ML25063A273

NRC-RES Sitewide, All Hazards Level 3 PRA Project: Integrated Site Risk for Two-Unit Site

ML25063A273
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Issue date:	03/04/2025
From:	Ball E, Keith Compton, Susan Cooper, Hathaway T, Alan Kuritzky, Brian Wagner Office of Nuclear Material Safety and Safeguards, NRC/RES/DRA/PRAB
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NRC-RES Sitewide, All Hazards Level 3 PRA Project:

Integrated Site Risk for Two-Unit Site Susan E. Cooper, Brian Wagner, Erick Ball, Keth Compton, Trey Hathaway, Alan Kuritzky US Nuclear Regulatory Commission, Rockville, MD, USA

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ABSTRACT The United States (U.S.) Nuclear Regulatory Commission (NRC) is performing a full-scope, sitewide Level 3 probabilistic risk assessment (PRA) project (L3PRA project) for a two-unit, pressurized-water reactor reference plant. The scope of this project includes all hazards (internal and external) and all radiological sources on the reference site: 1) two, nearly identical reactors; 2) two, hydraulically connected spent fuel pools (SFPs), and 3) a dry cask storage (DCS) facility. The NRC has published a series of reports documenting the results of all these PRA efforts.

In addition to the single unit PRAs performed for NRCs L3PRA project, a previously, never-performed task was added: integrated site risk. Integrated site risk is defined as the aggregation of risks from all relevant radiological sources. Of particular interest are those accident scenarios for which both reactors and the SFPs experience simultaneous radiological releases starting from at-power conditions.

Due to limited time and resources, the ISR task is focused on demonstrating a proof-of-concept, rather than full ISR quantification. Using the results of the sitewide dependency assessment with the multi-unit (MU) results from the Level 1 and Level 2 PRAs, a relatively large seismic event was identified as a good candidate scenario to analyze, since such an event can lead to a severe accident for both the reactors and SFPs. A scenario description was developed, including the timing of reactor and SFP behavior, the timing of plant conditions, and the timing of required operator actions. MU release category definitions were combined with SFP release categories to define release categories for the combination of the reactors and the SFPs. Then, MU consequence results calculated with the NRCs MACCS multi-source capability were combined with release category frequencies to generate MU risk.

Keywords: Integrated site risk (ISR), multi-unit PRA (MUPRA), multi-unit core damage frequency (MUCDF), MU release category frequency (MURCF), sitewide dependency assessment

1.

INTRODUCTION The United States (U.S.) Nuclear Regulatory Commission (NRC) is performing a full-scope site Level 3 probabilistic risk assessment (PRA) project (L3PRA project) for a two-unit pressurized-water reactor reference plant. The scope of this all-hazards, sitewide L3PRA project includes estimation of multi-unit (MU) risk and integrated site risk (ISR). This effort was directed by the Commission (see Staff Requirements Memorandum [1] that resulted from SECY-11-0089 [2]) and includes the following radiological sources on the reference site:

two (nearly identical) operating pressurized water reactor units (Unit 1 and Unit 2) two, hydraulically connected spent fuel pools (SFPs), one for each operating reactor unit (Unit 1 and Unit 2).

an independent dry cask storage (DCS) facility.

Previous NRC PRA efforts, such as NUREG-1150 [3], did not include consideration of risk from the SFPs or DCS, or risk of MU accidents. Consequently, ISR is a new task that has not been previously addressed in NRC PRA efforts. Note, for the L3PRA project, flexible mitigation strategies (FLEX) were addressed only in a sensitivity study, since the projects modeling freeze date is 2012.

In general, PRA has traditionally been performed for single units only, though the Seabrook PRA [4] was an early industry PRA effort that considered two-unit risk.1 In response to the 2011 Great Japan Earthquake and the events at the Fukushima Daiichi nuclear power station,2 there has been increased interest in MU risk in both the U.S. and internationally. Zhou and Modarres [10] summarize the history and current status of multi-unit PRA (MUPRA) development. In addition, there have been multiple responses to these events from the NRC, including lessons learned [11] and implications for PRA [12].

At present, public reports for the L3PRA project single unit reactor Level 1, 2, and 3 PRAs and DCS PRA have been published. However, the SFP PRA and ISR reports are still being finalized. A previous paper [13]

previewed some of the ISR task results to provide preliminary MU insights that might be applicable to advanced reactors.

2.

INTEGRATED SITE RISK APPROACH Due to limited time and resources, the ISR task focused on demonstrating a proof-of-concept, rather than full ISR quantification. Consequently, the goal of the ISR approach for the NRC's L3PRA project is to identify and describe a single scenario in which both reactors and the SFPs would have simultaneous (e.g.,

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of reactor trip) radiological releases.

The approach for the ISR task relies heavily on previously performed L3PRA project work, such as the single unit Level 1 and 2 PRAs, the SFP Level 1 and 2 PRAs, and multi-unit core damage frequency (MUCDF) results. At a high-level, the ISR approach consists of the following steps:

1. Perform a sitewide dependency assessment

2. Develop MUCDF results

3. Develop MU Level 2 and Level 3 PRA results

4. Define an illustrative integrated site risk scenario

5. Develop ISR Level 3 PRA results Step 1 is briefly summarized in this paper below. The remainder of the paper focuses on Steps 4 and 5.

Companion papers describe the NRCs MU Level 1 PRA results and the approach used to develop MU Level 2 PRA results.

Due to resource limitations, the ISR task was limited to at-power conditions. Furthermore, a single ISR scenario was developed due to resource and computational limitations for developing MU Level 2 PRA results.

In addition to the limited experience in performing MUPRAs described above, there is also limited guidance for developing Level 1 MUPRAs. Two reports by the International Atomic Energy Agency (IAEA) [14, 15]

1 Note that the second unit at the Seabrook site was never built.

2 See, for example, two reports released by the Government of Japan [5, 6], an IAEA report [7], and two Institute of Nuclear Power Operations INPO) reports [8, 9].

and an Electric Power Research Institute (EPRI) report [16] were used extensively in the L3PRA projects ISR task, especially for performing Steps 1 and 2. Similarly, there is little experience in integrating reactor risk with SFP risk. Two such efforts [17, 18] that integrate a single unit Level 2 PRA model with an SFP PRA were used by the L3PRA project.

3.

SITEWIDE DEPENDENCY ASSESSMENT The overall ISR task included a sitewide dependency assessment that is based upon the previous work by the IAEA [14, 15] and EPRI [16]. In particular, the L3PRA projects sitewide dependency assessment approach used a phased approach similar to that in the ERPI report and dependency categories that overlap those defined in both the IAEA and EPRI reports. The sitewide dependency assessment was performed from the perspective of both Level 1 and Level 2 PRAs.

For the L3PRA project, the three phases of sitewide dependency assessment were defined as follows:

Phase 1 Assessment: Sitewide and multi-unit initiating events Phase 2 Assessment: Shared physical resources, and shared or connected systems, structures, and components (SSCs)

Phase 3 Assessment: Identical components, proximity dependencies, human or organizational dependencies, accident propagation between units, potential hazards correlations The sitewide dependency assessment was performed first for the two reactors. Then, potential dependencies between the SFPs and the reactors were assessed.3 The results of this assessment identified that:

1. Only seismic events were important potential sitewide initiating events for both the reactors and the SFPs.

2. There were potential dependencies between the reactors in post-core damage conditions (i.e., Level 2 PRA) and the SFPs with respect to shared resources (e.g., portable pumps, water supplies) and human and organizational factors (i.e., sharing of available operators).

3. There were no other Phase 2 (e.g., shared physical resources or SSCs) or Phase 3 (e.g., identical components) dependencies.

Because of the general lack of dependencies between the reactors and the SFPs, a formal logic model did not need to be developed in order to evaluate sitewide risk.

4.

DEFINITION OF THE INTEGRATED SITE RISK SCENARIO When considering the objective of estimating sitewide risk, the L3PRA project team decided to focus on potential worst case scenarios that involved sitewide dependencies and consequences that occur close in time. For example:

What consequences could occur at the same time (e.g., nearly simultaneous releases from both reactors and SFPs)?

How (e.g., what dependencies and accident conditions must exist) could such a scenario occur?

4.1.

Approach for Defining the Illustrative ISR Scenario Because the sitewide dependency assessment (discussed in Section 3) identified a few sitewide dependencies between the reactors and SFPs, the ISR task focused on the identification and definition of 3 There were no potential dependencies identified between the reactors and the DCS facility.

an illustrative ISR scenario that involved both the reactors and the SFPs. For a variety of resource and scope limitations, a single illustrative ISR scenario was defined for a proof of concept analysis. Given that only one ISR scenario was going to be defined, the focus on a potential worst case scenario was further justified.

The approach used to select and define an illustrative ISR scenario involved all members of the project team. The project team used a variety of information sources, including:

the sitewide dependency assessment discussed in Section 3 the identification of MU and sitewide IEs the MU Level 1 risk results reported in Section 6 the details (e.g., time available for operator actions) of the (yet-to-be-published) single source SFP Level 1 and 2 PRA 4.2.

Selection of Seismic Bin 6 for the Illustrative ISR Scenario The illustrative ISR scenario that was selected by the project team was a large seismic event (i.e., seismic bin 6) for at-power conditions. The project team considered several factors in selecting seismic bin 6 for the illustrative scenario. Examples of these factors and their impact on the ISR scenario definition are:

From the Phase 1 sitewide dependency assessment, the ISR scenario should be a seismic event.

From the at-power, MU Level 1 PRA assessment:

o The combined MUCDF result for seismic events is over 50 percent of total MUCDF.

o Of the seismic events, bins 3 through 6 are the largest contributors to total MUCDF.

From the single source, SFP Level 1 and 2 PRA results for significant fuel uncovery frequency (SFUF):

o Seismic bins 5 through 7 are the most important for the at-power operating cycle phases of the SFPs.

o Mitigation is not credited for seismic bins 7 and 8 because: (1) the extreme sloshing in these cases will make the refuel floor immediately uninhabitable due to high radiation levels associated with low SFP water level (i.e., fuel uncovery is immediate), (2) the fuel handling building (FHB) itself may experience significant damage and thus be difficult to access, and (3) the potential for these extreme events to further degrade human performance.

o A simplified MELCOR model was used to calculate available times for operator actions for the SFP Level 1 and 2 PRAs, including the amount of time available to align makeup (depending on the size of the leak), the amount of sloshing (affected by the seismic bin),

and the operating cycle phase (OCP) (which affects decay heat).

4.3.

Description of Illustrative ISR Scenario The scenario that was selected and developed for the ISR task is summarized in Table 1. Table 1 consists of a timeline of major events, especially those that are important to modeling inputs used in the Level 1, Level 2, and Level 3 PRAs for both the reactors and the SFPs.

The accident progression for the reactors and the SFPs shown in Table 1 is set up by the initial conditions (i.e., both reactors at power and SFPs in nominal conditions), the specific seismic event (i.e., seismic bin 6), and response to the seismic event. For simplicity, both reactors are assumed to experience core damage and containment failure at the same time. Because only at-power conditions were addressed, only two

operating cycle phases needed to be considered for the SFPs (i.e., AAN1 and AAN54). Also, only three release categories (RCs)5 for the reactors were selected for the illustrative scenario:

containment isolation failure (CIF) late containment failure (LCF) intermediate containment failure due to burn (ICF-BURN)

The selection of seismic bin 6 for the illustrative ISR scenario prescribed many of the needed event details for quantifying the joint risk of both reactors and the SFPs. The most important of those details is occurrence of a large seismic event that:

is a significant contributor to MU and SFP risk causes widespread site SSC damage, including:

o widespread SSC damage that limits the effectiveness of reactor Level 1 PRA mitigation strategies o large SFP inventory losses that require mitigation o loss of offsite power (LOOP) and station blackout (SBO) conditions which limit the effective mitigation strategies for the SFPs, but do allow for the possibility of mitigation strategies being effective As noted above, the project team focused on the sharing of physical resources and the associated sharing of human resources between the reactors and SFPs, as well as the associated timing of potential mitigation strategies. The identified dependency involved the use of Extensive Damage Mitigation Guideline (EDMG) strategies that are modeled in the single unit, reactor Level 2 PRA. These reactor EDMG strategies require the same equipment, water resources, and operators that are modeled for the SFP makeup mitigation strategy.

In addition, required timing for both the reactor (i.e., approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after reactor trip) and SFP mitigation strategies (e.g., approximately 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> after reactor trip) is similar only for seismic bin 6. For larger seismic events, mitigation strategies are not feasible (i.e., no mitigation is possible for seismic bins 7 and 8). For less damaging seismic events (i.e., seismic bins 1 through 5), mitigation strategies are not needed until many days following the beginning of the accident scenario. The need for and timing of the SFP mitigation strategies is based on some unique modeling in the SFP Level 1 and 2 PRA, in which both liner failures and water sloshing out of the SFPs are projected to occur.

4 AAN1 and AAN5 involve both reactors at power and the SFPs in nominal conditions (i.e., no fuel handling activities).

They differ only in the time since one of the reactors was shut down for refueling.

5 MU Level 2 PRA is computationally challenging especially due to the potential number of RC combinations for two or more reactors. Consequently, a few of the more risk-important RCs were selected for the illustrative ISR scenario.

Table 1 Timing of Key Events in the Illustrative ISR Scenario Time Event(s)

Plant/SFP Conditions Notes/Assumptions

-0 Initial conditions Both reactors at power Both SFPs in nominal state6 0

Large seismic event Seismic bin 6 Loss of offsite power and SBO Sitewide event These losses are relevant to both reactors and the SFPs.

Significant water sloshing and liner failures Both SFPs These failures are assumed to be simultaneous.

Extensive structural damage from seismic event Sitewide Buildings relevant to both reactors and SFPs are affected.

Extensive equipment failure from seismic event Sitewide Equipment relevant to both reactors and SFPs are affected.

3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> General Emergency is declared See event column Core damage at:

- CIF = 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />s7

- LCF = 3.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />

- ICF-BURN = 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />s2 Reactors Reactors are assumed to experience core damage at the same time for the same RCs.

12 1/2 hours Evacuation complete:

- 45 minutes for notifications

- Additional 8 3/4 hours to evacuate emergency planning zone SFPs and reactors The same evacuation model is used for all seismic bins.

~10 hours Time available for external makeup for SFPs SFPs This is an EDMG strategy that uses a portable pump, fire water storage tanks, etc.

~22 hours Time available for containment spray cooling8 Reactors This is an EDMG strategy that uses a portable pump, fire water storage tanks, etc.

See event column Containment failure:

- CIF = 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />

- LCF = 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

- ICF-BURN = 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> Reactors Containment failures are assumed to be simultaneous for the same RCs.

See event column Containment releases:

- CIF = 21 & 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />

- LCF = 55 & 158 hours0.00183 days <br />0.0439 hours <br />2.612434e-4 weeks <br />6.0119e-5 months <br />

- ICF-BURN = 28 & 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> Reactors Xenon and Iodine releases for each RC Containment releases are assumed to be simultaneous for the same RCs.

See event column SFP releases:

- 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> (AAN1)

- 153 hours0.00177 days <br />0.0425 hours <br />2.529762e-4 weeks <br />5.82165e-5 months <br /> (AAN5)

SFPs 6 No fuel handling activities.

7 Core damage occurs later for CIF and ICF-BURN because, for these RCs, AFW is assumed to operate for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before batteries are depleted.

8 LCF and ICF-BURN only

5.

INTEGRATED SITE RISK RESULTS Both the approach and consequence results for the illustrative ISR scenario are summarized below.

5.1.

Approach for Developing Results for Illustrative ISR Scenario Involving Two Reactors and Spent Fuel Pools To develop results for the combination of both reactors and the SFPs, the project team used MU Level 2 PRA results and SFP Level 2 PRA results. Since the illustrative ISR scenario that the project team defined involved essentially simultaneous releases from all these sources, these results were combined. In turn, these combined release category frequencies (RCFs) were used as inputs to NRCs MACCS code, using its multi-source capability [19] to develop ISR consequences.

There are well-known computational challenges associated with the large number of potential MU RCs.

Consequently, a limited number of the most consequential MU RCs were addressed in the MU Level 2 PRA task (e.g., CIF-CIF and CIF-LCF). Table 2 shows the selected combinations of MU and SFP RCs9 and their combined RCFs.

Table 2 Frequency-Weighted MU Consequences RC combination Combined RCF (/rcy)

CIF-CIF-SFU5-AAN1 6.65E-10 CIF-CIF-SFU5-AAN5 1.97E-09 LCF-ICF-BURN-SFU5-AAN1 1.29E-09 LCF-ICF-BURN-SFU5-AAN5 3.82E-09 CIF-CIF-SFU6-AAN1 1.99E-09 CIF-CIF-SFU6-AAN5 5.90E-09 LCF-ICF-BURN-SFU6-AAN1 3.88E-09 LCF-ICF-BURN-SFU6-AAN5 1.15E-08 For this analysis, explicit MACCS multi-source analyses involving SFP source terms were not available at the time of writing. Instead, drawing on the insights from MU consequence analyses, it was assumed that the multi-source RCs could be reasonably approximated by the summation of the reactor MU multi-source consequences and the SFP consequences from the SFP Level 3 PRA. Because all RC combinations were assumed to be initiated by a severe seismic event, the MU multi-source consequences (assuming a degraded evacuation) were used. Table 3 shows some of the results10 that were generated.

9 SFP RCs include a Level 1 PRA to Level 2 PRA transition bin (e.g. SFU5), similar to a plant damage state in a reactor PRA, and an operating cycle phase (e.g., AAN1), that comes from discretization of the reactor and SFP operating cycle.

10 Results for total economic costs are not shown in Table 3, whereas results presented later include this risk measure.

Table 3 Reactor At-Power MU and SFP Consequences for Selected Integrated Site RC Combinations RC combination Individual Latent Fatality Risk, 0-10 mi (cases/person)

Collective Total Effective Dose, 0-50 miles (person-rem)

MU SFP MU +SFP MU SFP MU +SFP CIF, CIF, SFU5-AAN1 9.54E-04 1.20E-03 2.15E-03 6.56E+05 1.70E+06 2.36E+06 CIF, CIF, SFU5-AAN5 9.54E-04 7.60E-04 1.71E-03 6.56E+05 4.20E+05 1.08E+06 LCF, ICF-BURN, SFU5-AAN1 8.98E-04 1.20E-03 2.10E-03 5.43E+05 1.70E+06 2.24E+06 LCF, ICF-BURN, SFU5-AAN5 8.98E-04 7.60E-04 1.66E-03 5.43E+05 4.20E+05 9.63E+05 CIF, CIF, SFU6-AAN1 9.54E-04 1.20E-03 2.15E-03 6.56E+05 1.70E+06 2.36E+06 CIF, CIF, SFU6-AAN5 9.54E-04 7.60E-04 1.71E-03 6.56E+05 4.20E+05 1.08E+06 LCF, ICF-BURN, SFU6-AAN1 8.98E-04 1.20E-03 2.10E-03 5.43E+05 1.70E+06 2.24E+06 LCF, ICF-BURN, SFU6-AAN5 8.98E-04 7.60E-04 1.66E-03 5.43E+05 4.20E+05 9.63E+05 5.2.

Summary of Results for Illustrative ISR Scenario Context for the magnitude of MU risk compared to single unit risk is provided in Table 4. This table shows the risk from all modeled RC combinations for MU weather-related losses of offsite power (LOOPWR),

MU seismic bin 6 (EQK-BIN-6), and integrated site (MU and SFP combined) for seismic bin 6, and compares these results to the single unit risk estimates from the reactor at-power, reactor low-power and shutdown, and SFP analyses. Caution should be used in this comparison, because the MU results do not include the full set of initiators analyzed in the single unit analyses. However, it may be noted that total MU LOOPWR and EQK-BIN-6 risks are substantially less than those for the reactor at-power or low-power and shutdown but are comparable to those for the SFPs. The integrated reactor-SFP risk is substantially less than that for either the reactor at-power, reactor low power and shutdown, or SFPs.

Table 4 Summary of MU Risk Measures Scope Piece Release Freq. (/yr)

Individual Latent Fatality Risk, 0-10 mi (/yr)1 Collective Total Effective Dose Risk (person-rem/yr), 0-50 miles1 Total Economic Cost Risk, 0-50 mi (2015$/yr)1 All MU LOOPWR 5.52E-07 3.35E-10 0.14 1,220 All MU EQK-BIN-6 2.47E-06 1.84E-09 0.94 9,630 Integrated reactors and SFP for EQK-BIN-6 3.10E-08 5.55E-11 0.041 589 SFP (all hazards) 5.8E-07 6.3E-10 0.52 8,400 LPSD (Internal events) 4.7E-05 1.4E-08 9.7 150,000 At-Power (Internal events and internal floods) 6.9E-05 2.5E-08 9.9 80,000 At-Power (All hazards2) 1.6E-04 6.4E-08 27 230,000

6.

CONCLUSIONS The NRCs L3PRA project developed proof-of-concept, ISR results for an illustrative scenario. Combined MU and SFP consequences were developed for an illustrative large seismic event for which some potential dependencies between the reactors and SFPs were identified. Eight combinations of MURCs and SFP operating cycle phases were addressed and results for three risk measures were provided in this paper.

individual latent fatality risk collective total effective dose risk total economic cost risk The combined MU and SFP consequence results cannot be directly compared with the single source results because the scope of the analyses is different (e.g., total SFP results include both at-power and shutdown conditions). It can be generally concluded, however, that the results for the combined reactor and SFP risk are substantially less than that for either the reactor at-power, reactor low power and shutdown, or SFPs.

ACKNOWLEDGMENTS The NRC staff is grateful for the voluntary participation of the reference site in NRC/RES' site-wide, multi-hazard L3PRA project. Also, the authors are grateful for the input and support of the following members of NRCs L3PRA project: Jonathan DeJesus, Chris Hunter, Jeff Wood, Michelle Gonzalez, Latonia Enos-Sylla, Selim Sancaktar (retired).

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