ML19318D080
| ML19318D080 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/02/1980 |
| From: | Goodwin C PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19318D076 | List: |
| References | |
| TAC-42192, TAC-59657, TAC-59658, NUDOCS 8007070307 | |
| Download: ML19318D080 (6) | |
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e-PORTLAND GENERAL ELECTRIC C0!!PANY EUGENE WATER & ELECTRIC BOARD AND PACIFIC POWER & LIGHT COMPANY TROJAN NUCLEAR PLANT Operating License NPF '
Docket 50-344 License Change Application 62 Licensee hereby requests an amendment to License NPF-1 to incorporate changes in the reactor coolant pump breaker position inputs to the reactor trip-logic.
PORTLAND GENERAL ELECTRIC COMPANY By C. Goodwin, Jr. 7 Assistant Vice President Thermal Plant Operation and !!aintenance Subscribed and sworn to before me this 2nd day of July 1980.
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hd Notary Public of O'regon My Commiss' ion' Expires:
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LCA 62 Pags 1 of 2 LICENSE CHANGE APPLICATION 62 The Solid State Protection System (SSPS) logic inputs for reactor trips caused by reactor coolant pump (RCP) breaker position are presently a function of power level. Between 10 percent (P7) and.36 percent power, 2 out of 4 breaker open indications are required before initiating a reactor trip. Above 36 percent power (P8), the logic is changed to a 1 out of 4 requirement. This modification will retain the 2 out of 4 logic for all power levels at or above 10 percent. Table 3.3-1 on Pages 3/4 3-2, 3-4 and 3-7 of the Trojan Technical Specifications will be modified as shown on the attached pages.
REASON FOR CHANGE Fluctuations in 120-V a-c inverter output voltage have resulted in nui-sance reactor trips in the past. Any maintenance, modification, or testing work in the electrical panels where the 120-V a-c preferred bus power appears has the potential of causing such a trip by triggering the RCP breaker open position input to the SSPS trip logic. Whenever this power source is grounded in a panel, the inverter voltsge will suddenly swing low (for a very short period) and the associated RCP breaker position relays will indicate an open condition. When operating above P8, one such indication can cause a reactor trip (1 cf 4 logic).
This change will alter the RCP breaker position logic inputs to the SSPS. A 2 out.of 4 logic rather than the present 1 out of 4 idll be used for all levels above P7.
This will permit required maintecance and modification work in the.ontrol room cabinets to complete 11N-2 Lessons Learned action items (e.g., Centainment pressure, reactor ves-sel level indication, etc.) to proceed while lowering the probability of these spurious trips and decreasing stress to the Reactor Coolant System (RCS) by reducing thermal cycling.
SAFETY EVALUATION Fluctuations in the RCP invertor output voltage have instigated nui-sance reactor trips when the input to the SSPS incorrectly identifies the condition as an open RCP breaker. This change will Laprove Plant availability and will reduce stress to the RCS from unneccessary ther-mal cycling by modifying the reactor trip logic. Presently, when the Plant is operating above P8 (36-percent power) a single RCP breaker open signal (1 out of 4) will cause a reactor trip.
Below P8 but above P7 (10 percent power) two such signals (2 out of 4) are required.
This change calls for retention of the 2 out of 4 trip logic when at or above P7 and P8 rather than switching to a 1 out of 4 logic when P8 is reached.
This change neither increases the probability nor the consequences of any accident presently analyzed in FSAR Chapter 15.
No new accidents are introduced either. The RCP breaker open trip logic was provided as an anticipatory signal and is not credited in the Chapter 15 analyses.
The loss of flow signal for each RCS loop is the relied upon input for accident analysis. Since no credit is taken for the present anticipa-tory signals,.the margin of safety is unaffected above P8 and this change does not affect the logic between P7 and P8.
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LCA 62 Paga 2 of 2 The reactor trip resulting f rom an open RCP breaker is an anticipatory trip for low reactor coolant system flow resulting from a loss of RCPs.
Changing the logic above 36-percent power (P8) to require two RCP breakers to open to initiate a reactor trip is not considered to decrease the margin of safety that presently exists. The loss of one RCP will still result in a reactor trip from the loss of actual flow in one loop (two of three flow transmitters). The anticipatory reactor trip signal for a loss c.
flow resulting from loss of one RCP is not as important as it is for a loss of flow resulting from loss of two or three RCPs. The significant nuclear parameters (flow, nuclear power, heat flow and DNBR) do not approach as close to their critical or limiting valves when one of four RCPs is tripped as they do when two or three of four RCPs are tripped; thus, the anticipatory feature for which credit is not taken in the partial loss of forced reactor coolant flow safety analysis is not necessary for a loss of one RCP.
On the other hand, the anticipatory feature still exists for loss of two or more RCPs. Specifically, for a loss of the 12-kV buses, equiv-alent protection still exists.
If power is lost to one 12-kV bus, two RCPs will be tripped and the anticipatory trip of the reactor will still occur. For any other case in which low flow results from a loss of two RCPs, the anticipatory reactor trip will still occur and provide the same margin of safety.
Since this change will not have any affect on the response times credited in Chapter 15, Accident Analyses, the environmental analyses contained in the FSAR, The Environmental Report, and the final Environmental Impact Statement will likewise be unaffected. Neither will this change intro-duce any new environmental matter not previously reviewed or evaluated by the NRC.
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e TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION E
MINIMUM E
TOTAL NO.
CHANNELS DIANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION 1.
Manuu! Reactor Trip 2
1 2
1, 2 and
- 11 l
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2.
Power Range, Neutron Flux.
4 2
3 1, 2 2
3.
Power Range, Neutron Flux 4
2 3
1, 2 2
liigh Positive Rate 4
Power Range, Neutron Flux, 4
2 3
1, 2 liigh Negative Rate 5.
Intermediate Range, Neutron Flux 2
1 2
1, 2 and
- 3
[
6.
. Source Range, Neutron Flux A.
Startup 2
1 2
2# and
- 4 B.
Shutdown 2
0 1
3, 4 and 5 5
7.
Overtemperature AT Four Loop Operation 4
2 3
1, 2 2
Three Loop Operation 4
1**
3 1, 2 9
8.
Overpower AT Four Loop Operation 4
2 3
1, 2 2
Three Loop Operation 4
1**
3 1, 2 9
m n
S >N 9 9.
Pressurizer Pressure-Low 4
2 3
1, 2 6
[
g 10.
Pressurizer Pressure-High 4
2 3
1, 2 6
l
- 11. Pressurizer Water Level-liigh 3
2 2
1, 2 7
- iligh Voltage to detector may be de-engerized above P-6.
g s.
TABLE 3.3-1 (Continued)
M REACTOR TRIP SYSTEM INSTRUMENTATION E3 MINIMUM E-TOTAL NO.
OfANNELS CHANNELS APPLICABLE FUNCTION >'.L UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION w
~
19.
Safety Injection Input from ESF 2
1 2
1, 2 1
- 20. Reactor Coolant Pump Bre' ker a
Position Trip. at or Above P-7 4-1/ breaker 2
1/ breaker 1 10 per oper-ating loop
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- 21. Reactor Trip Breakers 2
1 2
1, 2*
1 R
- 22. Automatic Trip Logic 2
1 2
1, 2*
1 a
Y
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a 8.
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LCA 62
,,7 Attacitunt Page 3 of 3 TABLE 3.3-1 (Continued) a.
Less than or equal to 5% of RATED THERMAL POWER, place
.the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and restore the inoperable channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after increasing THERMAL POWER above 5% of RATED THERMAL POWER; otherwise, reduce THERMAL P04ER to less than 5% of RATED THERMAL POWER within the following 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
b.
Above 5% cf RATED THERMAL POWER, place the inoperable chancel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; operation may continue until performance of the next reeuired CHANNEL FUNCTIONAL TEST.
ACTION 8 - With the number of OPERABLE channels one less than the Total Numbers of Channels and with the THERMAL POWER level above P-7, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.
ACTION 9 - With a channel associated with an operating loop inoperable, restore the inoperable chenel to OPERABLE status within hours or be in H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; howe ar, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1.l.
ACTION 10 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
ACTION 11 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restor the inoperable channel to OPERABLE status within 48 hour: or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.
REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION
_ CONDITION AND SETP0 INT FUNCTION P-6 With 2 of 2 Intermediate Range
' Prevents or defeats Neutron Flux Charnels < 6 x 10-11 the manual block of amps.
source range reactor tri p.
TROJAN-UNIT 1 3/4 3-7