HomeAbout NRCHow We RegulateResearchState-of-the-Art Reactor Consequence Analyses (SOARCA)The SOARCA Process

# SOARCA Process, Step 2: Modeling Onsite Accident Progression and  Mitigation Measures

In this step of the State-of-the-Art Reactor Consequence Analyses (SOARCA), the project team modeled the onsite progression, plant behavior, and mitigation measures for each of the severe accident scenarios selected in Step 1 of the SOARCA process. To learn more, please see the following topics on this page:
        - Defense-in-Depth Barriers and Their Importance to SOARCA
        - Onsite Accident Progression and Plant Behavior
        - Mitigation Measures

## Defense-in-Depth Barriers and Their Importance to SOARCA

After selecting the scenarios to model, the team determined whether those scenarios could possibly breach the three "defense-in-depth" barriers (illustrated in the figure to the right) and release any radioactive material from the reactor core to the environment. Specifically, the team considered the following barriers:
        - The  cladding  around the fuel in the reactor core
        - The  reactor coolant system  (including the  reactor pressure vessel  and associated components)
        - The  containment  building

To establish these barriers, nuclear fuel pellets are sealed in metal tubes (called cladding) to create fuel rods. These fuel rods are then bundled into fuel assemblies, which are loaded into a thick steel vessel, known as the reactor pressure vessel (RPV). Designed to withstand high pressures, the RPV is an important part of the reactor coolant system (RCS), which is housed inside a special containment building.

One type of reactor containment is a large cylinder-shaped building made of reinforced concrete with a steel lining. It is designed to withstand the pressures that might buildup as steam and gases escape from the reactor during an accident. This type of containment is used in the pressurized-water reactor (PWR) design of the Surry Power Station. Another type of containment, called a pressure-suppression containment, has a large water-filled pool to cool the steam and reduce the pressure buildup in the containment in the event of an accident. This type of containment is used in the boiling-water reactor (BWR) design of the Peach Bottom Atomic Power Station. Each of these types of reactor containment is designed to hold (or "contain") radioactive material that might otherwise be released to the outside environment in the event of a failure of the first two "defense-in-depth" barriers.

A severe accident involves an uncontrolled increase in the temperature of the reactor core. This can lead to severe core damage, in which the fuel and other core internal structures melt and collect at the bottom of the reactor vessel. However, *all three barriers must fail* before a significant release of radioactive material can occur. (This includes bypass events, because a bypass of the containment is a type of containment boundary failure.)

## Onsite Accident Progression and Plant Behavior

The postulated severe reactor accidents from Step 1 were chosen based on their likelihood. In step 2, the SOARCA project team used MELCOR (an integral severe accident analysis code) to model the onsite accident progression and plant response to the severe accident scenario. This determined whether the selected accident scenarios could breach the defense-in-depth barriers and lead to a release of radioactive material.

For both Peach Bottom and Surry, the team modeled the following scenarios, called ***station blackouts (SBOs)***, which met the criterion of having a core damage frequency (CDF) higher than 10-6 (i.e., "one-in-a-million") per year and were assumed to be caused by an earthquake more severe than the plant was designed to withstand:
        - *Long-Term Station Blackout* (LTSBO) —  In this scenario, the station loses all alternating current power sources, but battery backups operate safety systems for about 4 hours until the batteries are exhausted.
        - *Short-Term Station Blackout* (STSBO) —  In this scenario, the site loses all power (even the batteries) and, therefore, all of its safety systems quickly become inoperable in the "short term."

Other events (such as power grid failure, floods, or fire) can also cause these scenarios; however, the SOARCA team modeled the scenarios that presented the most severe challenge to plant operators. Additionally, the team modeled the following two scenarios for the PWR design at Surry, which met the criterion for high consequence events (CDF higher than 10-7):
        - *Interfacing-Systems Loss-of-Coolant Accident* (ISLOCA) —  In this scenario, a random failure of check valves causes a rupture in the low-pressure system piping outside the  containment .
        - *Thermally Induced Steam Generator Tube Rupture* (TISGTR) —  This scenario is a low-probability variation of the STSBO. While the  reactor core  is overheating and boiling off the available water, extremely hot steam and hydrogen flow out and cause a steam generator tube to rupture. (For more information, see the  Backgrounder on Steam Generator Tube Issues .)

Both of these scenarios are considered "bypass events," in which radioactive materials reach the environment without a structural containment failure.

## Mitigation Measures

To ensure a realistic evaluation of the postulated severe accident consequences, the SOARCA project includes insights into the effectiveness and benefits of mitigation measures at operating reactors. Because such measures may affect accident progression, SOARCA modeled mitigation measures that included each site’s emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs). They also included security-related mitigation measures, which encompass the additional equipment and strategies that the NRC has required since the terrorist attacks on September 11, 2001, to further improve mitigation capability. In addition, as part of the effort to increase the realism of the analyses, the SOARCA team completed table-top exercises of the selected scenarios to glean insights into operator actions for implementation of the available mitigation measures.

The team also used discussions and plant walkdowns to develop timelines for each scenario, reflecting the anticipated operator actions and equipment lineup or setup times to implement the available mitigation measures. The team then used the mitigation measures as inputs for MELCOR to model the severe accident scenarios within each plant. Using those inputs, the project team modeled the following cases for each of the selected accident scenarios:
        - Mitigated Scenario —  The SOARCA team modeled what would happen if the operators successfully carried out the mitigation measures. The MELCOR calculations included this information to understand how the mitigation measures could affect accident progression if operators successfully execute these measures (based on a qualitative evaluation of the known accident constraints, procedures, equipment, and operator training).
        - Unmitigated Scenario —  The team modeled what would happen if the operators failed to carry out key mitigation measures to prevent the accident from progressing.

Page Last Reviewed/Updated Tuesday, July 7, 2020

Section Navigation

#### Navigation

          - Step 1: Selecting Accident Scenarios to Model
          - Step 2: Modeling Onsite Accident Progression and Mitigation Measures
          - Step 3: Modeling Offsite Release of Radioactive Material
          - Step 4: Modeling Emergency Response
          - Step 5: Modeling Health Effects

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Page Last Reviewed/Updated Tuesday, July 7, 2020