ML20095B952
| ML20095B952 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/15/1984 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20095B936 | List: |
| References | |
| NUDOCS 8408220316 | |
| Download: ML20095B952 (5) | |
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DESCRIPTION OF AMENDMENT REQUEST As a result of an NRC concern involving Low Temperature Overpressure Protection (LTOP), AP&L is proposing Technical Specifications which require additional restrictions and testing of the LTOP system.
Many of the actions required by the proposed Technical Specifications are presently performed per AP&L procedures.
BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The proposed Technical Specifications are more stringent as they are an addition to the present Technical Specifications and require more testing than was previously required by the Technical Specifications.
The proposed amendment request does not involve a significant Hazards Consideration as it does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Additionally, it does not introduce the possibility of a previously unanalyzed accident or involve a significant reduction in the margin of safety.
The Commission has provided guidance concerning the application of these standards by providing certain examples (40FR14870).
The proposed amendment matches example (ii) "A change that constitutes an additional limitation, restriction or control not presently included in the technical specifications:
for example, a more stringent surveillance requirement."
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3.1.2.7 Prior to reaching fifteen effective full power years of operation, Figures 3.1.2-1, 3.1.2-2 and 3.1.2-3 shall be updated for the next service period in accordance with 10CFR50, Appendix G, Section V. B.
The service period shall be of sufficient duration to permit the scheduled evaluation of a portion of the surveillance data scheduled in accordance with Specification 4.2.7.
The highest-predicted adjusted reference temperature of all the beltline region materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.8.
The provisions of Specifications 3.0.3 and 3.0.4 at' not applicable.
3.1.2.8 The updated proposed technical specifications referred to in 3.1.2.7 shall be submitted for NRC review at least 90 days prior to the end of the service period.
Appropriate additional NRC review time chall be allowed for proposed tecnnical specifications submitted in accordance with 10 CFR Part 50, Appendix G, Section V.C.
3.1.2.9' With the exception of ASME Section XI testing and when the core flood tank is depressurized, during a plant cooldown the core flood tank discharge valves shall be closed and the circuit breakers for the motor operators opened before depressurizing_the reactor coolant system below 600 psig.
3.1.2.10 With the exception of ASME Section XI testing, fill and vent of the reactor coolant system, and to allow maintenance of the valves, when the reactor coolant temperature is less than 280 F the four High Pressure Injection motor operated valves shall be closed with their opening control circuits for the motor operators disabled.
3.1.2.11 The plant shall not be operated in a water solid condition when the RCS pressure boundary is intact except as allowed by Emergency Operating Procedures and during System Hydrotest.
AMENDMENT NO. 57 18a ARKANSAS - UNIT 1
The heatup and cooldown rate stated in this specification are intended as the maximum changes in temperature in one direction in a one hcur period.
The actual temperature linear ramp rate may exceed the stated limits for a time period provided that the maximum total temperature difference does not exceed the limit and that a temperature hold is observed to prevent the total temperature difference from exceeding the limit for the one hour period.
Specification 3.1.2.9 is to ensure that the core flood tanks are not the source for pressurizing the reactor coolant system when in cold shutdown.
Specification 3.1.2.10 is to ensure that high pressure injection is not the source of pressurizing the reactor coolant system when in cold shutdown.
Specification 3.1.2.11 is to ensure that the reactor coolant system is not operated in a manner which would allow overpressurization due to a temperature transient.
REFERENCES (1) FSAR, Section 4.1.2.4 (2) ASME Boiler and Pressure Code,Section III, N-415 (3) ~FSAR, Section 4.3.10.5 (4) BAW-1440 (5) BAW-1698 (6) BAW-1547, Revision 1 (7) BAW-1511P (8) BAW-1436 AMENDMENT NO. 2, 22, 83 20 ARKANSAS - UNIT-1 7
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Table 4.1-1 (Cont'd)
Channel Description
-Check
' Test Calibrate Remarks 47.EFWActuationdontrolLogic-NA M
R
- 48. EFW Flow Indication R
NA R
- 49. RCS subcooling margin D
NA R
monitor
- 50. Electromatic relief valve D
NA R
flow monitor
- 51. Electromatic relief block D
NA R
valve position indicator
- 52. Pressurizer safety valve D
NA R
flow monitor
- 53. Pressurizer water level D
NA R
indicator
- 54. Control Room Chlorine Detector D M
R
- 55. Low Temperature Overpressure NA R
R Protection Alarm Logic Note:
S-Each Shift T/W-Twice per Week R-Once every 18 months W-Weekly Q-Quarterly PC-Prior to going Critical if not M-Monthly P-Prior to each startup done within previous 31 days D-Daily if not done previous NA-Not applicable week B/M-Every 2 months Amendment No. 25, 35, 58, 69 72b e
8
Table 4.1-2 (Continued)
Minimum Equipment Test Frequency Item Test Frequency
- 11. Decay _ Heat Removal Functioning Every 18 months System Isolation Valve Automatic Closure and Isolation, System
- 12. Flow Limiting Annulus Verify, at normal One year, two years, on Main Feedwater Line operating three years, and every at Reactor Building conditions, that a five years thereafter Penetration gap of at least measured from date 0.025 inches exists of initial test.
between the pipe and the annulus.
- 13. SLBIC Pressure Calibrate Every 18 months Sensors
-14. Main Steam Isolation _
a.
Exercise through a.
Quarterly Valves Approximately 10%
_ Travel b.
Cycle b.
Every 18 months
- 15. Main Feedwater a.
Exercise through a.
Quarterly Isolation Valves Approximately 5%
Travel b.
Cycle b.
Every 18 months
- 16. Reactor Internals Demonstrate Operability Each refueling shutdown Vent Valves by:
a.
Conducting a remote visual inspection of visually accessible surfaces of the valve body and disc sealing faces and evaluating any observed surface irregularities.
b.
Verifying that the valve is not stuck in an open position, and Amendment No. 4, 24, 25, Order 73a dated 4/20/81
Table 4.1-2 (Continued)
MINIMUM EQUIPMENT TEST FREQUENCY c;. Verifying through manual actuation that the valve is fully open with a force of 5, 400 lbs (applied vertically upward).
I 17.'PORV Exercise End of each refueling outage.
.g 73b