..................
B 3/4 4-2 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 TROJAN-UNIT 1 X
Amendment No. 93, 122
e REACTOR COOLANT SYSTEM ACTION (Continued) b)
Place the following reactor 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) Tavg -- Low-Low channel used in the coincidence circuit 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
~
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 $75% of RATED THERMAL POWER.
2.
THERMAL POWER is restricted to 570% of RATED THERMAL PO Below P-7:
With K pt >l.0, operation below P-7 may proceed provided at a.
e 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. ((, 122
s lREACTORCOOLANTSYSTEM HOT STANDBY 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:
With less than four reactor coolant loops in operation, immediately a.
de-energize 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 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> With no reactor cr.olant loop in operation, suspend all operations c.
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 REOUIREMENTS 4.4.1.2.1 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 operatio 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 l
alignments and indicated power availability.
4.4.1.2.3 At least one cooling loop shall be verified to be in operation l
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.
TROJAN-UNIT 1 3/4 4-2b Amendment No. 54,122
e 3/4.4 REACTOR COOLANT SYSTEM 4
J BASES 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 <38 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 P-7 (10 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-B (39 percent of RATED THERMAL POWER).
In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure con-siderations require that two loops be DPERABLE. 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 adequate 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 i
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 i
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 83/44-1 Amendment No. 5#, 58, 7g,122
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8 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 t
the valve's setpoint.
The relief capacity of a single safety valve is adequate to relieve any overpre', cure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability i
and will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be i
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' lif t settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
TROJAN-UNIT 1 B 3/4 4-la Amendment No. 5#, 7#, 122