NRC Inspection Manual 0609/Appendix H
https://www.nrc.gov/docs/ML2007/ML20078L336.pdf
text
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.
Table 1.1 Risk Significance Based on ΔLERF vs ΔCDF
Frequency Range/ry SDP Based on ΔCDF SDP Based on ΔLERF
≥ 10-4 Red Red
< 10-4–10-5 Yellow Red
< 10-5–10-6 White Yellow
< 10-6–10-7 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 risksignificant 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
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.
Issue Date: 03/23/20 4 0609, App. H
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
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)
ROP Reactor Oversight Process
SCFH Standard Cubic Feet per Hour
SDP Significance Determination Process
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.
Table 4.1 Containment-Related SSCs Considered for LERF Implications1
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
Large lines connecting containment airspace to
environment (e.g., vent/purge) can contribute to
– connecting RCS to environment or open
systems outside containment
Small lines (< 1–2 inch diameter) and lines
connecting to closed systems would not
generally contribute to LERF
– connected to closed systems
inside/outside containment
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 I’s
PWR containment sprays and fan coolers Impact late containment failure and source
terms, but not LERF
1 Some of the listed SSCs could affect the core damage frequency as well as LERF.
Issue Date: 03/23/20 8 0609, App. H
Table 4.1 Containment-Related SSCs Considered for LERF Implications1
Hydrogen control system
– igniters Important to LERF in Mark III and ice condenser
plants
For AP1000, a significant loss of function of
hydrogen igniters should be assessed for LERF
impacts (e.g., diffusion flames, deflagration-todetonate transition) until more experience with
that containment type is gained
– air return fans and hydrogen mixing
systems
Not essential to hydrogen control if igniters are
available
Suppression pool (SP) systems
– components important to SP
integrity/scrubbing (e.g., vacuum
breakers)
Important to LERF in all BWR plants
– suppression pool cooling Impacts late containment failure but not LERF
Ice condenser system
– ice condenser doors and ice bed Significant flow blockage can be important to
– air return fans Not important to LERF (similar to containment
sprays)
– ice mass air return fans Deviations in weight of ice not important to
– foreign objects in ice compartment 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: 03/23/20 9 0609, App. H
Table 4.1 Containment-Related SSCs Considered for LERF Implications1
ADS system (AP1000) The capability to depressurize the RCS in a
high-pressure transient mitigates the
consequences of having high RCS pressure
during melt progression and vessel rupture.
Such accidents have a potential to fail the
steam generator tubes or to lead to energetic
phenomena at the time of vessel rupture that
can challenge containment.
Operation of ADS stage 4 provides a vent path
for the severe accident hydrogen to the steam
generator compartments, bypassing the IRWST,
and mitigating the conditions required to
produce a diffifusion flame near the containment
wall.
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 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
finding and its impact on the risk evaluation.
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
2 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.
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.
• 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.
• 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,
Issue Date: 03/23/20 12 0609, App. H
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 CSGTR (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.
Issue Date: 03/23/20 13 0609, App. H
Figure 6.1 Road Map for LERF-based Risk Significance Evaluation for Type A Findings at
Power
Use Table 6.2 to
Identify LERF
Factors
Use Table 6.2 to
Identify LERF
Factors Use Worksheet
Table 6.3
Use Worksheet
Table 6.3
Use Table 6.1
None of the
Sequences
Important to
LERF?
Use Table 6.1
None of the
Sequences
Important to
LERF?
Issue Date: 03/23/20 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 PWR’s use caution
not to overlook High-Dry-Low sequences which could result in a C-SGTR that would be
significant for LERF.
• ATWS sequences are not important contributors to LERF for BWRs with Mark III
containment. Containment failure from ATWS sequences occurs due to gradual overpressurization 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.
• 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 subatmospheric 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
Issue Date: 03/23/20 15 0609, App. H
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 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 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.
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: 03/23/20 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
(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
creen Out
(Note 4)
PWR Large Dry and
SubAtmospheric
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)
PWR AP1000 Perform
Phase 2
Perform
Phase 2
Screen Out
(Note 5)
Screen Out
(Note 5)
Screen Out
(Note 5)
Screen Out
(Note 5)
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.
AP1000 should be screened for C-SGTR only if conditions described in section 05.01 are met.
Issue Date: 03/23/20 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
(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 AP1000 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: 03/23/20 18 0609, App. H
Table 6.3 Manual Worksheet for ΔLERF
(1)
Sequences
(2)
ΔCDF Score
(X)
(3)
Sequence
Attributes
(4)
LERF Factor
(Table 6.2 for
power, Table 6.4
for shutdown)
(F)
(5)
ΔLERF
Score
F * (1x10-X
)
Total ΔLERF Score
Issue Date: 03/23/20 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 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.
Issue Date: 03/23/20 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 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.
Issue Date: 03/23/20 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 6.3
Worksheet
Use Table 6.4 for
LERF Factors
Use Table 6.4 for
LERF Factors
Issue Date: 03/23/20 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 8 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)
Screen Out – Screened Out if igniters are
available
(Note 3)
PWR Large Dry and
Sub-Atmospheric Intact Screen Out (Note 3) Screened Out
AP1000 Intact Screen Out (Note 3) Screened Out
Ice Condenser Intact
1.0 – if igniters are not available
(Note 2)
Screen Out – Screened 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.
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 analyst must verify that the licensee’s plans for containment closure
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 RHR. (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: 03/23/20 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 7.2 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.
Issue Date: 03/23/20 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
Table 7.1
Table 7.1
Use
Table 7.2
Use
Table 7.2
Use
Table 7.2
Use
Table 7.2
Issue Date: 03/23/20 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
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
Not
Applicable1
Perform
Phase 2
Perform
Phase 2
PWR Large Dry and
Sub-Atmospheric
Perform
Phase 2
Not
Applicable
Not
Applicable
No
Applicable
Not
Applicable
Not
Applicable
PWR AP1000 Perform
Phase 2
Not
Applicable
Not
Applicable
Not
Applicable
Not
Applicable2
Not
Applicable
PWR Ice Condenser Perform
Phase 2
Perform
Phase 2
Not
Applicable
Not
Applicable
Perform
Phase 2
Perform
Phase 2
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)
Note 2: AP1000 is being treated akin to a large dry containment in the absence of operating experience, however, a
particular performance deficiency relating to local hydrogen effects (e.g., the potential for a diffusion flame near the
containment wall) could warrant further investigation.
Issue Date: 03/23/20 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 AP1000 Leakage from containment to environment >100 %
containment volume/day through containment
penetration seals, isolation valves or vent and purge
systems
White Green Green
Significant loss of function to hydrogen igniters Note 2
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
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 regard, and “two adjacent compartments” is 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.
Note 2: For AP1000, a significant loss of function to the hydrogen igniters should be assessed for LERF impacts.
Issue Date: 03/23/20 27 0609, App. H
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.
Issue Date: 03/23/20 28 0609, App. H
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 indepth 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 indepth 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: 03/23/20 29 0609, App. H
Figure 7.2 Road Map for LERF-based Risk Significance Evaluation for Type B Findings at
Shutdown
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
Issue Date: 03/23/20 30 0609, App. H
Table 7.3 Phase 1 Screening-Type B Findings at Shutdown
Reactor/
Containment Type
Containment
Status
(Note 1)
SSC Affected by Finding
Containment
Penetration
Seals, Isolation
Valves, Vent
and Purge
Systems
Suppression
Pool Integrity
Drywell/
Containment
Sprays
Igniters
BWR Mark I and II De-inerted No Type B Findings Important to ΔLERF (Note 2)
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 AP1000 Intact
Perform Phase 2
contact HQ for
further
assistance
Not
Applicable
Not
Applicable (note 4)
PWR Ice Condenser Intact Perform
Phase 2
Not
Applicable
Screen Out
(Not important to LERF)
Perform
Phase 2
All Open No Type B Findings Important to ΔLERF (Note 3)
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 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
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: Type B findings would be unimportant to ΔLERF because containment would be de-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: Type B findings would be unimportant to Δ LERF because containment is already open and cannot be reclosed. 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.
Note 4: For AP1000, a significant loss of function should be assessed for LERF impacts.
Issue Date: 03/23/20 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
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 the staff concludes 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 POS 2 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)
Issue Date: 03/23/20 32 0609, App. H
of 10 is assigned to pool scrubbing in the SDP. This DF results in the LERF-significant leak rate increasing 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 system would remain
effective at controlling concentrations in this regard, and “two adjacent compartments” is 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 designbasis 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: 03/23/20 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
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: 03/23/20 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
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
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: 03/23/20 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
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: 03/23/20 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
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: 03/23/20 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
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 Att 1-1 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).
Issue Date: 03/23/20 Att 1-2 0609, App. H
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 2-1 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 (PreDecisional, NonPublic Information)
n/a
04/21/2000
CN 00-007
Intial issuance N/A
n/a ML041340009
05/06/04
CN 04-010
Periodic update N/A
N/A ML18243A521
02/25/19
CN 19-008
Changes included:
• Removed all references to Phase 2 obosolete 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.
• Provided new definitions for “close in population” and “effective
evacuation”.
• Created a new Attachment 1 which is 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.
N/A ML18247A003
0609H-2225
Issue Date: 03/23/20 Att 2-2 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 (PreDecisional, NonPublic Information)
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.
03/23/20
CN 20-017
IMC 0609, Appendix H has been modified to assess AP1000 reactor
design.
0609H2 – Limitations and Precautions:
• Alerted analysts that this is the first introduction of AP1000 into
Appendix H, and if they have 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.
• Revised the line item about ISLOCAs that the path outside
containment is assumed to be not submerged, and nor does it
benefit from other means of fission product retention. This
sentence was revised based on insights from SOARCA that
substantial ISLOCA fission product retention could result from
means other than break submergence.
• A CDF value was assigned for AP1000 of 1E-6/ry. This value will
be re-visited as operating and PRA modeling experience is
gained with the AP1000 design.
• Shutdown definitions have been removed from IMC 0609,
Appendix H, they are more appropriately located in IMC 0609,
App G, Shutdown Operations Significant Determination Process.
• Reference #14 has been added to the references.
N/A ML19352E278
Issue Date: 03/23/20 Att 2-3 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 (PreDecisional, NonPublic Information)
• Section 5.01 has been added regarding C-SGTR in AP1000
reactors, and section 6.01 has been revised for C-SGTR.
• Tables 6.1, 6.2, 6.4, 7.1, 7.2, 7.3 have been revised to
accommodate AP1000 reactor design.
• A couple of abbreviations for AP1000 have been added to
section 3.01.
• Table 4.1 has been modified to include information about
AP1000 hydrogen igniters and the ADS system. The information
about stage 4 ADS came from the Vogtle Units 3 and 4 FSAR
table 19.59-18.
• Table 6.4 note 1 – the requirement to notify NRR/SPSB for open
containments per SRM 97-168 was removed because the
requirement to do so has been replaced by a change to 10 CFR 50.65(a)(4) (64 FR 38557) and the issuance of RG 1.160 that
imposed a requirement to manage risk maintenance activities
and clarified that the requirement applied during shutdown
states.
• The wording was changed on note 2 of table 7.4 since the SDP
is the NRC staff process not the licensee.
• Table 7.1 for PWR Ice Condensers was changed to Perform a
Phase 2 for issues with Igniters or Drywell / Containment Sprays.
This was mistakenly changed to Not Applicable in the 2019
revision to Appendix H, and is now being corrected.