ML18024A335

Browns Ferry Nuclear Plant Updated Final Safety Analysis Report (Ufsar), Amendment 27, 14.4 Table - Approach to Safety Analysis

ML18024A335
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Site:	Browns Ferry
Issue date:	10/05/2017
From:	Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
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Text

BFN-16 TABLE 14.4-1

(Sheet 1)

PLANT SAFETY ANALYSIS

SUMMARY OF ABNORMAL OPERATIONAL TRANSIENTS

Undesired Parameter Event Causing Variation Transient Scram Caused by

Nuclear system pressure Generator trip without Turbine control valve increase bypass fast closure

Nuclear system pressure Turbine trip without Turbine stop valve increase bypass closure

Nuclear system pressure Main steam line isolation Main st eam line isolation increase valve closure valve closure

Nuclear system pressure Loss of Condenser vacuum Turbine stop valve closure

increase Nuclear system pressure Bypass valve malfunction Reactor vessel high pressure increase

Nuclear system pressure Pressure regulator Reactor vessel high pressure increase malfunction Reactor water temperature Shutdown cooling malfunction High Neutron flux decrease decrease temperature

Reactor water temperature Loss of feedwater heater* None decrease

Reactor Water temperature Inadvertent pump start* None decrease

Positive reactivity Continuous rod withdrawal None insertion during power range operation*

Positive reactivity Continuous rod withdrawal High neutron flux insertion during reactor startup*

Positive reactivity Control rod removal error High neutron flux insertion during refueling

Positive reactivity Fuel assembly insertion High neutron flux insertion error during refueling

Coolant inventory decrease Pressure regulator Main steam line isolation failure - open** valve closure Coolant inventory decrease Open main steam relief valve**

Coolant inventory decrease Loss of feedwater flow Reactor vessel low water level

• This transient results in no significant change in nuclear system pressure.

• • This transient results in a depressurization.

BFN-16 TABLE 14.4-1 (Sheet 2)

PLANT SAFETY ANALYSIS SUMMARY OF ABNORMAL OPERATIONAL TRANSIENTS Undesired Parameter Event Causing Variation Transient Scram Caused by Coolant inventory decrease Loss of auxiliary power Loss of power to reactor system protection Core flow decrease Recirculation flow control None failure - decreasing flow**

Core flow decrease Trip of one recirculation None pump** Core flow decrease Trip of two recirculation None pumps** Core flow increase Recirculation pump flow High neutron flux control failure increasing flow* Core flow increase Startup of idle recirculation pump* None Excess of coolant Feedwater Controller Turbine stop valve closure inventory failure-maximum demand

• This transient results in no significant change in nuclear system pressure.

• • This transient results in a depressurization.

BFN-17 TABLE 14.4-2 PLANT SAFETY ANALYSIS RESULTS OF DESIGN BASIS ACCIDENTS Percent of Core Design Basis Reaching Cladding Peak Accident Temperature of 2200

°F System Pressure

Rod Drop Not applicable*** <1375 psig Accident

Loss of Coolant 0 Not applicable* Accident

Refueling Accident 0 Not applicable**

Main Steam Line 0 Not applicable* Break Accident

• This accident results in a depressurization.

• • This accident occurs wi th the reactor vessel head off. ***Peak fuel enthalpy is less than 280 cal/gm.