ML20206Q128
| ML20206Q128 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/10/1999 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20206Q126 | List: |
| References | |
| NUDOCS 9905190065 | |
| Download: ML20206Q128 (2) | |
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l Attachment to 2CAN079803 Page 3 of 3 1
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ANO-2 Technical Specification Bases, Page B 2-2 l
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i 9905190065 990510 PDR ADOCK 05000368 P
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i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Limiting safety system settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and limiting conditions for operation on DNBR and kw/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits (i.e., DNBR and centerline fuel melt temperature) are not exceeded during normal operation and design basis anticipated operational occurrences.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The Reactor Coolant System components are designed to Section III of the ASME Code for Nuclear Power Plant Components. The reactor vessel, steam generators and pressurizer are designed to the 1968 Edition, Summer 1970 Addenda; piping to the 1971 Edition, original issues and the valves to the 1968 Edition, Winter 1970 Addenda"8 Section III of this code permits a l
maximum transient pressure of 110% (2750 psia) of design pressure. The safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrence 3 and to assist the Engineered Safety Features Actuation System in udtigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.25 and 21.0 kw/ft, respectively.
Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip Setpoints.
"' Use of a later ASME Section III Code is acceptable, provided the Code section(s) is reconciled in accordance with Section XI.
ARKANSAS - UNIT 2 B 2-2 Amendment No. 34.M.-79,4M,