NRC Inspection Manual 0609/Appendix G
Shutdown Operations Significance Determination Process - https://www.nrc.gov/docs/ML1910/ML19101A289.pdf
- Att 1 Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings - https://www.nrc.gov/docs/ML1909/ML19094B822.pdf
- Att 2 Phase 2 Significance Determination Process Template for PWR During Shutdown - https://www.nrc.gov/docs/ML1910/ML19102A195.pdf
- Att 3 Phase 2 Significance Determination Process Template for BWR During Shutdown - https://www.nrc.gov/docs/ML1910/ML19102A206.pdf
0609G-01 PURPOSE
Appendix G provides guidance for Phases 1 and 2 of the Significance Determination Process
(SDP) for shutdown operations. The SDP Phase 1 guidance, presented as Attachment 1 to
Appendix G, is a separate document in the manual chapter and consists of screening criteria
developed specifically for shutdown operations. When inspectors identify findings that affect
shutdown conditions, they will use the Phase 1 guidance for both pressurized water reactors
(PWRs) and boiling water reactors (BWRs). Attachment 1 lists the criteria for types of findings
that would require further evaluation by Phase 2 or 3 of the SDP. After the inspector determines
that a Phase 2 or 3 analysis is needed, he or she will transition the evaluation to a Senior
Reactor Analyst (SRA) for further review. The guidance for Phase 2 is presented in Appendix G
as Attachments 2 and 3. Attachment 2 provides Phase 2 guidance for PWRs and Attachment 3
provides Phase 2 guidance for BWRs. These attachments are generic in nature for both reactor
types. Phase 3, detailed risk evaluation, analyses are referred directly to Headquarters staff to
be performed. The road map for shutdown inspection findings through SDP Phases 1 and 2 is
shown in Figure 1.
0609G-02 OBJECTIVES
Shutdown operations in both PWRs and BWRs introduce a different spectrum of vulnerabilities
that may not be applicable during at-power operations. A shutdown plant is in a safe condition
as long as certain key safety functions are maintained and managed adequately. Those
functions are:
• inventory control
• power availability
• reactivity control, and
• containment closure capability
Analysis of shutdown events has provided a better understanding of these vulnerabilities and
has informed the development of this SDP.
It is important to note that the scope of activities that each utility undertakes during a normal
refueling outage is large and diverse. Besides refueling, activities associated with preventive
and corrective maintenance, modifications, surveillance testing, in-service inspection, and the
administrative activities that support these tasks make outage planning and control a significant
challenge. The coordination of these activities, with the objective to manage risk and maintain
key safety functions, is essential and goes beyond compliance with technical specifications
requirements during shutdown. In addition, while the scope of activities for an unplanned or
forced outage is far less than that of a refueling outage, the same awareness of vulnerabilities
during shutdown conditions is required to safely conduct these outages. This SDP has been
developed to assist the agency in determining the safety significance of findings during
shutdown conditions taking into account the unique characteristics described above.
Issue Date: 01/08/20 2 0609 Appendix G
0609G-03 DEFINITIONS
03.01 ABBREVIATIONS
AC Alternating Current
BWR Boiling Water Reactor
CCDP Conditional Core Damage Probability
CCW Component Cooling Water
CD Core Damage
CDF Core Damage Frequency
CETs Core Exit Thermocouples
CST Condensate Storage Tank
DC Direct Current
ECCS Emergency Core Cooling System
ICCDP Incremental Conditional Core Damage Probability
IE Initiating Event
IEL Initiating Event Likelihood
IMC Inspection Manual Chapter
INDIC Indication
LOI Loss of Reactor Inventory Initiating Event
LOLC Loss of Level Control
LOOP Loss of Offsite Power
LORHR Loss of RHR Initiating Event
LOSDC Loss of Shutdown Cooling
LTOP Low Temperature Over-Pressurization
MS Mitigation Systems
OD Over Drain
OP Operator
OPDRV Operation with Potential to Drain Reactor Vessel
PORV Power Operated Relief Valve
POS Plant Operational State
PRA Probabilistic Risk Assessment
PTS Pressurized Thermal Shock
PWR Pressurized Water Reactor
RHRSW Residual Heat Removal Service Water
ROP Reactor Oversight Process
RV Relief Valve
RWST Refueling Water Storage Tank
SD Shutdown
Issue Date: 01/08/20 3 0609 Appendix G
03.02 DEFINITIONS
The following definitions apply to both PWRs and BWRs, unless otherwise specified.
Available - A piece of equipment is considered available if (1) it can be put into service within
half the time that is needed before the equipment will be used to perform its function, (2)
procedures or standing orders exist for using the equipment to meet its intended function, (3) all
necessary supporting systems (such as AC power, cooling water, and DC control power) can be
put into service within half the time that is needed before the equipment will be used to perform
its function, and (4) operators have been trained on using the equipment for the given situation.
Cavity Flooded - An RCS condition with the reactor head removed and the water level raised to
the refuel floor.
Core Damage - Core damage corresponds to a peak clad temperature above 1300 degrees
Fahrenheit. Above 1340 degrees Fahrenheit, phenomena such as clad oxidation and
ballooning affect core behavior. This definition is consistent with the definition of the onset of
core damage used in NUREG/CR 6144 Vol. 2, Part 1A, "Evaluation of Potential Severe
Accidents During Low Power and Shutdown Operations at Surry, Unit 1, Analysis of Core
Damage Frequency from Internal Events During Mid-Loop Operations."
Gravity Feed (PWR Only) - Gravity feeding is the process of adding water to the RCS from a
storage source (e.g., condensate storage tank or refuel storage tank) without an active
component (e.g., pump). It requires the water source to be higher than the reactor and the
reactor to be at or capable of reaching atmospheric pressure. Gravity feeding may be credited if
gravity feed is expected to be available AFTER RCS boiling initiates. To credit gravity feed, the
analyst needs to consider the following factors that can negate the elevation head provided by
the RWST or other sources of RCS inventory: (1) pressure drops in the surge line, (2) entrained
water accumulating in the pressurizer, (3) RCS vent paths that are restricted (to control loose
parts or control off gassing).
Mid-loop Operation (PWR Only) - Mid-loop conditions exist whenever the RCS water level is
below the top of the flow area of the hot legs at the junction with the reactor vessel.
Operation with Potential to Drain Reactor Vessel - A planned maintenance evolution that if it is
not conducted properly can lead to a loss of inventory event. Therefore, any issues with
operations with potential to drain the reactor vessel should be evaluated using the appropriate
LOI criteria.
SDP Significance Determination Process
SG PORV Steam Generator Power Operated Relief Valve
SRA Senior Reactor Analyst
TTB Time to Boiling
TW Time Window
TW-E Early Time Window, before refueling operation
TW-L Late Time Window, after refueling operation
Issue Date: 01/08/20 4 0609 Appendix G
Reduced Inventory Operation - An RCS inventory condition that results in a reactor vessel water
level lower than three feet below the reactor vessel flange. Mid-loop is a subset of reduced
inventory. Also, one or more fuel assemblies must be in the reactor vessel. PWR only.
RCS Vented - For PWR plants, the RCS is considered vented when (1) SG heat removal cannot
be sustained, and (2) a vent path has been established that is large enough to support feed and
bleed. Examples of vent paths include: open pressurizer manways, safety relief valve removal,
or vessel head removal. For BWR plants, the RCS is considered vented when (1) the vessel
head is removed OR (2) the vessel head is on, however, a sufficient RCS vent path exists for
RWST/CST Depletion - Occurs when RWST/CST level reaches the level that requires makeup
or recirculation to continue injection to the RCS.
Self-Limiting LOI - These are loss of inventory events where the leakage point is above the
location where the RHR system attaches to the RCS. Therefore, the leakage will stop without
human intervention before the RHR/SDC system is lost. For these types of LOIs there shall be
no reliance on manual or automatic actions.
Shutdown Operation - Shutdown operation exists during refueling outages, forced outages, and
maintenance outages starting when the plant has met the entry conditions for RHR/DHR and
cooling has been initiated, and ending when the plant is heating up and RHR/DHR has been
secured.
Phases of a Significance Determination
Phase 1 - Characterization and Initial Screening of Findings: Phase 1 is used to characterize the
important attributes of the inspection finding and to initially screen the finding to identify those
with very low significance (Green), or greater than very low safety significance. Findings
screened to be of very low significance can be dispositioned by the licensee’s corrective action
program.
Phase 2 - Initial Risk Significance Approximation and Basis: Initial approximation of the risk
significance of the finding and development of the basis for this determination for those findings
that are not screened out in Phase 1 screening.
Phase 3 - Risk Significance Finalization and Justification: Also known as a detailed risk
evaluation, this is a review and as-needed refinement of the risk significance estimation results
from Phase 2, or development of any risk analysis outside of this guidance, by an NRC risk
analyst (any departure from the guidance provided in this document or IMC 0609, Appendix G
for Phase 1 or Phase 2 constitutes a Phase 3 analysis and must be performed by an NRC risk
analyst).
Types of Shutdown Findings
Precursor Finding - Inspection findings that: (1) cause an event (e.g., a loss of the operating
train of RHR/DHR) or (2) increase the likelihood of an event.
Condition Finding - Inspection findings that involve a degradation of the licensee’s capability to
mitigate an event if an event were to occur. Findings affecting the standby train of RHR/DHR
are condition findings.
Shutdown Initiating Events
Loss of RHR (LORHR) - Includes losses of RHR/DHR resulting from failures of the RHR/DHR
system (such as a RHR/DHR pump failure) or failures of the RHR/DHR support systems other
than offsite power.
Loss of Offsite Power (LOOP) - Includes losses of offsite power which cause a loss of the
RHR/DHR function.
Loss of Reactor Inventory (LOI) - Includes losses of RCS inventory that cause, or could cause,
a loss of the RHR/DHR function due to automatic isolation of RHR/DHR on low level for BWRs
or loss of RHR/DHR pump suction.
Loss of Level Control (LOLC) - This initiating event category includes: (1) the operator
overdrains the RCS to reach mid-loop conditions such that RHR/DHR is lost, and (2) the
operator fails to maintain level or flow control while in mid-loop such that the RHR/DHR function
is lost. (PWR Only)
Overdrain (OD) – Overdrain is a subset of LOLC. It is intended to capture those events where
while the RCS is being drained, from one target level range to a second lower range, the
evolution is not stopped within the desired final range. For example, starting level is one foot
below the reactor flange and the target range is six to twelve inches above the top of the hotleg.
If the drain down evolution was not stopped until level reached the top of the hotleg, then an
overdrain event has occurred.
PWR Plant Operational States (POSs)
POS 1 - This POS starts when the RHR/DHR system is put into service. The RCS is closed
such that a steam generator(s) could be used for decay heat removal, if the secondary side of
each available steam generator(s) has sufficient inventory to be considered available as a heat
sink. The RCS may have a bubble in the pressurizer. This POS ends when the RCS is vented
such that the steam generators cannot sustain core heat removal. This POS typically includes
Mode 4 (hot shutdown) and portions of Mode 5 (cold shutdown).
POS 2 - This POS starts when the RCS is vented such that: (1) the steam generators cannot
sustain core heat removal and (2) a sufficient vent path exists for feed and bleed. This POS
includes portions of Mode 5 (cold shutdown) and Mode 6 (refueling). Reduced inventory
operations and mid-loop operations with a vented RCS are subsets of this POS.
NOTE: Findings occurring during vacuum refill of the RCS require use of the POS 1 event trees
POS 3 - This POS represents the shutdown condition when the refueling cavity water level is at
or above the minimum level required for movement of irradiated fuel assemblies within
containment as defined by Technical Specifications. This POS occurs during Mode 6.
BWR Plant Operational States (POSs)
POS 1 - This POS starts when the RHR/DHR system is put into service. The vessel head is on
and the RCS is closed such that an extended loss of the RHR/DHR function without operator
Issue Date: 01/08/20 6 0609 Appendix G
intervention could result in an RCS re-pressurization above the shutoff head for the RHR/DHR
pumps.
POS 2 - This POS represents the shutdown condition when (1) the vessel head is removed and
reactor pressure vessel water level is less than the minimum level required for movement of
irradiated fuel assemblies within the reactor pressure vessel as defined by Technical
Specifications OR (2) the vessel head is on, however, a sufficient RCS vent path exists for
POS 3 - This POS represents the shutdown condition when the reactor pressure vessel water
level is equal or greater than the minimum level required for movement of irradiated fuel
assemblies within the reactor pressure vessel as defined by Technical Specifications. This POS
occurs during Mode 5 (refueling).
Time Windows
Early Time Window (TW-E) - This time window represents the time before POS 3 is entered.
The decay heat is relatively high. The reactor is either in POS 1 or 2.
Late Time Window (TW-L) - This time window represents the time after POS 3 is entered. The
decay heat is relatively low.
0609G-04 GUIDANCE
04.01 SCOPE
Appendix G is applicable during refueling outages, forced outages, and maintenance outages
starting when the plant has met the entry conditions for RHR/DHR and cooling has been
initiated, and ending when the plant is heating up and RHR/DHR has been secured.
Note: If the licensee is in a refueling outage or forced outage and the plant is above RHR/DHR
entry conditions, then inspectors will use the full power SDP tools acknowledging: (1)
decay heat is less compared to full power, potentially allowing for more time for operator
recovery, (2) some mitigating systems may require manual operation versus automatic
operation, and (3) some containment systems may not be required to be operable
potentially increasing the likelihood of containment failure.
If the plant is shutdown and the entry conditions for RHR/DHR and RHR/DHR cooling have not
been met, then Appendix G does not apply. The inspectors shall contact the SRA for
assistance. Appendix G is used to evaluate two categories of findings: those that actually cause
an event or increase the likelihood of an event (i.e., precursor findings), and those that affect the
ability to mitigate an event (i.e., condition findings).
Typical events of interest are: losses of RHR/DHR, losses of RCS inventory, low temperature
over pressurization (LTOP) events, and reactivity events. Another category of events is losses
of level control, which are discussed in Section 4.3. Losses of RHR/DHR include (but may not
be limited to) those caused by RHR/DHR system isolation, LOOP, failure of the running pump,
failure of cooling to the respective RHR/DHR heat exchanger, failure of system flow (e.g., flow
divergence away from the RCS), etc. Losses of inventory may or may not progress to the point
of losing RHR/DHR. Regardless, all losses of inventory should receive an appropriate review.
Issue Date: 01/08/20 7 0609 Appendix G
Appendix G is used to screen shutdown findings for risk significance. When using this guidance
to assess a finding, there are two possible outcomes: (1) the finding requires quantitative
assessment (Phase 2 or Phase 3 analysis) to determine its risk significance, or (2) the finding
can be screened as having very low risk significance (Green). The road map for shutdown
inspection findings for SDP Phases 1 and 2 is shown in Figure 1.
04.02 MITIGATION CAPACITY
Attachment 1 of this appendix contains screening questions in Exhibits 2-5 for shutdown
operation to ensure that the licensee is maintaining adequate mitigation capability. The
screening questions were developed to encompass all plant operational states defined by:
operational mode, time to boiling, reactor coolant system level, and reactor coolant system
configuration. To complete Exhibits 2-5 the inspector will need to use Table G1 in Exhibit 1 of
Attachment 1. In Table G1, there is a set of equipment, systems, instrumentation, policies, and
procedures that the staff expects the licensee to maintain during shutdown. Table G1 is
grouped by the five shutdown safety functions identified by NUMARC 91-06: decay heat
removal, inventory control, power availability, reactivity control, and containment. The inspector
should check to ensure that each applicable screening question in Exhibits 2-5 is being met. If
the applicable screening questions are not being met, then follow the direction in the exhibits for
either a Phase 2 or Phase 3 quantitative assessment. Findings not requiring quantitative
assessment may be screened as Green.
04.03 LOSSES OF LEVEL CONTROL DURING SHUTDOWN
In addition to ensuring that the licensee minimizes events and maintains mitigation capability
during shutdown, the staff is also monitoring conditions or events that represent a loss of level
control. If the conditions, as described in Attachment 1 of this appendix occur, then the finding
needs to be quantitatively assessed. Losses of level control only apply to PWRs.
04.04 FINDINGS REQUIRING QUANTITATIVE ASSESSMENT
If a finding needs quantitative assessment, then the finding should be forwarded to the SRA.
The SRA will then decide if the finding should be forwarded to NRR for Phase 3 analysis or the
finding will be evaluated using the Phase 2 PWR and BWR templates located in Attachments 2
and 3, respectively. The SRA should be sent the completed Phase 1 evaluation associated with
the finding and a complete description of the finding.
0609G-05 REFERENCES
SECY-97-168, “Issuance for Public Comment of Proposed Rulemaking Package for Shutdown
and Fuel Storage Pool Operation.”
NUMARC 91-06, “Guidelines for Industry Actions to Assess Shutdown Management.”
Generic Letter (GL) 88-17, “Loss of Decay Heat Removal – 10 CFR 50.54(f).”
Federal Register (dated February 4, 1999, Vol. 64, no. 23)
IMC 0308, Attachment 3, Appendix G, “Technical Basis for Shutdown Operations Significance
Determination Process.”
Information Notice (IN) 88-36, “Possible Sudden Loss of RCS Inventory During Low Coolant
Level Operation.”
EPRI NSAC-164L, “Guidelines for BWR Reactivity Control during Refueling,” 1992.
EPRI NSAC-183, “Risk of PWR Inadvertent Criticality during Shutdown and Refueling,” 1992.
EPRI 1003113 “An Analysis of Loss of Decay Heat Removal Trends and Initiation Event
Frequencies, (1989-2000),” 2001.
EPRI 1003465, “Low Power and Shutdown Risk Assessment Benchmarking Study,” 2002.
EPRI 1021176, “An Analysis of Loss of Decay Heat Removal and Loss of Inventory Event
Trends (1990–2009),” 2010.
NUREG/CR-5820, “Consequences of the Loss of Residual Heat Removal Systems in
Pressurized Water Reactors,” 1992.
NUREG/CR-6143, “Evaluation of Potential Severe Accidents during Low Power and Shutdown
Operations at Grand Gulf, Unit 1.” 1994
NUREG/CR-6144, “Evaluation of Potential Severe Accidents during Low Power and Shutdown
Operations at Surry, Unit 1.” 1994.
Issue Date: 01/08/20 9 0609 Appendix G
Figure 1: Road Map for Shutdown Findings
Issue Date: 01/08/20 Att1-1 0609 Appendix G
Attachment 1: Revision History Page
Commitment
Tracking
Number
Accession
Number
Issue Date
Change Notice
Description of Change Description of
Training
Required
And Completion
Date
Comment Resolution and
Closed Feedback Form
Accession Number (PreDecisional, Non-Public
Information)
N/A 04/21/00
CN 00-007
IMC 609 supports the new Reactor Oversight
Program for the significance determination of
findings. The significance determination process
detailed in the manual chapter is designed to
characterize the significance of inspection findings
for the NRC licensee performance assessment
process using appropriate risk insights.
N/A N/A
N/A 02/27/01
CN 01-005
IMC 0609 App G has been revised to reflect changes
based on comments by regional inspectors. These
include definition for entry conditions and instructions
to bring issues to the attention of the SRAs.
N/A N/A
N/A 05/25/04
CN 04-015
IMC 0609 App G is revised to move the Phase 1
operational checklist that were previously in the main
body of Appendix G and place it in a new manual
chapter document. The new document is
Attachment 1 to Appendix G. The separation was
done to make Appendix G consistent with other SDP
appendices in the manual chapter. The main body of
Appendix G was changed to provide a “road map” of
how to navigate through the various attachments to
Appendix G. Attachments 2 and 3, which provide
SDP Phase 2 guidance for PWRs and BWRs,
respectively, are referenced in the revised Appendix
G.
N/A N/A
Issue Date: 01/08/20 Att1-2 0609 Appendix G
Commitment
Tracking
Number
Accession
Number
Issue Date
Change Notice
Description of Change Description of
Training Required
And Completion
Date
Comment Resolution and
Closed Feedback Form
Accession Number (PreDecisional, Non-Public
Information)
N/A 02/28/05
CN 05-007
IMC 609 App G is revised to clarify the definition of
available. The revised definition states that
necessary support systems can be put into service
within half the time needed for the equipment to
perform its function.
N/A N/A
N/A ML13050A933
05/09/14
CN 14-011
IMC 0609 App G is revised to enhance the usability
of this appendix, based on feedback received from
the SRA. The formatting was updated to be
consistent with IMC 0609 Appendix A. The
abbreviations section is a new addition and the
definitions section is updated to include many
additional useful terms. Figure 1 is updated to be
consistent with the revisions to 0609 Appendix G,
Att. 1. Table 1 was removed and the information is
included in Attachment 1. Incorporated feedback
from ROPFF 0609G-1323 and 0609G-1932. This is
a complete reissue no red line.
N/A ML13162A640
0609G-1323
0609G-1932
N/A ML19101A289
01/08/20
CN 20-004
Clarified loss of control is loss of level control and
only applies to PWR plants.
Revised the definition for Shutdown Operations to
align it with the scope of Appendix G.
Updated Figure 1 to clarify when analyses get sent
to headquarters for review. The updated Figure 1
now includes guidance on issues with open cold leg
penetrations.
No training is
required.
Issue Date: 01/08/20 Att1-3 0609 Appendix G
Commitment
Tracking
Number
Accession
Number
Issue Date
Change Notice
Description of Change Description of
Training Required
And Completion
Date
Comment Resolution and
Closed Feedback Form
Accession Number (PreDecisional, Non-Public
Information)
Added additional detail to the RCS Vented definition
to cover BWRs – taken from the POS 2 definition
for BWRs.
Removed the statement that LOOP findings are not
assessed in POS 3 since Attachment 1 still assess
them depending on timing of offsite power recovery.
Eliminated the Caution prior to section 04.02, it was
not necessary.
Revised the definition of “available” to make it clear
that equipment can be considered available if it can
be put in service within half the time before it will be
needed. Not to be confused with the PRA mission
time. This is also not the same “available” definition
as used for maintenance rule purposes.