NRC Inspection Manual 0609/Appendix A

• https://www.nrc.gov/docs/ML2022/ML20226A093.pdf

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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 AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. 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knowledge transfer purposes, the pre-solved tables, riskinformed notebooks, and associated ROP guidance documents have been archived.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 HProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609</br></br>Appendix H" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., “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.

SAPHIRE and SPAR Models:

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 JProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix J" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. 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)

ML042600558

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

ML043560116

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

ML052790196

11/22/05

CN 05-030

Enhanced guidance to help meet timeliness

requirements for finalizing the SDP for inspection

findings.

N/A

ML063470288

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)

ML070720624

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)

ML063060377

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

ML12142A091

Closed FBF:

0609.04-1458

ML12171A225

0609A-1575

ML12171A231

N/A ML19198A183

7/17/19

CN

Made draft publicly available to discuss at the July

31, 2019 ROP monthly public meeting

N/A N/A

ML19011A338

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.

ML19014A063

Closed ROP FBFs

0609A-2070

ML19014A104

0609A-2134

ML19014A205

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

ML19014A106

0609A-2260

ML19014A107

0609A-2311

ML19014A108

0609A-2318

ML19014A109

0609A-1924

ML19253A002

0609A-2085

ML19253A003

0609A-2290

ML19253A004

0609A-2355

ML19253A005

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)

ML20308A592

11/03/2020

CN DRAFT

Made draft publicly available to discuss at the

November 18, 2020, ROP monthly public meeting.

N/A N/A

ML20226A093

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.

ML20226A181

Closed ROP FBFs

0609A-2407

ML20226A221

0609A-2408

Editorial - rejected

from the feedback

process upon receipt

0609-2179

ML20226A217

0308.03A-2178

ML20226A207