ML20154F731
| ML20154F731 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/07/1988 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20154F462 | List: |
| References | |
| NUDOCS 8809200102 | |
| Download: ML20154F731 (4) | |
Text
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TABLE 3.3-11 T7e cm 5
~
1 TO < *
- M ACCIDENT MONITORING INSTRUMENTATION DS * '
82 00 t~ TOTAL NO. MINIMUM CHANNELS E OF CHANNELS OPERABLE O$
go a 1. Pressurizer Water Level 3 2 TOO c 8d" a
5 2. Auxiliary Feedwater Flow Rate 1 per steam 1 per steam e generator generator
- 3. Reactor Coolant System Subcooling Margin Monitor 1 1 l
- 4. PORV Acoustical Detector Position Indicator 2/ valve
- 1/ valve m
g 5. PORV Limit Switch Position Indicator 1/ valve 0/ valve
- 6. PORV Block Valve Limit Switch Position Indicator 1/ valve 0/ valve M
Y 7. Safety Valve Acoustical Detector Position Indicator 2/ valve
- 1/ valve o U
-g 8. Safety Valve Temperature Detector Position Indicator 1/ valve 0/ valve s
@ 9. Containment Sump Wide Range Water Level 2 1
- 10. Containment Wide-Range Pressure 2 0
- 11. In-Core Thermocouples (Core-Exit Thermocouples) 4/ core 2/ core quadrant quadrant
- 12. Reactor Vessel Level Indicating System 1 1
- One Detector Active, Second Detector Passive
P-TABLE 4.3-7 . ,.
M Q ACCIDENT MONITGRING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
& CHANNEL CHANNEL E CHECK CALIBRATION
+<
M R
' l. Pressurizer Water Level S/U(1) R S
~ Auxiliary Feedwater Flow Rate 2.
- a R M
- 3. Reactor Coolant System Subcooling Margin M R
- 4. PORV Acoustical Detector Position Indicator M R N PORV Limit Switch Position Indicator o 5.
o M R y " 6. PORV Block Valve Limit Switch Position Indicator O A M R 3- w 7. Safety Valve Acoustical Detector Position Indicator M R
- 8. Safety Valve Temperature Detector Position Indicator E M R
') 9. PORV Control Pressure channels (PT-RC-444, 445)
M R
- 10. Containment Sump Wide Range Water Level N/A R
- 11. Containment Wide-Range Pressure M R
- 12. In-Core Thermocouples (Core-Exit Thermocouples)
M R
- 13. Reactor Vessel Level Indicatine System (1) Channel check to be performmi in conjunction with Surveillance Requirement 4.7.1.2.a.9 following an extended plant outage.
.I J
., s , t ATTACHMENT B Safety Analysis Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 154 Description of amendment request: The proposed amendment would add the Incore Thermocouples and Reactor Vessel Level Indicating System (RVLIS) to Tables 3.3-11 and 4.3-7. In addition, the minimum channels operable for the Reactor Coolant System Subcooling Margin Monitor has been increased from 0 to 1. These changes are provided in response to NRC letter dated May 12, 1987 which recommended that we revise our technical specifications to include the Inadequate Core Cooling Instrumentation (ICCI). The ICCI was installed during the sixth refueling outage as described in your letter and the proposed technical specifications are similar to the guidance provided in Generic Letter 83-37 with the following exceptions. The Reactor Coolant Inventory Tracking System identified in Generic Letter 83-37 is proposed as RVLIS. This is consistent with the terminology used in the NRC evaluation provided by your letter dated May 12, 1987.
The minimum channels operable for the Reactor Coolant Subcooling Margin Monitor and the Reactor Vessel Level Indicating System proposed for Table 3.3-11 is 1 which is consistent with Generic l Letter 83-37. Two Reactor Coolant System Subcooling Margin Monitor channels and two RVLIS channels are installed and available to satisfy the requirements of this specification. However, Table 3.3-11 only specifies 1 for the total number of channels. This deviates from the 2 specified in Generic Letter 83-37. This is based on the fact that these instruments perform no automatic actions and any procedural actions keyed to specific values readable on these l instruments have alternate means avail &ble for determining subcooling l margin and the emergency operating procedures instruct the operators l ,
on what actions to take either with or without RVLIS, therefore, continued plant operation need not be lirJted when one channel of these instruments are inoperabic. Though those instruments provide additional information, they are not critical to the operation of the ,
plant. They have been added to the technical specifications to ensure periodic calibration is performed and at least one channel is -
available to satisfy NRC requirements, therefore, no additional considerations need be required.
The proposed changes are consistent with the system design which provides additional instruments and display equipment for monitoring l relevant plant parameters. The instruments do not initiato any !
automatic functions to mitigate accidents, however, they can be used during accident events to provide the plant operators with additional information. These changes will not affect thc FSAR accident analyses and the NRC has evaluated the system design and determined i that the regulatory requirements of NUREG-0737 item II.F.2 have been i satisfied. Therefore, based on the above, the proposed changes have been determined to be safe.
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e, ATTACHMENT C No Significant Hazard Evaluation Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 154 Basis for proposed no significant hazard consideration determination: The Commission has provided standards for determining whether a significant hazards consideration exists in 10CFR50.92(c).
A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
The proposed changes do not involve a significant hazards consideration because:
- 1. The proposed changes incorporate additional controls and surveillances to reflect the instruments and display equipment installed to comply with the requirements of NUREG-0737 Item II.F.2. The NRC has evaluated the system design and determined that the design changes satisfy the requirements of the regulations. These changes will not affect the FSAR accident analyses and the function of the accident monitoring instrumentation will not be reduced, therefore, a significant increase in the probability or consequence of an accident previously evaluated will not be involved.
- 2. The proposed changes have been incorporated to reficct the plant modifications installed in accordance with NUREG-0737. These instruments do not initiate any automatic functions during an accident, but may be used during accident conditions to provide information on relevant parameters to the plant operators. The additional controls and surveillances are consistent with other accident monitoring instrumentation listed on these tables and will not introduce any new or unique plant configurations that would create the possibility of a new or different kind of accident from any accident previously ovaluated.
- 3. The additional information provided by those instruments is consistent with the requirements of NUREG-0737 and will not affect any of the plant sotpoints or margins to the accident analyses or technical specification limits. Adding these instruments to the accident monitoring specification is consistent with the bases which is to ensure that sufficient information is available on selected plant parameters to monitor l and assess these variables during and following an accident.
Thorofore, the plant safety margins will not be reduced as a I
result of those changes.
Therefore, based on the above, it is proposed to detonnine that those changes do not involve any significant hazard considerations.