View source for Plant Fire-Induced Risk Model (Task 5)

==Task Overview==
===Background===
This task discusses steps for the development of a logic model that reflects plant response following a fire. Specific instructions have been provided for treatment of fire-specific procedures or preplans. These procedures may impact availability of functions and components, or include fire-specific operator actions (e.g., self-induced-station-blackout).

===Purpose===
This section describes the procedure for developing the Fire PRA Model to calculate CDF, CCDP, LERF, and CLERP for fire events. The procedure addresses the process of implementing temporary or permanent changes to the Internal Events PRA to quantify fire-induced core damage frequency (CDF), conditional core damage probability (CCDP), large early release frequency (LERF), and conditional large early release probability (CLERP), and for developing special models to address fire emergency procedures (FEPs). The procedure also addresses the transition from temporary changes to permanent changes to the Internal Events PRA Model during the development of the Fire PRA Model.

===Scope===
This procedure addresses the following major steps for developing the Fire PRA Model for calculating CDF/CCDP and LERF/CLERP for fire events.
* Step 1–Develop the Fire PRA CDF/CCDP Model.
* Step 2–Develop the Fire PRA LERF/CLERP Model.

==Related Element of ASME/ANS PRA Standard==
Fire PRA Plant Response Model (PRM)

==Related EPRI 1011989 / NUREG/CR-6850 Appendices==
None

==Supplemental Guidance==
With the exception of main control room abandonment scenarios and plant trip probabilities as discussed below, the process for developing the Fire PRA logic model to calculate CDF/CCDP and LERF/CLERP model is largely unchanged from EPRI 1011989 / NUREG/CR-6850. However, it should be acknowledged that the addition of multiple spurious operations and/or alternate shutdown in the PRA model needs careful consideration to be appropriately placed into the model (see [[Fire PRA Component Selection (Task 2)#Supplemental Guidance|Task 2]]). 

===[https://www.epri.com/#/pages/product/000000003002009215/ EPRI/NRC-RES Fire Human Reliability Analysis Guidelines: Qualitative Guidance for Main Control Room Abandonment Scenarios (NUREG-1921 Supplement 1 / EPRI 3002009215)]===

Fire PRAs analyze a wide variety of fire-induced scenarios, one of which is fire damage rendering the main control room (MCR) either inhabitable or ineffective. As a result of this fire damage, operators cannot stay in the MCR and the command and control of the plant is transferred from the MCR to another location. This is commonly referred to a main control room abandonment (MCRA). MCRA is analyzed as a special case of fire HRA. While NUREG-1921/EPRI 1023001 briefly addressed abandonment, additional guidance and inputs are needed to properly address the unique contexts of abandonment scenarios. 

EPRI  3002009215 / NUREG-1921 Supplement 1 provides more detailed consideration of the elements to model for main control room abandonment, including:
*Adding logic in the model to capture when the conditions requiring abandonment occur, whether due to a loss of control room habitability or a loss of the capability to reach a safe-and-stable state from the control room. i.e., loss of control (Page 3-2)
*Incorporating random failures of equipment required for remote shutdown 
*Model logic for incorporating mitigatable fire-induced failures of equipment required for remote shutdown
*Modeling non-mitigatable fire-induced failures of equipment required for remote shutdown
*Model logic for capturing accident scenarios that create conditions beyond the capability of the remote shutdown equipment and/or procedures.
Refer to Section 3 of EPRI 3002009215 / NUREG-1921 Supplement 1 for more information on modeling alternate shutdown capability in the PRA logic model.

===[https://www.epri.com/#/pages/product/3002016053/ Methodology for Modeling Plant Trip Probabilities in Nuclear Power Plant Fire Probabilistic Risk Assessment (EPRI 3002016053)]===
Typically, a Fire PRA assumes a reactor trip for fire scenarios unless it can be demonstrated that no Fire PRA components are impacted and the reactor will not trip. As a result, very few Fire PRA scenarios are screened from quantification. The operating experience in the United States was reviewed and showed that only one in six events results in a plant trip, and the events are strongly correlated with a few ignition source bins mainly related to electrical distribution and the turbine generator. This report documents a methodology and screening process that can be applied to fire scenarios for ignition source bins where a plant trip is unlikely to occur. For qualifying scenarios, a 0.1 conditional trip probability can be applied.