**NRC INSPECTION MANUAL**
APOB

INSPECTION MANUAL CHAPTER 0609 APPENDIX H

CONTAINMENT INTEGRITY

SIGNIFICANCE DETERMINATION PROCESS

Effective Date:  04/30/2020

0609, App F		Issue Date: 02/28/05

TABLE OF CONTENTS0609H-01	Introduction	101.01	Applicability	101.02	Entry Conditions	201.03	Appendix H Outline	201.04	Use of SAPHIRE Software to Calculate LERF	201.05	Use of Licensee Models for LERF	20609H-02	Limitations and Precautions	30609H-03	Abbreviations and Definitions	403.01	Abbreviations	403.02	Definitions	50609H-04	Overview of the Approach and Procedure for Significance Determination	604.01	Types of Findings	604.02	LERF Based Significance Determination Process	60609H-05	Consequential Steam Generator Tube Rupture (C-SGTR)	1005.01	Evaluation of C-SGTR in AP1000 Reactors	100609H-06	Procedure for Type A Findings	1106.01	Approach for Assessing Type A Findings at Power	1106.02	Approach for Assessing Type A Findings During Shutdown	190609H-07	Procedure for Type B Findings	2307.01	Approach for Assessing Type B Findings at Power	2307.02	Approach for Assessing Type B Findings at Shutdown	270609H-08	References	37

List of Tables

Table 1.1  Risk Significance Based on ΔLERF vs ΔCDF	1

Table 4.1 ****Containment-Related SSCs Considered for LERF Implications	7

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 Worksheet for ΔLERF	18

Table 6.4  Phase 2 Assessment Factors -Type A Findings at Shutdown	22

Table 7.1  Phase 1 Screening-Type B Findings at Power	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  LER-based Significance Determination Process	9

Figure 6.1  Road Map for LERF-based Risk Significance Evaluation for Type A Findings at Power	13

Figure 6.2  Road Map for LERF-based Risk Significance Evaluation for Type A Findings at Shutdown	21

Figure 7.1  Road Map for LERF-based Risk Significance for Evaluation Type B Findings at Power	24

Figure 7.2  Road Map for LERF-based Risk 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...............................................................................Att 1-1

Attachment 2	Revision History for IMC 0609, Appendix H....................................Att 2-1

Issue Date:  03/23/20	5	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.

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 very low risk significance, given  the findings are entered into the licensee’s corrective action program.

## 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.

# 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:

· This revision incorporates the AP1000 reactor design into Appendix H. Since this is a new reactor design that hasn’t been previously assessed by the SDP, if an analyst has a basis for why this procedure is not adequately capturing the risk, they may depart from this procedure and perform a Phase 3 detailed risk evaluation.

· 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 there is greater variability regarding when evacuation would begin.

· LERF determinations depend on the containment design, 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, nor does it benefit from other means of fission product retention (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 were10-4/ry for PWRs,10-5/ry for BWRs, and 10-6/ry for AP1000 plants.

· 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.

# 0609H-03	ABBREVIATIONS AND DEFINITIONS

## 03.01	Abbreviations

ADS	Automatic Depressurization System (AP1000)

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

PRHR	Passive Residual Heat Removal System (AP1000)

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

## 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 Commission’s Safety Goal Policy if there are no individuals residing within a mile of the plant boundary, an individual for evaluation purposes, 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.

Definitions related to shutdown plant conditions can be found in IMC 0609, Appendix G, Shutdown Operations Significance Determination Process.

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 licensee’s corrective action program.

· 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: Also known as a detailed risk evaluation, this is a 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). 

# 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 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 to shutdown operations to determine their contributions to ΔCDF.

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 based on ΔCDF.  If the total ΔCDF for the finding is less than 1E-7 per reactor year, then the finding should be assigned a Green significance level.

If the total ΔCDF ≥ 1E-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 increase in risk and the risk significance based on ΔCDF applies.  If one or more of the affected sequences is identified as a LERF contributor, an assessment is performed to estimate ΔLERF and determine the increase in risk significance based on LERF considerations as discussed in detail in Section 05.

Type B Findings:

Type B findings have no impact on the determination of ΔCDF and therefore will not have been processed through the CDF based SDP.  These findings, however, are potentially important to ΔLERF contribution and have to be allocated an appropriate risk category based 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.

Figure 4.1  LER-based Significance Determination Process

# 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 CE plants than Westinghouse plants[footnoteRef:3]. 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 finding and its impact on the risk evaluation. [3:  Two of the primary factors driving this difference 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-bottom 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.] 

Therefore, findings 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.

## 05.01	Evaluation of C-SGTR in AP1000 Reactors

For AP1000 reactors, conditions that may significantly affect the conditional probability of having a consequential (a.k.a., severe accident-induced) steam generator tube rupture should not be screened out. Generally, such conditions would involve an increase in the likelihood of accident sequences associated with the onset of core damage at high pressure, coincident with one or more steam generators having boiled dry. Such instances may include station blackout or transients with failure to depressurize the RCS (e.g., due to ADS and PRHR failures). For accident sequences when core damage occurs with high RCS pressure, a dry SG, and low secondary side pressure, it is likely that full-loop natural circulation conditions will develop, leading to creep damage to both the RCS piping (hot leg and surge line nozzles) and steam generator tubes. The order and timing of failure of these components dictates whether LERF is a concern. These accident sequences could have a greater contribution to LERF, similar to the other containment bypass events that have been screened in (e.g., ISLOCA). AP1000 is not subject to loop seal blockage conditions that can tend to mitigate the threat to SG tubes for other Westinghouse designs, though it is estimated to be less likely to incur such accident sequences to begin with. Additional experience with C-SGTR modeling for AP1000 design is necessary before these findings can be more efficiently screened.

# 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 determination of the finding significance to ΔCDF based on all contributing 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 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 a reminder, SAPHIRE can also be used to calculate LERF of Type A findings, and is the preferred method.

STEP 1 – Finding Characterization

Determine the total ΔCDF of the finding and identify the associated CDF sequences which may be 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, as well as AP1000, 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 in an exsting PWR reactor, is a station blackout with a loss of auxiliary feedwater.  A typical example of such an accident scenario for an AP-1000 reactor would be a station blackout or transients with failure to depressurize the RCS (e.g., due to ADS and PRHR failures).  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 contibutors to LERF.  A C-SGTR is more of a concern for Combustion Engineering (CE) plants.  Consult the Risk Assessment for Operational Events (RASP) Manual Volume 5 for more information.

· For the PWR plants with ice condenser containments, findings related to ISLOCA, SGTR, and SBO accident categories.

Use Table 6.1

None of the Sequences

Important to 

LERF?

Use Table 6.1

None of the Sequences

Important to 

LERF?

Use Worksheet Table 6.3

Use Worksheet Table 6.3

Use Table 6.2 to Identify LERF Factors

Use Table 6.2 to Identify LERF Factors

Figure 6.1  Road Map for LERF-based Risk Significance Evaluation for Type A Findings at Power

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 PWR’s 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 containment.  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

If the total ΔCDF (i.e., sum of all sequences) is <1E-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

Compare the attributes of all core damage sequences with a ΔCDF of ≥1E-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, those LERF sequences that are ≥1E-8 (sequence result of 8 or less) are evaluated for the overall LERF contribution.  If none of the sequences impacts LERF, the risk significance obtained from the ΔCDF assessment is used for the significance of the finding and no further LERF-related evaluation is necessary.  If ΔCDF sequences are identified as having the potential to affect LERF[footnoteRef:4], proceed to Step 3. [4: No 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.] 

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 by multiplying the ΔCDF score (column 2) by the LERF factor (column 4).  For example, if a sequence has a ΔCDF score of 7 (i.e., 1E-7) and the associated LERF factors is 0.4, the LERF score is 4×10-8.

Step 4 – LERF Significance

Sum the scores for all of the LERF contributing sequences associated with the finding and enter the total ΔLERF score in the space below Column 5.  Use the numerical result to determine the ΔLERF significance (color), using Table 1.1.

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.  

****

## 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 0609, 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.

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 ΔCDF score (column 2) by the LERF factor (column 4).  For example, if a sequence has a ΔCDF score of 7 and the associated LERF factor is 0.2, the LERF score is 2×10-8.

STEP 4 – LERF Significance

Sum the scores for all of the LERF contributing sequences associated with the finding being evaluated and enter the total ΔLERF score in the space below column 5 of the completed Table 6.3.  Use the numerical result to determine the ΔLERF significance (color), using Table 1.1.

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.

Use Table 6.4 for

LERF Factors

Use Table 6.4 for

LERF Factors

Use Table 6.3

Worksheet

Use Table 6.3

Worksheet

Figure 6.2  Road Map for LERF-based Risk Significance Evaluation for Type A Findings at Shutdown

# 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 7.2 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.

Use

Table 7.2

Use

Table 7.2

Use

Table 7.2

Use

Table 7.2

Table 7.1

Table 7.1

Table 7.1

Table 7.1

Figure 7.1  Road Map for LERF-based Risk Significance for Evaluation Type B Findings at Power

## 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 consistent with NRC expectations (GL 88-17) and Industry expectations (NUMARC 91-06)?  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?

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.

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

Use

Table 7.4

Use

Table 7.4

Use

Table 7.3

Use

Table 7.3

Figure 7.2  Road Map for LERF-based Risk Significance Evaluation for Type B Findings at Shutdown

# 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 Basis.” January, 2018.

2. PRAB–02-01 “Assessment of BWR Main Steam Line Release Consequences.” ML062920249. October 2002.

3. NRC, Memo from Barret to Haag, SPSB significance Determination Process, December 7, 2001.

4. NUREG-1765 “Basis Document for Large Early Release Frequency (LERF) Significance Determination Process (SDP)” Inspection Findings That May Affect LERF December, 2002.

5. NUREG-1150 “Severe Accident Risks: an Assessment for Five U. S. Nuclear Power Plants” December, 1990.

6. NUREG-1560 “Individual Plants Examination Program: Perspectives on Reactor Safety and Plant Performance” December, 1997.

7. NUREG/CR-6595 “An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events”.  January, 1999.

8. NUREG/CR-5432 “The Probability of Liner Failure in Mark-I Containment” August, 1991.

9. NUREG/CR-6427 “Assessment of the DCH Issue for Plants with Ice Condenser” April, 2000.

10. NUREG/CR-4330 “Review of Light Water Reactor Regulatory Requirements” June, 1986.

11. NUREG/CR-1493 “Performance- Based Containment Leak-Test Program” September, 1995.

12. 51 FR 28044 “Safety Goals for the Operations of Nuclear Power Plants; Policy Statement; Republication” August 1986.

13. NUREG-2195 “Consequential SGTR Analysis for Westinghouse and Combustion Engineering Plants with Thermally Treated Alloy 600 and 690 Steam Generator Tubes, Final Report” May 2018.

14. Westinghouse AP1000 Design Control Document Rev. 19 – Tier 2 Chapter 19 – Probabilistic Risk Assessment – Sections 19.59 PRA Results and Insights. ML11171A411. June 2011.

15. Vogtle, Units 3 and 4, Updated Final Safety Analysis Report, Revision 2, Chapter 19, Probabilistic Risk Assesment. ML13206A152. June 2013.

Issue Date:  03/23/20	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 progressing 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, when warranted, may be credited on a probability basis. Example: EALs provide for SRO judgement in some circumstances. 50% probability that SRO declares event early given the plant damage state.

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 purposes, but they often represent the “best available information” with respect to 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 timings 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 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 are aspects of the protective action implementation that are outside of the licensee’s control. For this reason, it is appropriate to perform a sensitivity study that shows how ∆LERF would differ if a more optimistic or pessimistic set of assumptions are employed.

Issue Date:  03/23/20	Att 1-1	0609, App. H

Attachment 2: Revision History for IMC 0609, Appendix H

Issue Date:  03/23/20	Att 2-1	0609, App. H

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