ML18243A521

IMC 0609 Appendix H Containment Integrity Significant Determination Process

ML18243A521
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Issue date:	02/25/2019
From:	Matthew Leech NRC/NRR/DRA/APHB
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Leech M
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CN 19-008, DC 18-020
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NRC INSPECTION MANUAL INSPECTION MANUAL CHAPTER 0609 APPENDIX H CONTAINMENT INTEGRITY SIGNIFICANCE DETERMINATION PROCESS APHB Issue Date: 02/25/19 i 0609 App H TABLE OF CONTENTS 0609H-01 Introduction ......................................................................................................... 1 01.01 Applicability ......................................................................................................... 1 01.02 Entry Conditions ................................................................................................... 2 01.03 Appendix H Outline .............................................................................................. 2 01.04 Use of SAPHIRE Software to Calculate LERF ...................................................... 2 01.05 Use of Licensee Models for LERF 0609H-02 Limitations And Precautions ................................................................................ 3 0609H-03 Abbreviations and Definitions .............................................................................. 4 0609H-04 Overview of the Approach and Procedure for Significance Determination ........... 6 0609H-05 Consequential Steam Generator Tube Rupture (C-SGTR) ................................11 0609H-06 Procedure for Type A Findings ...........................................................................11 06.01 Approach for Assessing Type A Findings at Power ............................................ 12 06.02 Approach for Assessing Type A Findings During Shutdown .............................. 19 0609H-07 Procedure for Type B Findings ...........................................................................23 07.01 Approach for Assessing Type B Findings at Power ............................................ 23 07.02 Approach for Assessing Type B Findings at Shutdown ...................................... 27 0609H-08 REFERENCES ..................................................................................................37 List of Tables Table 4.1 Containment-Related SSCs Considered for LERF Implications ................................. 8 Table 6.1 Phase 1 Screening-Type A Findings at Full Power ...................................................16 Table 6.2 Phase 2 Assessment Factors -Type A Findings at Power ........................................17 Table 6.3: Manual ..................................................................................18 Table 6.4 Phase 2 Assessment Factors -Type A Findings at Shutdown ...................................22 Table 7.1 Phase 1 Screening-25 Table 7.2 Phase 2 Risk Significance -Type B Findings at Power .............................................26 Table 7.3 Phase 1 Screening-Type B Findings at Shutdown ....................................................30 Table 7.4 Phase 2 Risk Significance -Type B Findings at Shutdown ........................................31 Table 7.5 BWRs With Minimal Shutdown Mitigation Capability ................................................33 Table 7.6 BWRs With In-depth Shutdown Mitigation Capability ................................................34 Table 7.7 PWRs With Minimal Shutdown Mitigation Capability ................................................35 Table 7.8 PWRs With In-depth Shutdown Mitigation Capability ................................................36 List of Figures Figure 4.1 - LERF-based Significance Determination ................................................................10 Figure 6.1 - LERF-based Significance Evaluation for Type A Findings at Power .......................13 Figure 6.2 - LERF-based Significance Evaluation for Type A Findings at Shutdown .................21 Figure 7.1 - LERF-based Risk Significance for Evaluation Type B Findings at Power ...............24 Figure 7.2 - LERF-based Significance Evaluation for Type B Findings at Shutdown .................29 List of Attachments Attachment 1 Guidance for Assessing the Timing of Protective Actions in Detailed Risk Evaluations ..................................................................................... Att1-1 Attachment 2 Revision History for IMC 0609, Appendix H ............................................... Att2-1 Issue Date: 02/25/19 1 0609 App H 0609H-01 Introduction Core damage accidents that lead to large, unmitigated releases from containment in a time frame prior to effective evacuation of the close-in population have the potential to cause early health effects, e.g. prompt fatalities. The frequency of all accidents of this type is called the large early release frequency (LERF) as described in Regulatory Guide 1.174 (reference 1). Such accidents include unscrubbed releases associated with early containment failure at or shortly after reactor vessel breach, containment bypass events, and loss of containment isolation. The relationship of LERF thresholds to core damage frequency (CDF) thresholds found in Regulatory Guide 1.174 provides the basis for the risk significant characterizations found in Table 1.1 below. The LERF based approach is one order of magnitude more stringent than the CDF based approach. Therefore, it may be necessary under some circumstances to characterize the risk significance of an inspection finding using the LERF based approach. The purpose of this appendix to provide guidance for assessing the impact of inspection findings in relation to the containment barrier cornerstone of safety. The basis for the guidance presented in this appendix is discussed in IMC 0308, Reactor Oversight Process (ROP) Basis Document. Frequency Range/ry -4 Red Red < 10 10-5 Yellow Red < 10 10-6 White Yellow < 10 10-7 Green White <10-7 Green Green The significance determination process (SDP) assigns a risk characterization to inspection findings based on LERF considerations. This process is designed to interface directly with the SDP for Type A findings, derived from IMC 0609, Appendix A (at power) and Appendix G (shutdown), that are important LERF contributors. In addition, the guidance addresses findings related to structures, systems, and components (SSCs) that do not influence CDF determinations but can impact the containment function (i.e., Type B findings). It is recommended that inspectors, working with senior reactor analysts (SRAs) as needed, evaluate both Type A and Type B findings for at power findings. It is further recommended that SRAs evaluate both Type A and Type B findings for shutdown. Note: Type A and Type B findings are defined in Section 03.02, Definitions. 01.01 Applicability The guidance in this SDP is designed to provide NRC inspectors, SRAs, and NRC management with a simple probabilistic risk framework for use in identifying which findings are potentially risk-significant from a LERF perspective. Appendix H also helps facilitate communication of the basis for significance between the NRC and licensees. In addition, it identifies findings that do not warrant further NRC engagement, due to their very low risk significance.

Issue Date: 02/25/19 2 0609 App H 01.02 Entry Conditions The entry conditions for the containment integrity SDP described in this document are related to: Findings evaluated under IMC 0609, Appendix A (at power) or Appendix G (shutdown) that potentially increase LERF, or Degraded conditions affecting containment barrier integrity (that can potentially increase LERF without affecting CDF). Appendix H provides simplified risk-informed guidance for estimating the increase in LERF associated with inspection findings related to deficient licensee performance during full power (see IMC 0609, Appendix A) and shutdown operations (see IMC 0609, Appendix G). 01.03 Appendix H Outline The guidance presented in this appendix is based on a number of assumptions and modeling approximations. Section 02 presents the limitations and precautions that must be considered when evaluating inspection findings. Abbreviations and definitions used in this appendix are presented in Section 03. Section 04 is an overview of the approach and the procedure. Section 05 describes consequential steam generator tube ruptures (C-SGTR). Section 06 presents the procedure for analyzing those findings that have an impact on CDF (i.e., Type A findings) and Section 07 presents the procedure for analyzing those findings that only impact the containment function (i.e., Type B findings). Findings related to power operation and findings related to shutdown operations are both addressed. 01.04 Use of SAPHIRE Software to Calculate LERF Although this manual chapter provides the methods to estimate LERF manually, LERF can now be calculated automatically with Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) software. The LERF assessment factors for Type A LERF findings have now been programmed into all standardized plant analysis risk (SPAR) models for all plants using global linkage rules. The use of SAPHIRE to calculate LERF for Type A findings is the preferred method since it eliminates the need to manually list sequences and sum them using the worksheets and methods in this manual chapter. SAPHIRE can also provide values for LERF in Type A findings using the SDP analysis tool. It is important to note, however, that even the SAPHIRE results will produce a Phase 2 initial risk significance approximation and further refinement might be appropriate. 01.05 Use of Licensee Models for LERF If provided, LERF risk insights from the licensee risk model can be a source of risk information. The SRA should determine if the PRA model in question is capable of adequately evaluating the risk associated with the finding (e.g., licensee PRA may not model C-SGTR or Type B findings). Any evaluation using licensee provided information should be done by an SRA during the detailed risk evaluation.

Issue Date: 02/25/19 3 0609 App H 0609H-02 Limitations and Precautions Appendix H generates a reasonably conservative, order-of-magnitude assessment of the risk significance of inspection findings. The intent of Appendix H is to provide guidance for NRC inspectors to easily obtain a quick assessment of risk significance. If appropriate, a more detailed assessment may be performed in a SDP Phase 3 evaluation. The approach in this appendix has numerous assumptions and limitations which include the following: Since this SDP is focused on LERF, i.e., early fatality risk, long-term risk effects such as population dose and latent cancer fatalities are not addressed in this guidance. In addition, long-term accident sequences that involve failure of containment heat removal and ultimately progress to containment failure, e.g., loss of containment heat removal sequences in BWRs, are assumed not to contribute to LERF. It is assumed that effective emergency response actions can be taken within the long time frame of these accident sequences. For the evaluation of risk significance during shutdown, only the period within eight days of the beginning of the outage is considered. After eight days, it is assumed that the short-lived, volatile isotopes that are principally responsible for early health effects have decayed sufficiently such that the finding would not contribute to LERF. In addition, all core damage sequences are considered as candidate LERF sequences, because it is not known when evacuation would begin. LERF determinations depend on the containment design, and plant-specific attributes and features, which have considerable variability. It was conservatively assumed for all interfacing system loss-of-coolant-accidents (ISLOCAs) that the path outside containment is not submerged (i.e., the release is not scrubbed). It was conservatively assumed for all steam generator tube ruptures (SGTRs) that the secondary side is open so that a path outside containment exists and the release is not scrubbed. For those findings that impact the containment function (i.e., Type B findings), baseline CDFs for full power were assumed in order to simplify the calculation of the change in risk. The baseline CDFs for full power assumed were 10-4/ry for PWRs and 10-5/ry for BWRs. It was assumed, conservatively, that a main steam isolation valve (MSIV) leakage rate in excess of 10,000 scfh in BWRs (Reference 2) with Mark I and Mark II containments is significant to LERF.

Issue Date: 02/25/19 4 0609 App H 0609H-03 Abbreviations and Definitions 03.01 Abbreviations ATWS Anticipated Transient Without Scram CAP Corrective Action Program CCFP Conditional Containment Failure Probability CCW Component Cooling Water CD Core Damage CDF Core Damage Frequency CE Combustion Engineering C-SGTR Consequential Steam Generator Tube Rupture DF Decontamination Factor DHR Decay Heat Removal ECCS Emergency Core Cooling System IMC Inspection Manual Chapter LER Licensee Event Report LERF Large Early Release Frequency LOIA Loss of Instrument Air Initiator LOOP Loss of Offsite Power LORHR Loss of RHR Initiating Event LOSW Loss of Service Water Initiator LTOP Low Temperature Over Pressure Events POS Plant Operating State PRA Probabilistic Risk Assessment RCS Reactor Coolant System RHR Residual Heat Removal ROP Reactor Oversight Process RPV Reactor Pressure Vessel SCFH Standard Cubic Feet per Hour SDP Significance Determination Process SG Steam Generator SGTR Steam Generator Tube Rupture SPAR Standardized Plant Analysis Risk SSC Structure, System, or Component TS Technical Specifications TW Time Window TW-E Early Time Window, before refueling operation TW-L Late Time Window, after refueling operation Issue Date: 02/25/19 5 0609 App H 03.02 Definitions LERF: The frequency of those accidents leading to significant, unmitigated releases from containment in a time frame prior to effective evacuation of the close-in population such that there is a potential for early health effects. Close-in population: The population living or transiting within one mile of the reactor site boundary. This also includes nonessential plant personnel being evacuated from the site, any temporary population or local workforce, and any population that may be transiting through the area. Per the C, if there are no individuals residing within a mile of the plant boundary, for evaluation purposes, an individual should be assumed to reside 1 mile from the site boundary. Effective Evacuation: A set of actions by the licensee and local authorities that results in reasonable assumption that the close-in population has been evacuated. This does not require verification or certainty that every individual has left the area, only that all reasonable efforts have been completed and the population has had time to leave the area. Appendix H Phases of Significance Determination: Phase 1 - Characterization and Initial Screening of Findings: Precise characterization of the finding and an initial screening of very low-significance findings for disposition by the Phase 2 - Initial Risk Significance Approximation and Basis: Initial approximation of the risk significance of the finding and development of the basis for this determination for those findings that are not screened out in Phase 1 screening. Phase 3 - Risk Significance Finalization and Justification: Review and as-needed refinement of the risk significance estimation results from Phase 2, or development of any risk analysis outside of this guidance, by an NRC risk analyst (any departure from the guidance provided in this document constitutes a Phase 3 analysis and must be performed by an NRC risk analyst or SRA). Plant Operating States (POSs) During Shutdown: For BWRs POS 1: Starts when the RHR system is placed in service and RCS pressure is reduced below 135 psig with the MSIVs closed. The reactor pressure vessel (RPV) head is on. This POS covers part of hot shutdown (Mode 3) and cold shutdown (Mode 4) of the technical specifications (TS) modes. POS 2: Represents the shutdown condition when the RPV head is removed and reactor vessel water level is less than the minimum level required for movement of irradiated fuel assemblies within the RPV as defined by TS. This POS occurs during Mode 5. POS 3: Represents the shutdown condition when the RPV water level is greater than or equal to the minimum level required for movement of irradiated fuel assemblies within the RPV as defined by TS. This POS occurs during Mode 5.

Issue Date: 02/25/19 6 0609 App H For PWRs POS 1 - Starts when the RHR system is placed in service. The RCS is closed such that a steam generator could be used for decay heat removal, if the secondary side of a steam generator is filled. The RCS may have a bubble in the pressurizer. This POS ends when the RCS is vented such that the steam generators cannot sustain core heat removal. This POS typically includes Mode 4 (hot shutdown) and portions of Mode 5 (cold shutdown). POS 2 - Starts when the RCS is vented such that: (1) the steam generators cannot sustain core heat removal and (2) a sufficient vent path exists for feed and bleed. This POS includes portions of Mode 5 (cold shutdown) and Mode 6 (refueling). Reduced inventory operations and mid-loop operations with a vented RCS are subsets of this POS. Note: performance deficiencies occurring during a vacuum refill of the RCS require use of the POS 1 event trees. POS 3 - Represents the shutdown condition when the refueling cavity water level is at or above the minimum level required for movement of irradiated fuel assemblies within containment as defined by TS. This POS occurs during Mode 6. Early Time Window (TW-E): Represents the time before POS 3 is entered. The decay heat is relatively high. The reactor is in POS 1 or 2. Generally, TW-E represents the first eight days after shutdown. Late Time Window (TW-L): This time window represents the time after POS group 3. The decay heat is relatively low. The reactor is in POS 1, 2, or 3. Available: An SSC is considered available if it can be placed in service quickly enough to meet its function and all necessary supporting systems are functional (such as AC power, cooling water, and DC control power). Reduced Inventory Operations (PWR only): Exists whenever the RPV water level is lower than three feet below the reactor vessel flange. Shutdown Operations: Shutdown operation exists during hot shutdown, cold shutdown, and refueling when at least one fuel assembly is in the reactor vessel and the criteria for placing the RHR/DHR system into operation have been satisfied. 0609H-04 Overview of the Approach and Procedure for Significance Determination The guidance described in this section provides an assignment of a significance level (color) to inspection findings based on LERF considerations. This guidance considers findings resulting from deficient licensee performance during full power operations as well as shutdown operations. In Section 04.01, two distinct types of inspection findings that can potentially affect LERF are defined. Section 04.02 provides details of the overall approach taken to the assessment of their significance. 04.01 Types of Findings An inspection finding associated with a licensee performance deficiency during full power or shutdown operations is characterized by its potential impact on SSCs, by an estimate of the Issue Date: 02/25/19 7 0609 App H duration of this degradation, and by other information needed to assess the impact on accident likelihood or barrier cornerstone. Two types of findings are encountered: Type A Findings: Type A findings can influence the likelihood of accidents leading to core damage that are also identified as contributors to LERF. Such a finding will already have been processed using Appendix A of IMC 0609 for findings at full power, or IMC 0609, Appendix G for findings related Type B Findings: Type B findings are related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. Table 4.1 shows a list of SSCs (associated with maintaining containment integrity in different containment types). The LERF significance of these SSCs is also addressed in the table. 04.02 LERF-Based Significance Determination Process Figure 4.1 describes the process flow of typical inspection findings. Findings processed through a CDF-based SDP will be processed for potential LERF contribution as Type A findings. Findings that only impact the containment function without affected core damage sequences will be processed as Type B findings. Type A Findings For Type A findings, the CDF-based SDP guidance is used to determine the risk significance -7 per reactor year, then the finding should be assigned a Green significance level. -7 per reactor-year, then a screening is conducted using LERF screening criteria to assess whether any of the core damage sequences affected by the finding are potential LERF contributors. If none of the sequences is a LERF contributor there is no one or more of the affected and determine the increase in risk significance based on LERF considerations as discussed in detail in Section 05. Type B Findings Typprocessed through the CDF based SDP. These findings, however, are potentially important to ased on LERF considerations. As shown in Figure 4.1, an initial screening is conducted to determine if a finding is related to a containment SSC (see Table 4.1) or containment status that has an impact on LERF. If the answer is NO, the finding is Green. If the answer is YES, an assessment of the risk significance is performed using guidance provided in Section 06.

Issue Date: 02/25/19 8 0609 App H Table 4.1 Containment-Related SSCs Considered for LERF Implications1 SSC LERF Significance Containment penetration seals: BWR Mark I and II drywell or PWR containment BWR Mark III wetwell Failure of penetration seals that form a barrier between the containment and the environment can be important to LERF Containment isolation valves in lines: connecting BWR drywell or PWR containment airspace to environment connecting RCS to environment or open systems outside containment connected to closed systems inside/outside containment Large lines connecting containment airspace to environment (e.g., vent/purge) can contribute to LERF. Small lines (< 1-2 inch diameter) and lines connecting to closed systems would not generally contribute to LERF. Isolation valves connecting to RCS can contribute to ISLOCA Main Steam Isolation Valves Excessive MSIV leakage can contribute to LERF in high pressure accident sequences in BWR Mark I and II plants BWR drywell/containment sprays Mark I and II drywell sprays and Mark III containment sprays are important to preventing liner melt-through and mitigating suppression pool bypass Containment flooding system(s) Important to preventing liner melt-through in Mark Is PWR containment sprays and fan coolers Impact late containment failure and source terms, but not LERF Hydrogen control system - igniters - air return fans and hydrogen mixing systems Important to LERF in Mark III and ice condenser plants Not essential to hydrogen control if igniters are available 1 Some of the listed SSCs could affect the core damage frequency as well as LERF.

Issue Date: 02/25/19 9 0609 App H Suppression pool (SP) systems - components important to SP integrity/scrubbing (e.g., vacuum breakers) -suppression pool cooling Important to LERF in all BWR plants Impacts late containment failure but not LERF Ice condenser system ice condenser doors and ice bed air return fans ice mass air return fans foreign objects in ice compartment Significant flow blockage can be important to LERF Not important to LERF (similar to containment sprays) Deviations in weight of ice not important to LERF Not important to LERF (unless CDF is affected) Filtration systems Standby Gas Treatment System control room ventilation Not important to LERF due to unavailability in dominant sequences (e.g., SBO), plugging from high aerosol loadings in severe accident, and other considerations Spent fuel assemblies (individual) fuel handling accidents within pool fuel handling accidents outside pool Not important to LERF due to small fission product inventory contained in single fuel bundle. Scrubbing by water in the spent fuel pool further reduces releases.

Issue Date: 02/25/19 10 0609 App H Issue Date: 02/25/19 11 0609 App H 0609H-05 Consequential Steam Generator Tube Rupture (C-SGTR) Consequential steam generator tube rupture (C-SGTR) is an event in which steam generator tubes leak or fail as a consequence of the high differential pressure or elevated temperatures during accident conditions. The main accident scenarios of interest for C-SGTR are those that lead to core damage with high reactor pressure, dry steam generator, and low steam generator pressure (High-Dry-Low or HDL) conditions. A typical example of such an accident scenario is a station blackout with loss of auxiliary feedwater. Though other situations can lead to the potential for C-SGTR (e.g., over-pressure from ATWS, a large main steam line break, deliberate action to isolate feed to a faulted steam generator), these other sources are generally understood to be lower contributors to LERF. All of these situations are distinct from SGTR as an initiating event, which should continue to be treated as described elsewhere in this appendix. NUREG-2195 concluded that the overall contribution of C-SGTR scenarios to containment bypass is about a factor of 10 larger for combustion energy (CE) plants than Westinghouse plants2. Since C-SGTR is expected to contribute no more than 1-2% additional LERF for a typical Westinghouse plant, it is on par with other sources of LERF for these plants. Conversely, C-SGTR has the potential to be a much more significant contributor to LERF for CE plants, depending on the nature of the performance deficiency and its impact on the risk evaluation. Therefore, performance deficiencies that could significantly influence the likelihood of having high RCS pressure during core damage or that involve the reliability of feedwater for a CE plant should be evaluated for potential LERF findings from C-SGTR. The RASP Handbook provides the technical basis and a simplified worksheet to estimate LERF resulting from a C-SGTR. Westinghouse plants can also experience C-SGTR, but since the potential for it becoming a significant LERF contributor is lower, Appendix H does not require Westinghouse plants to be screened for C-SGTR. 0609H-06 Procedure for Type A Findings The CDF-based SDPs (Appendix A and Appendix G to IMC 0609) provide guidance for assessment of the significance of findings that impact CDF. This leads to identification of CDF sequences associated with each finding, evaluation of the increase in frequency of each of the contributing sequences, and dcontributing sequences collectively. Evaluation of the impact of the finding on LERF for these sequences is addressed using this appendix. Section 06.01 presents the procedure for Type A findings at full power, and 2 Two of the primary factors driving this different are hot leg diameter and steam generator inlet plenum design. CE plants tend to have larger hot legs that connect to the steam generator closer to the tube sheet, along with flat-bottomed steam generator inlet plenums. Westinghouse plants tend to have smaller hot legs that connect lower in the steam generator inlet plenum, along with rounded-bottomed steam generator inlet plenums. These design features tend to dictate the degree of mixing in the inlet plenum under high-dry-low conditions, resulting in a greater challenge to the tubes in the design typical of CE plants.

Issue Date: 02/25/19 12 0609 App H Section 06.02 presents the procedure for Type A findings at shutdown. 06.01 Approach for Assessing Type A Findings at Power This section provides the step-by-step process (as shown in Figure 6.1) for assessing the risk significance with respect to LERF of Type A findings at full power. As discussed previously, SAPHIRE can also be used to calculate LERF of Type A findings, and is the preferred method. STEP 1 Finding Characterization Determibe LERF contributors. Step 2 Accident Sequence Screening Generally, only a subset of those sequences contributing to CDF significance of a finding has the potential to impact LERF. A more detailed discussion of these sequences for each containment type is provided in IMC 0308, and briefly summarized below. BWRs For BWR Mark I and Mark II plants, findings related to ISLOCA, ATWS, and accidents with high RCS pressure (i.e., transients and small break LOCA). For BWR Mark I plants, accidents that involve a dry drywell floor at vessel breach regardless of whether the RCS is at low or high pressure also need to be evaluated in Phase 2 as indicated in Note 3 to Table 6.1. For BWR Mark III plants, findings related to ISLOCA, transients, small break LOCAs, and station blackout (SBO) categories. PWRs For PWR plants with large dry and sub-atmospheric containments, findings related to the accident categories ISLOCA and SGTR. Certain accident sequences that lead to core damage with high reactor pressure, dry steam generator, and low steam generator pressure (High-Dry-Low or HDL) conditions can lead to a consequential steam generator tube rupture (C-SGTR). A typical example of such an accident scenario is a station blackout with a loss of auxiliary feedwater. Though other situations can lead to the potential for C-SGTR (e.g. over-pressure from ATWS, a large main steam line break). A C-SGTR is more of a concern for Combustion Engineering (CE) plants. Consult Volume 5 of the RASP Manual for more information. For the PWR plants with ice condenser containments, findings related to ISLOCA, SGTR, and SBO accident categories.

Issue Date: 02/25/19 13 0609 App H Use Table 6.2 to Identify LERF Factors Use Worksheet Table 6.3 Figure 6.1: Road Map for LERF-based Risk Significance Evaluation for Type A Findings at Power Use Table 6.1 None of the Sequences Important to LERF? Use Table 6.1 None of the Sequences Important to LERF?

Issue Date: 02/25/19 14 0609 App H Accident categories that are screened out in Phase 1 include: LOOPs with successful emergency AC power operation (non-SBO events). LOOPs with failure of emergency AC power in which power is recovered prior to core damage. In general, sequences with late core damage (i.e., sequences that proceed to core damage due to loss of containment heat removal) will not contribute to LERF. Other sequences that are screened out are summarized below. When screening out these events for PWRs, use caution not to overlook High-Dry-Low sequences which could result in a C-SGTR that would be significant for LERF. BWRs ATWS sequences are not important contributors to LERF for BWRs with Mark III containments. Containment failure from ATWS sequences occurs due to gradual over-pressurization of containment prior to core damage. However, these sequences leave the drywell and suppression pool intact, hence the releases are scrubbed and a large early release does not occur. PWRs ATWS sequences are usually not significant contributors to LERF for PWRs. During a PWR ATWS, containment pressure increases slowly and is therefore a late failure mode. The risk significance determined by the CDF-based SDP for ATWS events in PWRs is sufficient. An exception to this would be an ATWS sequence coincident with a loss of feed that could lead to a C-SGTR. High and low pressure core damage sequences (in which the containment is not bypassed) are not significant contributors to LERF for PWRs with large dry and sub-atmospheric containments. An important insight from the IPE program and other PRAs is that the conditional probability of early containment failure is less than 0.1 for core damage accident scenarios that leave the RCS at high pressure. If the RCS is depressurized, the probability of early containment failure is less than 0.01. In PWRs with ice condenser containments, severe accident studies indicate that the most significant factor is the availability of hydrogen igniters and the ice condenser to mitigate severe accidents. If the igniters are available (i.e., non-SBO accidents), the conditional early containment failure probability is less than 0.1 even during accidents that leave the RCS at high pressure. Step 2.1 -7 per year, the LERF significance is Green and further LERF-related evaluation is not needed. Otherwise, proceed to Step 2.2. Step 2.2 -8 per year with those in Table 6.1 to identify those sequences which have the potential to affect LERF. Individual sequence results that are <1E-8 are not significant and are not evaluated further. However, -8 (sequence result of 8 or less) are evaluated for the Issue Date: 02/25/19 15 0609 App H overall LERF contribution. If none of the sequences impacts LERF, the risk significance LERF-to affect LERF3, proceed to Step 3. Step 3 Phase 2 Assessment For sequences needing Phase 2 analysis, risk significance determination is performed using the following two substeps: Step 3.1 LERF Factor Determination Identify the LERF factor associated with each of the sequences remaining after screening using Table 6.2. Document these sequences and their associated LERF factors as discussed in the next substep. Step 3.2 LERF Significance Evaluation Document details of LERF significance assessment using the LERF worksheet (Table 6.3). List each sequence assessed in Phase 2 in Column 1 together with its CDF score (in Column 2). Document the sequence attributes that make it a potential LERF contributor (e.g., high RCS pressure, drywell floor status for BWRs, etc.) in Column 3. Document the LERF factor (see Step 3.1) in Column 4. Document the LERF score in Column 5. The LERF score is calculated score (Column 2) by the LERF factor (Cscore of 7 (i.e., 1E-7) and the associated LERF factors is 0.4, the LERF score is 4x10-8. Step 4 LERF Significance Sum the scores for all of the LERF contributing sequences associated with the finding and enter .1. Step 5 Finding Significance Compare the CDF significance (color) with that for the LERF significance for the same finding. The higher (color) is the preliminary risk significance of the finding. 3No extra credit should be given for Severe Accident Management operator recovery actions (e.g., actions to depressurize the RCS or to flood Mark I drywell) unless recovery is explicitly modeled in the CDF sequence. Defer such recovery credit to Phase 3 assessment if needed.

Issue Date: 02/25/19 16 0609 App H Table 6.1 Phase 1 Screening-Type A Findings at Full Power Reactor Type Containment Type Attributes of Accident Sequence Related to Finding ISLOCA SGTR ATWS SBO (Note 1) High RCS Pressure (Note 2) All Others BWR Mark I Perform Phase 2 Not Applicable Perform Phase 2 Perform Phase 2 Perform Phase 2 Note 3 BWR Mark II Perform Phase 2 Not Applicable Perform Phase 2 Perform Phase 2 Perform Phase 2 Screen Out (Note 4) BWR Mark III Perform Phase 2 Not Applicable Screen Out (Note 4) Perform Phase 2 Perform Phase 2 Screen Out (Note 4) PWR Large Dry and Sub-Atmospheric Perform Phase 2 Perform Phase 2 Screen Out (Note 4) Screen Out (Note 4) Screen Out (Note 4) Screen Out (Note 4) PWR Combustion Engineering Plants Perform Phase 2 Perform Phase 2 (Note 5) (Note 5) (Note 5) (Note 5) PWR Ice Condenser Perform Phase 2 Perform Phase 2 Screen Out (Note 4) Perform Phase 2 Screen Out (Note 4) Screen Out (Note 4) Note 1: SBO is defined as a LOOP sequence with loss of emergency AC and failure to recover AC power. Note 2: High pressure is defined as greater than 250psi at the time of reactor vessel breach. Transients and small break LOCAs (smaller than about 2-inch equivalent break size in BWRs and 0.75 - 1 inch in PWRs) will usually result in pressures in the RCS greater than 250psi at the time of reactor vessel melt-through in the absence of manual depressurization. Consider a sequence to be low pressure in case of: Large or intermediate LOCA Sequences that include successful depressurization (DEP) Availability of low pressure injection (LPI) is questioned on sequence branch Consider a sequence to be high pressure in case of: The sequence includes failure of depressurization (DEP) None of the low pressure considerations identified above apply Note 3: A Phase 2 assessment should be performed for any sequences that are expected to proceed to reactor vessel breach into a dry reactor cavity. Therefore, all other transients with successful RCS depressurization should be assessed. Sequences involving LOCAs in the drywell or drywell spray operation are excluded because they result in a flooded drywell floor. LOCAs involving stuck open relief valve sequences do not result in flooded drywell. Note 4: Screen out means that the accident sequence related to the finding is not significant to LERF and is Green. Note 5: CE plants should be screened for C-SGTR. Refer to the RASP Manual Volume 5 for more information.

Issue Date: 02/25/19 17 0609 App H Table 6.2 Phase 2 Assessment Factors - Type A Findings at Power Reactor Type Containment Type Attributes of Accident Sequence Related to Finding ISLOCA SGTR ATWS SBO (Note 1) High RCS Pressure (Note 2) Low RCS Pressure (Note 2) BWR Mark I 1.0 Not Applicable 0.3 (Note 3) 0.6 If drywell is Flooded <0.1 If drywell is Flooded 1.0 If drywell is Dry 1.0 If drywell is Dry BWR Mark II 1.0 Not Applicable 0.4 (Note 4) 0.3 Screen Out in Phase 1 BWR Mark III 1.0 Not Applicable Screen Out in Phase 1 0.2 0.2 Screen Out in Phase 1 PWR Large Dry and Sub-Atmospheric 1.0 1.0 Screen Out in Phase 1 Screen Out in Phase 1 Screen Out in Phase 1 Screen Out in Phase 1 PWR Ice Condenser 1.0 1.0 Screen Out in Phase 1 1.0 Screen Out in Phase 1 Screen Out in Phase 1 Note 1: SBO is defined as a LOOP sequence with loss of emergency AC and failure to recover AC power. Note 2: High pressure is defined as greater than 250psi at the time of reactor vessel breach. Transients and small break LOCAs (smaller than about 2-inch equivalent break size in BWRs and 0.75 - 1 inch in PWRs) will usually result in pressures in the RCS greater than 250psi at the time of reactor vessel melt- through in the absence of manual depressurization. Note 3: If the RCS is at high pressure during the SBO then the Factors for the high pressure column apply. If the RCS is at low pressure during the SBO, the factors for the low pressure column apply. Note 4: If the RCS is at high pressure during the SBO then the Factor is 0.3. If the RCS is at low pressure during the SBO, the finding can be screened out.

Issue Date: 02/25/19 18 0609 App H Table 6.3: Manual (1) Sequences (2) (X) (3) Sequence Attributes (4) LERF Factor (Table 6.2 for power, Table 6.4 for shutdown) (F) (5) Score F * (1x10-X)

Issue Date: 02/25/19 19 0609 App H 06.02 Approach for Assessing Type A Findings During Shutdown This section provides a step-by-step process (shown in Figure 6.2) for assessing the risk significance with respect to LERF of Type A findings applicable to shutdown operation. STEP 1 Finding Characterization Step 1.1 Review the assessment performed using IMC 609, Appendix G, to identify the sequences affected by the finding, and the POSs and time windows (TWs) applicable to the finding. Step 1.2 Determine the status of containment when the finding occurred for each POS and TW: For PWRs and BWR Mark IIIs, the status of containment is either open or intact. For BWRs Mark I and IIs, the status of containment is either intact, de-inerted, or open. STEP 2 Accident Sequence Screening Step 2.1 For each shutdown core damage scenario identified in Step 1, determine if the following conditions were met: The finding occurred while the plant was in POS 1E or POS 2E. The finding occurred within the first eight days of the outage. Step 2.2 If both conditions in Step 2.1 were met, go to Step 3. Otherwise, the LERF significance is Green and further evaluation for LERF implications is not needed. STEP 3 Phase 2 Assessment For sequences needing Phase 2 analysis, risk significance determination is performed using the following two substeps: Step 3.1 Determine the LERF factor for each core damage scenario affected by the finding for the appropriate containment status using Table 6.4.

Issue Date: 02/25/19 20 0609 App H Step 3.2 Document details of LERF significance assessment for the finding being evaluated using the LERF worksheet (Table 6.3). List each sequence assessed in Phase 2 in Column 1 together with its CDF score (in Column 2). Since all core damage sequences are potential LERF contributors, Column 3 may be left blank. Document the LERF factor (see Step 3.1) in Column 4. Document the LERF score in Column 5. The LERF score is calculated by multiplying the olumn 4). For example, if a sequence has a -8. STEP 4 LERF Significance Sum the scores for all of the LERF contributing sequences associated with the finding being in the space below Column 5 of the completed Table 61.1. STEP 5 - Finding Significance Compare the CDF significance (color) with that for the LERF significance for the same finding. The higher (color) is the preliminary risk significance of the finding.

Issue Date: 02/25/19 21 0609 App H Figure 6.2: Road Map for LERF-based Risk Significance Evaluation for Type A Findings at Shutdown Use Table 6.3 Worksheet Use Table 6.4 for LERF Factors Issue Date: 02/25/19 22 0609 App H Table 6.4 Phase 2 Assessment Factors -Type A Findings at Shutdown Reactor/Containment Type Containment Status (Note 1) Accident Sequence Related to Finding Finding occurs: (1) in POS 1E or POS 2E within first eight days of outage. All Others BWR Mark I and II De-inerted 1.0 Screened Out BWR Mark III Intact 0.2 - if igniters are not available (Note 2) Screened Out Screen Out - if igniters are available (Note 3) PWR Large Dry and Sub-Atmospheric Intact Screen Out (Note 3) Screened Out Ice Condenser Intact 1.0 - if igniters are not available (Note 2) Screened Out Screen Out if igniters are available (Note 2 and Note 3) All Open 1.0 Screened Out Note 1: An intact containment is one in which the licensee intends to: (1) close all containment penetrations with a single barrier or can be closed in time to control the release of radioactive material, and (2) maintain the containment differential pressure capability necessary to stay intact following a severe accident at shutdown. When the RCS is open, an intact containment means that containment can be re-closed prior to boiling the RCS inventory. If the licensee does not intend to maintain an intact containment, then containment is open. If a PWR licensee is not maintaining an intact containment during POS 1E and POS 2E, then this should be reported to NRR/CPSB, so that it can be tracked to meet SRM 97-168. A de-inerted containment is one in which limits on the primary containment oxygen concentration as defined in TS are no longer met. Note 2: There are no TS for igniters to be operable during shutdown. However, it is possible that igniters could be recovered by operator action, in which case the finding could be screened out (i.e. not significant to LERF) Note 3: To screen out the finding, the analclosure consider: 1) time to boiling given a loss of RCS inventory which shortens time to boiling compared to a loss of the operating train of RH. (NOTE: selecting time to boiling based on RCS level at the bottom of the hot leg should always meet the recommendation) and (2) a potential loss of offsite power and a loss of all vital AC power.

Issue Date: 02/25/19 23 0609 App H 0609H-07 Procedure for Type B Findings Type B findings have no direct impact on the likelihood of core damage but have potentially important implications for containment integrity. This section provides the procedure for evaluation of LERF significance of Type B findings. Similar to the Type A findings approach, a step wise process (Figure 7.1) is used, which leads to a conservative estimate of LERF significance. Section 07.01 presents the procedure for findings at full power, and Section 07.02 for findings at shutdown. 07.01 Approach for Assessing Type B Findings at Power STEP 1 Finding Characterization Characterize the finding in terms of its relationship to the containment barrier function. Collect information needed for significance determination: SSCs affected and the nature of the degradation; the duration (i.e., >30days, 30-3 days, and < 3 days) of the degraded condition; information such as the magnitude of the leakage or number and location of inoperable hydrogen igniters. The type of information required can be inferred from Table 7.1 below. STEP 2 Screening of Finding Determine if the finding is associated with an SSC(s) important to LERF, using Table 7.1. If the finding is screened out then no further assessment is needed and the finding is Green. Otherwise, proceed to Step 3 below. Note that a detailed description of finding to be assessed in Step 3 is included in Table 7.2. STEP 3 Phase 2 Assessment Use Table 7.2 to provide a significance assignment to a Type B finding. For inspection findings involving leakage rates (e.g., MSIV leakage, containment leakage), if the as-found leakage rate is less than the values listed in Table 7.2, the finding is Green.

Issue Date: 02/25/19 24 0609 App H Figure 7.1: Road Map for LERF-based Risk Significance for Evaluation Type B Findings at Power Table 7.1 Table 7.1 Use Table 7.2 Use Table 7.2 Issue Date: 02/25/19 25 0609 App H Table 7.1 Phase 1 Screening-Type B Findings at Power Reactor Type Containment Type SSC Affected by Finding Containment Penetration Seals, Isolation Valves, Vent and Purge Systems Ice Condenser Flow Blockage Suppression Pool Integrity MSIV Leakage Drywell / Containment Sprays Igniters BWR Mark I Perform Phase 2 Not Applicable Perform Phase 2 Perform Phase 2 Perform Phase 2 Not Applicable BWR Mark II Perform Phase 2 Not Applicable Perform Phase 2 Perform Phase 2 Perform Phase 2 Not Applicable BWR Mark III Perform Phase 2 Not Applicable Perform Phase 2 Not1 Applicable Perform Phase 2 Perform Phase 2 PWR Large Dry and Sub-Atmospheric Perform Phase 2 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable PWR Ice Condenser Perform Phase 2 Perform Phase 2 Not Applicable Not Applicable Not Applicable Not Applicable Note 1: Some BWR Mark lll containments may have a safety-grade low-leakage Main Steam Shutoff Valve (MSSV) outside of the out-board MSIV. (This may have been abandoned in some plants) Reference (2)

Issue Date: 02/25/19 26 0609 App H Table 7.2 Phase 2 Risk Significance - Type B Findings at Power Reactor Type Containment Type Finding Risk Significance > 30 days 30 - 3 days < 3 days BWR Mark I and Mark II Leakage from drywell to environment >100 % containment volume/day through containment penetration seals, isolation valves or vent and purge systems Yellow White Green Failure of systems/components critical to suppression pool integrity/scrubbing (vacuum breakers or other bypass mechanisms) Yellow White Green Main steam isolation valve leakage >10,000 scfh through the best-sealing valve in any steam line (see Reference 2) Yellow White Green Mark I Drywell sprays unavailable Yellow White Green Mark II Drywell sprays unavailable White Green Green BWR Mark III (NOTE 1) Leakage from wetwell to environment >1,000 % containment volume/day through containment penetration seals, isolation valves or vent and purge systems White Green Green Failure of systems/components critical to suppression pool integrity/scrubbing (vacuum breakers or other bypass mechanisms) Yellow White Green Failure of multiple igniters such that coverage is lost in two adjacent compartments White Green Green Containment sprays unavailable White Green Green PWR Large Dry and Sub-Atmospheric Leakage from containment to environment >100 % containment volume/day through containment penetration seals, isolation valves or vent and purge systems Red Yellow White PWR Ice Condenser (NOTE 1) Leakage from containment to environment >100 % containment volume/day through containment penetration seals, isolation valves or vent and purge systems Red Yellow White Blockage of more than 15% of the flow passage into or through the ice bed Red Yellow White Failure of multiple igniters such that coverage is lost in two adjacent compartments Red Yellow White See notes on the following page Issue Date: 02/25/19 27 0609 App H Notes associated with Table 7.2 Note 1: For BWR Mark III containments and PWR ice condenser plants, the term compartments is used interchangeably with the term regions or zones and relates to the likelihood that hydrogen concentrations could rise to levels that could challenge containment. For a particular finding, the intent is to determine if the igniter system would remain effective at controlling concentrations in this -of-thumb. If it is not clear whether the igniter system would remain effective, the inspector should refer to the text in Section 5.2.3 (Mark III containments) or Section 7.2.4 (ice condenser containments) of NUREG-1765, and consult the design-basis analysis associated with the igniter system. If it is still unclear, the inspector should contact the regional SRA and headquarters staff knowledgeable in this area. 07.02 Approach for Assessing Type B Findings at Shutdown STEP 1 Finding Characterization Figure 7.2 shows the process flow for this approach. Characterize the finding in terms of its relationship to the containment barrier function. Collect information needed for significance determination, specifically the SSCs affected and the nature of the degradation, the duration of the degraded condition if less than the complete outage and if the condition had existed before shutdown (during power operation), or could exist upon change of plant/containment status (e.g. return to power) and information such as the magnitude of the leakage or the number and location of the inoperable hydrogen igniters. The type of information required can be inferred from Table 7.4 below. In addition, identify each POS(s) and time windows with which the finding is associated. STEP 2 Accident Sequence Screening STEP 2.1 Screen on the Basis of POS and Time Window If the finding occurs (1) in POS 1 or POS 2 AND (2) in TW-E, AND (3) within eight days of the start of the outage, THEN, go to Step 2.2. Otherwise, screen the finding as Green. STEP 2.2 Screen on the Basis of the Impact of the Finding Determine if the finding is associated with an SSC(s) important to LERF using Table 7.3. Consideration of Items A through D (as applicable) facilitates the use of Table 7.3. A. Did the finding involve the licensee failing to maintain the capability to close containment (maintain an intact containment) when the licensee planned to maintain an intact containment? This question applies to PWR and BWR Mark III licensees only. If yes, go to Table 7.3, containment status is intact. If no, continue with Step B. B. Did the finding involve hydrogen igniters in a BWR Mark III or a PWR ice condenser containment and the licensee maintained an intact containment? If yes, go to Table 7.3, containment status is intact. If no, continue with Step C. C. Did the finding occur when the containment was de-inerted for a Mark I or Mark II containment?

Issue Date: 02/25/19 28 0609 App H If yes, go to Table 7.3, containment status is de-inerted. If no, continue with Step D. D. Did the licensee intend to have an open containment without the capability to reclose containment? If yes, go to Table 7.3, containment status is open. NOTE: If a PWR licensee is not maintaining an intact containment during POS 1E and POS 2E, this may be a significant finding under the Maintenance Rule. Check with an SRA for further guidance. If no, screen out the finding. If the finding is screened out, it is assigned Green significance and no further assessment is needed. Otherwise, proceed to Step 3 below. STEP 3 Phase 2 Assessment Determine if shutdown mitigation capability is minimal or in-depth or closely resembles an in-depth or minimal capability. Use Tables 7.5 and 7.6 for BWRs or Tables 7.7 and 7.8 for PWRs to help make this determination. NOTE: For PWRs, if mitigation capability does not match with the tables, choose between in-depth or minimal capability based on: (1) availability of SGs and (2) availability of ECCS pumps and charging pumps Use Table 7.4 to determine color of finding. NOTE: Should the duration of a Type B finding exist for less than eight hours, then the color finding is reduced by one order of magnitude. NOTE: Findings that may have existed before shutdown (during power operation) or that could impact LERF upon change of plant/containment status (e.g. return to power) should be assessed. In case the finding is judged to impact power operation, Section 07.01 guidance should be used in the assessment.

Issue Date: 02/25/19 29 0609, App. H Use Table 7.3 Use Table 7.3 Use Table 7.4 Use Table 7.4 Level of Shutdown Mitigation Capability Tables 7.5, 7.6, 7.7 & 7.8 Level of Shutdown Mitigation Capability Tables 7.5, 7.6, 7.7 & 7.8 Figure 7.2: Road Map for LERF-based Risk Significance Evaluation for Type B Findings at Shutdown Issue Date: 02/25/19 30 0609, App. H Table 7.3 Phase 1 Screening-Type B Findings at Shutdown Reactor/ Containment Type Contain-ment Status (Note 1) SSC Affected by Finding Containment Penetration Seals, Isolation Valves, Vent and Purge Systems Suppression Pool Integrity Drywell/Contain-ment Sprays Igniters BWR Mark I and II De-inerted BWR Mark III Intact Perform Phase 2 Perform Phase 2 Screen Out (not important to LERF) Perform Phase 2 PWR Large Dry and Sub-Atmospheric Intact Perform Phase 2 Not Applicable Screen Out (not important to LERF) Not Applicable PWR Ice Condenser Intact Perform Phase 2 Not Applicable Screen Out (not important to LERF) Perform Phase 2 All Open Note 1: An intact containment is one in which the licensee intends to: (1) close all containment penetrations with a single barrier or can be closed in time to control the release of radioactive material, and (2) maintain the containment differential pressure capability necessary to stay intact following a severe accident at shutdown. When the RCS is open, an intact containment means that containment can be reclosed prior RCS boiling. A Type B performance deficiency results when a licensee intends to have an intact containment but cannot maintain that capability due to a performance deficiency. For Mark III containments, the definition of intact containment applies to primary containment. If the licensee does not intend to maintain an intact containment, then containment is open. If a PWR licensee is not maintaining an intact containment during POS 1E and POS 2E, then this observation could be risk significant under the Maintenance Rule and should be reported to an SRA. A de-inerted containment is one in which limits on the primary containment oxygen concentration as defined in TS are no longer maintained. Note 2: -inerted and expected to fail due to hydrogen combustion, regardless of Type B finding. However, findings that may have existed before shutdown or that could impact LERF upon change of plant/containment status (e.g. return to power) should be assessed. Note 3: LERF because containment is already open and cannot be re-closed. However, findings that may have existed before shutdown or that could impact LERF upon change of plant/containment status (e.g. return to power) should be assessed. If a PWR licensee is not maintaining an intact containment during POS 1E and POS 2E, then this observation could be risk significant under the Maintenance Rule and should be reported to an SRA.

Issue Date: 02/25/19 31 0609, App. H Table 7.4 Phase 2 Risk Significance -Type B Findings at Shutdown (For POS 1/TW-E and POS 2/TW-E in which the finding occurs during the first eight days of the outage) Reactor/ Containment Type Containment Status (NOTE 1) Finding Risk Significance (NOTE 2) Minimal Capability In-depth Capability BWR Mark I, II De-inerted Screened Out in Phase 1 N/A N/A BWR Mark III Intact Leakage from containment to environment > 1000 % containment volume/day through containment penetration seals, isolation valves or vent and purge systems with suppression pool integrity (NOTE 3) POS 1E -Yellow POS 1E- White POS 2E - Yellow POS 2E - Green BWR Mark III Intact Loss of suppression pool integrity (NOTE 4) POS 1E -Yellow POS 1E- White POS 2E - Yellow POS 2E - Green BWR Mark III (NOTE 5) Intact Failure of multiple igniters such that coverage is lost in two adjacent compartments given that primary containment is intact POS 1E - White POS 2E- Green POS 2E - White POS 2E - Green PWR Large Dry and Sub-Atmospheric Intact Leakage from containment to environment > 100 % containment volume/day through containment penetration seals, isolation valves or vent and purge systems POS 1E -Yellow POS 1E - White POS 2E - Red POS 2E - White PWR Ice Condenser (NOTE 5) Intact Leakage from containment to environment >100 % containment volume/day through containment penetration seals, isolation valves or vent and purge systems POS 1E - Yellow POS 1E - White POS 2E - Red POS 2E - White Failure of multiple igniters such that coverage is lost in two adjacent compartments POS 1E - Yellow POS 1E - White POS 2E - Red POS 2E - White All Open Screened Out in Phase 1 Green Green See notes on the following page Issue Date: 02/25/19 32 0609, App. H Notes associated with Table 7.4 Note 1: An intact containment is one in which the licensee intends to: (1) close all containment penetrations with a single barrier or can be closed in time to control the release of radioactive material, and (2) maintain the containment differential pressure capability necessary to stay intact following a severe accident at shutdown. When the RCS is open, an intact containment means that containment can be re-closed prior to RCS boiling. A Type B performance deficiency results when a licensee intends to have an intact containment but cannot maintain that capability due to a performance deficiency. For Mark III containments, the definition of intact applies to primary containment If the licensee does not intend to maintain an intact containment, then containment is open. If a PWR licensee is not maintaining an intact containment during POS 1E and POS 2E, then this observation could be risk significant under the Maintenance Rule and should be reported to a SRA. A de-inerted containment is one in which limits on the primary containment oxygen concentration as defined in Technical Specifications are no longer maintained. Note 2: The results assume that each shutdown scenario results in a LERF if the licensee fails to maintain an intact containment or the containment fails due to loss of hydrogen control in ice condenser and Mark III containments. In Phase 3 analysis, if a licensee can show that failures involving long term cooling can be eliminated from LERF because the licensee would have evacuated given successful short term cooling, then the color of the finding would be reduced. When using this table, there are no duration factors associated with findings at shutdown. The generic shutdown CDFs include the frequency and duration that POS 1 and PO2 are entered into per calendar year for both PWRs and BWRs. For BWRs, POS 1 is assumed to last four days; POS 2 is assumed to last two days. For PWRs, POS 1 is assumed to last two days; POS 2 is assumed to last six days. Should the duration of a Type B finding exist for less than eight hours, then the color finding is reduced by one order of magnitude. Note 3: As discussed in Regulatory Guide 1.174, releases that pass through the pool would be scrubbed and would not contribute to LERF. Rather than crediting the pool with completely eliminating LERF, a decontamination factor (DF) of 10 is assigned to pool scrubbing in the SDP. This DF results in the LERF-significant leak rate increased from 100% containment volume per day to 1000% containment volume per day Note 4: With the suppression pool unavailable, fission products will not be scrubbed and steam generated by decay heat is assumed to lead to gradual over-pressurization of containment and the need to vent prior to effective evacuation. Thus, the finding could be LERF significant even if leak rate is less than 100% containment volume per day. Note 5: For BWR Mark III containments and PWR ice condenser plants, the term compartments is used interchangeably with the term regions or zones, and relates to the likelihood that hydrogen concentrations could rise to levels that could challenge containment. For a particular finding, the intent is to determine if the igniter used as a rule-of-thumb. If it is not clear whether the igniter system would remain effective, the inspector should refer to the text in Section 5.2.3 (Mark III containments) or Section 7.2.4 (ice condenser containments) of NUREG-1765, and to consult the design-basis analysis associated with the igniter system. If it is still unclear, the inspector should contact the regional SRA and headquarters staff knowledgeable in this area.

Issue Date: 02/25/19 33 0609, App. H Table 7.5 BWRs With Minimal Shutdown Mitigation Capability Total Annualized CDF Head on: 3E-6 (per calendar year) Total Annualized CDF Head off: 9E-7 (per calendar year) Item Value RHR pumps 2 (shared with ECCS) Other heat removal pumps 0 ECCS pumps (in standby) 2 (Shared with RHR) SRVs for Power Operated Relief Mode 2 CCW pumps/trains 1 train with 2 pumps SW pumps/trains 1 train with 2 pumps Containment Spray pumps 0 Fire Water No SW Injection into RCS No Path to Suppression Pool Yes Suppression Pool Yes Other Water sources No Other means of removing heat None Offsite power sources 2 EDGs 1 other onsite power sources 0 Level instruments Yes Vessel Temperature Instruments No Level 3 RHR Isolation Sometimes Not Used Issue Date: 02/25/19 34 0609, App. H Table 7.6 BWRs With In-depth Shutdown Mitigation Capability Total Annualized CDF RCS Head on: 2E-7 (per calendar year) Total Annualized CDF RCS Head off: 4E-8 (per calendar year) Item Value RHR pumps 2 (Shared with ECCS Other heat removal pumps 0 ECCS pumps 2 (shared with RHR pumps) SRVs (in Power Operated Relief mode) 2 CCW pumps/trains 1 train with pumps SW pumps/trains 1 train with pumps Containment Spray Pumps 0 Fire Water Yes SW Injection into the RCS Yes Path to the Suppression Pools Yes Suppression Pool Yes Other water sources No Other means of removing heat None Offsite power sources 2 EDGs 2 Other onsite power sources 0 Level instruments Yes Vessel temperature Instruments Yes Level 3 RHR isolation Always Issue Date: 02/25/19 35 0609, App. H Table 7.7 PWRs With Minimal Shutdown Mitigation Capability Total Annualized CDF RCS open: 3E-5 (per calendar year) Total Annualized CDF RCS closed: 3E-6 (per calendar year) Item Value RHR pumps 2 Other heat removal pumps 0 ECCS pumps (in standby) 1 RCS vents and pressure control Yes CCW pumps/trains 2 trains SW pumps/trains 2 trains Containment Spray pumps (as back up to the RHR pumps) 0 Gravity Feed Yes Accumulators 0 Steam Generators Yes Containment sumps Yes, but not fully reliable other borated water sources 0 other means of removing heat 0 Offsite power sources 2 EDGs 1 other onsite power sources 0 Level instruments 2 some of time Vessel temperature Instruments 2 some of time Issue Date: 02/25/19 36 0609, App. H Table 7.8 PWRs With In-depth Shutdown Mitigation Capability Total Annualized CDF RCS open: 1E-7 (per calendar year) Total Annualized CDF RCS closed: 8E-7 (per calendar year) Item Value RHR pumps 2 Other heat removal pumps 0 Charging Pumps 1 ECCS pumps (in standby) 1 RCS vents and pressure control Yes CCW pumps/trains 2 trains SW pumps/trains 2 trains Containment Spray pumps 2 as piggy back to the RHR pumps Gravity Feed Yes Accumulators 0 Steam Generators Yes Containment sumps Yes, enhanced reliability other borated water sources 0 other means of removing heat 0 Offsite power sources 2 EDGs 2 other onsite power sources 0 Level instruments 2 at all times Vessel temperature Instruments 2 at all times Issue Date: 02/25/19 37 0609, App. H 0609H-08 REFERENCES 1. Regulatory Guide 1.174, Revision 3, An Approach for using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to the Licensing January, 2018. 2. PRAB02-01, . October 2002. 3. NRC, Memo from Barret tig December 7, 2001. 4. NUREG-1765, Significance Determination Process (SDP) Inspection Findings That May Affect LERF December, 2002. 5. NUREG-1150, ere Accident Risks: an Assessment for Five U. S. Nuclear Power Plants. December, 1990. 6. NUREG-1560, and Plant Performance. December, 1997. 7. NUREG/CR-6595, g the Frequencies of Various Containment F January, 1999. 8. NUREG/CR-5432, -I Containment. August, 1991. 9. NUREG/CR-6427, . April, 2000. 10. NUREG/CR-. June, 1986. 11. NUREG/CR-- Based Containment Leak-Test Program. September, 1995. 12. Policy Statement; Republication. August 1986. 13. NUREG-Engineering Plants with Thermally Treated Alloy 600 and 690 Steam Generator Tubes, Final Report. May 2018.

Issue Date: 02/25/19 Att1-1 0609, App. H Attachment 1: Guidance for Assessing the Timing of Protective Actions in Detailed Risk Evaluations SDP LERF treatment typically relies on a general, functional definition of LERF, rather than a more detailed accounting of accident progression timing against protective action timing. While this more general treatment is usually sufficient for an SDP, experience has shown that it is infrequently necessary to evaluate the timing of protective actions relative to radiological release on an accident sequence basis as part of a detailed risk evaluation. In these situations, the guidance below may be helpful and should be considered in tandem with other considerations specific to the SDP in question. This level of detail may not be warranted, particularly if the available information on core damage and containment failure timings is not well-characterized. 1. Early declaration of a General Emergency based on emergency director discretion may be credited on a probability basis, when warranted. For example, there may be circumstances where it is more realistic to assume earlier declaration for 50% of an accident sequence frequency, while assuming later declaration for the remainder, given 2. It should be assumed that emergency action level monitoring and protective action recommendations are made in a timely manner (e.g., declaration made within 15 minutes of relevant plant conditions and protective action recommendations made 15 minutes thereafter). 3. For external hazards well beyond the design basis (e.g., seismic bins in the upper end of the seismic hazard), some impact on response capabilities is possible but also beyond the state-of-the-practice to model. If these types of events are particularly important to a particular risk evaluation, a sensitivity study could be used to address this aspect. 4. Evacuation time estimate (ETE) studies have been performed, and are periodically updated for all sites. These studies are appropriate sources of information for use in SDP assessments. It is understood that these studies are developed for other evacuation timing, which meets the intent of SDP. They typically provide estimates of the time between the start of evacuation to the time the last of the individuals have cleared the 10-mile boundary for a range of conditions and assumptions. 5. In using ETE studies as part of the best available information for LERF determination in a detailed risk evaluation: a. ETE studies typically present timing for evacuations of emergency response planning areas (ERPAs). ERPAs are typically defined by compass sectors and distances of 2, 5, and 10 miles. For LERF, it is the 2 mile population that is most relevant. b. ETE studies typically present for the time to evacuate 90% and 100% of the population in particular ERPAs and combinations of ERPAs. The analyst should use the time estimates associated with evacuating 90% of the population, as this Issue Date: 02/25/19 Att1-2 0609, App. H represents a reasonable tradeoff between the inclusiveness of the 100% value, versus the fact that the timings represent the time to reach the 10 mile boundary (whereas dose levels will likely drop below those of concern for LERF prior to that distance). c. ETE studies present times for different scenarios (e.g., day time, night time, winter storms, and roadway impacts). When considering the spectrum of applicable scenarios. If the LERF determination is not sensitive to the range of time estimates of relevance, use the most inclusive time. d. ETE studies also present timings for different evacuation assumptions (i.e., keyhole evacuation, 5-mile 360 degree evacuations, and 10-mile 360 degree evacuations). A judgement should be made as to which of these is most applicable. 6. There are a number of aspects of the above assumptions that are uncertain, and there control. For this reason, it is appropriate to perform a sensitivity study that shows how Issue Date: 02/25/19 Att2-1 0609, App. H Attachment 2: Revision History for IMC 0609, Appendix H Commitment Tracking Number Accession Number Issue Date Change Notice Description of Change Description of Training Required and Completion Date Comment Resolution and Closed Feedback Form Accession Number (Pre-Decisional, Non-Public Information) n/a 04/21/2000 CN 00-007 Intial issuance. n/a n/a 05/06/04 CN 04-010 Periodic update n/a Issue Date: 02/25/19 Att2-2 0609, App. H Commitment Tracking Number Accession Number Issue Date Change Notice Description of Change Description of Training Required and Completion Date Comment Resolution and Closed Feedback Form Accession Number (Pre-Decisional, Non-Public Information) N/A ML18243A521 CN Changes included: Removed all references to obosolete Phase 2 system notebooks Added SRA to the list of people that can perform a Phase 3 analysis. Added a reminder that SAPHIRE can now be used to perform LERF calculations. Removed the step to multiply the LERF score by a factor of 3.3 from table 5.3 (now Table 6.3) since it is not correct to do so. Revised Figure 5.1 (now Table 6.1) so it no longer mentions SDP Phase 2 notebooks Created a new Attachment 1 to provide a Guidance for assessing the timing of protective actions in detailed risk evaluations Addressed feedback form 0609H-2225 Added a new section (Section 0609H-05) for C-SGTR and revised Table 6.1 to add a row for Combustion Engineering plants. As a result of adding section 0609H-05, subsequent sections and table numbers were changed. Table 7.1 has been revised to add a new column for ice condenser flow blockage. Added a note to Tables 7.2 and 7.4 to provide more information on failure of multiple igniters for Ice Condenser and BWR Mark III plants. Added a new Section 01.05 regarding the use of licensee-provided LERF information. Revised definition for shutdown operation to better align with IMC 0609, Appendix G. N/A ML18247A003 0609H-2225 ML18352A703