ML20215M854
| ML20215M854 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/29/1986 |
| From: | LOUISIANA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20215M807 | List: |
| References | |
| NUDOCS 8611030368 | |
| Download: ML20215M854 (2) | |
Text
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2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES I
2.1.1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by (1) restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or less than 21.0 kW/ft which will not cause fuel centerline melting in any fuel rod.
l First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surfa e temperature is only slightly greater than the coolant saturation temperatur.
The upper boundary of the n.cleate boiling regime is termed " departure fro nucleate boiling" (DN8).
At this point, there is a sharp reduction of th heat transfer coefficient, which would result in higher cladding temper ures and the possibility of cladding failure.
Correlations predict DNB and the location of DNB for axially uni orm and non-uniform heat flux distributions.
The local DNB ratio (DNBR), de ined as the ratio of the predicted DNB heat flux at a particular core loca on to the actual heat flux at that location, is indicative of the margin to N8.
The
/,.24 minimum value of DNBR during normal operational occurrences is mited to for the CE-1 correlation and is established as a Safety Limit.
Second, operation with a peak linear heat rate below that which would l
cause fuel centerline melting maintains fuel rod and cladding integrity.
Above this peak linear heat rate level (i.e., with some melting in the center),
fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods.
Volume i
changes which accompany the solid to liquid phase change are significant and require accommodation.
Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.
Because of the above factors, the steady state value of the peak linear heat rate' which would not cause fuel centerline melting is established as a Safety Limit.
To account for fuel rod dynamics i
I (lags), the directly indicated linear heat rate is dynamically adjusted.
I 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 are not exceeded during normal operation and design basis anticipated operational occurrences.
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 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, 1974 Edition, of the ASME Code for Nuclear Power Plant Components which permits a 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 occurrences and to assist the Engineered Safety Features Actuation System in mitigating 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, t.2 4 The DNBR - Low and Local Power Density - High are digital generated i
trip setpoints based on Limiting Safety System Settings of.
and 21.0 j
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.
To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR - Low and Local Power Density -
High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in CEN-li?(S) " " Function ! "::ign Sp::i'ica-tier fer : Cer oretectier C:!:u!:ter," January 1^81; CEM-liS'S) " "runcti:n:1
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