2.1.1 Reactor Core:

The reactor core safety limit prevents overheating of the fuel 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 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.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation; therefore THERMAL POWER, Reactor Coolant Temperature, and Pressure have been related to DNB through correlations developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis: uncertainties in the WRB-1 and WRB-2 correlations, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered statistically such that there is at least a 95% probability with 95% confidence level that DNBR will not occur on the most limiting fuel rod during Condition I and II events.

The safety limit curves show loci of THERMAL POWER, RCS pressure, and average temperature for which the minimum DNBR is no less than the design DNBR value, or the average enthalpy at the vessel exit equals the enthalpy of saturated liquid. When the axial power imbalance is not within tolerance, the axial power imbalance affect on the Overtemperature delta-T trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 RCS Pressure:

The RCS pressure safety limit of 2735 psig protects the integrity of the Reactor Coolant System from overpressurization and prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant, which permits a maximum transient pressure of 110% (2735 psig) of design pressure. RCS piping and fittings are designed to ANSI B 31.1 1955 Edition; valves are designed to ANSI B 16.5, MSS-SP-66-1964, or ASME Section III-1968, which permit maximum transient pressures of up to 120% (2985 psig) of component design pressure. The entire RCS is hydrotested at 3107 psig (125% of design pressure) to demonstrate integrity prior to initial operation. (Amendment No. 197)

The Trip Setpoints are the nominal values at which the bistables are set. A bistable is properly adjusted when the “as-left” value is within the band for CHANNEL CALIBRATION accuracy (± rack calibration + comparator setting accuracy).

Trip Setpoints and Allowable Values are conservatively adjusted with respect to the analytical limits to account for: calibration tolerances, instrumentation uncertainties, instrument drift, and severe environment errors for RPS channels that must function in harsh environments (10 CFR 50.49). The methodology is consistent with ISA-S67.04-1982, endorsed via NRC Regulatory Guide 1.105, Rev. 2.

The actual nominal Trip Setpoint entered into the bistable is more conservative than the Allowable Value to account for changes in random measurement errors (e.g., drift during the surveillance interval) detectable by a CHANNEL FUNCTIONAL TEST. If the measured setpoint does not exceed the Allowable Value, the bistable is considered OPERABLE.

Manual Reactor Trip: Redundant channel to automatic protective instrumentation; provides manual reactor trip capability.

Power Range Neutron Flux (High): Provides core protection against reactivity excursions too rapid to be protected by temperature and pressure protective circuitry.

Power Range Neutron Flux (Low): Provides redundant protection in the power range for a power excursion beginning from low power. May be manually bypassed when P-10 is active (two of four power range channels indicate above approximately 9% RTP); automatically reinstated when P-10 becomes inactive (three of four channels indicate below approximately 9% RTP).

Power Range High Positive Rate: Provides protection against rapid flux increases characteristic of rod ejection events from any power level. Complements Power Range High and Low trips for rod ejection from partial power.

Intermediate and Source Range Neutron Flux: Provide core protection during reactor startup. Source Range initiates reactor trip at approximately 10⁵ counts per second unless manually blocked when P-6 becomes active. Intermediate Range initiates trip at approximately 25% RTP unless manually blocked when P-10 becomes active. No credit taken in accident analyses for these trips; required to enhance overall RPS reliability.

Overtemperature delta-T: Provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided the transient is slow with respect to piping transit delays from the core to temperature detectors (approximately 4 seconds), and pressure is within the range between High and Low Pressure reactor trips. Includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays. With normal axial power distribution, this reactor trip limit is always below the core safety limit. If axial peaks are greater than design (as indicated by top-bottom power range detector difference), the reactor trip is automatically reduced.

Overpower delta-T: Provides assurance of fuel integrity (no melting) under all possible overpower conditions, limits the required range for Overtemperature delta-T protection, and provides a backup to the High Neutron Flux trip. As a result of the new AREVA steam generators, credit is taken for operation of this trip in accident analyses for protection of the reactor core following a main steam line break.

Pressurizer Pressure (High): Backed up by pressurizer code safety valves for RCS overpressure protection; set lower than the pressurizer safety valve set pressure (2485 psig).

Pressurizer Pressure (Low): Provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level (High): Ensures protection against RCS overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through pressurizer safety valves. No credit taken in accident analyses; required to enhance RPS reliability.

Loss of Flow: Provides core protection to prevent DNB in the event of a loss of one or more RCPs. Above 11% RTP (P-7 equivalent), reactor trip occurs if flow in any two loops drops below 90% of nominal full loop flow. Above 36% RTP (P-8), reactor trip occurs if flow in any single loop drops below 90% of nominal full loop flow. Three-loop operation above the 4-loop P-8 setpoint has not been evaluated and is not permitted.

SG Water Level Low-Low: Prevents operation with SG water level below the minimum volume required for adequate heat removal capacity. The setpoint provides allowance that sufficient water inventory will be in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.

Undervoltage and Underfrequency — RCP Busses: Provide core protection against DNB resulting from loss of voltage or underfrequency to more than one RCP. Specified setpoints assure a reactor trip signal is generated before the low flow trip setpoint is reached. Time delays are incorporated to prevent spurious trips from momentary electrical transients. For undervoltage: time delay set so signal reaches reactor trip breakers within 0.9 seconds following simultaneous trip of two or more RCP bus circuit breakers. For underfrequency: time delay set so signal reaches reactor trip breakers within 0.3 seconds after the underfrequency setpoint is reached.

Turbine Trip: Causes a direct reactor trip when operating above P-9. No credit taken in accident analyses; required to enhance RPS reliability.

Safety Injection Input from ESF: If a reactor trip has not already been generated by RPS, ESF automatic actuation logic will initiate a reactor trip upon any signal that initiates a safety injection. Provided to protect the core in the event of a LOCA.

RCP Breaker Position Trip: An anticipatory trip providing core protection against DNB resulting from opening of two or more pump breakers above P-7. Blocked below P-7. No credit taken in accident analyses; required to enhance RPS reliability. (Amendment Nos. 197, 261)