NRC Inspection Manual 0609/Appendix A
text
NRC INSPECTION MANUAL APOB
INSPECTION MANUAL CHAPTER 0609 APPENDIX A
THE SIGNIFICANCE DETERMINATION PROCESS
FOR FINDINGS AT-POWER
Effective Date: 01/01/2021
Issue Date: 11/30/20 i 0609 Appendix A
0609A-01 PURPOSE
The Significance Determination Process (SDP) described in this Appendix is designed to
provide staff and management with a simplified framework and associated guidance for use in
screening at-power findings. This Appendix aids the user in determining if a finding has a very
low safety significance (screens to Green) or directs the user to other applicable SDP
appendices or to perform a detailed risk evaluation.
This SDP is applicable to at-power findings within the Initiating Events, Mitigating Systems, and
Barrier Integrity Cornerstones. The SDP described in this Appendix is implemented by direction
from Inspection Manual Chapter (IMC) 0609, Attachment 4, “Initial Characterization of Findings.”
0609A-02 BACKGROUND
Over the years, maintaining the pre-solved tables and risk-informed notebooks from IMC 0609,
Appendix A proved to be a challenging task. As plants implemented equipment modifications
and associated revisions to the plant risk model, the accuracy of the pre-solved tables and riskinformed notebooks began to degrade. Instead of separately maintaining and updating the
plant-specific pre-solved tables and risk-informed notebooks, the agency decided to transition to
a software-based system called SAPHIRE (Systems Analysis Programs for Hands-on
Integrated Reliability Evaluations). Using SAPHIRE, a user can perform analyses on a regularly
maintained site-specific Standardized Plant Assessment Risk (SPAR) model. Updating sitespecific SPAR models provides an efficient and effective infrastructure that facilitates risk model
fidelity. For legacy, reference, and knowledge transfer purposes, the pre-solved tables, riskinformed notebooks, and associated ROP guidance documents have been archived.
In the transition from the pre-solved tables and risk-informed notebooks to SAPHIRE and the
site-specific SPAR models, it is important to note process differences. The pre-solved tables
and risk-informed notebooks, by process, provided a second layer of screening and an
estimation of the risk impact of the finding. In lieu of the pre-solved tables and risk-informed
notebooks, the SDP Workspace, a module within each SPAR model, was developed. The SDP
Workspace performs a delta CDF calculation similar in many respects to the risk estimate
performed by use of the risk-informed notebooks. However, use of SDP workspace is no longer
intended to provide a prescriptive additional layer of screening beyond that which is outlined in
Section 0609A-04, “Screening,” of this IMC. Rather, the SDP Workspace is one of many tools
the inspection staff and SRAs can utilize to support a detailed risk evaluation (see Section
0609A-05 of this IMC, “Detailed Risk Evaluation,” for more details).
0609A-03 GUIDANCE
This appendix is divided into two functional parts. The first part is a screening tool that uses a
series of logic questions to determine whether or not the finding can be characterized as having
very low safety significance (i.e., Green) and preclude a more detailed risk evaluation. The
second part provides guidance in determining the risk significance of a finding that did not
screen to Green in part one.
0609A-04 SCREENING
The screening questions are categorized by cornerstone. As such, there is one set of screening
questions for Initiating Events, one for Mitigating Systems, and one for Barrier Integrity (Exhibits
1, 2, and 3 respectively). If more than one cornerstone is affected, the screening questions in
all the affected cornerstones apply. In addition, under each cornerstone the screening
questions are categorized into sub-sections, so a finding and associated degraded condition
might be applicable to more than one sub-section. Typically, the inspection staff completes the
screening process with support from the regional SRAs, as needed. The screening questions
cover a wide range of instances and scenarios but are not intended to be all inclusive.
Therefore, if the inspection staff and/or SRA do not agree with the screening results, other risk
tools (e.g., the SDP Workspace) and guidance provided in Section 0609A-05, “Detailed Risk
Evaluation,” of this IMC can be used to confirm or challenge the screening results. The
screening process also directs the user to other applicable SDP appendices as needed (similar
to Table 3 of IMC 0609, Attachment 4).
The screening logic questions are designed to systematically determine whether a degraded
condition(s) resulting from a finding is of very low safety significance (i.e., Green) or not. If all
the logic questions under the applicable cornerstone(s) do not apply, then the finding is
screened as Green and the risk evaluation is complete (assuming that there are no unique
technical considerations that need to be assessed). Conversely, if any one of the logic
questions under a specific cornerstone is applicable to the degraded condition(s), the finding
cannot be screened as Green and further risk evaluation is warranted.
In applying the SDP screening questions, inspectors are evaluating the degraded condition in
the plant, for which the performance deficiency has been determined to be a proximate cause.
In defining the degraded plant condition, inspectors will need to use their judgment, in a
reasonable and realistic manner, consistent with previous similar findings. Inspectors are not
required to have proof of assumptions used in the SDP but must have a reasonable technical
basis. See IMC 0308, Attachment 3 for additional information on the basis of the SDP.
The duration of a plant degraded condition, i.e., the exposure time, is often an important
assumption in the SDP and is specifically used to assess the Mitigating Systems screening
questions. The exposure time is the duration or time period that the failed or degraded SSC is
reasonably known to have existed. The exposure time used in the SDP may not be equivalent
to that used for reportability or operability. Inspectors should consult with an SRA if there are
questions about determining the exposure time for a finding. The exposure time is often
evaluated against the duration of the Technical Specification (TS) allowed outage times, as
these periods are generally known to represent configurations of very low risk significance.
Also note that as a risk-informed tool, the at-power SDP is focused on initiating events,
mitigating system functions, and barrier integrity functions used in probabilistic risk assessments
(PRAs), which may differ from design basis transients and accidents as discussed in the
Updated Final Safety Analysis Report (UFSAR).
04.01 Initiating Events (Exhibit 1)
The Initiating Events screening questions are categorized into five sub-sections titled Loss of
Coolant Accident (LOCA) Initiators, Transient Initiators, Support System Initiators, Steam
Generator Tube Rupture (SGTR), and External Event Initiators. Below is additional guidance to
support answering the screening questions for each sub-section:
Issue Date: 11/30/20 3 0609 Appendix A
a. LOCA Initiators – Considers small, medium, and large LOCA initiating events. For SDP
purposes, a small LOCA is defined as a steam or liquid break in the reactor coolant
system (RCS), other than a SGTR, that exceeds the ability to makeup using normal
charging (PWR) or control rod drive (BWR) pump flow. Normal makeup flow may
include control room actions to start a standby pump or minimize letdown flow, if
appropriate for the situation.
b. Transient Initiators – A transient initiator is an event that results in a reactor trip or
scram. Some examples of transients are loss of main feedwater, loss of condenser
heat sink, and loss of offsite power (LOOP) events.
c. Support System Initiators – Support systems include SSCs needed to start, operate, or
control a front-line system, where the front-line system fulfills a critical safety function.
Support system initiating events are a subcategory of initiating events where the failure
not only causes a loss or challenge to a safety function, but also adversely affects one
or more systems needed to respond to shutdown of the reactor. These events not only
trigger sequences of events that challenge plant control and safety systems whose
failure could potentially lead to core damage or large early release, they also fail all or
part of those systems used for mitigation. Examples of support system initiators include
loss of service water, loss of vital AC/DC power buses, loss of cooling water and loss of
instrument air events. Site-specific support system initiators can be identified in the
Plant Risk Information eBook (PRIB). In the rare case that the degraded condition is
associated with a support system but does not increase the likelihood of a plant
transient or trip, then the finding should still be evaluated by considering its mitigation or
other PRA functions under the Mitigating Systems Cornerstone (Exhibit 2).
d. SGTR – Steam generator tube conditions that violate the structural integrity
performance criterion (typically three times the differential pressure across a tube
during normal full-power steady-state operation, 3ΔPNO) make the tube more
susceptible to failure during high pressure, dry steam generator core damage
sequences. Steam generator tube conditions that violate accident leakage limits may
not be able to meet 10 CFR 100 dose guidelines during design basis accidents.
e. External Event Initiators – In the Initiating Events Cornerstone, the external events of
interest are limited to fire and internal flooding. Other external events, in the context of
the Initiating Events Cornerstone, are not applicable because the licensee does not
have control over these events (e.g., tornado, hurricane). However, the licensee does
have control over the systems used to mitigate an external event and that is covered in
the Mitigating Systems Cornerstone (Exhibit 2).
04.02 Mitigating Systems (Exhibit 2)
The Mitigating Systems screening questions are categorized into five sub-sections titled
Mitigating Systems, Structures, Components (SSCs) and PRA Functionality (except Reactivity
Control Systems); External Event Mitigating Systems (Seismic/Flood/Severe Weather
Protection Degraded); Reactor Protection System; Fire Brigade; and Flexible Coping Strategies
(FLEX). Below is additional guidance to support answering the screening questions for each
sub-section:
a. Mitigating SSCs and PRA Functionality (except Reactivity Control Systems) – For the
purposes of this sub-section, the SSCs (and their associated functions) of concern are
Issue Date: 11/30/20 4 0609 Appendix A
those that provide a risk significant or risk relevant mitigating function in response to an
initiating event, i.e., the PRA function. Normally those SSCs that are in the risk model
provide a risk significant or risk relevant function; however, that is not always the case
(e.g., some SSCs are not modeled explicitly). There are several ways to determine
whether an SSC provides a risk significant or risk relevant mitigating function and below
are some sources of information to support this determination:
1) Plant Risk Information eBook (PRIB), Table 6 – Table 6 lists systems/functions
that are included in the SPAR model. It also provides specific success criteria
given a particular initiating event. See PRIB definition in Section 0609A-05 of
this IMC, “Detailed Risk Evaluation.”
2) PRIB, Table 7 – Table 7 lists the components included in the SPAR model with
their associated risk importance measures.
3) SDP Workspace – The SDP workspace contains risk significant and risk relevant
SSCs derived from the site-specific SPAR model.
4) UFSAR – Although the systems/functions described in the UFSAR might be
different than the systems/functions modeled in the SPAR, the licensed design
bases for systems/functions can provide useful information in determining safety
significance.
5) Licensee Risk Insights – If provided, risk insights from the licensee risk model
(e.g., importance measures, dominant sequences, delta CDF calculations, etc.)
and risk/safety significant SSCs from their maintenance rule program can be a
good source of risk information.
PRA function refers to the ways in which the SSC can be used in a PRA to prevent an
initiating event from resulting in core damage. An SSC may have more or different PRA
functions than those functions for which it is credited in the design or licensing basis.
For example, the design function of the core spray system may be limited to mitigation
of large loss of coolant accidents (LOCAs). As such, the accident analysis may define
a certain flowrate required to mitigate that accident. However, the core spray system
can be credited in a PRA to provide coolant injection in any scenarios in which coolant
injection is needed and pressure can be reduced such that the system can operate.
Thus, the PRA function of the core spray system is not limited to the mitigation of large
LOCAs and the system may be able to perform some of its other PRA functions without
meeting its design flowrate.
A key concept in assessing whether an SSC can perform its PRA function is mission
time. A 24-hour mission time is standard in PRA applications and should be considered
in SDP screening as a general rule. The 24-hour mission time used for the purposes of
SDP may be different than the time the SSC is required to operate as stated in the
accident analysis or design basis for the SSC. Inspectors should consult with an SRA
for unique situations or questions about mission time.
When the screening questions refer to a TS allowed outage time (AOT), the AOT is
being used to assess the impact of the exposure time during which the SSC could not
perform its PRA function. Although TS AOTs were not necessarily derived from risk
evaluations, operating experience has shown that an SSC that cannot function for less
than its AOT is generally not risk significant. Therefore, a detailed risk evaluation only
needs to be performed when the SSC could not function for a period of time greater
than that defined in the AOT. For single train systems or single trains within a multitrain system, the period of the AOT is used. For loss of function for two separate TS systems, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is used to determine if a detailed risk evaluation is warranted. For
risk-significant, non-TS SSCs, 3 days is used. For plants that have adopted TSTF-505
and implemented risk-informed completion times (RICTs), the frontstop AOT should be
used for screening purposes. RICTs may not be applied in retrospect after a degraded
condition occurs.
The screening question that refers to “loss of system and/or function” generally applies
to single train systems, or system/function as defined in the PRIB. A system/function is
modeled in the PRA but may not have a precise SSC definition. Examples include the
recovery of offsite power after a LOOP event, feed and bleed in a PWR plant after AFW
system failures, or various plant crosstie capabilities.
b. External Event Mitigating Systems (Seismic/Flood/Severe Weather Protection) – This
section is only applicable for findings related to seismic, flooding, and severe weather
protection. Findings related to fires or fire protection equipment are not included in this
question because those findings should be evaluated using IMC 0609 Appendix F, “Fire
Protection Significance Determination Process.”
c. Reactor Protection System (RPS) – The main focus of the screening question is to
screen findings that result in a minor functional degradation of RPS (e.g., one automatic trip from one instrument) but there are several redundant trips that provide the same
function (e.g., three other automatic functional trips). If there is a significant functional
degradation to RPS, a detailed risk evaluation is warranted. The determination of what
a “significant” or “minor” functional degradation of RPS should be based on reasonable
technical judgment of the inspectors, SRA, and management.
d. Fire Brigade – This section screens fire brigade findings to Green that are not expected
to result in additional fire growth. Fire brigade findings that are expected to result in
additional fire growth are evaluated further using IMC 0609 Appendix M, “Significance
Determination Process Using Qualitative Criteria.”
e. Flexible Coping Strategies (FLEX) - This screening section is intended for use in
assessing inspection findings that are associated with equipment, procedures, training,
and other programmatic aspects used specifically for satisfying the requirements of
Orders EA-12-049, EA-12-051, and EA-13-109 or for compliance with 10 CFR 50.155.
In the event that the equipment serves another function, a different and more limiting
SDP tool will be used. For example, if the performance deficiency concerns installed
plant equipment that is credited for Phase 1 mitigating strategies, but is also credited for
use under normal operating conditions or used to mitigate other transients or accidents
(e.g. reactor core isolation cooling pump, turbine-driven auxiliary feedwater pump), the
more limiting SDP (e.g., IMC 0609, Appendices A, G, and H) would be used to assess
the significance of the issue. This applies to all equipment, procedures, training, and
other programmatic aspects that are not credited for the sole purpose of satisfying the
requirements of Orders EA-12-049, EA-12-051, EA-13-109, or 10 CFR 50.155. This
section is used to screen findings related to all aspects of Phase 1 and Phase 2
mitigating strategies.
For the purpose of this SDP section, a FLEX function should be considered failed if the
strategy could not be implemented in accordance with existing plant procedures in the
time allotted. This could occur if multiple pieces of equipment fail or if one piece of
portable equipment was failed but its failure would not be discovered within enough
Issue Date: 11/30/20 6 0609 Appendix A
time for the licensee to install the backup piece of portable equipment. This could also
occur if the failure of a piece of equipment would result in additional equipment failures
that would require recovery actions or prevent completion of the strategy in accordance
with existing procedures within the time allotted.
04.03 Barrier Integrity (Exhibit 3)
The Barrier Integrity screening questions are categorized into five sub-sections titled Fuel
Cladding Integrity, Reactor Coolant System (RCS) Boundary, Reactor Containment, Control
Room/Auxiliary/Reactor Building or Spent Fuel Pool Building, and Spent Fuel Pool. Below is
additional guidance to support answering the screening questions for each sub-section:
a. Fuel Cladding Integrity – The purpose of this section is to screen findings to Green that
do not challenge fuel cladding integrity. For the purposes of this SDP, issues that meet
any of the following four criteria represent a challenge to fuel cladding integrity and
require further evaluation: (1) placed the plant in an unanalyzed condition, (2) adversely
impacted any fundamental assumptions regarding fuel failure used in the accident
analysis (such as fuel failure temperature or oxidation rate), (3) resulted in reactor
coolant activity exceeding TS limits, or (4) resulted in automatic actuation of an SSC
necessary to protect against exceeding thermal limits. If degradation of fuel cladding
could result in a substantial potential for overexposure, the finding should also be
evaluated using IMC 0609 Appendix C, “Occupational Radiation Safety Significance
Determination Process.”
b. Reactor Coolant System (RCS) Boundary – All issues which address potential violations
of regulatory requirements for protection of the reactor pressure vessel against fracture
(e.g., pressure-temperature limits, pressurized thermal shock (PTS)) are addressed
under the Barrier Integrity Cornerstone and should be reviewed by the applicable
technical group in NRR (NRR/DNRL/NVIB). Findings related to RPV fracture toughness
requirements must be evaluated in accordance with the ASME Code,Section XI,
Appendix E, “Evaluation of Unanticipated Operating Events” which provides
deterministic acceptance criteria for evaluating the impact of the out-of-limit condition on
the structural integrity of the RPV to determine whether the plant is acceptable for
continued operation. All other RCS boundary issues (i.e., leakage) are evaluated under
the Initiating Events Cornerstone (Exhibit 1).
c. Reactor Containment – The purpose of this section is to refer findings that primarily
impact large early release frequency (LERF) for further evaluation using IMC 0609
Appendix H, “Containment Integrity Significance Determination Process.”
d. Control Room/Auxiliary/Reactor Building or Spent Fuel Pool Building – Findings that
impact control room habitability require further evaluation. Findings related to the
radiological barrier functions of the control room, auxiliary building, reactor building, and
spent fuel pool building are not expected to impact CDF or LERF. If degradation of the
radiological barrier could result in a substantial potential for overexposure, the finding
should also be evaluated using IMC 0609 Appendix C.
e. Spent Fuel Pool – Findings that challenge spent fuel pool design criteria require further
evaluation. Further evaluation is performed using IMC 0609 Appendix M because the
NRC does not maintain PRA models for spent fuel pools.
0609A-05 DETAILED RISK EVALUATION
The inspection staff and regional SRAs should coordinate efforts, using their specific skills,
training, and qualifications, to arrive at an appropriate risk evaluation given the specific
circumstances associated with the risk impact of the degraded condition(s) that resulted from
the finding. Typically, inspectors develop the finding and the associated functional impact on
the equipment and gather plant information to support the detailed risk evaluation. Then the
inspectors and SRA collaborate to develop appropriate input assumptions while the SRA
normally performs the detailed risk evaluation using the SPAR model, the RASP handbooks,
and other risk information as necessary. If the finding does not screen to Green, the regional
branch chief responsible for the issue and the SRA shall determine if an Inspection Finding
Review Board is warranted using the guidance in IMC 0609 Attachment 5, “Inspection Finding
Review Board,” to ensure alignment on the performance deficiency, the inspection finding, any
proposed violation(s), and the actions needed to determine the preliminary significance,
All detailed risk evaluations should be peer reviewed by an SRA or Reliability and Risk Analyst.
A peer review is recommended but not required for straightforward detailed risk evaluations for
Green findings. A peer review is highly recommended for more complicated detailed risk
evaluations for Green findings in order to verify reasonable modeling assumptions have been
made. Peer reviews are required for any detailed risk evaluations performed for greater than
Green findings, as discussed in IMC 0609 Attachment 1, “Significance and Enforcement Review
Panel Process.” When the internal events detailed risk evaluation results are greater than or
equal to 1.0E-7, the finding should be evaluated for external event risk contribution. Any
internal events results that are less than 1.0E-7 can be evaluated for external event risk
contribution at the discretion of the regional SRA.1
If an inspector uses the SDP Workspace to
perform a detailed risk evaluation, a regional SRA must review the results to determine if any
additional analyses need to be performed.
If more than one cornerstone is affected by the finding and associated degraded condition(s),
the risk evaluation of the finding should take into account all of the associated degraded
condition(s) from all of the affected cornerstones. However, for the purposes of the power
reactor assessment program, the cornerstone which captures the majority fraction of the overall
risk evaluation should be identified as the affected cornerstone. The risk tools and guidance
available to the staff to perform the detailed risk evaluation are discussed below:
NOTE: The risk tools (e.g., SDP Workspace) and guidance to support the SDP are designed to
have users engaged in the process and avoid a “blackbox” approach in determining the risk
significance of deficient licensee performance. Users need to be aware of the limitations and
specific capabilities of each risk tool and associated guidance to preclude misapplication.
1) SDP Workspace – The SDP Workspace provides the user with a change in core
damage frequency (delta CDF), and change in large early release frequency (delta
LERF) calculation with a comprehensive report of results. This tool only accounts for
risk associated with internal events (i.e., does not account for external event risk
1 Until operating experience is gained for AP1000 plants, the finding should be evaluated for external
event risk contribution when the internal events detailed risk evaluation results are greater than or equal
to 1.0E-8.
Issue Date: 11/30/20 8 0609 Appendix A
contributions) and cannot be adjusted to change the model (e.g., recovery actions,
common cause failure).
2) Event Condition Assessment – A workspace that is used by the SRA that allows the
analyst more flexibility in adjusting basic events.
3) General Analysis – A workspace that is used by the SRA that allows more flexibility in
adjusting both basic events and model logic.
4) Specific SPAR Model Changes – The SRA can alter the SPAR model logic and create
a set of changed basic events to reflect the degraded condition(s) and/or event. This
approach provides the most flexibility in performing a delta CDF calculation.
5) Plant Risk Information eBook (PRIB) – The PRIB is a summary document associated
with the site-specific SPAR model that provides a variety of risk insights.
Changes to SAPHIRE and SPAR Models:
Identified Errors or Discrepancies – Identified errors or discrepancies with SAPHIRE or the
site-specific SPAR model should be discussed and vetted by the inspection staff and SRA
and then reported to Idaho National Laboratory (INL) via the SAPHIRE webpage at
https://saphire.inl.gov/. On the SAPHIRE webpage there is one module to request changes
to SAPHIRE (i.e., software) and a separate module to request changes to the SPAR models
(which includes changes to the PRIB).
Timely SDP Evaluations – To support the SDP timeliness goal, an SRA may make changes
to the SPAR model of record, as appropriate, based on information from the inspectors
and/or the licensee, to accurately reflect the risk significance of the finding. The SRA should
consult with INL on SPAR model changes. These changes must be documented in the
associated inspection report and/or SERP package. The SRA should subsequently review
the model changes made to determine if those model changes should be incorporated into
the plant SPAR model of record.
Guidance Documents:
1) RASP Handbook Volumes 1 (Internal Events), 2 (External Events), and 4 (Shutdown) -
These handbooks provide standardized risk guidance and best practices to support
determinations across a variety of NRC programs (SDP, Accident Sequence Precursor
(ASP), and Management Directive (MD) 8.3, “Event Evaluation”).
2) NUREGs – There are many NUREGs that can provide useful information when
performing a detailed risk evaluation (e.g., initiating event and failure data, common
cause failure modeling techniques).
END
Exhibit 1 - Initiating Events Screening Questions
A. Loss of Coolant Accident (LOCA) Initiators
1. After a reasonable assessment of degradation, could the finding result in exceeding the
reactor coolant system (RCS) leak rate for a small LOCA (leakage in excess of normal
makeup)?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, continue.
2. After a reasonable assessment of degradation, could the finding have likely affected
other systems used to mitigate a LOCA (e.g., Interfacing System LOCA)?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, screen as Green.
B. Transient Initiators
Did the finding cause a reactor trip AND the loss of mitigation equipment relied upon to
transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss of
condenser, loss of feedwater)? Other events include high-energy line breaks, internal
flooding, and fire.
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, screen as Green.
C. Support System Initiators
1. Did the degraded condition result in an actual complete or partial loss of a support
system (e.g., component cooling water, service water, instrument air, AC power, DC
power)?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, continue.
2. Did the degraded condition increase the likelihood of a complete loss of a support
system that would result in a plant trip?
Issue Date: 11/30/20 Ex1 - 2 0609 Appendix A
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, screen as Green.
D. Steam Generator Tube Rupture
1. Does the finding involve a degraded steam generator tube condition where one tube
cannot sustain three times the differential pressure across a tube during normal full
power, steady state operation (3ΔPNO)?
□ a. If YES ➛ Stop. Go to IMC 0609, Appendix J.
□ b. If NO, continue.
2. Do one or more SGs violate “accident leakage” performance criterion (i.e., involve
degradation that would exceed the accident leakage performance criterion under design
basis accident conditions)?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section and refer to IMC 0609,
Appendix J as applicable.
□ b. If NO, screen as Green.
E. External Event Initiators
Does the finding impact the frequency of a fire or internal flooding initiating event?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, screen as Green.
Issue Date: 11/30/20 Ex2 - 1 0609 Appendix A
Exhibit 2 – Mitigating Systems Screening Questions
A. Mitigating SSCs and PRA Functionality (except Reactivity Control Systems)
1. If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does
the SSC maintain its operability or PRA functionality?
□ a. If YES ➛ Screen as Green.
□ b. If NO, continue.
2. Does the degraded condition represent a loss of the PRA function of a single train TS
system (such as HPCI/HPCS) for greater than its TS allowed outage time?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, continue.
3. Does the degraded condition represent a loss of the PRA function of one train of a multitrain TS system for greater than its TS allowed outage time?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, continue.
4. Does the degraded condition represent a loss of the PRA function of two separate TS
systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, continue.
5. Does the degraded condition represent a loss of a PRA system and/or function as
defined in the PRIB or the licensee’s PRA (such as recovery of offsite power or the
ability to feed and bleed) for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, continue.
6. Does the degraded condition represent a loss of the PRA function of one or more nonTS trains of equipment designated as risk-significant in accordance with the licensee’s
maintenance rule program for greater than 3 days?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, screen as Green.
Issue Date: 11/30/20 Ex2 - 2 0609 Appendix A
B. External Event Mitigating Systems (Seismic/Flood/Severe Weather Protection)
Does the finding involve the loss or degradation of equipment or function specifically
designed to mitigate a seismic, flooding, or severe weather initiating event (e.g., seismic
snubbers, flooding barriers, tornado doors) for greater than 14 days?
□ a. If YES ➛ Go to Exhibit 4.
□ b. If NO, screen as Green.
C. Reactor Protection System (RPS)
Did the finding affect a single RPS trip signal to initiate a reactor scram AND the function of
other redundant trips or diverse methods of reactor shutdown (e.g., other automatic RPS
trips, alternate rod insertion, or manual reactor trip capacity)?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, screen as Green.
D. Fire Brigade
1. Does the finding involve fire brigade training, qualifications, drill performance, or
staffing?
□ a. If YES ➛ check if the following applies:
□ The finding would not have significantly affected the ability of the fire brigade to
respond to a fire.
□ b. If the above is checked ➛ screen as Green.
□ c. If NO, continue.
2. Does the finding involve the response time of the fire brigade to a fire?
□ a. If YES ➛ check if one or more of the following apply:
□ The fire brigade’s response time was mitigated by other defense-in-depth
elements, such as area combustible loading limits were not exceeded, installed fire
detection systems were functional, and alternate means of safe shutdown were not
impacted.
□ The finding involved risk-significant fire areas that had automatic suppression
systems.
□ The licensee had adequate fire protection compensatory actions in place.
□ b. If at least one of the above is checked ➛ screen as Green.
□ c. If NO, continue.
Issue Date: 11/30/20 Ex2 - 3 0609 Appendix A
3. Does the finding involve fire extinguishers, fire hoses, or fire hose stations?
□ a. If YES ➛ check if one or more of the following apply:
□ There was no degraded fire barrier and the fire scenario did not require the use
of water to extinguish the fire.
□ The missing fire extinguisher or fire hose was missing for a short time and other
extinguishers or hose stations were in the vicinity.
□ b. If at least one of the above is checked ➛ screen as Green.
□ c. If none of the boxes under D.1.a, D.2.a, or D.3.a are checked ➛ Stop. Go to IMC 0609, Appendix M.
E. Flexible Coping Strategies (FLEX)
1. Is the inspection finding associated with equipment, training, procedures, and/or other
programmatic aspects credited for the sole purpose of satisfying the requirements of
Order EA-12-051 or 10 CFR 50.155 for spent fuel pool instrumentation or EA-13-109 for
containment venting (i.e., not credited for satisfying EA-12-049 or other portions of
10 CFR 50.155 as well)?
□ a. If YES ➛ Screen as Green.
□ b. If NO, continue.
2. Does the inspection finding involve equipment, training, procedures, and/or other
programmatic aspects credited in any Phase 1 or 2 FLEX strategy such that any FLEX
function (such as extended HPCI/RCIC/AFW operation, providing FLEX DC power,
FLEX AC power, or FLEX RCS feed) could not be completed in accordance with existing
plant procedures within the time allotted for an exposure period of greater than 21 days?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, screen as Green.
Issue Date: 11/30/20 Ex3 - 1 0609 Appendix A
Exhibit 3 – Barrier Integrity Screening Questions
A. Fuel Cladding Integrity
1. Did the finding involve control manipulations that unintentionally added positive reactivity
that challenged fuel cladding integrity (e.g., inadvertent boron dilution, cold water
injection, two or more inadvertent control rod movements, recirculation pump speed
control)?
□ a. If YES, ➛ Stop. Go to IMC 0609, Appendix M.
□ b. If NO, continue.
2. Did the finding result in a mismanagement of reactivity by operator(s) that challenged
fuel cladding integrity (e.g., reactor power exceeding the licensed power limit, inability to
anticipate and control changes in reactivity during crew operations)?
□ a. If YES, ➛ Stop. Go to IMC 0609, Appendix M.
□ b. If NO, continue.
3. Did the finding result in the mismanagement of the foreign material exclusion or reactor
coolant chemistry control program that challenged fuel cladding integrity (e.g., loose
parts, material controls)?
□ a. If YES, ➛ Stop. Go to IMC 0609, Appendix M.
□ b. If NO, continue.
4. Did the finding result from fuel handling errors, a dropped fuel assembly, a misplaced
fuel bundle, or crane operations over the core or anywhere in the refueling pathway that
challenged fuel cladding integrity or resulted in a release of radionuclides?
□ a. If YES, ➛ Stop. Go to IMC 0609, Appendix M.
□ b. If NO, screen as Green.
B. Reactor Coolant System (RCS) Boundary
Does the finding involve potential non-compliance with regulatory requirements for
protection of the reactor pressure vessel against fracture (e.g., pressure-temperature limits
or pressurized thermal shock issues)?
□ a. If YES ➛ Stop. Go to IMC 0609, Appendix M and consult the appropriate technical
branch in NRR (NRR/DNRL/NVIB).
□ b. If NO, screen as Green.
Issue Date: 11/30/20 Ex3 - 2 0609 Appendix A
C. Reactor Containment:
1. Does the finding represent an actual open pathway in the physical integrity of reactor
containment (valves, airlocks, etc.), failure of containment isolation system (logic and
instrumentation), failure of containment pressure control equipment (including SSCs
credited for compliance with Order EA-13-109), failure of containment heat removal
components, or failure of the plant’s severe accident mitigation features (AP1000)?
□ a. If YES ➛ Stop. Go to IMC 0609, Appendix H.
□ b. If NO, continue.
2. Does the finding involve an actual reduction in function of hydrogen igniters in the
reactor containment?
□ a. If YES ➛ Stop. Go to IMC 0609, Appendix H.
□ b. If NO, screen as Green.
D. Control Room, Auxiliary, Reactor, or Spent Fuel Pool Building:
1. Does the finding only represent a degradation of the radiological barrier function
provided for the control room, auxiliary building, spent fuel pool, SBGT system (BWR), or
EGTS system (PWR ice condenser)?
□ a. If YES ➛ Stop. Screen as Green.
□ b. If NO, continue.
2. Does the finding represent a degradation of the barrier function of the control room
against smoke or a toxic atmosphere?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, screen as Green.
E. Spent Fuel Pool (SFP)
1. Does the finding adversely affect decay heat removal capabilities from the spent fuel
pool causing the pool temperature to exceed the maximum analyzed temperature limit
specified in the site-specific licensing basis?
□ a. If YES ➛ Stop. Go to IMC 0609, Appendix M.
□ b. If NO, continue.
Issue Date: 11/30/20 Ex3 - 3 0609 Appendix A
2. Does the finding result from fuel handling errors, dropped fuel assembly, dropped
storage cask, or crane operations over the SFP that caused mechanical damage to fuel
clad AND a detectible release of radionuclides?
□ a. If YES ➛ Stop. Go to IMC 0609, Appendix M (refer to IMC 0609, Appendix C as
applicable).
□ b. If NO, continue.
3. Does the finding result in a loss of spent fuel pool water inventory decreasing below the
minimum analyzed level limit specified in the site-specific licensing basis?
□ a. If YES ➛ Stop. Go to IMC 0609, Appendix M.
□ b. If NO, continue.
4. Does the finding affect the SFP neutron absorber, fuel bundle misplacement (i.e., fuel
loading pattern error) or soluble Boron concentration (PWRs only)?
□ a. If YES ➛ Stop. Go to IMC 0609, Appendix M.
□ b. If NO, screen as Green.
Issue Date: 11/30/20 Ex4 - 1 0609 Appendix A
Exhibit 4 – External Events Screening Questions
1. If the equipment or safety function is failed or unavailable, are ANY of the following three
statements TRUE? The loss of this equipment or function by itself during the external
initiating event it was intended to mitigate:
▪ would cause a plant trip or an initiating event;
▪ would degrade two or more trains of a multi-train system or function;
▪ would degrade one or more trains of a system that supports a risk significant system
or function.
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, Continue.
2. Does the finding involve the total loss of any PRA function, identified by the licensee through
a PRA, IPEEE, or similar analysis, that contributes to external event initiated core damage
accident sequences (i.e., initiated by a seismic, flooding, or severe weather event)?
□ a. If YES ➛ Stop. Go to Detailed Risk Evaluation section.
□ b. If NO, screen as Green.
Issue Date: 11/30/20 Att1-1 0609 Appendix A
ATTACHMENT 1
Revision History for IMC 0609 Appendix A
Commitment
Tracking
Number
Accession
Number
Issue Date
Change Notice
Description of Change
Description of
Training Required
and Completion
Date
Comment Resolution
and Closed
Feedback Form
Accession Number
(Pre-Decisional, NonPublic Information)
04/21/00
CN 00-007
Initial issue
12/28/00
CN 00-029
Revised to incorporate changes based on inspector
feedback. Enhancements generated by IIPB and
SPSB risk analysts based on initial implementation
experience to date have also been added. A
significant change is the credit given for operator
actions in step 2.3 of the document. Clarification
changes have also been made to the phase 1
screening worksheets. Phase 2 worksheets are in
the process of being updated to include plant and
site-specific information. This document is an
integral part of the Significant Determination Process
for reactor inspection findings for At-Power
operations and will be used by resident and regionbased inspectors as well as by SRAs.
02/05/01
CN 01-003
Revised to correct formatting problems with charts
and tables, and to make minor editorial changes.
03/18/02
CN 02-009
Revised: 1) to correct identified problems with the
appendix, 2) to incorporate the rules for using the
site specific risk-informed inspection notebook, 3) to
simplify the process of accounting for external
initiators in characterizing the risk significant
inspection findings, and 4) to provide guidance on
evaluating concurrent inspection findings.
Issue Date: 11/30/20 Att1-2 0609 Appendix A
Commitment
Tracking
Number
Accession
Number
Issue Date
Change Notice
Description of Change
Description of
Training Required
and Completion
Date
Comment Resolution
and Closed
Feedback Form
Accession Number
(Pre-Decisional, NonPublic Information)
09/10/04
CN 04-023
Multiple editorial changes to enhance user
friendliness of the document. For example, re-format
action steps, provided additional examples, added
the reference to Appendix J for steam generator
issues.
N/A
12/01/04
CN 04-027
Corrected two errors on page 4 of the worksheet,
under MS cornerstone for screening issues and
under BI cornerstone guidance for question 3 for
screening to Green.
N/A
11/22/05
CN 05-030
Enhanced guidance to help meet timeliness
requirements for finalizing the SDP for inspection
findings.
N/A
03/23/07
CN 07-011
Incorporate references to the site-specific inspection
notebooks and associated Pre-Solved Tables; In
Attachment 2, update the site-specific risk-informed
inspection notebooks usage rules; Attachment 3,
provide user guidance for screening of external
events risk contributions.
1. Training has been
provided to the
SRAs at last two
SRA counterpart
meetings, and the
SRAs have provided
training to the region
based and resident
inspectors (10/2006)
2. Formalized
training will be
introduced through
the P-111 course
(FY 2008)
Issue Date: 11/30/20 Att1-3 0609 Appendix A
Commitment
Tracking
Number
Accession
Number
Issue Date
Change Notice
Description of Change
Description of
Training Required
and Completion
Date
Comment Resolution
and Closed
Feedback Form
Accession Number
(Pre-Decisional, NonPublic Information)
01/10/08
CN 08-002
Removed the Phase 1 Initial Screening and
Characterization of Findings process to create the
new IMC 0609, Attachment 4. Added clarification
statement to Step 2.1.2 and Usage Rule 1.1 that the
maximum exposure time used in SDP is limited to
one year.
N/A ML073460588
ML101400574
06/19/12
CN 12-010
Updated the guidance to reflect the transition from
the pre-solved tables and risk-informed notebooks to
SAPHIRE and the site-specific SPAR models.
Moved the Initiating Events, Mitigating Systems, and
Barrier Integrity screening questions from IMC 0609,
Attachment 4 to this appendix. Incorporated
feedback from ROP FBFs 0609.04-1458 and 0609A1575. This is a complete reissue.
Senior Reactor
Analysts and
headquarters staff
provided detailed
instructor-led
training to resident
inspectors, region
based inspectors,
and other regional
staff.
June 2012
Closed FBF:
0609.04-1458
0609A-1575
N/A ML19198A183
7/17/19
Made draft publicly available to discuss at the July
31, 2019 ROP monthly public meeting
N/A N/A
12/13/19
CN 19-040
Updated guidance to direct users to contact NRR for
issues with pressure-temperature limits (ROP FBF
0609A-2070), moved some of the reactivity control
questions to the Barrier Integrity Cornerstone exhibit
to align with IMC 0612 (ROP FBF 0609A-2134),
revised the fire brigade and support system initiator
questions for clarity (ROP FBFs 0609A-2167 and
No required training
on specific changes
to this revision.
Closed ROP FBFs
0609A-2070
0609A-2134
Issue Date: 11/30/20 Att1-4 0609 Appendix A
Commitment
Tracking
Number
Accession
Number
Issue Date
Change Notice
Description of Change
Description of
Training Required
and Completion
Date
Comment Resolution
and Closed
Feedback Form
Accession Number
(Pre-Decisional, NonPublic Information)
2311), added a question regarding fuel cladding
integrity, separated and revised the mitigating
systems questions to account for single train systems
and PRA functions (ROP FBFs 0609A-2260 and
2318), and incorporated FLEX questions from IMC 0609 Appendix O. Questions were modified to
screen FLEX findings that are solely related to EA13-109 and containment pressure control systems to
Appendix H (ROP FBF 0609A-2355). IMC 0609,
Attachment 4 was reviewed to align with this
appendix regarding support system initiators and
spent fuel pool applicability (ROP FBF 0609A-2290
and 2085). Inspector training related to use of riskinformed thinking and tools is planned (ROP FBF
0609A-1924). Document was reviewed and minor
changes were made to allow for use with new reactor
designs (AP1000).
In accordance with Management Directive 8.13 and
COMSECY-16-0022, the Commission was notified of
the described changes via SECY-19-0037, “Reactor
Oversight Process Self-Assessment for Calendar
Year 2018,” (ADAMS Accession No. ML19042A100).
The Commission was also notified of the revisions in
a Commissioner Assistants’ Note (ADAMS
Accession No. ML19302F254).
General training is
planned as part of
the Regional Risk
Informed DecisionMaking action plan.
0609A-2167
0609A-2260
0609A-2311
0609A-2318
0609A-1924
0609A-2085
0609A-2290
0609A-2355
Issue Date: 11/30/20 Att1-5 0609 Appendix A
Commitment
Tracking
Number
Accession
Number
Issue Date
Change Notice
Description of Change
Description of
Training Required
and Completion
Date
Comment Resolution
and Closed
Feedback Form
Accession Number
(Pre-Decisional, NonPublic Information)
11/03/2020
CN DRAFT
Made draft publicly available to discuss at the
November 18, 2020, ROP monthly public meeting.
N/A N/A
11/30/20
CN 20-066
Combined and revised the FLEX screening questions
to clarify when a detailed risk evaluation should be
performed (ROP FBF 0609A-2407). Revised the
FLEX screening questions and background
information to incorporate issuance of
10 CFR 50.155 and to move information to the basis
document, IMC 0308 Att 3 App A. Updated correct
branch to contact for findings related to protection
again reactor pressure vessel fracture (Exhibit 2,
Question B) after NRR/NRO merger (ROP FBF
0609A-2408). Changed the method of further
evaluation for these findings from a detailed risk
evaluation to IMC 0609 Appendix M based on SRA
feedback. Added guidance to Section 0609A-05 to
recommend peer reviews for all detailed risk
evaluations (ROP FBFs 0308.03A-2178, 0609-2179)
and to refer to IMC 0609 Att 5. Added a new
question to Exhibit 3 for fuel handling errors to align
with a change made to IMC 0609 Att 4 that routes
those types of findings to App A. Revised the LOCA
initiator screening question (Exhibit 1, Question A.2)
based on SRA feedback. Added screening guidance
consistent with revisions to the basis document for
those areas that had no additional guidance in
Section 0609A-04.
No training required
on specific changes
in this revision.
Closed ROP FBFs
0609A-2407
0609A-2408
Editorial - rejected
from the feedback
process upon receipt
0609-2179
0308.03A-2178